NUMO-TR-21-01

The NUMO Pre-siting
SDM-based Safety Case

November 2021
Nuclear Waste Management Organization of Japan (NUMO)

2021 年 11 月 初版発行
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〒108-0014

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NUMO-TR-21-01

The NUMO Pre-siting
SDM-based Safety Case

November 2021
Nuclear Waste Management Organization of Japan (NUMO)

CONTENTS
1. INTRODUCTION........................................................................................................................... 1-1
1.1 Basic concept of geological disposal ......................................................................................... 1-1
1.2 Background to deep geological disposal in Japan ..................................................................... 1-2
1.3 Revised strategy for geological disposal.................................................................................... 1-3
1.4 Overview of this report .............................................................................................................. 1-7
1.4.1 Purpose of the report ........................................................................................................... 1-7
1.4.2 Production of the report....................................................................................................... 1-8
References ...................................................................................................................................... 1-17
2. SAFETY CONCEPT ...................................................................................................................... 2-1
2.1 Requirements to be considered when planning geological disposal in Japan........................... 2-1
2.1.1 Radioactive waste subject to geological disposal ............................................................... 2-1
2.1.2 Required safety functions of the repository ........................................................................ 2-5
2.1.3 Meeting the requirements of laws and regulations ........................................................... 2-11
2.2 Site selection strategy ............................................................................................................... 2-15
2.2.1 Siting aspects to ensure operational safety........................................................................ 2-15
2.2.2 Siting aspects to ensure post-closure safety ...................................................................... 2-16
2.2.3 Spatial scales during stepwise site selection ..................................................................... 2-17
2.2.4 Development of site descriptive models (SDMs) ............................................................. 2-19
2.2.5 Approach adopted for site assessment and SDM construction ........................................ 2-20
2.3 Repository design strategy ....................................................................................................... 2-22
2.3.1 Stepwise repository design approach ................................................................................ 2-22
2.3.2 Availability of engineering technology............................................................................. 2-23
2.3.3 Retrievability, environmental protection and monitoring considerations ........................ 2-23
2.3.4 Approach for the development of repository design ........................................................ 2-24
2.4 Safety assessment strategy ....................................................................................................... 2-25
2.4.1 Operational safety assessment........................................................................................... 2-25
2.4.2 Approach for operational safety assessment ..................................................................... 2-26
2.4.3 Post-closure safety assessment.......................................................................................... 2-26
2.4.4 Approach adopted for post-closure safety assessment ..................................................... 2-27
2.5 Management systems ............................................................................................................... 2-29
2.5.1 Uncertainty management .................................................................................................. 2-31
2.5.2 Quality management (QM) ............................................................................................... 2-33
2.5.3 Requirements management............................................................................................... 2-34
2.5.4 Knowledge management................................................................................................... 2-35
2.5.5 Management of scientific and technology development .................................................. 2-36
2.5.6 Management of human resources ..................................................................................... 2-36i

2.6 Summary .................................................................................................................................. 2-37
Supporting Reports (SRs) .............................................................................................................. 2-39
References ...................................................................................................................................... 2-40
3. SELECTION AND MODELLING OF GEOLOGICAL ENVIRONMENTS SUITABLE FORGEOLOGICAL DISPOSAL .............................................................................................................. 3-1
3.1 The role of the geological environment in ensuring repository safety...................................... 3-2
3.1.1 Requirements on the geological environment and associated influencing factors............. 3-2
3.1.2 Approach to using influencing factors in site selection ...................................................... 3-5
3.1.3 Selecting a geological environment suitable for geological disposal in Japan................... 3-7
3.2 Process for identifying environments suitable for geological disposal ................................... 3-11
3.2.1 Basic strategy and methodology ....................................................................................... 3-11
3.2.2 Stepwise implementation of site investigations ................................................................ 3-18
3.2.3 Site characterisation technology........................................................................................ 3-28
3.3 Development of SDMs for representative host rock settings ................................................. 3-39
3.3.1 Model aims and development procedure .......................................................................... 3-39
3.3.2 Classification of representative host rocks ........................................................................ 3-40
3.3.3 Development of a representative SDM............................................................................. 3-42
3.4 Probability of occurrence and impacts of natural perturbing phenomena .............................. 3-85
3.4.1 Volcanic/igneous activity................................................................................................... 3-85
3.4.2 Earthquake and fault activity............................................................................................. 3-85
3.4.3 Uplift and erosion .............................................................................................................. 3-87
3.5 Summary and future perspective ............................................................................................. 3-88
3.5.1 Summary ........................................................................................................................... 3-88
3.5.2 Future perspective ............................................................................................................. 3-90
Supporting Reports (SRs) .............................................................................................................. 3-97
References ...................................................................................................................................... 3-99
4. REPOSITORY DESIGN AND ENGINEERING TECHNOLOGY............................................. 4-1
4.1 Approach to development of repository concepts ..................................................................... 4-1
4.1.1 Capturing requirements in design factors ........................................................................... 4-2
4.1.2 Setting and utilising design requirements ........................................................................... 4-4
4.1.3 Development of design options .......................................................................................... 4-5
4.2 Repository design procedures and boundary conditions ........................................................... 4-7
4.2.1 Repository design procedures ............................................................................................. 4-7
4.2.2 Waste characteristics............................................................................................................ 4-8
4.2.3 Disposal concepts and components .................................................................................... 4-9
4.2.4 Relationship between safety functions and repository components................................. 4-14ii

4.2.5 Features of the SDM relevant to repository design .......................................................... 4-23
4.3 Setting of repository depth ....................................................................................................... 4-25
4.4 EBS design ............................................................................................................................... 4-26
4.4.1 EBS for HLW .................................................................................................................... 4-27
4.4.2 Design of the TRU EBS.................................................................................................... 4-49
4.4.3 Construction of the EBS and waste emplacement............................................................ 4-62
4.5 Design of underground facilities .............................................................................................. 4-67
4.5.1 Concept of repository design ............................................................................................ 4-67
4.5.2 Tunnel design .................................................................................................................... 4-71
4.5.3 Design of backfill and plugs.............................................................................................. 4-83
4.5.4 Design of the underground facility layout ........................................................................ 4-91
4.5.5 Design of underground services...................................................................................... 4-122
4.5.6 Safety measures during construction and operation ....................................................... 4-127
4.5.7 Engineering technology for construction, operation and closure of underground facilities...................................................................................................................................... 4-136
4.6 Design of surface facilities ..................................................................................................... 4-138
4.6.1 Assuring safety in the design of surface facilities ........................................................... 4-138
4.6.2 Design of waste reception, inspection and encapsulation facilities ................................ 4-142
4.7 Reversal and retrieval ............................................................................................................. 4-151
4.7.1 Assuring practicality of retrieval ..................................................................................... 4-152
4.7.2 Retrieval methods............................................................................................................ 4-156
4.8 Summary and future perspective ........................................................................................... 4-158
4.8.1 Summary ......................................................................................................................... 4-158
4.8.2 Issues for specific geological environments ................................................................... 4-160
4.8.3 Future perspectives .......................................................................................................... 4-162
Supporting Reports (SRs) ............................................................................................................ 4-168
References.................................................................................................................................... 4-170
5. OPERATIONAL SAFETY ASSESSMENT.................................................................................. 5-1
5.1 Strategy for operational safety assessment ................................................................................ 5-1
5.1.1 Procedures for developing operational safety assessment scenarios .................................. 5-1
5.1.2 Procedures for assessment................................................................................................... 5-3
5.2 Premises for the assessment ....................................................................................................... 5-4
5.2.1 Operational processes assumed in assessment.................................................................... 5-4
5.2.2 Waste inventory and planned throughput............................................................................ 5-5
5.3 Assessment of normal operational scenarios ............................................................................. 5-6
5.4 Assessment of abnormal operation scenarios ............................................................................ 5-6
5.4.1 Development of abnormal operation scenarios .................................................................. 5-7
5.4.2 Consequence analysis for abnormal operation scenarios ................................................. 5-12
5.4.3 Potential for radionuclide release from the repository ...................................................... 5-22iii

5.5 Mitigation of, and recovery from, accidents ............................................................................ 5-23
5.5.1 Lessons from accidents at WIPP....................................................................................... 5-23
5.5.2 Mitigation and recovery plans for a Geological Disposal Facility (GDF) ....................... 5-25
5.6 Summary and future perspective ............................................................................................. 5-25
5.6.1 Key findings of this chapter .............................................................................................. 5-25
5.6.2 Future perspective ............................................................................................................. 5-26
Supporting Reports (SRs) .............................................................................................................. 5-28
References ...................................................................................................................................... 5-29
6. POST-CLOSURE SAFETY ASSESSMENT................................................................................ 6-1
6.1 Basic framework of safety assessment ...................................................................................... 6-1
6.1.1 Safety assessment procedure............................................................................................... 6-1
6.1.2 Prerequisites for safety assessment ..................................................................................... 6-2
6.1.3 Treatment of spatial scales ................................................................................................ 6-15
6.1.4 Treatment of evaluation period ......................................................................................... 6-16
6.1.5 Risk-informed safety assessment and evaluation criteria ................................................. 6-17
6.2 Methodology of safety assessment.......................................................................................... 6-20
6.2.1 Basic procedure of safety assessment ............................................................................... 6-20
6.2.2 Treatment of uncertainty in the safety assessment............................................................ 6-23
6.3 Developing scenarios and establishing analysis cases ............................................................ 6-23
6.3.1 Description of system behaviour after repository closure ................................................ 6-24
6.3.2 Developing scenarios ........................................................................................................ 6-34
6.3.3 Setting analysis cases ........................................................................................................ 6-58
6.4 Radionuclide migration analysis and dose assessment ........................................................... 6-85
6.4.1 RN migration analyses and dose evaluations of base scenario ........................................ 6-85
6.4.2 Dose evaluation for variant scenarios ............................................................................. 6-138
6.4.3 Evaluation of low probability perturbation scenarios..................................................... 6-149
6.5 Evaluation of inadvertent human intrusion scenarios ........................................................... 6-157
6.5.1 Development of human intrusion scenarios ................................................................... 6-157
6.5.2 Results and discussion..................................................................................................... 6-162
6.6 Summary and Future Perspective.......................................................................................... 6-164
6.6.1 Summary of safety evaluations ....................................................................................... 6-164
6.6.2 Extension of assessment to coastal sites ......................................................................... 6-165
6.6.3 Potential to improve future assessments ......................................................................... 6-166
Supporting Reports (SRs) ............................................................................................................ 6-173
References.................................................................................................................................... 6-174
Appendix......................................................................................................................................6-A-1

iv

7. SAFETY CASE SYNTHESIS........................................................................................................ 7-1
7.1 Objectives of this chapter ..........................................................................................................7-1
7.2 Integration of arguments in a safety case ..................................................................................7-2
7.2.1 Summary of the basis for safety evaluation ........................................................................ 7-2
7.2.2 Evaluating the safety of geological disposal ..................................................................... 7-16
7.2.3 Constraints on, and scope of, the safety case presented in this report .............................. 7-32
7.3 Assuring the safety case is “fit for purpose” (confidence building) ........................................7-34
7.3.1 Technical basis for Assessment ......................................................................................... 7-34
7.3.2 Dealing with uncertainty in the safety case....................................................................... 7-39
7.3.3 Structured management of safety case development........................................................ 7-43
7.3.4 Acceptance of safety case content ..................................................................................... 7-49
7.4 Outlook for future safety cases ................................................................................................7-49
7.4.1 Building an evolving safety case....................................................................................... 7-49
7.4.2 Fundamental safety case improvements ........................................................................... 7-53
7.5 Summary..................................................................................................................................7-56
Supporting Reports (SRs) ..............................................................................................................7-57
References......................................................................................................................................7-58
8. SUMMARY AND CONCLUSIONS ......................................................................................... 8-1
References........................................................................................................................................8-5v

1 INTRODUCTION
This report is a translation of the original Japanese Safety Case (SC) Report, however it isimportant to note that it does not represent an exact 1:1 translation. The content has beenslightly modified in places to make it more easily understandable to an international audience,
who may not be fully familiar with the Japanese radioactive waste management programme.

1.1 Basic concept of geological disposal
The use of nuclear power generates a range of radioactive wastes. It is the ethicalresponsibility of the current generation to identify specific measures for the safe disposal ofsuch waste and to ensure steady progress towards implementation of disposal projects.
In Japan, where energy resources are scarce, the basic policy involves a nuclear fuel cycle
in which uranium and plutonium from spent fuel are reused after reprocessing. Reprocessing
results in highly radioactive liquid waste, which is vitrified to produce a stable, solid,borosilicate glass high-level radioactive waste (HLW) that is suitable for disposal. In someother countries with nuclear programmes, direct disposal of spent fuel is also adopted as amethod of waste management. Although the radioactivity of HLW decays over time, itremains significant over an extremely long time, as shown in Figure 1.1-1. However, after a
few tens of thousands of years, this becomes equivalent to the radioactivity of the uranium orefrom which it was produced.
11

11

10

10

Discharge
Discharge
10

10

10

10

9

9

10

10

Reprocessing
Reprocessing//vitrification
vitrification

8

8

10

10

7

7

10

10

6

10

Enrichment
Enrichment

5

5

10

10

4

10

3

3

10

10

Mill tailings
Depleted uranium

2

10

2

10

1

1

0

-1

10

Fabrication

In the form ofuranium minerals in
the geological environment

1010Conversion

10

-1

10

-2

10

0

10

Storage of spent fuelStorage of vitrified waste

Uranium
Uraniummining
mining && millingmilling

4

10

Loading
Loading

Operation of nuclear power plant

Radioactivity(GBq)
Radioactivity[GBq]

6

10

-2

10
8

10

6

4

2

10
10
10
Timebefore
beforeloading
loadingTime(y)[y]Uranium
Uranium oreore (1%)(1%)(750 tonnes)(750 tonnes)

0

10

0

0

10

2

10

4

6

10
10
Time after loading [y]810Time after loading (y)Uraniummaterial(10 tonnes)(10 tonnes)

Nuclear fuelNuclear fuelVitrifiedVitrifiedwaste
waste(1MTU)(1 MTU)(400kg)(0.5 tonnes)

Figure 1.1-1 Temporal change of radioactivity for vitrified HLW(modified from JNC, 2000 [1], 1 MTU = 1 metric tonne uranium). Note that the weight of thevitrified waste generated by JFNL is 400 kg (see Section 2.1.1) and its canister loading is 0.8. 1
MTU corresponds to 500 kg of glass for fuel reprocessed by JNFL

1-1

As a result of numerous studies on methods for disposing of long-lived radioactive wastes,
mainly by countries using nuclear power, it is internationally recognised that geologicaldisposal is the most promising strategy that has no significant impact on the living
environment and requires no direct human control (OECD/NEA 1995 [2]).
Repositories for the disposal of higher-level radioactive wastes generally adopt a system ofmultiple engineered barriers (engineered barrier system - EBS), within a suitable deep
geological setting. Together, the EBS and the natural geological barriers ensure isolation of
the waste from the biosphere for a significant period of time (1 to 10 ky or more) after
repository closure. During this period, most radionuclides (RNs) will have decayedcompletely or to a level where any radiological risk is acceptably low. In addition, theenvironment deep underground is less susceptible to human activities and various natural
perturbing phenomena affecting the surface, such as floods and landslides.

1.2 Background to deep geological disposal in Japan
In Japan, the Japan Atomic Energy Commission (JAEC) promulgated the “Basic concepts
for high-level radioactive waste disposal” in 1998 [3]. Further to this, “H12: Project toestablish the scientific and technical basis for HLW disposal in Japan”, (hereafter the “H12Report”), produced by the Japan Nuclear Cycle Development Institute (JNC; now Japan
Atomic Energy Agency - JAEA) in 2000 (English edition) [1], compiled the results ofresearch and development (R&D) conducted since 1976 and established the fundamental
feasibility of geological disposal.
Based on these reports, the “Act on Final Disposal of Specified Radioactive Waste”(hereafter the “Final Disposal Act”) was enacted in 2000. “Specified Radioactive Waste1”
initially referred to vitrified high-level radioactive waste resulting from the reprocessing ofspent fuel, but was extended to include what is termed TRU waste, produced from both
reprocessing and MOX (mixed U/Pu oxide) fuel fabrication. Although some TRU waste
contains significant quantities of transuranic RNs, it also contains a wide range of other RNs
and the term TRU waste as used in this report refers specifically to the fraction of suchmaterial intended for geological disposal, unless otherwise specified. “The second progress
report on research and development for TRU waste disposal in Japan” [4] (hereafter the“TRU-2 Report”), compiled by JAEA and the Federation of Electric Companies of Japan(FEPC) in 2007, formed the basis for this revision of the Final Disposal Act.
Based on the Final Disposal Act, the Nuclear Waste Management Organization of Japan(NUMO) was established in 2000 as the organisation responsible for project implementation.
The site selection process specified in the Act consists of three phases of investigation:
literature surveys (LS), preliminary investigations (PI) and finally, detailed investigations (DI).
In 2002, NUMO issued a nationwide call for volunteer municipalities to initiate the repository
siting process.
Since its establishment, NUMO has been conducting R&D to enhance the reliability of thetechnical basis presented in the H12 report, with the focus on safety and feasibility. In
addition, JAEA has also been carrying out research in order to advance the Japanesegeological disposal programme – in particular in two underground research laboratories
The term “specified” radioactive waste was changed in 2015 when the Basic Policy on Final Disposal ofHigher-level Radioactive Wastes was revised (see Section 1.3). The term now used is “designated” radioactive
waste. N.B. This footnote is not included in the Japanese version of the report.
1

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(URLs), Mizunami (in crystalline rock) and Horonobe (in sedimentary rock) - to further
improve geological disposal technology and develop safety assessment methodology.
Results of the research performed by NUMO were compiled and published in a series oftechnical documents (in Japanese) leading up to publication of the report “Safety of thegeological disposal project 2010” [5] (hereafter the “2010 Report”), which summarises
NUMO’s progress in establishing technologies for safe geological disposal. NUMO has been
continuously developing the technologies required for safe implementation of the disposal
project and, in coordination with National Government, the electricity utilities and otherrelevant organisations, has made significant efforts in communication, in order to build public
acceptance for geological disposal (e.g., public hearings and public information activities).
Early efforts resulted in Toyo town in Kochi prefecture coming forward to make anapplication for a literature survey in January 2007. However, soon after the results of amayoral election held in April 2007, the application was withdrawn and no literature survey
was initiated.
The accident at the Tokyo Electric Power Co., Ltd. (TEPCO) Fukushima Daiichi Nuclear
Power Plant occurred as a result of the Great Tohoku Earthquake in March 2011, causing lossof public trust in both the nuclear industry and associated organisations. For the geologicaldisposal project, the Science Council of Japan recommended in 2012 that the policy on
disposal of HLW should be fundamentally reviewed. The Science Council also suggested that
extended storage of waste should be introduced in order to allow more time for both progressin research to enhance the safety of geological disposal and building of social consensus. In
particular, based on the lack of consensus among geoscience experts regarding the concept oflong-term safety and the presence of suitable geological environments in Japan, the Science
Council insisted that the limitations of scientific knowledge and associated technology should
be recognised and a forum for specialist deliberations be established [6]. In response to this,the Japan Atomic Energy Commission (JAEC) proposed research on the feasibility ofgeological disposal that reflected the latest findings in the field of geosciences and sharing ofresults with the public [7].

1.3 Revised strategy for geological disposal
Based on the situation described in the previous section, two expert working groups were
set up in 2013 by the National Advisory Committee for Natural Resources and Energy; one
covering radioactive waste and the other geological disposal technology.
The Radioactive Wastes Working Group (hereafter the Radioactive Wastes WG) discussed
re-establishing policies on geological disposal with the aim of highlighting areas to beaddressed in advancing the project [8]. Key conclusions include:••

It is internationally recognised that geological disposal remains the most appropriate
final disposal method for higher activity radioactive waste. Even in Japan, despite thechallenging boundary conditions, geological disposal is based on an extensive
scientific knowledge base and its feasibility has been demonstrated. Geologicaldisposal is the sole implementation method defined in law.
A mechanism for reversibility (capability to reverse or change decisions later) andretrievability (capability to recover the emplaced waste from the repository) should beensured to allow reconsideration of the decision on final disposal by current and future
generations, considering both existing uncertainties and enhanced social consensus on
geological disposal.
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•

•

To improve the site selection process, the Government should provide an explanationof the required geological environment characteristics from a scientific point of view
and promote understanding of site selection, showing areas that are considered to be
more suitable from a scientific viewpoint.
It will be important for the implementing organisation and the scientific community toestablish the technical reliability of geological disposal by continuously reviewing andutilising the latest scientific knowledge, establishing R&D on alternative disposal
options in parallel with site selection and promoting stepwise social consensusbuilding regarding the acceptability of geological disposal.

Characteristics of the geological environment in Japan and its long-term stability werediscussed by the Geological Disposal Technology WG [9], based on the latest knowledge
obtained since the H12 report. Key conclusions include:••

From the viewpoints of thermal, rock-mechanical, hydrogeological and geochemical
characteristics, potentially favourable geological environments that provide thenecessary conditions for hosting a geological disposal system have been identified andare widely distributed throughout Japan.
Through appropriate stepwise site investigations, sites with a favourable geologicalenvironment and assured long-term stability can be selected, despite the effects oflong-term evolution in all such settings.

Taking such input into account, the Government revised the “Basic Policy on Final
Disposal of Designated Radioactive Wastes” (hereafter the “Basic Policy on Final Disposal”)
in May 2015, and the following points were clearly stated:
•

•
•
•
•

Clarification of the current generation's responsibilities and assuring flexibility for
future generations for decision-making (e.g., technological developments, including
alternative options, and making progress in geological disposal are the responsibilityof this generation (i.e., burdens should not be passed on to the next generation), while
ensuring reversibility and retrieval options for future generations)2.
The need to foster both nationwide and regional understanding of the issues involved(e.g., by establishing regular, open dialogue).
The National Government is to take the lead in key areas, such as presenting regionsthat are considered to be scientifically more suitable, making proposals to relevant
local governments to support their understanding and cooperation in research, etc.
Supporting relevant regions (e.g., establishing dialogue for building consensus).
Improving the implementation system (e.g., strengthening NUMO’s management
system, increasing the involvement of JAEC in order to carry out continuous
evaluations of progress in technological development and thus ensure reliability, andto support the Nuclear Regulation Authority (NRA) in gradually developing
appropriate safety guidelines).

It has thus been confirmed by the Government that geological disposal is the mosttechnologically promising approach to managing the designated wastes and, in the light of thelatest scientific findings, there is a good prospect of finding an environment suitable for2

Technology development is to include alternative options, as indicated in the Basic Policy on Final Disposal,
which involves, e.g., research on the feasibility of direct disposal of spent fuel [10] and on the partitioning andtransmutation of longer-lived radionuclides [11], implemented by JAEA and other organisations. NUMO focuseson geological disposal of designated wastes as defined in the Final Disposal Act and hence carries out no R&D
on such alternative options.

1-4

geological disposal in Japan. Advancing this to implementing such disposal is the
responsibility of the current generation.
In response to the revision of the Basic Policy on Final Disposal, the Radioactive Wastes
WG decided to summarise the scientific characteristics of regions in the whole of Japan anddiscussed how to best establish dialogue and promote consensus building on this basis. The
purpose is both to contribute to the selection of suitable sites and to provide an opportunity
and materials for the general public to recognise and understand the need for final disposal
and the significance of the project [12]. In addition, the Geological Disposal Technology WG
discussed the requirements and criteria for scientific features of suitable regions fromgeoscientific and technical viewpoints.
As a result of these discussions, the Geological Disposal Technology WG published areport on the scientific features relevant for geological disposal [12]. In this report, as shownin Figure 1.3-1, the requirements and criteria for assessing site characteristics based on current
knowledge prior to initiation of site investigation are described.

Figure 1.3-1 Relationship between sharing the scientific features of regions with municipalities
and the site selection process, as defined in the Final Disposal Act [from 12]

This report considers the degree of confidence that safe geological disposal can bedemonstrated based on future on-site work. Considering long-term stability, safety during
construction and operation, safety during transport, and implementation practicality,
requirements and criteria for identifying “preferred” and “non-preferred” sites are developed.
For areas where there is a high probability that desirable characteristics can be confirmed, an
additional positive criterion is that the distance from the coast is sufficiently short,recognising the advantages (e.g. relating to logistics and nuclear security) of waste transport
from a coastal port to the repository site [12].
These requirements and criteria [12] led to publication in July 2017 of the ‘nationwide mapof “scientific features” relevant for geological disposal’ [13] (Figure 1.3-2), hereafter the“Nationwide Map”. This classified the whole of Japan into the following four regions:•Those probably unsuitable from the viewpoint of long-term stability, such as thosenear volcanoes or active faults,
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•
•
•

Those with mineral resources such as oil fields and gas fields, which are undesirablefrom the viewpoint of future potential human intrusion,
Those with a relatively high probability of being suitable due to lack of the abovementioned undesirable requirements and criteria,
Of the likely suitable areas, those preferable also in terms of transport advantages.

Figure 1.3-2 Nationwide map of “scientific features” relevant for geological disposal(Source: Ministry of Economy, Trade and Industry, 2017 [13])1-6

In October 2020, the town of Suttu in Hokkaido prefecture applied for a literature survey tobe undertaken. Additionally, another village in Hokkaido prefecture, Kamoenai, accepted a
proposal for a literature survey by the Japanese Government. NUMO initiated literature
surveys for both municipalities in November 2020. Currently, the Japanese Government and
NUMO are continuing to promote dialogue throughout Japan to deepen public understandingof geological disposal based on the Nationwide Map. This may lead to further municipalities
accepting to undertake a literature survey.

1.4 Overview of this report
1.4.1 Purpose of the report
When explaining site3 selection activities based on the revised Basic Policy on Final Disposal,
it is important for NUMO to be an organisation that is trusted by society and, in particular, by
local communities that may accept a literature survey. Emphasis is thus on establishingdialogue that will convey how NUMO will assure safe geological disposal in a manner thatwill gain public confidence.
In addition, regarding the technical reliability of geological disposal, continuous evaluation
based on the latest knowledge is required, as described in the Basic Policy on Final Disposal.
Describing the methods for assuring safe disposal for relevant geological environments is thus
an important requirement for NUMO.
Therefore, based on the latest scientific findings and technological developments, the study
reported here was conducted for sites typical of those with a relatively high probability of
being suitable, as indicated in the Nationwide Map. In accordance with the IAEA glossary, asafety case at this early stage of development should acknowledge the existence of any
unresolved issues and should provide guidance for work to solve these issues during futurestages of development. More specifically, this report has the purpose of:
•

•

•
•

Illustrating the technology for surveying and evaluating sites with conditions
necessary for isolating and containing designated radioactive wastes and ensuring no
significant impacts on the human living environment for a variety of relevant
geological environments in Japan.
Presenting site descriptive models (SDMs) that capture key characteristics ofinformation obtained from studies of deep geological environments in Japan,
illustrating repository designs and engineering safety measures tailored to these which
fulfil required safety functions.
Demonstrating the safety of a potential repository site, both pre- and post-closure of
the repository.
Based on integration of the above input, identifying technical issues for further
improvement and the required R&D needed for geological characterisation, repository
design and safety assessment.

Based on this work, a methodology is outlined which focuses on assuring the safety of theimplemented repository and involves stepwise tailoring to site-specific conditions. Thismethodology will be refined as the staged siting process progresses and updated to reflect
future advances in science and technology.

NUMO uses “site” as a term broadly referring to the area to be surveyed and, thereafter, the area finallyselected as the location of the repository.
3

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1.4.2 Production of the report(1) Compilation as a safety caseIn order to continually increase the confidence in the safety of geological disposal
throughout the project period, at programme milestones we will summarise the evolving“safety case”, which reflects the boundary conditions and incorporates the knowledge base atthat time. The safety case4 is a concept that involves structured demonstration of important
repository safety aspects, providing this information in a manner accessible to key
stakeholders [14]. This safety case has been prepared by NUMO as the project implementer.
Its conclusions will be judged by Japanese stakeholders (regulatory bodies, National
Government, local residents, general public, etc.), complemented by an international technicalreview, in terms of the credibility of the repository safety case presented.
This approach is consistent with the stated objectives of gaining the public confidence
needed to advance the project to the next phase of siting by showing both the feasibility ofsafe geological disposal and NUMO’s technical preparations for implementation. NUMO
plans [5] to develop the safety case further, based on the characteristics of specific sites after
the preliminary investigation phase. It will be updated again at later project milestones: site
selection, disposal facility construction and subsequent licence applications. Thus, this reportforms a basic safety case before any site has been identified, presenting the framework andinformation base for further safety cases to be produced in the future. NUMO has thus named
this report “the NUMO Pre-siting SDM-based Safety Case” (hereafter the “NUMO SC”). The
approach to safety case development has been described by several international organisations[14] [15] [16], resulting in the general structure shown in Figure 1.4-1 [16].
After establishing the “purpose and context” for a specific phase of the project, a “safety
strategy” presents the approach to achieving the required safety level in terms of managementof the whole project, site selection, repository design and safety assessment. Next, the“assessment bases” are documented, which include the disposal system concept (sitecharacteristics and associated repository design), the scientific knowledge and methodologythat form the basis for this, and the models, codes and databases supporting design and safetyassessment. Safety will be demonstrated in a logical manner by way of a “safety assessment”,
together with “supporting evidence”, such as that provided by natural analogues, that willreinforce/support the assessment results. Synthesis of both the safety assessment and the
supporting arguments will demonstrate that NUMO can build/develop a safety case that is fit
for purpose given specified boundary conditions.
This report is generally consistent with the role and basic structure of safety cases
presented by a number of international organisations [14] [15] [16], considering not only theperiod post-closure of the repository but also that pre-closure, to ensure consistent treatmentof safety as a whole.

The IAEA (2012) [14] defines the safety case as follows: “The safety case is the collection of scientific,
technical, administrative and managerial arguments and evidence in support of the safety of a disposal facility,
covering the suitability of the site and the design, construction and operation of the facility, the assessment ofradiation risks and assurance of the adequacy and quality of all of the safety related work associated with thedisposal facility.”4

1-8

Figure 1.4-1 Structure of a safety case as shown by OECD/NEA [from 16](2) Structure of the report
For Japanese boundary conditions, the NUMO SC is modified from the basic structure
shown in Figure 1.4-1, resulting in Figure 1.4-2, which also shows the relationship between
each component of the safety case and the section of this report in which it is covered, as
discussed further below.
Chapter 2 describes specific strategies and concepts devised by NUMO to facilitate
progress towards safe implementation of a deep geological repository, managing site selection,
repository design and safety assessment. Regarding the assessment basis, in response to thestrategies shown in Chapter 2, methodologies and techniques for analysis along withassociated models and data are used to illustrate the selection of appropriate sites, repository
design and evaluation of pre-/post-closure safety.
Chapter 3 describes the geological investigation techniques used for characterisation of
relevant sites and development of the site descriptive models (SDMs) that incorporate
resulting site understanding. From these SDMs, tailored designs for the disposal system andassociated safety assessment can be developed.
Chapter 4 describes repository design and engineering (including waste packages), basedon a design philosophy and associated methodology to ensure that the required safety
1-9

functions are both provided and are effective. This is illustrated by tailoring designs to the
SDMs for the representative host rock types outlined in Chapter 3. The assumed construction,
operation and closure procedures are illustrated and the engineering feasibility based on
current technology assessed.
Chapter 5 evaluates operational safety, based on the repository designs presented in
Chapter 4, and the procedures for repository construction, operation and closure.
Chapter 6 presents the required methodology and uses it to assess safety after repository
closure for the concepts outlined in Chapter 4 in geological settings defined by the SDMs
from Chapter 3.
Chapter 7 summarises input from previous chapters and describes its integration into asafety case. Based on this, it highlights topics for future R&D, with special emphasis on
overarching issues not captured in the specialist chapters, and outlines how the safety casewill develop further as siting progresses.
Chapter 8 concludes by summarising the key messages and perspectives in the report.

Figure 1.4-2 Adaptation of the NEA safety case outline to fit NUMO programme boundary
conditions (with indication of the chapters of this report in which issues are addressed). QA:
quality assurance, QM: quality management, RMS: requirements management system(3) Report documentation
The safety case documentation is prepared at several hierarchical levels, as shown in
Figure 1.4-3, so that the information can be accessed by readers depending on their needs andthe level of technical detail they require.
1-10

Figure 1.4-3 Overall structure of the safety case documentation

The safety case arguments are supported in more technical detail, progressing from theupper level to the lower level, while links and cross-references ensure traceability. The
present report corresponds to the “Main Report”, forming the central part of the safety case,but is also linked to detailed information contained in the “Supporting Reports”.
A “Summary Report” has also been produced (in Japanese) in order to allow less technicalreaders to quickly and easily grasp the main points of the safety case. Many individual
technical reports and scientific papers are referred to in this report, forming the technicalfundamentals of the safety case. The Main Report and Supporting Reports together, arecollectively called “the NUMO Pre-siting SDM-based Safety Case Report” as mentioned
previously. It should be noted that this technical report assumes a readership with a certain
level of technical expertise and background knowledge of geological disposal.

(4) Organisation of report preparation
For the production of this report, which required input from relevant research institutes anddomestic and international experts, the systematic approach illustrated in Figure 1.4-4 was
implemented. This ensured that the latest findings and technical development results wereeffectively captured, together with technical review and advice from experts outside of
NUMO. These efforts were aimed at ensuring the technical quality of this report.

1-11

Figure 1.4-4 Overview of safety case report production

In addition, NUMO published the ‘NUMO Safety Case Report for Review’ in October
2018 and the report was then peer-reviewed by the Atomic Energy Society of Japan (AESJ),in order to obtain an objective evaluation of its technical content. Based on the comments andsuggestions from the AESJ review team in December 2019 [17], the report was subsequently
revised (resulting in the present document).

(5) Abbreviations
The abbreviations given in Tables 1.4-1 to 1.4-3 are used throughout this report, for laws,
report titles and organisations relevant to geological disposal (both Japanese and international).
In principle, the units used in this report are SI (SI base and derived units). However, non-SI
units are also used as needed. A list of these is given in Table 1.4-4.

1-12

Table 1.4-1 International organisations relevant to geological disposal
Full nameInternational Atomic Energy Agency

IAEA

International Commission on Radiological Protection

ICRP

Organisation for Economic Co-operation and Development/Nuclear Energy Agency

International organisations

Abbreviation

OECD/NEA

Nuclear Waste Management Organization (Implementer, Canada)

NWMO

Säteilyturvakeskus (Regulator, Finland)

STUK

Posiva Oy (Implementer, Finland)

PosivaAutorité de sûreté nucléaire (Regulator, France)

ASN

Agence Nationale pour la Gestion des Déchets Radioactifs (Implementer, France)

ANDRA

Bundesministerium für Umwelt, Naturschutz und Reaktorsicherheit(Implementer, Germany)

BMU

Gesellschaft für Anlagen und Reaktorsicherheit (Regulator, Germany)

GRSStrål Säkerhets Myndigheten (Regulator, Sweden)

SSM

Svensk Kärnbränslehantering AB (Implementer, Sweden)

SKB

Eidgenossisches Nuklearsicherheitsinspektorat (Regulator, Switzerland)

ENSI

Nationale Genossenschaft für die Lagerung Radioaktiver Abfälle(Implementer, Switzerland)

Nagra

Environment Agency (Regulator, UK)

EARadioactive Waste Management, a subsidiary of the
Nuclear Decommissioning Authority (Implementer, UK)

RWM
NDA

Nuclear Regulatory Commission (Regulator, USA)

US NRCUS Department of Energy (Implementer, USA)

US DOE

1-13

Table 1.4-2 Domestic organisations relevant to geological disposal

Domestic organisations

Full name

Report abbreviation

Nuclear Waste Management Organization of Japan

NUMOJapan Atomic Energy Agency
Previously: Japan Nuclear Cycle Development Institute

JAEA

Before that: Power Reactor and Nuclear Fuel Development Corporation

JNC
PNC

Radioactive Waste Management Funding and Research Center

RWMC

Central Research Institute of Electric Power Industry

CRIEPI

National Institute of Advanced Industrial Science and Technology

AIST

National Institute of Radiological Sciences

NIRS

Japan Nuclear Energy Safety Organization

JNES

Federation of Electric Companies of JapanFEPCAtomic Energy Society of Japan

AESJ

Nuclear Regulation Authority

NRA

Japan Atomic Energy Commission

JAEC

Japan Nuclear Fuel Limited

JNFL

1-14

Table 1.4-3 Abbreviated list of key Japanese reports, laws and associated documents relevant
to geological disposal

Reports

Laws etc.

Full name

Abbreviation

Designated Radioactive Waste Final Disposal Act

Final Disposal Act

Law Enforcement Regulations on Final Disposal ofSpecified Radioactive Waste

Final Disposal Law Enforcement
Regulations

Act on the Regulation of Nuclear Source Material,
Nuclear Fuel Material and Reactors

Nuclear Reactors Regulation Act

Basic Policy on Final Disposal of DesignatedRadioactive Wastes

Basic Policy on Final Disposal

Designated Radioactive Waste Final Disposal
ProgramFinal Disposal Program

Regulations for the Safe Transport of Radioactive
Material*

Transport Regulations

H12 Project to establish the scientific and technicalbasis for HLW disposal in Japan – second progress
report on research and development for thegeological disposal of HLW in Japan

H12 Report

H17: Development and management of the technicalknowledge base for the geological disposal of HLW

H17 Report

Second progress report on research and development
for TRU waste disposal in Japan

TRU-2 Report

Safety of the geological disposal project 2010 - safe
geological disposal based on reliable technologies

2010 Report*Defined by NRA referring to ‘Regulations for the Safe Transport of Radioactive Material 2012 Edition,
IAEA Safety Standards Series No. SSR-6’.

1-15

Table 1.4-4 SI units and non-SI units used in the NUMO SC
QuantityTime
LengthMass
Thermodynamic temperature
Amount of substanceForce
Pressure, stress
Energy, work, heat
Power, radiant flux
Potential difference
Celsius temperatureRadioactivity
Dose equivalent
Time

Plane angle, phase angle
VolumeUnit system
SI base unit
SI base unit
SI base unit
SI base unit
SI base unit
SI derived unit
SI derived unit
SI derived unit
SI derived unit
SI derived unit
SI derived unit
SI derived unit
SI derived unit
Non-SI unit
Non-SI unit
Non-SI unit
Non-SI unit
Non-SI unit
Non-SI unit

1-16

Unit symbol
s
m
kg
K
mol
N
Pa
J
W
V
℃
Bq
Sv
min
h
d
y
°
l

References[1] JNC (Japan Nuclear Cycle Development Institute) (2000): H12: Project to establish thescientific and technical basis for HLW disposal in Japan; Project overview report, JNCTN1410 2000-001.
[2] OECD/NEA (1995): The environmental and ethical basis of geological disposal of longlived radioactive wastes, A collective opinion of the Radioactive Waste Management
Committee of the OECD Nuclear Energy Agency.
[3] JAEC (Japan Atomic Energy Commission) (1998): Basic concepts for high-level
radioactive waste disposal, Special Committee on High-level Radioactive Waste Disposal,
JAEC (in Japanese).
[4] JAEA (Japan Atomic Energy Agency) and FEPC (The Federation of Electric Power
Companies of Japan) (2007): Second progress report on research and development for TRU
waste disposal in Japan, JAEA-Review 2007-010, FEPC TRU-TR2-2007-01.
[5] NUMO (2013): Safety of the geological disposal project 2010, NUMO-TR-13-05.
[6] SCJ (Science Council of Japan) (2012): Response: Concerning high-level radioactive
waste disposal (in Japanese).
[7] JAEC (Japan Atomic Energy Commission) (2012): Future approaches to high-level
radioactive waste geological disposal (Opinion) (in Japanese).
[8] Advisory Committee for Natural Resources and Energy (2014): The interim report of theRadioactive Wastes WG, Radioactive Wastes WG, Nuclear Energy Subcommittee, Electricity
and Gas Industry Committee, Advisory Committee for Natural Resources and Energy (inJapanese).
[9] Advisory Committee for Natural Resources and Energy (2014): Reassessment ofgeological disposal technology based on the latest scientific knowledge: Geologicalenvironment characteristics and long-term stability of the geological environment, Geological
Disposal Technology WG, Nuclear Energy Subcommittee, Electricity and Gas Industry
Committee, Advisory Committee for Natural Resources and Energy (in Japanese).
[10] JAEA (Japan Atomic Energy Agency) (2015): Preliminary assessment of geologicaldisposal system for spent fuel in Japan - First progress report on direct disposal -, JAEAResearch 2015-016 (in Japanese).
[11] JAEA (Japan Atomic Energy Agency) (2008): Present status and future plan of researchand development in JAEA on partitioning and transmutation technology for long-lived
nuclides, JAEA-Review 2008-074 (in Japanese).
[12] Advisory Committee for Natural Resources and Energy (2017): Results of theexamination of the requirements and criteria for the presentation of regional scientific
1-17characteristics of geological disposal (summarised by the Geological Disposal Technology
WG), Geological Disposal Technology WG, Nuclear Energy Subcommittee, Electricity and
Gas Industry Committee, Advisory Committee for Natural Resources and Energy (inJapanese).
[13] METI (Ministry of Economy, Trade and Industry) (2017): Explanation material of thenationwide map of "scientific features" relevant for geological disposal,
https://www.enecho.meti.go.jp/en/category/electricity_and_gas/nuclear/rwm/pdf/explanation_material.pdf (accessed 2021-01-12).
[14] IAEA (2012): The safety case and safety assessment for radioactive waste, Specific
Safety Guide, IAEA Safety Standards Series, No. SSG-23.
[15] OECD/NEA (2004): Post-closure safety case for geological repositories, NEA No. 3679.
[16] OECD/NEA (2013): The nature and purpose of the post-closure safety cases forgeological repositories, NEA/RWM/R(2013)1.
[17] AESJ (Atomic Energy Society of Japan) (2019): Report of the Special Review
Committee on the NUMO safety case, Special Review Committee on the NUMO Safety Case,https://www.numo.or.jp/en/what/pdf/The_review_of_the_NUMO_safety_case.pdf (accessed2021-01-12).

1-18

2 SAFETY CONCEPTAs described in Section 1.4.2, NUMO’s current strategy for developing the safety casedescribes the basic approach to ensuring safety, and shows how it intends to achieve safe
geological disposal during each phase defined in the implementation programme [1]. In thischapter, after first setting out the requirements to be considered at the current stage of
geological disposal planning, the technical considerations for site characterisation, repository
design and safety assessment are described. Finally, the required management concepts andtools for integrating this work in an effective and quality assured manner are described andthe structure of the following, more detailed documentation is summarised.

2.1 Requirements to be considered when planning geologicaldisposal in JapanThe required safety features for geological disposal of radioactive waste at specific sites
depend on the characteristics of the waste and the specifications in laws and regulations, as
outlined in this section.
2.1.1 Radioactive waste subject to geological disposal
As mentioned in Section 1.2, radioactive waste for geological disposal in Japan includes
vitrified HLW and TRU waste1 (see Supporting Report 2-1 for details of the definition ofthese “designated wastes”). Immediately after production of HLW, both activity and theresulting thermal output are significant, so the waste is first safely stored for about 30 to 50
years before it can be accepted for disposal. In addition, TRU wastes are solidified, sealed andstored in containers before disposal, as specified in the Basic Policy on Final Disposal.
In order to carry out the design and safety assessment of the repository, further informationon the amount of waste generated and its characteristics are required, as provided below.
(1) Basic information on vitrified HLW
There are four sources of HLW: vitrified waste reprocessed and returned from overseas (byOrano (formerly AREVA NC) in France and Sellafield Ltd. (formerly BNFL) in the UK) andvitrified waste produced in Japan by JAEA and JNFL (Japan Nuclear Fuel Limited). Table
2.1-1 shows the standard specifications of such HLW [2] [3]. Further details of the properties
of HLW are given in Supporting Report 2-2.

1

The definition of TRU waste to be disposed of in a geological repository is slightly different according to the
Final Disposal Act and the Nuclear Reactors Regulation Act. The treatment of this is summarised in Supporting
Report 2-1.

2-1

Table 2.1-1 Standard specifications of HLW and fabrication canisters2 and number currently instorage (based on [2], [3], [4] [5], [6], [7])Producer
Totalradioactivity(Bq - at time of
production)

JNFL

JAEA

Orano

Sellafield Ltd  ≤ 2.17 × 1016
α ≤ 1.29 × 1014

  ≤ 1.5 × 1016
α ≤ 2.6 × 1014

(Representative)  ≤ 2.8 × 1016
α ≤ 1.4 × 1014

(Representative)  ≤ 4.5 × 1016
α ≤ 3.5 × 1014

Thermal power
(kW)

≤ 2.3

≤ 1.4

< 2.0

< 2.5

(at time of production)(at time of production)(at time of transportation)(at time of transportation)

Dimensions offabrication
canister (mm)Height: 1,340
Height: 1,040
Outer dia.: 430
Outer dia.: 430
Canister thickness: 6 Canister thickness: 6

Height: 1,340
Outer dia.: 430
Canister thickness: 5

Height: 1,340
Outer dia.: 430
Canister thickness: 5

Weight of filled
HLW canister
(kg)

500

380

492

550

No. of HLW
canisters instorage(December 2020)

346

316

1,310

520

Information on the evolving heat generation rate and radioactivity inventory of waste fromthe time of receipt in a repository is necessary for repository design and safety assessment. In
this report, such properties are established based on the following assumptions, with details
given in Supporting Report 2-3.
•••

When the JNFL reprocessing plant is operational3, the generated HLW fromexisting spent fuel (SF) or the SF generated by future nuclear power plants willaccount for the majority of the total inventory. Therefore, in this report, as in the
H12 report, JNFL specifications of typical vitrified HLW [8] will be used as areference for repository design and safety evaluation. This can be used to define thethermal output and radionuclide (RN) inventory of the waste at the time of disposal.
The standard specification of SF is that used for the design of the JNFL
reprocessing plant (fuel type PWR, burnup 45,000 MW days, initial enrichment
4.5%, specific power 38 MW, initial cooling period four years until reprocessing4)
in order to calculate the thermal output and RN inventory.
The storage period after fabrication of the HLW until it is received at the repository
is expected to range from 30 to 50 years according to the Basic Policy on Final
Disposal, but it is difficult to be more specific at the present stage. For this reason,
two storage periods, 30 and 50 years, are considered.

2

N.B. During the reprocessing of spent fuel, liquid HLW is calcined and then vitrified within a fabrication
canister – together referred to as a HLW canister for subsequent handling.
3
Construction of the JNFL reprocessing plant is planned to be complete in the first half of FY 2022 [9]. 346
HLW canisters are in storage as of December 2020 (produced during active tests of the reprocessing plant).
4
JNFL is currently planning to specify an initial cooling period 15 years or more until reprocessing [10]. For
studies of repository design and safety assessment in this report, a cooling period of 4 years is considered to be
the current standard specification for spent fuel. The study regarding the influence of these different cooling
periods on the inventory is shown in Supporting Report 2-3.

2-2

(2) Basic information on TRU waste
TRU wastes are low-level radioactive wastes generated during the operation anddecommissioning of SF reprocessing facilities and MOX fuel plants at JAEA and JNFL, andduring the reprocessing of SF outsourced to other countries. These wastes include various
types of materials such as metal, mortar, bitumen etc., which differ in their form and RN
content. For this reason, wastes are classified into four groups based on their characteristics[11] [12], as shown in Table 2.1-2, referring to the TRU-2 report [13]5.
Table 2.1-2 Group classification and characteristics of TRU waste(expanded from NUMO, 2011 [12])The following characteristics of each group should be taken into account in the design of arepository and associated safety assessment.
•Group (Gr.) 1 comprises the silver adsorbent used to capture off-gas iodine, which isimmobilised with mortar and has extremely low heat output. It contains a significant
amount of radioactive iodine (I-129), a long-lived radionuclide with assumed highsolubility and low sorption.

The TRU wastes generated in the reprocessing process of spent fuel in France have been returned as “CSD-B”,
containing vitrified residues from low-level enriched liquid waste, and “CSD-C”, containing solidified materials
such as hulls. These were included in Gr.2, which has a similar waste form, considering convenience of
operation [12]. The TRU wastes generated in the course of reprocessing in the UK will be substituted for
vitrified high-level radioactive wastes with equivalent radiation effects [14], and these wastes will be included in
the HLW inventory.
5

2-3

•

•
•

Gr.2 comprises compressed metal cladding stripped from SF (hulls and ends), sealed
in a stainless steel canister. Its thermal output is relatively high and it contains a large
amount of radioactive carbon (C-14), which is an intermediate half-life nuclide withassumed high solubility and low sorption.
Gr.3 results from the solidification of liquid waste generated during reprocessing. It is
conditioned in a matrix of bitumen or mortar and has a relatively low heat output. It
contains nitrate, which can affect the engineered barrier and host rock containment.
Gr.4 consists of miscellaneous wastes generated during reprocessing and MOX fuelfabrication, which are solidified or encapsulated in mortar, etc. There are two types ofsuch waste: one with relatively low (Gr.4L) and the other with relatively high (Gr.4H)
thermal output [11].

More details on the characteristics of TRU waste are provided in Supporting Report 2-4.
The thermal output and RN inventory at the time of waste production were documented by
NUMO in 2011 [12] [15]. Further, since the TRU storage time before disposal is not defined,
in line with the TRU-2 report [13] thermal output and RN inventory are calculated at 25 years
after production [9] [11], which is taken as the assumed emplacement time (see Supporting
Report 2-3 for more background).
The thermal output and radioactivity inventory of HLW and TRU waste described above
will be regularly reviewed to reflect production, storage volume and storage period of basedon the evolution of nuclear energy utilisation and the operational status of reprocessing
facilities.
(3) Waste acceptance criteria
In order to construct a safe repository, NUMO will establish the acceptance criteria for
HLW and TRU waste, based on the progress of site selection and the design of the repository,
with sufficient time before the application for the project licence. The relevant laws andregulations that should be taken into account in setting these acceptance criteria, and theacceptance criteria for HLW in similar projects for reference, are documented in Supporting
Report 2-5. It is understood that part of the HLW is already vitrified, whereas most of the
TRU waste is not yet conditioned. There is no decision as yet on how to condition thesewastes and whether additional conditioning (e.g. of bituminised waste) or repackaging would
be needed. Such potential needs will strongly depend on what repository concepts are finallyselected.
(4) Required repository capacity
As shown in Table 2.1-1, as of December 2020 a total of 2,492 HLW canisters are storedsafely at reprocessing facilities at both JNFL in Rokkasho-mura, Aomori prefecture, and
JAEA in Tokai-mura, Ibaraki prefecture. However, it should be noted that spent fuel,corresponding to approximately 25,000 canisters of HLW, has already been generated.
According to the “Plan for the Final Disposal of Designated Radioactive Wastes” (hereafter
the “Final Disposal Plan”) published in 2008, about 3,231 m3 of TRU waste is stored at both
JAEA and other facilities in Japan as of 2007.
The Final Disposal Plan requires that a disposal site can accommodate of the following
volumes of waste:•

HLW: 40,000 canisters or more (to be disposed of at a rate of about 1,000 per year).
2-4

•TRU waste: ≥ 19,000 m3.

It is difficult to estimate the total generation of radioactive waste to be disposed of becausesuch numbers depend on future nuclear power policies to be established in Japan. In thisreport, based on the above, 40,000 canisters of HLW is assumed for repository design andsafety assessment. For TRU waste, the total estimated amount generated in Table 2.1-2 is18,084 m3, so this is modified to obtain a disposal volume 19,000 m3 or more specified by the
Final Disposal Plan.
(5) Co-disposal of HLW and TRU waste
There are clear practical advantages of co-locating repositories for HLW and TRU waste –for example in terms of reducing required efforts for site characterisation and sharing of someinfrastructure, although the potential interactions between these wastes needs to be considered[12]. With a focus on safety, assuming that conclusions would be applicable also to separate
disposal facilities, in this report co-located repositories for HLW and TRU waste are assumed.
2.1.2 Required safety functions of the repository(1) Safety features of geological disposal
As described in Section 1.1, a deep geological setting has the function of physicallyisolating waste from the biosphere whilst also suppressing release and migration of RNs
contained in groundwater. A repository comprises both engineered and natural barriers in asuitable geological setting, which function together to safely isolate the radioactive wastefrom the human environment, and contain RNs until long after closure, during which time
most will have decayed and any residual radiological risk is acceptably low. Such a safetyconcept is based on a multi-function/multi-barrier system and is common to all advanced
national programmes (in addition to being the international standard), and is explicitly
specified in the Basic Policy on Final Disposal.
The basic concept of ensuring post-closure safety by geological disposal can be specified
in terms of the key safety functions of “isolation” (removal from humans/the biosphere) and“containment”6. These functions can be related to the roles of different components of therepository system in ensuring safety. In recent years, this concept has proven important for
directly demonstrating the link between conceptual explanations of the repository as
mentioned above and practical actions, such as repository design and safety evaluation, within
the framework of a safety case [16] [17].
In order to ensure safety during construction and operation to the point of closure, thesafety features required are generally similar to those necessary for other nuclear facilities orunderground structures. The post-closure, multi-barrier safety features described here arefundamentally the same as those commonly used in other repository concepts for HLW andTRU waste developed in other national programmes.

In this report the word “containment” includes both complete containment and the ability to constrain/retard
releases. In many other safety cases the latter is called “retention”. See also further discussion in this section.
N.B. This footnote is not included in the Japanese version of the report.
62-5

(2) Required operational safety functions
During the operational phase, safety functions ensure the safety of local residents andrepository workers from both radiation risks and hazards other than those associated withradiation.
Those safety functions related to radiation protection during the operational phase aresummarised in Table 2.1-3. Operational containment refers to the confinement of radioactive
material in a restricted area to prevent its release outside the facility during the operational
period of handling the waste until closure.
Table 2.1-3 Required operational radiological safety functions(based on NUMO, 2011 [18])Basic concept

Containment
duringoperation

Radiation
shielding

Safety function

Description

Prevention of leakage of RNs fromwaste

Complete containment7 of the waste during
the operational period

Prevention of release of RNs from therepository

Prevention of release of radioactive material
due to any perturbations during waste
handling operations

Reduction of radiation dose

Effectively complete shielding of externalradiation from the waste

Containment of radioactivity within the waste results primarily from the wasteconditioning and packaging, complemented by the robust systems used to transport waste to
and within the repository.
Although transport, storage and handling systems are designed to minimise the risk ofperturbations, these can never be completely precluded and hence a further safety function iscontainment within the surface or underground facilities of the repository in case of any
accident that breaches the waste package. For such cases, containment measures will beestablished for all relevant facilities after evaluation of both the probability of perturbations
that could lead to leakage of radioactivity and the consequences of such an event.
Even when RNs are contained within the waste package, there is always a potential risk
due to the penetrating nature of particular radiation (gamma rays and high energy neutrons).
Any such health hazard to workers is reduced by assuring that sufficient shielding is present
to reduce external radiation dose rates to acceptable levels, and that this is complemented byappropriate use of monitored, radiation-controlled zones and monitoring of worker doses. The
layout and design of the repository assures that there is no dose to surrounding populations
from this source.
Incidents not involving radiation, referred to as industrial accidents, include those
impacting the public around the facility and those impacting only the workers engaged in theconstruction, operation, and closure of the repository. The former can include secondary
impacts around the facility due to fire or other incidents (smoke, fumes, etc.) and also traffic
accidents involving off-site vehicles during the construction and operation of the facility.

7

Complete containment is generally assured when the container and/or matrix is intact. N.B. This footnote is notincluded in the Japanese version of the report.

2-6

With regard to industrial safety of workers, the repository is designed to prevent accidents
from developing and ensuring maintenance of a healthy working environment during theperiod from the geological survey, construction, operation until final closure of the site. The
safety functions related to industrial safety of the repository during the pre-closure phase areshown in Table 2.1-4.
Table 2.1-4 Required operational occupational safety functions(based on NUMO, 2011 [18])Basic concept

Prevention ofindustrial accidents

Maintenance of
healthy working
environment

Safety function

Description

Prevention of occurrence andpropagation of incidents

Establish counter-measures to prevent
occurrence of events which could lead towork-related accidents

Evacuation routes
established in case of
accidents

Safe havens and evacuation routes
established for all relevant accident
scenarios

Maintain conditions
appropriate to worker healthand safety

Ensure comfortable and healthy working
conditions

Prevention of industrial accidents includes safety measures to reduce the impact of natural
perturbations caused by earthquakes, tsunamis, pyroclastic flows, landslides, etc., as well as
possible operational incidents such as rock-falls, fires and explosions. Included here are alsomeasures taken to respond to incidents (so that they do not develop further to accidents) andprotective actions in the event that accidents do occur, such as safe underground evacuationroutes. Special activities related to maintaining a healthy working environment are
particularly associated with controlled surface or underground facilities, where temperature,humidity, particulate concentration, noise level, etc. are set at levels established to be
comfortable for workers.
Required pre-closure safety functions can be subdivided in terms of safety features ofrepository design components, as discussed further in Section 4.2.4 (1). The safety functions
to be assigned to each component and the design requirements for the safety of the repository
before closure are given in Chapter 4.
(3) Required post-closure safety functions
After closure, the repository will have isolation and containment roles, as summarised in
the safety functions shown in Table 2.1-5.

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Table 2.1-5 Required post-closure safety functions(based on IAEA, 2011 [19])Basic Concept

Isolation

Safety function

Description

Protection from significant
effects of natural perturbing
phenomena

Assuring sufficient depth to avoid risks due to surface
perturbations for the time when waste toxicity is highReduction of the likelihood of
human intrusion

Emplacement deep in a suitable geologicalenvironment reduces the risk of inadvertent human
intrusion

Restriction of RN leaching

Retention8 within the waste package for the period ofhighest toxicity and slow release thereafter, assured byhydrogeology and geochemistry of the deep
underground setting

Restriction of RN migration

Delay and reduction of releases to the biosphere due toretardation during geosphere transport

Containment

In terms of isolation, the assurance of protection from geological perturbations isparticularly important in Japan due to its location relative to active tectonic plates, which
results in significant potential for volcanism, fault movement and uplift/erosion at somelocations. Careful siting and disposal at sufficient depth is thus required to ensure that the
selected geological setting would not be significantly perturbed over the assessment timescale.
In addition, the geological environment should contribute to reduction of the risk ofanthropogenic perturbations – predominantly by excluding areas containing mineral resourcesthat might be exploited in the future, when knowledge of the repository has been forgotten.
The depth of disposal is also a factor here, ensuring that it is sufficient to avoid potential
impacts due to conventional civil engineering activities at times when there is no longer
institutional control of the disposal site. These safety functions related to isolation are one roleof the geological environment, which will be confirmed through investigation and evaluation
during site characterisation, as discussed further in Chapter 3.
Containment will be ensured by safety functions that involve, firstly, containment of mostRNs in and around the engineered barriers for an extended period and, secondly, restriction of
the rate of RN release thereafter. In terms of containment of RNs, the low water flow at depth
and suitable geochemical conditions (e.g. chemically reducing groundwater) support thelongevity of containers that completely contain the waste. This ensures a decrease in toxicity
and thermal output due to radioactive decay of shorter-lived isotopes [17] and, thereafter,
together with buffer, backfill and plugs, limits the rate of RN release from the engineered
barriers for the expected natural barrier properties.
After release from the engineered barriers, the site geology plays further roles in delaying
and reducing RN release concentrations due to slow groundwater flow rates, long transport
paths, retardation and dispersion during transport (see Chapter 3, Section 3.1.1 (2)) anddilution at the geosphere-biosphere interface.
The safety functions assigned to each component and the design requirements for the longterm post-closure safety of the repository are discussed further in Chapter 4.

8Here the definition of retention means highly effective immobilisation (containment) within a specific barrieror zone, but allows for trace releases. N.B. This footnote is not included in the Japanese version of the report.

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(4) Spatial scales over which safety functions apply
The size of the planar repository area (footprint) for disposal of 40,000 HLW canisters wasestimated in the past to be about 3 km × 2 km and that for 19,000 m3 of TRU waste to beabout 0.5 km × 0.5 km, with the surface facilities requiring about 1 × 1.6 km, assuming codisposal [12]. Although the actual footprint will depend on site geological and environmental
conditions, on the basis of past study cases this would be in the order of a few km2, evenassuming co-disposal.
Depending on the geological setting, the surface facility may be directly above orsomewhat displaced from the disposal panels, which themselves may lie at a set depth (below
300 m and up to depth of a km or so) or be distributed between multiple levels within thisrange. Taking into account the barrier roles of surrounding rock, the investigations to confirm
the isolation and containment functions would cover an area of several km × several km and a
depth of around 1 km or more.
(5) Safety function timescale
Considering that the radioactivity of the waste remains significant for an extremely longtime (Figure 1.1-1), it is necessary to consider the timescales over which repository safetyfunctions should operate. The disposal system will be put in place during the construction andoperational phases and then slowly evolve post-closure, with recovery of the original
saturated hydrogeological and reducing chemical environment, long-term degradation of theengineered barriers and gradual alteration of the geological environment due to tectonic
movements. It is thus necessary to set the repository safety functions considering such
temporal changes, as discussed below.
The time required for the disposal project will depend on site environmental conditions,but can be roughly estimated as follows: 20 years for site investigation, 10 years forrepository construction, 50 years for operation, and 10 years for repository closure [18].
During the period from repository construction until closure, the expected radiological safety
and general occupational health and safety functions, as given in Tables 2.1-3 and 2.1-4
respectively, will be assured by appropriate site selection, design, construction and operationof the repository, and confirmed by operational safety assessment.
Following backfilling and closure of the repository, any open void space will gradually resaturate and trapped oxygen will be consumed, to allow recovery of original reducing
conditions. The extent of the transient phase in the natural and engineered barrier system(EBS) depends on the amount of heat generated and the specification of particular engineered
barriers, with past studies suggesting periods of several tens to a few hundred years [12][20].
When considering the potential for gas generation, these times could be even longer.
Quantification of the performance of the multi-barrier system for high thermal output waste
during the initial transient period is complex and, to reduce uncertainties, it is thus preferable
that the EBS has the function of excluding contact between the waste and the groundwater
during this period, so that release and transport of RNs do not have to be assessed. However,it should be noted that this may not be the case for Gr.4H – releases could occur before thethermal transient has passed.
Only after the complete containment function is lost, will waste come into contact withgroundwater and RNs begin to dissolve. In practice, the containment functions of the EBS
will gradually evolve, with changes in performance being very slow and the retardation
function of the engineered barriers lasting for a very long time, if the favourable geological
2-9environment persists. In a properly selected geological environment with favourable
characteristics (e.g., low groundwater flow), it is considered that the migration-inhibiting
function of the engineered barriers can be assured for a very long period of time, especially as
the multiple barriers provide robust performance even when the functions of some individualcomponents are degraded. During this time, most RNs will remain within, or in the vicinity of,
the engineered barriers s. Furthermore, any RNs released will be retarded further in thegeological environment. Thus, during site selection, it will be important to verify (in addition
to its isolation functions) that the long-term retention function of the geological environmentis assured. However, it will still be necessary to evaluate the transport behaviour of RNs fromthe EBS to the surface, and to assess the impact (dose) of RNs released to the biosphere, inorder to confirm whether the safe functioning of the repository can be ensured.
Continuous plate tectonic movement which causes phenomena such as volcanic activity,
faulting, and uplift and erosion, is expected to proceed at a uniform speed in one direction,and the occurrence of such natural phenomena and their effects on the geological environment
are likely to continue for a period of up to 100 ky [21] [22] (see Supporting Report 3-1). Suchan assumption, which was also in the H12 Report [20] and reviewed by the Government [23],is considered to be acceptable [24], and its validity has been reconfirmed in the latest
compilation of findings [25] since the H12 Report [20]. Therefore, it is considered that thefavourable characteristics of the deep geological environment at appropriately selected sites
can be maintained, avoiding significant effects of perturbing phenomena, and that the longterm isolation and containment functions expected of the geological environment can beensured (see Section 3.1).
Between around 100 ky to 1 My, evolution of the geological environment characteristics
will be assessed by models or extrapolations of paleo-geological observations, based onevidence for the long-term continuity of events and processes governed by tectonic platemovement over the last several million years [26] [27]. Current tectonics show a generally
consistent picture over a period of several 100 ky to 1 My, although regional differences may
be found [22] [28]. The safety of a future repository can thus be evaluated on the basis of thisknowledge base, bearing in mind the inherent uncertainties associated with geologicalevolution over such a long period of time.
The occurrence of natural perturbing phenomena and their effects on the geologicalenvironment at a site can be assessed, together with their uncertainties, using a combination of
extrapolation, analogy and probabilistic methods (see Section 3.2.3 (2)). In areas that have notbeen significantly affected by disruptive natural phenomena, it has been found thathydrogeological and chemical conditions favourable for geological disposal have been
maintained for more than a million years [29] [30], despite the effects of inevitable fault
movement, uplift, erosion and sea-level change (see Section 3.1.3 (2)). Such paleohydrogeological evidence indicates that there is little likelihood of sudden or abrupt changes
in the characteristics of the deep subsurface environment, because disturbances areconstrained by the inherent buffer functions of the geological environment [31]. By
integrating these findings with the results of the previous assessments, it can be argued thatthe expected isolation and containment functions of the geological environment are likely to
be maintained for several hundred thousand years or more and this can be confirmed byrigorous site-specific assessment. In addition, it is necessary to assess the transport behaviourof RNs in the geosphere, taking into account associated uncertainties, and to assess the effectsof radiation on the biosphere (dose), in order to confirm that the required safety functions of
the repository can be assured.

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For longer timescales, beyond a few hundred thousand to a million years, the uncertainties
associated with the assessment of the required safety functions of a repository are even greater.
Both the scientific evidence for the occurrence and impacts of natural perturbations and thepersistence of driving tectonic plate movements is inherently more limited. Over such very
long timescales, it is less useful to assess future human safety by calculating doses, and more
important to discuss safety in terms of other indicators, such as the relative toxicity ofremaining radioactivity in the repository and that naturally occurring in the host rock.
The basic concepts for site selection, repository design, and safety assessment aredescribed in detail in Sections 2.2, 2.3 and 2.4 respectively, taking into account such safetyfunctions for different timescales.
2.1.3 Meeting the requirements of laws and regulations
Geological disposal of radioactive waste in Japan is regulated by the Final Disposal Act,
the Basic Policy on Final Disposal and the Final Disposal Plan set by METI. In addition,
safety regulations will be separately determined by the Nuclear Regulation Authority (NRA),
which is the regulatory body. NUMO is responsible for formulating and implementing the
project in accordance with these laws and regulations.
(1) Phased implementation
Figure 2.1-1 illustrates the stepwise site selection process as regulated by the Final
Disposal Act. This involves Literature Surveys (LS) of volunteers, selection of Preliminary
Investigation Areas (PIAs), from these, selection of Detailed Investigation Areas (DIAs) and,
finally, a site for application for a repository construction licence (which also covers theoperational phase).

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Figure 2.1-1 Japan's phased implementation process for geological disposal(Based on Advisory Committee for Natural Resources and Energy, 2014 [32])When selecting an area to move to the next stage of investigation, it has to be ensured thatthe selection requirements (statutory requirements) specified in the Final Disposal Act (Table
2.1-6) are met. In addition, it is necessary not only to satisfy such statutory requirements, butalso to comply with any requirements for ensuring safety that will be specified by regulatory
bodies in the future. Progressing from one investigation stage to the next involves intensive
interaction with stakeholders, in particular aimed at obtaining the understanding of the region.
For implementation, METI will determine the acceptability of sites selected forinvestigation, taking account of opinions of the prefectural governor and the mayor of themunicipality with jurisdiction over the site. Also, NUMO must provide the results of each
investigation stage in the form of a report for review by relevant prefectures, and, if issues areraised, it is stated that sites should be selected with consideration of these. Thus, NUMO will
not proceed to the next siting stage if the mayor of the municipality or prefectural governor is
opposed, regardless of the result of the investigations.
In order to increase the transparency of the selection process and enhance safety-related
considerations, NUMO will establish selection criteria for both the preliminary investigations(PIs) and detailed investigations (DIs) to judge eligibility, based on the statutory requirements.
The selection criteria established will be announced before the start of each survey stage.

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Table 2.1-6 The selection requirements specified in the Final Disposal Act
Siting stage

Selection requirements

Selection of PIA(during LS)• There should be no record of significant movement in geologicalformations due to earthquake or fault activity, igneous activity, uplift,
erosion and other natural phenomena• The possibility of significant movement in the future due to earthquake orfault activity, igneous activity, uplift, erosion and other natural phenomena
should be small• There should be no record of unconsolidated Quaternary deposits at
appropriate depths• There should be no record of mineral resources that are economicallyvaluable

Selection of DIA(during PI)• The host geological formation has not been subject to significant
geological change for a long period of time due to natural phenomena suchas earthquakes• The host geological formation is suitable for excavation• If there are active faults, fracture zones, or groundwater flowing in thesubject stratum, etc., there is little risk that these will adversely affect thetunnels and other underground facilities• Other matters specified by an Ordinance of the Ministry of Economy,
Trade and Industry

Selection of repository site
(during DI)• The underground facilities are not likely to be subjected to abnormalpressure in the target formation, and its physical properties are expected tobe suitable for repository implementation• The underground facilities are not likely to be subjected to abnormal
corrosive effects in the target geological formation, and the chemical
properties are expected to be suitable for repository implementation• There is no risk of groundwater flow interfering with the functioning of
the underground facilities• Other matters specified by an Ordinance of the Ministry of Economy,
Trade and Industry

For the LS phase, in 2002 NUMO published siting factors for selecting PIAs [33]. Since
then, as described in Section 1.3, the Nationwide Map was published which, together with theassociated reports produced by the Geological Disposal Technology [21][34] and Radioactive
Waste Working Groups (WGs) [32], provides confirmation on how to proceed with the LS
stage and subsequent site selection. If a municipality applies for a LS, this will be initiatedbased on the report of the WG on Geological Disposal Technology [34] and an establishedprocedure [35].
After the selection of a repository site, NUMO will move to a licensing process, startingfrom project permitting, through subsequent safety reviews for construction, operation,
closure, management after closure and, finally, discontinuation of any management role inaccordance with the Act for the Control of Nuclear Source Material, Nuclear Fuel Materials
and Nuclear Reactors (hereafter the “Nuclear Reactors Regulation Act”). Afterdiscontinuation of NUMO’s role, the Japanese government will assume all responsibility forsite access control, record-keeping, etc9. Currently, project implementation is estimated to
Regarding “post-closure monitoring” illustrated in Figure 2.1-1, there is no established scientific rationale forcontinuing monitoring after the completion of closure measures and associated confirmation of long-term safety.
9

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take over 100 years. Therefore, NUMO identifies intermediate milestones and implementation
items to achieve the goals of each phase, based on an overview of the total project. This
allows required technology development to proceed in a systematic and stepwise way. In the2010 technical report [18], a project implementation plan for ensuring safety and developmentof required technology was presented in the form of a road map.
(2) Assuring reversibility and retrievability
The OECD/NEA has compiled an international assessment of disposal concepts and thecurrent status in terms of reversibility and retrievability [36]. Reversibility is defined as ameans of leaving options open for future generations and providing flexibility in waste
management, thus leaving the possibility open to reverse decisions if required. In addition,
this option is also considered to increase confidence in the implementation process.
Retrievability is a technical measure to provide reversibility of waste emplacement, inparticular the ease and safety of such a reversal, were it required. However, any increase insocial acceptance of geological disposal resulting from technical modifications to ease waste
retrieval needs to be balanced against both a potential decrease in operational or post-closure
performance and financial costs. Thus, efforts to improve reversibility and retrievability will
require trade-offs with other fundamental project requirements, which differ depending on theboundary conditions of different national programmes [36].
Also, on the basis of such international debates, in Japan there has been discussion on theappropriate level of reversibility and retrievability for geological disposal projects – forexample, by the Nuclear Safety Commission (2000) [37] and the Advisory Committee for
Natural Resources and Energy (2008) [38]. There is a consensus that it is fundamentallypossible to recover waste until repository closure but, in order to carry this out safely andefficiently, adopting a design that explicitly includes this functionality is considered important.
As a result, NUMO will develop designs that maintain practical retrievability until the closure
plan is approved [18]. As noted by the Radioactive Waste WG [33], this provides an option toreview and include future generations in decision-making related to final disposal. Further, in
2015, the revision to the Basic Policy on Final Disposal specified that reversibility andretrievability should be assured until closure of the repository, in order to give future
generations the ability to implement an improved disposal concept should one arise. Thus, thegovernment and various research institutes conduct studies on the effects of maintaining such
retrievability until final closure for all designated radioactive wastes.
NUMO is aware that assessing reversibility and retrievability for such a long project period
has to consider the possibility of changes in social conditions, policies and stakeholder
requirements. As the implementing agency, it is thus important to ensure the robustness andflexibility of NUMO’s organisational system, including considering the human and economic
resources required.
(3) Safety regulations
Basic policy on the formulation of safety standards and guidelines for geological disposalof radioactive waste was issued by the Nuclear Regulation Authority [37], noting that it
would be important to formulate required detailed safety standards and guidelines in steps, inHowever, monitoring may be carried out as necessary to improve public confidence in the geological disposal
system [18].

2-14response to the progress of site selection, site-specific situations and developments of scienceand technology. In the amended Nuclear Reactors Regulation Act (2007), a regulatory
framework was established in accordance with progress of the project, as shown in Figure2.1-1, involving obtaining a permit for the implementation of the project; authorisation ofdesign and construction methods and the closure plan; and regular reviews of theimplemented repository (safety reviews). The 2011 accident that occurred at TEPCO’s
Fukushima Daiichi Nuclear Power Plant has resulted in a review of the entire safety
regulatory regime in the nuclear industry: a major revision of the Nuclear Reactors Regulation
Act was carried out in 2012 and regulations concerning measures to prevent serious accidents
have been implemented for nuclear facilities.
In a revision of the Nuclear Reactors Regulation Act in 2017, a system has been
established to limit the underground disturbance of land above and around the repository.
Currently, the Nuclear Regulatory Authority is developing new regulatory standards that are
expected to be available at the time of selection of PIAs and will be extended to cover later
project milestones. NUMO will follow such developments together with relevant international
trends to ensure the availability of the necessary technology to be able to respond
appropriately.
With regard to the environmental impact aspects of the geological disposal project, it
should be noted that, as currently specified, a repository based on the Final Disposal Act is
not subject to the Environmental Impact Assessment Act [39]. Nevertheless, in each phase of
the investigation, environmental impacts will be assessed and NUMO will take all actions
required to protect the environment around the repository.
Based on the requirements noted in Sections 2.1.1 to 2.1.3, Section 2.2 expands on site
selection, Section 2.3 on design of the repository, Section 2.4 on safety assessment and,
finally, Section 2.5 on the management tools needed for effective implementation of the
programme.

2.2 Site selection strategy
Site selection aims to ensure that the geological setting provides the key safety functions ofisolation of radioactive waste from the human environment, ensuring that engineered andnatural barriers form an effective multi-barrier system to ensure RN containment and avoiding
or preventing the effects of natural perturbations on the safety of construction and operationof the repository. The long-term stability of the geological environment is a key factor,
assured by extrapolating site-specific geological evolution from the past to the present andthus build understanding of how it will develop in the future.
2.2.1 Siting aspects to ensure operational safety
The integrity of the repository needs to be maintained in order to secure its operationalsafety functions. Factors that influence the integrity of both underground and surface facilities
include natural hazards, such as earthquakes, tsunamis, pyroclastic flows and landslides. In
addition, factors such as rock-falls could impact underground facilities. These will be
evaluated with respect to the legal requirements to exclude areas with:•

Unconsolidated Quaternary sediments at a depth of 300 m or more, considered topose a problem for construction, maintenance and operation of the undergroundfacilities.
2-15

•
•

Likelihood of perturbing events such as rock bursts, significant flooding, gas
inflow, etc., which influence the practicality and safety of construction andoperation over a repository lifetime of about a century.
An unacceptable likelihood that major natural perturbations, such as tsunamis or
pyroclastic flows, could occur.

After recognising the level and extent of the possible influences at a proposed site,
engineering measures will be considered based on those already developed for related nuclear
facilities, underground civil engineering structures, etc. In cases where significant
perturbations are likely, and these cannot be managed by engineering counter-measures
should they occur, the location would be excluded from further investigations.
2.2.2 Siting aspects to ensure post-closure safety(1) Assuring isolation functions
Features that assure the isolation of waste in a repository from the human environment are
the depth below the surface ensuring siting within a stable rock formation and the absence of
natural resources that could lead to inadvertent human intrusion. Factors that could impair
isolation include uplift/erosion and magma intrusion, the extent and probability of which will
be very site specific.
In order to acquire sufficient information about such perturbing phenomena, natural eventsand processes with the possibility of impairing the isolation function are investigated overwide areas (approximately tens of km × tens of km) surrounding any site being investigated.
Areas with risks of significant perturbations over the next 100 ky are excluded. From thiswide area, based on literature review and site investigations, a potential site with minimal
probability of such hazards will eventually be selected to host the repository (several km ×
several km). It also has to be shown that the risk of human intrusion is low, due to the absenceof potentially exploitable natural resources. If it cannot be concluded that a sufficiently large
potential repository construction site is free from such risks, even considering tailored
repository concept options to reduce the areas of favourable rock required (e.g., compact or
multi-level repositories), the location would be excluded from further consideration.
(2) Assuring containment functions
The geological environment is intended to effectively confine RNs, i.e., most RNs remain
or decay away within the repository EBS and its vicinity over relevant timescales. The small
fraction of RNs that migrate to the biosphere are released in sufficiently low concentrations
that they do not have a significant radiological impact on humans or the environment, being
well below the permitted levels. Such a containment function results from the mutuallycomplementary action of the natural barrier provided by the geological environment and thespecific engineered barriers of the repository, as discussed in Section 2.3. The strategy forrepository design is presented in the following section but, from the viewpoint of containment,a site needs to be selected with a geological environment which provides both conditions thatallow the designed EBS to fulfil required safety functions and also has favourable featuresthat limit migration of RNs (with a focus on groundwater release scenarios).

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Thermal, hydraulic, mechanical and chemical (THMC10) conditions, including theirpotential evolution over time, will be characterised during investigation of the geologicalenvironment [21]. The conditions of the geological environment favourable to engineered
barrier performance and/or restricting RN migration include:••••

Lower temperatures at repository depth (e.g., ensuring buffer longevity).
Slow groundwater flow (reducing solute transport rates).
Sufficient mechanical strength (reducing deformation of tunnels).
Favourable groundwater chemistry (reducing corrosion or waste degradation rates).

A repository is designed such that it sufficiently fulfils multi-barrier safety functions. If
such a design is possible, even when considering various other design requirements andrestrictions, a specification of a repository concept (or concepts) needs to be assessed toensure that long-term safety will be provided.
For any repository design, the safety functions of the multi-barrier system should beverified by safety assessment (see Section 2.4). In addition, if necessary, the results of thesafety assessment are fed back to the design of the repository, to iteratively improve specificsafety functions. If it is shown that repository safety can be demonstrated, the site may be
selected as a potential repository host and specification of a disposal system tailored to it
developed. If more than one suitable site is found, several sites may go forward to the DI
stage. The procedure for subsequent site selection is not further discussed in this document;
the findings of the safety assessment will be a key, but not the only, input to such a selection
decision.
2.2.3 Spatial scales during stepwise site selection
Chapter 1 presented the Nationwide Map (Figure 1.2-2), which identifies regions withpreferred geological characteristics and also those suitable from the viewpoint of wastetransport. This is a good starting point for NUMO’s site surveys, but these need to besystematically developed from a nationwide scale to regional comparisons. The regional
databases also include material that is not more widely available, such as chemical properties
deep underground, which could not be assessed in the map [34]. The stepwise assessment of
all relevant geological conditions during the LS, PI and DI stages is shown in Figure 2.2-1,
indicating how they are analysed in terms of repository construction and safety in order todetermine the scientific suitability of sites.

10

N.B. C also implicitly includes biological processes. N.B. This footnote is not included in the Japanese versionof the report.

2-17Figure 2.2-1 Stepwise characterisation of the geological environment, design of the repositoryand safety assessment(Modified from Advisory Committee for Natural Resources and Energy, 2017 [34])At the LS stage, the focus is on using information available on a nationwide scale to assess
exclusion factors, such as distribution and activity of Quaternary volcanoes and active faults,
extent of likely uplift and erosion, and presence or absence of natural resources. In addition,
for non-excluded areas, the geological environment characteristics are synthesised in a threedimensional site descriptive model11 (3D SDM) which also schematically represents its
expected long-term evolution.
At the PI stage, a series of geological investigations will be systematically implemented,
targeting both the PIA and its surroundings [33]. Specifically, geophysical and borehole
surveys will determine the local extent of the influence of Quaternary volcanoes, active faults,
and uplift and erosion to confirm the LS assessment of eligibility. In addition, geologicalstructures and characteristics of the deep environment will be determined to improve site
understanding and evaluate its long-term evolution. This allows the 3D SDM to be refined,
and a conceptual model for the long-term evolution of the geological environment to be
developed (“4D SDM”).
In the first half of the DI phase, more detailed information about the candidate host rock(s)is obtained by surface-based investigations, to further refine the 4D SDM. In the second half
11

A SDM is a synthesis of geology, rock mechanics, thermal properties, hydrogeology, hydrogeochemistry andthe surface system, describing geological structures and various properties, considering past and on-goingprocesses”. However, the SDM has several different parts (e.g. geological structures, hydrogeological properties,
thermal properties,…), that can be differently represented on different scales (with higher resolution in the moredetailed scales), and aspects of the SDM could be represented by numerical simulation models (like a DFN
groundwater flow and migration model for a part of the system). Still all these components need to be consistent(e.g. features of different hydraulic properties in the hydrogeological representation should reflect the geologicalfeatures of the site).

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of this phase, geological characterisation is extended by constructing an undergroundinvestigation facility (UIF) to confirm that the candidate host rock meets the legal
requirements, and to obtain detailed information important for repository design and safetyassessment, such as that related to RN migration and retardation. Based on the results of theseinvestigations, repository designs and safety assessments are updated to support selection of arepository construction site.
Figure 2.2-2 schematically illustrates the scope of such three-phase site characterisation,
narrowing in to the preferred repository location as it progresses from LS to PI to DI, while
the level of detail increases. During all stages, the surface environment and other “nongeological” factors of importance for the site selection will also be characterised.

Figure 2.2-2 Three stages of site investigation and their relationship to site descriptive model
scales (regional, repository and panel). Not illustrated in the figure is that regional studies (e.g.,of tectonics and seismicity) may need to continue also in the DI stage.
(Modified from Advisory Committee for Natural Resources and Energy, 2017 [34])2.2.4 Development of site descriptive models (SDMs)
NUMO has to integrate key properties of the geological environment into a representativemodel, incorporating attributes of a site that extend over large areas and depths and which are
inherently heterogeneous on all scales. Characterisation needs to consider all relevant spatial
scales, although the level of detail required is different, being greatest for the volume of rockcontaining the repository and, in particular, the most critical parts of the geological barrier
from a safety assessment viewpoint. The key features of the geosphere and their spatial extent
are site- and design-specific and are determined from the output of the associated safetyassessment. These are captured in a site descriptive model (SDM), which is presented in theform of a series of nested representations on different scales and levels of detail.
From the viewpoint of the design of the repository, SDM representations on the following
scales will be required (see Figure 2.2-2):

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•
•
•

Tens of km × tens of km around potential sites (regional scale), aimed at determining
suitable isolation functions of potential repository locations in terms of natural hazards
and the regional groundwater flow field.
Several km × several km (repository scale), containing the entire potential repository
footprint and all underground facilities, to assess constraints on the layout of
emplacement zones.
Several hundred metres × several hundred metres (panel scale), defining the near-fieldenvironment as required for design and configuration of engineered barriers andassociated emplacement zones.

Meanwhile, from the safety assessment viewpoint, the following scales will be required:•••

Tens of km × tens of km (regional scale) containing the potential site for identifying
transport paths from the repository to the discharge point of RNs to the biosphere.
Several km × several km (repository scale), for evaluating nuclide migration in therepository according to the layout of emplacement zones and the impact of repositoryinduced features such as the excavation damaged zone (EDZ).
Several hundred metres × several hundred metres (panel scale), for evaluating nuclide
release and migration depending on specifications of the EBS and the characteristicsof the host rock immediately around the repository (near-field).

Thus, the SDM is developed on three spatial scales; regional scale (tens of km × tens ofkm), repository scale (several km × several km), panel and near-field scale (several hundred
metres × several hundred metres). For the sake of completeness, it is noted that, for safetyassessment, “micro-scale” models of flow paths are also required in order to quantify RN
migration processes. These nested models are developed to ensure sufficient consistency of
THMC conditions within and between them, with a level of detail of geological information
and relevant processes and their couplings that is appropriate to the investigation phases
considered, the technology availability, the features of analytical models and capabilities ofcalculation codes, etc.
As discussed in Section 3.2, the LS, PI and DI stages of site characterisation and evaluation
are planned and implemented based on such a hierarchical model.
2.2.5 Approach adopted for site assessment and SDM construction
Prior to the availability of candidate sites, establishing the basis for demonstration of thecapacity to model relevant geological environments at a level appropriate for site selection as
discussed in Chapter 3 is based on:•••

Systematic appraisal of the current state-of-the-art of the science and technology
required for site suitability assessment in order to establish a basis for the approach
adopted.
Presentation of a concept for stepwise site characterisation along with demonstrationof the technology for interpretation and integration of expected output to form thebasis for design and safety assessment.
Illustration of “representative host rock types”, as preparation for the literature study,
taking account of the requirements and standards of scientifically preferable andunpreferable areas [32] and utilising a compilation of geological information on anationwide scale, with a focus on selecting areas that are potentially suitable among
Japan’s diverse geological environments.
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•

For each representative host rock type, constructing an illustrative SDM that includes
all key features (e.g., major fault zones) that are required for realistically assessing its
advantages and disadvantages in terms of repository design and safety assessment.

Construction of representative SDMs is based on the following considerations.
1.

For the construction of the regional scale SDM, as indicated in Figure 2.2-2, surface
topography and underlying geological structures are represented for a large enougharea to capture all key factors influencing a repository – generally including an entire
river catchment basin.

2.

The SDM should take into account the long-term changes in topography andgeological structure due to uplift and erosion and the associated changes in thecharacteristics of the deep environment. These processes are strongly dependent onsite-specific conditions and, in this report, appropriate site selection as described in
Sections 2.2.2 and 2.2.3 should ensure that the repository host rock considered issufficiently deep to be unaffected by uplift and erosion, and that favourable geologicalcharacteristics will be maintained over a long period of time. The SDM is constructed
for such a deep host rock, which is assumed to change slowly but have little impact on
the design and safety assessment of the repository. In the assessment of safety after
closure, the possible effects of uplift and erosion on the safety function of the hostrock in the long term, relative to the depth considered in the design, would beconsidered (see Section 2.4.4).

3.

In actual site selection, investigations are carried out at sites that do not clearly failwith respect to the exclusion criteria [32]. In this report, the SDM is constructed basedon the geological environment information collected on a nationwide scale as
described above, despite the fact that some of this information is obtained from theareas that would be excluded, such as in the vicinity of active faults, where significant
uplift/erosion have occurred and where mineral resources exist. Although these areas
are not acceptable, they were included in order to represent the characteristics of deep
rocks that are widely distributed in Japan. Nevertheless, the SDM is developed
excluding geological information from areas within a radius of 15 km of the centre of
Quaternary volcanoes and areas with extensive depths of Quaternary sedimentary andvolcanic rocks - which have clearly unfavourable characteristics for geologicaldisposal (see Section 3.3.2).

At the panel scale (several hundred m × several hundred m), assessment of engineered
barrier specifications and tunnel shapes/layouts requires a detailed model of the threedimensional fracture distribution in the surrounding host rock. To provide the input for the
calculational models used, a representation of the SDM of 100 m × 100 m × 100 m is
specified. In this report, this is called the near-field scale” to distinguish it from the panel
scale representation.

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2.3 Repository design strategy
Repository design for a specific geological environment should assure safety (radiologicaland general occupational), during construction, operation and for a long period of time after
closure, while minimising impacts on the surrounding environment.
2.3.1 Stepwise repository design approach
In a repository design, the specifications of the repository (engineered barriers, surface
facilities, underground layout, etc.) are determined so that the safety functions of therepository are assured for the defined SDM. A particular focus is containment of RNs andrestriction of their migration.
In addition to ensuring safety (radiological safety, general occupational safety), the designof the repository should also take into account environmental protection of the area around thefacility. The engineered barriers, which play an important role in ensuring safety after closure,should be designed to be robust by assigning important safety functions to each of the systemcomponents and providing sufficient margins for these functions to allow for inherent
uncertainties. In order to ensure the robustness of the repository, even if the expected safetyfunctions of one barrier element are degraded, the complementary safety functions of otherbarriers will ensure sufficient safety.
The design of a repository takes into account various factors, such as safety andengineering feasibility, based on the increasing geological knowledge base obtained duringsite investigations, together with associated general developments of scientific knowledge andavailable technology. Design is also constrained by regulatory requirements for the repository,
which will be refined as the project progresses. In addition to scientific and technologicaladvances, it is recognised that the socio-political boundary conditions surrounding thedisposal project may change. In designing a repository, the design should adapt to thesechanging conditions and be gradually optimised from an illustrative concept to a conceptual
design and then to a detailed design, in line with the phased investigation, as shown in Figure
2.2-1. To enable the design of a repository to be flexible enough to take this into account, thefollowing approach is taken [40] [41] [42]:•••Incorporating multiple requirements as design factors, so that the repository can be
designed and optimised in a consistent manner.
Repository design options will emphasise flexibility in terms of their ability to be
tailored to the variety of geological environments expected and the progress of scienceand technology during the long implementation period.
Designing a repository with detail appropriate to the level of understanding of thegeological environment, which increases during stepwise narrowing down of theinvestigated area.

The term “design factors” denotes the features and capability that the repository design isrequired to have, e.g. long-term post-closure safety, operational safety, engineering feasibility,
retrievability and economic efficiency [40]. This is further elaborated in Chapter 4.
Design options include layouts of the underground facilities and the emplacement zones,
waste emplacement configurations (e.g. vertical, horizontal) and materials of the EBS
components, such as the overpack and buffer (see Section 4.2.3). A HLW repository concept
catalogue including examples of such design components has already been developed [41].
Such a range of design options facilitate tailoring to geological conditions better thanconventional reference repository designs. Even technologies without demonstrated

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engineering feasibility at present can be considered, given expected future progress of scienceand technology over the period before the repository is licensed. Such a design concept is
consistent with the aim of improving reliability of geological disposal based on Best
Available Technology (BAT) [43].
2.3.2 Availability of engineering technology
In order to ensure implementation feasibility on the basis of technology currently available
or under development, it is necessary to assess practicality of construction, operation andclosure based on the specifications of particular designs.
For surface facilities, relevant technology is well established and experience has been
gained in the construction, operation and waste transport of existing nuclear facilities, such as
those for HLW interim storage, in Japan and abroad. For underground facilities, theknowledge base includes experience in the construction of other large-scale undergroundstructures and testing of technology in domestic and international underground laboratories.
This will be complemented by demonstration tests of technology related to construction,
operation and closure in a UIF constructed at candidate sites in the second half of the detailed
investigation phase. In addition to technology development, quality control checks prior toand during operation (e.g. pre-service inspections, regular facility inspections), together withdemonstrations in the UIF, will provide the verification necessary to ensure safety of
construction and operation.
2.3.3 Retrievability, environmental protection and monitoring considerations
As stated previously, disposal facilities should allow for reversibility and possible waste
recovery during the stages of construction and operation, while assuring environmental
protection that is confirmed by monitoring, as noted below.
(1) Assuring reversibility and retrievability
As described in Section 2.1.3 (2), the Basic Policy on Final Disposal specifies that therepository operator shall ensure project reversibility, to enable future generations to provide
input on decisions on the implementation approach and respond to advances in science andtechnology. NUMO will also ensure practicality of waste retrieval during the period before
repository closure, without compromising other safety requirements. Also as required by the
Basic Policy on Final Disposal, research institutes will develop safe and technically practicalmethods to recover all designated radioactive wastes should this ever be required.
(2) Environmental protection
Studies of potential repository impacts on the atmosphere, water, soil, biodiversity, etc.
will aim to introduce appropriate measures to reduce the burden on the environment (forexample, reduce emissions of greenhouse gases). This will particularly consider theconstruction of the UIF in the second part of DI and subsequent repository construction,
operation and closure and will capture experience in other major construction projects. In
addition, recent laws and regulations related to the enforcement of environmental
conservation will be fully reflected in the geological disposal implementation plan. For moredetails, see Supporting Report 2-6.
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(3) Monitoring
In a geological disposal project, various types of monitoring are carried out from theinitiation of site investigations to final closure to contribute to safety and environmental
protection. Monitoring is an important means of confirming that the project is being carriedout properly, and is essential for enhancing confidence in it. The closure of a repository is aprerequisite for achieving a passively safe condition that does not require active management,such as monitoring, from the point of view of safety. However, monitoring may continue asnecessary, based on the needs of society and other stakeholders, for the period between
closure and the end of any subsequent institutional control [18].
Monitoring objectives are classified into the following four categories [44], assuring:••••

Radiological safety during operation.
A suitable working environment during construction and operation.
Environmental protection of the surroundings.
Long-term safety after closure, by confirming the expected behaviour of theengineered barriers (for example, saturation of buffer material after backfilling) andsurrounding host rock (for example, recovery of groundwater level).

Monitoring commences at the stage when field work is initiated, tailored to site properties(surface environment, social conditions of the region, geological setting) and associatedrepository design and construction/operation plans. The monitoring programme also takes
into account the needs of the safety assessment to confirm or characterise specific phenomena
and, if required, provide feedback to improve design and operational methods. Monitoring
focuses on quantifying temporal changes in parameters and thus it is important to fully
characterise the initial, undisturbed state (baseline) along with its inherent variability before
invasive actions commence on site.
Monitoring technology for environmental protection is well established for similar major
construction projects, while that for radiological safety can be taken from the long history ofsuch work in the nuclear industry. Confirmation of post-closure safety is more challenging,
and will possibly require modification of existing technology or development of new
approaches - for further details, see Supporting Report 2-7.
2.3.4 Approach for the development of repository design
As the conditions of the geological environment cannot be specified in detail before
candidate sites come forward, development of repository design is an iterative process andinvolves the following (described in more detail in Chapter 4):•••

Demonstrating the availability of state-of-the-art design technology for establishing asafe repository with flexibility to respond to a variety of site conditions and social
environments (systematic design requirements methodology, reference designs anddesign options).
Starting from the H12 [20] and TRU-2 [12] reports, focusing technology development
to target SDMs, illustrating repository design specifications that meet pre- and postclosure safety requirements and also demonstrate engineering feasibility forconstruction, operation and closure.
Applying a methodology in which specific features of SDMs, such as major fault
zones, are specifically assessed to determine their impacts – both positive and negative– on different disposal concepts.
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•

•

Assessing the specifications for illustrative repository designs to determine if
implementation (including any required reversibility/retrievability) is practical, basedon existing engineering technology or that reasonably expected to be developed in the
future.
Developing plans for environmental protection and monitoring, based initially on
information about the surface environment and social conditions of the site
accumulated during the LS phase. As this is very site-specific, it is not specifically
discussed in this report.

2.4 Safety assessment strategy
Operational and post-closure safety will be assessed based on the information on selectedsites and associated repository designs and available scientific and technological knowledge,in the light of relevant regulatory standards and the requirements of stakeholders.
2.4.1 Operational safety assessment
The objective of the operational safety assessment is to confirm, on the basis of regulatory
standards to be developed, that the safety of the workers and local residents around therepository is ensured during the construction of the repository (surface and undergroundfacilities), handling operations such as waste transport, acceptance/inspection/encapsulation,
waste emplacement and closure. It should be noted that disaster prevention and industrial
safety measures against perturbations without radiological impacts will be addressed byconventional risk management throughout the periods of construction, operation and closureof the repository.
The safety of disposal facilities is generally assessed, as in other facilities that handle
radioactive waste, based on regulations contained in the Nuclear Reactors Regulation Act.
However, it is noted that the mining operations involved in constructing and operating therepository may involve hazards not usually considered for nuclear facilities. Since theaccident at the Fukushima Daiichi Nuclear Power Plant, new regulations aimed at reviewing
the design of reactors and other nuclear facilities have been formulated. This has resulted in
new regulations for HLW management facilities, enacted as “Rules Concerning the Criteria ofLocation, Structure and System of the Waste Management Facility” (hereafter the“Programme Licensing Rules”). The Programme Licensing Rules require not only designing arepository that ensures normal safety requirements, such as shielding and containment, butalso demonstrating that it has taken into consideration major perturbations, such as
earthquakes, tsunamis and other external impacts (other natural hazards, anthropogenic
hazards such as plane crashes), when defining the location of the repository and designingappropriate counter-measures. It also requires installing measures to protect against illegal
access or introduction of hazardous materials such as explosives into the repository. In theevaluation of a maximum design-basis accident, it is required to evaluate the likelihood thatthe local residents around the repository site could be exposed to radiation by accidents, suchas dropping of waste packages, in spite of the various safety counter-measures implemented.
The IAEA [45] has also published a guide on safety case development and safetyassessment for radioactive waste conditioning and storage facilities and their operations. In
this guide, the approach to the safety assessment is based on the analysis of the design andoperation of the facility, the establishment of a series of scenarios of conditions and events

2-25

that may lead to radiation exposure of humans and contamination of the surrounding
environment during operation, and the development of a safety case based on these scenarios.
Based on the set of scenarios, the radiological effects on humans and the surrounding
environment are to be analytically assessed and the results of these analyses are to becompared with radiation protection standards to confirm that the repository and the operating
method are designed to ensure sufficient safety.
NUMO will develop technologies to identify events that may affect radiological safety
before repository closure, to establish safety assessment scenarios, and to develop models anddata sets for analysis and assessment, taking into account these international guidelines andrelevant future safety regulations. Based on this, NUMO will identify appropriate measures tomitigate these risks and incorporate these into the repository design.
2.4.2 Approach for operational safety assessment
At present, criteria for geological disposal facilities have not been formulated within the
Programme Licensing Rules. Further, in the absence of a specific repository site, the extent towhich the relevant surface environmental conditions can be defined is very limited. This, inturn, constrains sensible evaluation of the impact of site-specific external perturbations, suchas earthquakes or tsunamis. In addition, with regard to prevention of illegal entry into therepository, as implementation will occur only in the future the emphasis is on monitoring
developments in setting standards for geological disposal facilities and relevant measures
implemented in other nuclear facilities. Nevertheless, evaluation of safety is possible for
maximum design basis accidents.
Therefore, as described in detail in Chapter 5, this report will:••

Illustrate the concepts and methodology of radiological safety assessment for
operational processes on both workers and local residents with reference to the safety
regulations for other relevant nuclear facilities.
Present results of the illustrative operational safety assessments for specified disposal
concepts, while assessing the practicality of potential counter-measures to reduce
associated risks.

2.4.3 Post-closure safety assessment
NUMO’s evaluation of long-term post-closure safety adopts methodology compatible withthat used internationally. For repository designs tailored to geological environments atselected sites, such assessments will define representative scenarios to evaluate theradiological impact on the surrounding populace, with consideration of uncertainties involved.
Such analysis should confirm that the repository meets safety standards over the assessment
period and will not have a significant impact on the biosphere.
Safety assessments are not intended to predict future repository evolution or the associatedhuman exposure to radioactivity, but rather to comprehensively assess whether the required
isolation and containment of radioactive waste can be achieved. Inherently, such safetyfunctions will gradually degrade with time and eventually be lost, but representative credible
future evolutions (scenarios) can be defined that capture current scientific knowledge alongwith associated uncertainties.

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Such scenarios allow potential future release and migration of RNs to be quantified
although, due to particularly large uncertainties in the future surface environment and human
lifestyles, idealised biosphere representations need to be used to convert radioactivity releases
into possible radiological impacts. Numerical values of calculated doses need to be used withgreat care, but can serve as indicators of expected repository performance, which can becomplemented by other supplementary indicators and arguments.
In accordance with the aim of safety assessment, it is important to consider all possible
conditions that may occur in a repository, but focus on likely conditions that are important for
determining the safety of the repository. For this purpose, a “FEP” catalogue is used, which
describes the characteristics of each element of the repository that may be relevant to its
safety function (Feature), the events that affect these characteristics (Event), and the processof the repository's evolution over time (Process). In order to estimate resulting dose,mathematical models and data sets are used to describe the release and migration of RNs fromthrough the EBS and host rock, and into the biosphere. The future human exposure to RNs
migrating into the biosphere depends on the transport of RNs depending on surface conditions(topography, land use, etc.) and the mode of exposure as a result of human lifestyles (sourcesof drinking water, extent of consumption of local crops and livestock, etc.).
NUMO has been working on technology development related to fundamental componentsof the safety assessment: scenario development, modelling and preparation of required data
sets [18]. Additionally, for the PI stage of site characterisation, a preliminary safetyassessment manual that systematically outlines required procedures and methods has been
produced [46]. With a view to subsequent site selection, extension and refinement of thissafety assessment methodology and confirmation of its applicability for specific sites will benecessary.
Preparation of an appropriate safety assessment analysis toolkit is especially important toallow objective evaluation and comparison of the performance of alternative disposal
concepts, where the methodology must also handle different uncertainties in terms of site
conditions and repository specifications.
2.4.4 Approach adopted for post-closure safety assessment
The regulatory framework for safety assessment in Japan will be developed in the future
and thus, in this report, guidelines on safety standards and indicators indicated byinternational organisations (e.g. IAEA Safety Standards Series [17] [19] [43] and ICRP
recommendations [47] [48]) and safety regulations in other countries with similar projects areused to provide reference performance targets for future human exposure.
In recent years, the need for radiation protection of both the environment and non-humanorganisms has been discussed (e.g. ICRP recommendations [49]). In addition, non-radioactive
hazardous substances contained in the waste could have post-closure impacts on humans andthe environment. Such issues are not considered in this report, but will be addressed in the
future.
For the scenario framework, a risk-informed approach that combines the disaggregatedlikelihood of occurrence and the significance of the consequence of the scenario has become
popular among international organisations and regulatory bodies since 2000 (see Section
6.1.5(1)). While the scenarios leading to radiation exposure via groundwater flow are themain focus of the assessment, scenarios that include events that lead to the loss of the

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isolation functions or a significant reduction of the containment function are developed asthese may need to be considered for specific sites.
The safety evaluation period can be determined by consideration of the following points ingeneral, referring to IAEA [43], OECD/NEA [31] and the safety regulations in relevant
countries:
•
•
•
•
•

The time of the maximum estimated dose to the general public.
The time at which the potential hazards of radioactive waste have decayed to anegligible level.
The time at which uncertainty of evaluation becomes too large for models to bemeaningful.
The time required to assess impacts of slow processes and occurrence of extremelyrare events.
Timescales of stakeholder concerns.

In some countries, the period for quantitative assessment, taking into account the above
aspects, is specifically indicated in safety regulations and, in many cases, a time frame of 1
My is adopted (e.g. [50] [51] [52]). In Japan, the safety evaluation period for geologicaldisposal is under discussion and will be specified in future regulatory standards. Therefore, inthis report, the safety evaluation period is not selected on a scientific basis, but is taken to belong enough that it will cover the maximum impact from credible groundwater release
scenarios.
Based on the above, post-closure safety assessment, as discussed in more detail in Chapter
6, includes the following:
•

•
•

•

•

A risk-informed approach, introduced in order to appropriately deal with various kindsof uncertainties. The focus is on an appropriate concept and methodology toqualitatively assess safety on the basis of international guidelines and otherconsiderations.
A safety assessment analysis method is developed in order to allow the objective
evaluation and comparison of performance of different geological environments andassociated repository designs.
Post-closure long-term safety assessments for the SDMs developed in Chapter 3 andthe repository designs tailored to them in Chapter 4, considering the likelihood of
relevant scenarios describing their future evolution. Results are discussed in the lightof tentative performance targets, accepting current limitations on available knowledge
and hence the models and databases used.
By appropriate site selection, it is assumed that the host rock is little affected by upliftand erosion, and that favourable geological characteristics will be maintained for a
sufficiently long time (see Section 2.2.5). However, to ensure this, the potential effectsof long-term uplift and erosion on the safety functions of the host rock are considered(see Supporting Report 6-10).
Matters that depend on the site-specific geological environment and construction andoperating methods, such as the degree of disturbance of conditions during construction
and operation of the repository and the process of their recovery, are presented as
issues to be considered after site identification, unless they could be conservatively
disregarded in the safety assessment.

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2.5 Management systems
Responding to the challenges faced when developing a safety strategy under therequirements described above is facilitated by appropriate management tools andmethodology. These integrate all the actions carried out and ensure consistency in therepository concepts developed for a variety of geological environments, with respect toconstruction, operational and post-closure safety, feasibility, economic efficiency, and social
acceptance requirements, whilst also considering trade-offs among them.
A geological disposal programme continues for a long period of time and needs to beflexible so that it can respond not only to progress in technology but also to changes insocietal boundary conditions. This has been discussed at a conceptual level in the “NUMO
structured approach” [53] and in the 2010 report [18]. The key components of the strategy forensuring safety while maintaining flexibility can be summarised as follows:
•

•

•

•

Development of management methodologies for close coupling of the geologicalinvestigations and site evaluation, repository design and safety assessment work,
facilitating good communication between the staff involved in each field (Figure 2.51). This coupling will evolve naturally as the programme progresses, from site
selection and the subsequent licensing phases, through construction, operation andclosure.
Comprehensive assessment of conventional and radiological risks to both workers andthe general public at all programme stages. Setting and analysing appropriate
scenarios allows identification of potential hazards from both normal and perturbed
operations. Counter-measures are thus identified that could be implemented wherever
required.
Regular reviews to ensure that technology is maintained at state-of-the-art levels,reflecting developments in system understanding and advances in science. Anytechnical work performed by NUMO, including site selection and technology
development, is subject to regular checks and review by external experts.
Re-evaluation and updating of the safety case, including major decisions made basedon it, in each phase of the programme. This is conducted to check that current
regulatory guidelines and standards are complied with and also to provide continuous
reassurance to the general public. Technical issues are also identified that should bereflected in subsequent updates of the safety case.

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Figure 2.5-1 Coupling of site characterisation, repository design and safety assessment(based on Figure 4.2 in the 2010 report [18])The design of the repository should be based on this information and incorporate sufficient
safety margins. The safety assessment should confirm that radioactive material in therepository will not have a significant effect on humans, taking into account the various
uncertainties inherent in the SDMs and in the repository designs. If it is judged that therequired level of safety is not met, or if the uncertainty in the results of the assessment is so
great that confidence in the results needs to be improved, the design and safety assessment of
the repository should be re-examined, for example by expanding the geological information
available to reduce the uncertainty.
If information is added or revised during the course of these activities, it is necessary tounderstand the extent to which this affects the investigation and assessment of the site, thedesign of the repository and the safety assessment and the integrated safety case. In order toachieve this, it is important for NUMO to have the overview required to lead integration of
results and determine the direction of the work, as well as to ensure close communication
between project managers in different disciplines, so that changed information and data canbe passed between them and the decisions made reflect the consensus of all involved.
Coordination between technical disciplines, preliminary assessment of conventional andradiological risks and the preparation of response measures, repeated review of geologicaldisposal technology, and the development of safety cases are all essential management tasks
for ensuring safety. At the same time, the following issues should be addressed:
•
•
•
•
•

How to deal with changes in the social environment and uncertainties in safety overthe long term after repository closure.
How to ensure the quality of all technical studies.
How to share the vast amount of accumulated knowledge, information and data
within and between generations.
How to promote continuous technical development.
How to ensure the availability of human resources in a variety of technical fields.

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2.5.1 Uncertainty management
As already mentioned, the handling of uncertainty is extremely important, particularly
because geological disposal involves unusually long timescales, both in terms of the durationof the project and in terms of ensuring safety after closure, and because the geologicalenvironment has inherently heterogeneous characteristics on all spatial scales.
The four basic approaches to dealing with uncertainty in geological disposal have beentermed “identification”, “avoidance”, “reduction” and “assessment” [54]. Uncertainties are“identified” and “avoided” or “reduced” in the safety case developed at each stage of the
project. For example, uncertainties in properties or events that are difficult to understand or
predict, such as the distribution of geological characteristics over large spatial scales or thepredicted evolution of a future geological disposal system over time, can be addressed bydesigning a robust repository that assures sufficient safety despite such uncertainties as
determined by realistic and/or conservative assessments. Uncertainty should also be “reduced”
by continual extension and improvement of the technical and scientific knowledge base.
Nevertheless, uncertainty cannot be completely eliminated, so it is necessary to “assess”
remaining uncertainty and decide whether it is acceptable while ensuring the safety of therepository. In staged site selection, uncertainty is quantified and reduced by refining thegeological knowledge base as captured in SDMs, tailoring design of the repository to themand specifying the engineering techniques for implementation, while the remaining
uncertainty is taken into account in the safety assessment, introducing conservatisms whenrequired (see Supporting Report 2-8 for details).
In general, uncertainty is defined as aleatory uncertainty, due to inherent differences orvariations in observed values of the system considered, and epistemic uncertainty, due to lackof knowledge or information about it [55]. In the site investigations, for example, in addition
to aleatory uncertainty due to the inherent temporal and spatial heterogeneity of the geologicalenvironment, there is epistemic uncertainty involved in its conceptualisation within SDMs
due to uncertainty in data (e.g. caused by measurement errors) and its resulting interpretation.
Over relevant spatial scales in the order of several kilometres and timescales of hundredsof thousands of years or more, it is clearly impossible to obtain enough observations to fully
quantify all uncertainties [54]. Nevertheless, quantification of the heterogeneity of geologicalcharacteristics is carried out to the extent practicable (e.g., statistical analysis to define means,variances and standard deviations of data). In the case of important geological structures andnatural events/processes that impact repository implementation and post-closure safety, thetypes, causes and degrees of uncertainty and the factors associated with these (e.g., thenetwork structure of faults and fractures) are identified and responded to through theknowledge communication as shown in Figure 2.5-1.
In the design of a repository, the specifications of engineered barriers and undergroundfacilities will be set in such a way that the required functions can be assured even if
uncertainties in the site geology and in the analytical models and parameters used in thedesign process are taken into account. In addition, the quality of fabrication and its variability
will be assessed through demonstration tests (e.g., mock-up tests, large-scale tests in URLs,etc.), and associated quality assurance protocols defined to ensure that as-built repository
components meet specified requirements.
Uncertainties in safety assessment are often categorised and treated as being associatedwith scenarios, models and data. Scenarios are developed for credible evolutions of thedisposal system, based on information integrated into the SDMs and associated repository
design, taking into account uncertainties arising from lack of understanding or knowledge of
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phenomena and processes. Specifically, scenarios are classified into those that are likely tooccur based on best current understanding, those that take into account uncertainties that canbe reasonably established based on current scientific knowledge, and those that consider
events with a low probability of occurrence. In this way, some impacts of uncertainty on thesafety of a geological disposal system are taken into account.
As a starting point, the FEP catalogue is expanded based on the latest domestic andinternational knowledge and, to the extent possible, realistic scenarios should be derived fromthis. If realism is not practical, scenarios should be sufficiently conservative, utilising expert
opinions on their appropriateness and completeness, if required. In this way, the risks ofoverlooking safety-critical scenarios or developing inappropriate scenario descriptions can be
reduced.
In addition, since it is inherently impossible to predict human activities in the distant future(as noted in Section 2.4.3), or even assess any uncertainties involved, the biosphereassessment approach is based on internationally accepted concepts of stylised scenarios (e.g.,
IAEA (2012) [43], ICRP (2013) [48]). Stylised scenarios based on current human activities
are also developed for potential future perturbations (human intrusion scenario).
Key uncertainties in the safety assessment models and associated databases for radiologicalconsequence analysis of defined scenarios should be compensated for by including sufficientconservatism. For example, if it is difficult to establish sufficiently reliable models/databases,
analyses should be carried out in such a way that conservative results can be obtained, e.g. byignoring some of the expected positive functions of barriers or by using models and data thatclearly give pessimistic results despite these uncertainties.
Important uncertainties are identified in the safety case developed at each siting stage andreflected in the investigation and technology development plan for the next stage. This stepby-step approach to uncertainty reduction ensures that the repository is constructed in a safemanner. Prior to the closure of the repository, periodic safety reviews, as stipulated in the
Nuclear Reactor Regulation Act, will be carried out to capture the latest knowledge and theresults of monitoring, thus reducing uncertainties associated with operational optimisation andpost-closure safety.
During the 100-year lifetime of the project, there may be changes in socio-political andeconomic boundary conditions that affect its implementation. In the light of such uncertainties,
a range of technical options will be developed (e.g., various disposal concept options) to allow
for flexibility in responding to changes and minimise associated management risks for
NUMO.
In general, the variability of geological characteristics is site-specific, while uncertainties
about the quality of engineered barrier systems depend on the materials and manufacturingmethods involved. Therefore, in this report, representative characteristics of the host rock are
set by assessing nationwide literature, and uncertainties are generally not directly captured in
the SDMs (exceptions involve groundwater chemistry, which plays an important role inperformance of both engineered and natural barriers). Clearly, this is a limitation that would
be addressed in updates to the SDMs. Similarly, for the physical properties of engineered
barrier materials, representative values are set based on reported experimental data anduncertainties in these properties are currently not considered, but will be in the future. A
summary of the treatment of the uncertainties in this report is given in Section 7.3.2.

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2.5.2 Quality management (QM)
Quality management plays a very important role in both developing a safe repository andin the production of a report like the present one, which integrates the key components
involved [56]. NUMO has developed a quality management system (QMS), which builds on
ISO 9001 in order to manage the site investigations (LS, PI and DI), repository design andassociated safety assessments together with the development of technology required tosystematically implement a repository (construction, operation, and closure). In the NUMO
QMS, procedures to ensure quality of repository implementation include:
•
•
•
•
•

Quality goals and requirements for repository implementation and associated technicaldocuments.
Processes to check that such goals and requirements are met, including quality audits,monitoring and verification activities.
Assurance of availability of required infrastructure and human resources to meet
quality levels.
Comprehensive production of records and long-term archiving of these.
Correction of any non-conformity and associated preventative actions to prevent
recurrence.

To ensure safety of repositories, technical quality management has to be tailored to key
activities. For example, in preparation for geological investigation and evaluation of potentialsites, NUMO has already produced a site investigation planning manual and a quality control(QC) guide for activities up to the stage of initiation of preliminary investigations [57]. This
specifies a system in which the quality of the geological investigation methodology andprocedures, as well as the primary data and synthesised output, are reviewed by both NUMO
internal and external experts. In addition, a geological information management support
system has been developed to assist in ensuring that databases produced are traceable andcomplete. Similar procedures and methods for the design of a repository and safetyassessment have been compiled as guidebooks, to allow teams working under different site
conditions or on facilities with different designs and safety assessments to achieve the same
quality level.
The repository and components of the EBS are assured to meet the quality of design
specifications by appropriate testing and inspection procedures, extended to the extent
possible to include input information on geological environments, characteristics of waste
packages, etc. The quality of the waste material will be assured by setting criteria for its
acceptance in accordance with the approach described in Supporting Report 2-5. Specificallyfor repository design, a focus at the current stage is on verification that available technology
can construct the engineered barriers to required quality levels, given the constraints ofoperational logistics (e.g. expected waste emplacement rate) and the challenges of working
underground, predominantly using remote handling/tele-operated systems. This includes
developing monitoring systems, to directly confirm the quality of the resultant EBS, andtesting procedures to respond to any cases where required quality is not met.
Inherently, however, the quality of the EBS performance after closure cannot be directly
verified and is dependent on the safety assessment analysis – and hence on the quality of
approaches, models and databases used to implement this on the basis of the SDM and a
representative range of future evolution scenarios. Individual system models and databases
can be subject to expert review, to check quality of fundamental assumptions, and verified byinter-calibration with alternative models and formal methods of code testing. However,validation is limited to short-term tests, using laboratory or in-situ experiments, or more
2-33

uncertain natural analogues. For more details of quality assurance (QA) of analytical codesand databases used in this report, see Supporting Report 2-9.
During the three phases of site characterisation, enormous volumes of data will be
produced and the design and safety assessment analysis will be frequently updated. Updating,change management and communication of the huge multi-disciplinary knowledge baselinking site investigation, design and safety – including the management of tacit knowledge
accumulated by key staff – is recognised to be an important challenge and is a focus for
systematic QMS and knowledge management system (KMS) development (see also Section
2.5.4).
The quality assurance of the results of technology development is important in order todevelop a safety case based on BAT and the quality of technology development is promoted
by a programme of obtaining technical reviews and advice from experts in all appropriate
fields. Appropriate QA is also required when incorporating R&D results from otherorganisations into the geological disposal programme. Thus, NUMO employs establishedmethodology to achieve the necessary quality of technical work. This includes production ofscientific papers for peer-reviewed publications and technology development reports, which
are reviewed by third parties.
The quality management framework described above has been applied to the establishmentof the SDMs, design and safety assessment of the repository in this report. In addition, thesystem shown in Figure 1.4-4 in Section 1.4.2 ensures that the process and results of studies
in this report are reviewed by experts to ensure the quality of the safety case. As described in
Section 1.4.2, this report has been improved based on the review comments [58] from the“Special Review Committee on the NUMO Safety Case” of the Atomic Energy Society ofJapan on the NUMO Safety Case Report (Review Version) (published in November 2018).
2.5.3 Requirements management
The implementation of a geological disposal programme requires the hierarchicalconsideration of a wide variety of requirements. These include technical requirements for
ensuring the safety of a repository and requirements for quality management as mentioned
above, as well as requirements from laws, regulations and policies defined by the national
government, regulatory bodies, or requested by various stakeholders. In order to manage this,NUMO has developed a Requirements Management System (RMS) [59], aimed at
systematically identifying the wide variety of requirements and managing them in an effective
and transparent manner.
Such requirements and associated decisions need to be consistent in each phase of the
programme, and also throughout the diverse range of technical work carried out by NUMO
managers and technical teams. The RMS ensures completeness of the requirements andprovides a mechanism for change management, with transparency and traceability, as
requirements evolve with time associated with the stepwise progress of the programme,
impacting decisions made based on these requirements (e.g. design specifications). This iscomplemented by structured management of associated knowledge (see Section 2.5.4),
allowing appropriate responses for any case where significant changes in boundary conditions
occur, with reference to the basis of the recorded background to past decisions or judgments.
The current requirements for the characterisation of geological environments, repository
design and safety assessment were re-examined during the compilation of this report. This
information will be recorded as necessary in an expanded and updated RMS. As further
2-34

discussed in Chapter 7, development of the RMS will be essential, but also complex since itwill involve interaction between all the technical areas and groups working to implement
NUMO’s programme.
2.5.4 Knowledge management
The Japanese geological disposal programme will develop over a long period of time,extending over a century or more, comprising site characterisation, design, construction,
operation and closure of the repository, including the preceding R&D period. The vast amountof knowledge, information and data produced inside and outside Japan at each step of the
programme needs to be compiled systematically and also transferred to future generations.
Knowledge, information and data can be defined as follows, referring to the OECD/NEA
study [60]:•••Knowledge: know-how, understanding, experience, insight, grounded intuition andcontextualised information.
Information: contextualised, categorised, calculated and condensed data.
Data: facts and figures related to specific topics, but not organised in any way.

Management of knowledge, information and data (hereafter collectively termedknowledge) includes, as mentioned in the QM section (Section 2.5.2), procedures to process
information and data reliably in multi-disciplinary, coordinated work, including site
investigation, design and safety assessment. It also needs management of knowledge andinformation used as a basis for various decisions in each project phase, as described in therequirements management section (Section 2.5.3). Recognising this, a KMS is underdevelopment, which securely integrates, synthesises and archives the knowledge base, while
ensuring transparency, traceability and retrievability.
Given that the knowledge forming the basis for geological disposal has been increasing
exponentially with time, there are clear limitations in the traditional “report-based”
knowledge management systems. This was recognised by JAEA [61], which led to thedevelopment of an advanced KMS which can be linked to associated quality and requirements
management systems [61] [62]. A web-based communication platform, CoolRep [63], wasalso developed. The JAEA KMS has been refined further, with the intention of integrating it
into the NUMO RMS described above, reflecting the role of JAEA as an independent R&D
organisation providing both NUMO and regulatory agencies with fundamental scientific andtechnical knowledge. As the concept for the development of the JAEA KMS is based on the
structure of a safety case, it is well suited to systematically managing, quality assuring,
communicating and archiving knowledge accumulated during the compilation of this reportand presenting it to all stakeholders.
The safety case is developed based on integration of technical knowledge and informationfrom a wide range of fields of expertise. For this reason, it is not easy to overview the entire
safety case, even for experts in the specific fields included. Therefore, special communication
efforts will be required for the explanation of the safety arguments included at different levels,
responding to needs of different audiences, ranging from experts in different scientific fields
to non-technical stakeholders.
In order to promote such safety communication, several documents with different content
tailored to the knowledge level of readers, as well as a web-based communication platform(related to CoolRep mentioned above), will be produced. These increase the opportunities forusers to access and easily view the knowledge and information that they are interested in, at a2-35

level defined by hierarchical electronic documents, as indicated in Figure 1.3-2. In the future,such material on the safety case will be available via the NUMO website.
2.5.5 Management of scientific and technology development
To support implementation of geological disposal, NUMO and related research institutions
collaborate in strategic technical development and periodic evaluation of progress, as requiredby the Basic Policy on Final Disposal. This interaction between NUMO and research
institutions is facilitated by meetings of the Coordination Council on Basic R&D of
Geological Disposal (hereafter referred to as the “Coordination Council”), established by
METI in 2005, which link R&D programmes [64] [65] to NUMO’s technical development
plan and are updated to respond to changes in it. This Coordination Council was later
reorganised to also consider the practical R&D needed for implementation in addition to more
fundamental R&D [66]. This will promote the development of technology needed to safely
implement geological disposal in an economic and efficient manner.
NUMO already has a structured programme for developing and testing the technology thatwill be required for the LS and initial PI stages, as outlined in the medium-term R&D plan[67] implemented by NUMO over the five fiscal years from 2013 to 2017. However, thisprogramme will need to be adapted to the findings obtained as the siting programme
progresses, as discussed in Section 7.4 of this report.
The technical issues identified in this report (see following chapters), as well as the issues
identified from the results of the technical development and fundamental R&D, were reflectedin the general research and development plan for geological disposal (Fiscal Years 20182022) [68], which was compiled on the basis of the Coordination Council. In line with thisoverall plan, NUMO has developed a mid-term technical development plan covering theperiod from FY2018 to FY2022 [69].
NUMO will continue to develop geological disposal technology and, after a site is
identified, the necessary technology will be implemented in a timely and phased manneraccording to the characteristics of the site, reflecting the latest scientific knowledge (seeSection 7.4 of this report for further details). The management of the project will be facilitatedby a technology development plan, which is formulated by repeatedly summarising the safetycases at appropriate times to clarify the technical issues to be addressed in the future. In thisprocess, a PDCA (Plan-Do-Check-Act) cycle12 is applied at various levels to promote
constant improvement and upgrading of the geological disposal technology.
2.5.6 Management of human resources
Geological disposal involves a complex programme that includes many disciplines. The
personnel required are highly qualified staff, including both experts in each relevant field ofexpertise and experienced generalists who can conduct cross-cutting overviews of the output
from all included disciplines. Recruiting and training the required human resources andinstilling in them an appropriate working culture for an implementing body is a key
component of both quality and knowledge management, as described above, and is facilitatedby the toolkits developed in these areas. Planning and implementation of technology
12

The PDCA cycle is a management method in which the business processes are continuously improved byrepeatedly performing four fundamental activities: planning (plan) → execution (do) → evaluation (check) →improvement (act).

2-36

development is inherently constrained by the available manpower – both in NUMO and in
relevant R&D organisations – and hence the management of human resources is thus
recognised as being extremely important and explicitly included in the R&D plan [69]. As apart of this plan, NUMO has established training courses to develop young personnel incollaboration with relevant organisations in Japan [70].
In addition to domestic resources in Japan, it will be important to encourage theparticipation of external human resources from the international community. Especially whenensuring technical reliability, the participation of an international network is essential in, forexample, the preparation of qualified English documents and review of the safety case toensure that it meets international standards.
Together with the utilisation of tools for quality, requirements and knowledge management
as described above, NUMO will formally manage human resources, for example, whenplanning systematic recruiting and training, as well as in cooperation with relevantorganisations inside and outside Japan through:
•

•

•
•

Recruitment of both new graduates from universities and experienced workers in thefield of geological disposal, based on the programme schedule and required expertise
identified in the implementation plan. This will be coupled to cooperation inmaintenance of the key infrastructure (e.g. URLs) and knowledge transfer fromrelevant R&D organisations by long-term attachment of staff and collaboration in joint
research projects.
Improvement of tacit knowledge and practical abilities through staff attachment to, or
participation in, cooperative programmes between NUMO and other
implementing/R&D organisations involved in radioactive waste disposal, both insideand outside Japan.
Ensuring that NUMO’s technological capabilities and tacit knowledge base are
maintained, improved and passed to future generations of staff through the periodic
development of safety cases and participation in joint research projects.
Assessment and introduction of programmes to foster human resources with a wide
range of multi-disciplinary knowledge and project management skills (generalists).

In the compilation of this report, next-generation human resources were cultivated throughcollaboration of younger staff with relevant research institutions, discussions with external
experts, participation in documentation of the project, and participation in the review processinside and outside Japan.

2.6 Summary
In this chapter, based on the current status of Japan's geological disposal programme, thebasic implementation policy to ensure effective site selection, design of the repository andassociated safety assessment has been outlined. In addition, the approach to development of
the safety case described in this report has also been outlined, highlighting boundary
conditions and constraints at the present pre-siting stage.
In Chapters 3 to 6 of this report, the specific approaches to the selection and modelling ofgeological environments suitable for geological disposal, repository design and engineering
technology, and the assessment of operational and post-closure long-term safety are set out. In
each of these areas, application of the uncertainty, quality and knowledge management

2-37

approaches described in Section 2.5 of this Chapter is included. Chapter 7 summarises theseresults and discusses integration of these as a safety case.

2-38

Supporting Reports (SRs)
SR 2-1SR 2-2SR 2-3SR 2-4
SR 2-5
SR 2-6
SR 2-7
SR 2-8
SR 2-9Definition of specified waste for geological disposal
Characteristics and quantity of waste (HLW)
Waste inventory for design and safety assessment
Characteristics and quantity of waste (TRU)
The concept of waste acceptance criteria
The approach to environmental conservation
The concept for monitoring
Managing uncertainty
Assessment of the reliability of numerical analyses

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evolution for deep hydrochemical conditions in Horonobe area, Hokkaido, Japan, Journal of
Geography, Vol. 118, pp. 700-716, doi: 10.5026/jgeography.118.700 (in Japanese).
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ASME 2011 14th International Conference on Environmental Remediation and Radioactive
Waste Management, Reims, France, 25-29 September 2011, ASME, pp. 1477-1481, doi:
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and Gas Industry Committee, Advisory Committee for Natural Resources and Energy (inJapanese).
[33] NUMO (Nuclear Waste Management Organization of Japan) (2004): Evaluating site
suitability for a HLW repository: Scientific background and practical application of NUMO's
siting factors, NUMO-TR-04-04.
2-42

[34] Advisory Committee for Natural Resources and Energy (2017): Results of theexamination of the requirements and criteria for the presentation of regional scientificcharacteristics of geological disposal (summarised by the Geological Disposal Technology
WG), Geological Disposal Technology WG, Nuclear Energy Subcommittee, Electricity and
Gas Industry Committee, Advisory Committee for Natural Resources and Energy (inJapanese).
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Committee Concerning Basic Matters Based on the Environmental Impact Assessment Act,
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repository concepts for volunteer siting environments, NUMO-TR-04-03.
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siting process, Proceedings of the IHLRWM 2019, Knoxville, TN, USA.
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Safety Guide, IAEA Safety Standards Series, No. SSG-23.
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(2010): Study on monitoring and primary baseline for geological disposal of radioactive waste,
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2-43[45] IAEA (2013): The safety case and safety assessment for the predisposal management ofradioactive waste, General Safety Guide, IAEA Safety Standards Series, No. GSG-3.
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Guide, IAEA Safety Standards Series, No. GS-G-3.4.
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2-44[60] OECD/NEA (2019): Preservation of records, knowledge and memory (RK&M) acrossgenerations, Final Report of the RK&M Initiative, NEA No. 7421.
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concept of knowledge management system of geological disposal technology, JAEAResearch 2006-078 (in Japanese).
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2-45

3 SELECTION AND MODELLING OF GEOLOGICALENVIRONMENTS SUITABLE FOR GEOLOGICAL DISPOSAL
Based on the fundamental strategy for site selection described in Section 2.2, this chapter
is compiled for the following purposes:

To demonstrate that the concept for site suitability assessment and associatedinvestigation and evaluation methods have been developed systematically, based on
the latest scientific knowledge and technological developments relevant to Japan's
geological environment.



To demonstrate the technology for interpreting and integrating information on thegeological environment, synthesising this in a stepwise manner into a model of thehost rock in its geological setting, which is the basis for the design and safetyassessment of the repository.

In Section 3.1, in addition to the relevant safety features expected for the geologicalenvironment, the basic approach to addressing the factors that can affect these safety features
is described. Additionally, an overview of the geological characteristics widely observed indeep environments in Japan is given, which shows that those with the required safetyfunctions can be selected.
In Section 3.2, based on the basic concept for site selection described in Chapter 2, anapproach is presented for excluding geological environments with significant effects ofnatural phenomena that could impair the safety features, such as volcanoes/igneous activity,
earthquake/fault activity, and significant uplift/erosion. This allows focusing on those which
maintain favourable characteristics for a sufficiently long period of time. Required
investigation and evaluation methods for such a selection process are also summarised.
Section 3.3 outlines how, for potentially suitable host rocks that might result from site
selection (and for which the safety functions are expected to remain intact for a long time), toconstruct and show the applicability of a Site Descriptive Model (SDM). The SDM is derived
by interpreting and integrating information on the geological environment on the spatial
scales required for the design and safety assessment of the repository described in Chapter 2.
This model provides an important basis for efforts related to repository design andengineering (described in Chapter 4) and evaluation of long-term, post-closure safety(described in Chapter 6).
Section 3.4 describes the scientific knowledge that serves as the basis for examining rare
perturbing event scenarios for the evaluation of post-closure safety, with emphasis on theprobability of occurrence of relevant natural phenomena and their impact on the geologicalenvironment.
Finally, Section 3.5 summarises key conclusions from this chapter and discusses future
actions.

3-1

3.1 The role of the geological environment in ensuring repository
safety3.1.1 Requirements on the geological environment and associated influencing
factors(1) Requirements and conditions for ensuring safety before closure (during
construction and operation)
Starting from the literature survey (LS), selection of the repository site proceeds throughstaged siting investigations (for details, see Section 3.2.3), which are then followed by a seriesof tasks related to the construction, operation and closure of the repository (see Chapter 4 fordetails). These tasks include construction of the surface and underground facilities, acceptanceand inspection of the waste and encapsulation in overpacks, emplacement of the waste andconstruction of an engineered barrier system (EBS), and, finally, backfilling and sealing of therepository. Table 3.1-1 summarises the requirements1 ([1] and [2]) for the geologicalenvironment from the viewpoint of performing these tasks, with safety as the highest priority.
Table 3.1-1 Requirements and influencing factors for the geological environment until closure
Influencing factors*1

Requirements for the geological environment
The host rock must not be unconsolidated
Quaternary sediments

Distribution of unconsolidated Quaternary
sediments at depths of 300 m or more … (1)Events that affect the safety and operability of theunderground facilities are unlikely to occur

Geothermal activity/hot springs*2, swelling ground,rock burst, mud volcanoes, high water inflow, toxic
gas, earthquake activity... (2)Events that would disrupt the safety of the surface
facilities are sufficiently unlikely to occur

Pyroclastic flow, lava flow, lahar... (3)Events that could negatively impact the safety ofthe surface facilities are unlikely to occur

Distribution of soft ground, ground deformation anddisplacement, occurrence of earthquakes, tsunamis,
landslides, debris flows, floods... (4)*1The policy for selecting a site considering each of the influencing factors (1) to (4) is presented in Section
3.1.2.
*2
Refers to warm or hot water upwelling from underground.

(2) Requirements for ensuring long-term post-closure safety
The basic concept for ensuring long-term safety after closure of the repository involves“isolation” and “containment”, as described in Sections 2.1.2 (1) and (3). In order to achievethis, the geological environment is expected to maintain the following functions over a long
period of time [3] [4] (as shown in Table 3.1-2.):


Protection from significant effects of natural perturbing phenomena.

Reduction of the likelihood of human intrusion.

1

Not only the requirements defined in the Final Disposal Act and the Advisory Committee for Natural
Resources and Energy (2017) [2], but also the conditions considered to be favourable, are included as“requirements” in this report.

3-2



Restriction of RN leaching.

Restriction of RN migration.

In particular, from the viewpoint of the functions related to the restriction of the releaseand transport of radionuclides, the requirements on the geological environment for assuring
EBS performance and acting as a natural barrier [5] [6] can be summarised in terms offavourable thermal, hydraulic, mechanical, and chemical (THMC) conditions [1] [2], with afocus on HLW.
Table 3.1-2 Requirements for the long-term post-closure geological environment(Modified from IAEA, 2011 [3]; 2011 [4] and Advisory Committee for Natural Resources andEnergy, 2017 [2])Isolation

Basic
concept

Safety function

Requirements for the geological environment

Protection fromsignificant effects
Natural phenomena that would cause the waste to approach, or beof natural perturbing exposed to, the biosphere will not occur over relevant timescales
phenomena
Reduction of thelikelihood of human
intrusion

Mineral resources with high economic value are not identified

Containment

Thermal environment: The rock temperature is sufficiently low so that aloss of function resulting from thermal degradation of the buffer material
due to long-term exposure to temperatures higher than 100 ºC will not
occur
Restriction of RN
leaching

Hydraulic environment: Groundwater flow is slow due to a low hydraulicgradient and/or low hydraulic conductivity, which assures EBS longevity,
increases the transport time of radionuclides with groundwater flow andreduces the concentration of radionuclides due to radioactive decay

andRestriction of RN
migration

Mechanical environment: Rock deformation is small due to highcompressive strength and elastic modulus and small creep deformation toprevent damage to the overpack
Chemical environment: Groundwater satisfies conditions of no extreme
pH, is chemically reducing and carbonate concentration below 0.5 mol/l[2] to suppress: dissolution of glass, local corrosion and passivation of
overpack, alteration of buffer material a decrease in sorption capacity of
buffer material and bedrock, an increase in solubility of radionuclides

Japan is located in the Circum-Pacific Orogenic Belt, as described in Section 3.1.3 (1), so
natural disruptive phenomena, such as volcanic/igneous activity, earthquake/fault activity, anduplift/erosion occur, repeatedly or continuously on a geological timescale of hundreds ofthousands to millions of years. The characteristics of these phenomena, based on scientificknowledge and the results of case studies obtained up to the time of JNC’s H12 report [7],have been reconfirmed [1] in the latest compilation of scientific knowledge [8]. Such natural
phenomena may affect the safety functions expected for the geological environment [1] [2][7]. Especially when there is a “significant impact” [1] [9], which exceeds the range that canbe handled by the repository design, the safety functions expected for the geologicalenvironment could be significantly reduced or lost [10].
Even in geological environments where such natural phenomena are not significant,various perturbations could occur within the repository during relevant timescales, such as the3-3slow and cumulative effects of unavoidable phenomena, such as uplift/erosion and sea-levelchanges, on timescales exceeding tens of thousands of years. Such intrinsic or extrinsic
perturbations and associated effects are suppressed by the natural buffering function of thegeological environment [5] [6]. Deep underground, geological characteristics change slowly
over time, and sudden or fast changes are thus unlikely [11]. Taking into account the extent ofsuch changes, the geological environment can be considered to show sufficient stability when
favourable characteristics for repository safety are maintained for a long enough time [5] [6].
As described in section 3.1.3 (2), if this stability can be justified based on scientific knowledgeof relevant events and processes that have occurred from the past, it is considered that theprobability of reducing or losing the function of restricting the release and migration ofradionuclides is extremely small [11] [12] for periods exceeding around 100 ky in the future,by which time the hazard presented by the waste has decreased significantly due toradioactive decay.
In Table 3.1-3, factors affecting the expected long-term post-closure safety function of thegeological environment and key processes involved are summarised [1] [2].
Table 3.1-3 Impact on long-term post-closure safety functions expected for the geologicalenvironment and influencing factors (edited from Advisory Committee for Natural Resources
and Energy, 2014 [1]; 2017 [2])Influencing factors and main processes*1
Volcanic and igneous
activity

Magma intrusion and eruption at thesurface... (5)Migration and inflow of Movement and inflow of volcanic
volcanic hydrothermalhydrothermal and deep-seated fluids...
and deep-seated fluids*2 (6)Displacement of a fault reaching thesurface from deep underground … (7)Earthquake and fault
activity

Release of waste to the biosphere
Rise in ground temperature
Decrease in groundwater pH or increasein carbonate species concentration
Breaking and crushing of the rock mass

Increased hydraulic conductivity of the Change in groundwater flow paths androck mass around a fault and infiltration change of groundwater to oxidisingof oxidising surface water... (8)conditions
Significant decrease in overlying rockmass thickness … (9)Approach of waste to the biosphere

Continuous changes in topography andsea-land distribution … (10)Slow, long-term changes ingroundwater flow paths andgroundwater chemistry

Mining of high economic value mineral
resources... (11)

Human intrusion into the repository

Uplift/erosion*3

Existence of mineral
resources

Effects on safety functions*1

Policy on site selection corresponding to each of the factors (5) – (11) is described in Section 3.1.2.
Movement and inflow of volcanic hydrothermal and deep-seated fluids includes geothermal activity associatedwith magma and hot rock.
*3
Effects of climate and sea-level changes are considered as erosion factors.
*2

The table also lists factors that may result in unintentional human intrusion in the future,for example as in Section 3.1.2 (2), although the existence of mineral resources is notconsidered as a “disruptive phenomenon”.

3-4

3.1.2 Approach to using influencing factors in site selection(1) Factors influencing pre-closure safety
For factors affecting the pre-closure safety of the repository, the basic approach describedin Section 2.2.1 is applied, as outlined in Table 3.1-4 (together with those factors impacting
long-term post-closure safety presented in (2) below).
Table 3.1-4 Basic approach to site selection for factors influencing safety

Distribution of
unconsolidated
Quaternary sediments Exclude areas where unconsolidated
Quaternary sediments are distributed at a
depth of 300 m or more

(2) - (4)High geothermal
gradients, hot springs,
earthquakes, tsunamis,
distribution of softrock, rock deformationand displacement

 Understand the extent of impacts during theoperational phase Apply suitable construction technologies
and consider the location of the surface
facilities in order to reduce potential
impacts(3)

Pyroclastic flow, lava
flow, lahar

 Understand the possible extent of impacts
during the operational phase Consider the location of the surface
facilities to avoid an impact
 Exclude areas with significant effects
 Exclude areas with poor geologicalenvironment characteristics(5) - (9)(11)

Detailed
Investigations

Spatial scale■ Regional■ Repository■ Panel(1)(6) - (8)(10)

Preliminary
Investigations

Basic approach to site selection

Literature
study

Influencing factors(See Tables 3.1-1 and 3.1-3 for
numbers)

Volcanic and igneous
activity, presence of
volcanic hydrothermal
fluids and deep-seatedfluids, seismic andfault activity, upliftand erosion

Existence of mineral
resources Understand long-term changes in thegeological environment characteristics due
to slow effects
 Examine appropriate engineering measures
minimising such impacts
 Confirm that the function of restricting therelease and migration of radionuclides ismaintained for a sufficiently long time Exclude the site from the selection if thefunction of restricting the release andmigration of radionuclides cannot be
assured Exclude areas where minerals of higheconomic value exist

In terms of the factors in Table 3.1-1, for (1), areas with unconsolidated Quaternary
sediments at depths greater than 300 m are excluded because it is considered that construction,
3-5

maintenance and operation of the underground facilities will be difficult [2] in such cases. For(2) and (4), responses and safety measures in facility design are examined based on pastmeasures for existing nuclear facilities and other relevant underground structures in order toreduce impacts on the target formation, surrounding rocks or the site in general (for details,
see sections 4.5 and 4.6). For (3), if it is impossible to resolve issues by robust design of thesurface facilities [2], the location of these facilities is determined considering the extent towhich events related to volcanic activity, such as pyroclastic flows, are likely to occur during
the operational period.
(2) Factors influencing post-closure safety
For factors affecting long-term post-closure safety, the stepwise site investigation processdescribed in Section 3.2.2 will be conducted according to the basic concept presented in
Section 2.2.2. Based on the results, when significant effects are expected from processes (5)and (9) in Table 3.1-3, or when their impacts may become significant in the future, the areasof influence will be excluded [1] [2] because the geological environment cannot provide longterm protection from disruptive events. Similarly, when significant effects are expected as aresult of processes (6) - (8), or when their impacts may become significant in the future, theareas involved are excluded [1] [2] because the geological environment cannot be expected toperform the function of long-term restriction of the release and migration of radionuclides.
For areas where no significant effects from natural perturbations are expected, now or in
the future, impacts of processes (6) - (8) are assumed. As for the slow and cumulative effects
of unavoidable natural phenomena ((10) in Table 3.1-3), appropriate engineering measuresdepending on the degree of the impact will be considered, taking the expected long-termchanges in the geological environment into account. It is then confirmed that the function of
restricting the release and migration of is maintained for a sufficiently long time. If the results
indicates that long-term confinement in the geological environment cannot be assured, the site
is excluded from the selection candidates.
As noted in (11), areas where economically valuable mineral resources exist undergroundare excluded because exploration and mining that may be conducted in future could lead tohumans coming into contact with the waste as a result of inadvertent human intrusion. The
economic value and existence of mineral resources are judged by considering the issue ofmining rights according to the applicable Mining Law, but additional considerations may be
applied later, as it is recognised that the future value of mineral resources is difficult to judge.
In addition to valuable mineral resources, evaluation of effects from geothermal resources, hotsprings and groundwater resources will continue after the Preliminary Investigation Areas
have been selected [2].
Based on the policy for site selection described above, the selection process is conductedin three stages as shown in Table 3.1-4. Suitable geological environments are thus identified.
The spatial scale corresponding to each stage of investigation is as described in Section 2.2.3
and Figure 2.2-2.

3-6

3.1.3 Selecting a geological environment suitable for geological disposal inJapan(1) Features of geological environments in JapanThe Japanese islands form an arc-shaped archipelago that extends from northeast to
southwest on the eastern margin of Eurasia. It is located on a plate convergence boundary
where oceanic plates (Pacific and Philippine Sea plates) subduct below continental plates(Eurasian and North American plates). The current consensus is that the formation of theJapanese archipelago in its present form [13] was closely related to changes in the plate layout
and movement patterns. The early form of the archipelago was formed around 30 to 15 My
ago, along with marginal seas such as the Sea of Japan, the Chishima Basin and the Shikoku
Basin. Subsequent plate subduction continued in almost the same positions, with the directionof movement of each plate set in the current pattern about 2 My ago [13] [14]. In the Japanesearchipelago, volcanic/igneous activity, seismic/fault activity and uplift/subsidence occur due
to such plate movements. These natural phenomena are typical for a tectonic belt distributed
along a plate convergence boundary and occur repeatedly or continuously on a geologicaltime scale of hundreds of thousands to millions of years. In addition, global climate and sealevel changes impact erosion and sedimentation.
The following summarises the main scientific understanding of natural phenomena inJapan. Supporting Report 3-1 summarises the situation as presented in the H12 report [7].
(i) Volcanic and igneous activity
Volcanic and igneous activity due to plate subduction is unevenly distributed. On thearc/volcanic line scale (hundreds to tens of kilometres), the position of the volcanic front has
not moved significantly over the past several My [7] [8]. In northeast Japan, Quaternary
volcanic activity occurs repeatedly in the back-arc region of the volcanic front, with areas
where volcanoes are densely located clearly distinguished from areas where there are no
volcanoes [15] [16]. Areas of volcanic activity are controlled by finger-like, high-temperatureareas with a width of about 50 km (called “hot fingers”), which are considered to be stable fora long time and associated with magma bodies deep underground [17] [18]. In southwest
Japan, although the volcanic front is not as clearly defined, Quaternary volcanic activity tends
to be restricted to areas closer to the Sea of Japan than to the Chugoku Mountains [19].
Regarding the extent of magmatic activity, with the exception of huge caldera volcanoes
and monogenetic volcanoes, individual volcanic vents (side volcanoes) that make up the
Quaternary volcanoes [20] [21] are generally distributed within a radius of about 15 km
around each Quaternary volcanic centre [22]. However, cases in which dyke magmas migrate
more than 30 km from the centre of Quaternary volcanoes have also been identified [23].
(ii) Migration and inflow of volcanic hydrothermal and deep-seated fluids
According to correlation diagrams, the pH of groundwater and hot spring water [24],within around 15 km from the centre of a Quaternary volcano is often acidic, with pH below 4
due to dissolved volatile components such as SO2, H2S and HCl released from magma.
Geothermal water convection is closely related to volcanic and other igneous activity. Hightemperature anomalies with a geothermal gradient exceeding 10 °C/100 m is highly correlatedwith the uneven distribution of the aforementioned Quaternary volcanoes [25] [26].
3-7However, in non-volcanic areas such as the Joban region, the Noto Peninsula, the Niigata
Plain, the Kii Peninsula, Arima Onsen and the northern Shikoku region, thermal water of nonvolcanic that does not originate from underground magma and non-meteoric fluids with highsalinity or high carbonate concentrations are distributed [27] [28] [29] [30] [31] [32]. These
deep-seated fluids include waters of deep subducted slab or mantle origin that have arisen
through fracture systems and mixed with groundwater (for example, Arima type hot water) or
oilfield porewater, in a formation that has been stagnant for a long time.
(iii) Earthquake and fault activity
Seismic/fault activity has not only occurred repeatedly in response to regional
compressive stress fields since the Quaternary [33] [34], but has also occurred due to theeffects of stress changes associated with mega-thrust earthquakes such as the 2011 Tohoku
Earthquake [35] [36]. “Active faults” with large displacements that are repeatedly active andcan be confirmed at the ground surface are compiled in geological databases [37] [38] [39][40] [41]. Given that new active faults are sometimes detected after the occurrence of anearthquake, it is possible that an active fault exists underground, although not confirmed atpresent [39], because any surface displacement is not evident or has been overlooked [26]. Asa specific example of this, surveys conducted after the occurrence of large-scale inland
earthquakes (such as the Tottori Prefecture Seibu Earthquake in 2000 and the Iwate-Miyagi
Nairiku Earthquake in 2008) confirmed relevant underground structures related to surface
ruptures and earthquake source faults in areas where no active fault had been identified
previously [42] [43] [44].
The range of influence due to fault movement is defined by the width of the process zone,
which is about 1% of the fault length. In this area, the fracture density is higher than thesurrounding host rock outside the fault zone [45] [46] [47]. Around a large active fault, otherfaults nearby may be reactivated, or areas of fault activity may expand to include multiple
faults [48] [49] [50].
(iv) Uplift, subsidence and erosion
Uplift and subsidence are ongoing over the last several hundred thousand years with apattern resulting from regional tectonic activity. Especially in areas where crustal deformationis active, significant linear erosion can be caused by rivers (exceeding 50 m in the last 100 ky).
In the Central and Southern Alps, it has been confirmed that significant denudation (in theorder of kms) has occurred since the Late Pliocene (≈ approximately 3 My ago) [51] [52] [53].
The current uplift rate in the Northern Alps can be as high as 5-6 m/ky [54].
In general, erosion and uplift rates in the mountains and hills reach equilibrium in 100s ky
to about a My [55]. In the upper to middle basins of rivers, erosion and uplift in a glacial
cycle almost balance out [56]. In coastal regions, sedimentation has continued since the
Neogene in some areas, while in other areas this was replaced by uplift in the Late Quaternary[57] [58] [59] [60] [61]. Close to river estuaries within uplift areas, since the Late Pleistocene
the undercutting rate during one glacial cycle exceeds the amount of uplift, with a maximum
value of about 100 m [62]. In terms of climate and sea-level changes that cause erosion [63], a
key factor is the drop in sea-level of up to 150 m during glacial periods [64].
As an oceanic plate subducts below a continental plate, accretionary complexes(wedges/prisms) of various ages accumulate along the boundary of the plates, the position of
3-8which may change with time. The Japanese archipelago is generally characterised bybasement that is older on the continental side and younger on the Pacific side [26]. The
geology of the archipelago is composed of various types of strata and rocks (GeologicalSociety of Japan, 2011) [26]. From the viewpoint of clarifying geological features that areimportant for site selection, the geology (Geological Society of Japan, 2011) [26] can be
subdivided into seven categories; Quaternary sedimentary rocks, Quaternary volcanic rocks,
Neogene and Pre-Neogene sedimentary rocks, Neogene/Pre-Neogene volcanic rocks,
Neogene/Pre-Neogene plutonic rocks, and metamorphic rocks, as shown in Figure 3.1-1. The
Pre-Neogene sedimentary and metamorphic rocks and the Neogene/Pre-Neogene plutonic
rocks that intrude into them, form the basement, which is covered by Neogene and Quaternary
sedimentary rocks and volcanic rocks that erupted at various times, or is exposed in someareas.

Quaternary sedimentary rocks

Quaternary volcanic rocks
Neogene sedimentary rocks
Pre-Neogene sedimentary rocks
Pre-Quaternary volcanic rocks
Pre-Quaternary plutonic rocks
Metamorphic rocks
Quaternary volcanoes

Figure 3.1-1 Geological map of the Japanese archipelago(Based on AIST 1: 200,000 Japan geological map [65])Due to the tectonic conditions noted above, Japan has large topographic relief, with mostof the archipelago covered by mountains and hills. At many locations, the mountains/hills
extend very close to the coast. With a relatively warm climate, abundant precipitation andgeographical features typical for a marine archipelago, the groundwater table is generally high.
Underground formations and rock masses are thus saturated with groundwater up to near the
ground surface. The groundwater chemistry varies over a wide range, depending strongly on
its origin, the type and hydraulic conductivity of formations and rocks in contact withevolving water, as well as the spatial distribution of topography and geological structures that
control groundwater flow [66]. In inland areas, not only groundwater of meteoric origin but
also magma-derived groundwater in volcanic areas and deep-seated fluids rising through fault
systems are identified [28] [29] [30] [31] [32]. In coastal areas, both meteoric groundwater
extending under sea areas and intrusion of saline groundwater into freshwater aquifers are
seen. Deep, old seawater (fossil seawater or oilfield brine), reflecting stagnation in formations
3-9

for tens of thousands to millions of years, generally shows a complex areal distribution [67][68] [69].
Closely related to the evolution of geological environments, resources such as coal, oiland natural gas, as well as various metal and non-metal minerals, are known in specificlocations [70] [71] [72] [73].
(2) Environments with favourable characteristics from the viewpoint ofgeological disposal
Characteristics of the deep geological environment in Japan, in terms of requirementsfrom the viewpoint of waste confinement shown in Table 3.1-2, are summarised in Table 3.15 [8].
Table 3.1-5 Favourable geo-environmental characteristics widely found deep underground inJapan (Modified from Advisory Committee for Natural Resources and Energy, 2014 [1])Common characteristics of the geologicalenvironment

Requirements for the geological environment

Low rock temperature

Geothermal gradient is generally 3 - 5 °C/100 m,
except for higher values associated withvolcanic areas

Slow groundwater flow

Hydraulic gradient at relevant depths isgenerally in the order of 0.001 to 0.01.
Hydraulic conductivity is in the order of 10 -12 to
10-6 m/s and decreases with depth. Larger
hydraulic gradients may occur in lower
hydraulic conductivity rockMechanicalenvironment

Small rock deformation

Deep rock generally has high compressive
strength and elastic modulus, such that longterm creep is within the range that can behandled by repository design

Chemicalenvironment

Groundwater pH not extreme,
Deep groundwater generally has a pH of 6 to 9,
chemically reducing, carbonate
with reducing conditions and carbonate
concentrations less than 0.5 mol/l concentrations less than 0.1 mol/l

Thermalenvironment

Hydraulicenvironment

Geological environments in Japan have been identified in which favourable geologicaldisposal characteristics have been maintained over a long period of time. Through researchbased on natural analogues and paleo-hydrogeology, focusing on evidence preserved ingroundwater, rocks and minerals (record of paleo-hydrogeological evolution), favourable
hydraulic and chemical conditions have been shown to persist for a long time, despite
continuous or periodic uplift/erosion and climate/sea-level changes.
Examples of relevant studies are summarised below.


At several hundred metres below sea-level in the Horonobe area of Hokkaido,
Quaternary sedimentary rock contains groundwater recharged during the glacial
retreat, tens of thousands of years ago [67] [68] [69]. This is interpreted as being
related to movements of the saltwater/freshwater boundary caused by glacial cycle
sea-level fluctuations [61]. In addition, deeper Neogene sedimentary rocks have a low
hydraulic conductivity [74] and retain paleo-oceanic formation water with an age(residence time) of several My [75]. In such a closed environment, a reducing
3-10

environment would persist as a result of continued water-mineral-organicmicroorganism interactions [76].
In Neogene sedimentary rocks in the Yokosuka area of Kanagawa Prefecture and in
Cretaceous sedimentary rocks (classified as Pre-Neogene sedimentary rocks) in theKushiro area of Hokkaido, ancient groundwater is found with a residence time of atleast ≈ 7 My and ≈ 2 My or more, respectively [77] [78]. These formations mainly
consist of mudstone and shale, with groundwater flow being extremely slow in such
low hydraulic conductivity formations.



Neogene sedimentary rocks, including the uranium deposits in the Tono area, Gifu
Prefecture, have been affected by natural phenomena such as faulting, uplift/erosion,and sea-level changes. Despite this, uranium deposits have been preserved for as longas 10 My [79]. It is considered that the uranium did not dissolve because groundwater
pH and a reducing environment were maintained as a result of water-mineralmicrobial interactions so that the redox front near the surface is almost static. [80].

Cretaceous granites (classified as Neogene and Pre-Neogene plutonic rocks) in the
Tono area of Gifu Prefecture indicate some changes in the chemical environment withtime. This is confirmed based on changes in calcite crystals on the fracture surface(groundwater flow paths) and the heterogeneous distribution of some trace elements.
However, no significant change in the redox environment was observed and reducing
conditions were assumed to have been maintained deep underground [81].

At the Adera fault2 in Central Japan, although oxidising surface water infiltrated due topast uplift, its hydraulic and chemical impact was limited [82] [83]. Iron hydroxide
and calcite were formed in the damaged zone around the fault, which is a focus forgroundwater flow and mass transfer. Such iron hydroxide and calcite lead to a
reduction in hydraulic conductivity due to clogging (blocking rock porosity) andresults in persistence of the pH and reducing environment due to chemical buffering[82] [83].

Thus, there are areas deep underground in Japan where geological environments withfavourable characteristics have been confirmed, and therefore it is reasonable to assume thatsuch areas are widely distributed. By avoiding areas showing significant effects of natural
perturbations and assessing the temporal and spatial stability of key characteristics of thevolunteer sites3, it is considered possible to identify environments suitable for geologicaldisposal in Japan, despite its location in an orogenetic belt. The specific procedure fordetermining suitable geological environments is described in the next section.

3.2 Process for identifying environments suitable for geologicaldisposal
3.2.1 Basic strategy and methodology
In geological disposal, it is essential to select a geological environment where the safetyfunctions will be maintained for a sufficiently long time. For this purpose, NUMO will
2

Although the Adera fault is an active fault, it is described as one of the cases related to the buffering ofhydrological and chemical changes observed in the vicinity of the fault.
3
In this report, the areas that applied for a literature survey and the areas that accepted a proposal for a literature
survey from the Japanese government are collectively described as “the volunteer sites”.

3-11

proceed with stepwise site selection based on the concept of building understanding ofgeological events and processes that have occurred in the past, thus forming the basis forunderstanding of those occurring now and in the future. Specifically, following a site
assessment policy based on factors affecting safety as described in Section 3.1.2, the spatial
scale of potential siting areas and the relevant geological events and processes to beinvestigated are narrowed down through the stages of literature survey (LS), preliminary
investigations (PI), and detailed investigations (DI). The geological investigations andevaluations are conducted with a gradually increasing level of detail. Site selection is thus
implemented as follows:

Clearly unsuitable sites, owing to distribution of Quaternary unconsolidated deposits,
existence of mineral resources or risks of perturbation due to impacts of volcanic
activities, migration and inflow of volcanic hydrothermal and deep-seated fluids,
earthquake/fault activities and uplift/erosion, are excluded. The basic concept forinvestigation/evaluation of natural perturbations is presented in Supporting Report 3-2,
and the specific approaches involved for volcanic/igneous activity, earthquake/fault
activity, and uplift/erosion are presented in Supporting Report 3-3, Supporting Report
3-4, Supporting Report 3-5. The current status and research and development (R&D)
issues related to the methods for evaluating the effects of natural perturbations, as wellas the combination of individual technologies that make up theinvestigation/evaluation programme, are presented in Supporting Report 3-6.

For sites where significant effects of natural phenomena can be avoided, thecharacteristics of the geological environment from the viewpoint of confinement of
the waste are evaluated, and areas that are clearly less suitable can be excluded. It isnecessary to consider the effects of slow, cumulative processes, such as uplift anderosion, such characteristics. For this reason, past temporal and spatial changes of thegeological environment will be examined. By quantifying the range of variations
through modelling, NUMO aims to confirm that favourable characteristics are stableenough to support sufficient barrier performance in excess of 100 ky. The basic
approach and workflow for investigation and evaluation of the geological environment
are presented in Supporting Reports 3-7 and 3-8. The current status and research anddevelopment (R&D) issues relating to methodology, as well as the combination ofindividual technologies that make up an integrated programme are identified in
Supporting Report 3-9. An approach for assessing long-term changes in geologicalcharacteristics is presented in Supporting Report 3-10. In addition to the effects ofnatural perturbations, individual investigation and evaluation technologies related togeo-environmental characteristics and their long-term changes are summarised in
Supporting Report 3-11.

Particularly for the 100 ky timescale, it is necessary to consider events that could havea significant effect on the safety functions of the repository even if their likelihood of
occurrence is extremely small. The uncertainties in the occurrence of events and theireffects on the geological environment are evaluated using a combination of
extrapolation and stochastic methods, taking into account limitations of the associatedknowledge base. This is described in detail in Section 3.2.2 (2).

At each stage of the site selection process, following assessment of eligibility in terms ofstatutory requirements and siting factors, NUMO will examine repository concepts tailored tothe site to confirm engineering feasibility and post-closure safety. This will allow
identification of further information to be acquired on the geological environment, the level of
3-12

detail needed, and the methods to be applied in order to reduce uncertainties resulting fromthe lack of information or how it is interpreted. Although the items to beinvestigated/evaluated will depend on site conditions, examples are listed in Table 3.2-1 basedon experience accumulated in Japan and overseas [84] [85] [86].

3-13Table 3.2-1 Examples of geological investigations/evaluation topics for sites
Purpose of
survey/evaluation

Avoiding impactsof natural
perturbations toensure operational
and post-closure
safetyConfirmation ofengineering
feasibility

Survey/evaluation items
Volcano andigneous activity

 Probability and extent of magma movement from Quaternary volcanoes
 Probability and extent of new volcanoes (magma chamber development)

Volcanichydrothermaland deep-seatedfluids* Potential extent of thermal and chemical effects of future thermal fluid
inflow

Earthquake andfault activity

 Distribution of active faults and potential future activity
 Extent of hydraulic and mechanical impacts associated with fault movement
 Changes in the geological environment due to seismic activity

Uplift anderosion Future uplift/subsidence rate Future erosion rate and long-term topographic changes
 Climate/sea level change

Geologicalstructure Geometry of geological structures relevant to groundwater flow paths Geometry of strata and rock bodies as potential host rocks
 Spatial distribution of lithological discontinuities and fractures in strata androck bodies
 Geometry of water-conducting microstructures influencing nuclide
migration

Thermalenvironment Geothermal gradient Thermal characteristics of strata and rock bodies

Hydraulicenvironment

 Spatial distribution of hydrogeological structures
 Spatial distribution of groundwater flow
 General hydraulic characteristics of strata and rock bodies
 Hydraulic properties of water-conducting structures
 Gas migration characteristics of strata and rock bodies
 Gas migration characteristics of preferential flow structures
 Spatial distribution of hydraulic head and water temperature Solute migration characteristics of water-conducting microstructures

Mechanicalenvironment

 Physical characteristics of strata and rock bodies
 Physical characteristics of faults and fractures
 Mechanical characteristics of strata and rock bodies
 Mechanical characteristics of faults and fractures

Chemicalenvironment

 Spatial distribution of groundwater chemistry and tracer isotopes
 Spatial distribution of groundwater pH/Eh (oxidation-reduction potential) Properties and distribution of gas in groundwater Properties and distribution of gas in strata and rock bodies
 Chemical effects of colloids, organic substances and microorganisms
 Solute retardation characteristics of water-conducting microstructures

Long-term
evolution ofgeologicalcharacteristics

 Evolution of regional stress field Topographical and geological structure evolution
 Temporal changes in groundwater flow and hydraulic properties
 Evolution of factors controlling groundwater chemistry Temporal changes in thermal, physical and mechanical properties of strata
and rock bodies

andAssurance of postclosure safety*The movement of volcanic hydrothermal fluids include geothermal activity associated with both magma andother deep rocks.

Information obtained through site investigations is interpreted using a multidisciplinary
approach to construct a common conceptual model of the site geology (SDM). The current
geological environment, such as spatial distribution and geometries of geological
3-14

formations/rock bodies/geological structures, their thermal/hydraulic/mechanicalcharacteristics, and the spatial distribution of groundwater chemistry (as listed in Table 3.2.1)
are represented (visualised) within the SDMs. Temporal changes in such characteristics arealso conceptually modelled. Figure 3.2-1 shows an example of a SDM.
Such a SDM allows repository concepts to be developed, in terms of both repository
design and the associated safety assessment, and further requirements for future geologicalcharacterisation to be identified. Thus, NUMO will conduct iterative site investigation while
repeating the examination of the repository concepts based on the updated knowledge basecaptured in the SDM.

Figure 3.2-1 Example of a SDM constructed based on geological investigations & evaluations.
Left: Geological model (Ota et al., 2007 [74]), Right: Conceptual model of long-term changes of
the geological environment (Niizato et al., 2010 [87])When conducting the iterative site investigations described above, the reliability of thesafety assessment depends on the range, quantity and quality of the acquired geologicalinformation [4]. In the site investigations, as noted in Section 2.5.2, quality management of
the series of work processes from acquisition to interpretation of geological information isindispensable, as is recognised internationally [88]. Specifically, applying practical quality
control/assurance concepts as captured in documents [89] [90] currently being developed,
consistent quality management can be implemented during each of the stages of the site
selection process. At each stage, NUMO will confirm that not only that the report of theinvestigation and evaluation results is prepared in accordance with the Final Disposal Act
Enforcement Regulations, but also the resultant SDM meets the specified requirements.
Supporting Report 3-12 describes, for the example of the PI stage, an outline of the
documents related to quality control/assurance for investigations on land (inland/coastal) and
offshore areas, as well as the associated construction of the SDM.
The details of the basis for the staged site investigations are described below.

3-15(1) Reduction of uncertainty using an iterative approach
The investigations target geological environments with heterogeneous characteristics,
evolving over an extremely long timescale, and hence data and their interpretation include
temporal and spatial uncertainties. Handling such uncertainties is one of the most important
issues, not only in evaluating the probability of occurrence of natural perturbations and theirrange of influence, but also in evaluating the spatial distributions of geological characteristics
and their temporal changes [4]. Geological factors with high uncertainty (e.g. spatial
distribution of faults and hydraulic characteristics) are thus a special focus during site
investigations. When SDMs are developed for specific sites, information on key uncertainties
will be passed on to the repository design and safety assessment teams as an inherent
component of the SDM (and associated data set). Feedback from these teams will thus allow
NUMO to focus future work on the more significant geological factors. Based on this, andfollowing the basic concept for planning site investigations and evaluations [91], a plan isdeveloped to reduce such uncertainties by refining the targets and priorities set for the nextstage (as indicated in Sections 3.2.2 (2) and 3.2.2 (3)).
In the JAEA Mizunami Underground Research Laboratory (URL), the effectiveness ofsuch an iterative approach for reducing uncertainty has been illustrated, with developing site
investigations leading to refinement of the associated SDM [74] [92]. In terms of thegeological structures present, more detailed information on the distribution of formation
boundaries, faults and fracture zones is acquired as the investigations progress and visualisedin the evolving SDM. Examples in which the understanding of the geological structures has
been developed through this approach are presented in [92] [93] [94] [95]. For groundwaterflow in particular, the influence of the uncertainty in fault hydraulic conductivity is evaluated,
and the results are reflected in the investigation plan for the next stage, allowing such
uncertainty to be gradually reduced during the progress of investigations [92].
Furthermore, such an iterative approach has been adopted in site investigations overseas.
For example, during Posiva's investigation at the Olkiluoto site in Finland, studies from theground surface and subsequent investigations in the Onkalo underground rockcharacterisation facility (RCF) were implemented in stages, with progressively increasing
detail. In conjunction with updating the SDMs in areas such as structural geology of the hostrock, rock mechanics, hydrogeological structures and groundwater geochemistry, theassociated reductions in uncertainty and improvements in reliability have been evaluated [96][97] [98] [99]. A similar approach was followed for investigations conducted by SKB at theForsmark site in Sweden [100] [101] and those conducted by Nagra at the Wellenberg site in
Switzerland [102] [103].
As site investigations progress, the significance of residual uncertainties in terms ofdesign or safety generally decreases. However, it is possible that new facts and new
uncertainties will be identified by iterative geological investigations. Therefore, a decision to
end site investigations is made only if sufficient quality and quantity of geologicalenvironment information has been gathered to meet requirements and any newly acquired
information does not impact the associated safety assessment.
(2) Use of the SDM to integrate repository design and safety assessment
It is important to store the information obtained from the geological investigations in aformat that can be used for disposal site design and safety assessment, sharing the resultant
site understanding as well as associated uncertainties. As described earlier, SDMs are
3-16

constructed in which the current state of the geological environment is visualised in three
dimensions. The characteristics include the spatial distribution and geometry of geologicalformations, rock bodies, and geological structures, together with associated thermal, hydraulic,and mechanical properties, as well as the spatial distribution of groundwater chemistry.
Temporal changes in these geological characteristics are also modelled.
In order to effectively and efficiently progress from the site investigations to the
production of the SDM, a “geosynthesis” methodology will be used [85] [92] [104].
Specifically, for the geological environment on the target spatial scale, relevant issues related
to its characteristics will be defined (as shown in Table 3.2-1) and these assessed, together
with their long-term evolution. These are then incorporated into the SDM that include asufficient range of credible evolution scenarios. To understand these items properly,
geological information obtained from the investigations is interpreted based to produce
conceptual models of geology and geological structures, considering both multi-disciplinary
consistency and continuity between different scales. Based on these, the SDM is further
developed, incorporating the spatial distribution of formations and rock bodies in 3D, together
with major faults and lithostratigraphy. Using this as a common basis, models ofhydrogeology, groundwater chemistry and rock mechanics models are integrated within it. By
applying the iterative approach described above, the effects of uncertainty of data and lack ofunderstanding of the geological environment on the design and safety assessment of therepository at any stage (or phase) can be assessed. Thus, the focus for geologicalcharacterisation in the next stage (or phase) can be narrowed down, and the investigation plan
optimised.
Such geosynthesis allows the comprehensive knowledge base developed by site
investigations [74] [84] [92] [105] to be effectively captured for site-specific conditions. Thisknowledge base can be managed by advanced tools – such as the next-generation sitecharacterisation Information Synthesis and Interpretation System (ISIS) developed by JAEA
[105]. For each investigation scale and stage, this includes not only integrated data flow
guidance to systematically organise links between various investigation, data acquisition, datainterpretation, and data integration (modelling/analysis) processes, but also to summarise thetechnical knowledge and know-how obtained.
Such a method of integrating geological environment information (or methods based on asimilar approach) has been widely applied to overseas site investigations [99] [101] [103][106] [107] [108]. It is considered to be particularly useful in facilitating use of geologicalinformation for constructing a safety case [104], by linking site investigations to repository
design/safety assessment. This method was applied in JAEA's URL projects at both Horonobe
and Mizunami, and its effectiveness has been shown during optimisation of the undergroundconstruction on the basis of SDMs [74] [92]. Figure 3.2-2 shows a concrete example of a
SDM constructed by integration of geological information from JAEA's URL project inMizunami.

3-17Figure 3.2-2 Example of a SDM for the area around the Mizunami Underground ResearchLaboratory. Left: geological model showing the spatial distribution of geological structures.
Right: hydrogeological model showing the spatial distribution of hydraulic conductivities forgeological structures (Modified from Saegusa et al., 2007 [92])In Section 3.3, the processes and results of stepwise SDM construction are described indetail.
3.2.2 Stepwise implementation of site investigations
NUMO has developed workflows based on the staged site investigation proceduredescribed in Section 3.2.1 [109]. As noted in Section 1.2, the Geological Disposal Technology
Working Group (hereafter “WG”) stated that safe geological disposal has to be established byfield investigations, in which site suitability is confirmed in steps that show conformity withlaws and regulations. The resulting Nationwide Map of Scientific Features for Geological
Disposal (hereafter “Nationwide Map”) was then developed based on specific requirements
and criteria [110]. In response, NUMO reviewed the processes for implementing the LS ofvolunteer sites [111].
In this section, based on input from the WG together with documents related to the LS
stage prepared by NUMO [111], the technical processes4 for site investigations in the stagesof LS, PI and DI defined by the Final Disposal Act are described. Figure 3.2-3 shows theoverall work flow for the LS and PI stages, reflecting the basic concept and proceduredescribed in Section 3.2.1.

4This section describes only the processes for the site investigations from a technical perspective.

3-18

Figure 3.2-3 Workflow for LS and PI stages

3-19

(1) Literature survey stage
In the LS stage, literature information is collected based on statutory requirements andrequirements/criteria related to the presentation of the scientific characteristics of thevolunteer sites and their surroundings. The information collected includes publications (suchas geological maps), scientific papers and research reports, which need to be checked in termsof quality and reliability and managed using appropriate systems [109]. The eligibility of
application areas is assessed based on review of the collected information, with a focus on
identifying the likelihood and extent of potential impact of natural perturbations in order to
exclude areas that are obviously ineligible. In addition, an assessment of the distribution of
unconsolidated Quaternary sediments and the presence of mineral resources is conducted;
again, areas that are clearly ineligible are excluded. When factors are difficult to evaluate
based on literature information, the necessity of exclusion will be judged after the subsequent
PI stage.
For areas where eligibility to proceed to the PI stage is confirmed, the regional-scale SDM
is refined and long-term changes in geological characteristics are evaluated. From theviewpoint of confirming the engineering feasibility of the repository and its long-term postclosure safety, the geological characteristics in Table 3.2-1 are investigated and evaluated,along with their potential evolution. A preliminary study is conducted on the specifications of
the engineered barriers, the concept of the disposal facility, and the post-closure safety of therepository. With prioritisation of geological information to reduce uncertainties that could
have a significant effect on the repository concept, a PI implementation plan5 is formulated
based on the geological investigation and evaluation planning manual [91].
(i) Exclusion of areas significantly impacted by natural perturbations(a) Volcanic and igneous activity
Based on the literature information, the risk of volcanic activity in the coming tens of
thousands of years results in exclusion of areas within a radius of 15 km from the centre of
Quaternary volcanoes as well as areas of calderas larger than 15 km with Quaternary volcanic
activity [1] [2]. Additionally, the activity history of Quaternary volcanoes, with considerationof temporal and spatial changes related to the history of geological structure development,
will be evaluated. Further areas clearly identified as likely to have magma intrusions and/or
eruptions at the surface over the next tens of thousands of years are also excluded.
Thus, areas that are clearly expected to have a significant risk of such impacts can be
excluded. However, if volcanic rocks (eruptive or intrusive) or volcanic terrain with anunclear age are found in the vicinity, it may not be possible to make an appropriate judgement.
The uncertainties resulting from this are examined in detail in the PI stage.
(b) Migration and inflow of volcanic hydrothermal and deep-seated fluids
Extending from the assessment above, the relationship between underground structures
and the location of magmatic activity is assessed. If significant effects from geothermal
activity are expected, such as hot water convection due to magma or hot rock, the area
involved is excluded [1] [2]. When the presence of high-temperature, low-pH or high5

It is possible that part of the PI implementation plan is formulated in the PI stage.

3-20

carbonate concentration water is confirmed or is likely in the future, areas where impacts are
considered to be significant will be excluded [1] [2].
If the formation and age of alteration zones due to geothermal activity are unknown, or
when there is little literature information on deep-seated fluids, these factors are examined
after obtaining detailed data in the PI stage.
(c) Earthquake and fault activity
Literature information for both land and coastal sea areas allows exclusion of those withknown faults with a large displacement that have been repeatedly active in recent geologicaltimes [1] [2]. The fault influenced zone is conservatively defined as about 1% of the fault
length, based on the literature data that gives values of about 1/350 to 1/150 of the fault length[112]. On the basis of a more detailed evaluation of earthquake and fault activity, thefollowing areas are excluded (Figure 3.2-4): (1) areas included in damaged zones outside
main fault zones, (2) areas with evident large folds and flexures that continue to be active, (3)areas where there is a clear probability that new deformation may occur due to the activity of
the underlying fault structures [1] [2].

Figure 3.2-4 Influence of fault activity (Modified from Advisory Committee for Natural
Resources and Energy, 2017 [2])In the evaluation, a proper judgement cannot be made for faults with a small displacement,
faults with unknown activity history, or faults with surface signatures but unknown
underground structure. These are examined in detail in the PI stage.
(d) Uplift/erosion
Based on literature data, areas where the determined erosion in the last 100 ky exceeds
300 m are excluded [1] [2]. Additionally, areas such as inland mountains where uplift is
expected to exceed 300 m over the next 100 ky, and coastal areas where erosion is expected to
exceed 300 m over the next 100 ky are excluded. This is determined either by an equivalent
uplift rate in inland mountainous areas or a combination of such uplift with the impact of
expected future erosion in coastal areas.

3-21

In cases where there is more than one interpretation for the age of the topographic surfaces
used for calculating uplift, or where the uplift estimated by multiple methods differs, more
information is collected and evaluated in the PI stage.
(ii) Exclusion of areas with unconsolidated Quaternary sediments and mineral
resources
Areas are excluded as outlined in Section 3.1.2 (2) if unconsolidated Quaternary
sediments are present at depths of 300 m or more. In addition, areas are also excluded if
mineral resources with significant economic value exist [2].
If the literature information is inconclusive, or cases where materials present are notcurrently considered as mineral resources but could be considered valuable in the future (suchas geothermal/hot springs and groundwater resources), these factors are examined further in
the PI stage.
(iii) Evaluation of geological environment characteristics and their long-term
evolution
Based on comprehensive interpretation of literature information, geological characteristicsof the volunteer areas are determined. From an understanding of the spatial distribution andgeometry of formations and rock bodies, together with their hydraulic/mechanicalcharacteristics, a regional-scale SDM that visualises these factors is constructed. As stated in
Section 2.4.4, because it is important to consider groundwater migration scenarios whenassessing post-closure safety of the repository, groundwater recharge and runoff areas will be
determined and the regional groundwater flow field evaluated. This, together with anassessment of potential impacts of natural perturbations, allows extrapolation to determine
future evolution of key characteristics, such as hydraulic and chemical conditions, using
paleo-hydrogeological techniques [61] [87], together with evidence of long-term stability insimilar geological environments [113].
In the LS stage, it is likely that sufficient information on the geological environment
cannot be obtained and the resultant SDMs and their assumed evolution involve large
uncertainties. By assigning priority to geological information that will reduce uncertainties
relevant to the repository concept and associated safety case, the required supporting
information can be obtained during the PI stage.
(2) Preliminary investigation stageDuring the PI stage, according to the investigation plan formulated during the LS stage, aseries of systematic investigations is conducted for the site and surrounding areas, aimed toconfirm LS conclusions regarding site eligibility. Based on the information obtained, thespatial distribution of key characteristics and their long-term evolution are captured withinmodels at regional and repository scales. Further information on geothermal/hot springs,
earthquakes, ground deformation/displacement and pyroclastic flows acquired at this stage(corresponding to influencing factors (2) - (4) in Table 3.1-1), allows the extent of potential
perturbation impacts to be examined, as well as possible countermeasures to reduce or avoid
their impacts (for further details, see Sections 4.5 and 4.6).

3-22

The PI stage involves stepwise implementation in multiple phases to ensure that thepriority geological information specified during the LS stage can be obtained reliably andefficiently [91] [109]. An example of the implementation protocol for such PI is presented
below, but may be modified depending on the specifics of the site as found during the LS.
Preliminary regional investigations at the volunteer site include surface exploration,hydrological surveys, observation of crustal deformation and microearthquakes, together withgeophysical exploration from the air, ground and sea surface. If necessary, relatively shallow
borehole investigations (up to a depth of several hundred metres) can be included (e.g.
borehole drilling, physical/fluid logging, hydraulic/mechanical tests, and laboratory tests ofcores). Supplementary investigations may also be conducted in surrounding areas tounderstand the regional effects of natural perturbations and the development history ofregional geological structures [1] [2] [22].
Based on the results of such initial investigations, LS conclusions with regard to theeligibility of a volunteer site may be confirmed. Within the preliminary investigation areas(PIAs), the potential for significant future impacts from natural perturbations is assessed in
the light of the limitations of the investigation and evaluation techniques available. After
reconfirming the geological characteristics and processes to be considered, the regional-scale
SDM from the LS is updated, and a more detailed repository-scale SDM is constructed basedon it. A long-term conceptual model of geological evolution is then refined from the
expanded knowledge base and priority information is identified to reduce key uncertainties inthese models. This leads to a review of the preliminary investigation plan to optimise direct
investigation of the deep underground environment as described next.
Deep subsurface investigations (phase 2 of the PI stage) include borehole investigations
and high-density geophysical surveys targeting depths of 1,000 m or more, thus obtaining
information that cannot be derived via surface investigations. The locations and depths of theboreholes to be drilled are determined based on the results of the preliminary repository
concept, updated based on the SDMs noted above. From these results, and with considerationof statutory requirements for the selection of detailed investigation areas (DIAs), NUMO will
evaluate the range of influence and possible future impacts of natural perturbations on therepository, associated tunnel excavations (e.g. distribution of unconsolidated sediments orrocks with insufficient mechanical strength) and the safety functions of a repository (e.g.
water-conducting features, risks of significant groundwater flow and/or associated chemical
disturbances). Areas that are clearly poorly qualified (based on results of evaluation of therepository safety functions) are excluded. The SDM representations on regional andrepository scales are refined and the conceptual model of long-term geological evolution at aregional scale is improved.
From the PIAs (or options within a PIA), those that are more appropriate from theviewpoint of being a candidate (or prospective) repository site are selected. Included in theconsideration of candidate sites are the ease with which further site characterisation andconstruction can be carried out, whilst also minimising environmental impacts. For the target
host rock formations, understanding of their spatial distribution (depth and extent), geologicalcharacteristics and long-term evolution is captured within the repository-scale SDM.
Suitability is then evaluated by taking into account also the efficiency and economic aspectsof required detailed investigations. For potential host rocks, the depth at which the repository
can be constructed and the resulting available area are determined (for details, see Sections
4.3 and 4.5.4 (1)).

3-23The detailed investigations are again planned with a focus on geo-environmental
information that should be acquired with priority in order to reduce uncertainties that have alarge impact on the design of the engineered barriers and repository layout, as well as theresults of the safety assessment. In addition to a repository design and construction plan
aimed at reducing the environmental disturbance due to the construction and operation of therepository, a monitoring plan will be produced in order to determine the baseline conditions
used for safety assessment, as described in Section 2.3.3 (3).
(i) Exclusion of areas where effects of natural perturbations are significant(a) Volcanic and igneous activity
Based on newly acquired geo-environmental information, the evaluation results from theLS stage can be confirmed, ensuring that areas where significant impacts of natural
phenomena are expected to occur in the future, are excluded [2].
Specifically, in the PIA and its surroundings, information from surface exploration,geochemical analyses of surface springs, and geological/geochemical investigations using
boreholes is synthesised to determine the presence or absence of indications of volcanic
activity and, in particular, high temperature anomalies that may indicate hidden magma
chambers. Surface exploration, geological analyses (such as radiometric dating and chemical
analysis of volcanic rocks), and geophysical surveys such as airborne/terrestrial
electromagnetic surveys and microearthquake observations are combined to obtain
information on the extent and frequency of volcanic and igneous activity over the Quaternary
and any associated potential for future magma chamber formation.
(b) Migration and inflow of volcanic hydrothermal and deep-seated fluids
The expanded knowledge base will allow confirmation of exclusion of areas where
significant thermal or chemical effects from inflow of volcanic hydrothermal and/or deepseated fluids are expected in the future [2].
Specifically, investigations will target information on geological structures that canpotentially form flow paths for such fluids, the temperature distribution of deep waters, andthe spatial variation of groundwater chemistry. Together with the assessment of volcanic
activity described above, the existence and distribution of hydrothermal fluids and their range
of influence are evaluated.
(c) Earthquake and fault activity
The PI stage extends knowledge on the area of influence of faults with repeated activity
and large displacements, the area of their associated deformation zones, the presence of active
folds and flexures, to assure that areas likely to be significantly affected by such activity inthe future are excluded [2]. In particular, this confirms that faults and folds that haveundergone significant deformation in the Quaternary are unlikely to have significant effects inthe future.
Specifically, surface fault exploration, terrace topographical surveys, micro-topographicalsurveys, geophysical surveys, borehole investigations, and radiometric dating of materials in
fault and fault zones are conducted. This provides information on the presence or absence ofevidence of fault activity, the width of any faults and/or fault zones, relevant geological3-24

structures at depth, the history of fault activity in the Quaternary, the structure of folds andflexures and their activity pattern, and the deformation mode of relevant formations. Throughsynthesis of this information, the understanding of the range of influence of such features is
improved. The probability that faults are concealed underground, that faults branch (seeFigure 3.2-4) and extend further, and that new fault displacement occurs in the vicinity ofexisting faults, will also be investigated. Based on these results, it is possible to determine
areas where such activity occurs and where their effects, such as regional deformation or fault
movement, are likely to be significant, making engineering counter-measures difficult.
Although activity in recent geological times may be unknown in some cases, the possible
effects of future displacements of formations and expansions of fault zones is evaluated for
those where the possibility of activity cannot be ruled out, assessing possible impacts on key
safety functions. For places where strata and rock bodies with quite different properties are incontact with each other on a large scale, such as the edges of sedimentary basins, based on
information from regions with similar conditions, areas are identified where future
displacements may have significant effects on safety functions environment. This includes
taking into account the onset of future fault activity and the possibility of fault activity that isin the opposite direction to the present (inversion tectonics).
(d) Uplift/erosion
The PI stage will confirm that areas are exclude if a significant impact over relevanttimescales is likely for the expected repository depth [2].
Specifically, topographic surveys, geological exploration, geophysical surveys andborehole investigations are carried out. These include assessments of: the distributions andformation ages of terraces and eroded surfaces, sediment distributions; depths, formation ages
and shapes of dissected valleys and riverbed slopes. Through comprehensive interpretation ofthis information, future uplift and associated erosion, considering climate and sea-level
fluctuations, is estimated. In particular, the extent and distribution of erosion in the future isestimated based on consideration of contributions from general denudation, lateral andvertical erosion by rivers (including consequences of river capture) and sea erosion in coastal
areas. The specific impacts of uplift/incision of rivers due to sea-level change, glaciation, andpermafrost in the past several hundred thousand years is also considered. As background, it isimportant to evaluate long-term crustal movements determined from geological observations
and relate these to short-term crustal movements as observed geodetically.
(ii) Exclusion of areas with unconsolidated Quaternary sediments and mineral
resources
The PI stage will extend investigations of distribution of unconsolidated Quaternary
sediments at the assumed repository depth, as well as for the existence of mineral resources ofhigh economic value and, if they exist, these areas are excluded as described in Section 3.1.2(2) [2].

3-25

(iii) Evaluation of geological environment characteristics and their long-termchangesDuring the preliminary regional investigations, information on the distribution andgeometry of strata and rock bodies, faults and fracture zones at or near the surface, as well as
their thermal and mechanical properties, is obtained mainly by surface mapping, geophysicalsurveys from the air/ground/sea, trench surveys, surface hydrological investigations, andshallow borehole investigations. Groundwater recharge and runoff areas are identified,
together with information related to surface hydraulics and geochemical properties of thesurface waters and shallow groundwater.
During the subsequent investigations, information in three dimensions from deep
boreholes can be obtained. The information obtained includes: structural geology(faults/fractures, folds, facies distribution, etc.), rock thermal and mechanical properties(thermal conductivity and mechanical strength, etc.), hydraulic properties and solutetransport/retardation characteristics of strata/rock bodies/faults and fractures and thegeochemistry of groundwater. Through a comprehensive interpretation of the acquired
information, the extent of geological formations and rock bodies with favourable
characteristics is confirmed from the viewpoint of tunnel excavation and radionuclide
containment. By obtaining paleohydrogeological evidence (e.g. isotope ratios, crystal
structures of fracture-filling minerals) from groundwater, rocks and minerals, the long-term
evolution of the hydro-geochemical environment is evaluated.
(3) Detailed investigation stageDuring the detailed investigation (DI) stage, a series of systematic investigations arecarried out in two phases; the first phase from the ground surface and the second phase withina special underground investigation facility (UIF) [109].
During the surface investigations, which may be carried out at one or more detailed
investigation areas (DIAs) and/or a number of sites within these DIAs, information on thegeological environment around the proposed repository site is acquired according to an
Investigation Plan (IP) formulated during the PI stage. Based on this newly acquired
information, the regional- and repository-scale representation of the SDMs and the conceptualmodel of long-term evolution are re-assessed and updated. To understand in detail thehydraulic and mechanical characteristics of representative host rock(s) and the nuclide
migration and retardation characteristics of any water-conducting features, smaller scale
representations of the SDMs are constructed that focus on the size of disposal panels and also
the water-conducting microstructure of faults and fractures. Furthermore, geologicalinformation to reduce the uncertainty of parameters that have a high impact on these models
is prioritised. Specific items to be considered in further studies within the UIF is thendetermined, together with the required level of detail of such investigations.
For the second phase of DI, a UIF is constructed in a suitable area of the selected hostrock and investigations conducted to understand in more detail the underground geologicalcharacteristics and the temporal and spatial changes associated with this construction [114].
Existing experience and technology from domestic and overseas URLs can be extended asneeded. Based on results obtained, in the light of the statutory requirements for selecting thesite for a final repository, confirmation that the chosen host rock has favourable geologicalcharacteristics from the viewpoint of waste confinement is is obtained. The SDMs, including
the conceptual model of long-term evolution, will be refined to build confidence that the
3-26

favourable geo-environmental properties of host rock are maintained for a sufficiently longtime. From associated assessment of operational and post-closure safety, the selected layoutof the repository and the location of surface facilities are confirmed as meeting requirements.
It is assumed that areas with significant risks of future natural perturbations have been
avoided based on PI studies. During investigations conducted in the UIF, however, there may
indications of perturbations that need to be reconsidered. When these are encountered, theprobability of them having significant impacts will be examined, taking into account thelimitations of applicable investigation and evaluation technologies.
(i) Assessing the potential of significant effects of natural perturbations(a) Volcanic and igneous activity
When encountering intrusive rocks such as dikes or large-scale alteration zones, theirformation age is estimated by dating constituent minerals, which complements detailed
geological assessments of their occurrence. The rock temperature gradient/spatial distribution,and dissolved gas in groundwater are investigated in detail to understand the relationshipbetween intrusive rocks, alteration zones and associated magmatic activity.
(b) Migration and inflow of hydrothermal fluids
When groundwater with high temperature, low pH or high carbonate concentration isobserved, its origin and source contributions from volcanic hydrothermal or deep-seatedfluids are identified based on chemical and isotopic data (e.g. Li/Cl, 3He/4He, etc.). Based on
information on geothermal conditions, resistivity, microearthquake occurrence, etc., thepotential extent of groundwater migration and associated inflow routes to locations where therepository will be constructed is investigated.
(c) Fault activity
When encountering large-scale faults or fault zones, the probability of these extending
from the surface to relevant depths is evaluated, taking into account the geologicalinformation from surface investigations. Radiometric dating of minerals in the fault zone, as
well as mineralogical and chemical/isotopic analyses, are conducted to determine if thesewere formed as a result of repeated activity in the Quaternary and the probability that a fault
might be reactivated.
(ii) Evaluation of geological environment characteristics and their long-termchanges
Based on investigations from both phases, more detailed information on the representative
host rock setting is obtained. In addition to detailed characterisation within the UIF itself, insitu tests using shafts and tunnels excavated to construct it are conducted. The spatial
distribution of geological structures (faults, fractures, lithology, etc.),
thermal/mechanical/hydraulic properties of all relevant rocks, nuclide migration/retardation
properties of water-conducting features (fractures, water-conducting microstructures, etc.),geochemical properties of groundwater, as well as changes in the geological environment due
to this construction, are investigated in detail. From a synthesis of this information, areas with
3-27

favourable geological characteristics from the viewpoints of repository construction,
operation and assured post-closure performance are identified. Extending the PI knowledge
base, information such as paleohydrogeological evidence from groundwater/rocks andminerals to gain understanding of long-term site evolution is obtained.
3.2.3 Site characterisation technology
Stepwise site-specific geological investigations are conducted according to the procedures
described in Section 3.2.2, based on the best technology available. In addition to assuring
efficiency and cost-effectiveness of site investigation, for each study item listed in Table 3.2-1,
it is necessary that these technologies enable information to be acquired with the required
level of quality.
Based on these requirements, NUMO has summarised the status of relevant technology
development, testing and improvement by relevant research organisations, with a focus onneeds for PI and DI investigations (e.g. Ota et al. 2007 [74]; Saegusa et al. 2007 [92]; JAEA,
2010 [115]; 2013 [105]; 2015 [116]; 2018 [117]; CRIEPI, 2013 [118]; 2013 [119]; 2013[120]; 2018 [121]; 2018 [122]; AIST, 2013 [67]; 2013 [123]; 2016 [124];
AIST/JAE/RWMC/CRIEPI, 2016 [68]; 2017 [69]; 2018 [125]). In particular, from theviewpoint of accurate characterisation of relevant environments, the objectives, methods,
application examples, effectiveness and technical issues for individual technologies have beensummarised in Supporting Report 3-11. In the future, NUMO will work with, and coordinate
roles of, relevant national research institutions in order to systematically develop required
technologies prior to the start of each siting stage (described in detail in Section 3.5.2). In
parallel, as described in Section 3.2.2 (3), based on experience in domestic and overseas
URLs, the applicability of associated technologies for the UIF will be assessed and refined as
required, to resolve any technical issues identified as site investigation progresses.
In the next section, development of relevant research and evaluation technologies sincethe H12 report is overviewed. In addition, the probabilistic evaluation of the occurrence andconsequences of future natural perturbations on a timescale exceeding 100 ky is described.
(1) Development of investigation and evaluation technologies
Since the H12 report, and particularly since 2004, R&D roadmaps and five-year overallplans have been formulated through coordination meetings with the Agency for Natural
Resources and Energy (ANRE) and relevant research institutions [126] [127] [128] [129][130]. The following sections overview the current status of both technology to better assess
the potential impacts of natural perturbations and that for characterising geologicalenvironments together with their long-term evolution.
(i) Technology related to the impacts of natural perturbations
Investigation and evaluation technologies for the PI stage, aimed at understanding pastand future occurrence of natural perturbations and their associated effects, especially forinland applications, are being further developed and assessed via case studies [115] [116][131].
Specifically, technologies are being developed for determination of:
3-28



The existence and spatial distribution of deep underground faults and magma
chambers, using a combination of geophysical and geochemical methods [16] [132][133].



Active faults that are difficult to identify at the surface using airborne laser
measurements (that dramatically improve detection under dense vegetation) and use ofgeochemical indicators to confirm the presence of deep-seated fluids [133] [134][135].

The activity of faults based on mineral shapes and radiometric age for infill within
fault zones [136] [137].

Uplift and erosion rates using river terraces and thermochronological methods [138][139] [140] [141].

Extending from coastal land to offshore areas, AIST, JAEA and CRIEPI are developing
technologies for assessing uplift and erosion [68] [69] [125], improving investigation
techniques to confirm the presence of magma and associated fluids, extrapolating land upliftand erosion patterns to offshore areas and improving terrace evaluation and chronology
methods using empirical indicators [142].
For quantifying the probability of future natural perturbations and associated changes ingeological environment characteristics, the following topics are being promoted:
Development of long-term prediction models for occurrence of new volcanoes andfaults [143] [144].



Evaluation methodology for changes in the groundwater flow systems due to fault
movements [145].

Simulation methodology for topographic changes up to about 100 ky in the future[140].



Groundwater flow analysis considering topographic changes and climate perturbations[146] [147].

For time periods exceeding 1 My, a model capable of simulating the long-term changes of
the geological environment in three dimensions has been constructed. In addition to thevisualisation/numerical techniques and uncertainty evaluation methods required, developmentof fundamental technologies for establishing the geological database for model construction
and validation is also in progress [117].
Furthermore, utilising the latest analytical methods for trace elements and isotopes, JAEA
and CRIEPI have been pursuing improvements in the technology for accurately ascertaining
past occurrence times of volcanic/igneous activity and fault movements, the rate of uplift anderosion, the age of formation of strata/rock bodies/minerals and the residence time ofgroundwater [75] [120] [125] [133] [148] [149] [150] [151] [152] [153] [154] [155]. This isessential to support arguments for the long-term stability of relevant geological environments.
(ii) Investigation and evaluation methodology for geological characteristics andtheir long-term evolution
JAEA's work at the Horonobe and Mizunami URLs forms the core of this R&D
programme. In these facilities, initial investigations from the ground surface followed bystudies at the time of their excavation, corresponds to site characterisation PI and DI stages.
3-29

As described in Section 3.2.1, this resulted in uncertainty reduction in an iterative manner andcoordination of engineering design and safety assessment through integration of geologicalenvironment information. This work established the effectiveness of fundamental
technologies for staged site investigations of both Neogene sedimentary rocks and granitic
rocks (which represent plutonic rocks) for near-coast areas [61] [74] [85] [87] [92] [105].
Essential knowhow and the basis for judgments related to site assessment have been compiled
to serve as a knowledge base for investigation/evaluation technologies, as well as improving
their practicality and reliability [105] [115] [116].
In parallel with such work, a R&D programme led by AIST and CRIEPI is being
conducted for coastal areas for which relatively little site investigation experience is available.
This includes the following topics:Interpretation and modelling of geological structures based on literature information[123].



Geophysical exploration from land to sea and assessment of the spatial distribution of
freshwater/saltwater mixing areas [67] [124].



Exploration of groundwater seepage into the sea floor [156].



Controlled directional drilling and borehole investigations [118] [122].



Examination of technologies for drilling offshore [123] [124].

Geophysical technology to continuously profile coastal land to sea is recognised as an
issue and improvements are under development, such as shown in Figure 3.2-5. As noted in
Section 3.1.3, improved understanding of the origin, age, and flow of coastal groundwater has
been obtained around the Horonobe URL [67] [68] [69] [75]. With the aim of development ofcomprehensive sub-seabed characterisation technologies for coastal areas, R&D is ongoing on[68] [69] [123] [124]:
Geological structure.



Determination of regional groundwater flow.

Relationship between faults and groundwater flow.



Information on long-term changes in groundwater flow that can be obtained byborehole investigations.

3-30

Figure 3.2-5 Modelled distribution of subsurface resistivity based on a sea-land
electromagnetic survey in the Horonobe area, Hokkaido(Modified from AIST, 2013 [67])For sedimentary rocks that make up the accretionary prisms widely distributed in coastal
areas, and for Neogene sedimentary rocks, NUMO working with CRIEPI to confirm theeffectiveness of systematic investigation/evaluation technologies developed by research
institutions (for example, drilling technologies and borehole investigations), so that these will
be applicable in the PI stage [109] [157] [158]. Through this initiative, it is aimed toaccumulate the technical knowledge and practical experience to implement required
investigations efficiently and effectively.
Based on the current state of technology development for coastal areas, a study group fortechnical issues related to geological disposal beneath the coastal seafloor pointed out that,
although it is necessary in the future to advance specific technologies through verification of
their applicability, it is possible to conduct stepwise geological investigations in coastal areas
by combining relevant technologies for land areas (inland and coastal areas) [159].
For the technologies required in the latter half of the DI stage, based on experience in JAEA's
URLs, the following may be considered:

Evaluation of the function of faults encountered in tunnels as water-conducting
features (e.g. relationship to repeated fault movement and formation of infill minerals)[160].

Investigation and evaluation of water-conducting microstructures (e.g. the geometry
and connectivity of the pores that impact the migration and retardation ofradionuclides) [121].



Demonstration of techniques for assessing changes in the groundwater flow field andwater chemistry due to tunnel excavation and damage to rock around the tunnel [161].



Development of technology to examine the long-term behaviour of rock around thetunnel after backfilling.

Long-term monitoring of groundwater pressure and chemistry [162].

3-31

The current status of the technologies noted above, particularly considering PI
implementation, is summarised according to the topics previously listed in Table 3.2-1 andpresented in Table 3.2-2 (those related to the effects of natural perturbations) and Table 3.2-3(those related to the geological environment characteristics and their long-term evolution).
The aim will be to resolve most of the issues described in these tables during the PI stage (seeSection 3.5.2). Details of the investigation and evaluation technologies are provided in
Supporting Reports 3-6 and 3-9, with Supporting Report 3-11 comprehensively summarisingdetails of the individual methodologies.

3-32

Table 3.2-2 Current status of investigation and evaluation technologies related to the effects of natural phenomena (1/2)

Movement and inflow of
volcanic hydrothermal anddeep-seated fluids

Volcanic and igneous activity

Site investigation/evaluation
techniquesProbability and extent ofmagma migration from
Quaternary volcanoesProbability and extent of
new volcanic (magmasupplying) occurrences

Extent and scale of future
thermal and chemical
impacts of hydrothermal
deep-seated fluids







Current status
Proven technology
Available characterisation techniques and their applications
Geological characterisation of magma activity, and history ofdevelopment
Geophysical and geochemical characterisations of deep
underground structure of volcanic bodies
Confirmation of the existence of magma chambers for inland areasusing geophysical survey
Geological characterisation of the temporal and spatial evolutionof volcanic activity
Geophysical and geochemical characterisation, including seismic
velocity and thermal structures underground
Model of the uneven distribution of high temperature areas in thecrust and mantle






Issues
Methodology to be developed and refined
Knowledge to be developed and refined
Methods for confirmation of the existence of coastal sea area magma
chambers
Magma activity ranges according to activity types, such as
stratovolcano and caldera



Confidence of model related to time changes and continuity of hightemperature areas in the crust and mantle
Characterisation of effects of T-H-C conditions of volcanichydrothermal fluids (The technologies developed in the field ofgeothermal resources can be applied)Geophysical characterisation of the existence of deep-seated fluids
and related geological structures for inland areasDistribution of deep-seated fluids, along with chemical/isotopic
composition, determination methodology and relationships tofaults/tectonic structures
Method for confirmation of the existence of deep-seated fluids forcoastal sea areas
Comprehensive technologies for characterisation of distribution andflow paths of deep-seated fluid, together with potential impacts

3-33Table 3.2-2 Current status of investigation and evaluation technologies related to the effects of natural phenomena (2/2)
Site investigation/evaluation
techniques

Earthquake and fault activityDistribution of active faults
and their present and future
activity

Extent of hydraulic andmechanical effects associatedwith fault movements
Changes in the geologicalenvironment associated withearthquake activity

Uplift and erosionRate of future uplift andsubsidence
Future erosion rate andlandform change

Methodology for assessing the
likelihood of natural perturbing
phenomena* occurring over a long
period of time, e.g. >100 ky in the
futureMethods such as topographic interpretation, geodetic observation,
surface geological survey, trench survey, geophysical survey,
borehole investigation for inland areas
Detection of active faults after large earthquakes occurring in areas
where active faults are unconfirmed on the surface, using a
combination of the above techniques
Methods such as seabed topography surveys and seismic surveys
for coastal sea areas
Model experiments and numerical analyses on the branching andextension of active faults and on the reactivation of faults

Width, distribution and properties of fault zones and process zones,
their effects on hydrology and mass transport, and changes withtime

Acquisition of the knowledge to support claim that water pressurechanges associated with large earthquake are small and temporary
Characterisation of water pressure changes associated with large
earthquakes using volumetric strain analysis of the crustClimate and sea-level change

Current status
Proven technology
Available characterisation techniques and their applicationsComprehensive technologies for characterisation of rate of upliftand erosion








Issues
Methodology to be developed and refined
Knowledge to be developed and refined
Improvement of resolution of methods such as seabed topography
surveys and seismic surveys for coastal sea areas
Comprehensive technologies for characterisation of active faults
distributed around shorelines, based on the developed investigation
methods for inland areas and coastal sea areas
Characterisation of distribution and activity of topographically obscureactive faults based on geochemical methods, and accumulation of thecaseCharacterisation of the activity of faults in the absence of overlying
strata, based on mineralogical investigation and dating techniques formaterials in fault zones, and accumulation of the caseCharacterisation of the extent of the influence of active faults, andaccumulation of the caseOutflow of hot spring associated with the Tohoku earthquake andrecovery of hydraulic conductivity of faults associated with earthquakes

Characterisation methods for inland areas where effective indicators of
uplift and erosion have not been developed
Characterisation methods for coastal sea areasComprehensive technologies for characterisation of global climate
and sea-level change
Individual technologies for assessing regional climate change
Evaluation of the probability of future perturbations byextrapolation based on the assumption of uniform continuity ofcrustal movement
Regional characteristics of future perturbations in the scientifically
predictable periodProbability of future natural perturbations
Regional characteristics and temporal/spatial changes in theoccurrence/variation of such perturbations
Methods of classification of perturbations for development of safetyassessment scenarios, and assessment of associated probability anduncertainties* Encompasses volcanism and igneous activity, the movement and inflow of volcanic hydrothermal and deep-seated fluids, earthquake and fault activity, and uplift anderosion.

3-34

Table 3.2-3 Current status of investigation and evaluation technologies for geological environment characteristics and their long-term changes(1/3)

Characteristics to be understoodGeological structureSpatial distribution and geometryof groundwater flow paths
Spatial distribution and geometryof geological strata and rockbodiesSpatial distribution of lithologies
and fractures within geologicalstrata and rock bodies







Geometry of water-conducting
microstructures that contribute tonuclide migration and retardation

ThermalenvironmentSpatial distribution of geothermal
gradient
Thermal properties of strata androck bodies







Current status
Proven technology
Available characterisation techniques and their applications
Comprehensive method combining geophysical surveys, borehole
investigations and tunnel mapping for inland areas
3D geophysical survey for the characterisation of spatial
distribution of faults (including technology developed in the fieldof resource exploration)
Geophysical survey methods for faults distributed near the
shoreline
Geological characterisation for coastal sea areas (including
technology developed in the field of resource exploration andacademic research)
Characterisation of the connectivity of faults and waterconducting fractures at the repository scale
Borehole drilling and investigations
Detection of microstructures
Development of geological models at various spatial scales











Issues
Methodology to be developed and refined
Knowledge to be developed and refined

Available information on 3D geophysical surveys
Improvement of data analysis methods for nearshore geophysicalsurveys
Test applicability of site investigation technology for coastal sea
areas
Technology for comprehensive characterisation of theconnectivity of faults and water-conducting fractures.
Borehole drilling and investigation technologies for rock masses
including weaker layers
Improvement in accuracy of microstructure detection in
boreholes by mud drilling and boreholes drilled from tunnels

Characterisation of the geometry of water-conducting
microstructures using in-situ tracer tests and laboratory tests
Knowledge on conceptual modelling of water-conducting 
microstructures for Neogene sediments and granite rock

Methods for estimating the aperture of water-conducting
microstructures using high-viscous fluids and radon
concentrations
Information for conceptual modelling of water-conducting
microstructures in rocks other than Neogene sediments andgraniteCharacterisation of geothermal gradient and thermal properties of 
rock mass based on borehole investigations and laboratory tests
Borehole investigation methods

Borehole investigation technology for rock masses including
weaker layers

3-35

Table 3.2-3 Current status of investigation and evaluation technologies for geological environment characteristics and their long-term changes(2/3)

Characteristics to be understoodHydrogeological conditionSpatial distribution ofhydrogeological structures
Spatial distribution of advection
and diffusion domains
Hydrogeological and gas
migration characteristics of strata,rock bodies and hydrogeologicalstructures
Spatial distribution of hydraulichead and temperature










Rock mechanicalcondition

Nuclide migration and retardation
characteristics of waterconducting microstructures
Petrophysical and rockmechanical characteristics ofstrata, rock bodies, faults andfractures

Current status
Proven technology
Available characterisation techniques and their applicationsHydraulic properties and hydraulic head distribution using
borehole investigations in inland areas
Systematic investigation methods using boreholes
Hydrogeological characterisation for coastal sea areas based on
resource exploration and academic investigations
Hydrogeological modelling and groundwater flow simulation atrelevant spatial scales
Hydrogeological characterisation of water-conducting featureswith high transmissivity (> 10-8 m2/s)
Detection and hydrogeological characterisation of waterconducting features
Characterisation of gas migration in sediments and the amount ofgas inflow into tunnels

Characterisation of nuclide migration, retardation processes andsorption/diffusion using in-situ tracer tests and laboratory tests
Knowledge to support conceptual modelling of nuclide migration
and retardation characteristics for Neogene sediments and graniticrockComprehensive methodology combining geophysical surveys,
borehole investigations from both the surface and tunnels andtunnel mapping
Systematic investigation methodology using boreholes
Characterisation of the physical and mechanical properties of the
excavation damaged zone, and associated spatial distribution
Rock mechanical modelling methodology at relevant spatial
scales

3-36











Issues
Methodology to be developed and refined
Knowledge to be developed and refined
Borehole investigation technology for rock masses, including
weaker layers
Test applicability of hydrogeological investigations technologies
using boreholes for coastal sea areas
Knowledge on the consistent interpretation of the results ofgroundwater flow simulation, the age of groundwater and the
distribution of groundwater chemistry
Knowledge on the existence and formation conditions of stagnant
groundwater
Methodology for hydrogeological characterisation of waterconducting features with low transmissivity (< 10-8 m2/s)
Improvement of borehole hydrogeological investigations using
mud drilling and boreholes drilled from tunnels
Knowledge on the characterisation of diffusion domain using
natural stable isotopes
Knowledge to support conceptual modelling of water-conducting
microstructures in rocks other than Neogene sediments andgranitic rock

Borehole investigation technologies for rock masses, including
weaker layers

Table 3.2-3 Current status of investigation and evaluation technologies for geological environment characteristics and their long-term changes(3/3)
Characteristics to be understoodLong-term evolution of geologicaland environmental characteristics

Hydrochemical conditionSpatial distribution ofgroundwater chemistry, isotope
ratiosSpatial distribution ofgroundwater pH and Eh values
Properties and distribution ofgases in strata, rock bodies andgroundwater

Chemical effects of colloids,organic matter andmicroorganisms
Nuclide migration and retardation
characteristics of waterconducting microstructures
Evolution of the regional stressregimeDevelopment and formation ofgeological structures
Evolution of advection domains
and their hydrogeologicalcharacteristics
Formation mechanisms ofgroundwater chemistry andassociated evolution
Evolution of the petrophysical,rock mechanical and thermal
characteristics of strata and rockbodies
















Current status
Proven technology
Available characterisation techniques and their applications
Comprehensive methodology for understanding groundwaterchemistry, isotope ratios, pH and Eh in groundwater using
borehole investigations for inland areas
Understanding of spatial distribution of high-salinity groundwater
for inland areas, using electrical survey methods
Systematic investigation methods using boreholes
Hydrochemical characterisation for coastal areas, based on
resource exploration and academic investigations
Methodology for hydrochemical modelling and analysis atrelevant spatial scales
Characterisation of properties and distribution of gases in strata,rock bodies and groundwater
Methodology for sampling and analysis of colloids, organic
matter and microorganisms
Characterisation of nuclide migration, retardation processes andsorption/diffusion, using in-situ and laboratory tracer tests
Knowledge to support conceptual modelling of nuclide migration
and retardation characteristics for Neogene sediments and granitic
rockCharacterisation of regional stress field evolution, crustal
movements, evolution processes for geological structures andtopographic change
Methodology for modelling and analysis of long-term
topographic evolution and geological structures under simplified
conditions
Methodology for modelling and analysis of long-term evolutionof groundwater flow conditions
Groundwater dating methodology using multiple isotopes forinland areas
Analysis and evaluation methods for long-term evolution of thephysical and mechanical properties of rock masses due to theevolution of the regional stress field and geological structures

3-37









Issues
Methodology to be developed and refined
Knowledge to be developed and refined

Methodology for analysis of electrical survey data for coastal sea
areas
Borehole investigation techniques in rock masses including
weaker layers
Test applicability of hydrochemical investigation technologies
using boreholes for coastal sea areas
Test applicability of hydrochemical characterisation techniques
for site characterisation of coastal sea areasKnowledge to support conceptual modelling of waterconducting microstructures for rocks other than Neogene
sediments and granitic rock.



Methodology for modelling long-term evolution of large scale
heterogeneities in both topography and the geologicalenvironment with high resolution
Knowledge to support modelling of long-term evolution ofgroundwater flow conditions
Knowledge to allow consistent interpretation of the results ofgroundwater flow simulation, groundwater ages andgroundwater chemistry distribution
Knowledge supporting groundwater dating using multiple
isotopes for coastal sea areas
Data to test models of long-term evolution of petrophysical androck mechanical characteristics






(2) Evaluation of the probabilities and impacts of natural perturbation
phenomena
As described in Section 3.2.1, it is essential that the impacts of perturbations which cansignificantly affect the future performance of the geological barrier are avoided to the extent
possible. However, uncertainties accompany the assessment of future occurrences due tolimitations in scientific knowledge. Such uncertainty generally increases as the evaluation
period becomes longer and its treatment is important in assessing potential impacts on
repository safety functions.
Methods for evaluating the occurrence and impacts of such phenomena are roughlyclassified as: extrapolation, analogy, numerical analysis using a phenomenological model andprobability theory [163]. Extrapolations from the geological knowledge base to assess
potential future perturbations may be purely empirical, based on statistical analyses, or
supported by more fundamental mechanistic models (e.g. [163] [164]), but all options involve
considerable uncertainties. This is particularly so when the driving forces resulting fromtectonic plate movements significantly diverge from current conditions (needs to beconsidered at times > ≈ 100 ky).
NUMO considers that stochastic methods are particularly useful for managing such
uncertainties, in line with international recommendations for similar assessments, e.g. the riskof volcanic disturbances at nuclear sites [165]. For assessment of seismic risks to US nuclear
power plants and volcanic risks to the Yucca Mountain Project, several different
phenomenological models and parameter combinations have been developed and aprobabilistic evaluation method is applied by weighting the likelihood of occurrence using
expert elicitation [166].
Based on this background, NUMO developed the ITM (International Tectonics Meeting)–TOPAZ (Tectonics of Potential Assessment Zone) method for region-/site-specific risk
assessment [167] [168]. To develop models of potential changes in the regional tectonics
based on plate movements, the current tectonic situation in the vicinity of the site is assessedin the context of the large-scale geological environment. This method involves (1) creating
scenarios and logic trees, (2) determining confidence by soliciting expert opinion in a
structured manner, and (3) evaluating the likelihood of scenario occurrence. Figure 3.2-6
shows the basic structure of the resulting logic tree.

Figure 3.2-6 Information flow of the ITM-TOPAZ method and basic structure of the logic tree
(Modified from Goto et al., 2014 [173])

3-38

So far, the method has been applied to volcanic/igneous activity, seismic/fault activity,and uplift/erosion, with a resolution of about 5 km × 5 km (corresponding to the repository
scale) and for times up to about 1 My [169] [170] [171] [172]. After summarising thescientific knowledge necessary to evaluate the effects of such phenomena on post-closure
safety, the results of the risk assessment for future volcanic and igneous activity as describedin Section 3.4.1 are used for dose evaluation for the "new volcano occurrence case" in Section
6.4.3 (1).

3.3 Development of SDMs for representative host rock settings
3.3.1 Model aims and development procedure
In the staged siting process, as discussed in detail in Section 3.2.1, information obtained
through investigations at the volunteer sites is synthesised into site descriptive models(SDMs), following a consistent synthesis across the many disciplines involved. The
engineering practicality of repository implementation and its safety assessment based on the
SDMs are then performed, with these steps repeated iteratively as siting proceeds. As no sites
had come forward at the time of preparation of this report, “representative host rocks” are thus
illustrated, based on the current geological knowledge base for Japan. These focus on
geological formations in environments which may be eligible for site selection, having
required safety functions that are expected to persist for sufficiently long time periods. A
series of studies to illustrate repository design and safety assessment tailored to these settings(for details, see Chapters 4, 5 and 6) are then be conducted.
Selection and modelling of the representative host rocks proceeds according to the work
flow shown as (A) to (F) in Figure 3.3-1.

Figure 3.3-1 Stepwise procedure for developing SDMs for representative host rocks(Modified from Ota et al. 2015 [9]; Copyright (11 August, 2016) by the American Nuclear
Society, La Grange Park, IL, USA)In particular for (A), after selecting representative host rocks based on widespread
distribution in Japan (Section 3.3.2), a conceptual SDM is developed for each rock type

3-39

(Section 3.3.3 (2)). Based on this conceptualisation, the (B) geological model and (C)hydrogeological model are developed (see Sections 3.3.3 (3) and (4)). In addition, based on
the hydrogeological model, a groundwater flow analysis is performed (D) to determine thespatial scale of the area to be studied in more detail (see Sections 3.3.3 (3) and (4)). Taking
into account that the quality and quantity of the geoscientific information obtained in each
investigation stage can vary, the regional scale model mainly uses the nationwide database. Inmodelling from the repository scale to the panel scale, the geoscientific information forparticular sites where the deep geological and hydraulic characteristics of the rock mass havebeen comprehensively studied is used. The SDMs are developed so that the regional,
repository, and panel scale representations form a nested structure.
For the selected host rocks, based on the results of case studies in the particular sites, (E) aconceptual model of water-conducting microstructure to allow quantification of the migration
and retardation of solutes is developed (see Section 3.3.3 (5)). Additionally, based onnationwide dataset, (6) hydrochemical and rock thermal/mechanical models are developed(see Sections 3.3.3 (6) and (7)). The SDMs constructed for representative host rocks are thus
clearly more realistic than those used for previous safety assessments in Japan [9].
3.3.2 Classification of representative host rocks
From the viewpoint of geological disposal, seven rock groups have been identified(Section 3.1.3 (1)) and mapped (Figure 3.1-1) [26].
Quaternary sedimentary and volcanic rocks are widely distributed at the surface, but when
unconsolidated below 300 m these are excluded from consideration [2]. Further, according toYasue et al. (2014) [174], the distribution of these two rocks at reference depths of 500 m and1,000 m, is greatly reduced compared to their distribution at the surface (approximately 6%
and 1%, excluding areas within 15 km from the centre of Quaternary volcanoes, within which
80% of Quaternary volcanic rocks are found) [110]. Although eligibility as a host rock for therepository is determined through the staged site investigation process, compared to the otherfive rock types it is reasonable to assume that these are relatively unlikely to be representative
host rocks. In this report, therefore, Quaternary sedimentary and volcanic rocks are excluded
from further consideration.
For each of the other five rock types, the geoscientific information compiled in the H12report [7] has been updated (e.g. NUMO, 2013 [8]) with a focus on characteristics importantfrom the viewpoint of repository design and safety assessment. An overview is presented inTable 3.3-1, including thermal conductivity, uniaxial compressive strength and effective
porosity, which are important from the viewpoint of repository design, as well as structures ofwater-conducting features, hydraulic conductivity and effective porosity, which are importantfrom the viewpoint of safety assessment.
Excluding areas within 15 km from the centre of Quaternary volcanoes and the areas of
Quaternary sedimentary and volcanic rocks, the abundance of each rock type has been
calculated using the 1:200,000 Japan geological map [65], while abundance at 500 m and1,000 m below ground level has been calculated based on Yasue et al. (2014) [174]. The
geoscientific information collected and analysed for these rock types and the statisticallyprocessed data are summarised in Supporting Report 3-13.
For Neogene sedimentary rocks and plutonic rocks (mainly granites), typical of
sedimentary and crystalline systems in Japan, various geological environment investigations
have been carried out as part of studies for geological disposal safety assessments [7]. Based3-40

on these, both Neogene sediments and plutonic rocks are considered to have generally
favourable characteristics from the viewpoint of waste confinement. The latest geoscientificfindings obtained in JAEA's URLs projects [74] [92] [115] [116] are considered to support
this [8].
Table 3.3-1 Characteristics of the five rock types from the viewpoint of geological disposal

Sedimentary
41
40

PreQuaternary
Volcanic
18
15

PreQuaternary
Plutonic
18
20

Metamorphic
8
10

10

45

10

25

10

Pore, Fracture

Fracture,
Beddingcleavage

Fracture

Fracture,
Dyke

Fracture,
Schistosity

2.9 × 10-7

4.7 × 10-7

2.1 × 10-7

5.5 × 10-8

4.3 × 10-8

25-27

3.5-6.8

5.4-7.9

0.8-1.5

1.2-6.8

1.6-1.8

1.4-1.5

2.4-2.5

2.8-2.9

3.3

9-28

74-90

92-106108-110

55-66

Age

Neogene

Pre-Neogene

Rock typeSurface Abundance*1 (%)
500 m depth abundance*1 (%)
1,000 m depth abundance*1(%)

Sedimentary
15
15Structure of water conducting
features
Logarithmic mean of bulk
hydraulic conductivity*2 (m/s)
Average and median ofeffective porosity*3 (%)
Average and median ofthermal conductivity*3 (W/mK)
Average and median ofuniaxial compressive
strength*3 (MPa)*1

For geological environment excluding the area within 15 km from the centre of Quaternary volcanoes.
Logarithmic mean as representative value*3
Average and median as representative values, giving an indication of the distribution of data
*2

Pre-Neogene sedimentary rocks are important basement rocks and are widely distributedin Japan [175], estimated to occupy almost half of the area at around 1,000 m depth, as shownin Table 3.3-1. These occur as aggregates of blocks of different sizes, separated by faults.
Structures are extremely variable, e.g. not primarily stratified, including allochthonous or
autochthonous rock blocks composed of various lithological facies, or stratified but lacking
continuity. This rock is generally highly consolidated and hard, thus with water-conducting
features and mechanical strengths similar to plutonic rocks and quite different to Neogene
sedimentary rocks. Therefore, Pre-Neogene sedimentary rocks are also considered in thisreport as a representative host rock.
In addition to the features analysed for design and safety assessment of the repository,
consideration of a wider spectrum of characteristics (e.g. mean values and ranges given in
Supporting Report 3-13), shows that Neogene and Pre-Neogene volcanic rocks havecharacteristics similar to plutonic rocks from the viewpoint of repository design, and similar
to Pre-Neogene sedimentary rocks from the viewpoint of safety assessment. From the
1:200,000 Japan geological map [65], about 90% of the metamorphic rocks distributed at thesurface are low grade crystalline schist (e.g. phyllite) and high grade gneiss (e.g. amphibolite).
There is a clear difference between the characteristics of crystalline schist and gneiss, theformer with characteristics similar to Pre-Neogene sedimentary rocks and the latter similar toplutonic rocks (see Supporting Report 3-13 for details). For Neogene/Pre-Neogene volcanicand metamorphic rocks, therefore, design and safety of a repository can be assessed by
3-41

applying the concepts and methods applied to Pre-Neogene sedimentary rocks and plutonic
rocks.
Thus, in this report, plutonic rocks, Neogene sedimentary rocks (hereinafter Neogene
sediments), and Pre-Neogene sedimentary rocks (hereinafter Pre-Neogene sediments) areconsidered as representative host rocks. By developing concepts and methods for the designand safety assessment of the repository for these three rock types, the key characteristics ofthe five representative host rocks (including Neogene/Pre-Neogene volcanic and metamorphic
rocks) can be covered.
3.3.3 Development of a representative SDM(1) Basic concept
The repository scale area is about several km × several km and should have favourable
geological environment characteristics and long-term stability. It includes the repository hostrock together with surrounding formations and associated structures. The larger regional scale
area (several tens of km × several tens of km around the repository footprint) capturesfeatures that may impact perturbation phenomena and includes regional groundwater flow
system from recharge to discharge, as well as hydraulic boundaries such as groundwater
divides. A smaller panel scale area is around several hundred m × several hundred m within
the repository scale area, encompassing a single waste disposal panel. In this area, thecharacteristics that affect the design of the engineered barriers and assessment of any
radionuclide migration from them are evaluated in detail.
For the regional scale SDM, it is necessary to consider the topography and sea-land
distribution, in addition to the geological environment extending from deep underground (≈ 2km) to the surface. As discussed in Section 3.2.1, it is also necessary to conceptualise impactsof evolution of topography and geological structures caused by regional uplift and erosion,
sea-level change, etc. Based on established exclusion criteria [2] to assure safety of both
surface and underground facilities, these representative SDMs are developed that do notinclude any characteristics that would lead to their rejection during siting.
(2) Development of conceptual SDMs(i) Plutonic rocks
More than 90% of plutonic rocks distributed at the surface are granites and these havebeen extensively studied at depth in JAEA's URLs at Mizunami and Kamaishi [92] [115][116] [176], and a number of international URLs (e.g. Stripa and Äspö in Sweden andGrimsel in Switzerland). Therefore, granites are chosen as representative plutonic rocks.
The regional scale area was selected to be 50 km × 50 km (× 3 km depth), considering thatgranite bodies of this scale are widely distributed [174] and that the basin area of a large-scaleriver (considered to be the largest area within a groundwater hydraulic boundary) averages
about 2,200 km2. For the repository scale area, an area of 5 km × 5 km (× 1.5 km depth) was
selected within the regional scale area, considering the scale and distribution of faults, thegroundwater flow flux and the travel time from the repository to the area boundary. For this,the heterogeneity of the geological and hydrogeological characteristics of the rock mass
surrounding the repository was modelled in more detail. For the panel scale area, based on an
approach similar to that for the repository scale area, an area of 800 m × 800 m (x 100 m

3-42thickness) was specified in even greater detail in order to support design and safetyassessment of the repository.
The conceptual SDM for plutonic rocks, covers an area that includes regional
groundwater flow system from recharge to discharge areas and defines hydraulic boundaries,
as shown in Figure 3.3-2. It includes basement granite, an upper highly fracture domain, (orweathered zone) with relatively high hydraulic conductivity (up to about 200 m in thickness),
and a younger sedimentary overburden (about 100 to 200 m thick).

Figure 3.3-2 Conceptual SDM for plutonic rocks

Large scale faults in the granitic basement (discussed further in (3) (i) (a) below),
generally include fault gouge with relatively low hydraulic conductivity surrounded by fault
breccia and damage zones with a relatively high fracture density and high hydraulicconductivity. Anisotropy is often observed in their structural and hydraulic characteristics[47] [83] [92] [176] [177] [178] [179]. These features are found in all rocks in Japan, although
their detailed characteristics will vary for different geological settings [177][180] [181] [182].
Deep groundwater has a downward flow in recharge areas, with upward flow associatedwith structural discontinuities and in freshwater-saltwater mixing areas/discharge areas near
the coastline. For a regional groundwater flow system, groundwater flow at depth in central
regions is predominantly horizontal, without being greatly affected by the surface topography[183]. Such horizontal groundwater flow may be affected by the spatial distribution of faults
with anisotropic hydraulic conductivity.
The dataset used for plutonic rock is presented in Supporting Report 3-14, withinformation for establishing modelling areas and constructing associated SDMs presented in
Supporting Report 3-15.
(ii) Neogene sediments
The regional scale area was selected to be 30 km × 30 km (× 2 km depth) taking into
account distribution at the ground surface in the 1:200,000 Japan Geological Map [65], as
well as the maximum thickness obtained from the 1:50,000 geological map and existing
literature for 57 sedimentary basins where Neogene sediments are widely distributed. The
repository scale area was selected to be 5 km × 5 km (× 1 km depth) considering the layout of
the repository and spatial distribution of geological structures and groundwater flux/travel
times. The panel scale area within the repository scale area was selected to be 800 m × 800 m(× 100 m thickness) similar to the plutonic rock case.

3-43

When constructing a SDM for Neogene sediments, again an area that includes
groundwater flow system from recharge to discharge as well as hydraulic boundaries isrepresented, as shown in Figure 3.3-3. In this SDM, Neogene sediments overlie basement
granite and older sediments and are covered by Quaternary sediments, which include a rangeof horizontal, vertical and fold structural elements.
In the Neogene sediments, large scale faults are represented and considered to be similar
to those for plutonic rocks [74], as are the constraints on deep groundwater flow [183].

Figure 3.3-3 Conceptual SDMs for different Neogene sediment layers (shown in terracotta, lightbrown, and green). The basement (dark brown) is here assumed to consist of Paleogene
sedimentary rocks

Supporting Report 3-16 presents the dataset used to construct the SDMs for Neogene
sediments, and Supporting Report 3-17 presents basic information on the setting of modellingareas and construction of a conceptual SDM.
(iii) Pre-Neogene sediments
As most Pre-Neogene sediments exposed at the surface [67] correspond to accretionary
complexes, these were the focus for SDM development. Considering the general features ofaccretionary complexes, including the development of thrusts and rupture/mixing of strata[184] [185], both simple coherent facies and more complex mélange facies were modelled.
A focus was placed on the literature information on the Ashio, Mino, Tamba, Chichibu,
and Shimanto Belts [186] [187] [188] [189] [190], estimated to represent about 80% of thearea of all accretionary bodies at the ground surface according to the 1:200,000 Japan
Geological Map [65]. The regional scale area was selected to be 40 km × 40 km (× 3 km
depth) taking into account the average area and thickness of these accretionary complexes,
which can be considered as representative for Japan. The repository scale and panel scaleareas were set at 5 km × 5 km (× 1.5 km depth) and 800 m × 800 km (× 100 m thickness),
respectively, similar to plutonic and Neogene sediments as described above.
As for the other rocks, the SDM for Pre-Neogene sediments represents regional
groundwater flow system from recharge to discharge areas as well as hydraulic boundaries, as
shown in Figure 3.3-4. In addition, Pre-Neogene sediments consisting of units with different
lithologies and stratigraphic boundaries, together with associated thrusts, faults and folds are
represented.

3-44Figure 3.3-4 Conceptual SDM for Pre-Neogene sediments consisting of units with different
lithologies and stratigraphic boundaries, together with associated thrusts (indicated bydifferent colours)

Many thrusts of different scales develop within accretionary complexes [191]. In the PreNeogene sediments, in addition to thrusts, a range of other large faults will be present, thestructural and hydraulic features of which are considered to be similar to those in plutonic
rocks [181] [192]. Deep groundwater flow is considered, as for the other cases, to be largely
horizontal except for recharge and discharge areas [183].
Supporting Report 3-18 presents the dataset used to construct the SDMs for Pre-Neogene
sediments and Supporting Report 3-19 presents basic information on selecting modellingareas and construction of the conceptual SDM.
(3) Concept for development of geological and hydrogeological models(i) Geological models(a) Plutonic rocks
For plutonic rocks, faults and fractures are considered to play a greater role in defining
groundwater flow and radionuclide migration than the rock matrix. Therefore, similar to cases
where SDMs were developed from outcrop, geophysical and borehole investigations inoverseas siting investigations [96] [97] [98] [99] [100] [101], the distribution of faults andfractures are modelled deterministically or stochastically within a discrete fracture network(DFN) model. Indeed, modelling of the Olkiluoto and Forsmark sites using DFN models has
been accepted by the relevant regulatory organisations as a basis for safety assessment withinapplications for repository construction licenses [193] [194].
When carrying out geological modelling using DFN models, from the viewpoint of theease of characterisation during the staged siting process, faults and fractures are classified bylength as (A) 10 km or more, (B) 1-10 km and (C) less than 1 km. For (A), active faults may
result in M ≥ 6.5 earthquakes, which could potentially have a significant impact onunderground facilities [195]. In this report, such faults would be excluded from the repository
area during the literature survey stage or, at the latest, during the first half of the detailed
investigation stage. Regarding (B), based on the results from the Mizunami URL programme[92][116], it was confirmed that it would be possible to roughly identify the position, structureand hydraulic characteristics of these at the LS stage or again, at the latest, during the firsthalf of the DI stage. The frequency of occurrence of (B) is higher and potential impacts offault movement less than (A), hence these may be permitted within the repository area, but
excluded from the panel scale area. Because (C) would be numerous in the host rock, these
are impossible to characterise completely during PI and DI stages and hence the purpose of
3-45

the investigations is to understand their repository-relevant properties and to utilise statistical
approaches to determine their integrated impact, e.g., assuming a fractal spatial distribution of
faults and fractures [196].
Thus, the regional scale DFN model was developed by stochastically modelling the spatial
distribution of all faults longer than 1 km, based on nationwide data. At the repository scale,based on groundwater flow analysis on the regional scale, a region with relatively long
groundwater travel times was selected from the areas where faults longer than 10 km are notpresent. For faults with length of 1-10 km in the repository scale area, fault locations, strikes
and dips (modelled stochastically at the regional scale area) were represented
deterministically. In the panel scale model, because faults and fractures less than 1 km in
length are distributed throughout, deterministic modelling was adopted as far as possible
based on observations from tunnel walls. The stochastic approach was, however, used where
the spatial distribution of the faults and fractures were not available. Details of the source
information and its application is given in Section (4) below. It should be noted that the“length” of faults and fractures in this report corresponds to trace lengths measured at
outcrops and tunnel walls.
For fault and/or fault zones with a scale of several km or more, a damage zone may
develop asymmetrically around the fault plane [197] [198]. In other cases [47] [83] [92] [176][177] [178], however, a core zone consisting of fault gouge and fault breccia with a thicknessof a few metres, has a damage zone consisting of cataclasite and fractured rock that develops
symmetrically on both sides (schematically shown in Figure 3.3-5).

Figure 3.3-5 Conceptual geological model of a fault(Edited from Lin and Yamashita, 2013 [198])

In particular, the density of fractures in the fault breccia and damage zone is higher than in
the surrounding rock. Based on this, the conceptual geological model includes such features
as the representative concept for faults with a scale of several km or more, assuming that thewidth of the fault impacted zone is about 1/100 of the fault length [46] [199]. These features
are also found in Neogene and Pre-Neogene sediments.
In Supporting Report 3-14, a dataset of relevant nationwide geoscientific information ispresented.

3-46

(b) Neogene sediments
The key geological features of Neogene sediments included in the geological model, as
shown schematically in Figure 3.3-6, include the spatial distribution of faults and fractures
together with their lengths and strikes/dips; lithology and thickness of sedimentary layers;
inclination of monoclinic shear zones; wave lengths/wave heights/axial lengths/axial
directions of folds; and lengths of horizontal/monoclinic/fold structures in a section
perpendicular to the strike or dominant direction of strata.

Figure 3.3-6 Parameters considered for the geological model of Neogene sediments

The parameter values (arithmetic mean) for each stratum, the fold structures andgeological structures were derived from the information on Neogene sediments in the
published 1:50,000 geological map. The abundance of Neogene sediments corresponds toabout 70% of the entire distribution in the 1:200,000 Geological Map [65], and the selectedparameter values were therefore considered to be representative of Japan. The process
involved and the treatment of faults at repository and panel scales is discussed further in
Section (4) below. In Supporting Report 3-16, the dataset of relevant nationwide geoscientificinformation is presented.
(c) Pre-Neogene sediments
Pre-Neogene sediments may comprise coherent facies, consisting of alternating sandstone
and mudstone layers with good continuity, or mélange facies, containing a large number ofrock blocks with mudstone as a matrix. Parameters for the geological model were determinedbased on the same approach as for Neogene sediments described above, using literature
information on the Ashio, Mino, Tamba, Chichibu, and Shimanto Belts [186] [187] [188][189] [190] [191] [200] [201] [202]. This allowed determination of accretionary complex
thickness and slope, together with the wave length, wave height, axial length, axial direction,and plunge of large-scale fold structures (see Figure 3.3-7).

3-47

Figure 3.3-7 Parameters considered for the geological model of Pre-Neogene sediments(Modified from Matsuoka, 1998© Geological Society of Japan [186])

The ratio of the matrix and rock blocks that make up the accretionary complexes wasdetermined using the 1:200,000 Geological Map of Japan [65]. Especially for the mélange
facies, the geological model included stochastic descriptions of rock types andabundance/geometry of contained rock blocks. The process involved and the treatment offaults at repository and panel scales is discussed further in Section (4) below. In Supporting
Report 3-18, the dataset of nationwide geoscientific information used is presented.
(ii) Hydrogeological models and groundwater flow analysis(a) Hydrogeological models
For all representative host rocks, a continuum porous media (CPM) model that describes
the spatial distribution of hydraulic characteristics was developed, capturing the heterogeneityof hydraulic conductivity due to differences in faults, fractures, and rock types that impacts
groundwater flow on the regional and repository scale areas.
For large faults, fault gouge (shown in Figure 3.3-5) has relatively low hydraulicconductivity, however the surrounding damaged zones generally have high hydraulicconductivity [92] [178][179]. The entire fault was modelled taking into account suchanisotropy, with hydraulic conductivity high parallel to the fault plane and low perpendicular
to it. Depending on the rock type and fault formation processes, however, there are cases
where there is little difference between the hydraulic conductivity of fault gouge and/or
damage zones and that of the bulk rock [177].
The panel scale hydrogeology for plutonic rocks and Pre-Neogene sediments utilised DFN
models including faults and fractures < 1 km in length, whereas that for Neogene sediments
utilised a continuum model similar to that for the regional and repository scales. Table 3.3-2
summarises the hydrogeological models for the three rocks at different spatial scales.

3-48

Table 3.3-2 Outline for geological and hydrogeological models for each spatial scale

Modelling method

Regional scale

Scale/method
Spatialscale

Modelling method

Repository scale

Spatialscale

ModellingmethodPanel scale

Spatialscale

Plutonic rocks

Neogene sediments

Pre-Neogene sediments

50 km × 50 km

30 km × 30 km

40 km × 40 km

・ Geological model- Fault (> 1 km)
distribution: Stochastic
DFN*1
・ Hydrogeological model- Continuum

5 km × 5 km

・ Geological model・ Geological model- Fault (> 10 km) distribution:- Fault (> 1 km)
Deterministic DFN
distribution: Stochastic
and deterministic DFN*2 - Fault (1-10 km) distribution:
Stochastic and deterministic- Lithological distribution:
DFN*3
Continuum- Lithological distribution:
・ Hydrogeological modelContinuum- Continuum・ Hydrogeological model- Continuum
5 km × 5 km

5 km × 5 km

・ Geological model・ Geological model・ Geological model- Fault (1 - 10 km)- Fault (1 - 10 km) distribution:- Fault (1 - 10 km)
distribution: Deterministic distribution: Deterministic Stochastic and deterministic
DFN*4
DFN*3
DFN*4
- Lithological distribution: - Lithological distribution:- Fault/fracture (< 1 km)
ContinuumContinuum
distribution: Stochastic
DFN
・ Hydrogeological model ・ Hydrogeological model- Continuum- Continuum・ Hydrogeological model- Continuum
800 m × 800 m

800 m × 800 m

・ Geological model- Fault/fracture (< 1 km)
distribution: Stochastic
DFN・ Geological model- Fault/fracture (< 1 km)
distribution: Stochastic
DFN

・ Hydrogeological model- Stochastic DFN

・ Hydrogeological model- Continuum

800 m × 800 m
・ Geological model- Fault/fracture (< 1 km)
distribution: Stochastic DFN
・ Hydrogeological model- Stochastic DFN*1

DFN: Discrete fracture network.
The locations of faults (>1 km) are treated deterministically, taking into account their geological evolution*3
Only the dominant orientation of the fault (1 - 10 km) is treated deterministically.
*4
The locations, strikes and dips of faults (> 1 km) are included in the regional scale geological model.
*2(b) Determination of hydraulic characteristics
In addition to nationwide dataset, for plutonic rocks and Neogene sediments, geoscientificinformation obtained from JAEA's URL programme [74] [92] [115] [116] was also used – inparticular the transmissivity of faults/fractures based on borehole investigations. For PreNeogene sediments, the transmissivity of faults and fractures was determined based onliterature information only.
(c) Groundwater flow analysis
Three-dimensional groundwater flow analysis was carried out on a regional scale for eachrock, as indicated schematically in Figures 3.3-2 to 3.3-4, with a hydraulic gradient deep
3-49

underground considered to be predominantly horizontal [183]. Based on the average hydraulicgradients in the lowlands, plateaus and hilly areas presented in the H12 report(0.01,0.02,0.04 respectively) [7] and the results of groundwater flow analyses conductednationwide [203], the hydraulic heads at the boundaries were fixed to give a conservative
hydraulic gradient of 0.05. The direction of the hydraulic gradient was set for two orthogonal
directions (X and Y directions) in the model plane, considering the relationship between the
groundwater flow direction and predominant directions of faults and fractures.
For repository scale three-dimensional groundwater flow analysis (see Sections (4) (ii) (b)and (c)), one of the two directions of the groundwater flow on the regional scale was selected.
To assign appropriate boundary conditions in more detail than on the regional scale, thenumerical model domains were selected to be larger than the repository scale area,
dimensions of 10 km × 10 km (×1.5 km depth) for plutonic and Pre-Neogene sediments, and7.5 km × 7.5 km (× 1 km depth) for Neogene sediments were selected for the analysis.
(4) Geological and hydrogeological models(i) Regional scale(a) Plutonic rock
Regardless of their scale, faults and fractures in granite tend to fall into two dominantvertical sets and a low-angle set, roughly orthogonal to each other [204]. Therefore, three fault
sets were established (1-3) based on literature information [92] [176]. The length distributionof faults was determined by analysing the relationship between the trace length and areal
density (number of faults per 1 m2) based on:Active faults (Behavioural segments6) [41].

Fault map of Japan [195].



Calculation results of fault and fracture lengths based on wall observations inunderground cavities [205].



Observations of an outcrop in the Tono area (cumulative frequency distribution of
fractures with a length of 10 m or less observed at the outcrop) [206].

Lineaments longer than 100 m estimated from aerial photographs [7].

By overlaying the geological map at a depth of 500 m [174], i.e. the potential depth of theunderground facilities, onto the information on the fault positions described in the fault mapof Japan [195], the fault length distribution and areal intensity at a depth of 500 m were
calculated.
As a result, it was confirmed that the relationship between the length and the cumulative
areal density (fault length in metres per unit area m2) of fault/fracture can be expressed by apower law (with an exponent of 4.0). The cumulative volumetric fault/fracture density wascalculated from this using the theoretical conversion method of Wang (2005) [207]. The
volumetric fault/fracture density longer than 1 km was determined as ≈ 0.001 m2/m3 and

6

A behavioural segment is a fault section that divides an active fault into segments based on the timing of past
activity, average displacement rate, average activity interval and the direction of displacement. N.B. This
footnote is not included in the Japanese version of the report.

3-50Figure 3.3-8 shows the relationship between the cumulative volumetric fault/fracture density
and length.

Figure 3.3-8 Relationship between fault length and volumetric density of faults and fractures inplutonic rocks

The key parameters used to develop the regional scale model are provided in Table 3.3-3.
Figure 3.3-9 shows an example of a stochastically generated fracture model.
Table 3.3-3 Parameters used for the regional scale model of plutonic rock
Orientation

LengthPower law exponent value
Fault/fracture set Dip direction Dip angle
Fisher
Minimum length (m)(°)(°)
coefficient7
Maximum length (m)1 (NE-trending)

171

85

7.8

2 (NW-trending)
3 (low angle)

080
203

87
01

7.5
8.4

4.0
1,000
70,000Volumetricdensity(m2/m3)0.001

Figure 3.3-9 Regional scale geological model of plutonic rocks
7

The Fisher coefficient indicates the variation in the orientation distribution. N.B. This footnote is not includedin the Japanese version of the report.

3-51

The regional scale hydrogeological model was developed using hydraulic parameters offaults > ≈1 km length and a background fractured rock that incorporates properties of faultsand fractures < 1 km long. For faults > 1 km, low hydraulic conductivity perpendicular to thefault plane and high hydraulic conductivity parallel to it were assigned based on theconceptual representation illustrated in Figure 3.3-5 (i.e., the hydraulic conductivity of fault
gouge is low and that of fault breccia and damage zones is high [92] [178] [179]).
The hydraulic parameters used are presented in Table 3.3-4, while the resulting model isshown in Figure 3.3-10.
Table 3.3-4 Hydraulic parameters used for the regional scale hydrogeological model of plutonicrock
Hydrogeologicalunit

Direction & Hydraulic conductivity (m/s)

Remarks

Fault

Parallel to fault plane: 1.6 × 10-6
Perpendicular to fault plane: 1.3 × 10-9

Only faults > ≈1 km

Rock matrix

1.4 × 10-8

Also includes contribution fromfaults and fractures < 1 km

Figure 3.3-10 Regional scale hydrogeological model of plutonic rocks
Left: Perspective view, Right: Horizontal cross-section at repository depth of 1,000 m

Groundwater analysis at a regional scale was conducted for two groundwater flow
directions in the horizontal plane to assess the relationship between the orientations ofgroundwater flow and predominant faults: Case 1 (flow in the X axis) and Case 2 (flow in the
Y axis). From the Darcy flux distribution calculated, groundwater flow paths were identified
and relative travel times were estimated by a particle tracking method, modelling movementof particles initially positioned at intervals of 100 m and at a depth of 1000 m (repository
depth in plutonic rock; for details, see Section 4.3) within the domain of the facility (5 km × 5km), groundwater flow paths and travel times were calculated to the downstream boundary.
It must be emphasised that the relative travel time (travel distance divided by Darcy flux)
is a useful hydrologic characteristic, but should not be confused with the actual water travel
time, which requires specification of the flow porosity of all water-conducting features. More
importantly, this parameter cannot be related to solute transport times, even for non-sorbing
species, as the latter requires specification of the characteristics of small-scale transport
pathways (e.g. as shown in Figure 3.3-5).
Figure 3.3-11 shows examples of the head distribution calculated within the horizontal
cross-section at a depth of 1000 m.
3-52

Figure 3.3-11 Head distribution at regional scale of plutonic rocks
Left: Case 1, Right: Case 2

Figures 3.3-12 and 3.3-13 show the Darcy flux distribution and the relative groundwater
travel time to the downstream boundary, respectively.

Figure 3.3-12 Darcy flux distribution at regional scale of plutonic rocks
Left: Case 1, Right: Case 2

Figure 3.3-13 Relative groundwater travel time to the downstream boundary at regional scale ofplutonic rocks; Left: Case 1, Right: Case 2

On the regional scale, although the iso-potential contour lines in Figure 3.3-11 are slightly
perturbed by the difference in hydraulic conductivity between the background fractured rockand explicitly modelled faults, a groundwater flow field with an effectively constant hydraulicgradient is observed. The Darcy flux is larger in the faults, but almost constant otherwise. The
relative groundwater travel time is short at positions where faults with a length ofapproximately 10 km or more are located or where multiple faults are connected in the
direction orthogonal to the flow direction. In other areas, this travel time gradually decreases
3-53

from upstream to downstream. This trend remains nearly the same, regardless of the directionof groundwater flow. The heterogeneity of hydraulic properties of plutonic rocks with a highdensity of faults and fractures thus has a small impact on groundwater flow, except in thevicinity of large or connected faults parallel to the groundwater flow direction.
Supporting Report 3-20 presents details of the development of the regional scalegeological and hydrogeological models of plutonic rocks together with the associatedgroundwater flow analysis.
(b) Neogene sediments
Table 3.3-5 shows the parameters used for developing the regional scale geological modelof Neogene sediments, based on the concept discussed in Section (3) (i) (b).
Table 3.3-5 Parameters used the regional scale geological model of Neogene sediments
Parameter

Set value

Length of geological structure(km)

Horizontal 8 ± 1: Monoclinic 6 ± 1: Folding 16 ± 1

Inclination of monoclinic structure

(°)

36±15

Wave length

(km)

3±1

Wave height(km)

0.6±0.3

Axis length

(km)

3

Axis direction(°)

8

Mudstone(m)

449, 449 (double layer)

Alternating
mudstone &sandstone(m)311 (single layer)

Sandstone(m)462, 62 (double layer)

Conglomerate(m)

268 (single layer)

Folding

Formation
thickness

For the predominant orientation of faults with a length of 1 km or more, four arbitrary
directions were determined based on the behavioural segments [41] and the fault map ofJapan [195]. As for the plutonic rock model, a power law (with an exponent of 3.3), which
was assumed based on the relationship between the length and the cumulative areal density offaults/fractures, was applied to describe the length distribution .
The volumetric density was determined as 0.0003 m2/m3 from the relationship between
the cumulative volumetric density and the length of the faults/fractures. Figure 3.3-14 shows
the relationship between the cumulative volumetric fault/fracture density and length. It isnoted that approximation straight line power law was applied though the fit line above theobserved data to ensure that the volumetric density is derived conservatively (overestimated).

3-54

Figure 3.3-14 Relationship between the cumulative volumetric density and length of faults andfractures for Neogene sediments

Parameters used to model the spatial distribution of the faults on a regional scale areprovided in Table 3.3-6.
Table 3.3-6 Parameters used for the regional scale geological model of Neogene sediments(spatial distribution of faults)
Fault/fractureset
12

Orientation
Dip
Dip
Fisherdirectionangle
coefficient
(°)
(°)
246
84
7.4
324
90
24

3

286

41

9.6

4

103

44

15

LengthPower law exponent value
Minimum length (m)Maximum length (m)Volumetricdensity(m2/m3)3.3
1,000
70,0000.0003In the development of the regional scale model, the impacts of changes in the regional
stress field since the Neogene [208] [209] were considered. As shown in Figure 3.3-15, thisconsiders rifting by normal faults under tensile stress in the early Neogene, prior to
sedimentation during the middle Neogene while the rift evolves, and then uplift/tilting/folding
associated with the movement of reverse faults under compressive stress from the late
Neogene to the Quaternary.

3-55

Figure 3.3-15 Developmental history of the geological structure of Neogene sediments

The geological structure and spatial distribution of lithofacies were modelled based on theparameters in Table 3.3-5, after uplift and tilt of the strata had been assessed deterministically.
Figure 3.3-16 shows the resulting regional scale model.

Figure 3.3-16 The regional scale geological model of Neogene sediments
Upper left: perspective view, Lower left: spatial distribution of faults, Right: Horizontal crosssection at repository depth of 500 m

The regional scale hydrogeological model was developed by determining the hydraulicparameters of each facies and also faults with a length > ≈ 1 km. In particular, for such faults,
3-56

fault gouge and fault breccia/damage zones are considered (as shown in Figure 3.3-5).
However, based on a case study indicating that the anisotropy of hydraulic conductivity is lessthan that in plutonic rocks [210], an isotropic hydraulic conductivity value was assigned forthe entire fault (including both core zone and damage zone).
The parameters used to develop the regional scale model are given in Table 3.3-7 and theresult presented in Figure 3.3-17.
Table 3.3-7 Parameters used for the regional scale hydrogeological model of Neogene
sediments
Hydrogeological unit

Hydraulic conductivity (m/s)

Quaternary

1.0 × 10-5

Mudstone

2.3 × 10-8

Alternating sandstone andmudstone

Vertical: 2.3 × 10-8
Horizontal: 5.3 × 10-7

Sandstone

5.3 × 10-7

Conglomerate

6.5 × 10-8

Basement

1.1 × 10-8

Fault

5.4 × 10-7

Remarks

Vertical: identical to mudstone
Horizontal: identical to sandstone

Length > 1 km

Figure 3.3-17 Regional scale hydrogeological model of Neogene sediments
Left: Perspective view, Right: Horizontal cross-section at cross-section depth of 500 m

The groundwater analysis using the regional scale hydrogeological model was conducted
for two horizontal groundwater flow directions as for plutonic rocks; Case 1 (flow in the Xaxis) and Case 2 (flow in the Y axis). Using the Darcy flux distribution of the groundwaterflow analysis results, the groundwater flow paths were identified using a particle tracking
method.
Figure 3.3-18 shows the head distribution obtained by the groundwater flow analysis at
the repository depth of 500 m.

3-57

Figure 3.3-18 Head distribution at the regional scale for Neogene sediments
Left: Case 1, Right: Case 2

Figures 3.3-19 and 20 show the Darcy flux distribution and the relative groundwater travel
time to the downstream boundary, respectively

Figure 3.3-19 Darcy flux distribution at the regional scale for Neogene sediments. Left: Case 1,
Right: Case 2

Figure 3.3-20 Relative groundwater travel time to the downstream boundary on the regional
scale for Neogene sediments. Left: Case 1, Right: Case 2

On the regional scale, for Case 1, the groundwater flow direction intersects the
predominant orientation of the faults (NNE-SSW to NE-SW) and formations (NNE-SSW) at alarge angle (tending towards perpendicular). In contrast, for Case 2, the intersecting angle issmall (tending towards parallel). This explains differences in the iso-potential contours and, inparticular, the Darcy flux for the two cases.

3-58Supporting Report 3-21 presents the details on the development of the regional scalegeological and hydrogeological models of Neogene sediments together with the associatedgroundwater flow analysis.
(c) Pre-Neogene sediments
In developing the regional scale geological model of Pre-Neogene sediments, theparameters for accretionary complexes, fold structures and lithology (see Figure 3.3-7) were
determined based on an analysis of the literature [186] [187] [188] [189] [190] [191] [200][201] 202], described previously in Section (3) (i) (c).
The distribution of coherent and mélange facies within a single accretionary complex, as
well as the rock fractions contained therein, were derived from the 1:200,000 Geological Map
of Japan [65]. For the coherent facies, parameters were determined with reference to the
Shimanto Belt, which is predominantly comprised of this facies. For the mélange facies, theparameters were determined with reference not only to the Ashio, Mino, Tamba and Chichibu
Belts, where mélange facies are dominant, but also the Tokoro, Hidaka, Idonnap, andNorthern Kitakami Belts, considering the heterogeneity of lithology described later.
For the lithology of the coherent facies, based on Nakae (2000) [188], mudstone-dominant
matrix (ms), sandstone-dominant matrix (ss) and rock blocks (chert) were assumed. For the
mélange facies, the type and abundance ratio of rock blocks and matrix were determinedbased on data from the accretionary complexes noted above.
The accretionary complex was modelled based on the parameters shown in Table 3.3-8.
Table 3.3-8 Parameters used for the regional scale geological model (geological structure andlithofacies) of Pre-Neogene sediments
Parameter
Accretionary
complex

Folding

Selected value

Thickness

(m)

2,500

Inclination

(°)

50

Matrix composition(%)

Coherent facies: 0-5,
Mélange facies: 20-40

Wave length

(km)

20

Wave height

(km)

6

Axial length

(km)

10

Axial orientation

(°)

90

Plunge(°)

0-25W

Matrix/composition(%)

Coherent facies: ms dominant 50, ss dominant 50
Mélange facies: Mudstone 100

Matrix/composition(%)

Coherent facies: Chert 100
Mélange facies: Greenstone 45, Chert 40,
Ultramafic rock 10, Limestone 5

Lithology

Different accretionary complexes come into contact at thrusts [186] [191], as shown in
Figure 3.3-4. Thrusts of 10 km or more that separate the accretionary complexes are
expressed deterministically as boundaries in the model. For faults 1 to 10 km long in coherent
facies, the cumulative areal density of faults/fractures was approximated by a power law witha conservatively assumed exponent of 4.0 (Figure 3.3-21) based on data from behaviour
segments [41], the fault map of Japan [195], and fault/fracture lengths observed on tunnel
3-59

walls [205]. The model was then developed with the azimuth and inclination determined
deterministically to match the entire geological structure (folding and thrust distribution).

Figure 3.3-21 Relationship between cumulative volumetric density and length of faults andfractures for Pre-Neogene sediments

For mélange facies, the model was constructed with the azimuth and inclination set
deterministically, but assuming that about 30% of faults 1 to 10 km long correspond to out-ofsequence thrusts, formed independently of the internal structure of the old accretionary
complex, based on the 1:50,000 scale geological map of the Chichibu Belt distributed in the
Kanto, Shikoku and Kyushu regions.
Figure 3.3-22 shows examples of regional scale geological models of coherent andmélange facies resulting from this process.

3-60

Figure 3.3-22 Regional scale geological models of Pre-Neogene sediments
Left: Coherent facies model, right: Mélange facies model

For the coherent facies, hydraulic parameters were determined for thrusts over 10 km in
length, mudstone-dominant layers, sandstone-dominant layers, and rock blocks (chert). The
geological information on Pre-Neogene sediments collected nationwide is, however,relatively limited compared to plutonic rocks and Neogene sediments, and is extremely
limited below 300 m. When deeper data were available, these were used, otherwise it was
assumed that the depth dependence of the data observed for other rock types would hold. For
the mélange facies, because no data on the differences in the hydraulic parameters as afunction of the rock block abundance were available, these were determined assuming that the
mélange facies has macroscopically homogeneous properties on regional scale, except forlarge thrusts. The hydraulic conductivity of thrusts longer than 10 km was determined fromthe permeability measured for large faults within the Pre-Neogene sediments constituting theaccretionary complex [181] [182] [192].
From this analysis, it was confirmed that hydraulic conductivity of deep Pre-Neogene
sediments could be less than that of background fractured rock of plutonic rock (including
faults and fractures of less than 1 km in length) or mudstone layers of Neogene sediments.
The hydraulic parameters used for the regional scale hydrogeological model are summarisedin Table 3.3-9 and the resulting models are shown in Figure 3.3-23.
Table 3.3-9 Hydraulic parameters used for the regional scale hydrogeological models for PreNeogene sediments
Hydrogeological unit

Hydraulicconductivity (m/s)

Remarks

Thrust1.0 × 10-8

Length > 10 km

2.0 × 10-9

ms dominant, ss dominant

Coherent
faciesMélangefaciesMatrixRock mass
Matrix

1.0 × 10

-8

Chert

2.0 × 10-9Rock mass

3-61

Figure 3.3-23 Regional scale hydrogeological models of Pre-Neogene sediments
Left: Coherent facies, right: Mélange facies

Again, the groundwater flow analysis was conducted for two groundwater flow directions.
Using the Darcy flux distribution of the groundwater flow analysis results, the groundwaterflow paths and travel times were estimated based on the particle tracking method.
Figure 3.3-24 shows the head distribution obtained by the groundwater flow analysis
within the horizontal cross-section at the repository depth of 1000 m. Figures 3.3-25 and 3.326 show the Darcy flux distribution and the relative groundwater travel time to thedownstream boundary, respectively.

Figure 3.3-24 Head distribution at the regional scale for Pre-Neogene sediments
Upper left: Coherent facies (Case 1), Upper right: Coherent facies (Case 2)
Lower left: Mélange facies (Case 1), Lower right: Mélange facies (Case 2)

3-62

Figure 3.3-25 Darcy flux distribution at the regional scale for Pre-Neogene sediments
Upper left: Coherent facies (Case 1), Upper right: Coherent facies (Case 2)
Lower left: Mélange facies (Case 1), Lower right: Mélange facies (Case 2)

Figure 3.3-26 Relative groundwater travel time to the downstream boundary at a regional scaleof Pre-Neogene sediments
Upper left: Coherent facies (Case 1), Upper right: Coherent facies (Case 2)
Lower left: Mélange facies (Case 1), Lower right: Mélange facies (Case 2)

On the regional scale, the hydraulic gradient from upstream to downstream was uniform
due to the low contrast of hydraulic properties between thrusts, matrix and rock blocks.
3-63

Although the Darcy flow flux in the coherent facies was small compared to mélange facies,
where the area between thrusts was narrow in the direction of groundwater flow, the travel
time was generally short because flow occurred at a relatively high flux along the thrust.
In Case 2 (flow in Y axis direction) where the groundwater flow direction intersects thethrusts with a high angle, there was a tendency for less flow along thrusts compared to Case 1(flow in X axis direction) where the intersection angle is low.
Supporting Report 3-22 presents details on the development of the regional scalegeological and hydrogeological models of Pre-Neogene sediments together with theassociated groundwater flow analyses.
(ii) Repository scale(a) Plutonic rocks
The regional scale groundwater flow analysis showed that Darcy flow flux distributions
were similar regardless of the direction of groundwater flow (Figures 3.3-12 and 3.3-13). The
repository scale area was thus determined after arbitrarily selecting the groundwater flow as
Case 2 (flow in Y direction). In particular, the regional scale area excludes faults with a lengthof 10 km or longer (represented by the green planes in the right figure of Figure 3.3-9 and the
red lines in the left figure of Figure 3.3-27) as well as the areas potentially impacted by them.
Thus, an area with a relatively low fault/fracture density and long groundwater travel time(see Figure 3.3-13) was selected (Figure 3.3-27).
Faults with a length of 1 km or more in the repository scale geological model weremodelled deterministically using fault locations, strike and dip derived from the regional scalegeological model (see Figure 3.3-9). Faults and fractures less than 1 km in length weremodelled stochastically, assuming that the minimum fracture length that can be determinedfrom boreholes is 1 m, using the same statistical distribution of faults and fractures (see Table
3.3-3) as the regional scale geological model. Figures 3.3-27 and 3.3-28 show the selected
repository scale area and the derived geological model, respectively.

Figure 3.3-27 Selection of the repository scale area for plutonic rocks
Left: Geological model, right: groundwater travel time distribution (from Figure 3.3-13 right)

3-64Figure 3.3-28 Repository scale geological model for plutonic rocks
Left: 3D visualisation, right: Horizontal cross-section at repository depth of 1,000 m

As for the regional scale, the repository scale hydrogeological model includes
deterministically modelled faults longer than 1 km and background fractured rock (including
faults and fractures shorter than 1 km) with appropriate hydraulic parameters, as given in
Table 3.3-10.
Table 3.3-10 Hydraulic parameters used for the repository scale hydrogeological model ofplutonic rocks
Hydrogeologicalunit

Hydraulic conductivity (m/s)

Remarks

Fault

Parallel to fault plane 1.6×10-6
Perpendicular to fault plane 1.3×10-9> 1 km in length

Rock matrix

2.7 × 10-8

Including faults and fractures < 1 km in
lengthFor the hydraulic conductivity of the background fractured rock, equivalent hydraulicconductivities for DFN model realisations were calculated and an average value used. The
resultant hydrogeological model is illustrated in Figure 3.3-29.

Figure 3.3-29 Repository scale hydrogeological model of plutonic rocks
Left: 3D visualisation, right: Horizontal cross-section at the repository depth of 1,000 m

Groundwater flow analysis at the repository scale was carried out by imposing fixed head
conditions at the boundaries, based on the head distributions obtained from the regional scale
analysis (Figure 3.3-11). Figures 3.3-30 and 3.3-31 show the resultant head and Darcy flux
distributions, respectively.

3-65

Figure 3.3-30 Head distribution at the repository scale for plutonic rocks
Left: Vertical cross-sections, right: Horizontal cross-section at a repository depth of 1,000 m

Figure 3.3-31 Darcy flux distribution at the repository scale for plutonic rocks
Left: Vertical cross-sections, right: Horizontal cross-section at a repository depth of 1,000 m

Figure 3.3-32 shows the distribution of relative groundwater travel times, which is used
for determining the area where the repository will be located (for details, see Section 4.5.4(1)). These times assume a distance of 500 m, set as the shortest distance from the repository
to the model boundary. This clearly shows that groundwater travel times tend to be relatively
short in and upstream of faults.

Figure 3.3-32 Groundwater travel time distribution (for 500 m) at the repository scale forplutonic rocks

3-66The details of the repository scale geological and hydrogeological models as well as thegroundwater flow analyses are provided in Supporting Report 3-23.
(b) Neogene sediments
The selection of the repository scale area took into account the need to make the areas
sufficiently large to allow adjustment of the layout to meet the requirements for engineering
feasibility and long-term post-closure safety, based on the regional scale geological model(see Figure 3.3-16) and groundwater flow analysis (see Figures 3.3-18 to 3.3-20). In particular,
as a requirement to ensure the stability of disposal tunnels, it is preferred that the axial
direction of such tunnels is aligned with the direction of maximum stress in the horizontal
plane. From the viewpoint of reducing the influence of groundwater flow on the release ofradionuclides, however, it may be preferred that the axial direction of the disposal tunnel isorthogonal to the groundwater flow direction (for details, see Section 4.5.4 (3)). The two
groundwater flow directions considered, Case 1 (flow in the X direction) and Case 2 (flow in
the Y direction), need to be assessed for compressive stress aligned in the X direction as
defined in the regional scale model. For Case 1, the boundary conditions are more challenging
as the two requirements conflict when tunnel orientation is being decided. In order to examine
the issues involved, this is selected as the reference case for the SDM.
Areas with relatively long groundwater travel times were first selected, based on theregional scale groundwater flow analysis (see Figure 3.3-20). Then, taking account thecomplex geological structures that may be encountered in this rock, an area with a foldstructure and multiple lithological units was arbitrarily selected to allow assessment of thelayout of disposal areas and other underground facilities (for detail, see Sections 4.5.4 (1), (3)and (6)) and associated post-closure safety assessment (see Sections 6.4.1 (4) and (6)).
Figures 3.3-33 and 3.3-34 show the selected repository scale area and resultant geologicalmodel, respectively.

Figure 3.3-33 Selection of the repository scale area for Neogene sediments
Left: Geological model (from Figure 3.3-16 right), Right: Groundwater travel time distribution(from Figure 3.3-20 left)

3-67

Figure 3.3-34 The repository scale geological model for Neogene sediments
Left: 3D visualisation, right: Horizontal cross-section at a repository depth of 500 m

In the repository scale hydrogeological model, different hydraulic conductivity values
were assigned to the mudstone layers above and below the 62 m thick sandstone layer. The
hydraulic parameters used for developing the repository scale hydrogeological model areprovided in Table 3.3-11.
Table 3.3-11 Hydraulic parameters used for the repository scale hydrogeological model of
Neogene sediments

Quaternary
Upper mudstone

Hydraulic conductivity(m/s)
1.0 × 10-5
2.0 × 10-9

Lower mudstone

4.4 × 10-8

Alternating sandstone
and mudstone

Vertical: 2.3 × 10-8
Horizontal: 5.3 × 10-7

Sandstone
Fault

5.3 × 10-7
5.4 × 10-7

Hydrogeological unit

Remarks

Calculated from the panel scale DFN model
Logarithmic mean of equivalent hydraulicconductivity distribution
Vertical: logarithmic mean of upper and lower
mudstone layers
Horizontal: same as sandstone layers> 1 km in length

The developed hydrogeological model is shown in Figure 3.3-35.

Figure 3.3-35 The repository-scale hydrogeological model of Neogene sediments
Left: 3D visualisation, right: Vertical cross-section

Groundwater flow analysis at the repository scale imposed fixed hydraulic heads at theboundaries based on results of the regional scale groundwater flow analysis (Figure 3.3-18).

3-68

As shown in Figure 3.3-36, the head distributions indicate downwards flow in the shallow
part (upper mudstone layer) on the upstream side, while, on the downstream side, the flow
was almost horizontal with only a slight downwards component.

Figure 3.3-36 Head distribution at the repository scale for Neogene sediments
Left: Horizontal cross-section at depth of 500 m, right: Vertical cross-sections

The Darcy flux distribution obtained is shown in Figure 3.3-37.

Figure 3.3-37 Darcy flux distribution at the repository scale for Neogene sediments. Left:
Horizontal cross-section at a repository depth of 500 m, right: Vertical cross-sections

Figure 3.3-38 shows the distribution of relative groundwater travel time over a distance of
500 m (for details, see Section 4.5.4 (1)) calculated from the Darcy flux distribution in ahorizontal section at a repository construction depth of 500 m, which is used to select the area
where the repository will be located.

3-69

Figure 3.3-38 Relative groundwater travel time distribution at the repository scale for Neogene
sediments.

The details of the repository scale geological and hydrogeological models as well as
associated groundwater flow analyses are provided in Supporting Report 3-24.
(c) Pre-Neogene sediments
The repository scale area was selected based on the results of regional scale groundwaterflow analysis (see Figures 3.3-24 to 3.3-26), excluding thrusts with a length of 10 km or more(see Figure 3.3-22).
For coherent facies, the Darcy flux and groundwater travel time distribution obtained fromthe groundwater flow analysis (see Figure 3.3-25 upper left and upper right figures and Figure
3.3-26 upper left and upper right figures), show a tendency for fast groundwater flow along
the thrusts in Case 2 (flow in the Y direction) that is less apparent in Case 1 (flow in the X
direction). Thus, groundwater flow direction in Case 2 was selected as the reference forfurther study in this report.
Areas with relatively small Darcy flux were first selected, based on the distribution at therepository depth of 1,000 m from the regional scale groundwater flow analysis (see Figure
3.3-25, upper right). Then, taking account of the complex geological structures that may be
encountered in such rocks, an area where multiple facies are distributed was arbitrarily
selected for further design studies, (for detail, see Sections 4.5.4 (1), (3) and (6)) andassociated post-closure safety (for detail, see Sections 6.4.1 (4) and (6)). Figure 3.3-39 shows
the repository scale area selected for the coherent facies.

3-70Figure 3.3-39 Selection of the repository-scale area for Pre-Neogene sediments.
Coherent facies, left: Geological model (reshown, see Figure 3.3-22 left), right: Darcy flux
distribution (reshown, see Figure 3.3-25 upper right)

For mélange facies, as discussed in Section (4) (i) (c), an arbitrary modelling domain was
selected as the spatial variations in head, Darcy flux, and relative groundwater travel timefrom the regional scale groundwater flow analysis were small (Figures 3.3-24 to 3.3-26) due
to the homogeneous hydraulic properties assumed for all components, except thrusts (Figure
3.3-23).
Faults with a length of 1 km or more (whose spatial distribution could be derived from site
investigations) were modelled deterministically. Their strikes/dips were determined to beconsistent with those of thrusts that are 10 km or longer. For mélange facies, it is difficult to
support a deterministic model for the spatial distribution of rock blocks of various sizes andshapes [186] [188] [191] [200] [201] [211] [212], even following geophysical and borehole
investigations. The spatial distribution of rock blocks was thus modelled stochastically,
assuming that sufficient information on the types of rock blocks and their one- or twodimensional distribution could be obtained by site investigations.
Figure 3.3-40 shows the repository scale geological models for coherent and mélange
facies.

3-71

Figure 3.3-40 Repository scale geological models for Pre-Neogene sediments
Left: Coherent facies model, right: Mélange facies model

The repository scale hydrogeological model is the same as that used at a regional scale.
The heterogeneity of the hydraulic properties of the coherent facies is clearly greater than thatof the mélange facies, making it easier to identify issues related to engineering feasibility andpost-closure safety of the repository, and hence this is selected as the reference case.
Table 3.3-12 shows the hydraulic parameters used the coherent facies and the resultinghydrogeological model is shown in Figure 3.3-41.
Table 3.3-12 Hydraulic parameters used for Pre-Neogene sediments

Faults and fractures

Hydraulic conductivity(m/s)
1.0 × 10-8

Sediment

2.0 × 10-9Rock mass

1.0 × 10-8

Hydrogeological units

Remarks> 1 km in length
Sandstone dominant and mudstone
dominant formations are identical
Chert

Figure 3.3-41 Repository scale SDM for Pre-Neogene sediments
Coherent facies, right: 3D model, left: Horizontal cross-section at a depth of 1000 m

The groundwater flow analysis at a repository scale again imposed fixed hydraulic heads
at the boundaries based on the regional scale groundwater flow analysis (Figure 3.3-24). The
resulting Darcy flux distribution at a depth of 1000 m is shown in Figure 3.3-42 while Figure

3-72

3.3-43 shows the distribution of the relative groundwater travel time for a distance of 500 m(for details, see Section 4.5.4 (1)).

Figure 3.3-42 Head and Darcy flux distributions at the repository scale for Pre-Neogene
sediments. Coherent facies, left: head distribution, right: Darcy flux distribution

Figure 3.3-43 Relative groundwater travel time distribution for Pre-Neogene sediments on therepository scale

The results show that groundwater travel times tend to be shorter in the vicinity/upstreamof the faults and chert rock blocks. The details of the development of the repository scalegeological and hydrogeological models as well as the groundwater flow analyses for PreNeogene sediments is provided in Supporting Report 3-25.
(iii) Panel scale(a) Plutonic rocks
The panel scale geological model was developed following the concept given in Section(3) (i) (a), using the same statistical data on faults and fractures (see Table 3.3-3, Figure 3.3-8)
as for the regional and repository scale geological models. In particular, for faults andfractures shorter than 1 km, assuming a minimum length that could be determined from3-73

boreholes is 1 m, their volumetric density was determined based on one-dimensional density
and inclination data from borehole wall observations [207].
The panel scale hydrogeological model was developed using transmissivity values
obtained from a series of hydraulic tests lengths of 10 m or less at depths of 500 m or more,
conducted in the Mizunami URL programme [213] [214]. In particular, the logarithmic meanand standard deviation of the transmissivity distribution were first tentatively assigned tofaults and fractures modelled at a panel scale, with simulated results then compared withmeasurements [213] [214] for multiple DFM realisations. The transmissivity distribution thatbest described observations was then selected.
The hydraulic parameters used for developing the panel scale geological andhydrogeological models are provided in Table 3.3-13 and the resultant the panel scale models
shown in Figures 3.3-44 and 45, respectively.
Table 3.3-13 Hydraulic parameters used for panel scale geological and hydrogeological modelsof plutonic rocks
Orientation
Fault/fractureset

1 (EW trending)
2 (NS-trending)
3 (low angle)

Transmissivity
and distribution

LengthDip
azimuth(°)

Dip angle(°)

FishercoefficientPower law exponent
value
Minimum length (m)Maximum length (m)

171
80
203

85
87
1

7.8
7.5
8.4

4.0
1
1,000

3D density(m2/m3)

1.9
1.4
0.5

Figure 3.3-44 Panel scale geological model of plutonic rocks

Figure 3.3-45 Panel scale hydrogeological model of plutonic rocks

3-74

Logarithmic
mean (m2/s)
Logarithmic
standard
deviation
1.0 × 10-9
2.0

The details of the development of the panel scale models are provided in Supporting
Report 3-26.
(b) Neogene sediments
The panel scale geological model was developed for an arbitrary area of mudstone layerswith relatively low hydraulic conductivity in order to assess issues related to post-closure
safety. In particular, a case study [74] showed that the orientation, length and density offractures differ depending on the timing and environment of formation, mineral composition,
physical and mechanical properties of the mudstone. Thus, different sets of faults andfractures were generated for the upper and lower mudstone layers. Faults and fractures lessthan 1 km in length were modelled stochastically by applying a power law model (Figure 3.314) similar to that used for the regional and repository scale geological model.
The panel scale hydrogeological model describes the spatial heterogeneity of hydraulicproperties of both the mudstone matrix and fractures. The transmissivity values of fractures
were determined based on the results of borehole investigations carried out in the Horonobe
URL programme [74]. The parameters used for developing the panel scale model areprovided in Table 3.3-14, while resultant geological and hydrogeological models areillustrated in Figures 3.3-46 and 47, respectively.
Table 3.3-14 Parameters used for panel scale geological and hydrogeological models of
Neogene sediments

Fault/fracture
set

Upper
Mudstone
Formation
Lower
mudstone
layer

1
2
3
1
2
3
4

Orientation

LengthDip
azimuth(°)

Dip angle(°)

FishercoefficientPower law
exponent value
Min length (m)Max length (m)

323
203
118
341
215
69
227

52
42
11
57
34
26
12

9.8
52
3.1
16
20
51
3.1

3.3
0.15
1,000
3.3
0.15
1,000

3D
density(m2/m3)

Transmissivity
and distribution
Logarithmic mean
(m2/s)
Logarithmic
standard deviation

0.53
0.06
0.44
0.63
0.26
0.22
0.79

Figure 3.3-46 Panel scale geological model of Neogene sediments
Fault/fracture distribution in the upper mudstone layers

3-75

1.1 × 10-8
0.43

7.8 × 10-8
0.52

Figure 3.3-47 Panel scale hydrogeological model of Neogene sediments
Hydraulic conductivity distribution, Left: Horizontal (X) direction, Middle: Vertical (Z) direction.
Right: Horizontal (Y) direction

The details of the development of the panel scale models of Neogene sediments isprovided in Supporting Report 3-27.
(c) Pre-Neogene sediments
The panel scale geological model was developed using the same concept as that forplutonic rocks, with faults and fractures <1 km in length modelled using a power law basedon the results of the fault/fracture length measurements [205] from tunnel wall observations inJapan. The panel scale hydrogeological model was developed based on the hydraulicconductivity and density of water-conducting features used in the repository scalehydrogeological model. The parameters used for developing the panel scale model areprovided in Table 3.3-15.
Table 3.3-15 Parameters used for panel scale geological and hydrogeological models of PreNeogene sediments
LengthOrientation
Fault/fracture set

Dipazimuth
(°)

Dip
angle
(°)

Fishercoefficient

1 (NE-trending)

358

76

20

2 (NW-trending)

96

78

20

3 (low angle)

223

19

20

Power law
exponent value
Min length (m)Max length (m)
4.0
0.15
1,000

3D density(m2/m3)

Transmissivity anddistribution
Logarithmic mean
(m2/s)
Logarithmic
standard deviation

12.5
1.25

2.0 × 10-9
1.0

1.25

These parameters were used for both coherent and mélange facies. Following thediscussion in Section (4) (ii) (c), modelling at the panel scale was carried out for the coherent
facies. Examples of the resultant panel scale geological and hydrogeological models areshown in Figures 3.3-48 and 49, respectively.

3-76

Figure 3.3-48 Constructed panel scale geological model of Pre-Neogene sediments
Fault/fracture distribution for coherent facies

Figure 3.3-49 Constructed panel scale hydrogeological model of Pre-Neogene sediments
Coherent facies

The details of the development of the panel scale geological and hydrogeological models
for Pre-Neogene sediments is provided in Supporting Report 3-28.
(5) Conceptual modelling of water-conducting microstructure
For each of the representative host rocks, based on geological environment investigations
and in-situ tests, a conceptual model was constructed to describe the characteristics of waterconducting microstructures that control radionuclide migration/retardation on a scale of a few
cm to several tens of cm, as required for assessing post-closure safety (see Section 7.2.1 (2)(ii) (c) for further details). It can be noted here however that “flowing porosity” from suchmodels will allow Darcy fluxes to be converted into transport times, that can then be used forthe purposes of comparison with those calculated in previous sections.
(i) Plutonic rocks
Granites are taken as representative of plutonic rocks, allowing transfer of the results ofnatural analogue studies and in-situ tests conducted by JAEA at the Kamaishi mine [176][215] [216]. Analysis of water-conducting fractures shows that advective flow occurs in a
network of flow channels formed within fracture infill. It has also been confirmed that a
network of interconnected pores exists in the fracture filling layer and the rock matrix (which
has been altered), allowing solute diffusion into this material from the flow channels.
Taking these findings into account, a conceptual model of water-conducting
microstructure in plutonic rocks is derived, as shown in Figure 3.3-50.

3-77

Figure 3.3-50 Conceptual model of water-conducting microstructures in plutonic rocks(Ota et al., 1999 [216])

In addition, based on a dual porosity model [217], migration and retardation ofradionuclides in the water-conducting microstructure were conceptualised. The details of themodelling procedure are provided in Supporting Report 3-29.
(ii) Neogene sediments
Based on the findings from the Horonobe URL programme and the geoscientific researchat the Tono mine [218] [219] [220], faults and fractures are known to be infilled with clay androck fragments originating from the host rock, and contain a network of flow channels. This
infill lies within a micro-fractured crushed structure (damage zone). Like plutonic rocks, it isconsidered that advection occurs predominantly in flow channels in the water-conducting
fractures, with solute diffusion through the damage zone and into the rock matrix. As shownin Table 3.3-1, the large porosity and the developed network of connected pores in the matrix
allows also groundwater flow to some extent in the matrix.
The resulting conceptual model, illustrated in Figure 3.3-51, defines the key migration
scale features for Neogene sediments, which allows for dual permeability [217] migration ofradionuclides in both channels and rock matrix. For details, see Supporting Report 3-30.

3-78

Figure 3.3-51 Conceptual model of the water-conducting microstructure for Neogene
sediments(iii) Pre-Neogene sediments
In Pre-Neogene sediments, based on analyses of water-conducting fractures in coresamples [221], faults and fractures are known to be filled with calcite and clay minerals inwhich a network of flow channels occurs. As shown in Supporting Report 3-28 and Table 3.315, considering that the density of water-conducting fractures is high and their hydraulicconductivity is higher than that of the matrix, as for plutonic rocks it is assumed thatadvection occurs predominantly in the flow channels of water-conducting fractures, withsolute diffusion from these into the fracture-filling mineral layer and the rock matrix.
Thus, the conceptual model shown in Figure 3.3-52 illustrates the water-conducting
microstructure of Pre-Neogene sediments.

Figure 3.3-52 Conceptual model of the water-conducting microstructure for Pre-Neogene
sediments (the porosity and matrix permeability of these rocks is similar to that of plutonic
rock, thus advective flow in the matrix was not considered)In addition, based on the dual porosity model [217], the migration and retardation ofradionuclides in the water-conducting microstructure is conceptually shown. For details, see
Supporting Report 3-31.

3-79

(6) Modelling groundwater chemistry
For each representative host rock, a model of associated groundwater chemistry is
required for repository design (see Section 4.4 for details) and assessment of post-closure
safety (see Sections 6.1 to 6.5 for details).
Groundwater chemical data are generally obtained through sampling in boreholes, and canbe significantly affected by disturbances caused by both drilling (e.g. contamination due tothe drilling fluid) and sample collection/analysis (e.g. degassing and oxidation). Particularly
for assessing post-closure safety, it is essential to use groundwater chemical datasets ofguaranteed quality, with effects of perturbations avoided to the extent possible. Based on this,available groundwater chemical data were first screened, using the procedure for quality
assurance implemented for the Horonobe URL programme [222] [223]. For the groundwaterin potentially suitable host rocks, where long term stability is expected, representative waterchemistry datasets were then modelled to check for the above-mentioned disturbances. Here it
was assumed that water chemistry is determined by chemical equilibration processes with theminerals existing in the water-conducting channels, allowing consistency of observed data to
be assessed (even though disequilibrium is commonly observed in many deep groundwater
systems, especially in terms of redox).
The established representative water chemistry datasets were assumed to apply to theentire host rock, thus spatial variations in chemistry were not considered. Only limited
information on colloids, organic matter and microorganisms in groundwater is available fromthe JAEA URL programme [224] [225] [226] [227] [228] and thus, because potential impacts
on radionuclide migration are sensitive to the hydro-geochemical setting, these factors were
not considered further at the present stage at which no site has been identified.
Figure 3.3-53 summarises the procedure for determining representative groundwaterchemical datasets, with further details in Supporting Report 3-32.

Figure 3.3-53 Processes for determining representative groundwater chemistry datasets forrepresentative host rocks(i) Acquisition and screening of groundwater chemical data
In addition to the nationwide groundwater chemical database [229], additional data werecompiled from resource exploration and hot spring surveys conducted all over Japan; JAEA’s
3-80

URL programme [74] [92]; the Kamaishi in-situ Test site [176]; and the Yokosuka
demonstration project [78] (I in Figure 3.3-53). The references used are summarised in
Supporting Report 3-32.
For the collected groundwater chemical data, the rock type was first identified frominformation on the groundwater sampling locations and depths. The data were then screened
from the viewpoint of including only the three considered host rocks, as well as representing
depths of 300 m or more (II in Figure 3.3-53). Further screening was then carried out toassure required data quality (III in Figure 3.3-53). Specifically, groundwater chemical data
were selected based on (A) contamination by drilling fluid less than 10 %, (B) tritium in the
groundwater less than 1 TU (tritium unit) indicating negligible contamination by drilling fluidor surface water, (C) data allows determination that any increase in pH or changes inchemistry due to degassing effects during sampling are small, (D) the charge balance of
cations and anions conforms to the analysis guidelines [230], and thus the analysed values canbe confirmed to be reliable.
As a result, as shown in Figure 3.3-54, chemical data for groundwater with low and highsalinities for plutonic rocks and high salinity groundwater for Neogene sediments werecompiled. For groundwater with low salinity in Neogene sediments and data for Pre-Neogene
sediments, however, water chemistry data failed to meet the above-mentioned requirements(A) ~ (D), predominantly due to a lack of information to assess the degree of contamination
by drilling fluid.

Figure 3.3-54 Selection of groundwater chemistry data for geochemical modelling

As shown in Figure 3.3-54, qualified low salinity groundwater chemical data forshallower Neogene sediments were then considered. Such groundwaters have long residence
times in the formation, and have features that meet geochemical requirements such as being
neutral to weakly alkaline, chemically reducing and with low carbonate concentrations (seeTable 3.1-2) [79] [80]. As groundwater with a low salinity also exists within Neogene
sediments deeper than 300 m, selection of the shallower water appears justified (IIIa in Figure
3.3-53).
For Pre-Neogene sediments, even shallower groundwater chemical data satisfying quality
requirements could not be found. Therefore, considering that Neogene and Pre-Neogene
3-81sediments have similar ranges of groundwater chemistry (e.g. in terms of salinity) and theminerals present in the target mudstones, the water chemistry data selected for Neogene
sediments were also used here (IIIb in Figure 3.3-53).
(ii) Initial conditions for geochemical modelling
Initially, for each rock, the initial groundwater chemistry was defined (IV in Figure 3.353) from selected data with both low and high salinities, and consideration of the ambient
temperature at repository depth, minerals in the water-conducting channels that are expected
to be in chemical equilibrium with groundwater and chemical components considered
important from the viewpoint of safety assessment [231] [232].
The initial pH and concentrations of Na, K, Mg, Ca, total Fe, Al, F, Cl, Br, I, total S(calculated from the values of S2-, HS-, SO42-), total P, total N (calculated from the values ofNH4+, NO3-, NO2-), total inorganic C (TIC), Si, and B were determined by taking thearithmetic mean of available data. For temperatures, an average Japanese geothermal gradientof 3 °C/100 m and surface temperature of 15 °C, result in 45 °C for plutonic and Pre-Neogene
sediments (repository depth 1,000 m), and 30 °C for Neogene sediments (repository depth
500 m). For granites, considered typical of plutonic rocks, calcite, iron hydroxide, pyrite,
smectite, sericite, chlorite and quartz were selected as the minerals that are expected to be inchemical equilibrium with the groundwater in water-conducting fractures [82] [83] [92] [116][216]. For Neogene sediments, based on the observations of target mudstones [74] [233][234], calcite, dolomite, siderite, pyrite, smectite, illite, kaolinite, chlorite, zeolite, amorphoussilica, cristobalite, quartz, and plagioclase were selected. The same set of minerals was usedalso for Pre-Neogene mudstone.
(iii) Representative groundwater chemistry
Geochemical modelling was carried out using PHREEQC ver. 3.0 [235] withthermodynamic database JAEA β-TDB ver. 1.07[236]. As noted in Figure 3.3-53 VI, the pHEh conditions for groundwater were first determined, considering the defined reference
temperature of each rock. The concentrations of Al and Si, which tend to be overestimated inanalyses due to the influence of colloids [237] [238], were determined next by assuming
chemical equilibrium with relevant minerals and then charge was balanced.
In detail, pH was determined from CO2 gas partial pressure and TIC concentration,
assuming equilibrium between groundwater and calcite. Eh was then determined assuming
chemical equilibrium between the groundwater and pyrite, while assessing consistency within-situ Eh measurements included for some groundwater chemical data. The concentrations of
Al and Si for plutonic rocks were determined assuming chemical equilibrium between the
groundwater and smectite/quartz. For Neogene and Pre-Neogene sediments, Al and Si
concentrations were determined assuming chemical equilibrium between groundwater andsmectite/cristobalite. Finally, the representative groundwater chemistry was determined bycorrecting the charge balance resulting from the above chemical equilibrium calculations withNa, the highest concentration cation in the groundwater.
The representative groundwater chemistry for each rock is summarised in Table 3.3-16.

3-82

Table 3.3-16 Groundwater chemistry for representative host rocks
Host rock

Plutonic

Groundwater

Neogene

Pre-NeogeneLowsalinityHighsalinityLowsalinityHighsalinityLowsalinityHighsalinity

T

(°C)

45

45

30

30

45

45

pH

(-)

8.16

7.56

8.38

6.54

8.15

6.33

Eh

(mV)

-301

-259
-3

1.72 × 10-282-22.77 × 10-168-32.18 × 10-289-12.82 × 10-171-3

2.19 × 10-1Na

(mol/l)3.09 × 10

K(mol/l)1.58 × 10-5

1.02 × 10-4

3.01 × 10-5

3.18 × 10-3

3.01 × 10-5

3.18 × 10-3

Mg(mol/l)8.23 × 10-6

6.17 × 10-5

1.47 × 10-5

4.95 × 10-3

1.47 × 10-5

4.95 × 10-3Ca(mol/l)3.96 × 10-4

1.60 × 10-2

2.26 × 10-4

3.45 × 10-3

2.26 × 10-4

3.45 × 10-3

Total Fe(mol/l)8.95 × 10-7

4.92 × 10-7

8.45 × 10-7

3.25 × 10-5

8.45 × 10-7

3.25 × 10-5Al(mol/l)7.91 × 10-7

2.84 × 10-7

2.20 × 10-8

1.31 × 10-9

4.25 × 10-8

2.44 × 10-9F(mol/l)5.68 × 10-4

1.26 × 10-4

1.90 × 10-4

6.49 × 10-6

1.90 × 10-4

6.49 × 10-6Cl

(mol/l)2.34 × 10-3

4.90 × 10-2

1.11 × 10-3

2.08 × 10-1

1.11 × 10-3

2.08 × 10-1Br(mol/l)4.32 × 10-6

3.88 × 10-5

4.44 × 10-6

7.98 × 10-4

4.44 × 10-6

7.98 × 10-4I(mol/l)7.88 × 10-6

5.52 × 10-6

3.94 × 10-6

1.83 × 10-4

3.94 × 10-6

1.83 × 10-4

Total S

(mol/l)7.28 × 10-6

2.03 × 10-5

1.24 × 10-4

4.09 × 10-6

1.24 × 10-4

4.09 × 10-6

Total P

(mol/l)6.46 × 10-7

5.26 × 10-6

5.62 × 10-6

5.92 × 10-6

5.62 × 10-6

5.92 × 10-6

Total N(mol/l)2.03 × 10-5

2.77 × 10-5

2.54 × 10-6

1.01 × 10-2

2.54 × 10-6

1.01 × 10-2

Total C

(mol/l)9.48 × 10-4

2.16 × 10-4

1.66 × 10-3

4.00 × 10-2

1.70 × 10-3

4.66 × 10-2Si

(mol/l)3.20 × 10-4

3.05 × 10-4

7.47 × 10-4

6.64 × 10-4

1.03 × 10-3

9.18 × 10-4B(mol/l)4.62 × 10-6

2.73 × 10-4

4.62 × 10-6

1.02 × 10-2

4.62 × 10-6

1.02 × 10-2

Ionic
strength(mol/l)

0.004

0.065

0.004

0.238

0.004

0.238

The representative groundwater chemistry datasets for low and high salinities as definedbound the ranges of deep groundwater compositions in Japan for these rocks. Furthermore,
pH values are neutral to weakly alkaline and waters are reducing, which meets therequirements for a suitable geological environment after closure (as shown in Table 3.1-2). In
addition, when defining this chemistry, geochemical modelling allows correction to thespecified ambient rock temperature. Such modelling of reference groundwaters, and also
correction of temperature-dependent data obtained from laboratory experiments, involves
uncertainties that will need to be better defined in the future.
(7) Rock thermal and mechanical properties
For each rock, information collected on a nationwide scale is utilised to define thermal
and mechanical properties required for repository design (see Section 4.5.4 (1) for details) andpost-closure safety assessment (see Sections 6.4.1 (2) and (4) for details).
In particular, average and median values for thermal conductivity, effective porosity andunconfined compressive strength were defined as representative values for each rock, basedon the data presented in the H12 report [7] and complemented by more recent measurements.

3-83

To quantify the variability of the complied thermal and mechanical data, first and third
quartile values were also shown.
For Neogene sediments, the SDM assessment leads to selection of the upper mudstonelayer as the repository host rock (e.g. Figures 3.3-34 and 3.3-38) and therefore thermal
conductivity and unconfined compressive strength required for design studies are defined forthis formation. The effective porosity is required for the evaluation of migration andretardation of radionuclides and thus averaged over all rock types. Similarly, for Pre-Neogene
sediments, the thermal conductivity and unconfined compressive strength were specified for
mudstone and sandstone, while the effective porosity was averaged over all rock types. For
Pre-Neogene sediments, the selected reference is coherent facies, as discussed in Section (4)(ii) (c). It was, however, difficult to judge from the literature information whether mudstone
data reported corresponds to coherent facies or not, and hence no distinction was madebetween coherent and mélange facies.
The thermal and mechanical property datasets are shown in Table 3.3-17. The statistical
distribution of the compiled thermal and mechanical property data together with thereferences used are summarised in Supporting Report 3-33.
Table 3.3-17 Thermal and mechanical property datasets for representative host rocks
Parameter
Number of data
Thermalconductivity(W/m K)

52

30

1.6

1.7

Median

2.8

1.6

1.9

3rd quartile

3.0

2.0

2.2

1st quartile

1.9

1.0

1.4

1,647

1,230

420

1.5

27

6.8

Median

0.8

25

3.5

3rd quartile

1.1

39

10

1st quartile

0.5

13

1.4

Number of data
Uniaxial
compressive
strength(MPa)

Pre-Neogene
sediments

2.9

Mean value

113

Neogene sediments

Mean value

Number of data
Effective
porosity(%)

Plutonic

Mean value

805

1,057

110

592

24

88

Median108

668

3rd quartile

165

20

142

1st quartile

57

3

18

The values for thermal conductivity and uniaxial compressive strength for Neogene and
Pre-Neogene sediments in the table were defined for specific lithofacies as described above.
Therefore, these values are not necessarily identical to the representative values in Table 3.3-1,
which did not specify lithofacies.

3-843.4 Probability of occurrence and impacts of natural perturbing
phenomena
In this report, as will be described in Section 6.1.5 (1), NUMO assess the possible impactof natural perturbations on post-closure safety (see Section 6.4.3 for details). This is based ona risk assessment, with decoupled consideration of the probability of occurrence of selectedscenarios (see Section 6.3.2 (3) for details) and quantitative analysis of consequences (seeSection 6.3.3 (3) for details). As previously discussed, significant impacts of such phenomena
should be avoided to the extent possible through the staged site investigation process.
However, it is important to consider these, even if they have a very low probability ofoccurrence and/or could occur only in the distant future, in order to assess if they could have asignificant impact on the safety of the repository. The scientific understanding of relevant
phenomena, as required for safety assessment, is summarised below.
Inflow of volcanic hydrothermal fluids and deep-seated fluids is not described here
because the scientific knowledge necessary to assess the likelihoods and consequences of thisprocess is not considered sufficient at this stage.
3.4.1 Volcanic/igneous activity
Magma intrusion into, and eruption through, a repository could clearly impact postclosure safety of even a properly sited and designed facility [1], even if only occurring in thefar future. As stated in Sections 3.2.2 (1) and (2), even in areas where volcanic and igneous
activity are not presently evident, NUMO excludes areas where significant volcanic andigneous activity is expected to occur over the coming tens of ky.
As discussed in Section 3.1.3 (1) (i), locations of volcanic fronts in Japan are determinedby subduction of tectonic plates and have not significantly changed over a long period of time,with volcanic activities repeatedly occurring in the back-arc regions of these fronts [16] [26].
According to the Quaternary volcano distribution map [20] [21], it can be confirmed that
volcanic and igneous activity is very unevenly distributed and has not occurred in the fore-arc
region of the volcanic front in Tohoku, as well as the Shikoku region, for at least several My.
Volcanoes are also unevenly distributed in the back-arc region of the front, with regions ofhigh and no volcanic activity clearly distinguished [26].
From the viewpoint of long-term stability of magma supply systems, areas with anextremely low probability of volcano formation over the next 100 ky can be identified,
especially in the fore-arc region of the volcanic front [239]. In the back-arc region of the
volcanic front in Tohoku, however, cases of new volcano occurrence have been identified,even when there was no record of volcanic activity during last several 100 ky [16] [240] [241].
For any particular location, the probability of volcanic activity as a function of time overthe next 100 ky has been evaluated stochastically using such as the ITM-TOPAZ method
described in Section 3.2.3 (2) [143] [167] [169] [170] [171] [172] [239]. Relevant scientificknowledge to allow assessment of magma intrusion into the repository and eruption at thesurface is summarised in Supporting Report 3-34.
3.4.2 Earthquake and fault activity
As fault movement could cause displacement within the repository or significant changes
in hydrogeology, the impacts on post-closure safety need to be assessed, given that this cannot
3-85be precluded even for a properly sited facility [1]. Although the significance may be
exaggerated for real cases, a special concern is identified as faults extending from repository
depth to the surface. Fault investigations are conducted using an interdisciplinary approachacross various fields, such as geomorphology and geology. In this section, NUMO adopts thefault definitions of the Japan Nuclear Safety Institute (2013) [242].
In the Japanese archipelago, as discussed in Section 3.1.3 (1) (iii), faults formed underearlier regional stress field may be reactivated, depending on the regional stress field during
the Quaternary or local changes to this [33] [34] [35] [36] [49] [243]. Even in cases where theexistence of faults cannot be confirmed at the surface [39], potentially active faults may still
exist underground [26] and be reactivated due to evolution of the local stress field [48].
Therefore, potential impacts of this must be assessed, even for sites with no known active
faults.
Although cases of such faults have been recorded for inland earthquakes of M6.5 or more[244], no quantitative information has been obtained on faults that extend to the surface fromdeep underground. Therefore, using the active fault database of Japan [41], the probability
that a fault will appear on the surface in the future in the repository area was calculated for
active faults [33] [34] [37] that are considered to have been repeatedly active in the Japanesearchipelago for at least the past several 100 ky. The probability was calculated very
simplistically by extrapolating the frequency of fault occurrence throughout Japan, assumingthat faults appear on the surface due to a single fault movement that took place during the past
100 ky. The resulting probability of a future fault reaching the surface from deep undergroundin a repository area is on the order of 10-7/y.
As stated in Section 3.3.3 (3) (i) (a), identified active faults with a length of ≈10 km could
generate the equivalent of a M6.5 earthquake. In addition, there is a possibility of movement
significantly affecting underground facilities [195] and hence they are excluded from therepository-scale area. However, faults 1 to 10 km in length may be included in repositoryscale areas. Based on investigations at the Mizunami URL project [92], it was considered
possible to roughly determine the position, structure and hydraulic characteristics of such
faults by surface geophysics. In offshore areas, depending on the target geological setting, it
may be difficult to identify faults by geophysical surveys with the same resolution as on land[69]. Combined with borehole surveys, however, it is possible to understand the distribution
and properties of large-scale faults. Through such investigations, faults with a length of 1 to
10 km are avoided to the extent practical in panel-scale areas.
Regarding the hydraulic and mechanical effects of earthquakes and fault movements on
the surrounding rock mass, scientific knowledge is limited, but there are cases where thepermeability of the rock near a fault increased by about a factor of 5 immediately after fault
movement and recovered to the original state several years later [245] [246]. There are also
cases in which changes in permeability due to hydraulic fracturing, fluid rise/circulation were
found in the vicinity of the fault plane (≈ several metres) [247] [248]. Changes in groundwaterlevel due to changes in the volumetric strain of the rock mass were observed over a wide area
in Japan following the extremely large Tohoku Earthquake and subsequent aftershocks [145][249] [250]. These changes recovered to the original state at many locations within a year,
and show that there are no significant effects on the long-term hydraulic gradient [251]. It has
also been shown that enhanced groundwater outflows following the Tohoku Earthquake
continued for more than four years in some cases [252]. In addition, various formulae basedon the magnitude of the earthquake, length of the fault, width of the fault zone, width of the

3-86

process zone, and the displacement of fault have been proposed [45] [253] [254] and theirapplicability in underground environments has been tested [115].
Changes in groundwater chemistry due to earthquake and fault activity are considered tobe caused by mixing of groundwater of different origins, such as water rising from deeper
underground [255] [256] [257] and drawn down oxidising surface water [258], as observed
near fault zones. Chemical changes tend to be slower than changes in groundwater level ortemperature [259]. The movement of groundwater near a fault is assumed to be caused by the
opening and blockage of microcracks and changes in water-conducting features, due tomechanical changes from fault movements [256] [258]. In overseas studies [260] [261] [262],
similar processes have been identified as factors causing short term (up to a few years)changes in groundwater chemistry due to fault activity. Following the Tohoku Earthquake,studies showed that changes in groundwater chemistry persisted for several months [116][263], thought to have been caused by a temporary change in the mixing conditions ofgroundwaters with different chemistries, due to changes in hydrology following fluctuations
in water pressure [116].
Relevant scientific understanding to assess impacts on a repository due to fault movement
are summarised in Supporting Report 3-35.
3.4.3 Uplift and erosion
Over time, the repository may gradually approach the surface due to uplift and erosion
and, even if only the thickness of overburden is reduced, the impact on post-closure safety of
the repository needs to be assessed [1].
Both uplift and erosion are influenced by regional differences between inland and coastal
areas; differences in topography and geology; differences of scale, speed and continuity ofcrustal movement; and periodic climate and sea-level changes [264]. As the time scale
increases, the magnitude of the changes gradually increase from micro-topography (natural
levees and valley bottom plains) to small size terrain features (fans and deltas) and medium
sized terrain features (mountains and basins). Therefore, crustal changes are included within
the different spatial scales in order to assess the topography constraining uplift/erosion [265][266].
For most parts of Japan, changes in ground level result from the differences between therates of uplift and erosion. For uplift, average regional rates over last 100 ky have been
documented nationwide [26]. For times greater than 100 ky, the evolution of the plate tectonic
setting needs to be considered. In the TOPAZ study (e.g. TR-16-04), a range of bounding
scenarios for such evolution are described and their impact on the rate and consequences of
uplift assessed.
For erosion, local models assess the correlation between terrain statistics (e.g.,
topographic gradient) and erosion rates [140] [267] [268] [269] [270]. In particular, using
models that can treat erosion rate variability with time, erosion rates have been determinedfrom the amount of dam sediment and the area of dam catchment over periods of several
decades. It has been shown that there is a strong correlation between derived values and the
topographic relief of the catchments [140] [267] [268] [269].
In Japan, topographic relief increases with altitude, and the erosion rate over timescales of
decades tends to increase. Since topographic relief is defined with reference to sea-level [269],
it is necessary to consider changes in erosion rate due to the effective change of elevation
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caused by sea-level changes. As noted in Section 3.1.3 (1) (iv), over the past ≈My, global
glacial cycles have a duration of about 100 ky [63]. During glacial periods, precipitation
decreases along with a decrease in air temperature and, in mountainous areas, net erosion may
also decrease (glaciation is not significant for most of Japan). On the other hand, downwarderosion advances in coastal areas as the supply of sediment from upstream decreases and thesea-level drops. The opposite occurs during interglacial periods.
For many parts of Japan, except for areas such as active fold zones within the strain
concentration zone, the uplift rate for the last 100 ky [26], the uplift rate over millions ofyears [139] [264] [271] [272] [273] and the rate of change in topographic gradient [273] canbe confirmed to show very similar trends.
In order to evaluate potential impacts on post-closure safety, available scientificknowledge on uplift and erosion rates has been collated and is summarised in Supporting
Report 3-36. Because this requires consequence analysis of scenarios describing the coupled
evolution of the EBS and the geological setting over periods > 100 ky, such assessment is notincluded in the current safety case due to limitations of the existing toolkit. However,developing the required time-dependent models and databases is identified as a priority forthe future.

3.5 Summary and future perspective
3.5.1 Summary
In this chapter, advances made since H12 in the basic concept for selecting a stablegeological environment suitable for radioactive waste disposal as well as in the required site
investigation technology were summarised. Based on the latest scientific knowledge, suitablegeological environments were identified in plutonic rocks and Neogene plus Pre-Neogene
sediments. To capture key characteristics of the geological setting, SDMs were constructed
for each of these representative host rocks to serve as the basis for repository design andsafety assessment.
NUMO, as the implementer of the geological disposal project in Japan, has developed therequired technology and stepwise site selection plan for carrying out site investigations.
NUMO has also developed approached for interpreting and integrating resulting geoscientificinformation into SDMs. By iterating synthesis of geological knowledge within the SDMs withuse of these to focus associated repository design and safety assessment, NUMO has thetechnical basis that will allow transparent and technically-justified selection of favourable
geological environments at a suitable site.
(1) Investigation and evaluation technology
Based on the latest scientific knowledge, deep geological environments with favourable
long-term characteristics from the viewpoint of radioactive waste disposal are considered to
be widely distributed in Japan. Although located in an active tectonic setting, it is possible toselect suitable environments by avoiding sites with risks of significant natural perturbations,based on an understanding of their temporal and spatial distribution.
In particular through knowledge gained in JAEA's URL programme, the basic concept
and required methodology was established for selecting suitable geological environments
during the stepwise site selection process. At each stage of site selection, in terms of
3-88confirmation of statutory requirements and siting factors, investigation/evaluation items are
identified to assure long-term stability, engineering feasibility of safe construction andoperational / post-closure safety. The multidisciplinary geoscientific information obtained
through site investigations will be assessed for consistency and integrated into SDMs. These
illustrate the geological environment, including characteristics such as the geometry of thegeological formations and structures present, their associated thermal/hydraulic/mechanicalproperties and the spatial distribution of groundwater chemistry. Based on these results, longterm evolution of the geological environment is also conceptually assessed and taken into
account in site evaluation (captured within a 4D SDM). Factors related to the geologicalenvironment having significant effects on engineering feasibility and pre-/post-closure safety
are identified, serving to focus subsequent site investigation to reduce uncertainties associatedwith them.
NUMO aims to utilise the best available characterisation technology, tailored to sitespecific requirements and boundary conditions. Therefore, available technology has beenreviewed to confirm its applicability, based on either an established track record in fields suchas resource development or construction or as a result of research and development conductedby relevant research institutions. Especially for coastal areas, where there was little
experience in relevant site investigation, advances have been made in technology forcontinuous determination of profiles of geological setting and groundwater characteristics
from land to sea, as well as methods for modelling evolution of the geological environment asa result of Quaternary crustal deformation. Through the general acquisition and analysis ofadvances in geological knowledge, NUMO is working to strengthen the technical basis for
conducting site investigations and evaluations for the diverse geological environments foundin Japan.
In conjunction with the above efforts, NUMO has been developing approaches andmethods for stochastically evaluating the likelihood of occurrence, impacts and uncertainties
of natural perturbation phenomena for periods in excess of 100 ky. In addition, to support
planning and implementation in an effective and efficient manner, documents describing basic
concepts and application manuals that capture practical experience have been produced thatcover both site investigation and associated quality management.
Thus, systematic development and demonstration of effective investigation and evaluation
technologies for the diverse geological environments in Japan are steadily progressing. By
applying these, NUMO will be able to identify the geological environments suitable forgeological disposal through the future stepwise site selection processes.
(2) Modelling of representative host rocks
An approach of developing SDMs to integrate geoscientific information arising frominvestigations in the stepwise site selection process has been illustrated. In the absence ofvolunteer sites, representative SDMs serve as the basis to demonstrate capability to bothdesign and carry out a safety assessment of a repository.
Considering the requirements and standards set by the Geological Disposal Technology
WG, potential geological environments that may result from site selection were identified,
focusing on currently identified characteristics important to determine engineering feasibility(constructability and ease of disposal) and post-closure safety (groundwater flow, nuclide
migration/retardation). Plutonic rocks together with Neogene and Pre-Neogene sediments
were selected as representative host rocks, while Neogene/Pre-Neogene volcanic and3-89metamorphic rocks that may also come into consideration were confirmed to be covered by
the assessments for plutonic rocks and Pre-Neogene sediments.
For each of the representative host rocks, the SDMs were constructed from available
nationwide geoscientific information and that derived from JAEA's URL programme. Such
SDMs capture the 3D geometry of the geological setting, the length/density/orientations offaults, fractures and other key structural elements together with their associatedhydrogeological, physical and hydrochemical properties. Geological and hydrogeologicalmodels were developed as a nested set, including regional-scale (several tens of km × tens ofkm), repository-scale (5 km × 5 km), and panel-scale (800 m × 800 m) representations. Suchdevelopment takes into account the fact that the quality and quantity of the geoscientificinformation will evolve as site investigations progress, thus the characteristics of thegeological environment were modelled assuming that finer resolution of structures will beavailable as the requirement for smaller scale representations of the SDMs arises.
Furthermore, for each rock, models were constructed on a cm-dm scale to capture thecharacteristics of the water-conducting features that determine the migration and retardationof radionuclides. Required groundwater chemistry and thermal/mechanical property datasetsfor engineering and safety assessment were derived from available literature. In particular,
Japanese groundwater chemistry data were screened in terms of rock type, sample depth,
degree of contamination by drilling fluid and/or surface water, degree of degassing and chargebalance. For groundwater chemistry data meeting quality requirements, representative
datasets were determined on the basis of assumed equilibrium between groundwater andminerals present in water-conducting features. For each rock, the representative groundwaterchemistry datasets with low and high salinity comprehensively cover the ranges of chemistryin deep groundwater in Japan.
Through the above efforts, the effectiveness of NUMO's technology for integrating thelatest scientific knowledge on relevant geological environments, and capture of this within
SDMs, has been demonstrated, providing a sound basis for the literature survey andsubsequent siting stages. For Pre-Neogene sediments (especially accretionary complex) thathave been little studied from the viewpoint of geological disposal, interpretation/integration of
the limited geoscientific information available has confirmed that favourable hydraulic,mechanical and chemical conditions for geological disposal could be found in Japan.
3.5.2 Future perspective
Future issues that should be addressed have been identified by the government, NUMO
and related research institutions; these are summarised in this report (e.g. Tables 3.2-2 and3.2-3) as well as by the Advisory Committee for Natural Resources and Energy (2014) [1]and the WG on technical issues of geological disposal below the coastal seabed [159]. In allcases, particular emphasis was placed on rigour in conducting the preliminary investigations,although not all issues need to be resolved prior to that stage.
In order to select a suitable site in Japan, it is essential to improve the reliability ofassessments of the occurrence and impacts of natural perturbing phenomena. Required
understanding can be derived from case studies for large-scale areas and for time periods thatare as long as possible. In particular, it is important to enhance scientific knowledge related to
the flow of volcanic hydrothermal and deep-seated fluids, for which understanding is weakcompared to other potential perturbations. Also assessment of land uplift and erosion willneed more careful site-specific study. For the investigation and evaluation technologies
3-90required, the scope of application of new methods and combinations of them will be expanded
and their effectiveness, accuracy and resolution confirmed through further case studies,
allowing improvement of NUMO’s characterisation toolkit.
It is also essential to improve reliability of the technology for characterising the geologicalsetting of potentially suitable sites and its associated long-term evolution, providing thegeoscientific information necessary for assessing engineering feasibility and post-closure
safety of repository concept options. Borehole drilling and investigation techniques can berationalised and optimised, particularly in terms of obtaining accurate geological information
to meet these requirements, taking into account the diversity of geological environments inJapan. Rationalising/optimising borehole drilling and associated investigation technology will
be focused by feedback from engineers and safety assessors, which will be facilitated by
improved 4D SDMs that capture site evolution, including long-term changes inhydrogeological and geochemical characteristics that impact RN release and transport.
Furthermore, specific technology for coastal areas will be enhanced, so that the completenessand quality of the resulting geoscientific knowledge base would be similar to that for land
areas. All of this will be accompanied by improvement of the quality management system andstrengthening of the technical basis for site investigations through capture of practicalexperience (“tacit knowledge”).
These issues are, as discussed in Section 2.5, reflected in the Overall Plan for Geological
Disposal R&D [274], which was compiled by the Geological Disposal Research andDevelopment Coordination Council. Table 3.5-1 summarises future efforts in the areas of:occurrence and impacts of natural perturbations, geological characteristics and their long-term
evolution and strengthening the technical basis for site investigations.

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Table 3.5-1 Future topics for selection and modelling of suitable geological environments
Topic

Occurrence andimpacts of natural
perturbations

Determination ofgeologicalcharacteristics andtheir evolution

Strengthening thetechnical basis for site
investigations

What will be done Advancement of technology for assessing the occurrence and impact of
volcanic/igneous activity
 Development of technology to improve understanding occurrence and impacts ofmovement of deep-seated fluids
 Advancement of technology for assessing the occurrence and impact ofearthquakes and fault activity
 Advancement of uplift/erosion assessment technology based on topographical andgeological information
 Further development of stochastic methods to capture probabilities andconsequences of rare phenomena, together with associated uncertainties
 Development of technology for determining hydraulic and solute transfer
characteristics of water-conducting features
 Based on potentially suitable repository concepts, assess how groundwatercomposition and its evolution may affect stability of the EBS
 Development of investigation and evaluation technologies applicable under thecoastal seabed Advancement of models of long-term evolution of geological environmental
characteristics
 Development of borehole investigation/monitoring/sealing technologies
 Improvement of the scientific knowledge base on the occurrence and impacts ofnatural perturbations
 Improvement of the scientific knowledge base on geological environmental
characteristics and their long-term evolution
 Expansion of the technical knowledge base on geological investigation andevaluation
 Expansion of the quality management system
 Accumulation of relevant practical experience

For these topics, NUMO will work together with government and relevant research
institutions to implement a coordinated R&D programme to ensure that required methodology,
technology and experience is available as and when required. This is discussed further in thefollowing sections.
(1) Occurrence and impacts of natural perturbations(i) Advanced technology for assessing the occurrence and impact of
volcanic/igneous activity
To improve assessment of the potential for new volcanic and igneous activity and its
resultant impacts, development of required technologies will provide a better understanding of
the internal structure of the crust and mantle on a regional basis as well as the potential
impacts depending on the type of volcanic activity (stratovolcanoes, calderas, or monogenetic
volcanoes).
More specifically, investigations of underground structures linked to possible future
volcanic activity will benefit from expanded seismic databases using teleseismic earthquakes,
so that the distribution and movement of fluids in the crust and mantle can be studied up to adepth of several tens of kilometres. This will improve existing models for the generation andmigration of magma. For calderas and monogentic volcanoes in particular, case studies tobetter understand the range of potential magma impacts will be conducted.
3-92For coastal waters, investigation technology used to determine the presence anddistribution of magma and other deep fluids on land (particularly seismic techniques) will be
refined so that they can be applied offshore with the same level of accuracy and spatial
resolution (applies also to (ii) below).
(ii) Development of technology to improve understanding occurrence andimpacts of movement of deep-seated fluids
For non-volcanic deep-seated fluids (deep rising water and long-term stagnant water),understanding of phenomena related to formation and movement deep underground will be
refined, and technologies for understanding the potential of flow from deep underground to
the near surface will be developed.
The geological, geophysical and geochemical knowledge required to understand theformation mechanisms and distribution of deep fluids will be accumulated, together withgeological environment characteristics related to fluid migration. Based on this, appropriatetechnologies to assess potential upwards flow of deep fluids along with associated thermal,
hydraulic and geochemical effects will be developed.
(iii) Advanced technology for assessing the occurrence and impact ofearthquakes and fault activity
To improve understanding of the hydraulic and mechanical impacts of fault activity, as
well as identifying active faults without signatures at the surface or overlying formations,investigation technology will be further developed and approaches that combine several
different methods assessed.
More specifically, to detect and characterise active faults without clear signatures,
approaches that combine geodetic methods with topographic/geological assessments,
geophysical surveys, and crustal deformation simulations will be extended. Based on
establishing a history of geological structure development through examination and chemical
analysis of minerals in fault zones, as well as the improvement in radiometric dating methods,
evaluation approaches to assess reactivation of hidden faults will be developed. In addition,
simulation techniques to evaluate the displacement of faults due to changes in the stress field,
together with the range of the hydraulic and mechanical impacts in the vicinity of the fault,
will be developed. Furthermore, by accumulating and analysing case studies on long-termchanges in water pressure and chemistry of springs and groundwater following earthquakes,
associated impacts on site properties will be investigated.
For coastal areas, especially for active faults extending from land to sea, technicalinformation from case studies to improve the accuracy and spatial resolution of theinvestigations and evaluations involved will be collected.
(iv) Uplift/erosion assessment technology
In order to improve the reliability of the assessment of future uplift and erosion,
characterisation technologies will be improved through the expansion of applicable methods
over timescales ≈ 100 ky – 1 My.
More specifically, in addition to conventional topographical methods using marine/riverterraces and dating of sediments (e.g. chronology using regional volcanic ash),3-93thermochronological methods that combine mineral closure temperatures with radiometricdating, and methods combining chemical analysis and dating of locally distributed sediments,
will be further developed.
For coastal areas, especially those where indicators are scarce, the following will beassessed: case studies on exposure dating based on the accumulation of cosmogenic nuclides
in rock erosion terraces; correlation and chronology of terraces using empirical weathering
indices for gravel layers and soils; terrain analysis using high precision digital terrain data;and extrapolation of land uplift and erosion patterns to coastal marine areas based on
topographic and geological profiles. By combining these methods, determination of uplift anderosion in both coastal land and sea areas will be improved.
(v) Development of evaluation technology related to the probability and impactof long-term natural perturbation phenomena
To improve the reliability of assessments of the likelihood of occurrence and impacts ofrelevant phenomena for periods > ≈ 100 ky, for which uncertainties are currently considered
to be large, the established TOPAZ methodology will be expanded and improved.
More specifically, considering regional characteristics in terms of future tectonic platemovement, changes in climate and sea level and the characteristics of specific perturbations interms of their development over time, NUMO will systematically develop scenarios ofpotential future perturbations over time periods up to ≈ 1 My. Even for areas selected due tolow risk of such perturbations, scoping assessment of impacts occurring only in the far future(e.g. > ≈ 100 ky) will need to be assessed. Furthermore, to assure traceability, rigorous
methods will be used to capture the expert knowledge required to stochastically quantify thephenomena modelled along with associated uncertainties. This approach will be applied, inparticular, to earthquake/fault activity and uplift/erosion assessment.
(2) Determination of geological characteristics and their evolution(i) Development of technology for determining hydraulic and solute transfer
characteristics of water-conducting features
Determining the regional groundwater flow field, including recharge and discharge areas,
together with identification of areas where groundwater remains effectively stagnant for longtimes, is essential in the PI stage. To improve the investigation approach, as well as associatedanalysis of water and solute transport, a methodology tailored to the temporal/spatial scales ofinterest will be developed.
More specifically, for the regional groundwater flow field, the validity of flow analysis
models will be tested by assessing consistency of results with characteristics that are
constrained by such flow, such as the spatial distribution of groundwater chemistry andapparent age. Furthermore, a methodology will be developed for assessing the impacts ofheterogeneity of flow/transport, e.g. due to the distribution of water-conducting features infractured rock.
For the specific case of rocks containing fossil seawater, evaluation of paleohydrogeological changes, coupled to tailored approaches combining borehole and geophysicalinvestigations, will allow us to better understand if this can be taken as evidence of long-term
stability of hydrogeochemical conditions.

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(ii) Development of investigation and evaluation technologies under the coastal
seabedAs investigations extending from coastal land to the sea are likely to be required at the PI
stage, required technology has to be developed and tested, with special emphasis on
determining hydro-geochemistry and its long-term evolution.
Although technologies for three-dimensional geophysical exploration, borehole
investigations, etc. have been applied to coastal sea areas as part of resource exploration or
academic scientific studies, their direct applicability or ease of modification to meet therequirements of site investigations for geological disposal will be evaluated. As in the case ofland areas, a focus is on the geoscientific information and quality levels necessary to assessengineering feasibility and post-closure safety. Furthermore, through case studies in which a
combination of investigation technologies are implemented and their applicability assessed,
the specific technology needed to characterise long-term stable hydraulic and chemical
conditions - including the impacts of faults on groundwater flow – will be developed.
(iii) Advanced models of long-term evolution of geological environmental
characteristics
NUMO will develop long-term evolution models (4D SDMs) that form the basis forrepository design and post-closure safety assessment. This is essential during the PI stage andnecessary to support arguments of the suitability of specific geological settings.
More specifically, the validity of the 4D models of hydraulic/chemical conditions will betested through appropriate case studies that include relevant changes in geo-environmental
characteristics, e.g. due to fault movement. NUMO will also utilise this approach to develop
credible scenarios of future evolution of hydrogeochemical conditions. Using the SDMs
already developed for the three representative host rocks, the latest scientific knowledge tobetter represent the impacts of processes such as deep penetration of oxidising waters, longterm topographic changes, and climate/sea-level changes will be captured. Through these
efforts, incorporation of increased realism into safety assessments will be facilitated.
(iv) Development of borehole investigation/monitoring/sealing technologies
As a requirement for the PI stage, NUMO will systematically develop required borehole
drilling, site investigation, long-term monitoring and sealing technology.
More specifically, drilling during the PI stage may encounter rocks in which boreholestability is an issue and hence knowledge and technology available to optimise countermeasures or responses to any problems encountered (e.g. modifying drilling mud and/or
drilling technology used) will be assured. Borehole logging/investigation/testing as well as
studies of water and core samples will be tailored to conditions experienced and technology
used to preserve hole stability, minimise perturbations and assure that characterisation goals
are reached with the required quality levels.
Monitoring technology for characteristics such as rock deformation and groundwaterpressure/chemistry in boreholes will be improved using fibre-optic technology in order towiden applicability, practicality and accuracy for long-term in-situ monitoring with timescales
up to several decades. Applicability of this technology will be tested in advance in existing
boreholes.

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There is an internationally recognised need to ensure that boreholes do not become
preferential water-conducting pathways after closure of the repository. The required
technology for recovering test equipment/casing pipes from boreholes as well as
selecting/emplacing borehole sealing materials, will be developed and field tested for relevant
geological environments, ideally within an international collaboration framework.
(3) Strengthening the technical basis for site investigations
To assure effective and reliable site investigations can be carried out, required technicalinfrastructure will be established, e.g. by compiling relevant scientific knowledge, building on
focused R&D, and further development of a tailored quality management system.
More specifically, NUMO will prioritise expanding scientific knowledge on Pre-Neogene
sediments, particularly within accretionary structures, due to the limited information basecurrently available. Bearing in mind that site investigations are both complex and highly
sensitive, capture of direct experience in geological characterisation and evaluating long-term
evolution, as discussed in Section 3.5.2 (2) above, is a priority – e.g. taking advantage of theresources developed in JAEA’s URL programmes. Priority is also placed on experience ingeological investigations in relevant land and coastal sea areas, including planning andapplication in terms of methodology, effectiveness and other technical issues. Furthermore,
effective quality management and data management systems are needed, based on thosealready applied to site investigations and tailored to improve effectiveness based on practicalexperience gained by applying them to borehole investigations, as discussed in Section 3.5.2(2) (iv) above.

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Supporting Reports (SRs)
SR 3-1

Review of conclusions on the geological environment in Japan provided by the H12
ReportSR 3-2 Evaluation of the likelihood of natural perturbations (basic concepts)
SR 3-3 Evaluation of the likelihood of natural perturbations (volcanic and igneous activity)
SR 3-4 Evaluation of the likelihood of natural perturbations (earthquake and fault activity)
SR 3-5 Evaluation of the likelihood of natural perturbations (uplift and erosion)
SR 3-6 Evaluation of likelihood of natural perturbations (combinations of investigation
techniques)SR 3-7 Characterisation of geological environments (basic concepts)
SR 3-8 Characterisation of geological environments (information flow)
SR 3-9 Characterisation of geological environments (combinations of investigation
techniques)SR 3-10 Evaluation of long-term evolution of geological environments
SR 3-11 Summary of investigation techniques
SR 3-12 Quality management during site characterisation and SDM development
SR 3-13 Identification of potential host rocks
SR 3-14 SDM development for plutonic rocks (creation of dataset)
SR 3-15 SDM development for plutonic rocks (setting areas and conceptual models)SR 3-16 SDM development for Neogene sedimentary rocks (creation of dataset)
SR 3-17 SDM development for Neogene sedimentary rocks (setting areas and conceptualmodels)SR 3-18 SDM development for Pre-Neogene sedimentary rocks (creation of dataset)
SR 3-19 SDM development for Pre-Neogene sedimentary rocks (setting areas and conceptualmodels)SR 3-20 SDM development for plutonic rocks (regional scale model)
SR 3-21 SDM development for Neogene sedimentary rocks (regional scale model)
SR 3-22 SDM development for Pre-Neogene sedimentary rocks (regional scale model)
SR 3-23 SDM development for plutonic rocks (repository scale model)
SR 3-24 SDM development for Neogene sedimentary rocks (repository scale model)
SR 3-25 SDM development for Pre-Neogene sedimentary rocks (repository scale model)
SR 3-26 SDM development for plutonic rocks (panel scale model)
SR 3-27 SDM development for Neogene sedimentary rocks (panel scale model)
SR 3-28 SDM development for Pre-Neogene sedimentary rocks (panel scale model)
SR 3-29 SDM development for plutonic rocks (water-conducting microstructure model)
SR 3-30 SDM development for Neogene sedimentary rocks (water-conducting
microstructure model)
SR 3-31 SDM development for Pre-Neogene sedimentary rocks (water-conducting
microstructure model)
SR 3-32 Modelling of groundwater chemistry of representative potential host rock settings
SR 3-33 Modelling of thermal and mechanical properties of representative potential hostrock settings
SR 3-34 Likelihood and potential impacts of future natural perturbations (magma intrusionand eruption)

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SR 3-35 Likelihood and potential impacts of future natural perturbations (earthquake andfault activity)
SR 3-36 Likelihood and potential impacts of future natural perturbations (uplift and erosion)

3-98

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[10] Chapman, N. and McCombie, C. (2003): Principles and standards for the disposal of
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3-124

4 REPOSITORY DESIGN AND ENGINEERINGTECHNOLOGY
Based on the basic concept of repository design described in Section 2.3, this chapter will
present the following:

Repository design specifications to demonstrate the engineering practicality of
construction, operation and closure of facilities tailored to the three representative
host-rock site descriptive models (SDMs) presented in Chapter 3, which will provide
the basis for assessing safety, both pre and post-closure.
Illustration of a practical design methodology that can respond to the geologicalenvironments expected to be encountered at volunteer sites, such as the presence offaults, and determine pros and cons of different disposal concepts.
Analysis of such designs (see Section 4.2.4 (ii)) to show that they have the potential
to provide the required safety functions for specific SDMs and that they can be
implemented using engineering technology available at present, or that which will be
realised in the near future. However, the impact on the safety functions if conditions
change, e.g. due to natural perturbations, can only be assessed as part of the scenario
analysis presented in Chapter 6.

In Section 4.1, the development of repository concepts using a structured methodology is
described, based on design factors to ensure flexibility for the variety of geologicalenvironments and social boundary conditions. Section 4.2 explains the basic design procedure
for the repository, based on the characteristics of the waste and the SDMs, together withrequired pre- and post-closure safety functions. In Section 4.3, the repository disposal depth is
set while, in Section 4.4, the design of the engineered barriers is described. This leads tospecification and illustration of underground facilities in Section 4.5, and surface facilities in
Section 4.6, which provide the required safety functions. Technology related to construction,
operation and closure of these facilities is also discussed as a basis for subsequent safetyassessment. Section 4.7 assesses the engineering practicality of waste retrieval during theperiod until acceptance of the safety case is confirmed, prior to closure of the repository.
Finally, Section 4.8 summarises the chapter and outlines future activities in this field.

4.1 Approach to development of repository concepts
The purpose of repository design is to develop specifications of the facilities required toensure safe waste disposal during operation of the repository and ensure passive safety
thereafter without any active institutional control. With this aim, design technology
development is based on the following thinking [1]:In order to be able to focus design to both assure safety and also engineering
practicality during the process of stepwise site selection, the multifaceted repository
requirements are captured as “design factors”.
Disposal concepts and design options are developed in such a way that it is possible
to respond flexibly to the diverse geological environments that may result from thesiting process, together with progress in science and technology during the longproject implementation period (in the order of a century).
In accordance with the stepwise narrowing of the scope of investigation and therefinement of the geological knowledge base, the repository will be designed using
SDM information and data appropriate to the spatial scale involved.
4-1

In this section, a design methodology developed on the basis of these ideas is described.
Repository design will be optimised in steps; starting from a conceptual design, followed bya basic design and finally a detailed design that will be tailored to developing understandingof the site geological environment.
The conceptual design is a first repository outline based on the SDM resulting from surface
exploration or boreholes included in the preliminary investigations (PI), but design work
actually starts during the literature survey (LS) phase. Based on the conceptual design, safetyassessment will be carried out to confirm the practicality of geological disposal, e.g. based on
the mechanical and hydraulic properties at relevant depths underground. The design options
developed will be narrowed down for the demonstration of engineering practicality in the nextstep.
An underground investigation facility (UIF) will be constructed in the target host rock(s)during the DI phase to confirm important properties, such as mechanical, hydraulic, thermaland chemical suitability. In addition, construction and operational technologies will bedemonstrated in the UIF (although, due to time constraints, initial development work will be
carried out in existing URLs). The basic design will be tailored to the updated SDM, with
illustration of feasible engineering of repository construction and operation. The safety casewill be developed for this basic design and applied to the licence application.
After passing the licence application, the basic design will be used as a starting point,
allowing detailed design of the equipment related to construction and operation to be carriedout, with final design of the repository beginning after construction approval. In this way,
throughout the site selection process, the repository is designed in an iterative manner, with
optimisation on the basis of continual assessment of safety and practicality.
In the H12 report, examples of 2 conceptual disposal designs for HLW (in-hole vertical andin-tunnel horizontal (H12V/H12H) applicable to a wide range of geological settings in Japan
were presented, with limited consideration of site environmental characteristics. As further
discussed in Section 4.2.3, the selection of the design concept should originate from thedefined requirements. However, for the current version of the safety case, focus is on alreadyrelatively well-assessed concepts, such as presented in the H12 report for HLW and the TRU2 report for TRU waste, in order to show how requirements are used to further specify thedesign. Considering that the current stage is prior to starting the LS, the three representative
SDMs developed in Chapter 3 were used. The repository was designed with the objective ofshowing how it can be adapted to site conditions. Therefore, the amount of waste handled andthe operational methods are set provisionally, with equipment used for construction andoperation illustrative only. In the future, more detailed design work will be conducted inparallel, to confirm fundamental practicality through demonstration tests.
4.1.1 Capturing requirements in design factors
Based on nuclear safety regulations, since 2005, repository designs are required to be inaccordance with a performance code instead of the specification code previously used (see“Ministry Ordinance No. 62: order that establishes technical standards for power generation
nuclear facilities”). The performance code obliges NUMO to utilise best available technology(BAT) during the development of the repository over the next few decades. It is worth notingthat such flexibility of design methodology might facilitate tailoring the repository design tothe site.
4-2

“Design factors” provide a consistent focus for repository design that is applicable todifferent sites and different stages of project development [2]. The following issues are
considered when setting the design factors:

Those concerning the safety of the repository, as indicated by the regulations ofinternational organisations for related nuclear facilities.
Relevant Japanese policy, such as the Basic Policy on Final Disposal of DesignatedRadioactive Wastes (Basic Policy on Final Disposal – see Chapter 1).
Issues pertaining to the practicality and economics of the design, construction andoperation of repositories.

In this chapter, it was decided to consider five groups of requirements that constrain design
factors: operational safety, long-term safety after closure, retrievability, engineering
practicality and socio-economic issues1. The resulting design factors to meet theserequirements are shown in Table 4.1-1.
Table 4.1-1 Design factors and related requirements considered in this report
Design factor

OperationalsafetyPost-closure
safetyRetrievability

Engineering
practicality
Socio-economic
issuesSpecific requirements







Prevention of leakage of radionuclides (RNs) from waste
Prevention of release of RNs from the repository
Radiation shielding
Prevention of occurrence and propagation of incidents
Evacuation routes established in case of accidents
Maintain conditions appropriate to worker health and safetyProtection from significant effects of natural perturbing
phenomena
Reduction of the likelihood of human intrusion
Restriction of RN leaching
Restrict of RN migration







Maintain practicality of retrieval of waste
Avoid any reduction of safety margins as a result of maintaining
ease of retrieval

Fundamental practicality of construction, operation and closure of
all facilities
Application of proven technology

Cost-effectiveness of construction, operation and closure of
repository
Assured ability to procure all support materials/services required

Among the design factors, requirements in terms of operational safety functions are definedin relevant laws and regulations applicable to industrial and nuclear facilities, as summarisedin Tables 2.1-3 and 2.1-4 in terms of radiation protection and general occupational safety,
respectively. Post-closure safety requirements for geological disposal are derived from IAEA
[3] and a report on safety regulations [4], which are summarised in Table 2.1-5 in terms ofcontainment and isolation safety functions. Further, Section 2.1.3 (2) discusses assuring
practicality of waste retrieval in order to provide options for future generations to reconsider
1

Additional factors such as environmental conservation are not currently considered in this report, as theserequire more specific information about the geographical, geological and social conditions of the site.
Environmental monitoring of the impacts of repository implementation is, however, considered in Chapter 2 andwill form the basis of developing environmental design factors to be used in the future.

4-3

this choice of disposal method and hence “retrievability” was set as a design factor in NUMO
(2004) [2]. Finally, “engineering practicality” and “socio-economic issues” are set in order toassure that designs developed can actually be implemented based on real world constraints,
reflecting NUMO’s role as an implementer and the need to develop and manage a viable
project. Hence this is a significant development from the previous studies undertaken by R&D
organisations in Japan.
4.1.2 Setting and utilising design requirements
As shown in Figure 4.1-1, design factors are key to assuring the required performance ofthe repository, defining requirements for each component to allow specifications to be derived.
(Procedure of design requirement identification)
Design factor
Performance requirements of repository(Design procedure)
Fundamental requirements for repository
Requirements derived from design factors of pre- andpost-closure safety, practicality, etc.

Setting of repository components

Design requirements for components
Mandatory requirements and preferred requirements

Setting of specifications
Feedback asrequired

Evaluation items*

Evaluation of impacts and considerationof counter-measurese.g. Evaluation of long-term performance

Figure 4.1-1 Role of design factors in the design process

Design requirements can be categorised as those that must be fulfilled to ensure therequired performance (mandatory requirements) and those to further improve the performance(preferred requirements). Design is then performed using these requirements, and basic
specifications (materials, shapes, dimensions, etc.) are set.
For specifications designed to satisfy these requirements, evaluation items are selected by
identifying the influence factors that can lead to loss of function and/or decrease in long-term
performance before or after repository closure. The influence factors can be classified in tothermal, hydrogeological, mechanical and chemical impacts over the long-term (which may
be effectively permanent), as assessed and described in this chapter. Chemical behaviour
which may directly impact RN migration in the near-field is assessed in Chapter 6 (see Table
6.3-1). If these impacts are significant, either, counter-measures will be implemented, or
specifications for the repository component will be iteratively optimised to mitigate theimpact.
For quality assurance of design work, it is necessary to ensure the models and analysis
codes used are adequately tested, meet their defined specifications and conform torequirements to manage information properly. The codes used in this work are listed in
Supporting Report 4-1.

4-4

4.1.3 Development of design options
It is not necessary to narrow down to a single reference design at present, especiallyconsidering the diverse geological environments found in Japan. Rather, in this stage without
a specific site, maintaining a range of designs that meet the requirements of the design factors
allows flexibility to tailor design to site characteristics.
Even if technology is considered to have poorly-demonstrated engineering practicality atpresent, consideration of an extensive range of design options allows identification of
appropriate technology that can be developed further in the future. As a result, it is considered
possible to adapt designs to reflect progress in science and technology and improve resultant
performance. This idea, focusing on BAT, is consistent with international guidance on safe
implementation of geological disposal [5].
Based on this approach, NUMO and related research institutes have been developing a
number of design options. Specifically, H12 [6] provides a starting point for design
specification of the engineered barriers for HLW, together with the layout of the emplacement
and other underground facilities. This base has been expanded further by NUMO andorganised as a “repository concept catalogue” [7]. Several novel design options presented inthis catalogue, following subsequent technical development, have been shown to be based onpractical and promising technology. Based on progress in development of these design
options, repository concepts that form the basis of designs described in this report are selected(see Section 4.2.3).
Figure 4.1-2 classifies, in general terms, many of the repository concepts examined in Japanand abroad since H12. For example, most effort has been on options to place the waste intunnels, vaults or holes accessed by shafts and/or ramps.
Categorisation based onunderground accessGeologicaldisposalTunneldisposalCategorisation based on the heatgeneration/emplacement geometry

Distributed

Vertical emplacement

Applied to the HLW repository
as a design option

Horizontal
emplacement

Applied to the HLW repository
as a design option

Applied to the TRU waste
repository as a design option

Concentrated(e.g. in-vault)

Not applied

Deep borehole disposal

Figure 4.1-2 Classification of repository concepts

4-5Overseas deep borehole disposal2 has been discussed but, amongst other things, it isparticularly unclear whether this concept would be practical for Japan’s large waste inventory.
Tunnel-type disposal can be further classified as either distributed or concentrated, dependingon the density of waste emplacement. In order to reduce the temperature rise in the host rock,a distributed concept is generally applied to waste with high thermal output – such as HLW.
In this study two variants of this are considered, horizontal in-tunnel emplacement andvertical emplacement in holes in the tunnel floor. It is clearly recognised that these concepts
can be further optimised and that alternate concepts might be applicable in Japan.
To increase efficiency of operations and use of underground space, concentrated disposal
concepts are generally applied to waste with lower heat output, such as TRU waste, as
discussed further in the TRU-2 report [10] and adopted for our studies in this chapter.

2
Deep borehole disposal (DBD) involves waste emplacement in the lower parts of holes drilled from the surface [8]. Likeother forms of geological disposal, safety is provided in terms of forming multiple barrier systems with engineered barriers
and natural rock. Most countries planning geological disposal, including Japan, have considered tunnel-type disposal.
However, in recent years, DBD has been studied as an alternative concept, mainly in the United States. While this disposal
method can minimise underground excavation, it is considered that there are problems such as restriction of the borehole sizeand difficulty of recovery [9].

4-6

4.2 Repository design procedures and boundary conditions
4.2.1 Repository design procedures
The repository design work flow based on requirements was introduced in Section 4.1.2
and is shown in Figure 4.2-1. This requires specification of the characteristics of waste
assigned to geological disposal, as presented in Chapter 2, and the details of the representative
SDMs, presented in Chapter 3.
Repository Design
Design premises
Identification ofwaste characteristics
Requirements identification
Site descriptive model
based on site investigationand evaluation

Selection of repository concept andfundamental components
Definition of design requirements
and evaluation items

Decision on disposal depth

Design of EBS

Design of the surface facilities

Feedback asrequired

Design of the underground facilities

Feedback to SDM development

Pre andPost-closure safety assessment

Figure 4.2-1 Repository design work flow considering technical feedback loops. Some aspects(such as the potential revision of requirements due to interests of stakeholders) are notdirectly indicatedBased on the requirements indicated in Table 4.1-1, the components of appropriate disposal
concepts and associated facilities (engineered barriers, surface facilities, undergroundfacilities) can be developed. In particular, the requirements in Table 4.1-1 for pre- and postclosure safety (see also Section 2.1.2) are allocated to specific components on the basis ofsimilar facility designs in Japan and overseas. From this background, the disposal depth is set
with reference to specific SDMs, based on design requirements.
In the design process, it is normal practice to allow for a margin of safety to take into
account the uncertainties in the geological characteristics presented in the SDMs. In addition,
there may be other uncertainties to consider, such as those associated with the coupled
thermal, hydraulic, mechanical, chemical (THMC) processes in the buffer. However, safetymargins need to be set rationally to ensure practicality and prevent complications caused byover-design, so this uncertainty should be reduced in the design process, providing feedback
4-7

to improve the focus of geological characterisation. Further, if subsequent pre- or post-closure
safety assessment of the preliminary repository design indicates requirements cannot be metor safety margins demonstrated, such feedback helps focus design improvement.
At this time, as sites are not yet specified, it is possible only to provide guidance forrepresentative potential siting environments. Nevertheless, this exercise will demonstrate ourcapability to construct appropriate repository concepts for these. The associated guidance forrational repository design involves:Incorporating a common set of fundamental engineered barriers that are applicable to arange of geological environments, to which detailed specifications can be tailored on asite-specific basis.
Underground facilities are specified for the geological structures and physicalproperties of the rock set on the basis of the SDMs.
Surface facilities, including waste reception, inspection and encapsulation plants, aredesigned with an emphasis on work flow and radiation protection during theoperational period, which have a low sensitivity to the site geological environment.

In the following, boundary conditions for design, waste and geological characteristics thatset requirements for specific repository components are presented.
4.2.2 Waste characteristics
Specified waste for deep geological disposal includes HLW and TRU waste. This section
presents quantities of such waste and other characteristics relevant to design, such as heat
output and radioactivity.
(1) HLW
As mentioned in Section 2.1.1 (1), although HLW produced by different manufacturers has
different specifications, for the design of the repository, the specifications for HLW produced
by JNFL is taken as a reference. The final disposal plan specifies a requirement to dispose ofat least 40,000 HLW canisters (see Section 2.1.1 (4)), with this minimum value taken as thereference for design purposes.
High-level radioactive liquid waste is calcined, vitrified with a borosilicate glass frit andpoured into a cylindrical stainless-steel canister, with an outer diameter of 0.43 m and a heightof 1.34 m. Assessments of the radioactivity inventory and heat output of HLW are given in
Supporting Report 2-2. Heat output of a HLW canister, in the case of a four-year spent fuel(SF) cooling period before reprocessing and a period of 30 or 50 years storage before disposal,
is about 560 W or 350 W respectively, with total radioactivity being about 4 × 1015 Bq and 2
× 1015 Bq at these times. About 1 ky after disposal, the total radioactivity is around 1013
Bq/HLW canister.
As mentioned in Section 2.1.1 (1), interim storage time after HLW production is not yet
specified but is expected to lie in the range of between 30 and 50 years. Therefore, in thisreport, in order to highlight and quantify differences in potential representative host rocks,
designs were developed for storage periods of both 50 and 30 years (with the latter being
hotter and more radioactive). This will, however, be reassessed in the future when theinventory in general, and this parameter in particular, is better defined.
4-8

To assess operational safety functions (see Section 4.2.4 (1)), including those above andbelow ground during waste handling, encapsulation, transport and emplacement, shielding
and dose calculations are conservatively based on the radioactivity inventory of HLW stored
for 30 years.
Post-closure safety functions (see Section 4.2.4 (2)) are important in the design ofengineered barriers, such as overpack or buffer material specifications and the undergroundlayout of the disposal tunnels. For this purpose, it is necessary to set the radioactive inventoryof the HLW and its associated thermal characteristics. Generally speaking, for shorter preemplacement storage, the footprint of the repository and excavation volume become larger,
also increasing repository-driven perturbations – e.g. of hydrogeological conditions. To
ensure safety after closure, preferred characteristics of the geological environment (Section
3.1.3 (2)) should be disturbed as little as possible, thus the inventory of HLW with a storage
period of 50 years was considered as the design base. Nevertheless, it is technically possible
to design underground facilities for HLW with a storage period of 30 years, e.g. by increasing
temperature limits on materials, decreasing the disposal depth, etc. In the future, technology toallow design optimisation considering the variability of the inventory due to differences ininterim storage time will be developed.
(2) TRU waste
TRU waste is inherently very heterogeneous, with the defined groups containing different
waste streams from different producers, a significant quantity of which will arise only in the
future. For safety assessment in this report, representative package specifications are definedfor each group. In the future, however, the ranges of waste properties will be considered in
more detail.
TRU waste form (drum, canister, box, etc.), size, and weight, were described in Section
2.1.1 (2) based on the TRU-2 report [10] and NUMO (2011) [11]. As for HLW, the interim
storage time between waste production and disposal is not defined for TRU waste. The TRU2 design and recent safety assessment reports [11] [12] used a thermal and radioactivity
inventory at 25 years after production, which is also assumed here.
In the TRU-2 report, Group (Gr.) 1 waste is assumed to be grouted into 200 l drums. This
waste contains a large inventory of radioiodine (I-129), which can dominate impacts due to its
generally low sorption and high solubility, and led to work at relevant research institutes toreduce its leach rate by development of alternative solidification technology options. As aresult, various alternative conditioning processes have been shown to have the potential toreduce the leaching rate over a period of about 100 ky [13]. However, full-scale production
technology and assured performance under relevant geological conditions have yet to be
demonstrated for these approaches. Thus, this report conservatively assumes cement
conditioning of Gr.1 waste.
In addition, also for waste Grs.2, 3, 4L and 4H, heat output, radioactivity, volume, weight,
etc. were set to be conservative for design purposes (data summarised in Supporting Report 24).
4.2.3 Disposal concepts and components
Basically, the selection of the design concept should originate from the definedrequirements. However, for the current version of the safety case, focus has been on already
4-9

relatively well-assessed concepts, such as presented in the H12 report for HLW and the TRU2 report for TRU waste, in order to show how requirements are used to further specify thedesign. A more open and general approach, also considering other potential concepts, will beapplied in the future, especially if it turns out that the geological environments at potentialsites significantly deviate from the SDMs considered in this report. It is also recognised that
more optimal concepts might be derived, even within the range of siting environments studied
in this assessment.
(1) Setting of the basic disposal concept
Of the general geological disposal options indicated in Figure 4.1-2, requirements such asproven technology, ease of quality control, retrievability, etc. lead to selection of a mined
repository concept, as illustrated in the H12 report. For heat-emitting HLW, distributed
disposal is selected, with options of vertical or horizontal emplacement considered. Lower
heat TRU waste can be more efficiently emplaced using a concentrated emplacement concept.
As mentioned in Section 2.1.1 (5), in this report co-location of disposal facilities for HLW
and TRU waste is assumed. Such co-disposal has clear benefits in terms of efficiency of
implementation due to sharing site survey results and some surface facilities, provided theavailable site would allow the resulting footprint. However, it is necessary to keep any
interactions between the two disposal facilities as small as possible [10] [11]. In the following,
the engineered barriers, underground and surface facilities are discussed based on the disposal
concepts set out above.
(2) Engineered Barrier System (EBS)(i) EBS for HLW
The EBS variants considered in this report are all taken from existing designs and illustrated
in Figure 4.2-2. Their derivation from requirements is discussed in following sections of thischapter. The EBS comprises multiple barriers, including the borosilicate glass matrix (and thefabrication canister within which it is contained, although this is conservatively ignored), theoverpack and the buffer.
The overpack may be made from carbon steel, copper or titanium. Metal overpacks are
considered in this assessment because they are commonly used in many industries and much
experience in container manufacturing exists. In addition, metal overpacks are also chosen as
the reference material for waste containers by almost all international implementers ofgeological disposal of HLW or SF.
Bentonite, which contains high concentrations of swelling clay minerals (e.g. smectite), has
the capacity to provide physical and chemical buffering in the repository [6] and is apromising buffer material. Alternative materials, such as mortar or resin, could be also
considered, but they may require additional assessment to ensure all of the expected roles forthe buffer are provided. Bentonite is therefore chosen by most repository implementers.
For vertical in-hole emplacement (H12V), the option of in-situ buffer construction using
compacted bentonite blocks is selected on the basis of wide international experience [14]although other options have been studied in Japan [15]. For horizontal in-tunnel emplacement,a PEM (Prefabricated Engineered barrier system Module) option is selected to ensure ease of

4-10

emplacement of the bentonite buffer with the required quality by remote handling (even under
wet conditions).

(a) H12V method(b) PEM method
Figure 4.2-2 HLW engineered barriers

The PEM is similar in principle to the KBS-3H concept [16], which Sweden and Finland
are examining. This allows buffer construction within the handling shell on the surface, whichis advantageous in that quality control is easier and problems such as swelling of the bentonite
underground due to dripping water or high humidity can be avoided. Furthermore, as
described in Section 4.4.3 (1) (ii), PEMs have been fabricated at full scale andtransport/emplacement tests have been carried out.
The focus in this report is thus H12V using bentonite blocks and the PEM, but other buffer
emplacement options exist, for example, methods utilising horizontal pellet injection or insitu compaction [1] [17], together with other waste package emplacement approaches.
(ii) EBS for TRU waste
The EBS for TRU waste is illustrated in Figure 4.2-3. As for HLW, the EBS comprises
multiple barriers including: the TRU waste matrix and primary container (waste form), waste
package and infill, concrete structures and, in some cases, a bentonite buffer. However, since
the characteristics of the waste groups are different, the EBS is tailored to each of these. The
waste package manufactured for disposal encapsulates primary containers holding the waste(waste form), which are surrounded by infill, as also shown in Figure 4.2-3: this is the unit
considered for design of disposal vaults.

4-11

Vault support

Backfill

Waste package container

Waste package

Concrete cell

Bentonitebuffer

Disposal vault

InfillWaste form

Figure 4.2-3 TRU waste engineered barriers

In the TRU-2 report, a waste package (containing the waste form) fabricated from steel
with a mortar infill (termed waste package A) was illustrated; this was specifically designed
to facilitate stacking with a forklift in the disposal vault. In the same report, the concept of
emplacement using gantry cranes was also illustrated as a design option, but a problem
identified was to reduce the risk of waste package drops during handling. However,
emplacement by cranes is commonly used in low-level radioactive waste repositories in Japan[18] and overseas [19].
Since the TRU-2 report, the use of high strength/high density concrete, potential use of atitanium container for C-14 containing waste [20] and waste packages fitted with a lid (termed
waste package B) were amongst the variant design options considered [11] [20]. Thus, thedevelopment of container technology is progressing, with standard [21] rectangular steel
containers established for low-level radioactive waste disposal. In this report, the TRU-2
waste package A (emplaced with a forklift) and in addition, a more robust waste package B(emplaced by a crane) are both considered as options.
(3) Components of underground facilities
Figure 4.2-4 illustrates key components of underground facilities, including access ramps
and shafts, disposal tunnels for HLW and vaults for TRU and associated undergroundconnecting tunnels.
Repository closure structures (e.g. mechanical plugs and hydraulic plugs) are also shown.
For TRU waste in particular, extensive infrastructure is required in the disposal vaults fortransporting/emplacing waste packages and subsequent infilling/backfilling. Transport andemplacement equipment is also required for HLW (H12V or PEM), together with services forthe entire repository, including ventilation, drainage, lighting, communication, etc. All suchcomponents have to be included in site-specific designs.

4-12Figure 4.2-4 Sketch of key underground facilities. Note that the number of shafts and accessramps is only illustrative. In the future, efforts will be made to minimise the number of accesspoints(4) Components of surface facilities
The surface facilities provide the general services required for the practical implementationof this major industrial project, nuclear standard buildings for waste handling andencapsulation operations and the technical services required to support underground
operations. For the case of co-located HLW and TRU waste disposal, such facilities areillustrated schematically in Figure 4.2-5.
The surface facilities consist of buildings to receive waste transported from off-site, inspectand encapsulate the HLW into the overpack and receive, inspect, encapsulate, and package
the TRU waste. Also included are plants to provide ventilation, drainage and water treatment
required for the construction, operation and closure of the underground facilities; a power
substation (including backup power supply); an excavated spoil storage yard; and a buffer
manufacturing facility. The configuration of facilities will depend on the site, potentially
including port facilities and dedicated roads for receiving waste at a coastal site.

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Figure 4.2-5 Sketch of key surface facilities

4.2.4 Relationship between safety functions and repository components
The repository design sets the design factors for post-closure and operational safety as a toppriority (See Section 4.1.1). These requirements are referred to as safety functions (Section
2.1.2). Their relationship to the basic structure of a repository is outlined below.
(1) Relationship of operational safety functions and repository components
Operational safety considers risks to both staff and the general public during construction,
operation and final closure processes. Here, after outlining operational processes, thecomponents that are responsible for the safety functions of radiation protection and general
occupational safety will be assessed.
(i) Operational processes
Repository operation can be roughly classified into the processes on the surface and thoseunderground. These are further broken down into seven main sequential steps for operation
for the H12V concept, as schematically shown in Figure 4.2-6 and discussed below.

4-14Figure 4.2-6 Repository operational steps (H12V case)In the PEM case, an additional process of “assembling the PEM” is added between theprocesses of “encapsulation and testing” and “transporter loading”, with the PEM loaded onto
the transport vehicle instead of the overpack. For TRU waste disposal, the waste package andemplacement operations are different, but the seven steps are fundamentally the same, as also
described below.
(a) Transportation to surface facilities
After interim storage for 30 to 50 years, HLW is inspected to check that characteristics meet
specifications (e.g. in terms of heat output, radioactivity, waste canister characteristics) andthen placed in a dedicated transportation container (cask) for transport to the repository. If it istransported by sea in a specially designed ship, the transport cask will be off-loaded at a
dedicated port near the repository site and, after surface dose measurement and inspection of
appearance, taken to the surface facilities on a special transport vehicle travelling on a private
road. A similar process applies to TRU waste after 25 years interim storage.
(b) Reception, inspection and buffer storage
HLW/TRU waste is received at the surface facility in transport casks and placed in a
temporary store which may act as a buffer store from a logistical viewpoint. Afterconfirmation of identity and specifications, the transport cask is moved into a radiationcontrolled zone and unloaded by overhead crane. Waste canisters/containers are inspected tocheck for abnormalities, such as damage to the waste during transport. Accepted waste is thenplaced in a buffer store. All operations – lifting by cranes and movement on rail transporters –are tele-operated or automated, designed to assure the required throughput while avoiding
presence of staff in these controlled areas. Procedures for waste that does not meet theacceptance criteria have not yet been developed, but will be addressed in future R&D.
4-15

(c) Encapsulation of HLW in overpack/packaging of TRU waste
HLW canisters taken from the buffer store are placed into overpacks while TRU waste
containers are grouted into emplacement packages. After lid sealing (currently planned to bedone by welding), the overpacks/packages are subjected to non-destructive testing (e.g. using
ultrasonic methods) and, after acceptance, placed in buffer stores. In the case of the PEM
concept, HLW overpacks are placed into steel handling shells that are filled with buffer,sealed, tested and placed in another buffer store. Again, all processes are tele-operated or
automated and plans for any overpacks/waste packages that do not pass testing have not yet
been developed. However, procedures for this eventuality will be addressed in future R&D.
(d) Loading onto the underground transport vehicle
Overpacks/PEMs/TRU waste packages are taken from the buffer store and, after a check ofsurface contamination, moved out of the main surface facility to a transport vehicle in theaccess portal to the disposal facility radiation-controlled zone(e) Transportation underground
The overpack/PEM/waste package is transported underground via the access ramp by aspecially designed vehicle. At the bottom of the ramp, HLW is either placed in a buffer store
or directly transferred to an emplacement machine for transport to the disposal tunnel, while
TRU waste is taken directly to the waste group specific disposal vault on the transporter.
(f) Waste emplacement
HLW overpacks/PEMs are moved through connecting tunnels to the disposal tunnel for
vertical/horizontal emplacement, respectively, by specially designed machines that are teleoperated or automated. For H12V, emplacement of the overpack in the disposal hole iscoordinated with placing the blocks of compacted buffer around it, while the PEM is simply
placed on a plinth on the disposal tunnel floor.
In the case of TRU waste, at the entrance to the disposal vault, packages are removed fromthe transport vehicle with either a forklift (in the case of the waste package A) or gantry crane(in the case of the waste package B) and moved to a pre-determined position inside using
remote-handling technology.
(g) Backfilling and sealing
After waste emplacement, tunnels/vaults are backfilled and mechanical plugs emplaced toseal them. The operations involved depend on the waste type and disposal option, as outlinedin the operational description, although details will depend on assessments of quality
requirements and practicality that are ongoing at present.
(ii) Safety functions and components
Operational safety functions should take into account both expected repository operating
conditions (termed the normal state) and possible deviations from these (termed abnormal
states). In the normal state, the main components that contribute to radiation protection are
4-16

shown in Table 4.2-1, subdivided into those that provide RN containment and thosecontributing to radiation shielding.
Table 4.2-1 Relationship of the operational safety functions and repository components(radiation protection) (modified from NUMO, 2011 [1])Basic concept

Containment
duringoperation

Safety function

Component

Prevention of leakage of
RNs from waste

HLW matrix, fabrication canister, overpack, (PEM)
TRU Waste matrix, container (drum, canister), TRU
waste package, transport cask

Prevention of release of
RNs from the repository

Nuclear-standard surface facilities operating below
atmospheric pressure, ventilation system (including
exhaust filters), controlled drainage
Waste package self-shielding

Radiation
shielding

Reduction of radiation
dose

Shielding walls, layout of surface facilities, locating
activities below surface, access control, remotehandling
Transport shielding and shielding provided bytransport vehicles

Containment of RNs during operation is assured by the waste matrix, the primary container(canister, drum, etc.), the HLW overpack and the TRU waste package, together with any
additional packaging provided for transport to or within the repository. After the PEM isfabricated, the buffer and the handling shell provide further containment for HLW. After
receipt, all waste handling operations take place in zones designated as the radiation control
areas and, particularly those where there is a risk of contamination, operating under negative
pressure prevents the release of airborne contamination. Nuclear standard construction andsurface decontamination processes, ventilation and drainage prevent RNs from being released
from the facility in solid, liquid or gaseous form.
With regard to radiation shielding, a basic measure is to properly design the thickness of theshielding walls of relevant buildings to meet dose targets. By appropriately setting thelocation of the facility and the distance of key buildings from the site boundary, radiation
exposure to the public can be reduced to negligible levels. Shielding integrated into transport
vehicles or waste handling systems further reduces exposure to workers, which is controlled
by appropriate monitoring equipment.
Components providing safety functions, summarised in Table 4.2-1, are broken down interms of the relevant operational steps (as shown in Figure 4.2-6) in Table 4.2-2.
In addition to the facility design for the normal state, it is important to consider how
abnormal states can be generated and the countermeasures required to provide defence in
depth, as illustrated schematically in Figure 4.2-7.

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Table 4.2-2 Attribution of requirements of containment and radiation shielding (HLW) torepository components
Safety functions
and corresponding components

Containment during
operation

Surface facility
Repository

Emplacement
machineOn-site transport
vehicleWaste
reception/inspection& encapsulation
facility
Transport cask

✓

Waste
reception/inspection &encapsulation facility✓

Overpack

HLW matrix(including fabrication
canister)

1. Transportation to surface
facilities
2. Acceptance and inspectionof waste, temporary storage

Transport cask

Operation process

Radiation shielding

✓

✓

✓

✓

3. Encapsulation

✓

✓

✓

4. Loading on to transporter

✓

✓

✓

✓

5. Transportation underground

✓

✓

✓

6. Transportation andemplacement in disposal
tunnel

✓✓✓7. Backfilling and sealing ofdisposal tunnels

✓

✓

✓
✓
✓

✓: Components ensuring the safety function
Normal operation

Abnormal operation (Design basis)

Level 1

Level 2

Level 4

Accident management

Control of accident conditions
Success

Success
Limited within the design
basis

Recovery process

Fire
Waste package drop
Loss of power, etc.

Beyond design basis accidents

Control of accidents within thedesign basis

Design basis incidents

Success

Success

Natural hazards
Human factors

Level 3Control of abnormal operationand detection of failures

Initiating events

Prevention of abnormal
operation and failures

External hazard

Normal operation

Accident

Limited accidental radioactive
releases

Mechanical and/or thermal impact on
the waste package

Figure 4.2-7 Abnormal state development and its relationship to safety measures (based on
IAEA, 2005 [22], Nuclear Emergency Headquarters, 2011 [23])4-18In terms of design of facilities and operational processes, it is necessary to consider
measures to prevent the occurrence of any significant abnormal events, together withmeasures to prevent the expansion of abnormal events that cannot be excluded, measures tomitigate resulting impacts, as well as any required recovery measures. Together thesemeasures contribute to the resilience of the repository design, which has been identified as a
key requirement following the Fukushima Daiichi accident (discussed further in Umeki et al.,2017 [24] [25]).
Initiating events that can lead to abnormal states during construction and operation of
repositories include flooding, gas inflow, rock falls, etc., which are mainly related to general
occupational safety. Although the concept of defence in depth/resilient design described wasdeveloped for radiation protection, it can also be applied for general occupational safety. If
there is a need for recovery after incidents, this could also imply radiological considerations.
In such cases an assessment would be needed to determine whether specific designs should be
modified considering this aspect.
Table 4.2-3 summarises the relationship between safety functions and some components
related to general occupational safety.
Table 4.2-3 Relationship between safety functions related to operational safety and main
design components (modified from NUMO, 2011 [1])Basic concept

Prevention ofindustrial
accidents

Maintenance of
healthy working
environment

Safety function
Fire protection
measures
Prevention of
occurrence andMechanical stability of
propagation of
the tunnel
incidents
Additional accident
prevention measures
EvacuationroutesAssuring alternative
established in
evacuation routes
case ofemergency
Maintainconditions
appropriate toworker healthand safety

Ventilation
DrainageLighting

Components
Fire/smoke detection and fire-fighting
equipment
Support work: mesh/shotcrete/rockbolts, liners
Fall prevention devices, emergency
water storage tank
Access shafts for air intake, connecting
tunnels, emergency shelters, airlocks,
communication equipment
Tunnel ventilation circuits, ventilation
system (including cooling system andhumidity control, if required)
Tunnel drainage path, drainage system
Primary and backup lighting
equipmentThis is subdivided in terms of the requirements to maintain a healthy working environment,
building resilience by reducing risks and consequences of perturbations and assuring worker
safety in the event that an accident occurs.
(2) Relationship of post-closure safety functions and repository components(i) HLW
After closure, safety is assured by passive functions of HLW repository components,
provided the disposal tunnels are backfilled and the repository is closed as intended in thedesign, as indicated in Table 4.2-4. The safety functions relating to isolation are
predominantly provided by the geological setting (complemented by markers and/or record
keeping to reduce the risk of inadvertent human intrusion): the key design feature here
4-19

having an impact is the selected disposal depth, with greater depth generally providing moreisolation. Safety functions related to containment can be influenced by design, and are split
into those restricting RN release and those restricting subsequent migration to the biosphere.
At early stages, various THMC processes will impact performance and these will be assessedin detail in future studies.
Table 4.2-4 Relationship between safety functions related to post-closure safety for HLW andmain design components (modified from NUMO, 2011 [1])Isolation

Basic
concept

Safety function
Protection from significant effects of natural
perturbing phenomena

Geological environment

Reduction of the likelihood of human intrusion

Geological environment/depth

Restriction of
RN leaching

Containment

Component

Restriction of
RN migration

Prevention of contact
between waste andgroundwater

Overpack

Low waste matrix
dissolution

Glass matrix

Low solubility of RNs

Geological/EBS environment

Low migration rate due toslow groundwater flow

Geological environment

Limit release from EBS byadvection

BufferColloidal transfer inhibition

Buffer

Sorption of RNs

BufferGeological environment

Dispersion of RNs

Geological environment

Avoiding tunnels acting as
short-circuit flow paths

Backfill/plugs

However, provided no major disturbances occur at early stages the key post-closure
containment functions of the different repository components are:
○ EBS




The HLW borosilicate glass matrix is highly stable and will leach only very slowly incontact with water, limiting releases of RNs. Under stagnant conditions, glassdissolution is associated with formation of secondary products, which will also
incorporate some safety-relevant RNs. Radiolysis of water will occur, predominantlyby alphas at glass/secondary product surfaces, potentially leading to net oxidisingconditions.
The overpack assures complete containment for the period in which the EBS shows
evolving complex THMC conditions and has thus been designed to ensure that RN
releases commence only when all trapped air is consumed, the buffer is fully saturated
and all components are at rock ambient temperature. After failure, the overpack and its
corrosion products will buffer any radiolytic oxidants produced within it and may
4-20





sorb/incorporate some released RNs. As loss of containment will probably belocalised, the overpack may also act as a barrier to RN release to the inner side of thebuffer.
As the buffer material saturates with groundwater, it swells to form a homogeneous,plastic, low permeability, self-healing, microporous barrier that provides mechanicalprotection of the overpack, acts as a colloid filter and functions as a semi-permeable
membrane. The buffer thus prevents significant advective flow and retards the
diffusive transport of RNs.
Retention within the EBS will mean that a large part of the total radioactivity
inventory will decay within this zone and releases will be restricted to more mobile,
long-lived RNs (and their decay daughters).

○ Backfill and plugs
Repository access (ramps and shafts), connecting and disposal tunnels, and theirassociated excavation damaged zones, provide potential short-circuit flow paths after
closure. This is prevented by tailoring tunnel layout in relation to hydraulic gradientsand emplacement of appropriate backfill and plugs/seals.

○ Geological environmentThe large geological overburden effectively isolates the repository disposal panelsfrom the biosphere, assuring that, after closure, even major surface perturbations haveno impact on disposed waste. After loss of containment by the EBS, the slow flow ofgroundwater deep underground constrains subsequent migration, while the mineralogyof flow paths and the reducing chemical environment leads to restriction by
sorption/precipitation of RNs. Even non-reactive RNs can be retarded by diffusionfrom advective flow paths into non-flowing matrix porosity. During transport throughthe geosphere barrier, concentrations of RNs are reduced by both decay andhydrodynamic dispersion.

(ii) TRU waste
For definitions of TRU waste groups, see Table 2.1-2 in Chapter 2. After closure, safety is
assured by passive functions of the TRU waste repository components as indicated in Table
4.2-5. The isolation role is exactly the same as for HLW. As in the TRU-2 report, vaults forGrs.1 and 2 are assumed to include a buffer. The newly defined Gr.4H is also assumed toinclude a buffer due to its similarity to Gr.2 in terms of thermal output.

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Basicconcept

Safety function

Isolation

Table 4.2-5 Relationship between safety functions related to post-closure safety for TRU waste
and main design components

Protection from the significant
effects of natural perturbing
phenomena
Reduction of the likelihood of
human intrusion

Waste Gr.
1

2

3

4L

4H

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓
✓

✓
✓*3
✓*3

✓
✓
✓

✓
✓

✓*3
✓*3

Low solubility of
RNs

✓

✓

✓

✓

✓

Geologicalenvironment

Low migration rate
due to slow
groundwater flow

✓

✓

✓

✓

✓

Geologicalenvironment

✓

✓

✓

Buffer✓✓✓Buffer

Sorption of RNs

✓
✓
✓

✓
✓*3
✓

✓
✓

✓
✓

✓
✓*3
✓

BufferInfillDispersion of RNs

✓

✓

✓

✓

✓

Geologicalenvironment

Avoiding tunnels
acting as short-circuit
flow paths

✓

✓

✓

✓

✓

Backfill, plugs*3

Restriction of RN migration

Restriction of RN leaching

Containment

Component

Limit release from
EBS by advection
Colloidal transfer
inhibition

Geologicalenvironment

Waste matrix*1
Waste package*2✓Component to secure safety function in each waste group.
*1TRU waste is within a container or drum, which may prevent waste contact with groundwater for a period after
closure.
*2Waste package B has a lid and hence waste contact with groundwater is prevented until this package fails.
*3This is heat-generating waste, and it is necessary to confirm possible thermal impacts on the safety function
when developing the design.

Grs.3 and 4L have no buffer. Key containment functions of the different repository
components are thus:○ TRU waste EBSSome TRU waste streams have a matrix that limits RN release rates. This isparticularly the case for Gr.2 metallic hulls and ends and Gr.3 bitumen. It is possiblethat conditioning will also limit release rates of other waste streams, but the associatedknowledge base is limited.
TRU waste is conditioned within sealed drums or metal containers, which will prevent
RN release until they fail.
The TRU waste disposal package B (see Section 4.2.3 (2) (ii)) will provide complete
containment until it fails and, thereafter, like package A, may limit RN release bysorption onto corrosion products or package infill.
4-22





After failure of containment, releases of RNs will be constrained by both restriction ofgroundwater flow by structural concrete and vault infill/backfill and also sorption onto
these materials. The high pH environment assured by the large quantity of concretealso limits the solubility of some key RNs. It is possible that heat-emitting waste(Grs.2 and 4H) may cause thermal degradation of such properties.
When buffer material is present, it saturates with groundwater and swells to form ahomogeneous, plastic, low permeability, self-healing, microporous barrier that acts asa colloid filter and functions as a semi-permeable membrane. The buffer thus prevents
significant advective flow and retards the diffusive transport of RNs.

○ Backfill and plugs
As for HLW above.

○ Geological environmentAs for HLW above.

4.2.5 Features of the SDM relevant to repository design
The design of the repository is carried out based on the 3D SDMs for representative rocks
developed in Chapter 3. The SDMs include structural, lithological and hydrogeologicalmodels, together with physical, mechanical and thermal characteristics of the host rock (seeTable 4.2-6 below) and the chemical composition of groundwater required to develop
repository design. Note that for the Pre-Neogene sediments, the coherent facies that was used
to build the repository-scale hydrogeological model (Section 3.3.3 (4) (ii) (c)) was also used
for the repository design.
Thermal properties of rock generally depend on porosity and water content, which cansignificantly alter these parameters [26]. In the H12 report, thermal conductivity and specificheat are defined by contributions of water, air and rock matrix given by their volumetric ratios,
which is also assumed in this report. Based on data from Table 3.3-17, required thermal andmechanical property values have been derived and summarised in Supporting Report 4-2.
The uniaxial compressive strength can be used to define the rock stability index3 forunderground facilities. Low values of this parameter, especially for Neogene sediments,
indicate the need to provide mechanical support for tunnels or locate the facility at a shallower
depth.

3

The stability index is an indicator used as a guide for setting rock stability and is obtained by dividing uniaxial
compressive strength by initial stress. Generally, the smaller the value is, the more unstable a tunnel is. When the stability
index is less than 2, it is considered that rock damage is likely to occur around the tunnel. (N.B.This footnote is not includedin the Japanese version of the report.)

4-23

Table 4.2-6 Values of geological environment characteristics (physical, mechanical andthermal properties)
Rock type

Plutonic

PreNeogene

Neogene

2.69

2.64

2.28

2.7

2.7

2.7

0.8

3.5

25

Saturated density  (Mg/m3)
PhysicalTrue density R (Mg/m3)
characteristics
Effective porosity ne (%)
Uniaxial compressive strength qu
(MPa)

115

15

Modulus of elasticity E (MPa)37,0003,500

Poisson's ratio ʋ (-)

0.25

0.3

Cohesion intercept c (MPa)153Internal friction angle φ (°)

45

28

8

2.1

Tensile strength σt (MPa)
Mechanicalcharacteristics
Lateral pressure coefficient K0 (-)

164/h + 0.74 (h: depth (m))

Initial vertical stress σv (MPa)gh/1,000(: saturated density (Mg/m3),
g: gravitational acceleration (9.8 m/s2),
h: depth (m)Initial horizontal stress σh (MPa)K0σv

Elastic wave (P wave) velocity Vp(km/s)

2.1 + 2.9 {1-exp(-0.00792 ρh)}

Thermal conductivity λ (W/m K)

2.9

2.8

2.3

Specific heat C(kJ/kg K)

1.0

1.0

1.4

ThermalGeothermal gradient dT/dz (°C/100 m)
characteristics
Mean surface temperature T (°C)

1.8 + 1.4 {1-exp(-0.00057 ρh)}3
15

Table 4.2-7 shows the hydrogeological characteristics (hydraulic conductivity) used indesign, taken from repository-scale models (Section 3.3.3 (4) (ii)). Here, plutonic rocks are
assigned a single lithology, while Neogene and Pre-Neogene sediments are assigned hydraulicconductivities for each of the main lithological units present in the SDM.
Table 4.2-7 Design data of geological environment characteristics (repository scalehydrogeological properties)
Rock typeHydraulicconductivity (m/s)

Plutonic

Neogene

Pre-Neogene

2.7 × 10-8

Upper mudstone layer:
2.0 × 10-9
Lower mudstone layer:
4.4 × 10-8
Sandstone layer:
5.3 × 10-7

Sandstone dominant substrate:
2.0 × 10-9
Mudstone dominant substrate:
2.0 × 10-9
Chert rock mass:
1.0 × 10-8Reference groundwater chemistries for the 3 representative rocks are given in Table 3.3-16,
presenting high salinity and low salinity variants for each (more information in Supporting
4-24

Report 3-31). In Table 4.2-8, the key chemical parameters for design are presented: pH, Eh,salinity, total inorganic carbon and sulphur species concentrations. The pH is in the range of
6.3 to 8.4, near neutral to weakly alkaline, with Eh in the reducing range of -170 to -300 mV,
which would be favourable for minimising metal corrosion.
Table 4.2-8 Water chemistry used for repository design
Rock type

Plutonic

Neogene

Pre-Neogene

Groundwatersalinity

Low

High

Low

High

Low

High

Water T (°C)

45

45

30

30

45

45

pH

8.2

7.6

8.4

6.5

8.2

6.3

Eh (mV)

-300

-260

-280

-170

-290

-170

1.1 × 10-3

2.1 × 10-1

1.7 × 10-3

4.7 × 10-2

Cl concentration (proxy for salinity) (mol/l)-

Cl

2.3 × 10-3

9.5 × 10-4

4.9 × 10-2

1.1 × 10-3

2.1 × 10-1

Carbonate species concentration (mol/l)
Total C

2.2 × 10-4

1.7 × 10-3

4.0 × 10-2

Sulphur species concentration (mol/l)
H 2S

4.8 × 10-10

2.8 × 10-9

9.5 × 10-11

1.4 × 10-9

3.3 × 10-10

4.7 × 10-9HS-

1.3 × 10-8

2.2 × 10-8

2.9 × 10-9

8.5 × 10-10

8.7 × 10-9

2.5 × 10-9SO42-

6.9 × 10-6

1.4 × 10-5

1.2 × 10-4

2.6 × 10-6

1.2 × 10-4

2.5 × 10-6

CaSO4

2.3 × 10-7

5.8 × 10-6

2.0 × 10-6

8.1 × 10-8

2.3 × 10-6

8.5 × 10-8

NaSO4-

1.3 × 10-7

7.7 × 10-7

2.1 × 10-6

1.1 × 10-6

2.1 × 10-6

1.1 × 10-6

MgSO4

1.3 × 10-8

6.7 × 10-8

2.9 × 10-7

3.0 × 10-7

4.1 × 10-7

4.0 × 10-7

Even at its highest, the salinity is about a third of that of sea water, and thus lies within therange used during testing of overpack and buffer materials. In Table 4.2-8, the carbonate
concentration is noted to allow assessing the passivation of the overpack carbon steel andevaluation of effects on local corrosion.
Carbonate concentrations between 2.2 × 10-4 and 4.7 × 10-2 mol/l lie within the range of,preferable geological environment characteristics, being below a specified upper limit of 0.5
mol/l (Table 3.1-2). Sulphur species concentration is used to assess microbial corrosion of
metal.

4.3 Setting of repository depth
A key constraint on design of the EBS and underground facilities layout is the selected
disposal depth. Clearly, the depth at which potential host formations are present at a site plays
a key role but, in the following, it is assumed that this does not constrain choice (however thiswill need a thorough assessment during the PI stage). As a boundary condition for geologicaldisposal, a clear legal requirement of a depth below surface of at least 300 m is defined. In
addition, a further depth margin can be considered to account for the effects of uplift anderosion during the main evaluation period (100 ky), which would define a minimum depth.
As mentioned in Section 3.4.3, long-term uplift and erosion strongly depend on local
topography, geology, tectonic conditions, etc., which can be considered quantitatively when asite is available. Thus, rather than conducting a detailed quantitative assessment of erosion, a
4-25

rough estimate is made here using average data, showing that uplift rates in Japan generally
lie in the range of 0 - 30 m in 100 ky (see Supporting Report 3-35).
Although risks of perturbation due to uplift and erosion decrease with increasing depth, thisneeds to be balanced against greater lithostatic and hydrostatic loads and higher ambient rocktemperature, so choice of depth has to balance these conflicting requirements. For this reason,
the maximum depth at which the HLW repository can be reasonably constructed is defined
based on the design requirements shown in Table 4.3-1. Nevertheless, it should be recognisedthat hydraulic properties and hydro-geochemistry may also be depth dependent, and thiswould also be factored into final selection for actual sites.
Table 4.3-1 Design considerations for setting the HLW repository depth
Increasing repository depth:
Influenced components and impacts
Component

ImpactDesign considerations

TunnelLowering of the cavity stability
due to the increase in rock stress

Ensuring the mechanical stability of the cavity withreasonable shoring (Supporting Report 4-3)

BufferIncrease in the waste footprint
due to the increase of rocktemperature

For specified buffer temperature limit (100 °C), the waste
package separation has to be increased as required(Supporting Report 4-14)

Ventilation
systemIncrease in ventilation capacity
due to the increase of rocktemperature

Ventilation equipment has to be specified to limit
temperature in working areas as specified by Article 611
of the occupational health and safety regulations (37 °C)

Using Table 4.2-6 geological environment characteristics, reference depths of 1,000 m forplutonic rocks and Pre-Neogene sediments and 500 m for Neogene sediments was derived
using the simple, rather conservative assessment for HLW disposal described in detail in
Supporting Report 4-3. When selecting these depths, no specific consideration was maderegarding potential differences between the siting environments sensitivity to erosion, butsuch considerations would certainly be made at real sites. Assuming co-disposal of HLW andTRU waste, both disposal zones are assumed to be at the same depth, although depths foreach would be re-assessed independently based on conditions found at a specific site, using
less conservative assumptions.
More factors (e.g. construction practicality, costs, thermal dimensioning and risk of human
intrusion) than those considered above may need to be assessed in setting repository depth.
The required approach has been developed, so that it can be applied on a site-specific basis.

4.4 EBS design
The basic safety requirements for the HLW and TRU waste repositories are noted in
Section 4.1.1 and expanded upon to derive safety functions and the related design
requirements of specific components of the EBS in Section 4.2.4. Here, detailed designs aredeveloped based on these. In addition, their practicality is assessed based on the availability ordevelopment needs of technologies, with particular emphasis on manufacture/construction torequired quality levels.

4-26

4.4.1 EBS for HLW(1) EBS performance targets
HLW EBS components must meet both operational and post-closure safety functions (Table
4.2-1, Table 4.2-4), as indicated in Figure 4.4-1. The key EBS components, for the specified
glass waste form and steel fabrication canister, are a long-lived overpack and a low
permeability, plastic buffer that acts as a colloid filter.
Vitrified high level waste(Including a stainless steel fabrication canister)• Restriction of RN leaching;“Low waste matrix dissolution”

Overpack(Metal container)• Restriction of RN leaching;“Prevention of contact waste and groundwater”

Buffer(Compacted bentonite)

Buffermaterial• Restriction of RN migration;“Limit release from EBS by advection”“Colloidal transfer inhibition”“Sorption of RNs”

Host rock

Figure 4.4-1 EBS for HLW and illustration of some associated safety functions(adapted from NUMO, 2004 [2])HLW is specified in Section 4.2.2 (1) as an EBS component, but properties are given byprogramme boundary conditions and hence not subject to design (although other components
should prevent favourable HLW matrix properties from being degraded over relevanttimescales). The overpack and buffer are designed on the basis of the specified shape, size,
thermal output and radioactivity of the HLW canister. Also, as mentioned in Section 4.2.1,
vertical H12V and horizontal PEM systems with the same overpack/buffer specifications are
considered for the three defined SDMs.

4-27

(2) Overpack design(i) Overpack design requirements and evaluation items
The overpack design requirements are shown in Table 4.4-1. The main safety functions of
the overpack include prevention of leakage of RNs from waste during operations (Table 4.21) and prevention of release of RNs to groundwater for an extended period after closure(Table 4.2-4).
Table 4.4-1 Overpack design requirements
Designrequirements

Objectives

Low
corrosion rate

The safety function to prevent waste from coming into contact withgroundwater for a prescribed period of time after disposal is assured

Structural
integrity

The safety function to prevent the waste from coming into contact
with the groundwater is not impaired by the loads acting during
operation and for a prescribed period after disposal

Specifications
Material,
thickness,
lid seal
Material,
thickness, shape,
lid seal, lifting
features

No radiation
effects oncorrosion rate

No significant acceleration of corrosion by radiation field

Material,
thickness

Qualityassuredmanufacture

Structure and materials that can be manufactured by existing
technology or technology that will be realised in the near future

Material, shape

Remotehandled HLW
encapsulation

Processes and equipment capable of remote-handled encapsulation of
HLW by existing technology or that which will be realised in the near
futureRemotehandled
emplacement

Processes and equipment capable of remote handled overpack
emplacement using existing technology or technology that will be
realised in the near future

Materials,
dimensions,lid sealing
concept
Shape, weight,
lifting features(would not
apply for PEM
option)To meet these safety functions, design requirements include: corrosion resistance, structural
integrity and avoidance of radiation effects on corrosion. In addition, practicality offabrication to required quality, remote-handled encapsulation and remote-handled
emplacement are set as design requirements to meet the design factor “engineering
practicality”. These design requirements are used to develop basic overpack specifications.
The period required to maintain the overpack containment is considered as follows. At
early times after closure, HLW radioactivity and heat generation is large, and processes in thenear field are complicated by coupled saturation, swelling and chemical reactions in the buffer.
If RNs were released from the waste at this time, quantifying migration under these transientconditions would involve large uncertainties. Therefore, as in the H12 report, to simplify theassessment it is useful to assure complete containment during the period of significant heat
output, which is taken to be about 1 ky (after which time the thermal output from HLW is
only a few W/canister).
As indicated in Section 4.1.2, the impacts of loss of overpack function have to beconsidered in terms of performance of the entire system (coupling overpack design to that ofother EBS components), both before (Chapter 5) and after closure (Chapter 6). Guidance onsuch coupling is provided by the Features Events and Processes (FEP) list (Section 6.2.1,
4-28

Supporting Report 6-4), which identifies further issues to be considered, as shown in Table
4.4-2.
The resultant design specifications for the overpack include assurance of integrity in theevent of localised buffer swelling, weld defects or earthquake ground motions. Additionally,
the chemical impacts of high groundwater salinity and sodium in bentonite being exchanged
by calcium have been considered in the design. The details of setting these design
requirements and evaluation items are given in Supporting Report 4-4. Other potential
impacts on how the EBS stability may be altered by unfavourable changes in the chemical
environment are discussed in Chapter 6. If additional countermeasures against chemical
impacts are required as a result of the performance assessment (Chapter 6), these will befurther considered in the light of more detailed understanding of the evolving groundwaterchemistry at specific sites.
Table 4.4-2 Overpack evaluation items considered in this report
Evaluation item

Objectives

Structural integrity in case of
uneven buffer swelling

Load caused by uneven swelling of buffer material due
to variable water content and uneven expansion of
corrosion products does not impair structural integrity

Structural integrity in case of
weld defects

The presence of weld defects does not compromise
structural integrity

Structural integrity against
earthquakes

Structural integrity is not impaired even for large-scale
earthquake motions(ii) Setting of overpack material, shape and inner dimensions
The overpack design and the exact specification of overpack material, shape anddimensions can only be decided when a specific repository concept has been selected.
However, to illustrate the design steps involved, this study focuses on a steel overpack for asingle HLW canister, with a bentonite/crushed rock buffer. Although copper and titanium
were considered as alternative materials in the H12 report [6] [27], carbon steel is chosen here
as the reference case due to the large knowledge base for this material. Furthermore, with afocus on ease of assuring structural integrity, fabrication by carbon steel forging (SF) thatmeets the relevant Japanese standard (JIS G 3201) is selected, with a specification (SF340A)
that facilitates welding with a low risk of cracking. As described further in Section 4.4.1 (2)(viii), technological development is ongoing to confirm practicality of manufacturing andwelding this material to required quality levels. It should be noted here that selection of
welding as a sealing technique for the reference concept is again based on significant
industrial experience. It may be, however, as sites are gradually narrowed down and a final
concept is selected, that other options may be considered.
The shape of the overpack is taken to be cylindrical, from the viewpoint of the cylindricalshape of the HLW canister and structural integrity considerations. Furthermore, from thedimensions of the canister (diameter 430 mm × height 1,340 mm) and considering fabrication
accuracy and remote handled encapsulation, the inner overpack dimensions are currently set
as diameter 440 mm and height 1,350 mm.

4-29

(iii) Setting overpack thickness
In the H12 report, the thickness of the carbon steel overpack is set to 190 mm regardless ofthe geological environment [6]. In this report, the required thickness of the overpack is reevaluated. Table 4.4-1 indicates how overpack thickness contributions to corrosion resistance,
structural integrity and radiation shielding can be evaluated, based on the process shown in
Figure 4.4-2.
Structural integrity③ Required pressure
resistance thickness

Radiation shielding to avoid
accelerated corrosion④ Required shielding
thickness

Corrosion resistance② Required anaerobic
corrosion allowance

Corrosion resistance① Required aerobic corrosion
allowance

Required thickness of overpack① + Max {(② + ③), ④}

Figure 4.4-2 Setting overpack thickness

In the evaluation of corrosion resistance, stress corrosion cracking (SCC), hydrogen
embrittlement cracking and microbial corrosion are also explicitly considered. Further detailsof the approach to setting thickness are provided in Supporting Report 4-5.
(a) Corrosion resistance
Environmental conditions that affect overpack corrosion include temperature, redox
conditions and groundwater chemistry. The saturated buffer is expected to provide anisotropic compressive stress field and limit microbial activity, both of which are favourable
for overpack longevity.
Section 4.5 sets a limiting buffer temperature of 100 °C (which relates to past simplistic
assumptions and does not relate to current best-estimates to assure negligible alteration),
which will relatively quickly decrease towards rock ambient. Similarly, initial trapped air will
be rapidly consumed by reaction with the overpack, or minerals in buffer or host rock, to givea reducing environment. Groundwater chemistry (Table 3.3-16) will be changed by reaction
with buffer minerals to produce the porewater chemistry in contact with the overpack (seeSupporting Report 6-5), which is shown in Table 4.4-3.

4-30

Table 4.4-3 Buffer pore water chemistry components considered relevant to carbon steel
corrosion
Rock type
Groundwatersalinity

Plutonic

Neogene

Pre-Neogene

Low

High

Low

High

Low

High

pH

7.2 to 9.6

7.1 to 8.4

7.2 to 9.8

6.3 to 6.5

7.1 to 9.6

6.1 to 6.3

Eh(function of pH)(mV)-4.1 × 102
to-2.0 × 102-3.2 × 102
to-2.0 × 102-3.9 × 102
to-1.9 × 102-1.7 × 102
to-1.4 × 102-4.0 × 102
to-2.0 × 102-1.7 × 102
to-1.4 × 102

Cl(mol/l)

2.3 × 10-3
to
4.5 × 10-3

4.9 × 10-2to
5.1 × 10-2

1.1 × 10-3to
3.3 × 10-3

2.1 × 10-1

1.1 × 10-3to
3.3 × 10-3

2.1 × 10-1

Inorganiccarbonconcentration(mol/l)9.6 × 10-4
to
7.9 × 10-3

1.2 × 10-4to
4.6 × 10-3

1.6 × 10-3
to
9.1 × 10-3

3.1 × 10-2
to
4.0 × 10-2

1.7 × 10-3
to
8.6 × 10-3

3.6 × 10-2
to
4.7 × 10-2More information on the calculation of pore water composition is provided in Supporting
Report 6-15. Carbonate concentrations lie in the range of 1 × 10-4 to 5 × 10-2 mol/l which,
together with relevant pH values, are used for the examination of sensitivity to SCC. Salinities
lie in the range 1 × 10-3 to 2 × 10-1 mol/l, which is taken into account when assessing theresults of corrosion tests. Where there is the potential for microbial activity in the buffer, theconcentration of sulphur species may also be important.
For corrosion in the oxidising environment immediately after disposal, as in the H12 report,
an average corrosion depth is calculated assuming that all oxygen remaining in the near-fieldrock, the buffer and backfill is consumed by corrosion (certainly over-conservative for thePEM case). This assumes that no oxygen can reach the overpack from the open tunnel system.
From this, the maximum corrosion depth was estimated based on extreme value statisticalanalysis of corrosion data. From the analysis described in Supporting Report 4-6, based on thetunnel geometry given in Section 4.5.2, maximum corrosion depths of 5 to12 mm are derived,
as shown in Table 4.4-4. Future consideration of actual knowledge and uncertainties willallow more realistic estimates of such corrosion to be derived, which may be significantly
smaller.

4-31

Table 4.4-4 Required carbon steel overpack minimum thickness for 1 ky lifetime (units in mm)
1 Corrosion 2 Long-termallowance for corrosion
34
the initial
allowance Pressure Requiredoxidisingfor reducing resistance shielding
environment conditions

Conditions
Rock

Plutonic
and
PreNeogene

Emplacement
method
H12V

PEM
H12V

Part
Flat

11

6

104

80

121

Cylindrical

11

6

44

80

91

Flat

5

6

104

80

115

Cylindrical

5

6

44

80

85

Flat

12

6

78

80

96

Cylindrical

12

6

25

80

92

Flat

5

6

78

80

89

Cylindrical

5

6

2

80

85

NeogenePEM

Overpack
Required
thickness= 1 + Max{(2 + 3), 4}

Steel corrosion under anaerobic conditions in buffer saturated with artificial seawater (pH8.5, NaCl: 5.6 × 10-1 mol/l, NaHCO3: 2.4 × 10-3 mol/l) or artificial fresh water (pH 8.5, NaCl:
2.5 × 10-3 mol/l, NaHCO3: 2.5 × 10-3 mol/l) has been measured for timescales of around a
decade [28]. The average corrosion rates measured initially decreased over time regardless of
salinity, which is consistent with the observed surface films caused by precipitation of
secondary minerals such as iron carbonate (siderite) [28] [29]. However, as shown in Figure
4.4-3, the average depth of corrosion increased approximately linearly with time after 1 year,
in all cases, yielding average corrosion rates of 2 m/y or less [28], but there is also potential
for localised corrosion that must be considered.
Steel
母材AA
Steel
母材BB
Steel
母材CC
Steel
母材AA
Steel
母材BB
Steel
母材CC

(m)
corrosion depth
Mean
[µm]
平均腐食深さ40

Simulated pH: 8.5
Cl-: 5.6×10-1 mol/l
salinewater
HCO3-: 2.4×10-3 mol/l
Simulated pH: 8.5
flesh
NaCl: 2.5×10-3 mol/lwater
NaHCO3: 2.5×10-3 mol/l

351.95 µm/y
1.72 µm/y30
251.50 µm/y20

1.17 µm/y

150.92 µm/y0.87 µm/y

10SF340A, 80 ºC

SF340A, 80℃
Drydensity of1.6
bentonite
乾燥密度:
Mg/m3
specimen with 30 wt.%
ケイ砂混合率:30%3silicasand 1.6 Mg/m

5
0

0

2

SF340A, 80 ºC
Dry density of bentonite
specimen with 30 wt.% silica
sand 1.6 Mg/m3

4 6 8 10 12 14
浸漬期間［y］
Immersion
period (y)Figure 4.4-3 Evolution of the average corrosion depth of the carbon steel in buffer material
4
under an anaerobic environment [28]4

In the literature [28], corrosion tests using a base material SF340A for tests of TIG (Tungsten Inert Gas) welding, electron
beam welding and MAG welding are reported. Here, they were noted as base materials A, B, and C, respectively.

4-32Based on this result, the average corrosion rate was set as 2 m/y, independent of rock type,equivalent to 2 mm over the reference 1 ky period. Furthermore, considering potential unevencorrosion in this environment a maximum pitting factor was selected [30], leading to a
maximum corrosion depth of 6 mm. This corrosion allowance is shown in Table 4.4-4. Moredetails of the calculation and assumptions are given in Supporting Report 4-7.
SCC and hydrogen embrittlement cracking are phenomena that could potentially cause
early failure of containment. These depend on stress conditions, material properties andenvironmental conditions – in particular temperature and water chemistry [31]. As SCC
results from tensile stress, removal of residual stress by heat treatment after manufacture will
be carried out, but there is a possibility that tensile stress may remain following welding of
the lid. Thus, Section 4.4.1 (2) (viii) describes welding options that also include subsequent
heat treatment, which should greatly reduce any residual tensile stress and assure that the
overpack is subject to a predominantly compressive stress field.
Impacts of SCC and hydrogen embrittlement on the overpack are assessed in detail in
Supporting Report 4-8. Relevant SCC forms are those that occur (1) at high pH and highcarbonate concentrations; (2) at near neutral pH and low carbonate concentrations [32]; (3)associated with pitting/crevice corrosion of passivated steel at high salinities. Based on
chemical constraints alone, as set by the porewater chemistries shown in Table 4.4-3, SCC
forms (1) and (3) can be confidently excluded for relevant carbon steel, although the hightensile stresses required are also very unlikely. For near-neutral pH SCC, pore-waterchemistry is marginal (pH tending to be too high) and again residual tensile stresses are
expected to be too low. Thus, SCC is not considered a threat for carbon steel in compacted
bentonite but, in any case, heat treatment after welding is planned.
Hydrogen embrittlement cracking may occur when hydrogen generated by carbon steel
corrosion is absorbed and diffuses into the steel at a high enough concentration [6]. This
process occurs predominantly in high strength steel and, for the more vulnerable welds,
tensile strength is low enough that a relatively high diffusible hydrogen concentration isrequired (several ppm or higher), which is well above the values measured under relevant
conditions and hence hydrogen embrittlement cracking is considered unlikely.
In addition, microbial activity can influence corrosion of steel [33]. Section 4.4.1 (3) (iii)indicates that microbial activity is low in buffer material having a dry density above 0.56
Mg/m3, a situation expected to be maintained as long as the initially installed buffer density
does not decrease too much [34]. Therefore, the probability of significant microbial activity
on the overpack surface is considered low. However, considering that other studies [35]suggest much higher limits, work on this issue continues. Based on the more detailed
assessment in Supporting Report 4-9, regardless of the microbial activity level, limitations inthe supply rate of oxidants that can be used by microbes to catalyse corrosion (e.g. sulphate)in this hyper-oligotrophic environment result in total corrosion that is expected to be
negligible compared to inorganic processes. Nevertheless, the activity of microorganisms inbuffer materials is still being evaluated [35] and impacts will be reassessed in the future.
(b) Structural integrity
As the buffer re-saturates and swells, resultant pressure will add to the groundwaterhydrostatic pressure. Furthermore, buffer deformation will result from corrosive expansion of
the overpack together with rock creep (for soft rocks), with the combined force (referred to as
the compaction reaction force) considered to act on the overpack. As discussed in detail in
4-33

Supporting Report 4-10, the combination of these loads results in a calculated externalpressure acting on the overpack of 10.7 MPa for plutonic rocks and Neogene sediments (depth
1,000 m), and 6.8 MPa for Neogene sediments (depth 500 m). The required thickness towithstand these external pressures is set very conservatively using nuclear power plant
standards [36], specifically as applied to class MC containers. As a result, the thickness
required for structural stability is set for both the flat ends and the cylindrical body as 78 to
104 mm and 25 to 44 mm respectively, as shown in Table 4.4-4.
(c) Prevention of radiation effects on corrosion
Oxidising species, such as hydrogen peroxide, may be generated by the radiolysis of water
due to penetrating radiation from the HLW. Corrosion tests of carbon steel in artificial
seawater and low oxygen concentrations have been carried out to determine the impact of γirradiation [38] [39] [40]. Significant corrosion rate acceleration is observed only at doses
above 3 Gy/h. Thus, a shielding margin is specified such that the absorbed dose rate of theoverpack surface is equal to or less than 3 Gy/h. On the basis of a very conservative shielding
analysis, the required shielding thickness was set to 80 mm (Supporting Report 4-11).
However, it is recognised that more realistic analyses (e.g. taking into account decreases inradiation dose with time) could reduce this thickness significantly.
(d) Overpack thickness
Based on the above results, the procedure for setting the required thickness of the overpack inaccordance with Figure 4.4-2 is summarised in Table 4.4-4. The required thickness of theoverpack to assure 1 ky lifetime, even for the worst case plutonic H12V system, is a minimumof 121 mm. However, the thickness of the overpack is set to the H12 value of 190 mm,
defined to cover all emplacement methods and both the lid and the cylindrical body with alarge safety margin, although this could be rationalised in the future. As noted previously, thedimensions provided in this study are not optimised with respect to requirements, andespecially not adapted for site-specific conditions. However, this will be a topic for future R& D.
(iv) Design of the overpack lid
As indicated in Table 4.4-1, the lid seal is relevant in terms of corrosion resistance andstructural integrity. Two basic lid designs are currently considered, a drop-lid type and a flat
type, as illustrated in Figure 4.4-4. Here the focus is on a drop lid structure. However, in the
future, other designs, and even replacement of welding with a screw lid design could bepossible.

Drop-in lid

Flat lid

Figure 4.4-4 Schematic diagram of the lid structure

To set the weld depth, this was examined from the points of view of both corrosion
resistance and structural integrity. Incidentally, with this lid structure, radiation is shielded by
4-34

the base material peripheral to the weld and hence it is not necessary to set a radiation
shielding margin for the weld itself (Figure 4.4-5).

Weld

Cylindrical body

Lid

Radiation passage
through the gap
Direct exposure ofradiation

Figure 4.4-5 shielding of radiation around the weld(a) Corrosion resistance
Under aerobic conditions, the corrosion behaviour of the weld is expected to be the same asthat of base material in the case of Electron Beam Welding (EBW), which does not use any
welding rods. For other methods that use some kind of welding rod, such as TIG (Tungsten
Inert Gas) or MAG (Metal Active Gas) welding, the phenomenon of selective corrosion of theweld compared to the base material has been observed [41] [42]. However, it has been
confirmed that the use of an improved Ni doped welding material has the effect of restricting
such corrosion [43].
For long-term anaerobic corrosion, the average corrosion rate of welds in a reducing
environment has been measured for immersion periods of up to 10 years in fresh and saline
solutions, confirming that there is no significant difference in the corrosion behaviour of theweld and the base material [28] [37]. Thus, from Table 4.4-4, the sum of corrosion underoxidising and reducing conditions lies in the range of 11 to 18 mm in 1 ky.
Section 4.4.1 (2) (viii) defines the size limit for detection of weld defects as 2 to 3 mm. To
account for the fact that an undetectable weld defect may be present, the required corrosion
allowance of the base material is increased by 3 mm and thus the required corrosion
allowance of the weld was set to between 14 and 21 mm.
(b) Structural integrity
The minimum weld depth is required to ensure that structural integrity is not impaired byloads acting during operation and after disposal, which are assessed in detail in Supporting
Report 4-10 for the dominant post-closure loads summarised in Section 4.4.1(2) (iii) (b).
Based on this shear stress, the required welding depth was conservatively set by applying theallowable shear stress specified in the rules on design and construction for nuclear power
plants [36]. The resulting minimum weld depths are 20 mm for plutonic rocks and PreNeogene sediments and 12 mm for Neogene sediments.

4-35

(c) Setting the welding depth
As summarised in Table 4.4-5 summing the requirements gives minimum weld depths of 26
to 40 mm.
Table 4.4-5 Required minimum weld depth (mm) for the overpack
1
Corrosion
allowance forthe initial
Emplacement
oxidisingmethodenvironment

Conditions

Rock type3
4
Required2
welding depthLong-term
Consideration Pressure = 1 + 2 + 3 + 4
corrosionof welding
resistanceallowance for
defects
reducing
conditions

Plutonic
and
Pre-Neogene

H12V

11

6

3

20

40

PEM

5

6

3

20

34

Neogene

H12

12

6

3

12

33

PEM

5

6

3

12

26

Nevertheless, as in the H12 report, this parameter was conservatively set to the full
overpack thickness. Therefore, rationalisation of welding depth is possible in the future.
(v) Design of the overpack lifting feature

190

The grip of the overpack shown in Figure 4.4-6 is based on the H12 and past studies [44],with the height of the grip portion given as 150 mm. This design will be revisited in the future,but it has little significance for the present safety case.

190
210

190

Φ440
Φ820

Weld

20
150

210

Grip

1,350
1,750
1,900

190

Figure 4.4-6 Overpack specification(vi) Overpack specifications
Based on the above study, the overpack specification illustrated in Figure 4.4-6 is assumed
to apply for both disposal concepts and the 3 representative host rocks.
(vii) Evaluation of overpack integrity
In Table 4.4-2, further evaluation items to assure overpack integrity were noted, including
stability in the case of uneven buffer swelling, seismic motion and weld defects.
4-36

The net stress acting on the overpack is the sum of those resulting from expansion of
corrosion products, hydrostatic load/rock creep and swelling of the buffer. Particularly in thecase of uneven wetting of the buffer at early times, this could potentially lead to a bending
force acting on the overpack, which could be a cause for SCC. On the basis of the extreme
case of an uneven load on the overpack, evaluation results show that the plasticised zone does
not extend through the overpack and hence mechanical failure will not occur [6]. For the PEM
case, this phenomenon can potentially be prevented by including an outer geotextile layer, as
demonstrated in the FEBEX mock-up [45].
Impacts of seismic motion are considered in Section 4.4.1 (3) (iv), with a focus on stabilityof the buffer. From this analysis (described in detail in Supporting Report 4-17), themaximum stress acting on the overpack is only on the order of 1% for the yield strength andhence the risk of damage is negligible [46] [47]. In addition, the impact of secondary fractures
moving as a result of the seismic motion has been considered (Section 4.4.1 (3) (iv)), but thiscan be mitigated by selecting a sufficiently plastic buffer.
As assessed in detail in Supporting Report 4-12, the effect of weld defects on structural
integrity can be evaluated by consideration of welding residual stress and external forces
acting on the overpack. Conservatively assuming decreased fracture toughness due toembrittlement, the allowable defect size that still ensures mechanical stability is 55 mm,
reduced by a 10-fold safety factor to 5 mm. Therefore, since Section 4.4.1 (2) (viii) argues
that non-destructive inspection techniques can detect defects of 2 to 3 mm or more, any defect> ≈ 5 mm would certainly be picked up by weld inspection and hence this should not be aconcern. Because of the huge degree of conservatism involved, more realistic assessment of
maximum expected defect size would be useful to provide guidance on the response in termsof repair or overpack replacement in the event that small defects over the 5 mm limit arefound.
Based on the assessment above, it was concluded that the structural integrity of theoverpack was not threatened by any of these evaluation items.
(viii) Overpack production
The fundamental practicality of production of the overpack as specified in Figure 4.4-6 has
been confirmed by state-of-the-art fabrication demonstrations [6] [48]. Figure 4.4-7 illustrates
full-scale fabrication of the overpack by forging [49] [50].

4-37Figure 4.4-7 Production of a carbon steel overpack by forging (Source: (a), (b) [49]; (c) [50])In terms of welding the lid, full-scale demonstrations of TIG (Figure 4.4-8), MAG andelectron beam welding have been carried out. Here, TIG welding is described, which has been
confirmed to be able to be produce a defect-free 190 mm weld of a drop lid structure [51].

Figure 4.4-8 TIG welding test (Source: [51])The time required for welding the lid is between 24 and 33 hours, although this can besignificantly reduced by using up to four torches simultaneously [20]. For the design of thesurface facilities in Section 4.6, TIG welding of lids is assumed. In terms of alternative
welding options, a fusion failure occurred at the bottom of the lid gap in the full-scale test of
MAG welding [52]. To consider adopting this method, it would be necessary in the future todevelop technology to improve quality by optimising the welding conditions. For electron
beam welding, it has been confirmed that welds of up to 80 mm are possible for the drop lid
structure [53]. As described in Section 4.4.1 (2) (iv) (c), the actual required welding depth is
between 26 and 40 mm, so electron beam welding could also be applicable.
Section 4.4.1 (2) (iii) (a) noted that, since the environmental conditions for near-neutral
SCC cannot be precluded, reduction of the weld residual tensile stress would be beneficial.
Post-Weld Heat Treatment (PWHT) and External Induction Heating Stress Improvement(EIHSI) can provide such residual stress reduction, as shown in Figure 4.4-9 [44] [54].
4-38Residual radial stress (MPa)Residual radial stress (MPa)

Before PWHT
After PWHT (at 600 °C for 6 hours)
After PWHT (at 600 °C for 4 hours)

Before EIHSIPost welding heating at 600 °C
QuenchingAfter natural air-cooling down to 150 °C

Depth from lid surface (mm)Depth from lid surface (mm)(a) Example of residual stress reduction by
PWHT(b) Example of residual stress reduction
by EIHSI

Figure 4.4-9 Residual stress reduction examples ((a) [54], (b) [43])According to the results, in the corrosion allowance layer (14 to 21 mm from the surface),
the tensile residual stress (up to about 390 MPa in the overpack radial direction) occurring
before the post heat treatment can be reduced in the case of PWHT to about -30 MPa, whereas
in the case of EIHSI, this is converted to compressive stress field of about -330 to -200 MPa.
It can be noted that the temperature involved for heat treatment (~ 600 °C) [54] lies below thatat which devitrification of HLW is a concern [55] [56].
With regard to inspection technology for weld defects, confirmation of the applicability ofultrasonic flaw detection to welds with a depth of 190 mm is in progress. Based on
confirmation by destructive testing, ultrasonic tests detected defects with size 2 to 3 mm ormore with ≈ 100% efficiency in the weld with 190 mm deep [57].
(ix) Period over which overpack safety functions are assured
From the above, the overpack thickness of 190 mm specified is about 70 mm above that
required to maintain the function of the overpack to prevent contact waste and groundwaterfor 1 ky (in the worst case of H12V in plutonic rock this is 121 mm – see Table 4.4-4). As
discussed in detail in Supporting Report 4-13, the earliest estimated time to overpack failure
was approximately 17,000 years. However, since the assessed lifetime is much longer thanthat required, there may be room for optimising the overpack design, although such
optimisation must consider the performance of the other barriers as well. Modifications of thedesign will be carried out in future steps of the repository development programme.
(3) Buffer design
As for the overpack, this section is based on the H12 design. It is understood that these
designs may be modified in future NUMO design work.
(i) Design requirements and evaluation items for the buffer
The design requirements for HLW buffer which assure the safety functions (as discussed in
Section 4.2-4) are shown in Table 4.4-6. These result from an assessment (in Supporting
Report 4-14) of top-level requirements for limiting RN release, inhibiting colloid migrationand limiting RN transport.
4-39

Table 4.4-6 Design requirements for the HLW buffer
Designrequirements

Objectives

Specifications

Low
permeability

Restricting groundwater advection in the buffer,
reducing overpack corrosion and restricting
transport of RNs

Material, effective clay dry
density

Colloidal
filtration
capacity

Preventing RN migration in a colloidal phase

Material, effective clay dry
density

Self-sealing

Sufficient swelling to fill any gaps left after
construction

Material, effective clay dry
density,thickness

Self-healing

Ability to close any openings generated after
emplacement, e.g. due to gas breakthroughMaterial, effective clay dry
densityEngineering
practicality

Ability to manufacture and install to required
quality levels based on existing technologies or
those reasonably expected in the near future

Material, effective clay dry
density, fabrication plan

Prevention of
microbial
effects

Suppressing microbial activity in buffer material
that could increase overpack corrosion

Material, effective clay dry
density

PhysicalbufferSufficient plasticity to protect the overpack by
reducing the impact of mechanical perturbations

Material, effective clay dry
density, thickness

Requirements important for post-closure safety can be broken down into direct buffer
design requirements for low permeability, colloidal filtration, self-sealing (closing of originalgaps), self-healing (closing of gaps after perturbation of the buffer), RN sorption, etc.
Furthermore, in addition to those that contribute to overpack design requirements, such as
limitation of microbial activity and physical buffering, design requirements necessary toensure engineering practicality have to be added. The buffer material selection, design(geometry, thickness) and emplacement concept are then developed on the basis of these
design requirements.
As was the case for the overpack (Section 4.4.1 (2) (i)), the impacts of buffer loss offunction have to be considered in terms of performance of the entire disposal system(coupling buffer design to that of other EBS components), both before (Chapter 5) and after
closure (Chapter 6). Guidance on such coupling is again provided by the FEP list (Section
6.2.1, Supporting Report 6-4), which identifies further issues to be considered, such as thoseshown in Table 4.4-7.

4-40

Table 4.4-7 Evaluation items for HLW buffer in the current assessment
Evaluation item

Objectives

Mechanical stability during
evolution of the EBS

Demonstration that long-term evolution of stresses due tocorrosion, weight of the overpack and creep of rock does not
significantly degrade the required functions of the buffer

Mechanical stability in the eventof earthquakes

Demonstration that overpack and buffer do not losemechanical stability in the event of large-scale earthquake
motions

For example, the effect on the buffer due to heat from the waste is considered during
repository depth selection in Section 4.3 and tunnel design requirements in Section 4.5.2 (1).
From the standpoint of assuring long-term performance of both overpack and buffer, theevaluation items specified focus on mechanical stability, both in terms of evolution of the
EBS and impacts of major earthquakes. More information about these design requirements
and evaluation items is given in Supporting Report 4-14.
(ii) Choice of buffer material
Compacted bentonite is selected as a material that can satisfy all buffer requirements for therange of siting conditions assessed. A number of bentonites of different origins and properties
have been studied [58]. Kunigel V1, which is a Na-type bentonite, has been well investigatedin Japan and therefore selected as the buffer material for the reference specification. The
permeability and swelling properties of bentonite are affected by the salinity of groundwater.
As shown in Table 4.2-8, salinities in the model groundwater range from 1 × 10-3 to 2 × 10-1
mol/l, and thus, both “fresh” and “saline” conditions are considered when setting the buffer
specifications. Also, due to the diffusion of Ca from groundwater and concrete present in thedisposal tunnel, exchangeable Na ions may be replaced with Ca ions, causing the initial
performance to degrade with time. In this report, it is conservatively assumed that thisconversion may occur; however this assumption may be reassessed with a more detailed
analysis in the future if conversion to Ca bentonite is shown to have a large impact on
performance.
In addition, with regard to the composition of the buffer material, the H12 report considers
mixing of silica sand with bentonite. Such a mixture tends to lower the swelling pressure, but
is advantageous in terms of compaction characteristics, thermal conductivity, resistance toerosion, cost, etc. [59] [60]. However, this will be studied further and other compositions may
also be considered in the future.
(iii) Setting the density and thickness of the buffer
Design requirements to assure buffer performance (see Table 4.4-6) are effective clay dry
density (i.e. the dry weight of bentonite per unit of buffer) and thickness, for fresh and salineconditions, which are tailored to H12V and PEM concepts, as illustrated in Figure 4.4-10 anddescribed in more detail in Supporting Report 4-15.

4-41

(a) H12V(b) PEM
Figure 4.4-10 Specification ranges of buffer for current basic design requirements (Rs is thesilica sand content)

Here, Figure 4.4-10 plots performance objectives as a function of effective dry density ofclay and buffer thickness, assuming swelling clay fills any gaps left after buffer emplacement.

4-42

A gap of 40 mm was assumed between rock and buffer, and 20 mm between buffer andoverpack in the H12V case, while only the latter was assumed in the case of the PEM5.
To conservatively assess impacts of ionic strength, salinities up to that of seawater wereconsidered. In the formulation of the buffer, the mixing ratio of silica sand (Rs) was set to30% by weight on the basis of ability to produce high quality compacted blocks [61] (seeSection 4.4.1 (3) (v) (a)), while also ensuring that other requirements were met. The effective
clay dry density, used as an index of the buffer specification, is the dry density of bentonite in
the bentonite/sand mixture. It should be noted that the relation between effective clay dry
density and properties depends on the type of bentonite. Curves and relationships presented inthis chapter are for Kunigel IV, unless otherwise stated.
From Figure 4.4-10, the specification of effective clay dry density established for
freshwater conditions has to be above the design requirement ⑤ (lower limit of colloidal
filtration) and below ⑥ (upper limit for engineering practicality. For saline conditions (orCa-type of bentonite, as shown in Figure. 4.4-11), the range lies again below ⑥ but with thelower limit either set by either ① (lower limit of assured self-sealing) or ③ (lower limit ofassured self-healing), depending on thickness.

Figure 4.4-11 Relationship between effective clay dry density and the equilibrium swellingpressure of bentonite (from Suzuki et al., 1992 [62], Maeda et al., 1998 [63], Suzuki & Fujita,
1999 [64], Sugita et al., 1999 [65], Sasakura et al., 2003 [66], Kikuchi & Tanai, 2005 [67], Toida etal., 2005 [68])

To provide a common specification that would apply to a wide water chemistry range, themore restrictive range for saline conditions is used to select effective clay dry density of about
1.4 Mg/m3, which is in about the middle of the range. For this density, the required thicknessof buffer to meet the design requirements is seen to be (Figure 4.4-10) about 450 mm for
H12V and about 120 mm for the PEM.

5

The buffer is assumed to be packed in the PEM container such that there is no significant gap between thebuffer and the handling shell (see Section 4.4.1 (3) (v) (b)).

4-43

In addition to the above design requirements for buffer thickness, physical protection toassure the structural integrity of the overpack, has also to be considered during the period (> ≈1 ky) when its key safety function is ensuring complete containment (see Section 4.4.1 (2)(iii) (b)). As mentioned, in addition to buffer swelling pressure and hydrostatic head at
repository depth, stability assessment considers additional loads from expansion of overpack
corrosion products and bedrock creep (in the case of Neogene sediments). The details of theassessment are given in Supporting Report 4-16. Figure 4.4-12 (for 2 values of corrosion
allowance (17 and 18 mm) for H12V in Table 4.4-4) shows the relationship between thebuffer material thickness and the required pressure resistance thickness of the overpack lid.

Figure 4.4-12 Relationship between the pressure-resistant thickness required for the
overpack lid and the buffer thickness

From the viewpoint of safety, it is preferable to minimise the variation of the required
mechanical stability thickness and hence this is taken to be 400 mm or more. From Figure
4.4-10, the minimum buffer thickness for H12V is about 450 mm, so this also covers thisrequirement. However, for the PEM the minimum thickness was 120 mm, so this wasconservatively increased to 300 mm, bearing in mind the significantly lower corrosion in the
PEM case (11 mm).
Based on these results, the reference specifications derived are presented in Table 4.4-8. As
in H12, the buffer thickness was set at about 700 mm to have a sufficient safety margin over
the required minimum thickness (450 mm for H12V, 300 mm for the PEM), bearing in mind
uncertainties involved in quantifying long-term performance.

4-44

Table 4.4-8 Buffer material specifications (formulation and density) for HLW
Formulation(Dry weight ratio)

After saturated swelling(Set specification)*1

H12V
Bentonite 70%
Sand 30%
PEM

DensityDry density= 1.6 Mg/m3
Effective clay dry density *2≈ 1.4 Mg/m3

Production condition*3(reference)
Dry density ≈ 1.8 Mg/m3
Effective clay dry density ≈ 1.6
Mg/m3
Dry density ≈ 1.7 Mg/m3
Effective clay dry density ≈ 1.5
Mg/m3*1 Dry density: dry density of the mixture of bentonite and silica sand*2 Effective clay dry density: dry density of bentonite excluding silica sand
*3 State at the time of production: before buffer material swells following infiltration of groundwater

However, as in the case of the overpack, there may be prospects for optimising this design.
In the case of adopting the PEM option, the production constraint is less, since the impact ofgaps is smaller compared to H12V blocks.
(iv) Evaluation of long-term buffer stability
Based on Table 4.4-7, the buffer specifications given above need to be evaluated todetermine mechanical stability, both in terms of evolution of the EBS and impacts of major
earthquakes. Details of the analyses are given in Supporting Report 4-17, while key results arepresented below.
After repository closure, the density of the buffer changes as the overpack graduallycorrodes, with expansion of overpack corrosion products, and as a result of bedrock creep (inthe case of Neogene sediments). As buffer is deformed and shear stress increases, its
continued ability to support the weight of the overpack has to be assessed, considering also
consolidation (secondary consolidation) of the plastic buffer due to slow overpack sinking.
Figure 4.4-13 shows the results of such an analysis.

Figure 4.4-13 Overpack sinking over time

4-45

The amount of overpack settling is larger for freshwater conditions, as plasticity of buffer
decreases with increasing salinity. In the case of H12V, sinking is about 11 mm after 1 ky and
17 mm after 100 ky, while for the PEM it would be about 6 and 9 mm, respectively, at thesetimes. Even though sinking continues after 100 ky, the rate decreases with time due tosecondary consolidation and hence the increase in sinking up to 1 My would only be in theorder of several mm. Thus, the buffer material is believed to maintain the minimum necessary
thickness over the assessment timescale with a good safety margin. Ca bentonite is less plastic
and hence, even in freshwater, sinking would be less than the Na bentonite/saline water case[64], [67].
Figure 4.4-14 shows the stress ratio severity distribution for buffer in plutonic rock after
corrosion product swelling is complete (very conservatively assumed to be ≈ 20 ky - obtainedby corrosion product swelling analysis). The stress ratio severity is defined as normalising theeffective stress ratio (q/p', where q is the deviator stress and p' is the mean effective stress)
with the effective stress ratio at failure (the critical state parameter M) and it is an index
showing how close conditions are to failure through shearing – see Supporting Report 4-17
for further details).

Figure 4.4-14 Distribution of stress ratio severity of buffer around overpack(plutonic rock, fresh water conditions)

The stress ratio takes values from 0 to 1, with risk of failure increasing as it gets closer to 1.
For H12V, marginal regions are limited to overpack corners (actually rim, but looks likecorners in the 2-D profile) while the PEM does not approach this critical state. In any case,there seems no risk of damage extending throughout the entire buffer thickness. From theanalyses conducted, this conclusion applies also to other rock types.
To assess impacts in the event of large earthquakes, seismic response analysis of dynamic
EBS stability was carried out. This was based on observations of ground motions from Japan's
largest earthquakes, such as the 2011 Tohoku-Pacific Ocean earthquake, together withconservative conditions for the response analysis [46] [47]. This confirmed that, for all
analysed cases, sufficient safety margins assure that stresses generated would not significantly
damage either buffer material or the overpack.
In addition to ground motion from distant earthquakes, movement of existing faults needs tobe considered. As mentioned in Chapter 3, areas containing large active faults are excluded
during site selection. Furthermore, as shown in Supporting Report 3-35, the extent of
displacement due to a single fault movement is related to its length, being about 80 cm for a
fault length of 10 km, the size of largest faults assumed to be present at the repository site, but
avoided during layout (see Section 4.1.3). Thus, displacements large enough to make the4-46overpack contact the rock would not be expected to occur for a buffer thickness of 70 cm.
However, even if the probability is very low, Section 6.4.3 (2) considers such rare event
scenarios to assess if radiological impacts could be significant.
To improve understanding, modelling studies have been carried out for different positionsof the moving fault. These indicate that, as long as the displaced overpack does not contact
rock, a sufficient plasticity of the buffer will protect the overpack from serious damage [69][70] [71] [72]. As shown in Figure 4.4-15, shear deformation by faulting causes the overpack
to rotate in the buffer, with concomitant stress. In the case of the analyses shown in thisfigure, the von Mises stress produced on the inner surface of the overpack is at most 16 MPa,
additional to the total buffer and the pore water pressure of 25 MPa, giving a total stress of 41
MPa [71]. Since this is only about 25% of the yield stress of the overpack (carbon steel
SF340A), absence of failure can be assured with a sufficient safety margin.
Shear displacement
Case

Bufferdeformation

TensileDistributionof themaximum
principal
stress inbuffermaterial

Distributionof vonMises stressonoverpack

Figure 4.4-15 Shear stress sensitivity analysis of the HLW EBS (Source: JAEA, 2014 [71])The models showed that the structural integrity of the overpack can be assured even forshear deformation up to 140% of the buffer material thickness (1 m) [70]. For the buffer itself,
fractures may form due to the occurrence of shear bands along the shear surface, but selfhealing of such fractures (as with those generated by gas breakthrough) and recovery of
hydraulic barrier properties have been confirmed by experiments [73].
(v) Buffer production(a) Compacted blocks
Figure 4.4-16 shows a buffer block 35 cm in height being manufactured by compaction
with a 2000 t uniaxial press machine. It was verified that the dimensional change of the buffer
block after removal from the mould (dimensions change gradually after the compaction force
is released) can be controlled to some extent by varying the time, speed and holding time ofcompaction, based on the material conditions.

4-47

Figure 4.4-16 Buffer block fabrication by uniaxial press (Sources: (a) [61]; (b), (c) [74])It is known that cracks can occur in the buffer block during mould removal, so techniquesshould control such cracking. For a mixture of 70% bentonite and 30% silica sand with a 9%
water content, the goal of a dry density of 1.9 Mg/m3 was obtained in all test cases (10 cases)
and it was confirmed that a high-quality block, with no defects such as cracks, could be
produced [61]. In addition to the results from single-axis press machines, block manufacture
by cold isostatic pressing (CIP) has also been confirmed [61] [75].
(b) PEM buffer
PEM buffer manufactured by compaction using a pneumatic ram is shown in Figure 4.4-17[48]. In this method, buffer as powder is poured into a steel shell ring corresponding to asection of the PEM handling shell and compacted by the ram to produce ring-shaped buffer
segments.

Figure 4.4-17 PEM buffer production ([48])Compaction occurs in steps and, to date, a step size of ≈ 50 cm has been tested. Since buffer
compaction could occur within the actual PEM shell, there would be no gap at the interface.
Tests were conducted on material with 18% moisture content and a mix of 70% bentonite and30% silica sand. In the test, the buffer dry density goal was 1.9 Mg/m3, but the average value
after fabrication was slightly lower (at 1.8 Mg/m3), but still above the 1.7 Mg/m3 specified inTable 4.4-8. Low density measured near the inner side of the ring was attributed to loss of the
4-48

compacting pressure due to insufficient rigidity of the inner frame. Therefore, it wasconsidered that this could be solved by increasing the rigidity of this component. Althoughdemonstrating fundamental practicality, further R&D will focus on developing an optimised
approach to PEM buffer production, in terms of both operational logistics and quality
achieved.
(4) EBS specification
Sections 4.4.1 (2) and (3) provide specifications of the HLW overpack and buffer. Figure
4.4-18 brings these together (with specifications of disposal holes from Section 4.5.2 (2) (i)).
For H12V, the surface dose rate of the overpack is 11 mSv/h (30 years interim storage; see
Supporting Report 4-18). Therefore, in order to allow for the option of manual backfilling of
the disposal tunnel after waste emplacement, backfill material with a thickness of 1 m isrequired on top of the buffer (from H12).

(a) H12Vb) PEM
Figure 4.4-18 HLW EBS specifications

Details of the design and specifications of the PEM are given in Supporting Report 4-19.
The PEM is assumed to be fabricated on the surface and then transported and emplaced
underground. Accordingly, the handling shell must be able to withstand stress or impact
forces during transport and emplacement without loss of structural integrity, requiring 28 mm
thicknesses of carbon steel. In addition, the gap between the overpack and buffer is filled withbentonite pellets to prevent damage of the buffer due to vibration during transport. However,from the viewpoint of conservative specification to assure self-sealing (Section 4.4.1 (3) (iii))
20 mm gaps were assumed unfilled. Due to additional shielding by the buffer and handlingshell, the surface dose rate for the PEM is 3.1 Sv/h (after 30 years interim storage: see
Supporting Report 4-18).
4.4.2 Design of the TRU EBS(1) EBS objectives
TRU waste safety functions pre- and post-closure have been assigned to specificcomponents of the EBS (Table 4.2-1, Table 4.2-5), as summarised in Figure 4.4-19. These
4-49

provide input to design of waste packaging, infill between waste packages and buffer (whenpresent). The TRU waste components and primary containers are specified in Section 4.2.2(2). For disposal, these are conditioned within waste emplacement packages, as in the TRU-2
report and discussed further in Section 4.2.3 (2) (ii). Two types of waste package are assessed,
waste package A designed for emplacement with a fork lift and a more robust type, waste
package B intended to be emplaced by a gantry crane [76]. Alternative waste package designs(e.g. with greater longevity for Gr.2 wastes) are being considered but development is at an
early stage and hence only the well-studied packages A and B are assessed in this report.
Drums

Stainless steel
canisterBox container

Waste form (TRU waste)

InfillCompacted
hulls andend-pieces

The waste form is the waste matrix consolidated or
grouted into a drum, stainless steel canister orbox container• Restriction of RN leaching

Waste package
The waste package consists of the waste form, filler andwaste package container• Restriction of RN leaching;“Prevention of contact waste and groundwater”“Sorption of RNs” in cementitious infill
Note: The box container will be disposed of directly

Waste package container

Gap infill (mortar)• Restriction of RN migration;“Sorption of RNs”

Backfill
Structuralconcrete

Buffer (Compacted bentonite)• Restriction of RN migration;“Limit release from EBS by advection”“Colloidal transfer inhibition”“Sorption of RNs”

Figure 4.4-19 TRU EBS components and safety features(2) Design of TRU waste package
TRU waste is classified into 4 groups and conditioned in drums, canisters or box containers
as indicated in Table 2.1-2. This waste has a low thermal output compared to HLW, allowing
transport and emplacement of multiple containers to improve operational efficiency. The
TRU-2 report illustrated emplacement packages, each containing four 200 litre drums, fourcanisters or two BNGS 500 litre drums (now no longer being considered for disposal), withvoids filled with mortar. These emplacement packages, together with the box containers, canbe readily stacked in disposal vaults. In the following, after showing the design requirements
and evaluation items of the waste package, design examples of each component are shown.

4-50

(i) Waste package design requirements
The design requirements of the emplacement waste packages are shown in Table 4.4-9.
Operational safety requirements (safety functions) focus on preventing leakage of RNs during
all operations (Table 4.2-1). Corrosion resistance and structural integrity are set as design
requirements to ensure these safety functions. In addition, practicality of fabrication and theability to emplace the package by remote handling are set as implementation design
requirements. The key post-closure safety requirement (safety function) is restriction of
releases of RNs (Table 4.2-5).
Table 4.4-9 Waste Package Design Requirements
Designrequirements
Corrosion
resistanceObjective

Specifications

The function of preventing leakage of RNs during operation
is not impaired by corrosion

Structural
integrity

During operation, the function of preventing leakage of RNs
is not impaired by applied loads

Production
practicality

Package can be manufactured by existing technology or
technology which will be realised in the near future

Remote waste
packaging

Waste can be placed in a package, which is infilled and a lid
added by existing remote handling technology or
technology which will be realised in the near future

Container material
thickness
Container material
thickness, waste
package infill, handlingfeatures
Container material,
shape
Container material,
internal dimensions,waste package infill, lid,
lifting features

Remotepackageemplacement

Waste packages can be emplaced in disposal vaults byexisting remote handling technology or technology which
will be realised in the near future

Shape, lifting features

After the closure of the disposal vault, groundwater will infiltrate and gradually saturate
buffer, backfill and other porous materials, while the higher thermal output of waste Grs.2 and4H will decrease as short half-life radionuclides decay. This transient period is considered to
be up to around 300 years, based on previous studies [77]. As discussed in Section 6.3,
quantification of RN release (identified in Table 4.2-5) for evolving conditions of temperatureand saturation during this transient is associated with great uncertainties. Therefore, complete
containment is particularly valuable for the first 300 years. This is captured in Table 4.4-10,
along with a breakdown of the objective in terms of design requirements.
Table 4.4-10 waste package evaluation items
Evaluation item
Complete containment

Objective
Have corrosion resistance and structural integrity for a
specified period of time after closure

The details of the above design requirements and setting resultant evaluation items areprovided in Supporting Report 4-20.
(ii) Design of waste package containers(a) Container for waste package A
The TRU-2 report reference emplacement packaging design (now termed waste package A)
included a steel shell (rectangular cuboid) of thickness 5 mm, infilled with cementitious
mortar without a lid (used for all primary waste containers, except for the box containers).
4-51

Although here the containment function is unrelated to corrosion resistance, the primary waste
is enclosed and hence the Table 4.2-1 safety function of preventing leakage of RNs during
operations would be provided.
With respect to minimising loading that threatens structural integrity, the TRU-2 report and
NUMO [1] have illustrated remote-handled emplacement with a forklift, which is facilitatedby the gaps at the bottom of the waste package container (Figure 4.4-20).

Figure 4.4-20 Waste package A specifications (the container for MHHRW (miscellaneous higher
heat reprocessing waste) has yet to be specified)With regard to the load during stacking, the structural integrity is ensured by the strength of
the infill in the waste package as described later, not the waste package container [10].
(b) Container for waste package B
The waste package B container (Figure 4.4-21) was designed for lifting by a crane for
emplacement in the disposal vault. In the design, the shape of the waste package container issimilar to that designed for medium-depth disposal (Rokkasho L1), which also assumes
emplacement using a crane [21] [78].
Specifically, for SM400A/JIS G 3106 structural quality steel, the required corrosion
resistance and strength for lifting and stacking can be provided by a container thickness of
50 mm and appropriate lifting structures. To reduce the risk of releases in the event of drops,
the upper lid is sealed by welding. Studies of drops in Section 5.4.2 (1) (ii) show that such acontainer will maintain its integrity even for drops from the maximum lift height of 8 m.
Corrosion during the operational period, assuming exposure to air for a maximum about 8
years from waste packaging until vault infilling, would result in an average corrosion depth
is 0.5 mm, which would have little impact on the total wall thickness of 50 mm. Further
details of waste package B design and structural integrity studies are provided in Supporting
Reports 4-21 and 4-22.

4-52

Figure 4.4-21 Waste package B specifications (the container for (c) has yet to be specified)Grs.2 and 4H have higher radioactivity and could generate hydrogen gas by radiolysis ofwater contained in the mortar infill. In the case of waste package A, generated hydrogen gas
would pass through the mortar infill without a significant pressure build-up. For waste
package B, which is sealed by welding, the carbon steel container should be able towithstand any pressure from hydrogen gas build-up. To evaluate the pressure tightness, theinternal pressure from hydrogen gas build-up in this waste package was calculated from theamount of gas generated by radiolysis of water in the mortar infill between waste drums orcanisters, mortar infill or concrete matrix within the waste packages for all groups. If dry
sand is used instead of the mortar-infill in the packages, the pressure from hydrogen gas may
be reduced. This alternative infill material was also considered for Grs.2 and 4H.
Hydrogen gas build-up within the waste packages for Grs.1, 2, 3 and 4L calculated until
repository closure, was within that considered acceptable to ensure the structural integrity of
the waste packages (i.e., within the pressure threshold). In the case of Gr.4H, the applicationof an alternative design using dry sand and a high strength steel (e.g., JIS G3115 steel plate
SPV490 for pressure vessels) was more effective in ensuring the structural integrity. In thiscase, a separation steel wall should be included in order to prevent heterogeneous
distribution of dry sand in case of inclination during operation (see Figure 4.4-21 c). Detailsof hydrogen gas production and associated evaluation are described further in Supporting
Report 4-22.
The above evaluation can be updated if, at any point in the future, the properties of thegenerated waste change. It may be worth mentioning that reducing hydrogen gas generation
will also reduce the risk of rupturing the waste package container. A reduction of moisture
content in the mortar, or utilisation of dry sand instead for the infill, would be effective inthis regard and may be considered further in the future.
(iii) Design of waste package infill
Remote-handled waste package infill using a tailored mortar (for example, in terms of
fluidity, setting characteristics) is described in the TRU-2 report and [11].
4-53For structural integrity during stacking, an infill strength of 30 MPa is required [79]. Thisand other required characteristics can be obtained with the formulation shown in Table 4.411, which does not include minor components introduced to facilitate production.
Table 4.4-11 Waste package infill formulation [79]Mortar (kg/m3)

Water (kg/m3)

Fine aggregate (kg/m3)

483

266

1,449

In the case of waste package B, the steel container has a thickness of 50 mm and isdesigned to withstand loads during stacking. For this reason, it is not necessary to ensure thestructural support of the mortar infill against the stacked load. For groundwater pressureloads on the container after repository closure, the mortar infill may act as mechanicalsupport. However, in the case where dry sand is used as an infill for Gr.4H (see Section 4.4.2(2) (ii) (b)), structural integrity should be ensured by the waste package container alone.
The mortar infill may reduce groundwater penetration and hence slow down RN leaching
and, additionally, may retard RN migration by sorption [11]. Cracks that form due to mortar
shrinkage may affect RN migration and will be discussed further in Chapter 6.
Initially, the calculation the impact of gas was not considered. Subsequent calculations
suggest that the impact of gas generation due to corrosion is negligible, but this is still an open
issue, as discussed further in Chapter 6.
(iv) Waste package specifications
From the above, waste packages A and B, for each type of primary waste container, aregiven in Table 4.4-12 (illustrated in Figures 4.4-20 and 4.4-21, respectively). The external
dimensions of waste packages A and B for specific primary containers are effectively thesame and maximum thermal output and weight of these are provided in Supporting Report 421.
Table 4.4-12 Specifications of waste packages A and B
Material for waste
package container (JIS)

Waste
Gr.

Primary waste
container

Primary
containers
perpackage

1

Drum

4

MortarMortar

SM400A

SM400A

2

Canister

4

MortarMortar

SM400A

SM400A

3

Drum

4

MortarMortar

SM400A

SM400A

Drum

4

MortarMortar

SM400A

SM400ABox container

1

-

-

SM400A

SM400A

Drum

4

MortarMortar or
dry sand

SM400ASPV490MHHRW*

2

MortarDry sand

SM400ASPV490MHHRW

2MortarMortar

SM400ASPV490Waste package infill
Waste
package A

Waste
package B

Waste
package A

Waste
package B

4L

4H* Miscellaneous higher heat reprocessing waste

4-54

In the case of the waste package A, RNs can be contained by mortar infill, while the RN
containment during operation is improved in the case of waste package B due to the welded
lid. In addition, waste package B may also improve post-closure containment, even if this is
not rigorously quantified. The corrosion resistance and mechanical integrity of waste
package B during the post-closure phase are discussed further below.
(v) Evaluation of corrosion and structural integrity after closure(a) Waste package A
In the case of waste package A, first contact by infiltrating water occurs after surrounding
buffer and backfill have saturated and water then penetrates the structural concrete and interpackage filling. Even then, water must saturate the mortar and the primary waste container
must fail before releases of RNs from the waste will occur. It is reasonable to expect that thiswill take an extended period of time, but is very difficult to quantify. Thus, although somescoping calculations indicate this might take about the target 300 years [77], there is great
uncertainty in this – leading to the very conservative assumption made in safety assessmentsthat RN release starts soon after recovery from the initial transient, as described later in
Chapter 6.
(b) Waste package B
Although the initial conditions for waste package B are similar to those described above forwaste package A, as package B is thick-walled with a welded lid, it can be assured that noRNs are released until the package fails.
Because it is surrounded by mortar, the ambient pH is high and it is expected that corrosion
would be even lower than that considered in Section 4.4.1 (2) for the HLW overpack design.
Under highly alkaline conditions, the average corrosion rate of carbon steel in the initially
oxidising environment would be < ≈ 1 μm/y [80], while, in a reducing environment, < ≈ 0.1
μm/y has been reported [81]. Since it is difficult to estimate the time for transition fromoxidising to reducing conditions within the disposal vault, the average corrosion depth for 300
years was determined conservatively, using a corrosion rate of 1 μm/y. The average corrosion
depth would thus be 0.3 mm, which, taking account of non-uniform corrosion [82], would
give a maximum corrosion depth of 5.3 mm, indicating penetration of the 50 mm thick wall
would not occur.
Incidentally, the corrosion of carbon steel in an alkaline environment is not accelerated atradiation doses up to at least 25 Gy/h [83], so the maximum surface dose of about 1 Gy/h forwall thickness of 50 mm [11] indicates that this is not a concern. Further details of thisevaluation are provided in Supporting Report 4-23.
An average corrosion rate of 0.3 mm over 300 years is not expected to significantly impact
the structural integrity and hence mechanical analysis was carried out assuming the full wall
thickness of 50 mm. The stress applied to the waste package container walls was calculated
based on external and internal forces. The external force applied to the container, in the caseof plutonic rocks and Pre-Neogene sediments for a repository at 1,000 m depth, includes ahydrostatic pressure of 10 MPa (maximum case). The internal pressure caused due toradiolytic hydrogen gas within a 300 y period was also considered. As a result, the stressapplied to container walls was highest in the case of Gr.1 (no significant radiolysis) at 10 MPa.
4-55

From an elasto-plastic analysis, distortion of the side plate and bottom plate of the waste
package in contact with the solid mortar was shown to be extremely small. The top plate wasplastically deformed due to the 50 mm gap at the top of the container, but the amount ofdistortion did not reach the critical strain. From these results, it is considered that the waste
package container does not fail in this case. If required, it is possible to suppress deformation
by measures such as reinforcing the top plate and increasing the thickness of the waste
package container.
It is therefore assumed in the post-closure safety assessment for waste package B that RN
release starts at least 300 years after closure. More realistically, demonstration of greater
longevity may be possible – potentially on timescales in the order of thousands of years. For
this, more detailed evaluation of structural integrity as the corrosion depth increases would beneeded. It should also be noted that stress-corrosion cracking at areas where residual welding
stress occurs was ignored in the evaluation and hence R&D to reduce the risk of SCC may beneeded for this package. More information on this evaluation of the structural integrity is
included in Supporting Report 4-24.
(vi) Production of the waste package container
The waste package container can be manufactured by bending steel plates, as considered inthis report, or by casting. These methods of fabrication have been confirmed by full-scaledemonstrations associated with the L1 interim depth repository project [76] [84] (e.g. Figure
4.4-22) and hence required technology is assumed to be available.

Figure 4.4-22 Fabrication tests of the waste material container ([76])(3) Design of infill for the disposal vault(i) Design requirements for gap infill
The design requirements for vault infill are shown in Table 4.4-13. An operational safety
requirement is to provide radiation shielding for any case when workers are present (see Table
4.2-1). The infill complements self-shielding by the waste packages themselves, reducing
doses to any workers present, particularly at the time of construction of the buffer/backfillingbarriers above the emplaced waste.

4-56

Table 4.4-13 Gap infill design requirements
Designrequirements

Objectives

Specifications

Shielding

Reduces dose to workers, if present for backfilling work

Thickness of gap infill
around waste packages

Have sufficient fluidity for gap filling between waste
packages
Resistance to separation during pumping

Ingredient mix, infilling
plan

Separation of waste packages from each other and walls
during emplacement

Gap infill thickness

Remoteoperated
emplacement

Sorption of RNs was set as a post-closure safety requirement (safety function), (Table 4.25), although infill may contribute also to restricting RN solubilities. As noted in Section 4.2.4(2) (ii), there is no need to consider the thermal alteration of cement minerals for waste Grs.1,
3 and 4L when assessing this requirement, although this is a factor for the groups with higher
heat output.
In order for the safety function to be assured, mortar must be emplaced in a quality assured
manner that fills all void space and thus this is a consideration for remote emplacement,
which is set as a design requirement to assure practicality.
(ii) Infill material
The composition of the mortar infill must satisfy the above design requirements. As
described in TRU-2 and NUMO (2011) [11], mortar was chosen as an infill material that
could be easily emplaced by remote-handling, avoiding radiation exposure from TRU waste.
Reduction of RN release by the mortar in the post-closure phase is also required (See Table
4.2-5), while mechanical strength should be sufficient to assure stability of stacked A waste
packages. RN release reduction is provided by the low permeability and sorption capacity ofmortar, compared to an unfilled case. An example of composition of mortar infill which meetsthe mechanical strength criteria of 30 MPa has been reported [79] and is given in Table 4.4-11.
Workability with remote-handling was also checked for this mortar [79]. Fracture formationin mortar infill is likely but, if the water flow rate is sufficiently low, RN migration may belimited, as discussed further in Chapter 6.
For waste package B, RN release is constrained by both the metallic container and themortar infill. The mechanical strength for stacking is provided by the container walls rather
than the mortar infill, due to the void space that remains in the infilled package.
(iii) Thickness of infill between waste packages
Waste package clearances, between stacks and from the structural framework walls, areconservatively taken as 150 mm based on an assessment of swinging of waste packages
during emplacement using a gantry crane. The same clearance is also assumed for
emplacement using a forklift. There is certainly potential to reduce these values as equipment
designs are developed.
It is assumed that, after infilling, workers could be present to emplace the overlying
buffer/backfill. Buffer emplacement at the side of the structural framework is assumed to4-57

occur prior to commencement of waste package disposal (Section 4.4.3 (2)). However, if there
is a problem with the side buffer, manual removal and replacement is assumed.
For radiation shielding design, the effective dose limit for workers under normal conditions(50 mSv over 1 year, 100 mSv over 5 years) is used to calculate the infill thickness needed toprovide the required shielding. For backfilling the top of the vault, a 48 hour working week isassumed (equivalent to10 μSv/h), but only 10 hours per week working at the side (50 μSv/h).
Table 4.4-14 shows very conservative estimates of the required infill shielding thickness(assessment described in detail in Supporting Report 4-25). This distinguishes between waste
packages A and B due to the different container wall thicknesses (5 mm and 50 mm,
respectively), with the latter providing more self-shielding. Incidentally, Gr.1 requires noshielding but an infill thickness of 100 mm at the top is assumed for physical protection and
150 mm at the side due to emplacement considerations, as mentioned above.
Table 4.4-14 Gap infill thickness

Waste GroupTop (mm)
Required shielding
thickness
Package A

Package B

1

0

0

2

980

3
4L
(Drum)
4L(Box container)
4H(Drum)
4H(MHHRW)Set value*

Side (mm)Required shielding
thickness

Set value

Package A

Package B

100

0

0

150

860

1,200

220

110

300

510

550

600

0

0

150

900

940

1,050

250

140

300

440

430

450

0

0

150

700

740

800

80

0

150

780

960

1,000

220

120

850

* Although set values are all conservative, the safety margin is quite variable as thickness is also constrainedby vault geometry as discussed later.

4-58

(4) Design of buffer(i) Design requirements and evaluation items for the buffer
The TRU waste EBS includes a buffer for Grs.1, 2 and 4H, which have higher thermal
output and/or contain relatively high contents of poorly sorbing nuclides. As noted in Section
4.2.4 (2) (ii), buffer placed around the structural framework (Figure 4.4-19) has the functionof restricting RN transport. Design requirements and evaluation items, assuming use of
bentonite-based buffer, are shown in Tables 4.4-15 and 4.4-16. In the future, based on
feedback from post-closure safety assessment, the pros and cons of an external buffer will be
re-evaluated for all waste groups.
Table 4.4-15 Requirements for TRU waste buffer design
Designrequirements

Contents

Specifications

Low permeability

Restricting groundwater advection in the buffer,
reducing transport of RNs

Material, effective clay
dry density, thickness

Colloidal
filtration capacity

Preventing RN migration in a colloidal phase

Material, effective clay
dry density

Self-healing

Ability to close any openings generated after
emplacement, e.g. due to gas breakthroughMaterial, effective clay
dry density

Manufacturability
and engineering
practicality

Ability to manufacture and install to required quality
levels based on existing technologies or those which will
be realised in the near future

Material, effective clay
dry density,
emplacement plan

Table 4.4-16 TRU buffer evaluation items
Evaluation item

Objectives

Mechanical stability
during evolution of the
EBS

Demonstration that long-term evolution of stresses due to the weight of theconcrete emplacement structure and creep of rock does not significantly
degrade the required functions of the buffer

Mechanical stability inthe event ofearthquakes

Demonstration that overpack and buffer do not lose mechanical stability inthe event of large-scale earthquake motions

Note that there is no requirement for self-sealing of gaps as, based on full-scale tests, thebuffer construction process for TRU waste is not expected to result in such gaps [85]. TRU
waste buffer evaluation items and design requirements are discussed further in Supporting
Report 4-26.
(ii) Choice of buffer material
Compressed bentonite is selected as a material that can satisfy the various buffer
requirements. As for HLW, Kunigel V1 mixed with silica sand is selected as the buffermaterial to set the reference specification, with caveats as for the HLW case. As mentioned in
Section 4.4.1 (3) (ii), bentonite permeability and swelling pressure is influenced bygroundwater salinity, which varies as shown in Table 4.2-8. Potential alteration of Na-type
bentonite by exchange with Ca is also considered.
4-59

(iii) Setting buffer density and thickness
Table 4.4-17 summarises the ranges of buffer effective clay dry density that meet the basic
design requirements for low permeability, colloid filtration and self-healing for both fresh andsaline water (conservatively considering seawater) and also alteration to Ca-bentonite.
Table 4.4-17 Effective clay dry density requirements for TRU waste buffer
Effective clay dry density (Mg/m3)
Design requirements

Freshwaterconditions

Salineconditions

Ca typeLow permeability> ≈ 0.4> ≈ 1.1> ≈ 1.1

Colloidal filtration
capacity> ≈ 0.8> ≈ 0.8> ≈ 0.8

Self-healing> ≈ 0.6> ≈ 1. 2> ≈ 1. 2

Manufacturability andengineering practicality< ≈ 1.6 (field compaction)

The details for setting these limits are given in Supporting Report 4-27. In general, lower
density limits can be set for fresh water conditions, with higher limits being the same for bothsaline conditions and Ca-bentonite. All requirements can be met with effective clay dry
density of 1.2 Mg/m3, while the upper limit in terms of practicality of fabrication is 1.6
Mg/m3. Buffer thickness in the TRU-2 report was related to the RN release rate [10], with 1.0
m selected to provide a suitable performance margin. Additional studies on the thickness andbuffer density considering practicality and other safety functions will be carried out in the
future.
Average effective clay dry density is specified to be 1.4 Mg/m3 on the basis of sufficientperformance margins as well as other considerations. Bentonite is assumed blended with silica
sand at a mixing ratio of 30% to improve compaction performance. The dry density of 1.6
Mg/m3 thus gives the target effective clay dry density of 1.4 Mg/m3.
(iv) Evaluation of long-term buffer stability
After closure, the mechanical conditions in the EBS will slowly evolve as infiltrating
groundwater saturates all porous media; metals (especially waste packages and structuralrebars) corrode and swell; with also rock creep for the relatively low strength Neogene
sediments. Leaching of structural concrete will result in reduced strength, while interaction ofsuch leachate with buffer will cause its alteration. As for the HLW case, it is thus necessary toevaluate the long-term stability of the buffer in terms of both gradual evolution of the EBS
and also ground motion resulting from earthquakes. The details of these evaluations are given
in Supporting Report 4-28.

4-60

Figure 4.4-23 shows examples of the results after 100 ky of EBS evolution in terms ofresultant buffer dry density distribution for TRU waste Gr.2 in plutonic rock (EBS
specifications in Figure 4.4-25) and Neogene sediment. Geological data are taken from Table
4.2-6 and water chemistry from Table 4.2-8. In this simulation, the evolution of buffer
mechanical properties due to Ca exchange and smectite dissolution is considered in addition
to the evolving stress field resulting from loading by the structural concrete framework andswelling of metal corrosion products [86]. In addition, for Neogene sediments, bedrock creep
is taken into account.

Figure 4.4-23 Buffer dry density distribution (after 100 ky)

The buffer is compressed by expansion of inner EBS components and, for Neogene
sediments, external rock creep. Dry density of the buffer decreases only in very restricted
areas of tensile stress, with no significant volume below the minimum specified value of 1.4
Mg/m3 (effective clay dry density 1.2 Mg/m3). Thus, it is considered that the buffer material
can maintain the density necessary for assuring performance over a long period of time. Fromthe viewpoint of supporting the load of the filled emplacement cell, the stress ratio severity ofthe buffer was noted to approach 1 – with a risk of failure – only in restricted areas, hence lossof performance of the entire buffer thickness is not expected.
An example of the results of the assessment of the impacts of earthquake ground motion isshown in Figure 4.4-24, again showing that the stress ratio severity of the buffer limit of 1 isreached in a few areas but does not penetrate the entire buffer thickness.

Figure 4.4-24 Maximum stress ratio severity in buffer as a result of earthquake motion

4-61

Taken together these results indicate that the TRU buffer can be assured to perform overassessment periods of > ≈ 100 ky.
(5) Specification of the EBS
Based on the above, TRU waste EBS specifications are derived for all rocks and cases withand without buffer. Examples are shown in Figure 4.4-25 (based on specification of thedisposal vault developed in Section 4.5.2 (2) (i) (c)).

(a) With buffer (Gr.2)b) Without buffer (Gr.3)

Figure 4.4-25 EBS specification for TRU waste (for plutonic and Pre-Neogene sediments)

The buffer thickness required is 1.0 m, but at the bottom is set as 1.2 m to allow a sufficientmargin for consolidation settlement. In addition, since the side gap between the vault liner andthe concrete cell wall is filled with buffer, its thickness is > ≈ 1.0 m, depending on the shapeof the vault.
4.4.3 Construction of the EBS and waste emplacement
Currently, construction and waste emplacement technology is still under development. The
specifications given in this section are thus only illustrative examples, other options are being
assessed and decisions on preferred technology would be made only when designs of the EBS
become more advanced. Nevertheless, such basic concepts allow preliminary assessment offundamental engineering practicality.
(1) HLW(i) H12V
As described in Section 4.2.4 (1) (i), the overpack containing vitrified HLW is loaded on a
dedicated transport vehicle at the surface facility and transported underground via an accessramp [1]. In case of manual operation, the transfer vehicle is provided with radiation shielding.
Alternatively, if an access shaft is utilised, waste will be transported by a dedicated elevatorcapable of handling appropriate weights. The reference concept for transport of waste to therepository is by ramp, using either a road-based transport vehicle such as a truck or a railbased transport vehicle. To allow the ramp access route to be used for other purposes andfrom the viewpoint of transport flexibility, a road-based transport vehicle is selected as thereference [1] (although alternatives may be considered in the future).
4-62

After removal from the ramp transporter at a temporary storage location (see Section 4.5.4(4) (iii)), the overpack is transported through the connecting tunnels to the disposal tunnel, in
which it is emplaced in a hole already containing buffer (as shown in Figure 4.2-2 (a)). For a
prepared emplacement hole, lower buffer cylindrical blocks and surrounding rings arelowered into place (using a vacuum gripping device) before the overpack is inserted andupper buffer blocks added to fill the hole (Figure 4.4-26).

Figure 4.4-26 Buffer and waste emplacement procedure (H12V). These procedures weredeveloped assuming low water inflow rate into the holes

As noted in Section 4.4.1 (4), the surface dose rate of the overpack is ≈ 11 mSv/h, so
emplacement would involve remote-handling technology. Further studies will be required forthe installation sequence of overpacks (to ensure hole stability) and will probably also require
to be adapted to site conditions and the final selection of buffer composition.
Emplacement equipment for the overpack and buffer using rail transport has been
demonstrated at full-scale [87] [88]. A vacuum lifting technique is being developed forbentonite blocks that is effective in the limited space available and allows loading, unloading,transfer and emplacement of buffer blocks without damage. This has been shown to fail-safe(designed so that any credible accident does not lead to a significant perturbation) even in theevent of failure of the vacuum pump, based on full scale tests [87] [88] [89]. To allow
accurate block emplacement in disposal holes, the vacuum grip is incorporated into atelescopic arm for lowering into place. Supporting Report 4-29 overviews demonstration testsof buffer block fabrication and emplacement technology.
Since the buffer material starts swelling immediately when it comes in contact with highhumidity air or water, the inner surface of the buffer material may deform, preventing
emplacement of the overpack. Thus, buffer must be kept dry until the overpack emplacement
is completed, as described in the following.
To avoid buffer wetting, a method involving a protective rubber sheet has been studied [90].
This rubber sheet is assumed to be removed after successful emplacement of bentonite buffer
blocks, but this has not yet been examined to check practicality. Thereafter, inflowing water
into the hole contacts the bentonite buffer blocks, which start to swell. After saturation, thedesired properties of low permeability and colloidal filtration will be obtained. However, it
4-63

should be noted here that more studies will be needed to confirm whether such a method
would be practical or if other means of buffer protection would be needed.
If the water inflow rate is sufficiently large, there is a risk of piping-erosion (see 4.5.4 (5)),
which may lead to deterioration of performance. Either such holes would be rejected or atechnique used to reduce water inflow (e.g. by grouting or use of a hole liner). If water inflow
control measures were to be used, the post-closure safety implications would need to beassessed while, if some disposal holes are not useable, their likely number needs to be
estimated in order to allow for this, as described in Section 4.5.4 (5).
The above procedures require remote handling to reduce radiation exposure to workers.
Techniques related to accurate positioning and autonomous control thus become important,but, in recent years, such an approach is well supported by advances in relevant technology[91].
(ii) PEM
The PEM is assembled on the surface and transported underground, as for the H12V
overpack. The PEM is 3.4 m in length and 2.3 m in diameter, and has a weight of ≈ 37 Mg(specifications given in Figure 4.4-18 and more details in Supporting Report 4-19). The
surface radiation is reduced by self-shielding to ≈ 1μSv/h. Within the underground facilities(see Section 4.5.4 (4) (iii)), the PEM is transported through the connecting tunnels to thedisposal tunnel, in which it is emplaced by remote-handling technology (as shown in Figure
4.2-2 (b)).
The PEM is large and heavy, so special equipment is required for its transport andemplacement, and also for filling the gap between the PEM and the disposal tunnel; this iscurrently under development. One approach to PEM emplacement within a small tunnel
utilises an air caster system, which has been demonstrated in full-scale tests [92]. In particular,
SKB has demonstrated a form of PEM transport and emplacement (KBS-3H) at full scale inthe Äspö underground laboratory [14].
Although the air caster system reduces the gap between the PEM and the tunnel wall
allowing reduction of the diameter of the tunnel, the contact surface is required to be smoothand needs a special device for filling remaining gaps, requiring complex operations after each
PEM is emplaced. To avoid such problems, a larger diameter tunnel is considered in thisreport, together with PEM transportation/emplacement by a mobile lifting unit [1]. Moredetails on PEM fabrication, transport and emplacement are given in Supporting Report 4-30.
The effect of high humidity or groundwater in the disposal tunnel is much less for PEM
emplacement compared to H12V. However, the backfilling of the gap between the PEM andthe tunnel wall must be practical and have required quality, so water inflow control (grouting)
may be required and backfilling technology needs to be demonstrated [93].
These procedures require remote-handling, because the gap between the PEM and thedisposal tunnel wall is insufficient to provide a safe working environment for workers, even if
the radiation exposure is sufficiently low.
(2) TRU waste
TRU waste packages may be transported underground from surface facilities by an accessramp or shaft. When co-locating with HLW, dedicated TRU access and transport vehicles
4-64would be planned [1]. In case of manual operation, the transfer vehicle is provided withradiation shielding. As for HLW, road transportation vehicles such as trucks are assumed fortransportation in ramps and tunnels to the disposal vaults [1].
As an alternative method for transportation underground, an access shaft can be considered,
as in existing repositories for similar waste, such as the Waste Isolate Pilot Plant (WIPP), a
TRU waste repository in the United States [94], and the licensed Konrad repository inGermany [95]. Transportation by access ramp will, however, be considered as the referencecase in this report.
Waste packages are transported directly to the disposal vault (contrasting with HLW, which
involves transhipment at the end of the access ramp) [11]. As noted in Section 4.2.4 (1) (i) (f),emplacement is assumed to be by forklift for waste package A or by gantry crane for waste
package B (schematically illustrated in Figure 4.4-27). Due to the high radiation dose, thesetasks must be performed remotely. For this reason, techniques related to accurate positioning
and autonomous control become important. In recent years, such an approach is well
supported by advances in relevant technology [91].

Figure 4.4-27 TRU waste emplacement

Emplacement using a forklift has been considered in detail previously [10] [11], but thegantry crane system noted as a design option in the TRU-2 report is examined in more detailhere. In addition, if a forklift is used, partition walls need to be constructed when waste ispresent, and thus constrained by radiation exposure guidelines. The crane used for waste
package B avoids this problem, as construction of partition walls occurs before emplacementof waste packages.
The work flow for TRU waste emplacement by crane is shown in Figure 4.4-28. After
excavation, the bottom buffer is installed, with the side buffer installed after the constructionof the structural framework (2).

4-65

Figure 4.4-28 TRU waste disposal work flow

Figure 4.4-29 shows waste package transfer to the disposal tunnel, unloading by crane fromthe transport vehicle and moving to a predetermined position in the disposal vault (3).

Figure 4.4-29 Transport of waste packages to the emplacement vault.
Handling height, H, set as limit for waste packages (see Section 4.5.2 (2) (i) (c)).

After filling an emplacement cell with waste packages, it is infilled (4). After the entire
vault is filled, the upper buffer (5) is installed and remaining roof space backfilled (6) (seeSection 4.5.3 (1)). Full-scale demonstration tests of several of these operational steps havebeen carried out [20] [48] [85] and associated construction technology is summarised in
Supporting Report 4-31. From technological development to date, it is concluded that EBS
construction and associated quality control techniques to achieve the target performance isconsidered practical.
4-66

4.5 Design of underground facilities
In this section, the design of the underground facilities is described, based on the SDMs
developed in Chapter 3. In addition, the required engineering technology and safety measures
for underground construction, operation, and closure are assessed, for both normal operations
and potential perturbations (abnormal states), with a focus on countermeasures to provide
defence in depth.
4.5.1 Concept of repository design(1) Design procedure of underground facilities
The underground facilities of a repository for co-disposal of HLW and TRU waste include
the components summarised in Section 4.2.3 (3). As specified in Section 2.1.1 (4), the Final
Disposal Plan envisages emplacement of 40,000 canisters of HLW at a rate of approximately
1,000 per year. Thus, the HLW disposal requirements are similar to those in H12, withstepwise construction and operation assumed to run in parallel within specific panels ofdisposal tunnels. TRU waste packages are emplaced in a more compact manner in larger
vaults that have a much smaller total length than the HLW tunnels. For this reason, it isassumed that the TRU waste emplacement is started only after construction of all disposal
vaults and it is currently assumed that each waste group will be placed in a dedicated vault.
Nevertheless, since the actual TRU inventory is uncertain, allowance is made for later
expansion of the vault network. In addition to the design of disposal tunnels and vaults, allother tunnels and galleries required for construction and operation, plus the backfilling andplugs used to close them, must be considered.
For HLW, interim storage for 30 to 50 years prior to disposal has been considered. The
footprint of the repository is larger in the case of 30 years storage, in order to minimise therisk of thermal alteration of bentonite buffer. A larger repository may lead to a larger impact
to the deep geological environment during the construction and operation of repository. Thus,
to minimise such impact, a reference case assuming HLW stored for 50 years is selected (seealso Section 4.2.2 (1)).
The process of overall design of the repository is illustrated in Figure 4.2-1, which is
expanded in more detail for underground facilities in Figure 4.5-1. The underground facilities
will be designed in the following steps: (1) design of tunnels, (2) design of backfill materials
and plugs, (3) design of underground facility layout, and (4) design of infrastructure such as
drainage and ventilation systems.
The tunnel design defines the cross-sectional shape and dimensions of the different accessways, tunnels and vaults required, based on the installation depth set in Section 4.3 and anevaluation of cavity stability. In the design of backfill materials and plugs, the specificationsof materials, structures and installation positions are set based on consideration ofgroundwater flow characteristics of the excavation damaged zone (EDZ) around tunnels (andboreholes).

4-67

SDM at repository scale
Mechanical/thermal properties
Mechanics/hydrogeology

Design of tunnel cross section and support
Design of backfill and plugs

Geology/hydrogeology/
mechanical properties

Design of underground facility layout
Selection of suitable areas for repository

Shape and size of disposal panels
Layout of disposal panels

Connecting and access tunnel layout
SDM at disposal panel scale
Hydrogeology

Estimation of waste emplacement efficiency

SDM at repository scale
Geology/hydrogeology

Design of connecting/access tunnels

Hydrogeology/thermal properties

Design of entire repository

Figure 4.5-1 Repository design flow

For layout design, the tunnel, backfill and plug specifications are extended to include thesize and shape of the disposal panels and their spatial arrangement, together with requiredconnecting tunnels, access ramps, shafts, etc. As described in Section 4.2.2 (1), the layout
assumes co-location of disposal facilities for HLW and TRU waste, benefiting from sharingsite characterisation efforts and surface facilities. In this case, it is necessary to minimise any
interactions between these disposal areas [10] [11]. Finally, main services including
ventilation, cooling, drainage, power, etc., are considered for the set repository layout,
although only at a simple level appropriate to the current design stage.
(2) SDMs for the design of underground facilities
For the design of underground facilities as shown in Figure 4.5-1, key input is provided by
the SDMs shown in Section 3.3.3. The SDMs include geological structures, hydrogeology,
thermal and mechanical characteristics of bedrock and groundwater chemistry, specificallyintended to provide the necessary input for designing the repository.
For tunnel design, the geological characteristics summarised in Table 4.2-6 are used for theevaluation of cavity stability. As shown in Sections 3.3.3 (4) (i) (b) and (c), the SDMs for
Neogene and Pre-Neogene sediments are characterised by fold structures, which can beassumed to be associated with stress anisotropy, which has to be considered when designing
tunnels with respect to their orientation. Whilst not strictly true, plutonic rocks are treated as
4-68having no stress anisotropy in this assessment, as this would be expected to be less of an issue
here. Backfill plug designs are based on hydrogeological data presented in Table 4.2-7 andgroundwater chemistry from Table 4.2-8, in addition to the rock mechanical data from Table
4.2-6.
The repository layout is defined for each of the 3 SDMs from Section 3.3.3 (4), withparticular focus on the structural and hydrogeological model components. Section 3.3.3 (3)discusses the role of large faults and fractures as major groundwater migration pathways,
classifying the lengths of faults as an index of significance in the repository scale and panel
scale SDMs. Generally, the frequency of faults is inversely proportional to their length.
Classification of faults and their handling in design are summarised in Table 4.5-1.

4-69

Table 4.5-1 Classification of faults and fractures
Classificationof faults andcracksRepresentation in SDM(see Section 3.3.3 (3) (i) (a))

Handling in design of undergroundfacilities> 10 km

Assumed to be an active fault.
Treated as a fault excluded from therepository area during LS or PI.
Described in regional scale SDM, but
excluded from the model area at therepository scale.

From active fault definition, does not apply
to the examination of repository layout.

1 - 10 km< 1 kmDisposal tunnels are not placed in areas
Position and characteristics can bewhere such faults are found, as there may beroughly determined during LS or PI.
problems with excavation due to mechanicalFault distribution is described in regionalinstability, high water inflow, etc.
scale and repository-scale SDMs.
This can be classified as a layout determining
feature (LDF).
Considered in planning the disposal area.
The hydrogeological and mechanicalcharacteristics of such faults may varyBecause of wide distribution, not realistic
considerably. Since these may require
to detect all during LS or PI, but can besubstantial reinforcement or result in ingressdescribed in terms of a statisticalof water, tunnels/vaults around intersections
distribution.
of these may be unsuitable for waste
Faults described statistically in panel
emplacement, which needs to be consideredscale SDM.
when planning the repository layout.
This can be classified as an emplacement
determining feature (EDF).

Faults of 10 km or more in length are described in the regional scale SDM but, in this report,
they are assumed to be excluded from siting area considered on the basis of the initial LS or
the PI, and are thus not present in the repository scale SDM (see Section 3.3.3 (3) (i) (a)) andnot treated in design.
Faults with a length of 1 to 10 km are described in both the regional scale and repository
scale SDMs and it can be assumed that their locations will be determined during LS and PI
characterisation (Section 3.3.3 (3) (i) (a)). Concrete examples are shown in Section 4.5.4 (1),but, from the viewpoint of facilitating construction and operation, the layout should avoid
such faults to the extent possible, setting disposal panels in areas where they are not present.
These thus act as layout determining features (LDF).
Although faults and fractures less than 1 km in length can be located by observation oftunnel walls, they have a high density and it is not possible to detect and characterise them all(Section 3.3.3 (3) (i) (a)). In the repository-scale SDM, the effects of such small faults andfractures are included in the properties of the rock matrix, as these features should beconsidered in decisions on emplacement of waste and EBS design (so called emplacement
determining features: EDF). In the smaller, panel-scale SDM, these are represented by astatistically-generated discrete fracture network (DFN) model. This was used to assess to what
extent the variability of the inflows resulting from these fractures would affect the number ofsuitable disposal locations (see Section 4.5.4 (5) (iv) (b) for further details). The
hydrogeological models are used to characterise local variations in groundwater flux, flow
velocity and transport path to the accessible environments. In general, low fluxes/flow
velocities and long path lengths are favourable.
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4.5.2 Tunnel design
In the design of tunnels, key issues are the cross-sectional shape, the support system usedand pitch between tunnels. Here tunnel is used as a general term to cover all larger excavatedopenings – including shafts, ramps, galleries and vaults.
(1) Tunnel design requirements and evaluation items
Tunnel requirements (safety functions) for operational safety include design factors formaintaining a suitable working environment, prevention of occurrence/expansion of accidents,
avoiding risks to occupational safety and health, and ensuring alternative escape routes (seeTable 4.2-3). To assure the requirement of engineering practicality, the design should show
methods of construction, operation and closure based on existing (or soon to be realised)technology. These functions depend on the role of different types of tunnel, as summarised inTable 4.5-2 for the reference case.
Table 4.5-2 Roles of repository tunnels
Function

Accessramp

Accessshaft

Connecting/surrounding/approach tunnel

Disposaltunnel/vault

Transport of waste materials
and operational equipment

○

-

○

○

Pathway for workers

-

○

○

-

-

○

○

-

-

○

○

-

Ventilation route

○

○

○

-

Drainage/water supply route

-

○

○

-

Power supply andcommunication path

-

○

○

-

Emergency evacuation route

○

○

○

-

Pathway for construction
machineryTransportation path for
excavated rock

Considering the role of each of these tunnels, design requirements are defined as shown in
Table 4.5-3. The available space requirement depends on the role of each tunnel and isensured by setting an appropriate inner cross-section. For parallel construction and operation,access and connecting tunnels should be designed and laid out to allow movement ofmachinery, materials and workers without risk of inadvertent or deliberate bypassing of strictzoning between active and non-active areas.

4-71

Table 4.5-3 Tunnel design requirements
Designrequirements

Objectives

Specifications

Space

Assuring the space required for transport in the tunnel orthe installation of engineered barriers, plus equipment
necessary for construction and operation, and utilities
such as ventilation and drainage

Tunnel shape, innerdiameter, inner structures(e.g. drains, cable ways)

Securing safe
passage

A safe passage should be assured for sections where
workers are present

Cavity stability

The mechanical stability of the tunnel should be assured
so that construction and operation can be performed
safely and efficiently (also considers the thermal stressinduced by the waste)

Reduction ofthermalimpacts

The performance of the engineered barrier is not
significantly reduced due to the heat effect from thewaste

Tunnel shape, innerdiameter,layoutTunnel shape, innerdiametertunnel lining/support,
waste disposal pitch, pitch
between disposal tunnels
Pitch between disposal
tunnels, waste disposal
pitch

Prevention ofrock spalling

Prevention of rock fall

Tunnel support/lining

To meet these requirements, tunnel designs should ensure mechanical stability during
construction, operation and closure and also avoid any other potential perturbations, e.g. due
to high water inflow. For the specific case of heat-generating waste, reduction of thermal
impacts is set as a design requirement.
Table 4.5-4 identifies seismic stability as an evaluation item to assure mechanical stabilityof tunnels in the case of major earthquakes.
Table 4.5-4 Tunnel evaluation item
Evaluation item
Seismic stability

Objective
The mechanical stability of the tunnel is assured in the case of earthquake motions

In the design of the cross-section of the tunnels, firstly the required inner space will be set,
based on the equipment and services to be installed inside the tunnels, clearances for
operating machines or material transporters, and the required safe pathways for workers. Then
tunnel stability will be analysed, to determine the necessity of tunnel support and constraints
on excavation procedures [96] [97]. Here, in the case of the disposal tunnels, a number ofparallel tunnels are included in a disposal panel and their pitch must be set to assure bothmechanical stability and absence of significant thermal impacts. Specifically for HLW, thepitch of waste package emplacement has also to be considered.
The design of these tunnels is carried out for all three representative host rocks but, as
mentioned in Section 4.2.5, mechanical and thermal properties of plutonic rocks and PreNeogene sediments are effectively the same (see Supporting Report 4-3) and hence these canbe considered together. In terms of an excavation method that is adaptable in terms of crosssection, profile and geology, the cost-effective NATM (New Austrian Tunnelling Method) is
assumed. A TBM (tunnel boring machine) may be also applicable if the target rocks are
suitable for this technique. The tunnelling method will be chosen when tailoring to site
properties.

4-72

(i) Assessing the tunnel cross section
Standard tunnel profiles considered are horseshoe shaped or circular [98], based on
excavation practicality and mechanical stability constraints. For many tunnels a flat floor isrequired. However, depending on the combination of the size of the tunnel and the strength of
the rock, it may be difficult to assure stability at the intersection of the floor surface and thesidewall for a U-shape option. In such a case, stability considerations lead to a tailoredhorseshoe or circular profile, selected according to the specific conditions. When adopting
these cross-sectional shapes, a flat concrete floor is emplaced. Shafts with circular profile are
the norm for any deep excavation. Based on the such concerns, an appropriate tunnel crosssectional shape is selected.
Supporting Report 4-32 describes in detail the process of designing tunnel liners [96] [97].
Such design considers also the impacts of pitch on stability for the specific case of disposal
tunnels. The lining concrete specification is derived with explicit consideration of the lengthof time the tunnel is open, its dimensions, etc.
(ii) Assessing the disposal tunnel pitch
As noted above, a number of parallel tunnels are included in a disposal panel and their pitch
must be set to assure both mechanical stability and absence of significant thermal impacts
Therefore, tunnel stability analysis assessed the impact of pitch for different shapes and sizesof disposal tunnel cross section, together with the specifications of liners, as described in
Supporting Report 4-33.
In addition, for H12V, disposal holes are drilled at regular intervals and the pitch of thesehas to be selected so that they do not affect each other.
For higher thermal output TRU Grs.2 and 4H, in addition to confirmation of mechanicalstability, thermal impacts require consideration when setting vault pitch. For such waste, andalso HLW, thermal conduction analysis allowed the assessment of the impact of pitch on
maximum temperature, as described in Supporting Report 4-14 (100 °C limit assumed). Fromthese results, the larger of the values of the pitch required to meet stability and thermal
constraints is selected.
(2) Setting tunnel cross sections
Disposal tunnels and vaults comprise most of the repository volume and these, together
with other supporting tunnels, are now specified in terms of cross-sectional shape anddimensions, together with associated mechanical support structures/lining.
(i) Disposal tunnels(a) H12V
H12V involves placing overpacks containing HLW and surrounding buffer in regularly
spaced boreholes. Figure 4.4-18 (a) presented the specification of disposal holes with an innerdiameter of 2.22 m and a depth of 4.15 m. The factors that set the construction limits that
determine the cross section of the disposal tunnel are the size of disposal hole drilling rig andthe overpack/buffer emplacement machine, resulting in the profiles shown in Figure 4.5-2.
4-73

(a) Plutonic, Pre-Neogene sediments(b) Neogene sediments

Figure 4.5-2 Disposal tunnel size, cross section and support (H12V)

Mechanical support for Plutonic rocks and Pre-Neogene sediments is based on standard
guidelines [98], involving 50 mm thick shotcrete and rock bolts with a length of 2 m andseparation of 1.2 m in the circumferential direction and 1.5 m along the tunnel. In the case of
Neogene sediment, the same guidelines [98] and stability analysis lead to 200 mm shotcreteand high-standard steel support (HT 590: high strength H-shaped steel with 590 MPa tensile
strength) at 1 m intervals along the tunnel. Rock bolts 3 m long have a separation of 1 m inboth the circumferential direction and along the tunnel. Stability evaluation of HLW disposal
tunnels is described in detail in Supporting Report 4-34.
Disposal holes need to be mechanically stable until the placement of the buffer is complete.
Based on the mechanical analysis, such stability is questionable, particularly around the upper
part of the hole. It is possible that the design of the hole could be modified to address suchconcerns (Figure 4.5-3), including a shotcrete or steel liner, or an upper concrete reinforcing
ring.

(a) Shotcrete(b) Steel liner(c) Reinforcing ring
Figure 4.5-3 Hole support to ensure stability(Source: NUMO, 2011 [1])This would, however, both complicate hole excavation and require further materials to beconsidered in the post-closure safety assessment. For this report, it is assumed the hole would
be stable without such support.

4-74

(b) PEM
Assessment of the required disposal tunnel cross-section is based on the PEM specification(Figure 4.4-18) and the emplacement machine described in Supporting Report 4-30, withconsideration of clearances and space required for support services, resulting in the profiles
shown in Figure 4.5-4.

Plutonic rocks/pre-Neogene sediments

Neogene sediments(a) Plutonic, Pre-Neogene sediments(b) Neogene sediments

Figure 4.5-4 Disposal tunnel size, cross section and support (PEM)

Mechanical support specification and associated stability analysis are the same as for
H12V.
(c) TRU waste
TRU waste disposal vaults contain, in addition to waste packages and other engineered
barriers, structural framework and backfill (see Figure 4.5-5 for waste package B disposal
concept).

Figure 4.5-5 TRU waste disposal vault

4-75For this reason, the size of the vault is set from the considerations of the density of waste
emplacement, thermal constraints, mechanical stability and the design of the structures to beinstalled for waste handling and emplacement. The design of the TRU waste disposal vault
thus involves (1) selection of vault shape, (2) design of required structures (3), establishingconstruction limits, and (4) determining tunnel cross section shape and support withconsideration of the excavation procedure involved. A basic horseshoe shape cross section is
selected for plutonic rocks and Pre-Neogene sediments, for further analysis in Section 4.5.2(1) (i). For weaker Neogene sediments, cavity stability considerations lead to a choice of acircular shape.
As shown in Figure 4.4-28 and discussed further in Section 4.5.5, vault contents include
waste packages, infill between waste package and the structural framework, a buffer (in somecases), and a gantry crane required for emplacement. The following assesses the number ofwaste packages that can be emplaced for different designs of the structural frame and theconstraints set by the emplacement crane.
As the density of waste packages is not limited by heat generation for Grs.1, 3 and 4L,
waste package emplacement density focuses on meeting required storage capacity forminimum excavation volume. As shown in Table 4.5-5, different emplacement options interms of number and height of waste package stacks require different total excavated volumes,
allowing the most efficient to be selected.
Table 4.5-5 Low heat waste inventory and excavation requirements

Waste
Group
1*1
(Drum)
3(Drum)
4L(Drum)
4L*2 (Boxcontainer)

StacksHeightPackagespersection

4
4
5

4
5
5

16
20
25

(1)
Cross
sectional
area
(m2)
129
150
144

6
5
6
3
4
5

5
5
5
4
4
4

30
25
30
12
16
20

161
147
168
99
113
134

68.2
57.5
539.1

(3)
Volume ofadditional
excavation(m3) *3
5,364
7,122
-

459.0
460.2
390.7
65.0
54.5
48.5

3,168
4,401
6,047

(2)
Tunnellength(m)

Totalvolume(m3)
1×2+3Judgement

14,162
15,747
77,630✓73,89967,64965,638
9,603
10,556
12,546✓✓✓*1

Gr.1: includes buffer and its volume is also considered.
Gr.4L: since the dimensions of the waste package are different for drums and box containers, separate disposal
tunnels are considered.
*3
Gr.1, 4L: box container vaults are dead-end because the quantity of waste is small and the tunnel length isshort. Therefore, a widened section was provided between the connecting tunnel and the disposal vault toallow for this.
*2

Vault profiles for higher heat output Grs.2 and 4H are also assessed in the light of efficiency
of use of excavated volumes. However, as noted in Section 4.2.4 (2) (ii), infill between waste
packages could be thermally degraded (even if this is rather unlikely on the basis of current
knowledge) and a buffer outside of the structural framework is included to assure safetyfunctions. In addition, in order to prevent thermal degradation of the buffer, the waste package
density was set so that the temperature of the buffer material did not exceed 100 °C (again

4-76

very conservative based on current understanding). Further, the spacing between waste
packages was set to 150 mm in consideration of ease of mortar infilling.
The structural framework forming the disposal cells (manufactured from reinforced
concrete) is designed based on the selected waste package emplacement scheme and wall
thickness requirements to support the gantry crane (Section 4.4.3 (2)), providing structural
integrity against loads such as self-weight, fluid pressure at the time of infilling and swellingpressure of buffer. Details are provided in Supporting Report 4-35. From the designs, theheight of the side wall of the structural framework ranged from 4.05 to 7.95 m. Based on this,as described later in Section 4.5.6 (1), the maximum waste package handling height whenlifting by crane was set to 8 m.
Figure 4.5-6 illustrates examples of tunnel profile, dimensions, structural framework, bufferand structural support for Gr.2 in the hard and soft rocks.

(a) Plutonic rock, Pre-Neogene sediments(b) Neogene sediments

Figure 4.5-6 Examples of TRU waste disposal vault cross sections (Gr.2)

Specifications for the other waste groups are provided in Supporting Report 4-36. Cavity
stability evaluation results (see Supporting Report 4-37) show that the excavation method andprovision of mechanical support is important, particularly for Neogene sediments, when
excavation in a number of steps is required to assure release of drilling stresses without risk ofwall failure.
(ii) Other tunnels
Table 4.5-6 summarised the basic design assumptions for other types of tunnel, with an
example of resulting specifications for H12V/TRU in plutonic rocks (more details in
Supporting Report 4-38). A key constraint on dimensions is set by logistical aspects,
determining whether tunnels need to handle one- or two-way traffic. The size of transported
waste packages (especially H12 overpack or PEM) needs to be considered, along withadditional clearance for services. Such clearance would, for example, be needed for electricalsupply, ventilation ducts and safe passage for evacuation in case vehicles block the tunnel.
However, different technical solutions to meet requirements would impact the resultant
dimensions of the tunnel. The current dimensions are thus indicative of what is needed, but
may be optimised in the future by considering all necessary functions of the tunnel, as well as
the cost of excavation and eventual backfilling. Such studies would be needed for detailed,site-specific design of the repository.

4-77

Table 4.5-6 Assumptions and specifications for other tunnels (H12V and TRU waste inplutonic rock)Tunnel type

Design assumptions

Surrounding
tunnel(HLW)

Based on H12V with 5 disposal
tunnels operating in parallel toprovide the required 5
overpacks/day throughput
Daily transport rate is low and hence
1 lane is sufficient.

Connecting
tunnel(HLW)

Requires 2 lanes so that the
overpack and buffer emplacement
device can pass each otherApproachtunnel(TRU)

Because there is limited material
associated with loading waste
packages into individual vaults, onelane is sufficient
Because there is a large material
flow to all vaults, 2 lanes are
required to allow for waste delivery
and return of transport vehicles.

Surrounding
tunnel(TRU)
Access shaft(HLW ·
TRU)

Critical for material transport inlarge quantities at the time of
construction (e.g. removal of
excavation spoil and supply of
concrete). It is also important for
utility supply, such as ventilationand drainage. In addition, space isalso required to secure a work
platform for drilling the shaft itself.

Access ramp
(HLW ·TRU)

In order to transport over long
distances at low speed (safety
considerations), two lanes are
required to allow for waste delivery
and return of transport vehicles.

Specification example (plutonic rocks)

Same cross section as the surrounding tunnel(HLW)

Same cross section as the connecting tunnel (HLW)

Same cross section as connecting tunnel

4-78Since the specification of the connecting tunnel cross section here is the same as the HLW
disposal tunnel, the specifications of the mechanical support structure are also the same. The
access ramp and the connecting tunnel have a 3 m larger width than the surrounding tunnel,
and the standard support can be applied to ensure cavity stability. These tunnels operate for a
relatively long period, from the initial stage of construction to closure, which is explicitly
considered when specifying concrete lining requirements. In addition, in order to reduce
impacts of water inflow, a waterproof sheet is included between the initial shotcrete support
and the final concrete lining (details in Supporting Report 4-38).
(3) Setting the pitch of disposal tunnels and holes(i) Assessment of cavity stability
To assess mechanical stability, based on the H12 approach, the width of the plastic area ofrock surrounding a tunnel is obtained by two-dimensional or three-dimensional finite element
analysis, with consideration of the process in which mechanical support is emplaced during
construction (see details in Supporting Report 4-34). For H12V, meeting the stability criteria
requires a tunnel pitch of at least 2D (where D is the tunnel diameter) for plutonic rock and2.4D for Neogene sediments (i.e. 10 m and 12 m, respectively for D = 5 m). Disposal pitch is
set on the basis of approach to shear breakdown when the local safety factor is 1.2 or less.
Applying this criterion, the minimum hole pitch is 2d (d: disposal hole diameter) for plutonic
rock and 3d for Neogene sediments (i.e. 4.44 m and 6.66m, respectively, for d = 2.22 m).
For the PEM system, the same results apply in terms of tunnel pitch, but differences in the
diameter of the disposal tunnels need to be considered (see Section 4.5.4 (2) (ii) (b)). Thus, 2D pitch for plutonic rocks corresponds to 9.6 m while 2.4D for Neogene sediments
corresponds to 12.0 m.
Evaluation of the cavity stability of TRU waste disposal vaults was performed by elastoplastic analysis considering the stepwise excavation appropriate to such a large cross-sectional
area (details can be found in Supporting Report 4-37). For plutonic rocks, tunnel innerdiameter (D) ranges from 9.64 m ~ 12.69 m depending on the waste group, and a disposal
vault pitch of 2.5D was initially set to meet cavern stability criteria. For Neogene sediments, it
is difficult ensure stability compared to the case of the plutonic rocks. Thus, two specificcases, D = 13 m or 16 m, were analysed in detail for a vault pitch of 3D. In both cases
extensive support was required, despite drilling stress relief during stepwise excavation andsupport emplacement (2 steps for D = 13 m, 5 steps for D = 16 m). In both cases, 30 cm
thickness of shotcrete was required together with high standard steel support (H-200 and H250, for the small and large tunnel respectively).
(ii) Assessment of thermal effects(a) H12V
Thermal effects were assessed for plutonic rocks (depth 1,000 m) and Neogene sediments(depth 500 m), extending the stability studies from Section 4.5.2 (3) (i), with details
summarised in Supporting Report 4-39. For plutonic rocks, the minimum pitch to ensurestability of tunnels (2D, 10 m) and emplacement holes (2d, 4.44 m) were subject to heat
conduction analysis. It was confirmed that the buffer temperature limit of 100 °C was notexceeded for HLW that had been in interim storage for 50 years and hence these pitch values
met all criteria (see Section 4.5.1 (1) for further explanation).
4-79

For Neogene sediments, the minimum pitch to ensure stability of tunnels (2.4D, 12 m) andboreholes (3d, 6.66 m) were also subject to heat conduction analysis. Again, it was confirmedthat the buffer temperature limit of 100 °C was not exceeded and hence these pitch values met
all criteria. In addition, in this case the criteria were met for both 30 and 50 years of interim
storage, as shown in Table 4.5-7.
(b) PEM
It is assumed that PEMs are emplaced in contact with each other (effective waste pitch 3.36m) and hence the only variable considered for thermal analysis is the disposal tunnel pitch(details in Supporting Report 4-40). Figure 4.5-7 shows the relationship between disposal
tunnel pitch and the maximum buffer temperature for HLW stored for a 50-year period inplutonic rock. Minimum pitch to ensure the cavity stability (9.6 m) results in buffertemperatures slightly above 100 °C (Figure 4.5-7), while this limit is not exceeded for a pitchof 11 m.
104.5

104

Limiting temperature< 100 °C

Max temperature (°C)

102

100

99.7

98
96.9
96
94
92
90
9

10

11

12

13

Disposal tunnel pitch (m)Figure 4.5-7 Relationship between tunnel pitch and maximum buffer temperature(PEM, Plutonic rock)

It should be emphasised that this is an extremely conservative treatment, as the excess
temperature is minor and lasts only a few decades, during which the steel handling shellwould probably be intact – making the contained bentonite even less vulnerable to higher
temperatures. For Neogene sediments, the minimum pitch of 12 m to ensure stability oftunnels results in a maximum buffer temperature < 100 °C and hence this pitch met all criteria.
Again, the criteria were met for both 30 and 50 years of interim storage, as shown in Table
4.5-7.
(c) TRU waste
Thermal effects resulting from TRU waste disposal were assessed as for the HLW case
Based on the configuration of disposal vaults presented later in Section 4.5.4 (2) (iii) (b),4-80vaults for higher heat generating TRU waste are interspersed with those with a lower heatloading to reduce thermal effects on both the EBS and contained waste. The pitch of vaults
was initially determined on the basis of mechanical stability, and then assessed to check if
temperature limits set for buffer or mortar infill between waste packages are exceeded. The
thermal analysis is carried out in two steps, first simple heat-conduction calculations to assess
the sensitivity of temperature to vault pitch and then, when the pitch from mechanicalconstraints appeared suitable, more detailed analyses were performed to determine detailed
temperature profiles within the vaults for this pitch. The following presents examples ofresults obtained from the latter analyses (further details in Supporting Report 4-41).
Figure 4.5-8 shows the evolution of the thermal transient, evaluated at the location in thebuffer where the highest temperature was observed, for vaults containing the higher heat TRU
wastes in plutonic rocks (as the extreme case). The highest temperature was approximately
90 °C, for the highest thermal output TRU waste (Gr.4H, MHHRW). In the case of Neogene
sediments (not shown), the highest temperature was 84 °C. Thus, the buffer temperature limitof 100 °C was not exceeded in any case and the highest temperatures are apply to relatively
short transients with highest temperatures lasting only a few decades.
120
Gr.2 Canisters

110

Gr.4H Drums
Gr.4H MHHRW

Temperature (℃)

100
90
80
70
60
50
40
0.01

0.1

1

10
100
Time (y)1000 10000

Figure 4.5-8 Time-dependent temperature change in the bentonite buffer for the heatgenerating TRU wastes (plutonic rocks)

According to the TRU-2 report, mortar can be denatured at temperatures above 80 ℃,
although this may be rather conservative on the basis of more recent studies. Thus, in orderfor the vault infill to retain its sorption safety function, a requirement is set that thetemperature should be less than 80 °C. Gr.2 and 4H vault infill lies inside a buffer and, basedon the above analysis results, the mortar temperature limit is likely to be exceeded. Thus, theconservative treatment assumes loss of the sorption safety function described in Section 4.2.4(2) (ii). For other groups (1 and 4L), even taking into account the influence of heat from theGr.2 and 4H disposal tunnels, temperatures fall below 80 ℃. As indicated in Section 4.5.4 (2)(iii) (b), Gr.3 disposal tunnels are isolated from the other waste groups and hence not affectedby heat. Therefore, the sorption safety function is applied to vault infill only for Grs.1, 3 and
4L, even though the temperature excess is small with a short duration for the other waste
groups and hence this conservatism will be reconsidered in the future.
4-81

(iii) Results of setting disposal tunnel pitch
Table 4.5-7 summarises the calculation results of pitch between disposal tunnels describedabove.
Table 4.5-7 Results of minimum HLW disposal tunnel pitch evaluation (rounded values in m)
H12V
Condition
Plutonic Neogene
Based on diameter ofdisposal tunnel (≈ 5 m)

PreNeogene

Plutonic

Neogene

PreNeogene

10

12

10

10

12

10

-

-

16

16

22

16

Storage period
30 years

18

12

18

22

22

22

Storage period
50 years

10

12

10

11

12

11

10

12

16

16

22

16

Mechanicalstability Based on the diameter of thewidened entry of thedisposal tunnel (≈ 8 m1)
Thermaleffects

PEM

Chosen distance for reference design2
1

The diameter of disposal tunnel is widened from 5 m to 8 m at the entry, due to the turning radius of theemplacement machine for the PEM (Supporting Report 4-50).

2

Design in the present report focuses on interim storage of HLW for 50 years (see Section 4.5.1 (1)), so this isused to set the pitch and the 30-year storage data are used only to indicate the consequences of shorterstorage times.

From the table and the reference 50 years waste storage time, the chosen pitch is set bymechanical stability of tunnel. However, if HLW stored for 30 years is considered, thermal
constraints define the selected pitch. For TRU waste, the results of thermal and stability
analyses result in distributions of different waste groups and the vault pitches shown in Figure
4.5-9 (see Supporting Report 4-42).

4-82

(a) Plutonic rock and Pre-Neogene sediments(b) Neogene sediments

Figure 4.5-9 TRU waste group distributions and vault pitch(4) Evaluation of seismic stability
In Japan, where earthquakes often occur, it is necessary for the tunnel to ensure stability not
only at the time of excavation but also in the case of earthquake motion. Because the Neogene
sediments are the weakest of the three representative rocks, seismic stability was assessed forthese as representing the worst case. This showed that the seismic impacts on the local safety
factor and the support stress are small, even in the event of major earthquakes, so that themechanical stability of tunnels is assured during the excavation of the tunnel [99].
For HLW disposal tunnels, the largest observed seismic waves from the 2011 Tohoku
Pacific Ocean earthquake were considered for simulating ground motion, showing that theeffects of even such a huge earthquake during excavation would still meet the criteria fortunnel stability [100].
4.5.3 Design of backfill and plugs
Backfilling and hydraulic plugs can be considered together as shown in Tables 4.2-4 and4.2-5, providing the safety function of reducing RN migration by avoiding flow short-circuits.
In addition, mechanical plugs are installed at the end of disposal tunnels to constrain freeswelling, which decreases the performance of buffer and backfill. The concept of stepwise
backfilling and plug installation is illustrated in Figure 4.5-10 for the case of disposal inthrough tunnels (TT) and only slightly modified for dead-end tunnels (DET).

4-83

(a) Operational period(Backfilling of disposal tunnels and installation of
mechanical plugs)(b) Post-closure(Backfilling of all tunnels, installation of hydraulicplugs, access sealing)

Figure 4.5-10 Concept for backfilling and plugging for the case of TT

During the operational period, it is necessary to backfill disposal tunnels after waste
emplacement and to install plugs to prevent free-swelling of backfill out of the tunnel (Figure4.5-10 (a)). During repository closure operations all connecting tunnels, the access ramp andshafts are backfilled and hydraulic plugs installed at appropriate locations. In addition, accessseals are installed near the surface with the aim of preventing accidental human intrusion(Figure 4.5-10 (b)). Based on the permeability of the surrounding bedrock and EDZ aroundthe tunnel, the backfill material is designed to have sufficiently low permeability to ensurethat tunnels do not become preferential groundwater flow paths. More detailed studies of thisare ongoing.
For example, as shown in Figure 4.5-9, if the permeability of the rock is different in hostrock A and overlying formation B, backfill specifications will be tailored to each of these. To
prevent preferential hydraulic pathways along the access tunnel, hydraulic plugs are installedaround the ends of these. If there are boundaries between geological formations with different
hydraulic properties, hydraulic plugs are also installed near such boundaries. The specificationof backfill material and the installation positions of hydraulic plugs will be determinedaccording to the geological environment found.
Below, the design concept and specifications of backfill materials, mechanical plugs, andhydraulic plugs are shown for connecting tunnels, surrounding tunnels and disposal tunnels.
For access sealing, a structure combining dense concrete and coarse gravel, which is moreresistant to excavation, has been considered to reduce risk of human intrusion [101] [102].
(1) Design of backfill(i) Design requirements and evaluation items for backfill
Design requirements and evaluation items for backfill are listed in Tables 4.5-8 and 4.5-9.
A post-closure safety function relates to restriction of migration of RNs (Tables 4.2-4 and 4.24-845), which leads to low permeability being a design requirement, along with engineering
practicality.
Table 4.5-8 Backfill design requirements
Designrequirements
Lowpermeability
Engineering
practicality

Detailed description of requirement forconcepts assessed
Tunnel is not a preferential groundwaterflow path
Ability to manufacture and install torequired quality levels based on existing
technologies or those which will be
realised in the near future

Specifications
Material, effective dry density ofclay
Material, dry density, effective dry
density of clay, composition of
backfill material

Table 4.5-9 Backfill evaluation items
Evaluation item
Restriction of swelling of buffer

Specifications
To limit the swelling of buffer material and hydraulic plug
components (in case of clay plug) after saturation commences

Tunnels backfilled with suitable material should not act as preferential RN migration paths.
The location of backfill in disposal tunnels/vaults is illustrated in Figures 4.4-18 and 4.4-25,
for HLW and TRU, respectively. For H12V, in the current design without a borehole cap,
backfill has the additional function of restricting buffer swelling out of emplacement holes. A
similar function applies to backfill materials in contact with hydraulic plugs, whose
specification includes a clay barrier with properties similar to buffer (see Section 4.5.3 (3)).
Although possibly less critical for performance, backfill will also constrain swelling of PEM
buffer after failure of the handling shell. Therefore, as an evaluation item, restriction of buffer
swelling is selected to cover all these cases.
After selecting the material to be used for backfill, the density required to satisfy the design
requirements has to be specified. This allows it to be assessed for this specification, toconfirm that it is appropriate. Here, analysis was performed based on the evaluation of longterm stability of buffer material (more details in Supporting Reports 4-17 and 4-28).
(ii) Specification of backfill materials
In repository construction, a large amount of excavated spoil is generated. Spoil consistingof crushed host rock can be considered as potential backfill, particularly from the standpointof ease and economy of production [65]. However, it is extremely difficult for crushed rockbackfill to meet the goal of having the same low permeability as the original rock. Therefore,
mixing bentonite with crushed rock is assumed, which will produce the low permeability
required. The particle size of crushed rock is defined from the viewpoint of workability. At
this time, chemical changes during surface storage (e.g., acidification due to the oxidation of
pyrite) are not considered but, in the future, it will be necessary to confirm that this does not
impact performance.
As crushed rock will have specific properties that cannot be defined until specific sites are
identified, silica sand is considered as a substitute for backfill specification. Also, bentonite
will be represented by Kunigel V1, as for buffer definition. In setting the specification of
backfill materials, Na-bentonite alteration by Ca exchange was also considered. Groundwaterflow analysis to assess the impact of buffer permeability showed that, below a certain value,
4-85

flow was unaffected due to the role of the EDZ (details in Supporting Report 4-43). Thismeans that the impact of backfill is inherently limited and that, additionally, hydraulic plugs
are required in specific locations. Based on this analysis, the upper limit of hydraulicconductivity of the backfill was set as 10 times the average value of the host rock.

3 ]3)
(Mg/m
density[Mg/m
Dry
乾燥密度

In addition, for the engineering practicality design requirement, a dry density of 1.8 Mg/m3
was set based on the upper limit that can be obtained by in-situ compaction in a tunnel. Here,the lowest rock hydraulic conductivity (2.0 × 10-9 m/s; Table 4.2-7) is used in setting therequired density and composition generally applicable to all rocks. The details of the set
backfill specifications are given in Supporting Report 4-43, based on EDZ hydraulic andmechanical properties given in Supporting Report 4-44. Figure 4.5-11 illustrates compaction
properties of bentonite mixed with crushed rock (represented by sand or gravel). In general,the mixture becomes less compressible as the density and content of sand increases [103].

Compaction curve
Gravel content: 50 wt.%
Degree of saturation : 100%

85 wt%
80 wt%
90 wt%
95 wt%
100 wt%Content of sand or gravel
ケイ砂および礫の混合率
50 wt%

含水比[%] (%)
Water
content
Figure 4.5-11 Compaction curves of bentonite mixed with sand or gravel [65]From the viewpoint of restriction of buffer expansion, a high density is preferable, but otherrequirements on the backfill need also to be considered. From Figure 4.5-11, 15% bentonite
and 85% crushed rock would give the highest maximum dry density after compaction.
The backfill construction technology for disposal tunnels is considered in Section 4.5.7 (2),while backfilling of other tunnels at repository closure is covered in Section 4.5.7 (3). The
method of manually-operated, in-situ compaction is selected in cases without accessconstraints, while a remote-operated spraying method is considered for areas where worker
access is not allowed. To assess engineering practicality, these two methods have been
examined (see Supporting Report 4-43). From the above, the backfill specifications
summarised in Table 4.5-10 were set for these 2 construction options.

4-86

Table 4.5-10 Specifications of backfill
Construction
methodComposition

Dry density(Mg/m3)

Effective clay
dry density(Mg/m3)

Target areas

Compaction

Bentonite 15%
Crushed rock 85%

1.8

0.6

Disposal tunnel/vault(H12V/TRU), connecting
tunnels, approach tunnel (TRU
waste), surrounding tunnels,
access ramp, access shafts

Spraying

Bentonite 50%
Crushed rock 50%

1.6

1.2

Disposal tunnel (PEM)* These reference specifications are derived for silica sand and would be modified to account for the propertiesof specific host rocks.

For these backfill specifications, the effectiveness of restriction of buffer swelling wasassessed by a simple numerical analysis that considers buffer swelling during saturation,
assuming that backfill is emplaced before buffer swelling commences. It was confirmed that,in this case, the change in density of the buffer material was slight and remains within thespecification range for a long period (details in Supporting Report 4-45). However, thepracticality of the operations involved would need re-assessment to confirm acceptableperformance on a site-specific basis.
(2) Design of mechanical plugs(i) Mechanical plug design requirements
Mechanical plugs are planned to be installed at open ends of HLW disposal tunnels or TRU
vaults to prevent backfill material from swelling outwards (see Figure 4.5-12).

4-87

(a) HLW disposal tunnel (Example of H12V)(b) Disposal vault for TRU waste
Figure 4.5-12 Concept of the installation position and structure of mechanical plugs

Design requirements for mechanical plugs are structural integrity and engineering
practicality, which are given in Table 4.5-11 along with resulting specifications. In order tosatisfy these design requirements and to provide a solid structure that can withstand swellingpressure of backfill materials, the plug is assumed to be made of reinforced concrete, sincesimilar plugs have been demonstrated in URLs. The application of other materials (e.g. steelplug) may be considered after more detailed studies to determine installation technology
requirements for specific sites and disposal concepts.
Table 4.5-11 Mechanical plug design requirements
Designrequirements
Structural
integrity
Engineering
practicality

Detailed description of requirements forconcepts assessed

Specifications

Ability to withstand swelling pressure of backfill

Plug thickness, notch shape,
reinforcement

Ability to construct to required quality levels basedon existing technologies or those which will be
realised in the near future

Design, formulation andstrength of concrete

The mechanical plug concept currently considered does not provide a hydraulic barrier,which would require the ability to withstand high groundwater pressures during construction,although this option might be considered in the future. Thus, a permeable layer is placed
between the mechanical plug and the backfill material to facilitate drainage and to reduce thegroundwater pressure acting on the plug (see Figure 4.5-12), allowing drainage from thebackfilled tunnel during the entire operational period without a risk of backfill erosion.
As the basis for design, the load acting on the plug and the amount of water to be drained
from the disposal tunnel have to be defined. The structural specifications and material
4-88properties of the plug and associated drainage material are set for these conditions. With
regard to the plug body, the generated cross-sectional force and the reaction force from therock are calculated, with structural integrity examined with respect to the rock strength,
sliding force and stress level over the cross section. If the stress levels generated in the rockand plug are below the allowable limits, a concrete composition satisfying the design
requirements for construction practicality is set to satisfy the design requirement of structural
integrity. For drainage materials, the specifications should ensure sufficient drainage capacity
for expected conditions.
(ii) Mechanical plug specifications
Figure 4.5-13 illustrates the structure and dimensions of the reference mechanical plug,
although it is recognised that this would need to be tailored to specific sites and disposal
concepts.

Cross section of the disposal tunnel
Structural elements and dimensions
Figure 4.5-13 Design the mechanical plug

Drainage pipes (gradient 0.1%) are installed within the plug, and a permeable layer(crushed rock) is provided on the disposal tunnel side of the plug, enabling efficient drainage
during the operational period. At the boundary between the permeable layer and the backfill, afilter material (e.g. glass fibre fabric) is installed to prevent the backfilling material being
eroded. The design of this filter and confirmation of its function will require moredevelopment work. In addition, an embedded formwork (made of glass fibre reinforced
concrete) is used for the installation of the permeable layer and the manufacture of themechanical plug.
The load acting on the plug results from the swelling pressure of backfill and buffer, creepof rock and any hydrostatic pressure that would result if drainage is insufficient. In particular
for TRU waste vaults containing buffer, this underlies the backfill and the total swellingpressure of both acts on the plug. Structural stability calculations have been performed forthese set loads for the representative 3 host rocks, using the standard design strength of
concrete of 30 MPa, 2.5 m concrete thickness (reinforcement specification D32 @ 150 mm)
and a notch depth of 1.5 m. The details of the design of the mechanical plug are given in
Supporting Report 4-46.

4-89

(3) Design of hydraulic plugs(i) Design requirements for hydraulic plug
During closure, hydraulic plugs are set at key locations to prevent short-circuit flows of
RNs though tunnels. These provide an area with sufficiently low permeability and also blockpotential pathways through the tunnel EDZ. EDZ hydraulic and mechanical properties are
given in Supporting Report 4-44.
Design requirements for hydraulic plugs are shown in Table 4.5-12, which focus on postclosure restriction of RN transport (Table 4.2-4 and Table 4.2-5) due to low permeability,
together with the associated engineering practicality requirement.
Table 4.5-12 Hydraulic plug design requirements
Designrequirements
Low permeability
Manufacturability
and engineering
practicality

Detailed description of requirements forconcepts assessed
Limit the flow of groundwater along the tunnel andits surroundings in combination with the tunnel
backfill
Ability to manufacture and install to required
quality levels based on existing technologies or
those which will be realised in the near future

Specifications
Material, effective dry
density of clay,
Plug structure
Design, dry density,
mixing ratio of materials

After selecting materials, a specification is developed that satisfies the design requirements.
The material specifications (formulation and density) are provisionally set and then the
structure of the plug is examined by sensitivity analysis to develop a specification that meetsthe low permeability design requirement.
Here, by providing notches in the bedrock that exceed the width of the EDZ, the design
should block this transfer path of RNs. Since the installation of the notch in the bedrock itself
is a new tunnel excavation, a construction method is selected so that no significant EDZ isformed around the notch. The detailed design of this is described further in Supporting Report
4-47, but details would depend on the actual host rock properties.
(ii) Specification of hydraulic plug
The hydraulic plug currently considered is a clay plug in which silica sand is mixed withbentonite, with a material specification that can ensure a low hydraulic conductivity < ≈ 1.0 ×
10-10 m/s (blending of 70% Kunigel V1 bentonite and 30% silica sand, dry density 1.6
Mg/m3). Although the construction technology of hydraulic plugs is described in Section
4.5.7 (3), the design developed was based on a stack of compacted bentonite blocks. For thismethod, practicality and expected performance have been confirmed by full-scale in-situ
testing [104]. Further development work may be needed, especially to ensure that the specificplug design is applicable to actual host rock conditions.
Figure 4.5-14 illustrates the structure and dimensions of the hydraulic plug. It is
manufactured by cutting notches into the tunnel to form two walls (thickness 1.0 m; See
details in Supporting Report 4-47), with buffer sandwiched between them (see Table 4.4-8 forspecifications of bentonite buffer).

4-90

Figure 4.5-14 Structure and dimensions of the hydraulic plug

The notch depth of the ends of the hydraulic plug into the bedrock is set to 1.5 m (EDZ +0.5 m) for the assumed case in Figure 4.5-14, based on groundwater flow analysis withbentonite permeability coefficient and notch depth as variables. These depths may need to bemodified for actual sites, depending on their mechanical properties. In addition, the width ofthe plug is set such that the permeability of the entire plug is less than that of the host rockbefore excavation. From the groundwater flow analysis for plutonic rocks (highest hydraulicconductivity of the three rock types), this width was set to 3.0 m. As a result, the total lengthof the hydraulic plug is 5.0 m. The details on design and the related groundwater flow
analyses are given in Supporting Report 4-47.
4.5.4 Design of the underground facility layout
Section 4.5.1 (2) outlined the principles of design of the repository layout based on the
SDMs developed in Section 3.3.3 (4) (ii). To simplify the demonstration of the process, thereference repository depths derived in Section 4.3 are taken as fixed although, for specificsites, options for depth optimisation would be considered. Firstly, therefore, based on thespatial distribution of major geologic structures and groundwater flow characteristics at thedisposal depth, preferred areas for locating underground facilities are identified. Next,
disposal panels are laid out in such areas, to assure that they are of sufficient size and suitablegeometry. For a provisional layout of disposal panels, required access and transportationroutes will be arranged so that construction and operation can be performed efficiently andsafely with appropriate zoning to separate radiation-controlled and non-controlled areas. The
tunnel network will be refined to ensure work and material flows can be managed and accessto/from the surface provided as required for services such as power, ventilation and drainage.
Finally, based on estimates of the percentage of preferred areas where waste emplacement
will be difficult (e.g. due to higher water inflow), the layout is checked to ensure that it ispossible to find sufficient reserve sections so that the target inventory can be managed.
4-91

In the following, this design concept is illustrated for the 3 representative host rocks.
(1) Identifying preferred areas(i) Design requirements
Table 4.5-13 presents design requirements for preferred disposal areas. In terms of thesafety functions and repository components discussed in Section 4.2.4, no relevant safetyfunction is directly applicable to preferred areas, so these are chosen to meet the post-closure
safety function of the geological environment (Tables 4.2-4 and 4.2-5) in restricting migrationof RNs.
Table 4.5-13 Design requirements for preferred disposal areas
Detailed description ofrequirements for concepts assessed
Conditions suitable for excavation ofEase of excavation
tunnels
The amount of drainage during
Reduce the amountconstruction and operation is asof water inflow
small as possible
Relatively long travel paths in thehost rock and low groundwater flux
Restrict migration of(giving a relatively long transport
RNs after closuretime)* ○: Mandatory requirement, -: preferred requirement (see Section 4.1.2)
Design requirements

Classification*

Specifications
Setting of preferreddisposal areas
Setting of preferreddisposal areas
Setting of preferreddisposal areas

In addition, it is necessary to excavate many long disposal tunnels for HLW and hence,from the viewpoints of engineering practicality, safety and also cost of countermeasures(treating drainage water, grouting etc.), design requirements for low water inflow and ease of
excavation are included.
Key geological characteristics that need to be considered for these design requirements are
the layout determining features (LDFs) that bound the preferable areas (see discussion in
Supporting Report 4-48), as noted in Table 4.5-14. Although LDFs may include folds,
lithological discontinuities, etc., the focus here is on the major faults that are encountered inall host rocks in Japan. For example, as known from past tunnel construction, excavation
through such faults often causes perturbations such as rock bursts or flooding [105], whichshould be avoided to the extent possible. The width of the damaged zone affecting tunnel
construction is assumed to be within 1% of the fault length (see Supporting Report 3-35), andthis should be also included as part of the LDFs.
Table 4.5-14 Design requirements related to LDFs
Risks to consider

Geological characteristics of LDFs
to consider

Mechanical collapse of thetunnelMechanical properties

Flooding

Hydrogeological properties

Reduce the amount ofwater inflow

Increase in drainage
volume during operation

Hydrogeological properties

Restrict migration of
RNs after closure

Migration of RNs by fast
groundwater flow

Hydrogeological properties.
Bedrock permeability, groundwater
travel distanceDesign requirements

Ease of excavation

4-92

In addition, high water inflow into excavated tunnels presents problems from the viewpointof drainage water management [105]. In terms of migration of RNs, LDFs may serve as
preferential flow paths towards the surface and their impact depends on both thehydrogeological characteristics of the LDF and flow paths through the host rock to them.
Thus, LDF characteristics need to be assessed in terms of these three design requirements,
with particular consideration of the extent of rock influenced by the presence of the LDF andthe degree to which the LDF impacts release and transport of RNs from disposed waste (seeSupporting Report 3-35 for further details). In addition, as there is likely to be uncertainty inthe position of the fault, it was decided that tunnel emplacement within a few 10s m of thecurrent fault boundary (taking into account associated damaged zone) should be avoided. At
actual sites, the characteristics of the LDF at repository depth need to be assessed, as theassumption of preferential flow paths to the surface is inconsistent with general evidence ofvery old deep groundwater.
(ii) Concept of design within preferable areas
In the design of the underground layout, the 3 SDMs from Section 3.3.3 (4) (ii) are used
(Figures 3.3-29, 3.3-35 and 3.3-41). From these SDMs, maps of preferred areas at repository
depth are developed based on the conservative assumption that all LDFs are potential
preferential flow paths, as illustrated in Figure 4.5-15. A very simple measure of “relative
travel time” of groundwater through the host rock was calculated based on the specificdistance from a targeted point in a host rock divided by the modelled Darcy velocity. The
specific distance was tentatively determined to be 500 m based on the scoping calculations.
More information on the calculation these relative travel times is given in Supporting Report
4-49.
For plutonic rocks, Figure 4.5-15 (a), groundwater flow is driven by the hydraulic gradientand the relative travel times tend to be short up-gradient from LDFs. Therefore, the mapped
preferable areas are larger blocks that avoid such zones.
For Neogene sediments, Figure 4.5-15 (b) can be compared to the repository-scalegeological SDM (Figure 3.3-34); it is then clear that both sandstone layers and faults tend tohave higher permeability, giving short relative travel times. The preferable areas thus lie inthe lower conductivity mudstone layer. Similarly, for Pre-Neogene sediments, Figure 4.5-15(c) and can be compared with Figure 3.3-40 (left); relative travel times tend to be short nearboth faults and chert blocks. Since the permeability is equally low in mudstone-dominant andsandstone-dominant matrices, it was decided to prioritise the mudstone-dominant matrix as itis present over a larger area.

4-93

(a) Plutonic rock (1,000 m below surface)(b) Neogene sediment (500 m below
surface)(c) Pre-Neogene sediment (1,000 m below surface)
Figure 4.5-15 Distribution of relative travel times and selection of preferable areas(2) Setting of disposal panel shape
The shape of disposal panels (groups of disposal tunnels and the tunnels connecting them)is set following consideration of construction and operational methods.
(i) Design requirements
Design requirements considered to set disposal panel shape are captured in Table 4.5-15.
Here, main issues related to engineering practicality (see Table 4.1-1) are captured in terms ofoperational practicality, disposal volume and ventilation air speed.
4-94Table 4.5-15 Design requirements on disposal panel shape
Designrequirements
Panel shapefacilitates
operations
DisposalvolumeVentilation air
speed limit

Classification*

Detailed description ofrequirements valid for concepts
assessedSpecifications○

Ensure minimum turning radius ofequipment is allowed for

Tunnel junction angles
Junction area widening○

Able to dispose of specified waste
inventory○

Air speed in all tunnels is below
the specified limits

Number and length ofdisposal tunnels
Ventilation plan/requirements
Number of disposal tunnels
ventilated*○: Mandatory requirement, -: preferred requirement (see Section 4.1.2)(ii) Issues associated with disposal panels
Different disposal panel configurations are being considered in countries planninggeological disposal, as discussed further below.
(a) Through and dead-end tunnels
For HLW disposal, two fundamental design options involve either parallel groupings ofthrough tunnels (TTs: accessible from both ends, Figure 4.5-16 (a)) or dead-end tunnels(DETs: accessible from only one end, Figure 4.5-16 (b)).
TT layouts have been continuously studied in Japan since H12. Because they are accessible
from both ends, this may have some advantages in terms of work flow during construction
and operation [6]. Additionally, the surrounding tunnel provides additional characterisation
information to reduce the risk of surprises during emplacement tunnel excavation. Suchpanels were also previously examined in Switzerland (for crystalline [106] and sedimentary[107] host rocks).
Use of DETs allows the length of each to be tailored to the preferable rock area available
and avoids the need for a surrounding tunnel. Such flexibility is particularly important when
the rock structure is complicated but can be well characterised from the surface, e.g. incountries such as Finland [108], Sweden [109] and recently in Switzerland [110], or thoseconsidering alternative disposal options (in horizontal boreholes, France [111] or caverns
Canada [112]). It is also worth noting here that dead-end tunnels are of particular benefit ifthey can be laid out in a way that reduces the risk of access routes acting as preferential flow
paths. The disposal panel concept and shape for HLW and TRU waste has to be tailored to thepreferred disposal areas indicated in Figure 4.5-16.
For H12V, both through and dead-end tunnels were considered [6], [109]. As described in
Section 2.3.4, the H12 repository concept was the starting point for design and hence, for
H12V, TT as in H12 are prioritised. However, given that H12V is already rather inefficient interms of use of space [6], the additional requirements for the surrounding tunnel may cause
this option to be re-assessed for sites where available repository footprints are restricted.

4-95

(a) Through tunnels (TT)(b) Dead-end tunnels (DET)
Figure 4.5-16 layout example of underground facilities(Source: (a) NUMO, 2011 [1], (b) SKB, 2010 [108])

For the PEM case, the logistics of tunnel construction/waste emplacement are simpler than
H12V, so TT may have fewer advantages. Further, similar disposal concepts (SKB KBS-3H[14] and the Belgian super container [113]) are based on dead-end tunnels, so these are also
considered here.
For TRU waste, the disposal footprint is relatively small and hence TT as in TRU-2 aretaken as a reference. The disposal vaults involved use a large amount of concrete and mortar
for construction of the structural body, infill, etc., so the length of such vaults is constrainedby practical concrete pumping distances. For waste groups with small volume, however, DET
options could be much more cost-effective.
(b) Disposal tunnel junctions with the connecting tunnel
The crossing angle between disposal tunnels and the connecting tunnel has been designed inconsideration of the minimum turning radius of the type of equipment required to pass4-96

through the junction. In general, orthogonal junctions are favourable from the viewpoint of
mechanical stability (see Supporting Report 4-50).
As described in Section 4.4.3 (1) (i), the H12V concept is based on H12, which considered
TT and rail transportation of waste underground. This was the starting point for design andtechnical development of the EBS emplacement device. In this case, the junction between theconnecting tunnel and the disposal tunnel is a curve set by the practical limits on rail transport,
resulting in a panel with a parallelogram shape (see Figure 4.5-17 and Supporting Report 450).
For the PEM, road transportation was considered, which allows a smaller turning radius andthe option of widening access to the disposal tunnel to provide space to turn in from an
orthogonal tunnel junction. In addition, the option of dead-end disposal tunnels is considered,with dead-end tunnels/road transport also being considered for H12V in the future.
For TRU waste, as mentioned in Section 4.4.3 (2), road transport is assumed and, to assure
mechanical stability, an orthogonal junction between the connecting tunnel and the disposal
vault is assumed.
(c) Number of disposal panels
For HLW, following definition of the shape of disposal panels, the number of these is set byconsidering excavation and operational work flows (for these running in parallel) andpractical aspects such as ventilation wind speed. This results in between 6 and 8 panels, which
are currently assumed to contain the same quantity of HLW, although this could be further
optimised on a site-specific basis. Details of the setting of the number of HLW disposal
panels are shown in Supporting Report 4-51.
In the case of TRU waste, the entire repository is treated as one panel, although separated
into different areas for the different TRU waste groups.
(iii) Specification of disposal panel options(a) HLW
Examples of panel concepts for H12V and PEM are illustrated in Figure 4.5-17.

(a) TT (H12V)(b) DET (PEM)
Figure 4.5-17 HLW disposal panel shape and dimensions (plutonic rock)4-97

The details of the designs and associated analysis are given in Supporting Report 4-52 forthe 3 SDMs, the results of which are summarised in Table 4.5-16. For H12V in the case ofplutonic rock and Neogene sediments, parallelogram-shape TT panels are selected as areference (Section 4.5.4 (2) (ii) (a)). The disposal tunnel access radius of curvature, is set to
30 m [114] and the pitch between disposal tunnels is as set in Section 4.5.2 (3) (iii). Althoughcertainly a parameter that would be considered further for site-specific boundary conditions,the number of panels was set to 6 and the length and number of disposal tunnels thendetermined.
Table 4.5-16 HLW disposal panel options
Disposal conceptPlutonic rock

Neogene sediments

Pre-Neogene sediments

H12V

TT: 6 panels

TT: 6 panels

DET: 8 panels

PEM

DET: 6 panels

DET: 6 panels

DET: 8 panels

For Pre-Neogene sediments, the SDM is more complex and DET panels were selected (8 inthis case) in order to provide flexibility to make best use of preferable areas. For the PEM,
DET panels are chosen for all SDMs (Section 4.5.4 (2) (ii) (a)), with tunnel pitches as derived
in Section 4.5.2 (3) (iii).
(b) TRU waste
The TRU waste disposal area treated as 1 panel (Section 4.5.4 (2) (ii) (a)), with vault
lengths set on the basis of practical concrete pumping distances. Although pumping distances
have been reported that exceed 1,000 m, in order to assure quality, < ≈ 300 m is considered to
be a general limit [115]. With further consideration of limited clearances, remote operation
under radiation-controlled conditions, etc., additional conservatism leads to a limit for
concrete pumping to be set at about 200 m, which may be re-assessed during later
optimisation, especially as it may be practical to pump concrete from either end of the vault.
However, the complexity of TRU waste means that a very large range of processes impact
their evolution with time. The current assessment is illustrative only and focuses on relatively
well-defined impacts. However the current analysis will form the basis for morecomprehensive analyses in the future that include additional processes such as gas
release/transport and effects of microbes.
The layout of vaults was set in consideration of the following:

Wastes containing RNs that have high solubility and low sorption, especially I-129 andC-14 (Gr.1 and Gr.2), are positioned upstream from other groups, so that groundwatertransport distances will be maximised assuming no future changes in flow directions.



Gr.3 wastes, containing nitrate, are located to minimise potential effects of leachednitrates on other waste groups, and hence off the flow path/upstream side of all othergroups.



Wastes with higher thermal output (Gr.2 and Gr.4H) are placed as far towards theedges of the disposal panel as possible.

The impact of heat generated is minimised by alternating vaults containing higher andlower thermal output waste.

The general arrangement of vaults based on these considerations is shown in Figure 4.5-18.
4-98

Figure 4.5-18 Constraints on location of TRU vaults (container for MHHRW not yet specified)Resultant TRU waste disposal panel layouts are shown in Figure 4.5-19 (based on anassessment described in Supporting Report 4-42).

4-99

Gr.4H MHHRW
Gr4H_ハル缶・インナーバレル②

Gr3_ドラム缶③
Gr.3
Drum

Gr.4H MHHRW
Gr4H_ハル缶・インナーバレル①

Gr3_ドラム缶②
Gr.3
Drum
Gr3_ドラム缶①
Gr.3
Drum

Gr.4L
Box Cont.
Gr4L_角型容器
Gr.4H
Drum
Gr4H_ドラム缶

Gr4H_ハル缶・インナーバレル②
Gr.4H MHHRW

Gr.3 Drum
Gr3_ドラム缶③

Gr4H_ハル缶・インナーバレル①
Gr.4H MHHRW

Gr.3 Drum
Gr3_ドラム缶②

Gr.4L
Box Cont.
Gr4L_角型容器

Gr.4L Drum
Gr4L_ドラム缶②

Gr.2 Canister
Gr2_キャニスター④
Gr.2 Canister
Gr2_キャニスター③
Gr.2 Canister
Gr2_キャニスター②

Waste packagetransportation pathways
Disposal vault
shaft
立坑

Gr.2 Canister
Gr2_キャニスター①

shaft

立坑

Gr.1
Drum
Gr1_ドラム缶

Approach tunnel

Gr.4L Drum
Gr4L_ドラム缶①To access ramp

構造躯体
Concretepit
緩衝材
BufferGr.2 Canister
Gr2_キャニスター②

処分坑道
Disposal vault

Gr.2 Canister
Gr2_キャニスター①

埋め戻し材
Backfill

waste
廃棄体

Gr.2 Canister
Gr2_キャニスター⑤

Gr.2 Canister
Gr2_キャニスター③

Schematic view of cross section of disposal tunnel

力学プラグ
Mechanicalplug透水層
Drainage layer

Gr.2 Canister
Gr2_キャニスター⑥

Waste package transportation
pathways
Surrounding tunnel

Gr.2 Canister
Gr2_キャニスター④

Surrounding tunnel

EDZ

Principal Stress

Gr.4L Drum
Gr4L_ドラム缶②

アクセス斜坑へ

Tunnel access to 取付坑道
vault

Hydraulic gradient

Gr.4L Drum
Gr4L_ドラム缶③727.9m728 m

Gr.2 Canister
Gr2_キャニスター⑤

Gr.3 Drum
Gr3_ドラム缶①

Gr.4H
Drum
Gr4H_ドラム缶

Downstream ofhydraulic pathways

Gr4L_ドラム缶①
Gr.4L Drum

Hydraulic gradient461.5m461 m

584.8m585 m

Downstream of other vaults

547.2m547 mTunnel access to
vault
取付坑道
水抜き管
Drain pipe

Approach tunnel

Disposal vault
Shaft
立坑

Shaft
立坑

Gr.1
Drum
Gr1_ドラム缶

アクセス斜坑へ

To access ramp

Drain

排水
力学プラグ

Mechanical plug(a) Plutonic rock and Pre-Neogene sediment(b) Neogene sediment
Figure 4.5-19 TRU waste disposal vault layouts

Compared to plutonic rocks and Pre-Neogene sediments, the larger disposal area for the
Neogene sediments results from its lower strength, requiring larger vault spacing to assure
mechanical stability. As mentioned above, Grs.1 and 2 are upstream and Grs.4L and 4H
alternate to spread thermal load. Gr.3 was placed away from all other groups at thedownstream end of the panel.
The surrounding tunnels run around the parallel disposal vaults, with waste input from thesame end of each. Further, as mentioned in Section 4.5.4 (2) (ii) (b), waste package transport
is assumed to be road based and the junction of the vault with the surrounding tunnel is a rightangle. The link from the surrounding tunnel transporting waste to the vault is level with thevault floor, while the opposite surrounding tunnel is at roof level (see Figure 4.5-12 (b) or
Figure 4.5-19 inset), with more details in Supporting Report 4-53. It should also be noted here,that there are some DET, specifically for Gr.1 and some Gr.4L waste. This illustrative designis based on safety and efficiency considerations for construction of such a relatively large
cross-sectional area vault.
(3) Arrangement of disposal panels
Figure 4.5-15 shows the preferred locations for disposal in the different SDMs (outlined in
Section 4.5.4 (2) (iii)), which are now considered in more detail.

4-100

(i) Design requirements
The design requirements for laying out disposal panels are given in Table 4.5-17, based on
pre- and post-closure safety functions for repository components from Section 4.2.4 (Table
4.2-3 and Tables 4.2-4 and 4.2-5, respectively).
Table 4.5-17 Design requirements for panel layout
Design requirement

Classification*

Detailed description of requirements
for concepts assessed

Specifications

Restricting
groundwater flow
along the disposal
tunnel-The direction of the disposal tunnel isorthogonal to the current hydraulicgradient

Orientation of disposal
tunnels

Assuring stability ofthe disposal tunnel-

Managing
groundwaterdrainage

Reduction of waste
interactions in case ofco-disposal-○

The direction of the maximum
principal stress and orientation of thetunnels are optimal
The slope of the disposal tunnel andthat of the panel overall should be set
so that water draining from radiation
control areas will not move to otherdisposal areas.
Manage drainage by gravity flow tothe extent possible
In the case of co-disposal, arrange thedisposal sections so that the HLW andTRU disposal sections do not interact
thermally or chemically.

Orientation of disposal
tunnels, specificationsof tunnels (support etc.)
Slope of disposal
tunnels andsurrounding/connecting
tunnels,
Sump volumes andlocations

Arrangement of panels
for HLW and TRU* ○: Mandatory requirement, -: preferred requirement (see Section 4.1.2)

Requirements include restricting groundwater flow along the disposal tunnels and assuringstability of the disposal tunnel, which impact the location and orientation of the disposal panel.
Also, the requirement for engineering practicality leads to arrangement of tunnel slopes sothat gravity provides required drainage. Although forced drainage by pumps is possible, thistends to be more complex and is not failsafe in the event of power loss.
For the case of HLW and TRU waste co-location, interactions due to heat from HLW or
chemical plumes from TRU waste (e.g. high pH, organic substances and nitrates) should be
kept as low as possible [10] [11] [116]. Therefore, reduction of interactions is set as a design
requirement for post closure safety.
(ii) Determination of disposal tunnel orientation
From Table 4.5-17, orientation of tunnels needs to be considered in terms of bothmechanical stability and groundwater flow along tunnels. In terms of the former, theorientation of tunnels can be optimised for mechanical stability with respect to the local
horizontal stress field. However, it is suggested to be generally preferable that disposal
tunnels are orthogonal to the hydraulic gradient, to avoid them acting as preferential flow
paths, but additional numerical simulations may be required to assess this. If theserequirements conflict, a more detailed analysis is needed to determine how trade-offs lead toan optimal solution.
For the host rocks examined in this report, the Neogene sediments show stress anisotropy
with the stress maximum in the same direction as the hydraulic gradient. In this case, the4-101

safety of workers requires that stability is given more weighting, as post-closure safety can beshown even with non-optimal groundwater flow conditions (see Section 6.4). For other rocktypes, the two requirements are compatible and setting tunnel direction is straightforward. The
details of determination of the direction of the disposal tunnels for each SDM are given in
Supporting Report 4-54.
(iii) Slope of disposal tunnels
The HLW disposal tunnels are excavated with a slope to provide the gradient driving
natural gravity drainage during construction and operation. During excavation, work
progresses up slope to facilitate such drainage. The approach to drainage during backfilling
differs between TT and DET cases (Figure 4.5-20 (a), (b)). In the TT case, where both ends ofthe disposal tunnel link to connecting tunnels, it is necessary to install a mechanical plug at
one end before backfilling commences, so either up-slope or down-slope options are possible.
On balance, it was considered advantageous to backfill up-slope if drainage is assured throughthe mechanical plug (Figure 4.5-20 (a)). For the DET case, the only option is clearly
backfilling downslope, reversing the direction of original excavation (Figure 4.5-20 (b)).
However, simultaneously balancing all requirements for excavation, drainage, ventilation andbackfill installation logistics is tricky, especially for TT tunnels which allow fewer options.
More development work here will be needed in future stages of the design work.
Nevertheless, taking the above into consideration, the slope direction was set, with areference gradient of 1% assumed [98].

Figure 4.5-20 Tunnel gradients and direction of backfilling

The TRU vaults are currently designed to be horizontal, based on consideration of
construction, buffer emplacement, transportation of waste packages by gantry cranes andinfilling. A drainage channel with a slight gradient would allow natural flow to the lower
approach tunnel during construction and operation. Approach tunnels would slope down fromthe disposal vault towards the connecting tunnel, providing drainage while these arebackfilled (Figure 4.5-20 (c)).

4-102

(iv) Arrangement of panels for co-disposal
To reduce chemical and thermal interactions between disposal panels for HLW and TRU
waste, these are separated by as much distance as required, with TRU waste down hydraulicgradient from HLW [116]. This leads to an outline layout that is evaluated in the post-closure
safety assessment (see Section 6.4), allowing it to be modified if any issues are identified.
(4) Arrangement of connecting and access tunnels
After arranging the disposal panels, the necessary connecting and access tunnels (ramps andshafts) can be laid out so that work and material flows for construction and operation, together
with service routes for ventilation and drainage, are provided in an optimal manner. In
addition, provisions for transferring HLW overpacks/PEMs need to be arranged at thejunction of access and connecting tunnels.
(i) Design requirements
Table 4.5-18 shows the design requirements for layout of access and connecting tunnels.
For such tunnels, requirements relate to operational safety and engineering practicality, as set
in Table 4.5-2.
Because of planned parallel construction and operation (for HLW), zoning of undergroundareas is essential and thus independent work flow routes (material and manpower), ventilationand drainage are required for radiation-controlled (access restricted to nuclear workers) andnon-controlled zones (construction worker access). For each zone, practicality of construction(mechanical stability) and operation (limiting ramp gradients) needs to be complemented withappropriate measures for responding to relevant perturbations (e.g. assuring escape routes).

4-103

Table 4.5-18 Design requirements for layout of access and connecting tunnels
Designrequirement

Classification*

Detailed description of requirements
for concepts assessed○

Work flow lines for construction
materials, equipment and spoil, areindependent and do not intersect those forwaste emplacement and backfilling

Ensureindependent
ventilation○

Appropriate ventilation routes are
assured, which are independent in theradiation-controlled and non-controlled
zones

Ensureindependent
drainage○

Appropriate drainage routes are assured,
which are independent in the radiationcontrolled and non-controlled zones

Ensuring
independenceof work flowlines

Securing
evacuationroutes○Access slope
limit

○

In case of credible accidents, appropriate
escape routes are assured and, in casethese cannot be accessed, suitable
emergency shelters are provided
Set gradient within limits for thetransportation method used

Ensuring
Mechanical stability is assured in allmechanical○tunnelsstability oftunnels*○: Mandatory requirement, -: Preferred requirement (see Section 4.1.2)

Specifications
Number and layout oftunnels and theirrelationship toaccess routes (ramp/shaft)
Placement of tunnels,
Separation of air intake andexhaust routes, placement
and number of accessroutesArrangement and slopes ofcontact tunnels,
location and volume ofsumps
Evacuation routes,
emergency shelters
Access ramp slope
Pitch between connecting
tunnels, management oftraverses through major
structural features

The arrangement of connecting and access tunnels depends on the specific layout ofdisposal panels, as outlined in Section 4.5.4 (2) (iii), and varies significantly for HLW andTRU waste, as the latter does not involve parallel construction and operation. Since the panel
layouts are tailored to each SDM, final design can be developed only after such tailoring(Section 4.5.4 (6)). The following thus establishes only the basic design principles andconstraints and does not include implementation practicalities (e.g. re-zoning after completionof excavation of a panel).
(ii) Layout of connecting and access tunnels(a) HLW TT panels
A conceptual sketch of tunnel layout for HLW TT panels is shown in Figure 4.5-21,
highlighting the main features to allow parallel construction and operation. It should beemphasised that the work flows are illustrative only and would be specified on the basis ofsite-specific repository layout, along with required procedures for re-zoning panels after
construction is complete.

4-104Figure 4.5-21 Conceptual sketch of tunnel layout (HLW TT panels)

The associated sequence for construction and operation of 6 disposal panels is given in
Figure 4.5-22.

Emplacement
panel
Connecting
tunnelSurrounding
tunnelEmplacement
panel ①
Emplacement
panel ②
Emplacement
panel ③Timeline

Excavation

Transport・Ventilation ・Drainage
Construction Operation
Construction

Operation

Construction Operation
Construction

Emplacement
panel ④

Operation
Construction Operation

Emplacement
panel ⑤Construction Operation

Emplacement
panel ⑥

Figure 4.5-22 Sequence of construction and operation of panels (HLW 6 TT panels)

After construction of the first disposal panel, construction and operation are carried out inparallel, requiring appropriate zoning. Independent routes are provided for intake of
construction materials and removal of excavated spoil as shown in Figure 4.5-21, separating
panel (2) access and connecting tunnels from those used to dispose of waste in panel (1).
4-105

There are also separate ventilation circuits and drainage routes, with any required bulkheads
or tunnel plugs to prevent interaction between the 2 zones (see Supporting Report 4-60 for
details).
Within the operating panel, material flow routes are established to allow waste
emplacement and backfilling/sealing of filled tunnels to run in parallel, which may be
important for the more complex emplacement/backfilling operations for the H12V case. As
previously discussed, relevant connecting tunnels are dimensioned to handle 2-way traffic(see Table 4.5-6) and parallel connecting tunnels, with bulkhead doors to allow changing
connections, can be planned to facilitate re-zoning.
As shown in Figure 4.5-17 (a), the TT panel allows access of large equipment only in predetermined directions. Nevertheless, the complicated process of emplacing the H12V
overpack and associated buffer and then backfilling can be efficiently managed if these utilise
access to the disposal tunnel from different directions (backfill shown as brown and overpacks
in filled circles in the upper (operational) panel of Figure 4.5-21).
Ventilation and drainage routes are established to limit impacts of possible operational
perturbations, for example ensuring that the exhaust from the disposal tunnel in which HLW
is being emplaced does not flow into tunnels being backfilled. The air intake follows the
backfill transport, while exhausts follow the route of the waste emplacement equipment.
Although manual backfilling is currently assumed, remote handling of this operation could
both reduce ventilation constraints and simplify control of worker access to areas containing
waste.
Tunnel slopes are set so that drainage from the panel under construction and that underoperation could be collected independently in sumps below separate “active” and “inactive”
drainage shafts. As a result of this layout, there are sections where connection tunnels run inparallel, so an appropriate pitch is set to ensure mechanical stability.
Based on the above study, the required number and roles of access routes can be defined, as
shown in Table 4.5-19.
Table 4.5-19 Access roles (H12V TT panels)
No.

Name

FunctionRole of ventilation

Ramp

Operational ramp

Overpack and buffer transportation

Air intake

Shaft No. 1

Spoil removal shaft 1

Spoil removal, drainage (inactive)

Exhaust (inactive)

Shaft No. 2

Spoil removal shaft 2

Spoil removal, drainage (inactive)

Exhaust (inactive)

Shaft No. 3

Ventilation shaft

Drainage (active)

Exhaust (active)

Materials andEquipment Shaft 1
Materials andEquipment Shaft 2
Materials andEquipment Shaft 3

Construction materials, personnel,
equipment
Construction materials, personnel,
equipment

Shaft No. 4
Shaft No. 5
Shaft No. 6

Backfill materials, personnel, equipment

Air intake
Air intake
Air intake

The ramp provides a transport route for the overpack and buffer and, together with shaft 6(backfill support shaft), also acts as an air intake for radiation-controlled zones. Active
exhaust and drainage are provided by shaft 3. Shafts 4 and 5 support construction operations
and provide air intake for construction zones, with inactive exhaust and drainage provided byspoil removal shafts 1 and 2. It should be noted here however, that designs will need to beadapted to site conditions and therefore further studies would optimise use of access routes.
4-106

In terms of emergency evacuation routes, air exhaust shafts (No. 1 to 3) would not be usedas these may be smoke-filled in the event of a fire. Hence, independent routes would be
provided to shafts 4 to 6, potentially with the option of escape via bulkhead doors between
shaft 4/5 and 6, which would otherwise be kept closed to isolate these zones. In case of
accidents when evacuation routes cannot be accessed, emergency shelters will be set up, but
number, locations and design of these will be considered only in the future.

(b) HLW DET panels
A conceptual sketch of tunnel layout for HLW DET panels to assess access requirements isshown in Figure 4.5-23. Note that, if orientation of disposal tunnels perpendicular to thehydraulic gradient is not practical, all tunnels should be oriented so that dead ends are downgradient to avoid risks for preferential RN transport along the EDZ.

Figure 4.5-23 Conceptual sketch of tunnel layout (HLW DET panels)

It should be emphasised that all considerations of independent ventilation, drainage andescape routes for active and inactive zones are not included and hence parallel connecting
tunnels with crossover points closed by bulkhead doors would be incorporated in the final
design.
The associated sequence for construction and operation of DET panels is given in Figure
4.5-24.

4-107

Figure 4.5-24 Sequence of construction and operation of panels(HLW 4 DET option – only conceptual at this stage)

For the PEM option, when all emplacement operations are carried out by tele-operated or
automated equipment, backfill transported against the flow of active exhaust air and drainage
would not be problematic (although tunnel sealing may also need to be remote handled tocompletely remove this issue). However, this would require a more complex connecting
tunnel layout to provide worker access for H12V manual backfilling, which would
presumably require to be via shaft 4. Emergency escape would also require connections
between the active and inactive zones and, in case of a fire in the construction panel, a second
escape route would inevitably pass through a radiation-controlled zone. Such considerations
will form the basis of more detailed assessment of this option.
As for the TT case, tunnel slopes are set so that drainage of the construction and operational
panel is independently routed to the inactive and active drainage shafts, respectively. For thesections where connection tunnels run in parallel, an appropriate pitch was set to ensure
mechanical stability.
Based on this concept, the roles of access routes are shown in Table 4.5-20, based on 1
ramp which provides a transport route for the overpack and buffer/PEM and also acts as an air
intake, and 5 shafts for ventilation, drainage, material transfer, and access by workers. In thisoption, all ventilation exhaust and drainage from DETs is into a single connecting tunnel soflows are somewhat simpler but, during rezoning, either the functions of shafts 2/3 and 4/5
need to be exchanged or (probably much simpler) alternative connecting tunnel routes are
required for the A and B operational sides.
Table 4.5-20 Access roles (H12V/PEM DET panels)
No.

Name

Function

Ventilation role

Ramp

Ramp

Transport of overpack and bufferor PEM

Air intake

Shaft No. 1

Ramp ventilation shaft

Ramp drainage

Exhaust (active)

Shaft No. 2

Materials transport shaft A

Construction, backfill materials,
personnel and equipment

Air intake

Shaft No. 3

Spoil removal shaft A

Excavated spoil, drainage

Exhaust (active)

Shaft No. 4

Materials transport shaft B

Construction, backfill materials,
personnel and equipment

Air intake

Shaft No. 5

Spoil removal shaft B

Excavated spoil, drainage

Exhaust (inactive)

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As the background provided by H12 for TT is not available for the DET option, theassessment is currently at a much simpler conceptual level, which will be refined in the futureon the basis of issues identified.
(c) TRU waste
TRU waste disposal operations occur within a single panel (see Figure 4.5-19) after all
excavation and construction has been completed and hence there are no zoning issues here
(N.B. that this assumes the inventory is not modified after initial construction). Waste
packages are transported underground by the dedicated ramp, while infilling/backfillingmaterial is transported via shafts.
As described in Section 4.4.3 (2), waste packages are transported from the surface to thedisposal vaults directly, without any trans-shipment between ramp and undergroundtransporters. When emplacing with a remote-controlled forklift, packages are carried directlyfrom the transporter to the emplacement cell using established technology. When emplacingwith a gantry crane (see Figure 4.4-30), waste packages are lifted directly from the transport
vehicle to their disposal location within the operational disposal cell by remote control. It iscurrently assumed that waste transportation and backfilling/sealing vaults will be manual,
hence defining ventilation and emergency escape requirements, but remote handling here may
be an option in the future.
Drainage water will flow down the gradient of the tunnel floor in the TT and will becollected at a sump located at the lowest point and pumped up the associated shaft.
(iii) Design of trans-shipment chamber
The access ramp reference design has a square-spiral structure, allowing a relatively gentle
slope (for road transport) without an excessive footprint, but other designs (or rail transport)
may be considered in the future. Straight sections are 1,000 m long and each 90° corner has aradius of curvature of 30 m. In addition, the slope was set to 10%, based on precedents inother countries [117] [118].
For H12V, overpack and buffer are transferred at the bottom of the ramp to rail transport
vehicles for emplacement, as shown in Figure 4.5-25.

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Figure 4.5-25 Layout at bottom of trans-shipment chamber (H12V)

At this location, a shielded temporary store for overpacks and buffer serves to smooth
operational logistics. Transhipment and other movements are carried out using a remotecontrolled overhead crane and thus the height of the tunnel here is higher than that of either
the connecting tunnel or access ramp.
For the PEM case, at the bottom of the ramp the road transporter can directly transfer the
PEM to the rail-based transport and emplacement machine, as shown in Figure 4.5-26.

Control door

PEM
emplacement
systemPEM
Transportation vehicle
Control door with radiation shielding

Figure 4.5-26 Layout at bottom of trans-shipment chamber (PEM)

Further logistical analyses will be carried out in the future to assess whether temporary
stores for PEMs will be needed (e.g. when decisions are made on where the PEM (and buffer
components) will be manufactured and when site conditions/restrictions are better known).
This assumes that the emplacement machine can lift the PEM directly from the bed of thetransport vehicle with a gripping device having a twist lock mechanism. As a result, an
overhead crane is not required and it is possible to lower the height of this tunnel compared to
the H12 option.

4-110

(5) Setting up reserve areas
In the design of the repository layout, panels capable of disposing of a predeterminedinventory of waste are arranged in the preferred rock areas. The latter are constrained by
LDFs, which excludes faults and fractures greater than 1 km, but includes smaller features –which could impact the practicality of using particular parts of disposal tunnels/disposal holesthat were initially dimensioned to the reference inventory (Section 4.5.4 (2) (iii)). Here,
therefore, the required reserve disposal area to cover uncertainties in the distribution ofsmaller features is evaluated.
(i) Evaluation items
To define reserve requirements, the impact of smaller faults and fractures (or other highly
permeable features, such as sand layers in sediments) on acceptability of waste emplacement
locations has to be assessed in order to determine their impact on the panel requirement in
Section 4.5.4 (2) for sufficient disposal volume. Thus, Table 4.5-21 sets the evaluation item ofassuring sufficient disposal reserves.
Table 4.5-21 Reserve area evaluation items
Design requirement

Detailed description of requirements for concepts assessed

Assure sufficientdisposal reserves

Consider the effects of permeable features on utilisation of disposal panels to definereserve areas that will assure that the reference inventory can be disposed of(ii) Concept of design of reserve areas
As an example, potential features that could impact the usability of disposal area within apanel for the H12V case in plutonic rock are shown in Figure 4.5-26. To define Emplacement
Determining Features (EDFs), such as faults and fractures, the following characteristics
impacting practicality are considered:

If there is a large volume of water inflow, the effort required for drainage management
will increase.

When there is a locally large water inflow to disposal holes or tunnels from faults andfractures, it will be difficult to construct buffer or backfill to required quality levels.



Local water flow can cause piping erosion of buffer/backfill with reduction of its
performance.



If mechanical strength is reduced locally, there is a risk of borehole/tunnel collapse.

Figure 4.5-27 very simplistically represents some of the issues involved.

4-111

(1) Fracture with no water inflow, (2) high permeability fault or fracture, (3) low permeability
fault or fracture, (4) areas where faults and fractures decrease mechanical strength
Figure 4.5-27 Sketch of EDFs limiting disposal panel utilisation in plutonic rock(Example of H12V)

Relatively small, low permeability features (such as (1) in the figure) may not prevent waste
emplacement, but raise issues requiring consideration at the time (e.g. small breakouts fromthe borehole wall). In the case of more permeable features, these may have sufficientgroundwater inflow that emplacement is clearly precluded (e.g. (2)). Otherwise, flow may be
low enough to be potentially managed by special precautions during buffer emplacement or to
avoid subsequent erosion (e.g. (3)) – although such a hole might need to be abandoned if thiswas not practical. Similarly, larger weak zones (e.g. (4)) may also require a section of tunnel
to be excluded or a hole to be abandoned. In some cases, rock improvement techniques (e.g.
grouting) may reduce impacts of weak or water-bearing features, but the potential impacts on
post-closure performance need to be carefully assessed in such cases. For smaller problem
areas (especially the sensitive zone at the top of holes), additional stabilisation such as thatshown in Figure 4.5-3 could be considered although, again, impacts on post-closure
performance may need to be considered.
From such an assessment, features that are of potential concern can be assessed todetermine if they are actually EDFs and would preclude quality-assured borehole drilling (e.g.
low strength) or emplacement (e.g. based on water inflow rate and its impact on practicality of
buffer/backfill emplacement).
For the PEM, EDFs are less restrictive and relate to practicality of disposal tunnel
excavation and the emplacement of backfill (see Figure 4.4-29). For TRU waste, a number ofsteps involve emplacing buffer and/or backfill which would be sensitive to local water inflow,
but installation of liner and grouting surrounding rock can be considered without the postclosure concerns about high pH that exist for HLW. Thus, EDFs again relate to practicality ofdisposal vault excavation and the emplacement of buffer/backfill.
(iii) EDF characteristics
In the study of reserve areas, EDF characteristics are set as described below.
(a) EDFs for disposal holes
For H12V, buffer and backfill emplacement are coupled in terms of assuring quality andboth are very sensitive to early swelling/erosion by groundwater inflow (Section 4.4.3 (1) (i))
until final completion of the mechanical plug. Since the buffer is a key safety barrier and theextent to which its performance is expected to be degraded is proportional to the water inflow
4-112

rate into the borehole, this rate is selected as the criterion for characterising LDFs. Placing thedeposition hole in a more permeable zone would also imply severe problems with buffer
swelling prior to backfilling in the absence of a borehole cap.
In terms of practicality, when emplacing buffer blocks it is necessary to prevent the inflow
of groundwater to the disposal hole, which may require some form of in-hole water
management system. Therefore, the flux of water that can be managed by such a system sets a
constraint on EDFs, which is compared to a value set by the extent of possible piping erosion,
with the lower of the two values used as the criterion for defining an acceptable inflow rate.
Some initial assessments are made of such management techniques as described below, but
further development will be needed to reduce associated uncertainties.
It is assumed that piping erosion commences after removal of the sheet used to prevent
buffer wetting during emplacement, as outlined in Section 4.4.3 (1) (i), and will continue until
tunnel backfilling is complete and the mechanical plug is in place (see Supporting Report 455). Figure 4.4-10 shows the required buffer density to meet specific design requirements. If it
is assumed that piping erosion occurs only during re-saturation (see Supporting Report 4-56),
Figure 4.4-10 shows that the lower limit of density would be that to meet requirement (5)(colloidal filtration) at an effective dry density of clay of > ≈ 0.8 Mg/m3, which would allow a
total loss of 0.45 t per hole. Based on laboratory tests [119] [120], an empirical formula
relates the total groundwater flow to total buffer loss, setting a flow limit of between 3.4 × 102
and 8.7 × 104 m3. From the maximum time delay between emplacing buffer and completion of
backfilling, the minimum flow defining an EDF is 0.8 l/min, 0.8 l/min and 0.5 l/min inplutonic rocks, Neogene and Pre-Neogene sediments, respectively.
(b) EDF for disposal tunnels
For disposal tunnels, it is assumed that the impact of EDFs on their utilisation is determined
mainly by water inflow impacts on excavation of the tunnel, construction of buffer (for TRU
waste Grs. 1, 2 and 4H) and backfilling. In some cases, mechanical impacts could also be
significant, but these are not considered at present. Figure 4.5-28 indicates issues associatedwith EDF specification for the example of implementing PEM DET tunnels.
Before excavating a disposal tunnel, it is assumed possible to obtain information on the riskof encountering a problematic feature by drilling a hole in advance. If a problem area is
encountered, countermeasures such as grouting or modifying the excavation/support
procedures can be implemented. However, in the design stage, it is not possible to estimate
where in a panel such features would be found, so no quantitative assessment of the required
reserve area is set at present.
When the strength of the rock is locally low, it is possible to improve the mechanicalstability of the tunnel by measures such as pre-grouting, reinforcement of support systems andimplementation of alternative excavation methods. However, excavation becomes difficult
when water inflow at the face exceeds 300 to 500 l/min, based on past experience [121].
Therefore, to define EDF water inflow that prevents excavation, a limiting value of 300 l/min
is selected.

4-113

【Construction stage 1 】
Survey of emplacement panel by pre-boringfrom the connecting tunnel(Assumption)
 Discovery of features not previously
identified from characterisation work Preliminary borehole identifies theexistence of a feature with low strength
and/or high water inflow【Construction stage 2 】
Design change based on survey results anddrilling of tunnels (Assumption)
 To limit the drainage volume/ensure
worker safety, change the design of thepanel to avoid excavation through the
EDF【Operation stage 3】
Disposal of waste, backfilling andconstruction of mechanical plug at end(Assumption)
 Avoid placing waste in problematic
tunnel sections

Figure 4.5-28 EDF impacts on disposal tunnels (PEM example)

In disposal tunnels, backfilling will occur after waste emplacement. In addition, buffer will
be installed in-situ for TRU waste Grs.1, 2 and 4H. Therefore, the impact of water on theconstruction of buffer and backfill has to be considered. For the TRU vaults, thicker lining
will reduce water inflow, which can be complemented by localised grouting or drainage
behind the liner [93] [122] [123]. Emplacement techniques for buffer and backfill to required
quality levels will need to be developed to determine water inflow limits, hence no EDF
constraint is set at present.
(iv) Results of reserve area assessment(a) H12V
The definition of a relevant EDF is set by the rate of local water inflow, which can then be
used to determine the unusable fraction of a disposal panel.

4-114

From Section 3.3.3 (4) (iii), stochastic three-dimensional hydrogeological models at panel
scale allow inflow into disposal holes and tunnels to be determined. The hydrogeologicalmodel used represents three disposal tunnels and the surrounding rock mass, as shown in
Figure 4.5-29 for the case of plutonic rocks.

(a) Model of disposal tunnels(b) Hydrogeological DFN model
Figure 4.5-29 3D model to evaluate disposal hole groundwater inflow (plutonic rock)

The hydrogeology is represented by a discrete fracture network (DFN) with permeability
coefficients of fractures set probabilistically based on an assumed lognormal distribution (seeSupporting Report 4-57 for details). To assess inherent variability, 100 realisations of themodel were run. The determination of reserve areas was based on the distribution of inflow
into the middle disposal tunnel.
The cumulative probability distribution of the groundwater inflow into boreholes for eachof the 3 representative rocks is shown in Figure 4.5-30. It can be seen that these distributions
are quite wide in the cases of plutonic rocks and Pre-Neogene sediments, but narrower for
Neogene sediments, reflecting differences in the statistical distribution of fracture
permeability coefficients. Assuming no countermeasures to reduce water inflow, these
distributions can be compared to the EDF inflow limits set for each rock in Section 4.5.4 (5)(iii) (a) above.

4-115

(b) Pre-Neogene sediments(a) Plutonic rock(c) Neogene sediments

Figure 4.5-30 Cumulative probability distribution of water inflow (H12V holes)

Table 4.5-22 lists the disposal hole utilisation for each rock, ranging from 100% for
Neogene sediments to 80 – 90% for the other rocks.
Table 4.5-22 Disposal hole utilisation (H12V)
Rock type

Disposal hole utilisation rate (%)

Plutonic

79

Neogene sediments

100

Pre-Neogene sediments

90

These degrees of utilisation would evidently depend on the inflow limits and otherrequirements considered. Further studies may be needed to assess whether current limits areappropriate. Based on these numbers, reserve areas for about ≈ 21% and ≈ 10% of theinventory are assumed to be required for plutonic and Pre-Neogene cases, respectively (for a
total inventory of 40,000 HLW units, equivalent to a further 8,400 and 4,000 disposal holes,
respectively). Further, as the footprint of 1 unit in plutonic rock is 44.4 m2 (borehole pitch
4.44 m × disposal pitch 10 m) and in Pre-Neogene sediments is 71.0 m2 (borehole pitch 4.44
m × disposal pitch 16 m), for the given hole utilisation rates these reserves are equivalent to ≈470,000 m2 (plutonic rocks) and ≈ 320,000 m2 (Pre-Neogene sediments). These reserves areincorporated into the repository layouts shown in Section 4.5.4 (6).
(b) PEM and TRU waste
For the PEM and TRU waste, fundamentally the same analysis was carried out as describedfor H12V above (details in Supporting Reports 4-58 and 4-59, respectively), with the model
output given in terms of water inflow per metre of tunnel length. Results showed that, for
4-116

Neogene and Pre-Neogene sediments, no case exceeded the EDF cut-off value (300 l/min), set
in Section 4.5.4 (5) (iii) (b) above. However, for plutonic rocks, the EDF was exceeded in
0.3% (PEM) and 0.5% (TRU waste) of the tunnel length. For such a small fraction of thetunnel, it is considered practical to apply countermeasures so that, in all cases, it can beassumed that utilisation is effectively 100%.
(v) Expansion to include reserve areas for H12V
For H12V disposal in TT panels, existing panels cannot be expanded as they are surrounded
by a connecting tunnel, so a new panel is required. This could, for example, be constructed by
DETs excavated off a connecting tunnel, as illustrated in Figure 4.5-31.

Figure 4.5-31 Expansion to include an additional panel

However, in order to use the same emplacement equipment, an appropriate junction angle
between the connecting tunnel and the disposal tunnel is required.
(6) Repository Layout
The specifications of the repository layout are based on the considerations above for eachrock type. In addition, requirements from Table 4.1-1 in terms of economics with respect tocost-effectiveness of construction, operation and closure are considered for different disposal
concepts and disposal panel options, using the difference of the volume of excavated rock as asurrogate indicator of cost (which also correlates with environmental impact). As described in
Section 4.5.1 (1), the basis for illustrating and explaining the layout of the undergroundfacilities assumes 50 years interim storage of HLW, although the impact of emplacing waste
after a storage period of 30 years is also discussed.

4-117

(i) Plutonic rocks
For co-location of HLW and TRU waste disposal, layouts of the repository for H12V andPEM variants in plutonic rocks are shown in Figure 4.5-32.
+Y+Y

Reserved area
Panel no. (white)
Reserve no. (red)

0

-Y

500m

⑥

①

⑤

③

②

①

0

HLW repository

-X

+X

-Y

500m

⑤

②

①

③

⑥
④

Hydraulic gradient

④ ②

TRU
waste repository

Hydraulic gradient

TRU waste repository

Panel no. (white)

HLW
repository-X+X
Travel time in unit length

Travel time in unit length(a) H12V/ TT panels(b) PEM/DET panels

Figure 4.5-32 Layout of underground facilities (plutonic rocks, HLW interim storage 50 years)

From the analysis in Section 4.5.4 (1) (ii), preferable areas in the SDM are defined (seeFigure 4.5-15 (a)), in which the inventory of 40,000 waste canisters can be emplaced in 6
disposal panels as shown, together with a panel for TRU waste and required
access/connecting tunnels. The panel numbers relate to the assessment of work flow,
ventilation and drainage, as considered in Section 4.5.4 (4) and described in more detail in
Supporting Report 4-60.
As noted in Section 4.5.1 (2), stress anisotropy is not an issue for plutonic rocks and hence
panels can be laid out with disposal tunnels perpendicular to the hydraulic gradient. In
addition, the TRU waste panel is located 500 m to 1,000 m downstream of the HLW. Basedon the estimated disposal hole utilisation for H12V given in Table 4.5-22, the required reservearea is illustratively divided into two panels, which are placed adjacent to disposal panels 2
and 4 as shown in Figure 4.5-21 (a). In the cases of PEM and TRU waste, there is no need fora reserved area since the EDFs impacting buffer performance are so few that they could bemanaged by measures such as grouting. The panel numbers relate to the assessment of work
flow, ventilation and drainage, as considered in Section 4.5.4 (4) and described in more detailin Supporting Report 4-60.
For the case of HLW stored for only 30 years, disposal tunnel pitch would be increased asshown in Table 4.5-7. In terms of preferred areas shown in Figure 4.5-15 (a), sufficient
flexibility is available to accommodate the expanded footprint, especially if DET panels areused to make best use of better rock (details of this assessment shown in Supporting Report 452).

4-118

(ii) Neogene sediments
For co-location of HLW and TRU waste disposal, layouts of the repository for H12V andPEM variants in Neogene sediments are shown in Figure 4.5-33.
Hydraulic gradient+Y

Panel No. (white)①
HLW
repository②③④

TRU waste
repository

⑤
0

500m

⑥

-Y
-X

+X
Travel time in unit length(a) H12V/TT panels(b) PEM/DET panels(a) H12V/ TT panels(b) PEM/DET panels

Figure 4.5-33 Layout of repository (Neogene sediments, HLW interim storage 50 years)

From the analysis in Section 4.5.4 (1) (ii), preferable areas in the SDM are defined (seeFigure 4.5-15 (b)), in which the inventory of 40,000 waste canisters can be emplaced in 6
disposal panels as shown, together with a panel for TRU waste and required
access/connecting tunnels. The panel numbers relate to the assessment of work flow,
ventilation and drainage, as considered in Section 4.5.4 (4) and described in more detail in
Supporting Report 4-61.
As described in Section 4.5.1 (2), disposal tunnels were arranged to maximise stability withrespect to the direction of maximum principal stress (see Supporting Report 4-54). As forplutonic rocks, the TRU waste disposal panel was arranged to be downstream or out of flow
towards HLW panels. As noted in Table 4.5-22, Neogene sediments do not require reserveareas.
For the case of HLW stored for only 30 years, the disposal tunnel pitch would not need to
be increased because of the large pitch needed to ensure mechanical stability and the lower
impact of waste thermal output due to lower rock ambient temperature, as shown in Table 4.57.
(iii) Pre-Neogene sediments
The repository layout for Pre-Neogene sediments is shown in Figure 4.5-34. Since this rockhas a folded structure and many LDFs, DET panels were considered for both H12V and PEM
4-119

options in order to increase flexibility to tailor layout to preferred rock areas. In addition,
since these rocks have high strength and stress anisotropy is relatively small, the disposal
tunnels can be set perpendicular to the groundwater flow direction. TRU waste disposal isagain downstream from the HLW panels.
+Y

+Y

②
⑦
③

②

①

①

HLW
repository②

TRU waste
repository

TRU waste
repository

0

⑦
⑤
③
①

⑧
⑥
④

Hydraulic gradient④

Hydraulic gradient⑤
HLW
repository

⑧
⑥

0

500m

500m
Panel no. (white)

Panel no. (white)

-Y

-Y
-X

-X

+X
Travel time in unit length+X
Travel time in unit length(a) H12V/ DET panels(b) PEM/DET panels
Figure 4.5-34 Layout of repository (Pre-Neogene sediments, HLW interim storage 50 years)

Based on the estimated disposal hole utilisation for H12V given in Table 4.5-22, therequired reserve area is illustratively divided into two panels placed adjacent to disposal
panels 1 and 2, as shown in Figure 4.5-34 (a). For the PEM and TRU cases, no reserve areas
are required.
For the case of HLW stored for only 30 years, disposal tunnel pitch would be increased asshown in Table 4.5-7. In terms of preferred areas shown in Figure 4.5-15 (c), sufficient
flexibility is available to accommodate the expanded footprint, especially if DET panels areused to make best use of better rock (details of this assessment are given in Supporting Report4-52).
(iv) Comparison of excavated rock volumes and footprint
For the repository layouts described above for each rock type, the excavated rock volume
was estimated from the length and cross-sectional area of all tunnels as shown in Table 4.5-23(more details given in Supporting Report 4-62).
Comparison of H12V with the PEM clearly shows much larger excavated volumes (almosta factor of 2) for the former due to longer tunnels and the additional spoil from hole drilling.
In terms of host rock, the Neogene cases tend to have significantly larger excavated rockvolumes: this is due to factors such as the greater excavated volume to allow thicker tunnel
support and, for H12V, the greater length of disposal tunnels due to greater pitch between
disposal holes (see Section 4.5.2 (2)), despite lower volumes from access ramp and shafts dueto the lesser depth of the repository. Comparing H12V for plutonic rock and Pre-Neogene
sediments shows that the amount of excavated rock is greater for TT compared to DET panels.

4-120

Table 4.5-23 Excavated rock volumes for HLW disposal options
Panel type

Tunnel:DET type

H12V
Length (km)Volume (x 104 m3)
DisposalTunnelDisposal hole
Surrounding
TunnelConnecting
TunnelAccessramp
Accessshaft

PEM

Plutonic

Neogene

PreNeogene

Plutonic

Neogene

PreNeogene

Length

194

285

181

138

138

137

Volume

444

613

425

258

311

255

Length

166

166

166

-

-

-

Volume

64

64

64

-

-

-

Length

13

20

-

-

-

-

Volume

33

51

-

-

-

-

Length

13.6

16.7

6.4

4.9

5.8

5.4

Volume

78

123

36

28

43

31

Length

10.7

5.4

10.6

10.7

5.4

10.6

Volume

61

39

60

61

40

60

Length

6.0

3.0

5.0

5.0

2.5

5.0

Volume

25

13

21

21

10

21

Length

403

497

370

159

152

158

Volume

705

903

606

368

404

367

Total

The footprint is determined for the HLW and TRU repositories as shown in Table 4.5-24.
The footprint for HLW repository is in the range 3 to 12 km2 and is largest in the case of
H12V in Neogene sediments (disposal tunnel pitch is the largest of the three representative
host rocks). The area for the PEM is smaller than that for H12V in the same host rock. The
area for HLW repository is 11 to 28-fold that of the TRU repository.
Table 4.5-24 The footprint of the repository
Conditions

HLW

Area
ratio

TRU waste

Disposal
option(HLW)

Disposal
area
(HLW)

Length
(km)

Width
(km)

Area
(km2)

Length
(km)

Width
(km)

Area
(km2)

HLW
/TRU

H12V

Panel type

2.29

3.01

6.89

0.46

0.55

0.25

28

PEM

DET type

1.52

2.17

3.30

0.46

0.55

0.25

13

Neogenesediments

H12V

Panel type

4.97

2.37

11.78

0.73

0.58

0.42

28

PEM

DET type

3.19

1.50

4.79

0.73

0.58

0.42

11

PreNeogene
sediments

H12V

DET type

2.88

2.09

6.02

0.46

0.55

0.25

24

PEM

DET type

2.40

2.05

4.92

0.46

0.55

0.25

20

Rock type

Plutonic

4-121

4.5.5 Design of underground services
The design of the underground services is, to a large extent, very likely rely on existing
mining and tunnelling practice. However, some aspects and requirements will be unique to arepository and these are now assessed.
(1) Design of ventilation and cooling equipment
Ventilation and air conditioning/cooling equipment is required for both the construction andoperation of the repository. For the repository layouts developed in Section 4.5.4 (6) andassociated work-flows, the ventilation and cooling requirements can be specified. For HLW,
construction and operation will run in parallel and independent systems are required for these
two areas, with the latter established to be compatible with a radiation-controlled zone. In thissection, focus is on HLW as the repository is larger and independent ventilation for radiationcontrolled and non-controlled zones is required, as discussed further in Supporting Report 463 (which also considers TRU waste).
(i) Design requirements
The design requirements of ventilation and air conditioning equipment are outlined in Table
4.5-25. Conservatively assuming that workers are present in both construction and operationalzones during normal operation, ventilation and cooling are required for general occupationalsafety and maintaining a suitable working environment (see Table 4.2-3).
This is broken down in order to further quantify requirements; covering maintaining oxygen
concentration in working areas, dilution of combustible/hazardous gas or dust, temperature
control in working areas and ventilation rate limits.
Table 4.5-25 Design requirements for ventilation and cooling equipment
Designrequirements
Maintenance ofoxygenconcentration inwork areas

Detailed description of requirements for concepts
assessedSpecifications

The necessary amount of air can be supplied tomaintain the oxygen concentration in the work area

Ventilation and coolingsystem, airlocks

Dilution of
flammable gas

Required dilution of combustible gas (methane, etc.)
in tunnels (depends on the geology) and safe
discharge to the surface (relevant also for TRU
waste)

Ventilation and coolingsystem, airlocks

Dilution of harmful
gas and dust

Required dilution of harmful gas (e.g. gas after
blasting, exhaust from machinery) and dust generated
during excavation and safe discharge to the surface

Ventilation fans,airlocks, dust collectors

Maintain working
area temperature
Tunnel air flow rate
limitation
Containment of RNs
in case of
accidents

Required cooling of heat from rock and construction
equipment, so that temperature and humidity of work
areas are suitable
Ventilation in tunnels does not exceed the defined air
flow rate standard
For any accident involving leakage of RNs
underground, preventing release outside the facility

4-122

Cooling system
Ventilation fans,airlocks
Ventilation fans,airlocks, emergency
ventilationequipment/control
systemRelevant rules and guidelines, together with reference values for ventilation and working
temperatures, are given in Table 4.5-26, noting the special case for environments where
flammable gases are present. Although the Mine Safety Act is not applicable to geologicaldisposal, the air speed limit specified therein is considered to be a reasonable design
requirement.
Table 4.5-26 Regulations/guidelines for ventilation and cooling equipment
Compliance standards

Index

Reference value

Occupational Safety and Healthregulations the 611 Article

Temperature
Temperature

Ventilation technology guidelines
for underground construction

Air speed
Workerrespiration rate

Mine Safety Act enforcement
regulations

Air speed

Dry bulb temperature≤ 37 °C
Wet bulb temperature≤ 28 °C
0.3 m/s or more(≥ 0.5 m/s) *Summary
Law

Target value basedon guidelines

0.05 m3/s/person≤ 7.5 m/s

Although notcovered by Mine
Safety Act, set as areference value* When flammable gas is present

Depending on the geological environment, release of combustible gas (such as methane)
during tunnel excavation may or may not be an issue. In the former case, it is necessary totake measures, such as routinely monitoring gas concentration and assuring sufficientventilation to avoid any potential problems [124]. In addition, in such cases, it may benecessary to take additional precautions such as prohibiting the use of internal combustion
engines and making electrical equipment explosion proof. Also other gases, such as Rn, could
be of concern and the ventilation needs to be sufficient to keep concentration below maximum
allowed levels.
In the design of ventilation equipment, the required ventilation volume [125] is calculated
according to the work situation of each tunnel, with ventilation network analysis performed
for routes set by consideration of the repository layout. The capacity of the required surface
ventilation fans and the arrangement of baffles and regulators to adjust air flow are then set,so that the required ventilation volume can be provided where needed.
A cooling system is installed when the temperature of the work area cannot be maintained
simply by air flow. As a standard for the work areas, the dry bulb temperature shall be ≤ 37 °C
and the wet bulb temperature ≤ 28 °C [125]. This covers ventilation equipment design for
normal conditions, but if an accident with release of radioactivity occurs, it is also necessary
to assure containment within the facility. Therefore, an operational radiological safety
ventilation requirement is to prevent releases of RNs (see Table 4.2-1), as described in
Section 4.5.6. Section 4.5.6 also discusses ventilation needs if fire occurs, even if noradioactivity is likely to be released.
(ii) Specifications of ventilation and cooling equipment
The calculated required ventilation rates are given in Table 4.5-27 (details in Supporting
Report 4-63). Of the requirements in Table 4.5-25, dilution of harmful gases and dust during
construction leads to the highest ventilation rate. Because H12V requires longer tunnels and4-123

the additional drilling of boreholes, higher ventilation rates are required compared to the PEM
case.
Table 4.5-27 Required ventilation for HLW options
Disposal concept

H12V

PEM

Rock type

Plutonic

Neogene

Pre-Neogene

Plutonic

Neogene

Pre-Neogene

Panel type

TT

TT

DET

DET

DET

DET

Total required
ventilation (m3/s)

87

65

94

21

22

21

In setting the ventilation route, it is assumed that the main ventilation fans are installed at
the surface on exhaust shafts (see Section 4.5.4 (4)), which drives the circulation, as is normal
practice in mining [125]. Assuming manually-controlled excavation, ventilation of the rockface is particularly critical and a system based on air suction can be implemented as shown in
Figure 4.5-35. Here, a booster fan is installed at the tunnel entrance and fresh air supplied to
the face through a duct. An air suction duct provides an air curtain that restricts dust
movement into the tunnel and allows it to be captured in a dust collector.
For the construction zone, the ventilation circuit analysis allows the main components
required to be specified (wellhead fans, local fans, stoppings, etc.) as shown for the exampleof plutonic rocks in Tables 4.5-28 (H12V) and 4.5-29 (PEM). It should be emphasised thatsuch rock-specific analysis is based on more detailed air flow routes than the simple
conceptual representations presented in Figures 4.5-21 and 4.5-23.

Figure 4.5-35 Ventilation during disposal tunnel excavation (air-suction option) [125]Table 4.5-28 Ventilation equipment for construction zone (plutonic rock - H12V)
Equipment

Specification

Quantity

Summary
Exhaust for shafts 1 and 2 × 2
units on each3

Wellhead fan

Rated air volume 33 m /s
Rated total pressure 4.9 kPa

4

Booster fan

Rated air volume 17 m3/s
Rated total pressure 2.9 kPa

6Dust collector

Rated air volume 20 m3/s

6

Airlock

W (aperture) = 5 m

42

For disposal tunnels

Airlock

W=8m

26

For connecting tunnels

Soft ducting

φ 0.9 m4,000 m4-124

Table 4.5-29 Ventilation equipment for construction zone (plutonic rock - PEM)
Equipment

Specification

Quantity

Summary
Exhaust for shaft 1 × 1 units

3

Wellhead fan

Rated air volume 33 m /s
Rated total pressure 4.9 kPa

1

Booster fan

Rated air volume 17 m3/s
Rated total pressure 2.9 kPa

6Dust collector

Rated air volume 20 m3/s

6

Airlock

W=8m

37Soft ducting

φ 0.9 m

3,300 mFor disposal tunnels

To assure the required air flow (Table 4.5-27) for H12V, 4 wellhead fans with a capacity of
33 m3/s are required whereas, for the PEM a single 33 m3/s unit is required. Parallel
excavation of 6 disposal tunnels requires this number of booster fans and dust collection units
for each tunnel. The larger number of tunnels for H12V results in a complex ventilation path,
requiring a larger number of air locks. In practice, the ventilation analysis also needs toconsider how re-zoning after panel construction is organised, so that the reorganised flow thenmeets the radiation control requirements.
In terms of air-conditioning equipment, this would be installed only in locations where thewet bulb temperature exceeds the standard. For each layout, thermal analysis was carried out
based on rock ambient temperatures and heat radiation from tunnel walls, excavated rock,inflowing water, machinery and the heat of hydration of concrete (details in Supporting
Report 4-63).
Because of the 15 °C lower rock temperature at 500 m disposal depth, for Neogene
sediments cooling equipment is not required. The repository layout for Pre-Neogene
sediments required a high cooling capacity compared to plutonic rocks, due to the longer
distance from the air intake shaft to the disposal tunnels. In addition, because the total lengthof tunnels is greater, more cooling capacity is required for H12V compared to the PEM case.
In terms of the cooling method, supply of cooling water from the surface to an undergroundheat exchanger that cools the air is assumed, a method that is well proven in mines [126]. The
actual equipment needed can be specified based on the requirements in Table 4.5-30. Notethat this assessment focuses on the excavation areas and hence the impacts of heat-emitting
waste is not considered.
Table 4.5-30 Cooling equipment requirements
Disposal concept

H12V

PEM

Rock type

Plutonic

Neogene

Pre-Neogene

Plutonic

Neogene

Pre-Neogene

Panel type

TT

TT

DET

DET

DET

DET

Total cooling
capacity (kW)

1,200

Not
required

3,200

800

Not
required

1,600

For example, for H12V in plutonic rocks, by installing 200 kW capacity cooling systems at
six locations, the air temperature requirement can be satisfied. However, optimisation ofcooling system locations has not yet been considered, nor any options that could utilise thewaste heat.
4-125

(2) Design of drainage
Drainage requirements are assessed for the repository layouts set in Section 4.5.4 (6). It
should be noted here that management of drainage water is a very sensitive (and expensive)
issue in Japan. Thus, efforts to reduce water inflow to the extent possible are justified (anadditional benefit being that the environmental impact is also reduced).
(i) Design requirements
The design requirements for drainage are given in Table 4.5-31, which focuses on
operational safety related to maintaining a safe working environment (see Table 4.2-3).
Table 4.5-31 Drainage design requirements
Designrequirements
Assuring drainage
capacity

Detailed description of requirement for concepts assessed

Specifications

All water inflowing underground can be safely drained andpumped to the surface for disposal

Drainagecapacity

In order to ensure sufficient drainage capacity, the water inflow into specific layouts has tobe calculated.

(ii) Required drainage capacity
Required drainage (and the location of pumps and pipes) is defined for the layouts
established in Section 4.5.4 (6) (details are given in Supporting Report 4-64). In order toestimate the volume of water inflow, the method used accounts for both the EDZ depth andany grout improvement (grouting to reduce water inflow) associated with excavation oftunnels [127]. An appropriate cement-based grout is assumed, with a composition that will
not risk perturbation of EBS performance [128] [129]. For illustrative purposes, total inflows
are estimated for different layouts based on pressure heads experienced and discharge
reduction due to grouting, for an arbitrarily assumed grout depth of 3 m. Calculated
groundwater discharge rates are given in Table 4.5-32, but these would clearly need to be reassessed for site-specific layouts and implementation plans.
Table 4.5-32 Calculated groundwater discharge (grout depth 3 m)
HLW (m3/min)Rock type

TRU waste (m3/min)

H12V

PEM

Vault

Plutonic

17.3

10.4

4.6

Neogene

4.6

1.4

1.2

Pre-Neogene

3.2

3.0

2.3

For HLW, in all rocks, discharge is less for the PEM case compared to H12V because of
the shorter total tunnel length (see Table 4.5-23 for tunnel lengths). In addition, for both HLW
options and TRU waste, plutonic rocks give higher discharges than the other rocks because ofa higher average hydraulic conductivity.
In order to determine the capacity of drainage infrastructure (drainage pipe diameter,
number of pumps and required power), an allowance for variations of inflow volume(assumed to be a factor of three) is used to derive the specifications shown in Table 4.5-33.
4-126

Table 4.5-33 Drainage facilities capacity (for grout width 3 m)

Rock

Emplacement
methodDesignbasis
drainage

Head

Drain pipe innerdiameter(m3/min)(m)

Nominaldiameter(mm)

Numberofpumps

Required
power
(kW)

HLW repository (3 drainage shafts)
H12V

51.9

1,000

350A

318

6

20,100

PEM

31.2

1,000

250A

237

6

12,180

H12V

13.8

500

250A

249

3

2,700

PEM

4.2

500

150A

151

3

810

H12V

9.6

1,000

200A

191

3

3,780

PEM

9.0

1,000

200A

191

3

3,450

Plutonic

Neogene

Pre-Neogene
TRU waste repository (1 drainage shaft)
Plutonic

-

13.8

1,000

300A

284

2

5,380

Neogene

-

3.6

500

250A

249

1

680

Pre-Neogene

-

6.9

1,000

300A

284

1

2,630

It was assumed that pumping from the bottom of the drainage shaft to the surface utilises a
super high lift, large capacity pump. The material of the drainage pipe is carbon steel (JIS G
3454), which has a working pressure of about 10 MPa (equivalent to a head of 1,000 m).
4.5.6 Safety measures during construction and operation
The design of the underground services presented in Section 4.5.5 is based on normal
operational conditions. Normal conditions include inherent variability (e.g. of water inflow as
considered above), but not accidents or equipment malfunctions. This is a requirement for
nuclear facilities, which has been extended to repositories. Also abnormal conditions are
considered in the design and currently include:Drops during transportation and emplacement of waste.

Fires and explosions: caused by ignition of equipment used in underground facilities,
accumulation of flammable gas, etc.

External power loss: failure of services due to loss of electricity supply from external
power sources.



Other equipment failure: damage to equipment or loss of services due to malfunction,
incorrect operation by workers, etc.

The need to physically separate underground construction from operation is considered in
the layout of disposal panels and connecting tunnels, to separate radioactive areas from nonradioactive, but the current assessment focuses on issues during operation, with no assessment
4-127

of potential impacts from construction activities. Furthermore, the current assessment looks atincidents individually (without considering common mode failure). Such coupling will beconsidered in the future. Issues not directly considered in this Section include:Direct radiation exposure to the public from the repository, due to effectively completeshielding by the overburden.

Direct worker radiation exposure from transport and emplacement of the waste, as
these actions are assumed to be under remote control and zoning will prevent workers
approaching the waste, even in abnormal conditions (except for special cases during
recovery work).



Movement of active faults that cause direct disruption underground, as the presence of
these will be excluded by the site selection process (see Sections 3.2.2 (1) (i) (c), 3.2.2(2) (i) (c) and 3.2.2 (3) (i) (c)).

Seismic impacts on tunnels: these are provisionally designed to withstand expected
events, but more detailed assessment is only reasonable in the future for specific site
conditions.



External surface perturbations, such as tsunami, floods, typhoons, tornadoes, etc.,
would not directly affect the repository, but could be associated with risks such as
inundation or loss of surface services. Again, these are sensible to assess only after theenvironment of specific sites is defined.

Issues associated with illegal intrusion into waste management facilities, which will be
excluded by the design of surface facilities.



Relevant infrastructure related to abnormal event management (e.g. monitoring,radiation dose measurement, communications, etc.), which are also presently assumed
to be located at the surface.

Prevention of abnormal incidents involving radiation risks, which is covered in Chapter
5.

The identification and description of the abnormal states assessed are presented in moredetail in Supporting Report 4-65.
For the case of HLW, as mentioned in 4.2.4 (1) (i), all handling underground involves waste
within a massive overpack, possibly complemented by transport shielding or a surrounding
PEM. The assessment thus needs to consider if, for any of the abnormal conditions listed
above, the key safety function of containment during operation could be lost and RNs released.
Release of RNs requires both that the overpack is severely damaged and also damage to thestainless steel fabrication canister, so that the glass matrix is exposed and could releasecontaminants in a form transportable by air or water.
In the case of TRU waste, RNs may also be released if the waste package, container andwaste matrix are damaged. The situation here is more complex, as packages (especially waste
package A) are less protective than the HLW overpack. Gr.3 is potentially flammable andTRU waste matrices may be less robust than HLW glass. TRU waste will thus be a priority
for further studies of perturbation scenarios.
The stepwise implementation of safety measures to prevent abnormalities developing or
propagating into accidents is outlined in Figure 4.2-7, emphasising the concept of defence in
depth. Nevertheless, despite multiple safety measures to prevent serious damage to waste
4-128

packages, additional precautions are introduced to reduce the probability and consequences ofany scenario leading to possible RN releases from the repository.
(1) Package drops
In the repository, the main risk of a drop for H12V is currently considered to be associatedwith the process of lifting the overpack:At the bottom of the access ramp, overpack transfer from the transport vehicle tothe emplacement device using an overhead crane (see Section 4.5.4 (4) (iii)).



Within disposal tunnels, lowering overpacks into disposal pits by the emplacement
machine (see Section 4.4.3 (1) (i)).

Possible initiating events for overpack drops and countermeasures against them aresummarised in Table 4.5-34.
Table 4.5-34 Safety measures against HLW overpack drops
Possible abnormal
condition
Breaking of thelifting wire

Regular inspection and maintenance of equipment
Double lifting wire

Consequence reduction
measures
Handling height
limitation

Lifting withincomplete grip

Regular inspection and maintenance of equipment
Interlock that does not allow lift unless it is secure

Handling height
limitation

Accidental releaseduring lift

Regular inspection and maintenance of equipment
Interlock that does not release grip unless package iscompletely lowered

Handling height
limitation

Probability reduction measures

Drops could involve failure of a lifting wire, lifting without a fixed grip and accidental
release – the probabilities of which can be reduced by regular inspection and maintenance,
together with more robust designs including double wires and grip interlocks. In all cases,
consequences can be reduced by fixing a handling height limit, which is physically impossible
to exceed. For the PEM, there are no significant lifts involved - even when raised by theemplacement machine, the clearance is small and, even in the event of a drop, the risk ofdamage to the overpack is negligible due to protection by the PEM handling shell andcontained buffer.
TRU waste package B is emplaced in the disposal vault by an overhead crane gripping a
twist-lock mechanism (see Section 4.4.2 (2) (ii) (b) and Supporting Report 4-21). This lifts
waste packages from the transport vehicle, moves them within the vault and lowers them into
position on a waste stack (see Section 4.4.3 (2)). Fundamentally, issues associated with drops
are the same as in Table 4.5-34. For TRU waste package A, lifting and emplacement is byforklift truck, so an interlock mechanism that does not allow lifting of the waste package if
fork insertion is incomplete is necessary. Despite these safety measures, the evaluation ofwaste drops is covered in Section 5.4.2 (1).
(2) Fire
Fires require combustible materials, an ignition source and oxygen: eliminating any one ofthese will usually prevent the occurrence of fires. An exception involves materials that could
self-ignite, with TRU Gr.3 representing the only relevant case – which will be examinedfurther in the future.
4-129

Although the HLW overpack is clearly non-combustible, there is a possibility of othercombustible substances being present in the HLW repository:

Access ramp; fuel and wheels of the road transport vehicle (see Figure 4.5-36).



Bottom ramp; lubricant for overhead crane used for overpack transfer from ramp
transporter to emplacement machine (H12V only).

The fuel and wheels of the H12V emplacement machine.

Based on past underground fires [130], it is considered to be effectively impossible toexclude ignition sources completely, especially for transport vehicles. For the maintenance of
the working environment (Section 4.5.5 (1)) ventilation will ensure the presence of sufficient
oxygen. Possible initiating events for fires and countermeasures against them are summarisedfor HLW in Table 4.5-35.
Table 4.5-35 Safety measures against fires (HLW)
Possibleabnormalcondition

Probability reduction measures

Consequence reduction measures

Fire inequipment
handling waste

 Limitation of ignition sources Maintenance of vehicles and devices
that could serve as ignition sources
 Limited use of combustible material
 Use of non-flammable and flameretardant materials

 Use of non-flammable and flameretardant materials
 Thermal anomaly detection monitors,
fire alarm equipment
 Fire supressing equipment (automatic tothe extent possible) and fire-fighting
actions

Measures to prevent the occurrence of fires in underground facilities include limiting
ignition sources and combustible materials to the extent possible, together with regular
maintenance of vehicles and equipment that may be sources of ignition or fuel. When a fire(or precursor such as heat or smoke) occurs, rapid fire detection and response (automatic fire
extinguishing equipment or active fire-fighting) are important to minimise impacts andprevent the spread of fire. The widespread use of non-combustible and flame-retardantmaterials is also considered to be effective in preventing the spread of fire. Ventilation control
measures could also be effective and will be examined further in the future.
In terms of the assumed equipment in the operational part of the repository, the greatest firerisk is the road transporter for H12V overpacks, PEMs and TRU waste packages (illustrated
in Figure 4.5-36, along with a breakdown of the flammable components present). The spreadof fire to bituminised waste (Gr.3) must be prevented, but there is qualitatively no differencein the handling compared with other wastes.

4-130

Figure 4.5-36 Heat generated from a transfer vehicle fireThis vehicle has a longer travel time than the H12V emplacement machine and requires
fuel for descent and return to the surface, so would have a larger tank than the H12V
emplacement machine. Other sources of combustible material, such as lubricating oil foroverhead cranes, are estimated to be minor by comparison, so the assessment of firecountermeasures focus on transport vehicles in the access tunnel (see Supporting Report 4-66),
as captured in Table 4.5-36.
Table 4.5-36 Safety measures against transport vehicle fires
Possible abnormal
condition

Probability reduction
measures

Consequence reduction measures

Ignition of tyres (e.g. due
to abnormal heating
caused by brake failure) Inspection and maintenanceof transport vehicle
 Abnormal heat detection
device for wheels

 Protection provided by truck bed, steel
frame and transport shielding Abnormal heating detection device forvehicles, on-board fire extinguishingequipment
 Tunnel fire detection and extinguishingequipment

Ignition of light oil
leaking due to failure oftransport vehicle or
collision

 Inspection and maintenanceof transport vehicle
 Interlocks and protection
devices to prevent runaway
accidents or minimise theirconsequences As above

As noted in Table 4.5-36, credible fires could start in the tyres (e.g. as a result of brakeover-heating) or ignition of spilt fuel (e.g. as a result of a collision). Apart from reducing risksof both by rigorous inspection and maintenance of the transport vehicle, a countermeasure forthe former could involve installing a detector for abnormal heating of the wheels and for thelatter, could involve interlocks and protection (e.g. runaway capture zones) to reduce the risk
and consequence of collisions (see further subsection below on equipment failure). Fires on
ramps can be particularly problematic due to chimney effects, so ventilation controls (e.g.
automatic fire doors) will also be considered in the future. Future work will also consider
alternative means of transport to that of ramps.
Basically, the same transport vehicle is assumed for H12V overpacks, PEMs and TRU
waste packages, with only the specifications of the transport unit being different. For all 3
types of load, a preliminary assessment of simple fire and collision scenarios are considered in
Section 5.4.2 (2).

4-131

(3) Explosions
For the operational areas of the repository, explosions can result from the accumulation of
combustible gas, arising from the rock, waste or repository equipment/materials, in thepresence of oxygen and a source of ignition. A representative initiating event for explosions
and countermeasures against it are summarised in Table 4.5-37.
Table 4.5-37 Safety measures against explosions
Possible abnormal
condition

Probability reduction measures

Rise in flammablegas concentration ina tunnel Installation of flammable gas detection andalarm system at key points
 Dilution of flammable gas below explosion
point by ventilation
 Checks and maintenance of ventilationequipment and gas monitors
 Excluding potential ignition sources to theextent possible

Consequence reduction
measures Blast doors Ventilation controls Power cut-outs

As noted in Table 4.5-25, tunnel ventilation is specifically designed to keep problematicgases well below their explosion point. In addition, equipment and working procedures will
be designed to eliminate ignition sources to the extent possible. The above measures,combined with special protective equipment if needed (e.g. catalytic converters), should
reduce explosion risks to a negligible level. In any case, gas detection and alarm devices will
be installed in tunnels and response plans will be developed for scenarios in which this may
be a problem.
It should be noted that, in this report, the SDMs from Chapter 3 do not assume presence of
methane or other flammable gas. However, in some locations, Japanese Neogene sediments
contain significant quantities of methane. When investigating sites, starting from the LS, thepresence of combustible gases must be considered.
(4) Loss of external power
If all external power to the repository site is lost, it is assumed that this will impact manydifferent processes and services. Existing nuclear facilities reduce this risk by having several
independent power supply lines and, even in the event that all fail, backup power generators
on site that will provide sufficient power to run important services for an extended period oftime. In the worst case of total site blackout where both external and local power is lost,
batteries would allow key, safety-critical services to run long enough to bring the site to safe
shutdown conditions.
In terms of assessing impacts on underground facilities, the focus is on loss of power fromthe surface. Again, as for nuclear facilities, at least 2 independent power supply routes would
be planned, but it is possible that power could be lost due to either surface blackout or a
perturbation that disrupts all power supply routes (e.g. a major earthquake). It is notable thatthe impacts of these cases would be somewhat different (e.g. if surface power still available,
exhaust fans at the top of shafts would still drive ventilation) but, for present assessment, thefocus is on the former case. Table 4.5-38 thus focuses on impacts of the loss of external power
supply to underground facilities and associated countermeasures to reduce probability andconsequences of these, without considering the initiating event causing this loss. Morecomprehensive assessment of power loss scenarios will be carried out for specific4-132

sites/disposal concepts in order to develop countermeasures. Nevertheless, in general, theprobability of loss of such supply is reduced by increasing the number of independent powersources/supply routes.
Table 4.5-38 Safety measures against impacts of loss of external power supply
Possible resulting
abnormal condition

Probability reduction measures

Consequence reduction measures

Drop while handlingwaste Multiple independent powersources and distribution lines
 Mechanical gripping device (nopower required to maintain grip)
 Fail-safe lifting cable brakes in
event of power loss Backup power to allow lowering
to safe position
 Handling height restriction

Flooding of facilities
following loss ofdrainage Multiple independent powersources and distribution lines

Accumulation of
flammable gas due toloss of ventilation
Loss of fire detectionand extinguishingequipment Multiple independent powersources and distribution lines
 Catalytic convertors

Loss of control system Multiple independent powersources and distribution lines Multiple independent powersources and distribution lines Backup power for undergroundpumps
 Layout of drainage/capacity ofsumps
 Failsafe closure of bulkhead doors
 Backup power for undergroundventilation control
 Failsafe closure of bulkhead doors
 Backup power for key safetycritical components
 Failsafe closure of bulkhead doors
 Backup power for key safetycritical components

The possible abnormal conditions during the process for lifting and moving
overpacks/PEMs/waste packages are outlined in Section 4.5.6 (1). For the specific case of lossof power, it is important to include mechanical interlocks/brakes that fail-safe in the event thatpower is lost.
Ventilation requirements are outlined in Section 4.5.5 (1). In the case of loss of power,
workers would be immediately evacuated and, even if ventilation was lost, would be able toescape to surface or to a suitable-equipped safe refuge. If ventilation was lost, as considered in
Section 4.5.6 (3), the main concern would thus be accumulation of combustible gas leading toa fire or an explosion. To prevent this abnormal state from occurring, catalytic convertors
could be installed, while impacts can be reduced by backup power for key ventilation controls(e.g. booster fans) to facilitate the action of the convertors and failsafe bulkhead doors to
reduce consequences of an explosion.
Section 4.5.5 (2) outlines the drainage system concept which, in the event of failure, could
lead to flooding of at least some part of the repository. Although this may not lead to aradiological risk, flooding may damage underground infrastructure and recovery may be
costly and hazardous. To prevent flooding, backup power for pumps and a drainage layout
with large sumps that provide time for power recovery will be implemented. General layout,
potentially combined with the option of bulkhead doors that failsafe closed, could reduce risks
to more sensitive areas by diverting water to empty tunnels/vaults.
Section 4.5.6 (2) overviews fire protection measures, which can be assured, even in theevent of loss of external power, by backup power to safety-critical components. Again, theoption of bulkhead doors that failsafe closed could be considered to both protect sensitive
areas and smother fires in areas where no workers are present.
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Because of the multiple impacts of power loss, central control plays a key role in incident
management – assuring safety of all workers and bringing the site to a safe shutdown. It isthus important to identify and provide assured power to all safety-critical components of thissystem.
(5) Equipment failure
Many of the abnormal states considered above can be traced back to perturbations (impacts,
fires, floods, etc.) caused by equipment failure or malfunction due to operator error. There arecases, however, when complex abnormal states can develop which contain several of these
perturbations as a result of a type of common-mode failure. When such scenarios are
identified, special efforts must be made to provide defence in depth in terms of the initiating
event, reducing probability of occurrence to the extent possible and providing
countermeasures to reduce consequences if it does occur (as explicitly required by nuclear
regulations). This is illustrated for the case of loss of control of the waste transport vehicle in
the access ramp, as a high-speed collision could result in a physical impact on the waste
package, a fire, damage to the tunnel liner, loss of a ventilation inlet – all causing further
possible perturbations.
If a failure occurs in the braking system of the transfer vehicle while descending the accessslope shaft, a runaway could occur with speed continuously building up leading to an eventual
collision. Safety measures to prevent this or reduce consequences are shown in Table 4.5-39.
Table 4.5-39 Safety measures against runaway of the waste package transporter
Possible resulting
abnormal condition

Probability reduction measures

Consequence reduction measures

Transport vehicle
runaway due tomalfunction oroperator error

 Regular maintenance of transport
vehicles Independent failsafe speed
restriction mechanisms

 Speed restriction by horizontal ramp
sections
 Capture by runaway escape orarrestor sections
 Robust waste packaging

In order to prevent the occurrence of the abnormal state, regular maintenance of thetransport vehicle is complemented by provision of one or more independent, failsafe breaking
or speed-limiting mechanisms. In the event that the vehicle cannot be stopped by on-board
equipment, maximum speed can be limited by including horizontal ramp sections and it canbe captured in runaway escape or arrestor sections. Even though the probability of such a
runaway resulting in a collision with the tunnel wall may be very low, the consequences areevaluated further in Section 5.4.2 (5).
(6) Mitigation measures for radiological impact
Sections 4.5.6 (1) to (5) above have considered a range of abnormal conditions andmeasures that can reduce their likelihood of occurrence and their consequences, aiming toensure that the fundamental safety function of containment of radioactivity by the robust
waste package is not lost. However, as highlighted in the US WIPP radioactivity leakage case[130], failures in QA of waste conditioning and packaging can give rise to unanticipated,abnormal conditions leading to loss of containment. Although no particular scenario causing
this is identified for the NUMO case, radiation monitors will be located in all sensitive areas
and, in the event of detection of an anomaly, emergency drainage capture/ventilation through
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HEPA filters would be initiated in order to capture as much radioactive material as possible
and confine contamination to the site, thus preventing any significant release from therepository.
(7) Worker safety
For workers involved in construction and operation in underground facilities, safetymeasures in the event of fires, flooding, rock falls, rock bursts, etc. are described.
In terms of fire protection measures, limiting use of combustible material and presence ofignition sources, together with fire detection and extinguishing equipment, are the same aslisted in Table 4.5-34. However, when using explosives for excavation, absolute exclusion
cannot be applied and conventional mine safety measures are implemented to reduce
associated hazards. In addition to these measures, in order to ensure the safety of the workers,
it is planned to make fire-fighting activities remote-handled, to the extent possible. All
personnel not involved in firefighting operations require evacuation from undergroundfacilities; as noted in Section 4.5.4 (4) this is allowed for by independent evacuation routes
through connecting tunnels to access ramps/shafts with multiple emergency shelters in casethese cannot be used. In addition, considering that the construction and operation undergroundare carried out in parallel, zoning will strictly separate these activities and, for the case of fire,airlocks, fire doors will prevent both spread of fire and smoke penetration between thesezones.
The hydrostatic pressure at repository depth, as noted in Section 4.3, is 5 MPa or more.
Such high groundwater pressure presents a flooding risk, which is countered by measures
such as proactive grouting, rapid liner installation and assuring sufficient margins in thedrainage system (see Section 4.5.5 (2) (ii)) based on the current state-of-the-art in civilengineering technology for tunnel construction.
Because the lithostatic pressure is also high deep underground, there are risks of facecollapse and rock bursts during excavation. For both these cases, risks can be reduced byappropriate design of tunnel layout and excavation technology/support structures based on
information on the geological environment obtained during site survey and construction, withcharacterisation drilling ahead of face excavation and appropriate rock stabilisation measuresapplied as needed.
Since the time from the start of construction to the completion of closure extends for over
50 years, deterioration and potential failure of tunnel support materials and infrastructure
equipment installed underground has to be assumed. To reduce risks, regular inspections,maintenance, and replacement/upgrading of equipment will be considered. Furthermore, it isconsidered cost-effective to reduce risks by setting up backups for essential services such as
power supply, drainage and ventilation – assuring that, at the very least, requirements for safe
worker evacuation can be assured even for a worst-case scenario.
It is recognised that, in addition to the physical safety measures mentioned above,
administrative and organisational factors also play a role in assuring worker safety. This
includes implementation of a safety culture throughout NUMO and all support contractors
and specific actions such as formal risk management of all activities, education and training of
all relevant staff, implementing resilient communication systems to manage responses toperturbations and assuring that all experience with anomalies is captured and analysed to feedback to improvement of countermeasures and hence prevent recurrence.
4-1354.5.7 Engineering technology for construction, operation and closure ofunderground facilities
In this section, the state of development of engineering technology for construction,
operation and closure of underground facilities is summarised. A particular focus is qualityassured construction of the EBS, as outlined in Section 4.4.3.
(1) Construction of underground facilities
With regard to the excavation of the tunnels shown in Section 4.5.2, a construction methodwith a well-established track record of safe implementation for relevant civil engineering
work in similar geological environments will be selected. For all the tunnels, shafts and ramps,
the NATM can be selected as a standard method of excavation based on wide experience inJapan. This method uses steel supports, shotcrete and rock bolts in combination to maximise
the support capacity, is excellent in dealing with changes in rock quality, and is economical.
In addition, for the case of H12V, disposal holes are drilled following the completion of theoverlying disposal tunnel. To date, demonstration tests have been conducted using largediameter and full-section drilling machines developed specifically for this purpose [131] [132].
From these results, although drilling of such disposal hole can be implemented with existing
technology, it is considered important to improve drilling efficiency in the future. Details ofthese drilling techniques are given in Supporting Report 4-67.
During actual construction, it may be difficult to excavate deep tunnels under certain
conditions, such as encountering a large fault or structural discontinuity, where the rockstrength is low and/or water inflow is large. For this reason, with URL projects, theapplicability of relevant grouting methods before and after tunnel excavation has been
demonstrated [122] [123]. However, while underground engineering can handle mostsituations, encountering conditions where no clear engineering fix is at hand might result in
considerable delay in the excavation activities. Further studies of such problem areas from anengineering perspective will provide guidance to site characterisation, with the aim ofallowing layouts to be planned that avoid these to the extent possible.
(2) Backfilling of disposal tunnels and installation of plugs
Backfilling of disposal tunnels has to be assured to meet required specifications (seeSection 4.5.3 (1)), which would require confirmation methodology. From domestic andforeign experience to date, various methods utilising pre-compacted blocks, in-situ
compaction, spraying, pellet infilling and fluidised emplacement are potentially applicable.
The actual backfilling method and implementation (either manual or remote handled) will beselected in consideration of the characteristics of each disposal tunnel and construction
efficiency for specified geological environments.
Additionally, for H12V, a massive bentonite block, with the same specifications as buffermaterial, is emplaced using remote handled equipment in the upper part of disposal hole after
installing the EBS. As no borehole cap is currently considered, backfilling the overlying
tunnel must be carried out as quickly as possible thereafter to avoid loss of quality due toswelling or piping erosion.
TRU vault backfilling could be manual, due to radiation shielding of the already emplaced
EBS materials, so in-situ compaction with superior construction efficiency and effectivenessfor backfilling to relevant standards is assumed (although remote operation may be considered4-136

in the future to reduce worker risks). In addition, for areas with tight spatial restrictions, thiscan be combined with spraying methods, which have been seen to be practical in full-scaledemonstration tests in underground cavities [85].
For the PEM, the gap between the PEM and the disposal tunnel is assumed to be backfilled
by automated (possibly tele-operated) methods (pellet infill or spraying [13], [133]) due to thepractical challenges resulting from the confined spaces between both the PEM and the rockwall, as well as radiation protection considerations (assuring doses are as low as reasonably
achievable - ALARA).
At the end of the disposal tunnel, a mechanical plug made of reinforced concrete is assumed
to be installed in order to prevent the expansion of the backfilling material into the unfilled
space and resultant loss of quality (Figure 4.5-12). In order to construct the plug, it isnecessary to carry out tunnel support removal, excavation of the plug widening section(notch), rebar and formwork assembly and then concrete emplacement [134]. The concrete
plug construction technology is based on experience found in the national oil and gas storage
knowledge base [135]. At present, full scale demonstration tests of plugs based on various
design concepts are ongoing to assess engineering practicality in underground researchfacilities in Japan and abroad [136] [137] [138] [139]. Nevertheless, it is clearly undesirable
to carry out all such work after waste emplacement (especially for H12V as this requires
tunnel sealing as quickly as possible) and hence a concept needs to be developed in which as
much as possible of the structure is produced during the initial panel excavation.
(3) Repository closure
As described in Section 4.5.3, closure of the repository is intended to isolate the repository
from the surface, sealing all access (shafts, ramps) together with connecting tunnels and othertunnels used for construction and operation of the repository. All exploration boreholes arealso sealed and hydraulic plugs emplaced as required. After backfilling, to prevent human
intrusion, high strength concrete plugs are emplaced at key surface access points (see Figure4.5-10 (b)).
From the viewpoint of construction efficiency, the in-situ compaction method is effective
for backfilling of connecting and access tunnels. In the case of horizontal tunnels and ramps,
backfill material is poured onto the lower half of the tunnel and compacted to a predetermined
density with a roller compactor (such as a vibrating roller), with the upper half poured to forma slope which is compacted to a pre-determined density by a compaction vibrator [140] (seeFigure 4.5-37 b). For shafts, it is easy to backfill while sequentially compacting (See Figure
4.5-37 a).
For the construction of hydraulic plugs, stacking pre-compacted bentonite blocks andfilling interface gaps with blown bentonite pellets is effective and has been confirmed by fullscale in-situ testing [104], but other options may be more practical. Since the installation of anotch in the rock (to cut the EDZ) is a new excavation, it is necessary to select a construction
method that does not form a further damaged area outside the notch (for example, combining
line drilling and rock-splitting methods [134]).

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(a) Access shaft(b) Tunnels and ramps
Figure 4.5-37 Illustration of backfilling work

Boreholes from the surface are potential short circuits between the repository and thesurface if there is a significant upwards hydraulic gradient (e.g. artesian conditions), but must
be backfilled and sealed in any case due to regulatory requirements. Based on previous
empirical tests, backfilling with compacted bentonite in low permeability rock sections may
be effective, although such clay may be eroded in permeable fracture zones, in which case,concrete plugs may be more suitable [101] [141] [142]. Alternative methods demonstrated at
depths of several hundred metres or more involve bentonite emplacement within a perforated
copper pipe or as a cylindrical block which is pressed out of a container at depth, in both cases
sealing the hole as the bentonite swells following groundwater uptake [101] [142]. However,there is no standard methodology and the applicability of such methods will be confirmed byborehole tests. Finally, while borehole sealing requirements would be site specific, overseas
experience suggests (i.e. SKB TR-12-08) that using even relatively high permeability backfill(e.g. sand or rock spoil) for the purpose of borehole backfilling could be sufficient.

4.6 Design of surface facilities
The surface facilities receive and handle the waste transported from interim stores andsupport the work of construction, operation and closing of the repository (see Figure 4.2-5).
The functions of these facilities are diverse and they should be laid out appropriately based on
the characteristics of the site topography. It is necessary to construct many of the surface
facilities prior to construction underground and these will eventually be removed during
closure of the repository [2]. Surface facilities for both HLW and TRU waste were briefly
described in Section 4.2.3 (4). In this report, the focus is on facilities for receiving, inspection,and encapsulating/packaging HLW and TRU waste, for which radiation protection is an issue.
Other facilities that do not handle radioactive waste are similar to conventional industrial orcommercial buildings and will be considered only later in the site selection process.
4.6.1 Assuring safety in the design of surface facilities
Prior to the design of waste acceptance/inspection/packaging facilities, relevant backgroundon transportation of wastes, treatment of potential natural perturbation phenomena and safetyassessment of anthropogenic impacts will be described.
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(1) Waste transportation
Radioactive waste transportation from nuclear power plants, reprocessing plants, etc. is wellestablished both domestically and internationally, with international standards and regulations
such as those established by the IAEA [143] and additionally Japanese laws and regulations
based on these.
Transport of waste to the repository surface facility from an interim store utilises dedicatedtransport casks. The technical standards and quality assurance plan of such transport casks aredefined in the previously mentioned IAEA transportation regulations, based on extensive
testing programmes - drops, fire, immersion, etc.
Both sea and land transportation (road and rail) are well established for long-distances
although, if it is an option, exposure risk to the public is the lowest for marine transportationand this offers maximum flexibility for larger volumes of waste [144]. For this reason,
preferred sites in the Nationwide Map (Section 1.3) are coastal locations and it is assumed
that that marine transportation to a harbour near the repository site is followed by only local
transfer on land.
Radioactive waste transport ships currently in use are designed for safety for all credible
accidents, with double hull structures which are collision resistant, duplicate navigation
systems, ship tracking systems using satellite communication, etc. [145]. Although transportcasks for radioactive waste, specifications of transport vehicles, and transport routes are
undecided at present, experience with SF and returned vitrified waste from abroad indicatethat appropriate solutions can be readily developed as and when required (see Supporting
Report 4-68).
(2) Managing natural perturbation phenomena
The risks of natural perturbation phenomena are strongly dependent on the geographicaland geological setting of the site, which defines the design requirements necessary to assure
safety. Some extreme natural phenomena can be excluded by siting, as listed in Table 3.1-1,
including pyroclastic flows, lava flows, lahars, etc. Nevertheless, a range of natural
phenomena that affect the safety of surface facilities must be assessed, such as earthquakes,
tsunamis, landslides, floods, etc. as listed in Table 4.6-1, and appropriate safety measures
implemented to reduce potential impacts, as is normal practice for sensitive structures [146].
As an example of this approach, design countermeasures for tsunami risks are illustrated in
Figure 4.6-1. When locating sensitive facilities, the aim is to install them as high above sea
level as possible.

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Table 4.6-1 Safety measures for surface facilities considering natural perturbations
Naturalperturbation

Safety measures (assuming effects constrained by appropriate siting)

Earthquake

Seismic design to ensure structures can withstand credible seismic forces

TsunamiUtilise the topography of the area, e.g. installing key facilities as high as possible,
setting protective seawalls, installing inundation prevention doors, etc. towithstand credible tsunami impacts

Flood

Utilise the topography of the area, e.g. installing key facilities as high as possible,
installing inundation prevention doors, etc. to withstand credible floods

Wind (typhoon)

In consideration of the maximum wind speed observed around the site, the facility
is designed to have sufficient strength to resist them

Tornado

The facility is designed not to be damaged by tornado wind pressure and impact
load of flying objects

Freezing

Design based on the lowest temperature expected at the site

Precipitation

Design based on the maximum precipitation expected at the site. Establish adrainage plan to prevent flooding of buildings

Snowfall

In consideration of the maximum amount of snowfall expected at the site, thefacility is designed to withstand credible loads

Lightning
strikeDesign lightning protection equipment to provide required protection

Landslide

In consideration of the observation records and the topographic and geologicalsituation of the site, landslide countermeasures can be applied as needed

Volcanic effects

Even if at a distance from nearest active volcanoes, ash-fall must be consideredwith design to assure that credible events would not compromise facility safety

Biologicaleffects

Design to exclude animals, e.g. exclusion nets used to prevent bat/bird entry ofaccess shafts

Forest fire

Take protective measures if relevant, such as providing a barrier (fire break)between the facility and any surrounding forest

Figure 4.6-1 Illustration of tsunami countermeasures

4-140

If the tsunami risk is considered significant, it would be planned to take measures such as
installing a seawall and installing robust doors to prevent inundation. Furthermore, in order toprevent a tsunami flooding the repository, as illustrated and discussed in Section 4.6.2 (4), theaccess ramp will be located at basement level of the surface facilities (i.e. underground –which will also avoid potential animal access). For other accesses to the surface, such as
ventilation shafts, appropriate tsunami protection will be included in their design.
(3) Managing anthropogenic perturbations
Risks of anthropogenic perturbations depend on the geographical and social environmental
conditions around the site, which defines the design requirements necessary to assure safety.
Similar to the case of natural perturbations described above, appropriate safety measures are
implemented to reduce potential impacts, as is normal practice for sensitive structures [146]:
these are summarised in Table 4.6-2. It is recognised that this list is far from comprehensive(e.g. no consideration of climate change impacts, consequences of pandemics, etc.), but
provides a starting point based on requirements to meet regulatory guidelines.
Table 4.6-2 Safety measures for surface facilities considering anthropogenic perturbations
Human perturbation

Safety measures

Aircraft impact

Structural design of sensitive facilities should account for largest credible
impacts

Dam collapse

If this is a risk for a particular site, take safety measures as for tsunamis

Industrial explosion

Assessment of nearby industrial activities, evaluate risk from worst credible
external explosion and assure sufficient separation from/protection ofsensitive surface facilities

Industrial fire

Assessment of nearby industrial activities, evaluate risk from worst credible
external fire and assure sufficient separation from/protection of sensitive
surface facilities

Industrial release of
toxic gas

Assessment of nearby industrial activities, evaluate risk from worst credible
toxic gas release and assure sufficient separation from/protection of sensitive
surface facilities

Industrial
electromagnetic pulse

Assessment of nearby industrial activities, evaluate risk from worst credible
electromagnetic pulse and assure sufficient separation from/protection ofsensitive surface facilities

Shipping accidents

For coastal locations, a special concern is shipping accidents – especially ifthese could result in massive perturbations (e.g. oil tankers, LNG carriers,
bulk ore carriers). Sensitive facilities should be located far enough away fromdeep water that they are protected from worst credible accidents

Illegal intrusion

Implement rigorous security measures against cyber terrorism. Implement
rigorous access control, with additional measures to prevent intrusion into
sensitive areas. This is complemented by walls/fences with intrusion detectors
around the entire site, security patrols, etc. Off-site monitoring and close
liaison with local authorities also allows any major threats (e.g. large
demonstrations) to be identified well in advance and appropriate protection
organised

In the case of prevention of impacts of illegal intrusion, the following characteristics of thewaste are of relevance [147]:
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

Waste is sealed in massive overpacks/waste packages that provide physical protection
and make diversion of such material extremely difficult.

Surface waste handling facilities are remotely operated, with strict control of accesseven for workers. They are also heavily constructed with massive concrete shielding,making perturbation from outside buildings very unlikely.

Based on these, the level of protection against illegal intrusion is considered to be evengreater than that at nuclear power plants and other nuclear facilities and thus this could beclassified as a “hard” target, with very low risk of any radiological hazard from such a
perturbation [147].
Nevertheless, the nuclear industry recognises that cyber-security is a key issue for all
sensitive infrastructure [148], and state-of-the-art systems would be a design requirement, notonly at the time of construction but also as an ongoing need that would be managed bycontinuous performance review and updating of both hardware and software.
4.6.2 Design of waste reception, inspection and encapsulation facilities
The design of facilities for waste acceptance/inspection/encapsulation requires that thewaste inventory and reference throughput are clarified for both HLW and TRU waste. In
addition, since the natural and anthropogenic perturbations to be considered in the design ofthese facilities are so site-specific, designs presently focus on generic requirements forradiation shielding and containment.
The design of the surface waste handling facilities depends on the operational work flow,from first waste receipt to dispatch underground, constrained by key design requirements
under normal operational conditions, such as the wall thickness necessary for radiation
shielding and negative pressure management to assure containment of RNs in the event ofaccidents. Finally, designs will be refined to include countermeasures to respond to abnormal
conditions, such as falls, fires and explosions.
(1) Configuration of facilities
The handling process for HLW and TRU waste starts with receipt of dedicated transportcasks (as described in Supporting Report 4-68) containing the conditioned waste (see also
Section 4.6.1(1)). After receipt, waste is moved to a controlled zone and removed from thetransport cask, inspected, encapsulated in an overpack or waste package container (twodifferent options for TRU waste considered) and then prepared for transport underground. For
the PEM option, the HLW overpack and surrounding buffer are sealed in the handling shellunder strictly controlled conditions. All such operations in the radiation-controlled area are
rigorously monitored/electronically logged and tele-operated to the extent possible, which
covers all normal operational conditions. Worker access is expected only for routine
inspection and maintenance activities during breaks of operation, or as a result of specificperturbations – providing the basis design requirements for radiation protection.
(2) Waste inventory and throughput
As discussed in Section 2.1.1 (4), the Final Disposal Plan envisages emplacement of 1,000
HLW canisters per year for 40 years. HLW shipments in a dedicated ship are assumed to4-142

occur twice a year [149], in each case including 18 transport casks, each of which contains 28
HLW canisters. This defines the intake storage capacity for the temporary store [144]. Aftereach transport cask is opened, 28 canisters need to be processed and this is balanced against
the reference disposal rate of 5 overpacked canisters/PEMs per day by other buffer stores.
These plans would be revisited and optimised as part of the iterative site-specific assessments
to take placed during the LS, PI and DI stages.
The reference TRU waste inventory (90,584 primary containers with a volume of 19,018m3) [11] requires a disposal throughput of seven waste packages (mostly containing 4 but
sometimes 2 primary containers) per day for 25 years [11]. The maximum loading of atransport ship is 3,000 tons, with 6 transports per year assumed. Table 4.6-3 shows theaverage quantity of TRU in each shipment, to serve as a basis for planning. As for HLW,
these plans will need to be revised and optimised in the future.
Table 4.6-3 Reference shipment of TRU waste (6/year)
Number oftransport casks

Number of primary waste
containers per transport cask

Drum (Gr.3, 4L, 4H)

12

36

Canister (Gr.2)

828

Box container (Gr.4L)

1

2

Container for MHHRW (Gr.4H)

2

4

Drum (Gr.1)

2

8

Primary waste container (waste group)(3) Design requirements
Top-level design requirements for acceptance, inspection and encapsulation facilities are
given in Table 4.1-1, based on the interpretation of existing regulations and related legislation
covering other nuclear facilities.
The requirements (safety functions) for operational safety focus on radiation containmentand shielding, both related to radiation protection (see Table 4.2-1). Radiation protection isdetermined for both the public living near the site and all staff working in it. Under normal
conditions, radioactive waste should be confined within designated radiation-controlled zones
and associated shielding would ensure that doses received by personnel do not exceed thelimits stipulated by law and, indeed, are ALARA. Radiation protection of the public andworkers, and associated assurance of containment functions, are thus set as design
requirements. In addition, regulations for nuclear facilities require provision of facilities tomonitor radiation releases and exposure of workers while, as described above, it is necessary
to monitor operations in order to identify and respond to any abnormalities.
As described in Section 4.5.6 for underground operations, abnormal conditions are also
assumed in the design of the surface facility, including impacts, fires, explosions, loss ofexternal power, etc., so preventing loss of safety functions in these cases is also set as a design
requirement.
Further issues normally considered in relevant industrial operations that are not discussed
further in this report are:

As all waste handling is assumed to be remote control, even in abnormal conditions,associated radiation exposures to workers are not currently assessed (except for special
tasks such as recovery).
4-143



The impacts of natural perturbations described in Sections 4.6.1 (2) and 4.6.1 (3) are
extremely site-specific and designs to reduce these are not considered at present.

Designs will be required to withstand maximum expected seismic loading, but thecalculation of credible seismic forces needs to be carried out based on the geologicalenvironment of a specific site and hence not assessed further at present.

Prevention of illegal access, as described in Section 4.6.1 (3) is a requirement, but it isreasonable to consider only at a later stage of detailed design of site-specific facilities.



Surface facilities providing conventional infrastructure and support services (e.g.
administration, staff support…) will be examined at a later stage of detailed design.

In terms of radiation protection, as evaluated further in Chapter 5, design requirements toassure safety are summarised in Table 4.6-4.
Table 4.6-4 Operational safety design requirements
Designrequirements

Function

Specifications

Radiation
protection of the
publicShielding and other appropriate measuresassure sufficiently low doses around the site

Facility installation layout,
shielding walls, ventilation
management, shielded transport
cask

Radiation
protection ofworkers

Shielding and other appropriate measuresassure sufficiently low doses in controlledareas and other places where workers arepresent

Shielding walls, transport cask,
limited access to radiation control
areas

Ensuring radiation
containment

Areas where radioactive contamination
could occur are maintained under negative
pressure

Radiation
monitoring andmeasurement

Comprehensive monitoring of thecontainment function, automated alarm andresponse to any anomalies

Prevention offunctional loss inabnormalconditions

Designs include measures to prevent theoccurrence of abnormalities and avoid lossof containment function even in abnormal
conditions such as drops and impacts, fire,
explosion, loss of external power andequipment failure

Radiation control area layout
Ventilation and drainage systems(including filters/effluent
management).
Waste containers, overpacks,
waste packages, transport casks
Monitoring and measurement
equipment, automatedidentification and response toanomalies
Nuclear standard, resilient designof structures, equipment andprocesses, including a philosophyof defence in depth to ensure
safety for all credible perturbation
scenarios

Table 4.6-5 outlines design requirements to ensure engineering practicality (based on
general requirement noted in Table 4.1-1). Since the facility needs to be able to store alltransport casks received in a shipment and then process these to provide packaged waste
allowing the reference disposal rate, appropriate buffer stores and waste handling capacities
are required. In addition, all operations in the radiation-controlled zone should be carried out
by remote control (either strictly required due to high radiation or as commitment to keep
worker doses ALARA), this is also set as a design requirement.

4-144

Table 4.6-5 Design requirements to assure engineering practicality
Designrequirements

Function

Specifications

Capacity of
buffer storesTemporary storage of transport casks received in asingle shipment and of waste containers during
handling of the contents of a single cask

Layout/design of the waste
handling facility

Waste
packagethroughput

Ability to process waste package at a specified
rate

Waste handling processplan, equipment design andmaintenance plan

All waste handling processes within the facility
carried out by remote control, including routine
Remote control
maintenance and, to the extent possible, responses
to abnormal conditions

Waste handling processplan, equipment design andmaintenance plan,
perturbation response plan

Finally, it should be again noted that the current plans will need to be revisited andoptimised as part of the iterative site-specific assessments that will take place during LS, PI
and DI stages.
(4) Design of waste handling facilities(i) HLW
A schematic diagram of the HLW receipt, inspection and encapsulation facility is shown in
Figure 4.6-2. As discussed further in Supporting Report 4-69, this includes a transport cask
interim store and an inspection/buffer storage cell in which HLW canisters are removed fromthe transport casks and then those passing acceptance stored before overpacking. HLW
canisters are inserted into the overpack and the lid welded in place within the overpack
production and inspection room, where the seal performance is confirmed before transfer tothe overpack buffer store prior to transport into the repository. As yet plans have not beendeveloped for handling canisters or seals that fail acceptance criteria; this will however form apart of future R&D.

Figure 4.6-2 Schematic outline of the HLW receipt, inspection and encapsulation facility(H12V)

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The size and layout of the waste handling line is set for a reference throughput of 1000
packages/year, giving a target of 5/day (see Section 4.6.2 (2)). For example, Section 4.4.1 (2)(viii) described the TIG welding method which takes about one day to weld a single lid andhence, assuming this option is chosen, 5 welding machines will be installed, with anadditional spare to facilitate planning of maintenance and inspection. However, it should be
noted that the target outlined here is only an example and decisions for the encapsulation
facility have not yet been finalised. In addition, this schematic outline needs to be refined in
the future in order to assess how required services are implemented (e.g. assuring sensitive
areas are below atmospheric pressure).
After testing overpack seals, overpacks will be transported into the repository. In NUMO
(2004) [2], the transport vehicle travels on the surface to the entrance of the access ramp, but
this has been reconsidered to develop a design in which ramp access is directly from thebasement of the waste handling building, as illustrated in Figure 4.6-3.
When this structure is adopted (discussed further in Supporting Report 4-70), it will bepossible to both simplify the work flow and also provide protection from surface
perturbations, especially those presenting a risk from flooding. It may also be possible forworkers in radiation-controlled areas of the repository to evacuate to the surface through theaccess tunnel. A ventilation tower will be installed to ventilate the access tunnel
independently of the waste reception, inspection and sealing facilities.

Figure 4.6-3 Illustration of the HLW receipt, inspection and encapsulation facility and its linkto the access ramp (H12V)

Although the above was designed with the H12V concept in mind, for the PEM, the samefacilities are used as far as the overpack production and inspection room. After this, as
discussed in Section 4.4.1 (3) (v) (b), the overpack is remotely placed in compacted buffer
within a steel handling shell in a PEM assembly cell (Figure 4.6-4) and then, after quality
monitoring, placed in a buffer store until loading onto a transport vehicle for transfer
underground [150]. Design of the PEM production line will be carried out in the future.

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HLW reception and inspection

OP: Overpack
Compaction of bentonite buffer
into handling shell rings

Encapsulation & inspection of OP

PEM assembly and discharge area

PEM assembly
Reception of OP

Assembly of handling shell rings

PEM HS: PEM handling shell

Reception of handling shell rings

PEM discharge

Transport of PEM

Reception of overpack

Lower PEM to horizontal position

Emplacement of OP in PEM HS

Attach handling shell lid

Gap filling

Inspection of handling shell lid

Temporary storage of PEM

Inspection of gap-fillInspection of completed PEM

PEM moved to transporter

Transportation of PEM undergroundFigure 4.6-4 Flow diagram of PEM manufacture (Source: Kitagawa et al., 2013 [150])(ii) TRU waste
A schematic diagram of the TRU waste receipt, inspection and encapsulation facility isshown in Figure 4.6-5. As discussed further in Supporting Report 4-71, this includes atransport cask interim store and an inspection/buffer storage cell in which TRU waste
containers are removed from the transport casks and then those passing acceptance stored
before packaging.

Figure 4.6-5 Schematic outline of the TRU waste receipt, inspection and encapsulation
facility

Based on the throughput outlined in Section 4.6.2 (2), TRU waste containers are divided
into different groups and are inserted into appropriate waste packages, which are infilled withmortar using a remote filler device. As curing of mortar takes about 3 weeks, the waste
package is stored under controlled conditions during this time. It should be noted that, inregard to temporary stores, it has not yet been decided if the TRU waste groups will be stored
together or if individual stores for each will be required. As waste groups will placed in
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identical waste package containers, a system of waste labelling and individual package
tracking from receipt to final disposal will be required in any case.
In the case of waste package A, after inspection it is placed in a dispatch store prior totransportation underground. In the case of waste package B, after curing of the mortar, an
upper lid is welded in place in the waste package production and inspection cell, then the seal
performance is confirmed before transfer to the waste package buffer store prior totransportation underground. The direct connection from the waste handling building to theaccess ramp is also shown in Figure 4.6-3.
As for the HLW case, as yet the handling procedures in the case of waste/packages notmeeting specifications have not yet been defined.
(5) Radiation protection assessment
For radiation shielding design (more details in Supporting Reports 4-72 and 4-73 for HLW
and TRU, respectively), the effective dose received by workers under normal conditions
should not exceed the limit defined in nuclear power reactor regulations (50 mSv over 1 year,
100 mSv over 5 years). Although the operations illustrated in Figures 4.6-2 and 4.6-5 areremote handled, workers will spend time in surrounding areas, with the type of work involved
used to derive criteria for design dose rate targets and hence required shielding thickness.
For general public in the vicinity of the facility, in addition to shielding in the waste
handling facility and that provided by surrounding buildings, the distance to the site boundary
reduces doses and should ensure that they are below values specified in regulations (limit 1
mSv/y, target 50 μSv/y). The assessment to confirm this is outlined in Section 5.3.
The containment of radioactivity is assured by transportation casks, waste
canisters/containers and overpacks/waste packages. Even in the case of surface contamination(identified during acceptance inspection) or releases due to a fall during handling (describedin Section 4.6.2 (6) (i)), negative pressure in the facility and ventilation/drainage management
will prevent any releases from the radiation-controlled areas.
(i) HLW
Table 4.6-6 lists as examples the calculated shielding thicknesses for the transport cask
buffer store and the HLW inspection and buffer storage cell, assuming shielding walls of
reinforced concrete, with details given in Supporting Report 4-72.
Table 4.6-6 Required shielding thickness for HLW handling areas
Structure

Transport cask buffer Waste inspection/buffer
store (m)storage cell (m)

Wall

0.60

1.3

Floor

0.65

1.3

Ceiling

0.35

1.2

After entry of the transport cask into the facility, all operations occur by remote handlingunder negative pressure and strictly managed ventilation/drainage discharges. In addition,
radiogenic heat from the transport cask temporary store is designed to be removed by natural
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convection, like existing interim stores. However, HLW in internal buffer stores is cooled by
active ventilation, as needed due to the negative pressure control.
The above are the results of a study for H12V, but the process up to overpack encapsulationis common for the PEM and, thereafter, there is no need for additional radiation protection
measures.
(ii) TRU waste
As for HLW, Table 4.6-7 lists as examples the calculated shielding thickness for the TRU
waste transport cask buffer store and the waste inspection and buffer storage cell, assuming
RC shielding walls (details in Supporting Report 4-73).
Table 4.6-7 Required shielding thickness for TRU waste handling areas
Structure

Transport cask buffer
store (m)Waste inspection/buffer
storage cell (m)

Wall

0.75

1.25

Floor

0.65

1.20

Ceiling

0.35

1.15

The basis for assuring containment of radiation during handling is the same as for HLW,
although here heat production is not a concern.
(6) Examination of safety in the event of perturbations(i) Drops
For transport casks, their design resistance to high energy impacts ensures that they are
effectively immune to any drop that might occur during their handling. Drops would befurther prevented by operational procedures, barriers, warning lights and sounds, etc.
For HLW, lifting and moving of the transport cask, the HLW fabrication canister, theoverpack and the PEM is by gantry crane (Figure 4.6-2); therefore, the same measures
introduced to prevent accidents from such lifts underground are directly applicable (see Table
4.5-34). If drops occur, consequences can be minimised by limiting the lifting height andproviding impact limiters on surfaces onto which such falls could occur.
The most vulnerable unit lifted is certainly the HLW canister although, based on previous
studies [151] [152], it has been confirmed that, even for drops from a height of 9 m, thestainless-steel container only deforms and there is no loss of integrity. For this reason,
although the facility is designed to limit the lifting height to the minimum practical, an
absolute limit of 9 m is specified. Although they are certainly more robust, lifting height
limits for all other units are also set to 9 m. In addition, where practical, drop limitation andimpact reduction measures are incorporated into the designs.
For TRU waste, the situation is very similar to HLW with the key difference being therobustness of the various primary waste containers. Amongst these, drums (Gr.1, Gr.3, part of
Gr.4L) have been subjected to drop tests from heights of up to 9 m. For heights up to 6 m
drums are deformed without loss of integrity but, above this, drum lids may burst or
penetrative cracks may occur [153]. It is possible that measures could be implemented by the4-149waste producer to tailor infill (robust concrete or surface bitumen coating) so that RN releaserisks can be minimised even in the event of drum containment failure.
For Gr.2 wastes (in a canister similar to that for HLW), there is again evidence that integrity
is not lost even when dropped to a rigid floor from a height of 9 m [154]. Box containers for
waste assigned to medium depth disposal are similar to those used for Gr.4L waste; drop testsof these from heights of up to 9 m show deformation and penetration cracks occur, but no lossof containment [76] [78].
Although lifting heights are always minimised, based on these results, an absolute drum
lifting height is set as 6 m, while that for box containers is 9 m. Although there is no specificevaluation for MHHRW containers, the lifting height limit for drums is assumed until further
evidence is available.
Handling of the waste package assumes movement within the facility by conveyor, but,when loading onto the transport vehicle, waste package A is lifted by a forklift and waste
package B is lifted by gantry crane. The handling height of the forklift depends on its lifting
capacity, but will be around the height of the platform of the transport vehicle (about 2 m).
Although lifting heights by the gantry crane could be up to 8 m, these will be strictly
controlled to ensure that they are at the minimum needed and cannot exceed defined limits.
(ii) Fires and explosions
The concept of fire prevention is fundamentally the same as that outlined in Section 4.5.6(2) for the underground facilities, focused on limitation of the presence of flammablematerials and their substitution by flame-retardant or non-combustible alternatives to theextent practical (details of flammable materials present and fire prevention measures are given
in Supporting Report 4-74). With current technology, fire risk cannot be completely
eliminated and hence fire detection equipment, alarms and fire extinguishing equipment(remotely operated when feasible) will be provided to minimise consequences. In addition,
surface facilities will include a fire brigade, with a particular responsibility of ensuring thatfires outside radiation-controlled areas are quickly controlled and cannot spread to moresensitive areas.
For TRU waste, Gr.3 bituminised nitrate does not fall under the designated flammable
substances specified in the Fire Service Act, but spontaneous ignition is possible under someconditions, particularly following biodegradation if this occurs during storage prior todisposal. In order to minimise risks, Gr.3 buffer stores (incorporating appropriate fire
extinguishing equipment) will be separated from those handling heat-generating waste by
fireproof barriers, such as concrete walls, steel fire-doors, etc.
No credible explosion risk has been identified to date for HLW, but the moisture in themortar filling for TRU waste generates hydrogen gas due to radiolysis; or hydrogen/methane
could result from either radiolysis or biodegradation of bitumen. Therefore, to reduce
explosion risks, countermeasures as shown in Table 4.5-24 would be implemented, including
explosive gas monitoring and backups to ventilation (e.g. catalytic combiners). If any specificwaste is assessed to present a significant explosion risk, options to reduce the quantity of
explosive gas that can be produced by changing waste conditioning materials or procedures,or even reconditioning waste, could be considered. Nevertheless, pros/cons and costeffectiveness of any such measures need to be carefully assessed, taking into account all
aspects of waste production/conditioning/packaging, handling, storage, transportation anddisposal.
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(iii) Loss of external power
Within the nuclear industry, “uninterruptable” power supplies are standard for safetycritical operations so that, if normal power supply is lost, local backups (batteries, generators,
capacitors, etc.) provide enough electricity for operations to be brought to a safe shutdown.
This is complemented by failsafe designs that, even with complete loss of power, default to an
inherently stable situation (e.g. crane grips for waste packages that require power for opening
rather than closing, lifting mechanisms that lock in place on power loss or use mechanicalrelaxation to return to a stable state). In general, surface facilities would be inherently more
resilient to such power loss than those underground.
(iv) Common mode failure
Any facility located on the surface is vulnerable to perturbations that can give rise tocommon mode failure (CMF): the coupled collapse of many of the barriers providing defence
in depth. The initiating events are inherently impossible to describe in detail, but examples
directly impacting surface facilities would include large earthquakes accompanied bytsunamis (as for Fukushima Daiichi), a major volcanic eruption with a huge ash fall (e.g.
Mount St. Helens) or a national/regional socio-political collapse (e.g. as a result of worst-caseglobal warming), all potentially leading to long-term loss of external services and also
damage to structures and services on site. As noted above, the heavily engineered structuresof safety-critical buildings, fail-safe design and extensive on-site service back-ups make thesurface facilities inherently more robust than most other nuclear or industrial facilities.
Nevertheless, a more detailed assessment of credible scenarios and, if required, adoption ofrequired countermeasures will be carried out after candidate sites have been identified.
On a smaller scale, failure of key equipment (or operator error) can lead to cascades ofimpacts. As for the case underground, when such scenarios are identified, special efforts must
be made to provide defence in depth in terms of the initiating event, reducing probability ofoccurrence to the extent possible and providing countermeasures to reduce consequences if it
does occur (as explicitly required by nuclear regulations).

4.7 Reversal and retrieval
As described in Section 2.3.3 (1), a design requirement is that emplacement processes canbe reversed and waste retrieved if required [155]. This presents potential conflicts with preand post-closure safety requirements, so that the following can be distinguished:

Reversal – stopping the waste handling/emplacement process at any time (usually as aresult of an operational problem or failure of QA) and, with the equipment available,
returning it to an appropriate store until a decision on future management is made.



Retrieval – recovery of some or all waste that has already been emplaced, which will
generally require special equipment and handling depending on the reason for retrieval.

Thus, reversal is a requirement that is built into the design process and is a consequence of arigorous QA system, while retrieval – which is always possible in principle – is assessed fromthe point of view of practicality with existing technology, at least up to the point of repository
closure. This section focuses only on ease of retrieval; reversibility will be assessed in moredetail after concepts are developed for specific sites. The ease with which waste retrieval can
4-151be carried out will, of course, depend on the reason for waste recovery and how far theemplacement activities have proceeded at that time.
4.7.1 Assuring practicality of retrieval
Issues associated with waste retrieval depend on the stage of repository operation, and it isinternationally recognised that the complexity and hazard of the process increases as the stageof operation progresses [155]. Thus, in order to examine the technical practicality of specificretrieval methods, it is necessary to first understand the state of the EBS at the time ofrecovery. The following summarises the results of investigations for H12V, PEM and TRU
waste.
(1) HLW(i) H12V implementation
Based on the simple sketch in Figure 4.2-6, the evolution of repository implementation wasclassified into three states, as shown in Figure 4.7-1 (specifically for H12V).

Figure 4.7-1 Stepwise increase in retrieval challenges (H12V)

Figure 4.7-1 is extremely simplistic, but illustrates general increase in difficulty of waste
package recovery as work progresses. In the simplest case A, the overpack and buffer are
emplaced, but the disposal tunnel is not backfilled. In B, the disposal tunnel has been
backfilled and the plug emplaced, although access to the disposal panel is still open. In C, allconnecting tunnels have been backfilled and additional plugs emplaced, so that only theaccess ramp/shafts are open.

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(ii) PEM implementation
For the PEM, the equivalent three states are shown in Figure 4.7-2. A is then the case where
PEM is placed in the disposal tunnel and ease of recovery is assured as no backfilling ispresent. In practice, it may be that backfilling would be easier done in a stepwise manner as
the PEMs are emplaced but, even in this case, recovery is relatively easy before the plug isemplaced.

A:
After PEM installation

Access ramp & shaft

Connecting
tunnelDisposal tunnel

PEM

Lifting frame

B:
Disposal tunnel
backfilling/constructionof mechanical plugs

Access ramp & shaft

Connecting
tunnelMechanical plug

Disposal tunnel

PEM

Lifting frame

C:
Connecting tunnel
backfilling andinstallation of
hydraulic plugs

After PEM
emplacement

Access ramp & shaft

Connecting
tunnelBentonite blocks
Mechanical Plug

Connection to Connecting
tunnelDisposal tunnel

PEM

Open
connecting
tunnel after
disposaltunnelbackfill andmechanicalplug
emplacement

Open accesstunnel afterconnecting
tunnelbackfillingandhydraulicplug
emplacement

Figure 4.7-2 Stepwise increase in retrieval challenges (PEM)

In B, the disposal tunnel is backfilled and the mechanical plug installed, but the open
connecting tunnel will ease retrieval. For C, the connecting tunnel is backfilled and thehydraulic plug emplaced, with only surface access open to aid retrieval.

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(2) TRU waste
Fundamentally the same three states can be seen for TRU waste (Figure 4.7-3): A after
waste emplacement and infilling/buffer emplacement but with vaults still open; B after vault
backfilling and plugging and C after backfilling and plugging connecting tunnels.

Access ramp & shaft

Surrounding tunnel

A:
Waste
emplacement
along with any
infill/backfill

Disposal vault

Disposalvault open

Connection to vault
Waste
Structuralconcrete

Connection
to vault

Surrounding
tunnelBufferAccess ramp & shaft
Surrounding tunnel

B:
Disposal vault
backfilled andmechanicalplugs emplaced

Disposal vault

Backfill

Connection to
vault

Mechanicalplug

Connection to
vault

Waste

Connecting
tunnelopen

Structural concrete
BufferMechanical plug

Access ramp & shaft

Surrounding tunnel
Hydraulic plug

C: Connecting
tunnelbackfilled andhydraulic plugs
emplaced

Disposal vault

Backfill

Connection
to vault

Mechanical
plug

Waste

Connection to
vault

Structural concrete

BufferMechanical plug

Figure 4.7-3 wise increase in retrieval challenges (TRU waste)

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Only
surface
accessopen

(3) Retrieval issues for different operational states
The following is a summary of the issues associated with retrieval for different operational
states and waste disposal options:

A state: PEMs can be retrieved by simple reversal of emplacement process, but
H12V and TRU waste need special equipment to retrieve waste packages fromholes/vaults.



B state: in all cases mechanical plugs and disposal tunnel/vault backfill must be
removed before the actions required in A.

C state: in all cases hydraulic plugs and connecting tunnel backfill must be removed
before the actions required in B.

Some key issues associated with such retrieval for EBS components and the surrounding
host rock are illustrated in Table 4.7-1 for the specific case of H12V. Here the assessment of
impacts focuses on consequences if operations are modified to extend the A, B or C cases tomaintain ease of retrieval, as opposed to the reference case of backfilling/plugging as quickly
as possible. An additional issue not yet assessed is what to do with recovered waste packages,
as their integrity may have been compromised. This will be a topic for future work.
Factors concerned with maintaining retrievability
Ease of retrieval

Impact to hostrock in the nearfield

Impact to EBS

Maintenance oftunnels

Case A

Case B

Case C

Number of steps for retrieval① Removal of buffer② Retrieval of overpack① Removal of mechanicalplug② Removal of backfill③ Remove of buffer④ Retrieval of overpack

Cost for retrieval

Low

Lowering of groundwater level

Wide range

Water velocity in near-field

Fast

Redox potential in near-field

Aerobic

Risk of piping erosion of buffer

High

Low

Risk of aerobic corrosion of
overpack

High

Low

Maintenance items

Disposal，surrounding,connecting and accessramp & shaft

Water drainage

Large

Small

Maintenance cost

High

Low

① Remove of hydraulic plug
② Removal of backfill in connecting andsurrounding tunnels③ Removal of mechanical plug
④ Removal of backfill⑤ Remove of buffer⑥ Retrieval of overpack
High
Small range
Slow
Anaerobic

Surrounding, connecting
and access ramp & shaft

access ramp & shaft

Table 4.7-1 Relative ease of retrieval and impacts for different states (H12V)

The complexity of the recovery procedure and the associated costs and hazards increase in
the order A < B < C as described above. The recovery tasks considered to require remotehandling from the viewpoint of radiation protection of workers are mainly associated withremoval of waste and the surrounding EBS components.
In terms of the influence of the near-field host rock, the extent to which the groundwatertable is lowered and water chemistry is changed depends on the rock and site conditions aswell as the period of time that tunnels have been open. Continuing drainage to maintain easeof recovery may affect hydrogeological characteristics and, in some cases, long-term safety
after closure, with the impact considered to increase in the order of C < B < A.
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With regard to the effects on engineered barriers, in the case of H12V and TRU waste
including a buffer, during state A buffer will take up water and swell, making it difficult tomaintain EBS quality. In addition, for state A, with time, loss of buffer due to piping erosion
and the amount of corrosion of the overpacks/waste packages by air both increase, making it
preferable to reduce the duration of this state to the extent possible. The influence becomeslarger in the order of C < B < A. In the case of the PEM, the carbon steel handling shellcorrodes when exposed to air but, as long as it maintains its integrity, the contained buffer
will not swell.
With regard to the impact on tunnel maintenance, if tunnels are kept open for long periodsof time, deterioration of the lining/support system as well as on the long term EDZ properties
have to be considered and regular inspection and maintenance will be needed (with potential
complications if this is in a radiation-controlled zone).
To assess ease of retrieval, it is necessary to evaluate such issues in a comprehensive
manner but, at the present time, focus is on the B state to illustrate how these might behandled.
4.7.2 Retrieval methods
The recovery method according to the different options in state B is considered, illustrating
the development status of the related recovery technology.
(1) HLW(i) H12V
The recovery process is initiated by removal of the mechanical plug, assumed to be madeof reinforced concrete (see Section 4.5.3 (2) (ii)), using heavy machinery such as a remotehandled hydraulic breaker. Then the backfill, assumed to be a bentonite-rock mixture (Section
4.5.3 (1) (ii)), can be removed by equipment such as a remotely operated backhoe.
Since overpack corrosion during the period of maintaining recoverability, i.e. in the order of
40 years according to current plans, can be confidently assumed to be limited, it is possible torecover the overpack intact. Here, when removing the buffer, care must be taken not todamage the overpack. Therefore, to avoid risks associated with over-coring [1], a techniqueusing brine flushing has been developed [156] (Figure 4.7-4) that removes buffer in the formof a slurry.
Liftingmechanism
Gripping device

Saline water jet and slurry vacuum device(a) Prototype of buffer removal device(b) Buffer removal test
Figure 4.7-4 Demonstration of buffer removal by brine flushing((a) [119], (b) [158])

With regard to this technology, tests have been conducted in international [157] anddomestic facilities [119] [158], including using a remote controlled, full-scale device, and4-156

hence basic practicality can be assured, even if more specific technology for this, coupled toassociated lifting of the overpack, would need further development before they could beapplied under repository conditions.
(ii) PEM
For the PEM system, removal of the mechanical plug is as for H12V, so focus is on
methods of removing backfill and recovering the PEM as a unit (details are given in
Supporting Report 4-75). The clearance between the PEM and the bedrock/liner is relatively
large, so when rock conditions allow, most of the backfill could be removed by a road header,
and then, so as not to damage the PEM handling shell, thereafter using brine flushing to
remove the rest (see Figure 4.7-5).
1. Mechanical removal of surrounding backfill

2. Removal of backfill around PEM by saline water jet

回収段階 Ⅱ 埋め戻し材除去

回収段階 Ⅱ 埋め戻し材除去

処分坑道
埋め戻し材除去/ロードヘッダー（埋め戻し材除去装置）

処分坑道
埋め戻し材除去/ジャンボ（埋め戻し材除去装置）

Figure 4.7-5 PEM backfill removal (Left: using road header; Right: brine flushing)After removing backfill, a recovery device will enclose and lift the PEM from its pedestal(Figure 4.7-6).
1. Movement of PEM retrieval device

2. Installing PEM retrieval device

回収段階 Ⅲ

回収段階 ⅢPEM 搬出

PEM 搬出

処分坑道
PEM引抜き / PEM 引抜き装置 + 牽引装置

処分坑道
PEM引抜き / PEM 引抜き装置 + 牽引装置

3. Jacking up PEM

4. Retrieval of PEM

回収段階 Ⅲ

回収段階 ⅢPEM 搬出

処分坑道
PEM引抜き / PEM 引抜き装置 + 牽引装置

PEM 搬出

処分坑道
PEM引抜き / PEM 引抜き装置 + 牽引装置

Figure 4.7-6 PEM recovery

This particular method allows PEM recovery as a unit, even if the mechanical strength of
the handling shell or lifting points cannot be assured. Also, even if not strictly required, thiscan be implemented by remote control to keep worker doses ALARA.
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(2) TRU waste
In order to carry out the recovery work safely and efficiently, it is desirable that waste
packages are structurally sound and that the function of containment is assured for both waste
packages A and B.
As illustrated in Figure 4.7-3, in state B the initial removal of a mechanical plug andoverlying backfill is effectively as for H12V. Options thereafter are either excavation of infill
from the top (assumed left-hand plug in Figure 4.7-3 removed) or from the side (assumed
right hand plug removal). The choice between these may depend on the type of waste package,as waste package A is designed to be lifted from below by a forklift, while waste package B islifted from the top by a crane – although it is not clear that the associated lifting features will
be accessible after infilling.
Since the waste packages contain, and are surrounded by, cementitious infill, any porewater
present will be highly alkaline, so the corrosion rate of the carbon steel waste package can be
confidently assumed to be very small (1 μm/y or less) [80], and therefore any degradation due
to corrosion is considered to be negligible. Currently, the remote-handled technology toremove the hardened infill without damaging the waste package and then extract the waste
package does not exist, so future technological development is necessary. From this point ofview, it may be useful to consider disposal concepts that do not fill gaps between waste
packages, as shown by ANDRA [111].

4.8 Summary and future perspective
4.8.1 Summary
Chapter 4 has outlined repository concepts and designs tailored to the three SDMs, based on
depths of 500 m or 1,000 m depending on rock strength. Key results are summarised below.
(1) Repository design method
Repository design is based on specific design requirements that ensure safety as well as
engineering practicality and economic rationality. By applying this method during stepwise
site selection, it is possible to iteratively refine repository design so that it can ensure therequired safety functions while flexibly responding to evolving geological environmental andsocio-political boundary conditions.
(2) HLW EBSThe HLW EBS components are taken over from H12 and, for the 3 reference SDMs,
assessed to check that all design requirements were met. In particular, as a result of work
carried out since the H12 report, the specified 190 mm thickness for the carbon steel overpack
has been shown to have the potential to assure its containment function for at least 17 ky after
closure, exceeding the design requirement of preventing contact of groundwater and HLW for
1 ky. However it should be noted that a more careful assessment will be needed in the future
before full credit for this containment time can be assured. Therefore, depending on thegeological environment of the site, it is considered possible to reduce overpack thickness inthe future. In addition, the reference buffer is known to have reduced swelling performance inhigh salinity groundwater, but specifications have been developed to ensure the required4-158

performance (e.g. assuring diffusive solute transport, acting as a colloid filter) for a relevant
range of salinities. Furthermore, because the buffer provides mechanical protection, its
stability has been demonstrated so that key safety functions (e.g. protecting the overpack fromearthquake impacts) can be assured for relevant timescales.
Of the disposal options, practical construction of the reference H12V EBS is feasible only
under conditions of low humidity and low water inflow into disposal holes. Otherwise,
countermeasures to prevent perturbations resulting in buffer density loss must be
implemented, which have not been fully developed as yet. For the PEM, EBS emplacement isless problematic and, although this has been less well studied, demonstration including in-situ
tests in URLs are ongoing in Japan and abroad. In the future, based on the conditions found at
actual sites, alternative disposal concepts or EBS construction methods may be considered.
(3) TRU waste EBS
Based on TRU-2, two waste package options have been considered. Both are made out ofsteel, with cementitious infill, but one is thinner-walled and without a lid (designed for liftingby a fork lift) while the other has thicker walls and a welded lid (designed for lifting by acrane). The latter also provides an assured containment time of about 300 years after closure.
The EBS also includes a buffer to reduce the release and migration of long half-life, mobile
RNs, such as I-129 and C-14, present in waste Grs.1 and 2. It was also decided to install abuffer due to the possibility of thermal deterioration of vault infill for relatively high thermal
output Gr.4H. The buffer material has long-term mechanical stability and, despite interactions
with hyperalkaline cementitious leachate, the specification ensures that the expected safetyfunctions can be maintained for a sufficiently long time.
(4) Design of underground facilities
For the design of the underground facilities, a method of laying out the disposal panels wasdeveloped to take into account the distribution of Layout Determining Features (LDFs) in the
3 SDMs and ensure that there are no problems with excavation of tunnels and construction
and QA of the EBS. Furthermore, preliminary Emplacement Determining Factors (EDFs)
have been suggested, but it is noted that these would be very host rock, site and repository
concept dependent and, thus, will be better specified in the future. The post-closure safety
aspects of such layouts will be assessed in Chapter 6 and, in the future, will provide feedback
that will be reflected in design refinement. In addition, design requirements were set so thatradiation protection and general occupational safety are assured by the layout of connecting
tunnels and access ramps/shafts. In particular, work flow lines and ventilation/drainage routes
were organised so that the construction and operational zones are separated, with independent
provision of key services.
After repository closure, it is necessary to prevent tunnels acting as short-circuit transfer
routes for RNs. For this purpose, hydraulic plugs are installed at key locations and tunnels
backfilled with a low permeability bentonite-crushed rock mixture.
Multiple safety measures to minimise probability and consequences of abnormal conditions(drops, fires, explosions, loss of external power and even CMF) were also developed, withparticular reference to experience overseas. In case of any perturbation that could lead to lossof the containment function, measures are developed to prevent radiation releases outside of
the facility in either ventilation or drainage. These safety measures will become more specific4-159

as detailed design of underground facilities for the geological conditions at actual sites
proceeds.
Required construction technology is generally available and proven in major projects, suchas the construction of underground power plants. The technology for backfilling and plugconstruction is currently at the stage of equipment development, with demonstration tests atfull scale, using underground research institutes, already well advanced in Japan and overseas.
(5) Design of surface facilities
In this report, transportation of waste to surface facilities is considered in order to evaluate
logistical constraints on waste handling. There is wide experience in Japan of suchtransportation and, for a coastal location, marine transportation of both HLW and TRU waste
has demonstrated safety and allows for efficient movement of the relatively large inventories
of waste involved.
Design to date has focused on the most sensitive facilities, which directly handle wastefrom the point of its receipt to dispatch underground for disposal. Standard nuclear industry
guidelines ensure radiation protection of workers and the general public from all normal
operations and credible perturbations. In the design, selecting locations for the installation ofsuch facilities and their structural design is based on safety measures to reduce risks fromnatural perturbating phenomena.
For receipt, inspection, and encapsulation/packaging facilities for HLW and TRU waste,
design includes specification of wall thicknesses to provide sufficient radiation shielding andsafety measures to ensure containment of RNs in case of perturbations. It has been shownhow multiple safety measures can prevent the loss of containment even in the abnormal
conditions (drops, fires, explosions, loss of external power and CMF), complementing
working under reduced pressure with careful management of ventilation and drainage. These
safety measures will become more specific as the detailed designs for the geological andsocial conditions of specific sites proceeds.
(6) Assuring retrievability
Maintaining the option to easily retrieve waste after emplacement tends to conflict withother requirements to assure both pre- and post-closure safety (and also reduction of costs andenvironmental impact), so the trade-off of increasing effort for retrieval has to be accepted toallow stepwise stabilisation and isolation of disposal and connecting tunnels. The greatest
challenges are associated with the remote-handled removal of the waste from disposal
tunnels/vaults. The practicality of HLW recovery has been confirmed by technology
demonstration tests at large/full scale and in-situ in URLs. Although it is considered possible
to recover TRU waste using existing civil engineering technology, detailed design anddemonstration of practicality will be needed in the future.
4.8.2 Issues for specific geological environments
In addition to the generic points above, specific points for the three representative hostrocks considered and those for the special case of disposal under the coastal seabed aresummarised below.
4-160(1) Plutonic rocks
For plutonic rocks, it is particularly important to note that the frequency of water bearing
fractures in the reference SDM is high and the amount of inflow into tunnels could be greaterthan for the other host rocks. In particular for the H12V option, it is considered that humidity
control in disposal holes during installation of buffer and overpack will be difficult. In thiscase, it is also necessary to decide if disposal holes are practically unusable. As a result, theproportion of usable disposal holes may decrease, thus requiring more reserve volume thanfor the other representative host rocks.
The strength of the rock is good from the viewpoint of assuring mechanical stability oftunnels and the repository can be located at greater depth than in the case of softer rocks.
However, at greater depths, in order to provide suitable working temperatures, it may benecessary to take operational measures such as cooling ventilation air.
(2) Neogene sediments
Neogene sediments have relatively low strength and shallower disposal/more extensive
tunnel support may be required compared to stronger rocks. For H12V, in order to assurestability, a larger waste emplacement pitch is needed compared to the other host rocks andthe required footprint is much larger than that for the PEM option. In addition, since the
TRU waste vaults have a relatively large cross section, a complex excavation/support
process is required in order to assure the required mechanical stability.
Since some Neogene sediments are considered to have a relatively high risk of containing
methane, as noted in Section 4.5.6, special protection measures in terms of monitoring,
ventilation and explosion-proofing of equipment may be needed.
The frequency of fissures and faults in the SDM is relatively small, as is the amount ofgroundwater inflow, so quality control of emplaced buffer and materials is relatively easy.
(3) Pre-Neogene sediments
The characteristics of the Pre-Neogene sediments are close to those of plutonic rocks andhence the comments in (1) above apply here. In addition, in the case of Pre-Neogene
accretionary complexes, to design a repository layout in accordance with geologicalstructures such as folding and LDFs, a DET layout was chosen to provide more flexibility tochange the length of disposal tunnels compared to TT layouts.
(4) Sub-sea disposal
Discussion in this chapter does not explicitly specify whether the disposal footprint lies
under land or sea, although the hydraulic gradients derived from the SDM imply that it isunder land. Nevertheless, for preferred coastal locations specified in the Nationwide Map,
there may be advantages in sub-sea disposal.
Sites located below the seabed but accessed from land are not precluded by international
agreements constraining use of the sea and have the potential advantage of very low
hydraulic gradients and also being less affected by future climate change. The layout of therepository shown in this chapter assumes access by shafts and a ramp, which may beproblematic for coastal/subsea disposal. In particular, for deeper water without any islands,
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shafts over the disposal footprint may be impossible and all access may be by ramps fromland. For this reason, access ramps will be longer and take more time to excavate, while
ventilation and drainage capacities will also increase. For a coastal site, therefore, tailored
concepts for both onshore and offshore variants may be needed and a more detailed
assessment methodology developed to assess their relative pros and cons. This is identified
as a priority area for future R&D.
4.8.3 Future perspectives
A practical design technology for constructing a repository with the required safetyfunctions has been illustrated and the specifications of the EBS and other major structures
have been prepared. Proposed future efforts to increase safety and cost-effectiveness of therepository, based on assured practical, quality-controlled procedures and technology aresummarised in Table 4.8-1, organised from consideration of the EBS, surface andunderground facilities, and retrievability.
Table 4.8-1 Future major efforts related to engineering technology
Area

Main topics




EBS



Alternative EBS materials and design options
Improvement of the containment function of the TRU waste EBS
Development and demonstration of technology to evaluate
evolution of the EBS
Standardisation of EBS design methods and material property tests
Development and demonstration of EBS fabrication techniques andemplacement technology

Surface andundergroundfacilities

Development of holistic design technology for the entire repository
Development of technology to ensure safety during construction
Development of repository sealing/closure technology

Retrievability

Development and demonstration of waste recovery technology
Development of impact assessment technology to evaluate issues
arising from maintaining ease of waste recovery(1) EBS(i) Alternative EBS materials and design options
In Section 4.4.1, the HLW EBS comprising a steel overpack and a bentonite buffer was
shown to meet design requirements for both fresh and saline groundwaters. Here, the carbon
steel overpack assumes manufacture by forging, but cast steel could also be considered fromthe viewpoint of economics. Cost reduction may also be achievable by a thinner steel
overpack based on the more realistic assessment of the corrosion allowance, which may also
lead to a reduction of the volume of bentonite buffer, since the diameter of the overpack can
be decreased. This cost reduction may be more significant in the case of the PEM, since thesize of metal shell surrounding the bentonite backfill can be reduced.
It is also recognised that further R&D may allow alternative overpack materials or designs
to be considered, for example including protective coatings (e.g. thick copper electroplating[159]), with the aim of assuring the containment function for a wider range of conditions,4-162

while also reducing of costs. With regard to the bentonite used for the buffer, work to date has
focused on Kunigel V1 Na-bentonite. However, from the viewpoints of economics andprocurement diversity, it is necessary to confirm the applicability of other bentonite materials.
Thus, various alternative EBS materials and design modifications to allow tailoring to thegeological environments of actual sites will be evaluated.
In addition to safety, in the future it will also be important to develop design options taking
into consideration the efficiency of operations and the ease of recovery. For example,
although the overpack lid has a flat plate structure in the current specification of the overpack,a hemispherical structure [159] [160] can alleviate stress concentration and improve pressure
resistance.
From the viewpoint of operational efficiency, the PEM option has clear advantages but, as
this method is relatively new, further assessment of its design would be useful to determine
how performance can be better quantified and details modified to tailor to specific geologicalenvironments. Indeed, from the lessons learned in this chapter there seems to be potential toapply the PEM concept to TRU waste as a new design option.
(ii) Improvement of the TRU waste EBS containment function
In Section 4.4.2, TRU waste EBS specifications were developed, which were shown tomeet operational safety requirements under normal conditions and also, for a more robust
waste package, ensure containment of RNs for several hundred years after closure. In order toensure the containment performance of such design options, however, it is necessary tofurther confirm the robustness of the waste package under abnormal conditions (e.g. by drop
tests, fire resistance tests, etc.). In addition, further assessment of issues required forevaluation of long-term behaviour after closure, such as stress corrosion cracking in welds
and increases in internal gas pressure, will be carried out with the goal of strengtheningarguments for long-term containment by the waste package as a key part of the safety case.
Furthermore, from the viewpoint of further improvement of TRU post-closure performance,
ongoing R&D has focused on improved barriers for mobile, highly soluble radionuclides suchas I-129 (which contributes greatly to doses), both in terms of better immobilisation within
the waste package and higher performance buffers/backfills (e.g. incorporating anion
exchangers or getters).
(iii) Development and demonstration of technology to evaluate long-term
behaviour of the EBS
Most of the material property tests so far have been conducted for short periods, up to
several years; in the future, however, longer-term tests to refine models of behaviour that will
reduce uncertainty in temporal extrapolations will be included. For example, since thecorrosion rate tends to decrease as the test period becomes longer, corrosion tests lasting morethan 10 years in relevant environments could greatly improve understanding and allow more
realistic assessment of overpack/waste package lifetimes, serving as a basis for refining theirdesigns. In addition, a further aim is to advance the understanding of interactions within thenear field and evolution of the EBS, including impacts of the early THMC transients. Thus,
system evolution tests will be carried out, for example on the PEM and the different TRU
waste packages, both in the laboratory and in URLs. The acquired data will be used forevaluation of the post-closure performance of the EBS and verification or improvement of the
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associated models. However, for the complex coupled phenomena occurring in the EBS, it isnecessary to monitor behaviour of individual components and the interfaces between them in
a comprehensive manner. Thus, to maximise benefits of real-scale tests using simulated waste
in underground environments, it is necessary to develop a new generation of high-precision
sensors with long-term durability under in situ conditions.
(iv) Standardisation of EBS design methods and material property tests
In order to ensure the compatibility of technology and consistency in the design andconstruction of the EBS, an aim will be to improve the standardisation of both existingmaterial property test methods for engineered barriers and also new methodology that will be
developed in the future. This will lead also to standardisation of design methods (evaluation
methods for design requirements, etc.) and the management of data used by these. A useful
test case for such standardisation will be the planning, implementation and interpretation oflarge-scale, in-situ demonstration tests.
(v) Development and demonstration of quality-assured EBS fabrication
techniques and emplacement technology
Extensive development of the fundamental technology for overpack and buffer construction
has been carried out in Japan. However, in the future, targeted development will be carriedout to rationalise and test in full-scale demonstrations the technology for:

Waste package production and remote inspection.

Production and construction of the buffer.

PEM fabrication.

Production and emplacement of TRU vault infill.



Waste transport and emplacement, together with other EBS components, using
remote control or automated methods.

In terms of the manufacturing technology of overpacks, it is important to promote
development of lid bonding technology as well as manufacturing technology for alternative
materials. As welding technology, the applicability of TIG, MAG and electron beam welding
has been confirmed at full scale for steel overpacks, so it is necessary to now focus onrationalisation, such as shortening the welding time. Nevertheless, developments in suchtechnology will be monitored to check if any new developments would provide benefits forJapanese boundary conditions (e.g. friction stir welding, alternatives to welding). From theviewpoint of preventing stress corrosion cracking, it is important to reduce the residual stressin the weld zone, and hence work on the practical application of techniques such as heat
treatment after welding will be carried out. With regard to inspection techniques for welds,
the applicable techniques may be constrained by the radiation environment and therequirement that these are suitable for remote controlled operation. In general, these welding
and inspection techniques would be common for both HLW overpacks and TRU waste
packages. For all developed technologies and associated tests, strict quality control methods
will be established that include specifications, inspection standards and definition of
responses when specifications are not met.

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The PEM construction specification leads to a weight of about 37 Mg (see Section 4.4.3 (1)(ii)) and thus it is necessary to rationalise transport/emplacement techniques in order to assure
efficiency while also providing all pre- and post-closure safety functions. In addition, since
the PEM handling shell is an additional EBS component, examination of potential barrier
roles (e.g. providing containment for a period before it fails, reduction of uncertainty withrespect to buffer saturation during a THMC transient) and possibly increased systemcomplexity (shell-backfill and shell-buffer interfaces) is important. Further, as noted in
Section 4.4.3 (1) (ii), although PEM emplacement is relatively insensitive to water inflow, it
is necessary to demonstrate the practicality of backfilling the gap between the PEM and thedisposal tunnel wall surface to required quality specifications under realistic environmental
conditions. Development of H12V construction technology is supported by basic research,
including a full-scale demonstration in a URL. In the future, in addition to tests of alternative
bentonite materials and establishing quality control methods, it will be necessary to develop apractical moisture control technology in order to improve applicability in a wider range of
emplacement environments.
Technology for manufacturing TRU waste packages and their installation, together withassociated infill, is being further developed with the aim of improving performance in termsof containment after emplacement and ease of retrieval. A full-scale demonstration study will
support development and testing of EBS specifications and establishing quality control
methods for manufacturing and construction technology.
Waste handling equipment is a focus for advanced technological development, including
specific, fail-safe designs for gripping, positioning and placement for the different waste
options, both above and below ground. Again here, it is important to carry out demonstration
tests using simulated waste in surface and underground research facilities to support
development of prototypes and establishing quality control methods. Through these tests, it isaimed to improve practicality and efficiency by the second part of the PI stage. Additional
tests will be needed in a site-specific UIF within the DI stage, to confirm that the planned
handling and installation methods would be practical. The location of the UIF relative to theemplacement panels also needs to be considered in future design work. In addition, for all
equipment related to EBS manufacture and emplacement, it will essential to establish remote
operation and/or automation technology to meet radiation protection goals.
(2) Surface and underground facilities(i) Development of holistic design technology for the entire repository
The underground facilities are very large and complicated by parallel construction andoperation, with the requirement to assure independent ventilation and drainage of these twozones, including re-classifying non-controlled and radiation-controlled zones as the projectprogresses. In addition, due to the long operational time, a programme of inspection,
maintenance and, possibly, refurbishment will be needed. Clearly, there is a substantial
knowledge base that could be drawn from the design of other nuclear facilities, tunnels or
mines, but some issues will be unique to a repository, especially considering the potential for
unknown siting environments.
In Sections 4.5.4 and 4.5.5, outline designs for ventilation and drainage systems were
presented but, in the future, these will need to be refined based on site-specific assessments oflikely groundwater inflow (including abnormal flows as a result of perturbations) and4-165

ventilation needs for both construction and operation, with special consideration of
responding to fires.
Approaches to meet the design requirements for specific tunnels, shown in Section 4.5.2,
and for the layout of the repository, shown in Section 4.5.4, will need to be assessed and, if
required, modified based on experience gained from underground demonstration tests. This
will be complemented by continuous maintenance and updating the information/knowledge
base that is the foundation for judging suitability, developing specifications and setting
standards. When the geological environments of real sites are defined in the future, specificrisks of potential perturbations impacting operational safety can be identified and, based on
analysis of past accident cases and experience in other nuclear facilities, countermeasure canbe developed that are reflected in the design of both surface and underground facilities.
(ii) Development of technology to ensure safe repository construction
It is important to develop excavation techniques for disposal tunnels and vaults (plusdisposal holes if included) that minimise the perturbation of surrounding rock while ensuringsafety and cost-effectiveness. Section 4.5.7 (1) assessed excavation techniques, indicating thatthose applied in general tunnel construction would be suitable. However, with the goal of
improving safety and efficiency, it is intended to study tele-operation and/or automation of
excavation technology. Although relevant knowledge is being developed through tests aroundtunnels in deep URLs, there is a lack of knowledge about the long-term evolution of hydraulic,chemical and mechanical characteristics of the EDZ for different host rocks and hence more
extensive study in this area is required.
As shown in the examination of required reserve areas in Section 4.5.4 (5), judgement of
whether or not waste should be placed in specific locations depends on possible
countermeasures to undesirable conditions, particularly groundwater inflow. Managing
groundwater inflow is also very important in order to minimise the perturbation of local
hydrogeological conditions and reducing the cost of wastewater treatment. Thus, future work
will examine counter measures such as grouting with cementitious material, which is used ingeneral tunnel construction and its application has been confirmed under relevant highgroundwater pressure in Japanese URLs [122] [123]. However, in the future, reducing thechemical impact of grout will be examined, either by reducing the amount used to the extent
possible or by selecting alternative grouts. In this regard, it is important to recognise thatgrouting actions need to be considered in the context of networks of tunnels in a disposal
panel, where chemical interactions between them can be significant and need to be assessedby a panel-scale hydrogeological model, and hence improvements in assessment methodology
are required to guide grout optimisation approaches.
In terms of relevant knowledge capture, advances at the leading edge of tunnelling
technology will be followed. For example, in 2015, construction for the Chuo Shinkansen in
the Japanese Southern Alps involved one of the world's longest mountain tunnels (about 25km), with an overburden of up to 1,400 m. The construction technology required andgroundwater management countermeasures utilised may well fit repository design
requirements and hence developments in this and other similar projects will be kept underreview.

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(iii) Development of repository sealing/closure technology
With regard to sealing technology, including plugs and backfill to prevent tunnels fromacting as hydraulic short-circuits, evaluation is required of their role in the performance of theentire repository for a range of relevant geological environments, which will allow further
refinement of the design concepts outlined in Section 4.5.3 and assessment of theirpracticality. In terms of the performance of the tunnel seals, individual components of thehydraulic plug and backfill, as well as combinations of these, will be assessed in terms ofimpacts on post-closure repository performance under realistic geological settings. With
regard to construction technology, based on the design concepts examined so far and theresults of research in other countries, required technology will be developed that is capable offlexibly responding to diverse geological environments and disposal concepts. This
technology will be tested in full-scale experiments in surface and URLs in order to develop
detailed design specifications, which will also consider countermeasures against possible
perturbations and required quality assurance.
(3) Retrievability(i) Development and demonstration of waste recovery technology
In order to assure practicality of retrieval before final repository closure, it will be necessary
to develop and test the required technology. As described in Section 4.7.2 (1) (i), development
and testing of such technology for the H12V option is in progress. In the future, it will benecessary to confirm its applicability to the PEM option, noting what components can betaken over from H12V and what modifications/new technology would be needed. In addition,
retrieval technology is required for the more complex and less-studied TRU waste case,requiring highly flexible and reliable methods that could ideally be tailored to the different
waste groups and package types under consideration. In all cases, practicality needs to betested in large-scale demonstration experiments that simulate expected conditions for retrieval
at different times after emplacement and establish rigorous quality guidelines that can form abasis for later consideration of remote control/automation for actual implementation.
(ii) Development of impact assessment technology to evaluate issues arisingfrom maintaining ease of waste recovery
There can be clear conflicts between the desire to maintain ease of recovery (favouring
keeping tunnels open for as long as possible) and assuring operational and post-closure safety,
minimising environmental impacts and reducing costs (favouring closing and sealing tunnels
as quickly as possible). In Section 4.1.1, the requirement of maintaining ease of retrieval waspresented only in rather general terms, along with the need to avoid compromising theperformance of engineered barriers. In the future, it will be necessary to extend thisassessment to include quantitative analysis of the pros and cons of different options. As these
are likely to be site-specific, initial assessment will examine the impacts of maintaining easeof retrieval on safety and environmental impact for different boundary conditions, with a
special focus on the hydrogeological characteristics of the host rock.

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Supporting Reports (SRs)
SR 4-1
SR 4-2SR 4-3SR 4-4SR 4-5
SR 4-6
SR 4-7

Overview of analysis codes used for design and pre-closure safety assessment
Set values of the geological environment characteristics used for repository design
Setting of repository depth
Overpack design requirements and evaluation items
Setting minimum overpack thickness
Setting required corrosion allowance for overpack (initial oxidising environment)
Setting of required corrosion allowance for overpack (long-term reducing
environment)
SR 4-8 Evaluation of stress corrosion cracking and hydrogen embrittlement cracking of
overpacks
SR 4-9 Assessment of overpack microbial corrosion
SR 4-10 Setting of the thickness of the overpack for mechanical stability
SR 4-11 Setting the overpack shielding margin
SR 4-12 Assessing the impact of overpack weld defects on structural integrity
SR 4-13 Overpack longevity assessment
SR 4-14 Design of buffer and setting of evaluation items (HLW)
SR 4-15 Technical specifications for HLW buffer
SR 4-16 Technical assessment of compaction properties of HLW buffer
SR 4-17 Evaluation of long-term soundness of HLW buffer
SR 4-18 Calculation of surface and surrounding air dose of overpack and PEM
SR 4-19 PEM handling shell design
SR 4-20 Setting of TRU waste package design requirements and evaluation items
SR 4-21 Setting specifications of TRU waste packages
SR 4-22 Structural integrity evaluation during waste package operations
SR 4-23 Long-term corrosion resistance of TRU waste package B
SR 4-24 Hydrostatic pressure impact assessment for TRU waste package B
SR 4-25 Design of infill between TRU waste packages
SR 4-26 Design of buffer and setting of evaluation items (TRU waste)
SR 4-27 Technical specifications for the TRU waste buffer
SR 4-28 Evaluation of long-term soundness of the TRU waste buffer
SR 4-29 Fabrication and emplacement technology for buffer blocks
SR 4-30 Horizontal PEM: method of emplacement
SR 4-31 EBS construction technology for TRU waste disposal facilities
SR 4-32 Initial concept for design of tunnel liners
SR 4-33 Technical basis for setting disposal tunnel/hole pitch
SR 4-34 Evaluation of HLW tunnel stability
SR 4-35 Design of the disposal cell for TRU waste
SR 4-36 Setting of the TRU waste disposal tunnel cross-section
SR 4-37 Evaluation of TRU waste cavern stability
SR 4-38 Cross section design of other tunnels
SR 4-39 Impact of waste heat on HLW disposal tunnel pitch (H12V)
SR 4-40 Impact of waste heat on HLW disposal tunnel pitch (PEM)
SR 4-41 Impact of waste heat on TRU waste disposal vault pitch
SR 4-42 Setting disposal vault/panel geometry (TRU waste)
SR 4-43 Backfill material specifications
SR 4-44 Setting of the excavation damaged areas for backfill plug design
SR 4-45 Evaluation of long-term soundness of disposal tunnel backfill
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SR 4-46 Technical rationale of mechanical plug design
SR 4-47 Technical rationale for hydraulic plug design
SR 4-48 Impact of geological features on layout of the underground facility
SR 4-49 Groundwater flow analysis to support selection of disposal panel locations
SR 4-50 Setting of approach to disposal tunnel (HLW)
SR 4-51 Setting disposal panel dimensions (HLW)
SR 4-52 Setting disposal panel layout (HLW)
SR 4-53 Design of access to TRU waste disposal vaults
SR 4-54 Orientation of disposal panels
SR 4-55 Setting of piping erosion evaluation period
SR 4-56 Evaluation of piping erosion of buffer
SR 4-57 Evaluation of disposal hole utilisation efficiency (H12V)
SR 4-58 Evaluation of disposal tunnel utilisation efficiency (PEM)
SR 4-59 Evaluation of disposal vault utilisation efficiency (TRU)
SR 4-60 Study of work flow and ventilation paths (plutonic rock/HLW)
SR 4-61 Study of work flow and ventilation paths (Neogene sediments/HLW)
SR 4-62 Estimates of excavated spoil volumes (HLW)
SR 4-63 Underground facility ventilation and cooling
SR 4-64 Evaluation of groundwater inflow volume
SR 4-65 Relevant regulations for abnormal conditions assumed during design
SR 4-66 Conceptual designs of underground waste transport vehicles
SR 4-67 Repository excavation technology
SR 4-68 Safety measures during transportation to the repository
SR 4-69 Outline of HLW reception, inspection, and encapsulation facility (H12V)
SR 4-70 Design of connections between the surface facility and access tunnels
SR 4-71 Overview of TRU waste reception, inspection and packaging facilities
SR 4-72 Shielding design for the HLW reception, inspection and encapsulation facilities
SR 4-73 Shielding design for the TRU waste reception, inspection and packaging facilities
SR 4-74 Examination of safety measures for surface facilities
SR 4-75 Engineering feasibility of PEM recovery

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Specific Safety Requirements, IAEA Safety Standards Series, No. SSR-6.
[144] Advisory Committee for Natural Resources and Energy (2017): Results of theexamination of the requirements and criteria for the presentation of regional scientificcharacteristics of geological disposal (summarised by the Geological Disposal Technology
WG), Geological Disposal Technology WG, Nuclear Energy Subcommittee, Electricity and
Gas Industry Committee, Advisory Committee for Natural Resources and Energy (inJapanese).
[145] NFT (Nuclear Fuel Transport) (2013): A key player in the nuclear fuel cycle, NFT
Pamphlet (2013) (in Japanese).
[146] JNFL (Japan Nuclear Fuel Limited) (2014): The application form for permission tochange the waste management business at the reprocessing plant, 2013 Reprocessing Planning
Issue (SAIKEIHATSU) No. 507 (in Japanese).
[147] JAEC (Japan Atomic Energy Commission) (2007): Basic concepts of the safeguards ofvitrified high-level radioactive waste, Advisory Committee on Nuclear Security, JAEC (inJapanese).
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security, http://www.fepc.or.jp/enterprise/supply/security/ (accessed 2020-12-01) (inJapanese).
[149] NFT (Nuclear Fuel Transport): Spent fuel transport vessels,
http://www.nft.co.jp/equipment/equipment1.html (accessed 2020-12-01) (in Japanese).
[150] Kitagawa, Y., Suzuki, S., Kubota, S., Tada, H., Nakashima, H., Toguri, S., Tamai, M.,
Nakahara, Y., Kawakami, S., Iwata, Y., Tochigi Y., Yahagi, R. and Asano, H. (2013): Study on
the basic style of the PEM system (Part 3: Study of the operation system), The 68th Annual
Academic Lectures of the Japan Society of Civil Engineers, CS11-057.
[151] CRIEPI (Central Research Institute of Electric Power Industry) (1990): Drop tests forassessing the integrity of vitrified waste, Research material of CRIEPI, U90904 (in Japanese).
[152] Mano, T. and Otaka, H. (1991): Drop tests of vitrified waste, PNC TN1410 91-035 (inJapanese).
[153] CRIEPI (Central Research Institute of Electric Power Industry) (1980): Safety
demonstration test for dropping waste package (in Japanese).
[154] Advisory Committee for Natural Resources and Energy (2008): Study report on thesafety of consolidated waste packages (CSD-C) among returned low-level radioactive waste,
Waste Material Safety Subcommittee, Nuclear and Industrial Safety Subcommittee, Advisory
Committee for Natural Resources and Energy (in Japanese).
[155] OECD/NEA (2011): Reversibility and Retrievability (R&R) for the deep disposal ofhigh-level radioactive waste and spent fuel, Final Report of the NEA R&R Project (20072011).
[156] RWMC (Radioactive Waste Management Funding and Research Center) (2014):
RWMC FY2013 Annual Technical Report (in Japanese).
[157] Eng, A. (2008): Äspӧ Hard Rock Laboratory, Canister Retrieval Test, Retrieval phase,
Project report, IPR-08-13.
[158] RWMC (Radioactive Waste Management Funding and Research Center) (2015): Reporton advances and developments in retrieval technologies for geological disposal, FY2014
geological disposal technology survey project (in Japanese).
[159] Keech, P. G., Vo, P., Ramamurthy, S., Chen, J., Jacklin, R. and Shoesmith, D. W. (2014):
Design and development of copper coatings for long term storage of used nuclear fuel,Corrosion Engineering, Science and Technology, Vol. 49, No. 6, pp. 425-430.
[160] Nagra (2002): Project Opalinus Clay: Safety, report, Demonstration of disposal
feasibility for spent fuel, vitrified high-level waste and long-lived intermediate-level waste,
Nagra NTB 02-05.
4-182

5 OPERATIONAL SAFETY ASSESSMENT
Based on the strategy for operational safety assessment described in Section 2.4.1, thischapter will:

Develop the strategy and methodology for assessment of the radiological impact to the
public in the vicinity during operations, taking into account similarity to other facilities
handling HLW and TRU waste and existing safety regulations for nuclear facilities.
Preliminary assessment of operational safety for current design specifications of thedisposal facility.
Identify key findings and future perspectives.

Section 5.1 describes the strategy and methodology for operational safety assessment.
Section 5.2 describes the operational processes, waste characteristics and throughput asbackground for this assessment. Sections 5.3 and 5.4 describe assessment results of normal
and abnormal operational scenarios, respectively. Based on these assessment results, Sections
5.5 and 5.6 describe an approach to post-accident responses, together with a summary and an
outline of future work.

5.1 Strategy for operational safety assessment
As described in Section 2.1.2 (2), operational safety assessment covers both radiation
protection and conventional hazards. This chapter will focus on potential radiological impacts
to the public in the vicinity of the repository, based on the design and operational plan
presented in Chapter 4. As also described in chapter 4, the design measures and scenarios
described as event-trees are based on the defence-in-depth principle, which also means thatthe output of the operational safety assessment is a key source for further improving design.
Since safety regulations have not yet been developed for geological disposal facilities, thereference abnormal states assessed in this chapter were selected, and the assessment procedure
determined, by consideration of similarities to other facilities handling radioactive waste andassociated regulations, together with the safety case methodology and safety assessment guide[1] issued by the IAEA. Here it should be noted that lessons learned from the TEPCO
Fukushima Daiichi Nuclear Power Station accident led to development of new safety
regulations for nuclear reactors that have already been enforced and which emphasise
resilience, even in the event of potential accidents with low probability of occurrence.
The operational safety assessment will be performed for operational scenarios establishedby reviewing the geological disposal facilities and safety measures described in Chapter 4.
Assessment of conventional safety was not included in this Chapter, as it requires detailed
designs which will only be developed at later stages of the programme. Furthermore, safetyissues related to underground construction and operation could in the future, to a large extent,
rely on existing mining and tunnelling practice. Nevertheless, measures contributing toconventional safety, such as maintaining good work conditions, are summarised in Table 2.1-4
and described in more detail in Sections 4.5.5 and 4.5.6.
5.1.1 Procedures for developing operational safety assessment scenarios
Scenarios were developed for both normal and abnormal operations by referring to the
5-1

IAEA guide on safety case methodology and safety assessment [1]. The guide states that thesafety case must include measures to mitigate any risks to workers and the public in thevicinity of the repository. Such safety measures are described in Section 4.5.6 for undergroundfacilities and Section 4.6.2 (6) for surface facilities. The strategy for developing relevant
scenarios is described below.
(1) Procedure for developing normal operation scenarios
These scenarios describe the planned activities in the facilities during normal operation, as
outlined in Chapter 4, aiming to assess radiological consequences to the public when thesafety functions of radiation shielding and containment during operation (see Tables 4.2-1 and4.2-2) work normally.
Normal operational scenarios are developed taking into account the following:

As described in Section 4.6.2 (5), shielding by the walls of radiation-controlled areas
should ensure protection of workers, when combined with other radiation control
methods that limit and monitor exposure to radiation.
For HLW, the containment of radionuclides (RNs) is ensured by the glass matrix andthe welded stainless steel canister before encapsulation, with the additional overpackand transport cask after encapsulation as described in 4.6.2 (5) (i). Additional
containment would be provided by the PEM shell and buffer for the PEM case.
These constraints are basically similar for TRU waste, as shown in section 4.6.2 (5) (ii).

The following issues are excluded from the normal operation scenarios considered in thisreport:Internal radiation exposure of the public around the facility: RNs are assured to be
confined in the specific zones handling radioactive waste, as shown in the design of thewaste reception/inspection/encapsulation facilities in Section 4.6.2.
Radiological effects from underground facilities during normal operation: these aresufficiently shielded by bedrock that no exposure to the public is possible.
Total doses to radiation workers: as shown in Sections 4.5.6 and 4.6.2, control of accessto zones containing radioactive material, remote-handling methodology, shielding andother radiation control is assumed to assure safety.

(2) Procedures for developing abnormal operation scenarios
Abnormal operation scenarios are developed to assess the risk of RN release from thefacility due to perturbations of planned activities. To assess how the safety measures describedin Sections 4.5.6 and 4.6.2 (6) could be perturbed, the sequential transition from normal
operation is captured in an event tree diagram. The event tree in which all safety measures failis selected as a worst-case abnormal operation scenario.
As show in Figure 4.2-7, an abnormal operation scenario starts from initiating events, suchas external hazards (natural events and human-induced events) or internal initiating events(drop, internal fire, explosion, loss of off-site power and other system failures). Scenarios interms of external hazards involve seismic impacts, tsunami impacts, damage to off-site power,
forest fire, electromagnetic impacts, and other impacts on the facilities due to external hazards(see also Table 4.6-1). These impacts may also lead to internal initiating events [2], while themagnitude of impacts may strongly depend on the site environment. In the following section,
it is assumed that internal initiating events occur independently of the external hazards, and5-2

therefore the scenario will start from the internal initiating events in a site-generic andconservative manner. This approach also covers scenarios due to internal initiating events
which are not caused by external hazards
The following points were considered when creating the event trees used to develop theabnormal operation scenarios:






The event trees illustrate the process where assumed initiating events lead to abnormal
states (e.g. drop or fire) that result in mechanical or thermal impacts on waste.
Since the aim is to assess the possibility of RN release from the facility as a result ofaccidents, the event tree does not include measures to mitigate the effects and accident
management after loss of containment.
Safety culture and education of workers are important safety measures. However, theevent tree does not include such measures explicitly, although their failure may be aroot cause of perturbations leading to abnormal operation, while their contribution to
detection of failures leads to identification of disturbances and associated countermeasures to reduce impacts and allow recovery of normal operations.
Although multiple perturbations may occur simultaneously (common mode failure), forassessment of the direct effectiveness of counter-measures discussed in Sections 4.5.6
and 4.6.2 (6), this report does not consider these, but they are highlighted in Section
5.6.2 (1) as a future priority.

The developed abnormal operation scenarios are classified into five groups: drops, fires,
explosions, loss of external power, and other equipment failure. As discussed in Section 5.5,
this list may not be complete. For example, scenarios resulting from other external events or
impacts of irregular waste packages (failure of production quality control) may be consideredin the future.
5.1.2 Procedures for assessment(1) Procedures for normal operation scenario assessment
As described in Section 4.6.2, this section assesses the radiation exposure to the generalpublic from the reception/inspection/encapsulation facility for HLW and TRU waste. Once a
repository site is selected, the site boundary will be set based on consideration of local
geological, geographical and environmental conditions. Thus, assessment in this reportquantifies the annual effective dose, taking the distance from the facility to the boundary as avariable.
To assess the radiation shielding function, direct irradiation and skyshine from the
HLW/TRU waste handling facility shown in Section 4.6.2 is quantified, using the 50 μSv/ynuclear facility regulatory dose limit to the general public and determining the distance atwhich this limit is reached.
(2) Procedures for abnormal operation scenario assessment
Since the RN release from the facility is limited to cases where HLW and its
canister/overpack/transport cask, as appropriate, are significantly damaged (in the case of
TRU waste, the waste matrix, primary container, waste package and transport cask), an
evaluation of the robustness of these is included in the assessment. If damage did occur, boththe characteristics and concentrations of RNs released could be used to estimate the5-3radiological consequences. However, such assessment requires topographic, climatic andlocal lifestyle information, which can be determined only on a site-specific basis. For thisreason, this report assesses only the risks of damage to waste form that could lead to RN
releases.
A conservative approach is taken in which, before analysis, a key indicator of impact(explained further below) is evaluated for each scenario, and then the worst case selected foranalysis.

5.2 Premises for the assessment
As premises required for this assessment, this Section describes:The operational processes.
The waste specifications and throughput.
Specifications of the repository engineered barriers, surface facilities and repository.

5.2.1 Operational processes assumed in assessment
As shown in Section 4.2.4 (1) (i), during normal operation, wastes are transported to thesurface facility in transport casks and received at the reception/inspection/encapsulation
facility. Before encapsulation of HLW into an overpack, HLW canisters are temporarily stored
before inspection to confirm compliance with waste acceptance criteria. Such storage is in aradiation shielded cell, with negative air pressure to prevent RN release in case of any surface
contamination being present. After inspection, the HLW is moved to the encapsulation celland sealed into an overpack by remotely operated equipment. The waste handling equipment
for these procedures, such as overhead cranes or lifts, is designed to prevent waste drops as
outlined in Section 4.6.2 (6).
For the H12V case, the overpack is, in the design presented in Chapter 4, placed into atransport cask and moved underground via an access ramp (in the future other means ofaccess may be selected, e.g. shafts). The overpack is transferred to the emplacement machineat the bottom of the access ramp and then moved to, and emplaced in, a disposal tunnel. These
procedures are also carried out by remote-control, with strict radiation protection. Whentunnels are filled with waste, they are then backfilled and sealed. In the assessment, reference
is made to relevant studies performed for similar operations in other nuclear facilities, forexample drop tests of HLW canisters have been already documented elsewhere [3] [4]. This
report focuses on operational processes specific to geological disposal.
For the PEM case, the overpack is placed inside the PEM on the surface and thentransported down into the repository. The detailed technical design for this is not yet decided.
For the TRU waste case, waste package B is placed into a transport container and thentransported into the repository. The waste package is then transferred by overhead crane at theentrance to the disposal vault and emplaced at the designated location within the disposal cell.
Once a disposal cell is full, gaps are filled with mortar and, after all cells in a disposal vault
are full, itis backfilled and sealed. In the case of waste package A, procedures are the same as
for waste package B with the exception of emplacement, which is carried out using a forklift.

5-4

5.2.2 Waste inventory and planned throughput(1) HLW repository
Fundamental features of HLW and its overpack were presented in Sections 2.1 and 4.4.1(2), respectively. Interim storage of HLW before disposal may last for between 30 to 50 years
but, when conducting shielding and thermal analyses, 30 years storage is conservatively
assumed.
1,000 canisters of HLW are planned to be disposed of annually (see section 2.1.1 (4)). As
described in Section 4.6.2 (2), marine transport was assumed to occur twice a year, with 28
units of HLW contained in each transport cask and at least 18 casks in each shipment, so thewaste reception/inspection/encapsulation facility is designed with a store capacity for 20
casks.
In the case of H12V emplacement, overpacks are loaded into transport casks with aradiation shielding and physical protection functions, and then moved underground via theaccess ramp, two at a time, by the transport vehicle (see Supporting Report 4-66). At thebottom of the ramp, overpacks are transferred to emplacement machines, one by one (seeSection 4.4.3 (1) (i)). The emplacement machine transports the overpack to a disposal tunnel
and then lowers it into a disposal hole (see Section 4.4.3 (1) (i)). These procedures are also
carried out by remote-control with strict radiation protection. When tunnels are filled with
waste, they are then backfilled and sealed.
The PEM concept process is the same as for H12V until the overpack is sealed. Thereafter
the overpack is put into the PEM handling shell, together with buffer material. The PEM
provides radiation shielding and also physical protection of the overpack and hence notransport shielding is required.
(2) TRU waste repository
The fundamental features of TRU waste are described in Sections 2.1(2) and 4.4.2 (2),whilst the amount of waste handled annually is specified in Section 4.6.2 (2).
TRU waste transport casks are assumed to be delivered to the repository by ship [5], six
times a year, for 25 years of operation. For the case of drums, depending on size, two or fourare grouted into waste package containers in the surface facilities, placed into a single
transport cask, and then moved underground to the disposal vault. The Gr.4L box containers
are moved directly underground. Waste package unloading and emplacement in a concrete
vault utilises a forklift for waste package A or an overhead crane for waste package B (seeSupporting Report 4-66). After a specified number of waste packages are emplaced in adisposal cell, gaps between waste packages are filled with mortar. In the case of forklift
emplacement, the technology for this backfilling has not yet been decided. After completionof emplacement of all waste packages in the vault, the top backfilling material and plugs are
installed.
The radiological and thermal properties of TRU waste, which will be considered forshielding analysis, are described in Supporting Report 2-3. For the fire impact analysis,among the categories of TRU waste, bituminised waste requires particularly careful handling,
because runaway thermal reactions can lead to spontaneous ignition at temperatures higher
than ≈ 200 °C [6]. For the physical impact analysis of the drop scenario, waste groups of 1, 3,
4L and 4H, which utilise drums, are assessed.
5-55.3 Assessment of normal operational scenarios
Radiological exposure of the public from the repository is assessed based on the strategy
described in Section 5.1.
For the HLW repository, the inventory of radioactive waste at the surface is highest at thereception/inspection/encapsulation facility (see Figure 4.6-2) when 20 transport casks arepresent in the buffer store and 28 HLW canisters from a single cask are stored prior toinspection. For this case, the annual effective dose is calculated as a function of distance tothe site boundary. The radiation exposure from other areas is negligible in comparison, as only
a few HLW canisters are present and they are located in the basement of the facility, withradiation shielding provided by both concrete floors and surrounding soil, as shown in Figure
4.6-2.
In assessment of the temporary storage area for transport casks, since the specification ofthese casks [7] is in accordance with the transport regulations [8] shown in Section 4.6.1, it
can be assumed that the dose rate is less than 2 mSv/h and 0.1 mSv/h at the surface of thecask and at a distance of 1 m from it, respectively.
Exposure by direct and skyshine radiation is assessed to ensure it falls below the annual
effective public dose target of 50 μSv/y, which can be demonstrated if a distance of 200 m or
more from the facility to the site boundary is set. The details of the very conservativeshielding analysis are given in Supporting Report 5-1. For both the H12V and PEM options,the process from reception to encapsulation of HLW in an overpack is the same and therefore
there is no difference in the results of this assessment.
For TRU waste, the assessment was performed in the same manner as for HLW, theradiation exposure considering both the buffer stores for transport casks and waste packages
waiting for inspection. It is concluded that, if the distance from the facility to the siteboundary is more than 100 m, the annual effective dose falls below the target of 50 μSv/y. The
details of the analysis are given in Supporting Report 5-2. The processes until waste
packaging are the same for both waste packages A and B, as shown in Section 4.4.2 (2), andhence there is no difference in the assessment results.
From the above, when the HLW and TRU waste facilities are located at the same site, theeffects from each waste are summed in the assessment. Since the radiation effect from HLW
is significantly larger than that of TRU waste, by maintaining a distance of 200 m or morefrom facilities to the site boundary, the dose target of 50 μSv/y can be met. If the distance tothe site boundary is less than this at a specific site, the thickness of the shielding walls of the
facility could be increased to reduce the dose rate. It should also be noted that the dose
calculations are very conservative and do not take account of shielding by other surrounding
buildings. In addition, greater shielding can be obtained, if required, by locating buffer stores
underground.

5.4 Assessment of abnormal operation scenarios
The radiological consequences of abnormal operation scenarios were assessed taking
account of the safety measures developed (see Sections 4.5.6 and 4.6.2 (6)). All individual
procedures in which waste is handled are subject to a formal event tree analysis, whichassesses how the normal defence in depth to prevent occurrence of abnormal states could
possibly fail. The abnormal states considered can be grouped in terms of the operation
involved, as many of them have similarities that allow them to be treated together when
5-6

assessing their impacts on HLW or TRU waste.
5.4.1

Development of abnormal operation scenarios

The abnormal operations described in Sections 4.5.6 and 4.6.2 (6) result from perturbations
including drops, fires, explosions, loss of external power and other equipment failures. Eventtrees were developed for situations when safety measures fail or the perturbations exceed
those assumed during design. Event trees that resulted in potential loss of the containment
function were selected to develop abnormal operation scenarios.
(1) Drops(i) HLW
Assessment of drops with the PEM during operation has not yet been performed, althoughconsequences will be similar to or less than those for H12V. Operations that could lead topossible drops of the HLW transport cask, canister or overpack for the H12V option include:






The process of lifting and moving transport casks in the buffer store (see Figure 4.6-2).
The process when HLW canisters are lifted out of the transport cask by the overheadcrane and moved to the inspection and temporary storage area and then into theoverpack (see Figure 4.6-2).
After encapsulation, the process when an overpack is moved into a buffer store andthen loaded onto the transport vehicle by overhead crane (see Figure 4.6-2).
The process when the overpack is transferred from the transport vehicle to theemplacement machine by overhead crane at the bottom of the access ramp (see Section
4.5.4 (4) (iii)).
The process when the overpack is lowered into disposal holes (see Figure 4.4-26).

Figure 5.4-1 shows typical event trees for the drop of an overpack during handling by anoverhead crane, including the safety measures shown in Table 4.5-34. The other cases of dropof a transport cask or a HLW canister (or indeed a TRU waste package B) can be described bysimilar event-trees.
Possible abnormal state

Prevention of abnormal state

Incomplete grip

・Interlock to prevent lifting withincomplete grip
・Periodic check and maintenanceof lifting device

Result of counter-measures

・Interlock to prevent releaseduring lifting
・Periodic check and maintenanceof lifting device

・Periodic check and maintenanceof wires
・Use of double wires

Limiting lift height

Mechanical impact
on overpack

Limiting lift height

Mechanical impact
on overpack

Limiting lift height

Mechanical impact
on overpack

Recovery process
Unsuccessful

Success

Recovery process
Unsuccessful

Overpack drop

Wire failure

Consequences

Success

Overpack drop

Release during lifting

Limitation of abnormal state

Success

Recovery process
Unsuccessful

Overpack dropFigure 5.4-1 Event tree for drop of a HLW overpack

5-7

Drops could occur if the overpack is lifted with incomplete gripping, if gripping is wrongly
released during lifting or transport, or if the lifting wire is damaged. It should be noted that
this focuses only on the gripping mechanism and ignores other possibilities, such as failure of
the winch or derailing of the crane. Such aspects will be considered in future assessments.
In order to prevent such perturbations, lifting wires are duplicated and interlocks are
installed in the gripping device. However, if these safety measures fail, impact forces act on
the overpack and contained waste, although a fundamental constraint is set by physicallimitation of the handling height, which prevents excessive drops.
In Section 5.4.2 (1) (i), the effects of drop scenarios are assessed.
(ii) TRU waste
As described in Section 4.4.3 (2) and Figure 4.4-27, waste package A is handled by forklift
and waste package B by an overhead crane. As such, the following operations may be causesof a waste drop:






The process of lifting and moving transport casks in the buffer store (see Figure 4.6-5).
The process when TRU primary containers are lifted out of the transport cask by
overhead crane and moved to the inspection and temporary storage area (see Figure
4.6-5).
The process when TRU primary waste containers are lifted into the waste package byoverhead crane (see Figure 4.6-5).
After production, the process where waste packages are moved to a buffer store andthen to the transport vehicle (crane or forklift, only crane option shown in Figure 4.65).
The process that stacks and emplaces waste packages in the concrete cell of thedisposal tunnel (refer to Section 4.4.3 (2), Figure 4.4-27).

The event trees for drops are effectively the same as for HLW (e.g. Figure 5.4-1), withsafety measures (e.g. limitation of the handling height) as shown in Table 4.5-34. As describedin Section 4.4.2 (2) (ii), emplacement by overhead crane uses a twist lock gripping system(see also Figure 4.4-27) and, in case of failure, a drop may occur.
Waste package A is assumed to be moved by a waste package conveyor and lifted byforklift only during transporter loading and unloading/waste emplacement (as shown in the
TRU-2 report). Since the drop height is greatest when waste packages are stacked in thedisposal vault (See Figure 4.4-29), a waste package drop scenario is developed for thisprocess as described in Section 4.4.2 (2) (ii), with the specified number of layers given in
Table 4.5-5.
The forklift incorporates an interlock system which prevents lifting when fork insertion
into the lifting pockets is incomplete It also prevents withdrawal of the forks unlessemplacement is complete. If a malfunction of the interlock system occurs, e.g. in the casewhere the forklift backs away before emplacement is complete, a waste package could drop to
the floor of concrete vault.
Finally, it should be mentioned that there is also a potential risk of waste stacks toppling;
however, this risk has not yet been analysed.

5-8(2) Fires(i) HLW
The causes of, and counter-measures against, fires in facilities are described in Sections
4.5.6 (2) and 4.6.2 (6) (ii). In the following, only fires in underground facilities are discussed,
since there is already substantial experience in assessment of fire impacts in surface facilities.
Evaluation of the risks of, and countermeasures against, fires in surface facilities will be
included in later stages of the programme. Wherever possible, all construction in the facilities
utilises non-flammable or flame-retarding materials. Even if flammable materials are present,
limitation of ignition sources would prevent occurrence of fires (and explosions). In addition,
fire, gas and smoke detectors, together with alarms and fire extinguishing equipment, will be
installed in all sensitive areas.
The following equipment and devices described in Chapter 4 for the H12V option,
however, may include combustible materials:On the access ramp; transport vehicle, diesel fuel and wheels with rubber tyres.
At the bottom of the access ramp; hydraulic and lubricating oils used by the overheadcrane that transfers overpacks from the transport vehicle to the emplacement machine.
Diesel fuel and rubber tyres of the emplacement machine that transports overpacks todisposal tunnels and emplaces them into disposal holes.

The ramp transport vehicle is likely to have the largest quantity of combustibles, as
indicated in Figure 4.5-36, hence fire scenarios focus on it. It is also recognised that diesel may
not be used in the future, as electric vehicles are then more likely, and fire hazards for electricalvehicles will also need to be assessed. Event trees for fire scenarios are shown in Figure 5.4-2,

based on the design of the transport vehicle (Supporting Report 4-66).
Possible abnormal state

Ignition of rubber tyres
due to malfunction ofbraking system or
spreading fireFire due to ignition offuel caused by defector collision incident

Prevention of abnormal state

・Use of incombustible
materials surrounding waste
e.g. steel platform andsupport frame
・Detection system forabnormal heating events・Check and maintenance ofvehicles・Use of incombustible
materials surrounding waste
e.g. steel platform andsupport frame
・Detection system forabnormal heating events・Check and maintenance ofvehiclesResult of counter-measures

Limitation of abnormal state

Consequences

Success
Recovery process

Unsuccessful
Fire spreading to thevehicle・Fire detection system andextinguishing equipment on thevehicle・Fire detection system andextinguishing equipment in theaccess ramp
・Protection by transport
container

Success
Recovery process
Unsuccessful
Thermal impact on thevitrified waste

Success
Recovery process
Unsuccessful
Fire spreading to thevehicle・Fire detection system andextinguishing equipment on thevehicle・Fire detection system andextinguishing equipment in theaccess ramp
・Protection by transport
container

Success
Recovery process
Unsuccessful
Thermal impact on thevitrified waste

Figure 5.4-2 Event trees for HLW transport vehicle fire in an access ramp

Since diesel oil and rubber tyres have the highest potential heat production, these are set as
fire sources (see Figure 4.5-36).

5-9

The tyre fire scenario is described as:






During transportation on the access ramp, one of the transport vehicle rear wheels isabnormally heated due to a braking system malfunction, leading to ignition of a tyre.
Alternatively, a small fire, e.g. due to leakage of gear oil, spreads to ignite a tyre.
The operator would notice this fire, but attempted fire extinguishing operations fail,allowing the fire to spread to adjacent tyres.
Air heated by the fire and thermal conduction through the vehicle deck would warm thetransport cask and, by conduction, also its contents.
Even after the fire burns out, the deck temperature remains high and will continue toheat the cask and its contents by thermal conduction.
In the case of diesel oil pool fires, the scenario assumes fuel would leak to the tunnel
floor and be ignited due to an abnormal state involving the transport vehicle.

In the PEM case, the thermal impact to HLW will be lower than H12V, since the overpack
containing HLW will be surrounded by thick bentonite buffer which restricts heat conduction.
Thus, in Section 5.4.2 (2) (i), the thermal effects of fires on HLW are assessed for H12V only.
(ii) TRU waste
Fire scenarios for TRU waste are effectively the same as for HLW; so Figure 5.4-2 can beapplied.
As shown in Table 2.1-2, most of the TRU wastes comprise inorganic material (mortar andmetal) and non-combustible wastes, therefore spontaneous ignition (or explosion) is unlikely
and fire impact assessment is similar as to that for the HLW case. An exception is the Gr.3
waste, which is assumed to be stable at ambient temperatures in the underground facility(about 30 to 45 °C), but the bitumen and nitrate mixture has a risk of spontaneous ignition if
the waste is exposed to higher temperatures [6]. Fire scenarios thus focus on this waste, as
consequences are likely to be more severe than for other TRU waste groups.
(3) Explosions
As discussed in Section 4.5.6 (3), the SDMs used in this in this study do not indicate risks
from potentially explosive gases and their generation from the waste or other repository
components is assumed to be unlikely. Therefore, explosion scenarios are not considered, but
the methodology to assess them is under development and is introduced in Section 5.4.2 (3),assuming methane gas occurrence, and will be addressed further in future work.
(4) Loss of off-site power
In this scenario, shutdown of surface and underground facilities due to loss of off-site
power is described. In developing this scenario, it is recognised that just loss of off-site power
would have little immediate impact on operations, since all systems could be brought to a safe
shut-down condition using uninterruptable power supplies, auxiliary generators, batteries, etc.
However, as a limiting case and illustrating consequences of a situation that could lead tocommon mode failure of many safety functions, the consequences of a sudden total blackout,e.g. caused by a massive natural or anthropogenic electro-magnetic pulse (EMP), is worth
assessing in the future.
As described in Section 4.5.6 (4) and Section 4.6.2 (6) (iii), sudden loss of power may
5-10

cause a drop of a HLW canister or overpack being handled by an overhead crane. However,this is effectively the same as for the drop scenarios described in Section 5.4.1 (1).
Loss of power may also cause shutdown of drainage systems in underground facilities.
However, for HLW and type B TRU packages, the drainage water would not be contaminatedunless failure of radioactive waste containment occurs. For these waste types, the shutdown of
the drainage system is unlikely to cause direct RN release during loss of off-site power.
However, it should be noted here that longer term impacts of loss of drainage have not yet
been assessed, but are recognised as a future priority.
In the reception/inspection/encapsulation facility, negative pressure is established for areas
where the possibility of contamination exists, however high-efficiency particulate air (HEPA)
filters at air outlets should prevent RN releases even if negative pressure cannot be maintained(see section 4.6.2(6)(iii)). Similarly, in any event where a release of RNs from undergroundfacilities is detected, switching to an emergency ventilation system equipped with HEPA
filters prevents RN release from the facilities, if backup power is available (See Section 4.5.6(6)). However, in the case of total power loss, including both off-site and all backups,
shutdown of all ventilation and loss of part of the containment function will result. As long as
waste forms are not damaged, RNs will not be released. Therefore, the assessment of theradiological consequences of ventilation shutdown scenarios is included only when thepossibility of damage to waste forms results from assessment of other scenarios.
In addition, in the event of ventilation system shutdown, the temperature of HLW in thebuffer stores rise and hence this scenario is analysed. Although ventilation systems are notprimarily used to cool down HLW, these are assessed as an effect of loss of power.
Finally, it is noted that, apart from the systems assessed above, total loss of power could
also include loss of control functions, monitoring systems, lighting and communications. Such
impacts will be covered in future studies.
(5) Other system failures
External perturbations, system damage, malfunctions and human errors can result in drops,
fires, explosions and power loss, as described above. Many other possible events could impact
RN containment and, as an example, an abnormal state involving the ramp transporter vehicle
is considered, as shown in the event tree in Figure 5.4-3.
Possible abnormal state

Prevention of abnormal state

Runaway of thetransport vehicle in theaccess ramp due tomalfunction or humanoperation error

・Periodic check and maintenanceof the vehicle
・Provision of the mechanisms forautomatically limiting speed of
the vehicle

Result of counter-measures

Limitation of abnormal state

Consequences

Limiting length of straightslopes and provision of
capture zones / arrestors

Mechanical impact
on overpack

Success

Recovery process
Unsuccessful

Vehicle collision withthe ramp wall

Figure 5.4-3 Event tree for transport vehicle runaway on an access ramp (HLW)

As described in Table 4.5-38, regular maintenance and interlocks with multiple brakes toprovide speed limitation are design counter-measures against vehicle malfunctions. Failuresof these measures could lead to vehicle runaway and then collision with the tunnel wall. In
order to reduce the collision impact, the length of straight slopes could be limited andrunaway capture zones/arrestors of the type used on steep ramps included in the design.
5-11

This abnormal operation scenario for HLW would be directly applicable also to TRU
waste, with “overpack” in Figure 5.4-3 replaced by “TRU waste packages”.
5.4.2 Consequence analysis for abnormal operation scenarios
For the abnormal operation scenarios described in Section 5.4.1, consequence analyses are
conducted to determine if there is any risk of loss of containment function due to resulting
damage. Detailed designs of equipment in the facilities have not been developed at the current
generic stage of the programme, and hence failures are assumed without considering potential
resilience designed to limit either probability of occurrence or resultant impacts of abnormal
operations.
(1) Drops(i) HLW
Transport casks are designed to resist a drop from a height of 9 m based on transport
regulations, so the lifting height limit alone assures that no loss of containment would result
from a drop [7] [8].
For assessment of drops of a HLW canister, an analysis of past drop tests indicated that no
penetrating cracks occur when the lifting height was less than 9 m, though the canister might
be deformed [3] [4]. Thus, prevention of loss of the containment function is achieved bylimiting the lifting height to significantly less than this value.
For overpack assessment, a drop from the greatest height possible is assessed – i.e. in thereception/inspection/encapsulation facility, with a limit of 9 m (see 4.6.2 (6) (i)) by taking a
conservative approach as described in Section 5.1.2 (2). In underground facilities, as shown in
Section 4.5.4 (4) (iii), the lifting height of the overhead crane for transferring overpacks fromthe on-site transport vehicle to the emplacement machine is 3 times the height of the overpack(1.9 m) and hence significantly less than 9 m. Similarly, the maximum overpack drop fromthe emplacement machine into the disposal hole is about 4.2 m (floor surface of the disposal
tunnel to the bottom of the disposal hole - see Figure 4.5-2), although the presence of buffer inthe hole would reduce the drop height further.
For the maximum drop height of 9 m, the free fall velocity of the overpack at collision withthe floor surface is calculated, and then elasto-plastic analysis used to assess impact with theconcrete floor slab to determine the possibility of generating through-cracks.
The drop onto a concrete floor slab was assumed to occur at an angle, so that the impact
occurs on a more vulnerable edge. If the equivalent plastic strain exceeds the breaking strain
0.23 of the carbon steel (SF 340A) used [9] (see Section 4.4.1 (2) (ii)) and connects from theoutside to inside of the overpack, penetrating cracks could occur. Figure 5.4-4 illustrates theanalytical results, showing that the area exceeding the breaking strain limit was found only
near the collision point and thus no crack penetration would occur. The details of this analysis
are provided in Supporting Report 5-3.

5-12

Equivalent plastic strain

2.3×10-1
Breaking strain ofcarbon steel (SF340A)

Cross-section
Concrete floor

Concrete floor

Figure 5.4-4 Physical analysis of overpack drop

In the case of the PEM option, after emplacement in the handling shell, drop impacts onto
the overpack are physically reduced by the PEM handling shell and the buffer material, so
damage would be even smaller.
(ii) TRU waste repository
As for HLW, drops of transport casks are not a concern. To prevent loss of the containment
function, the lifting height is physically restricted in designs of equipment and systems as in
the HLW case [7] [8]. As described in Section 4.6.2 (6) (i), the lifting height of drums (Grs.1,
3, 4L, part of 4H) will be designed to be less than 6 m [10], while Gr.2 and Gr.4L (boxcontainer) lifts are less than 9 m.
The mechanical durability of wastes solidified into other containers has not been assessed,
because these waste containers have not yet been fully specified. After such specification, thesame assessment of drop consequences as for the other waste-forms will be made, allowing
lifting height limits to be set.
For type B waste packages lifted into disposal vaults by overhead crane (See Figure 4.4-27
and Section 4.6.2 (6) (i)), the maximum drop height was assumed to be the maximum liftingheight of 8 m, (the lifting height during stacking is less than 8 m). As described in Section
4.6.2 (6) (i), based on past drop tests, no failure of canisters of Gr.2 waste and box containersof Gr.4 waste were observed up to a height of 9 m while, in the case of drums, lid failure wasreported for drops from 6 m or more. As an example, an elasto-plastic analysis simulating adrop of waste package B containing drums was conducted, with the possibility of throughcracks assessed as for the HLW overpack, using the breaking strain 0.24 of the carbon steel(SM400A) used [11] (see Section 4.4.2 (2) (ii)).
The worst-case drop of the waste package was considered to be when a corner of the uppersurface, which has the twist-locks, lands on the concrete floor, so this was the case modelled.
Figure 5.4-5 illustrates the analytical results, showing that areas where the equivalent plasticstrain exceeds the breaking strain were found, however fracturing is considered unlikely
because the area where the failure criterion are exceeded was very limited.

5-13

Drum

Waste package container

Equivalent plasticstrainEquivalent plastic strain

2.4×10-1
Breaking strain ofcarbon steel(SM400A)

2.1×10-1
Breaking strainof carbon steel(SS400)Figure 5.4-5 Results of physical impact analysis for a type B waste package containing 4 drums

In addition, a maximum equivalent strain of 0.08 of the contained drums was calculated,which is far below the relevant breaking strain (0.21) for SS400 type steel [11] and occurs
only in the vicinity of the collision point. Thus, it is clear that there is no risk of RN release.
The details of analytical methods and associated calculations for type B waste packages are
given in Supporting Report 5-4.
Equivalent scenarios for waste package A assume that the drop occurs during emplacementin the disposal vault by forklift. As shown in Table 4.5-5, the maximum number of waste
packages containing drums in a stack is six, for both Gr.3 and 4L vaults. If a waste package
drops due to malfunction during the emplacement of the top layer, the drop height isequivalent to the height of the upper surface of the 5th layer (5.75 m). According to drum drop
tests [10], the drum could be deformed but the contents will not be released when dropped
from a height of < 6 m. In addition, since the drum is solidified with mortar within the waste
package, the impact force on it is significantly smaller than when dropped directly. It is thus
considered that there is a low probability of package failure, which will be confirmed in futurestudies.
(2) Assessment of fire scenarios
Fire scenarios considered in Section 5.4.1 (2) focus on the ramp transport vehicle, as
discussed in Section 4.5.6 (2), with determination of the resultant temperature increase of theoverpack (or TRU waste package). This vehicle has a rack installed on a flatbed deck to hold
either two overpack transport casks (for H12V, see Section 4.5.6 (2)) or a single cask for a
TRU waste package (see Supporting Report 4-66). The transport casks have a radiation
shielding function, but also protect the contained overpack/waste package from physicalimpacts and fire in case of abnormal states. Transport casks are installed near the centre of thedeck, located as far as possible from the fuel tank (see Supporting Report 4-66).
From the design of the transporter vehicle, diesel oil and rubber tyres were determined tobe the major combustibles present. Fire duration is calculated for each fire source andcompared, with the scenario that has the longer duration chosen for assessment.
(i) HLW(a) Duration of tyre fires
According to a report of causes of fire in road vehicles [12], a common case involves
abnormal wheel heating due to locking of the braking system until ignition of the tyre. The
5-14

associated uncertainties in this analysis have not been assessed at this point, since it wasconsidered sufficient to look at just one representative scenario to illustrate the assessment
process. At later stages, more thorough assessment may be carried out if design changes are
not made to eliminate this issue (e.g. change to electric-powered rail transport).
The calculation assumes that, in the event of a fire, ventilation would be stopped to limit
air flow, but did not consider fire doors to prevent exhaust of combustion gases through theaccess ramp, see Support Report 5-6 for details. The duration of fire was calculated assumingthat fire from one tyre spreads to those nearby as follows;

The right tyre of the third axle (denoted as tyre 3) closest to the two transport casks(shown in Figure 4.5-36) ignites.
As this tyre burns, the right tyre of adjacent second and fourth axles (denoted as tyre 2,
tyre 4) are heated and ignite.
As the right tyre of the fourth axle burns, the right tyre of the adjacent fifth axle
(denoted as tyre 5) is heated and ignites.

In order to calculate the fire duration, the combustion analysis used a numerical model ofthe vehicle, as shown in Figure 4.5-36. Figure 5.4-6 (left) plots the heat generation rate as afunction of elapsed time since ignition.
Overpacks in transport
casks

Tyre ②,④
Maximum heat generation at
3,460 sTyre ③
Maximum heat generation
at 4,940 s

Tyre ③
Maximum heat generation at
1,940 sSupportframe

Loading
platform

Drivingcabin

Flames

4,000
⑤

④

3,500

③

②

①

Heat generation (kW)

700 s elapsed
Tyres

3,000

2,500

2,000
⑤

1,500

④

③

End of heatgeneration at 5,700 s

1,000

Tyre ③
ignited at
0s

500

Tyres ②,④
ignited at
1,500 sTyre ⑤
ignited at
3,000 s
⑤

1,000

2,000

①
1,940 s elapsed

0
0

②

3,000

4,000

5,000

Elapsed time from ignition (s)

6,000

7,000

④

③

②

①
3,460 s elapsed

Figure 5.4-6 Combustion analysis for tyre fire of a ramp transport vehicle (HLW)

The ignition condition for rubber tyres was a surface temperature over 400 °C or a heat
flux of over 16.5 kW/m2 [13].
Initially, only tyre 3 on the right side of the vehicle starts burning and then neighbouring
tyres 2 and 4 reach the ignition condition and catch fire after1,500 seconds. Finally, tyre 5
catches fire after 3,000 seconds. In this simulation, tyre 1 on the right side of the vehicle andthe tyres on the left side did not reach the ignition condition.
The heat generation rate reached a maximum value of 3,300 kW after 3,460 seconds,
5-15

mainly due to tyres 2 and 4, and decreased to zero at 5,700 seconds when the fire burned out.
The fire duration was thus ca. 90 minutes. Figure 5.4-6 (right) shows the distribution of heataround the burning tyres from the numerical model. The flames, represented as orange, areshown to reach upper parts of the flatbed deck, but not the transport casks or the fuel tank.
The details of this combustion analysis are provided in Supporting Report 5-6.
The fire model is accepted to be very simple and will be developed further in the future,e.g. to assess the impacts of tyre burst during fire and potential for vehicle to tip, causing
flames to reach tyres on the other side.
(b) Duration of fuel fires
The analysis assumes that diesel fuel leaks from a fuel tank to the floor of the tunnel, e.g.
due to collision with the tunnel wall or failure of the fuel tank, with fire duration calculated
assuming a pool fire. The size of the flame was calculated based on the method described by
the Architectural Institute of Japan [14]. The area of the pool of leaked fuel (200 litre), see
Supporting Report 4-66) was assumed to be 50 m2, giving a theoretical flame height of ≈ 11m, so the flame will reach the 7 m high tunnel roof (see Table 4.5-6). Fire duration was
calculated to be 60 seconds, based on the burning rate of diesel fuel and the depth of the pool.
Details of the calculation of the fire duration are described in Support Report 5-5. Again,
potential uncertainties in these results have not been assessed at this point, since the focus was
only on illustrating development and assessment of a representative scenario. Nevertheless,the analysis suggests that a tyre fire would be the limiting event.
(c) Fire impact analysis for HLW
Since the fire duration was longer in the tyre fire case, this was assessed to evaluate
consequences. Carbon steel, used for the overpack and waste package, undergoes phasetransformation above 727 °C [15], while HLW borosilicate glass may devitrify attemperatures above 610 °C [16] [17]. If these temperatures are exceeded, the integrity of theoverpack may be compromised. These temperatures are therefore used as references to assess
the potential impact of heating on containment functions.
Figure 5.4-7 shows the distribution of temperature on the surface of the vehicle deck andtransport casks, as determined by thermal conduction analysis using the amount of heat input
to the lower surface of the deck from the tyre fire combustion analysis.

5-16Temperature distribution (1 hour elapsed)

Temperature distribution (2 hours elapsed)

Figure 5.4-7 Temperature distribution of the vehicle deck during the tyre fire (HLW)
Since tyres burn only on the right side, major temperature rise is limited to the area above
them. The colour legend represents temperature in °C

The temperature of the bottom plate of the deck on the right side of the vehicle where thetyre burns is the highest and rises to 649 °C, however the temperature of the top plate of thedeck only rises to 99 °C because of limited heat conduction due to the hollow structure of thedeck. In addition, the amount of heat input to the casks was about 0.4 kW at maximum,
significantly smaller than the heat generation rate from the fire presented in Figure 5.4-6. Thisis because flames from the tyres are blocked and dispersed by the deck (see Supporting
Report 5-6), as shown in Figure 5.4-6 (right).
Figure 5.4-8 plots the maximum temperatures of the transport cask, overpack and HLW as
a function of time. These temperatures were determined by the following procedure:

Steady-state thermal analysis was conducted to determine the temperature distribution
within the HLW overpack.
The heated overpack sits within a transport cask on the deck of the ramp transport
vehicle and transient thermal analysis was conducted assuming that it took 5 hours fromloading to moving to a depth of 500 m, where fire is assumed to occur.
The heat fluxes from the deck of the vehicle and heated air around the transport cask,
determined by the combustion analysis illustrated in Figure 5.4-6, were applied as heat
input.
Since heat is transferred from the vehicle deck even after the fire has burned out,
thermal analysis was continued for 12 hours from fire initiation.

5-17

Figure 5.4-8 Temperatures at various positions as a function of fire duration(30 years interim storage of HLW)

The initial temperatures of the centre and surface of the HLW and the surface and bottom of
the transport cask at the time of fire were 124 °C, 95 °C, 31 °C and 34 °C, respectively. Figure
5.4-8 shows that the temperatures of the centre and surface of the HLW were almost constantduring and after the fire. The temperatures at the surface and bottom of the transport cask
increased to 37 °C and 42 °C, respectively. These results are very much below the criticaltemperature criteria and this shows that RN release due to failure of containment is highly
unlikely for this scenario, even if all uncertainties have not yet been assessed. The details of
the fire impact analysis described above are provided in Supporting Report 5-7.
ii) TRU waste
The tyre fire scenario for the transport vehicle of TRU waste was assessed in a similarmanner, using the heat generation rate applied for the case of HLW shown in Figure 5.4-6.
Only waste package B containers were assessed, but findings are assumed to be similar for
waste package A.
As noted previously, bituminised waste in Gr.3 is assumed stable at normal temperatures.
However, at temperatures > 180 °C to 200 °C, nitrate reaction with bitumen is autocatalytic,
potentially leading to a runaway thermal reaction [6] and therefore this waste was targeted forassessment of fire scenarios.
Figure 5.4-9 shows the distribution of temperature on the surface of the deck and transport
cask calculated in a similar manner to the HLW case.

5-18Temperature distribution (1 hour elapsed)

Temperature distribution (2 hours elapsed)

Figure 5.4-9 Temperature distribution of the truck deck during a tyre fire (TRU waste)
Since tyres burn only on the right side, major temperature rise is limited to the area above
them. Colour legend shows temperature in °C

Heat input to the transport cask was at most 4 kW. This value was higher than that of thecase of HLW shown in Figure 5.4-8, because the surface of the transport cask is closer to theedges of the deck, but was again significantly smaller than heat input to the bottom of thedeck (Figure 5.4-6) for the same reasons as the HLW case. Similarly, although the lower
surface of the deck is in contact with the flames and has a temperature of up to about 640 °C,
the upper deck temperature increases only up to ≈ 80 °C.
The normal radiogenic heat production of bituminised waste is negligibly small (< ≈ 1 W,
see Supporting Report 2-4), and therefore the initial temperature was assumed to be 30 °C,
equivalent to the ambient rock temperature at a depth of 500 m. The analysis continued until
12 hours although heat generation effectively ends at 1.5 hours (see Figure 5.4-6). Figure 5.410 shows changes in the maximum temperature of the transport casks, waste packages, andbitumen matrix based on this analysis.

5-19

Figure 5.4-10 Temperatures at various positions as a function of fire duration (TRU waste)

The surface of the transport cask reaches a maximum temperature of 35 °C approximately 1hour after the fire and, thereafter, gradually decreases towards ambient temperature. The
temperature of the waste package and contained bitumen gradually increases, but the rate ofincrease slows after 12 hours. Thus, the temperature rise of the bitumen is far lower than the
runaway temperature. For details of this fire assessment, see Supporting Report 5-8 (but itshould be noted here that a more thorough assessment of fires will likely be needed in the
future).
(3) Explosions
As mentioned in Section 5.4.1 (3), Both monitoring and counter-measures (e.g., sufficiently
high ventilation rates, flammable gas catalytic oxidisers) will be implemented to reduce therisk of a gas explosion to the extent possible. Nevertheless, especially given that relatively
high methane inflow is experienced in some tunnels in Japan (e.g., at the Horonobe URL),
methodology is being developed to assess the consequences of such an abnormal operation
scenario. It should be emphasised, however, that here focus is on illustration of how models
could be applied to a highly simplified scenario, rather than consideration of how realistic the
scenario (and hence interpretation of results) may be.
For this evaluation of a gas explosion, it is assumed that methane concentration in a tunnel
results from outgassing from the host rock, and increases due to the failure of the ventilationequipment. It is assumed that such a problem is not picked up by monitoring and/or countermeasures cannot be implemented, possibly due to common mode failure of detection/alarm
equipment, loss of access, etc.
As a worst case, ignition from an unspecified source leads to an explosion in a disposal
tunnel, which occurs under the condition of a methane/air mixing ratio that maximises its
explosive power. This explosion then impacts a HLW overpack that is sitting on the floor of
the tunnel. For the specified H12V emplacement option, such overpacks would be within atransport container until lowered into the disposal hole, but the more complex analysis
required for this case will be introduced only in the future.
The maximum speed of the overpack displaced by the worst-case explosion was calculated
to be 20 m/s by numerical analysis of the blast wave. An elasto-plastic analysis was thenperformed, assuming that the overpack collides with the tunnel wall surface at this speed. As
5-20

shown in Figure 5.4-11, the equivalent plastic strain exceeded the breaking strain near thesurface of the overpack at the collision point, but the deformation of the inner surface of theoverpack was within the elastic region, so no through-cracks should occur. Details of this
impact assessment are given in Supporting Report 5-9.

Velocity (m/s)

相当塑性ひずみ
Equivalent
plastic strain―Direction along the longitudinal axis oftunnel―Direction normal to the tunnel axis

Elapsed time from explosion (ms)Figure 5.4-11 Results of impact analysis for a methane gas explosion scenario

This calculation was carried for the H12-V concept, but could readily be applied to the
PEM option. Nevertheless, it is clear that the consequences for the PEM would besignificantly less due to its greater mass and the protection of the overpack by the PEM
handling shell and surrounding buffer. In the future, a priority will be to assess explosion risks
for TRU waste, for which case explosive gasses produced from the waste packages
themselves (mainly hydrogen and methane) would also need to be considered. In all cases,
gas explosion impact assessments will aim to be more realistic and provide feedback to designthat improves resilience – decreasing both the probability of such abnormal states and theconsequences should one occur.
(4) Loss of power
Based on Section 5.4.1 (4), an illustrative scenario for loss of ventilation cooling of the
HLW buffer store is assessed here. During buffer storage immediately after unloading of atransport cask, 28 canisters of HLW may be present and loss of ventilation will result in
temperature increases. Thermal analysis was conservatively conducted by assuming heatgeneration of 559 W per HLW canister (i.e. after 30 rather than 50 years interim storage).
Figure 5.4-12 shows that the temperature rises rapidly until 5 days after loss of ventilation,and reaches about 200 °C at the surface of the waste. Thereafter, the temperature risesignificantly slows down, reaching a plateau of about 210 °C after about 30 days. Thistemperature is much lower than the sensitisation temperature (450 to 850 °C) [18] and melting
point (1,398 to 1,427 °C) [15] of austenitic stainless steel and hence damage of the HLW
canister and RN release is extremely unlikely. In addition, the centre temperature of the HLW
is sufficiently below the devitrification temperature limit of 610 °C [16] [17] that this is not aconcern. The details of this analysis are given in Supporting Report 5-10.

5-21温度高い（赤：230℃）

250
HLW 廃棄体中心（最高）centre (highest T)

Temperature
温度(℃) (℃)

200

HLW 廃棄体表面（最高）surface (highest T)

150
HLW
surface (lowest T)
廃棄体表面（最低）100
Air (average)
空気（平均）
天井（平均）
Roof (average)

50

distribution after 30 days
換気停止から30日後の温度分布
温度低い（青：20℃）Temperature

Concrete
floor (average)
コンクリート（平均）

0
0

10

20

30

Elapsed time (d)
換気停止後の経過時間(日)Figure 5.4-12 Assessment of temperature rise of HLW due to loss of buffer store ventilation

In the case of TRU waste, as shown in Supporting Report 2-4, the highest heat generation
rate results from Gr.4 waste (glass melting furnaces) at ≈ 60 W/container. This heat generation
rate is thus small compared with the HLW (about 10%). Further, only 8 containers are
assumed to be received in one transportation shipment and thus there is no need of temporarystorage before inspection. A moderate temperature rise is possible in the case of loss ofventilation, but currently this is considered to have negligible safety impacts, although will be
more rigorously assessed at a later date.
(5) Assessment of equipment malfunction scenarios
As noted in Section 5.4.1 (5), the case of the runaway of the ramp transport vehicle due to a
failure of its braking system is chosen as an assessment example. Even with the conservative
assumption that all abnormal state counter-measures fail, collision speeds were determined to
be insufficient to damage the overpack severely due to a penetration crack [19].
The actual speed at the time of collision in a runaway scenario needs to be set on a site- anddesign-specific basis, taking into consideration the length of possible runaway sections andthe effects of any catch lanes, arresters, etc. These counter-measures will be specified in the
future, as part of the repository design of access ramps for specific sites.
5.4.3

Potential for radionuclide release from the repository

A wide range of abnormal operation scenarios were assessed with a focus on the potential
failure of containment and resulting RN release. Partial deformation by the force of drops wasnoted, but sufficient damage to the overpack (or waste package) that could lead to RN releaseis considered to be unlikely. In addition, even if the temperature of HLW or TRU waste iselevated due to fires or failure of ventilation cooling, current analyses suggest it is insufficient
to give a risk of thermal damage.
For the specific case of PEM emplacement, the PEM shell and buffer material provide
additional mechanical and thermal protection of the overpack compared to the H12V case.
In the future, in order to enhance the reliability of the assessment, more detailed equipment
designs will be developed based on actual site environmental conditions, allowing
development and evaluation of a more complete range of perturbation scenarios.
5-22

5.5 Mitigation of, and recovery from, accidents
According to the assessment results in Section 5.4, the properties of the waste itself,
primary containers and, after encapsulation, overpacks or disposal packages with considerable
physical and thermal resistance, makes any loss of RN containment very unlikely, even underabnormal operational conditions.
Even if unlikely, it is important to consider any credible scenario that could potentiallyresult in RN releases. Mitigation of, and recovery from, significant accidents, whether or not
RN releases actually occur, is therefore considered. A perspective on this is provided by therecovery plan [20] developed after the accidents at the US Waste Isolation Pilot Plant (WIPP)
site.
5.5.1

Lessons from accidents at WIPP

Within the WIPP facility, two separate accidents occurred in February 2014: a rocktransport vehicle fire and a radiological release from emplaced waste. A summary of theaccidents and associated recovery work, based on the reports of US Department of theEnvironment (DOE) and related research institutes, is given below.
(1) Outline of the WIPP accident
According to the accident investigation report [21], the vehicle fire on February 5, 2014,
started near the engine compartment of a truck carrying excavated rock salt. The cause was aleak of oil (lubricating fluid or fuel), which contacted the hot exhaust system and led toignition. The worker driving the vehicle initially tried to extinguish the fire, without success,and within 10 minutes the front tyre ignited and the fire burned fiercely for 20 to 40 minutes(this sequence of events demonstrates that the fire scenarios analysed by NUMO could indeed
occur, but also that propagation of fires from fuel to tyres would need to be considered in the
future).
Evacuation of workers from the underground facilities was completed within 50 minutes ofdetecting the fire. The fire continued to smoulder for about 12 hours, until finallyextinguished by a team using foam extinguishing agents. The work zone in which the accident
occurred did not handle radioactive waste and thus there was no RN release, but six workers
were treated for smoke inhalation.
Independent of the vehicle fire, on February 14, 2014, there was an event in which RNs
were released from a drum after emplacement. In this accident, low pH wastes containing
nitrate and organic matter, which did not meet the waste acceptance criteria, were mistakenly
received and emplaced. After emplacement in the disposal panel, due to a runaway chemical
reaction, heat and gas generation occurred, bursting the lid of a drum and causing RN release
into the facility [22] [23] [24].
In normal operation, air was not filtered prior to discharge. On the day of the accident, acontinuous air monitor detected radioactivity and therefore the system automatically switched
to High Efficiency Particulate Air (HEPA) filtration, capable of removing most RN from theexhaust air. As a result, it was possible to prevent a major release to the outside, however aslight increase in the radioactivity in the air was observed outside the facility, due to leakageof valves that redirect ventilation to the filtration system [23].

5-23Due to the radiological release accident, US DOE formulated the WIPP Recovery Plan[20], suspending operations until December 2016 in order to carry out recovery work, such as
decontamination of the facilities.
(2) Recovery work
The recovery work at WIPP was carried out in two phases and included the steps below[20].
Recovery Plan - Phase I








Isolate ventilation flow-through: two bypass dampers, which had allowed limited
airflow to bypass HEPA filters, were sealed with high-density foam.
Monitoring instruments were lowered into the air intake shafts: data from the monitors
showed no radiological contamination in the area of the shafts underground andestablished safety for personnel entry.
Recovery teams entered and surveyed conditions to establish two usable egress
locations — a requirement for future underground work. The teams confirmed
communications with the surface base station and established an underground base ofoperations.
Recovery teams surveyed conditions further, moving towards the suspected releaselocation and identified a breached waste container. A Technical Assistance Team ofexperts was established to review photos and video, as well as analyse samples ofdebris, in order to determine the cause of the release.
Contaminated filters were replaced: for operations to continue, a properly functioning
ventilation system was required and hence two filtration units were replaced.

Recovery Plan - Phase II

Mitigate the contamination source: this was accomplished according to a plandeveloped from the knowledge gained during Phase I.
Restore conditions that will support operations: this includes equipment for radiologicalmonitoring, ground stabilisation activities (e.g., bolting) for mine safety, equipment andsystems maintenance, cleaning, and upgrades.
Incorporate corrective actions: lessons learned from the fire and RN event are used toenhance programmes and procedures prior to resuming operations.

In parallel with the above recovery work, identification of the causes of the accidents was
carried out and results were disseminated on a website [25]. It is noted that the root causes
were fundamentally due to failures of waste production/characterisation QA and waste
acceptance testing – in turn resulting from poorly defined regulations and acceptance criteria,
together with weaknesses in staff training and safety culture. Corrective actions are compiled
as a safety assessment report [26], where the following improvements are described:Fire protection measures (maintenance and inspection of equipment used in nonradiation work, installation of automatic fire extinguishing equipment, etc.).
Emergency response plan (continuous assessment of emergency response requirements,
education and training of workers, etc.).
Waste confirmation (approval of wastes on reception, confirmation of certified waste,5-24

etc.).
5.5.2

Mitigation and recovery plans for a Geological Disposal Facility (GDF)

Based on the assessment results shown in Section 5.4.2, RN releases are unlikely in theperturbation scenarios examined but, considering the WIPP experience summarised in Section
5.5.1, the installation of an emergency ventilation system is considered prudent as a measure
to mitigate the effects of any RN release that could possibly occur (see Section 4.5.6 (6)). In
addition, conditions not relevant to WIPP, like the need for drainage management, should also
be considered.
If RN leakage occurs, recovery using the following procedure can be considered, referringto the WIPP Recovery plan.










Contain the released RNs within the facility.
Implementation of measures to prevent the spread of contamination (ventilation anddrainage).
Sharing and disseminating information on accident.
Investigation of the site where the accident occurred and confirmation of radiologicalrelease, identification of the cause of the accident.
Formulation of a recovery work plan.
Establishment of a response base for the recovery work; training and education of a
remediation team in the recovery work.
Recovery of damaged waste.
If there is contamination, decontamination of the contaminated area.
Confirmation of site recovery.

Detailed mitigation and recovery plans will be specified in the future when formulating
operational plans based on specific site conditions and then, as shown in Section 5.6.2 (3),relevant accident response techniques will be developed.

5.6 Summary and future perspective
5.6.1 Key findings of this chapter
This chapter focussed on the assessment of potential radiological consequences during theoperational period of a repository, based on the repository designs and operational processes
developed in Chapter 4 and with reference to guidelines and regulations for safety assessmentof similar facilities handling radioactive materials. Key findings in this chapter are
summarised as:

In assessment of normal operation scenarios, the annual effective dose to the public was
below the 50 μSv/y target, due to appropriate design of shielding and distance to thefacility boundary.
Abnormal condition scenarios were assessed using event-trees describing the transitionof perturbations to abnormal states that may damage the waste and cause loss ofcontainment. The assumed abnormal operations are classified as drops, fires,
explosions, loss of power and other equipment malfunctions. The robustness of waste
containment was assessed under the conservative assumption that design countermeasures fail.
5-25



Even under abnormal operational conditions, the containment system described wasshown to be sufficiently robust, given the fundamental safety measures incorporated
into the design, such as limitation of handling heights. No scenario considered in thecurrent study led to damage that would result in RN release.

Based on the above results, the likelihood of occurrence of an accident that releases RNsfrom the facility design developed in Chapter 4 is assessed to be extremely small.
Nevertheless, the preliminary stage of this operational safety assessment must also be noted.
Several issues should be further explored, as noted in Section 5.6.2, and the assessment will
also need to be updated when the repository designs are further developed.
5.6.2 Future perspective
To improve the confidence in operational safety assessment, key R&D issues noted in Table
5.6-1 will be considered further.
Table 5.6-1 Major future technical efforts related to operational safety

Classification

Main action items

Operational safetyassessment
Scenario construction••

Construction of scenarios that include complex event chains,such as common-mode failure
Hazard database update

Development ofoperational safetyassessment technology
Development ofaccident response
technology••

Acquisition of important data to support safety evaluation
Evaluation methodology for complex events•

Examine measures to deal with accidents and associatedrecovery(1) Development of scenarios for operational safety assessment
Although illustrative abnormal operation scenarios were developed in this chapter, assuring
comprehensiveness of scenarios is key to improving the confidence in assessments required
for tailoring repository design to given site environments. In particular, single failure modes
were investigated, while multiple failure modes or common mode failure scenarios need to be
developed and analysed to support detailed design of both the repository and associatedoperational processes. Furthermore, specific perturbations with potentially large impacts (e.g.
flooding) have been identified that would justify a more thorough assessment.
The procedure of scenario development needs to be coupled to continuous reviews ofincidents in relevant environments, to reduce as much as possible the risk of completely
unexpected abnormal states (sometimes termed “black swans”), which is a lesson learned
from many abnormal states in the past. A review of relevant international and domestic
experience of abnormal states will be performed and results recorded in a database, as part ofa structured operational safety knowledge management system. The combination of such a
database and constant review of details of scenarios will help to ensure theircomprehensiveness.
Finally, it is noted that conventional accidents during repository construction are very
important hazards to consider in repository development, but this is not formally seen as acomponent of operational safety from the nuclear regulatory perspective covered in this
Chapter. Instead, actions to handle requirements related to workers protection are identified asa future action in Section 4.8.3 (2) (ii).
5-26

(2) Development of analytical methods for operational safety assessment
The mechanical analyses of drops and collisions, together with the thermal analyses offires, are critical to the assessment of operational robustness presented in this chapter.
Verification of the analytical codes, quality assurance of models and data used and thevalidation of analytical results are all required. In particular, for validation, full scale
simulation of relevant drops and fires need to be carried out under relevant conditions.
Although this chapter focussed on radiological safety, conventional safety of any workers
present in radiation-controlled areas will also need to be assessed for the abnormal conditions
noted. If a fire occurs underground, for example, damage to facilities and equipment may besignificant, leading to risks to workers. For such cases, the applicability of analytical models
of fires underground will be investigated, based on best international practice.
(3) Development of mitigation and recovery techniques
As shown in Section 5.5, decontamination plans will be developed for scenarios of
perturbations that could lead to RN contamination. Lessons learned from accidents, such as
those in the WIPP facility [20], will be used (although it is recognised that site and repository
conditions will be different in Japan). For example, the design of the underground facility
layout and key supporting infrastructure specifically aimed to minimise the impact of such
accidents may be effective as a mitigation measure. Nevertheless, it will be necessary toestablish and document recovery procedures and to develop required technology, especiallyfor actions carried out by remote control.
In addition, in terms of the treatment of any waste subject to perturbed conditions, criteria
to determine the necessity of recovery or required treatment thereafter will be considered in
the future.

5-27Supporting Reports (SRs)SR 5-1 Evaluation of direct irradiation and skyshine from the HLW surface facility
SR 5-2 Evaluation of direct irradiation and skyshine from the TRU waste surface facility
SR 5-3 Evaluation of HLW overpack impacts
SR 5-4 Evaluation of TRU waste package impacts
SR 5-5 Assessment of fuel fires
SR 5-6 Evaluation of tyre fires
SR 5-7 Impact of a tyre fire on HLW overpacks
SR 5-8 Impact of a tyre fire on TRU waste packages
SR 5-9 Calculations of physical impact by a methane gas explosion
SR 5-10 Evaluation of temperature rise in the HLW buffer store due to loss of ventilation

5-28

References[1] IAEA (2013): The safety case and safety assessment for the predisposal management ofradioactive waste, General Safety Guide, IAEA Safety Standards Series, No. GSG-3.
[2] Yamashina, K., Suzuki, S. and Kubota, S. (2017): Pre-closure safety assessment in safetycase development, the proceedings of IHLRWM 2017, Charlotte, NC, April 9-13, 2017, pp.
741-749.
[3] CRIEPI (Central Research Institute of Electric Power Industry) (1990): Drop tests forassessing the integrity of vitrified waste, Research material of CRIEPI, U90904 (in Japanese).
[4] Mano, T. and Otaka, H. (1991): Drop tests of vitrified waste, PNC TN1410 91-035 (inJapanese).
[5] NFT (Nuclear Fuel Transport): Spent fuel transport vessels,
http://www.nft.co.jp/equipment/equipment1.html (accessed 2020-12-24) (in Japanese).
[6] JAEA (Japan Atomic Energy Agency) (2018): Report on technological developments forassessing and verifying disposal systems, FY2017 Technological development project forgeological disposal of high-level radioactive waste (in Japanese).
[7] Kikura, H., Watari, M., Mizobuchi, H. and Takatsuki, H. (2013): Fundamentals of safe
transportation of nuclear fuels, revised, Aritomi, M. and Shimura, S. (supervisors), ERC
Publishing Company (in Japanese).
[8] IAEA (2012): Regulations for the safe transport of radioactive material 2012 Edition,
Specific Safety Requirement, IAEA Safety Standards, No. SSR-6.
[9] Japanese Standards Association (2017): JIS handbook, Iron and steel I (Glossary/
qualifications and certification/ inspections and tests/ steel for special purposes/ cast andforged products/ others) (in Japanese).
[10] CRIEPI (Central Research Institute of Electric Power Industry) (1980): Safety
demonstration test for dropping waste package (in Japanese).
[11] Japanese Standards Association (2017): JIS handbook, Iron and steel II (sticks, molds,
plates and belts/ steel pipes/ wires and secondary products) (in Japanese).
[12] Brousse, B. (2005): Technical report: part 2 fire safe design, Thematic network FIT - Firein tunnels.
[13] Babrauskas, V. (2003): Ignition handbook, Fire Science Publishers.
[14] Architectural Institute of Japan (2008): AIJ (Architectural Institute of Japan)
recommendations for fire resistant design of steel structures, Maruzen Publishing.

5-29

[15] Japan Society of Mechanical Engineers (2014): JSME mechanical engineers’ handbook,
DVD-ROM, Maruzen Publishing (in Japanese).
[16] AESJ (Atomic Energy Society of Japan) (2011): Quality management report onradioactive waste for geological disposal: Functions of and requirements for vitrified highlevel radioactive waste necessary for final disposal, Special Technical Committee for the
Quality Management of Radioactive Waste Subject to Geological Disposal (in Japanese).
[17] Kawamura, K., Koide, M. and Matsushita, K. (2014): Evaluation of temperaturedependence of viscosity and crystallization time for Japanese HLW glass, Transactions of theAtomic Energy Society of Japan, Vol. 13, No. 1, pp. 27-33 (in Japanese).
[18] Japan Institute of Metals and Materials (1985): Modern metallurgy course: Materials
edition, Vol. 4, Steel materials, Maruzen Publishing (in Japanese).
[19] Suzuki, S., Kubota, S., Hyodo, H. and Fujihara, H. (2015): Assuring operational safety ofa Japanese geological repository, Proceedings of International Nuclear Fuel Cycle Conference(Global 2015), Paris, France, 20-24 September 2015, Paper 5290, pp. 2286-2291.
[20] U.S.DOE (2014): Waste Isolation Pilot Plant recovery plan.
[21] U.S.DOE (2014): Accident investigation report, Underground salt haul truck fire at the
Waste Isolation Pilot Plant February 5, 2014.
[22] Savannah River National Laboratory (2015): Waste Isolation Pilot Plant technicalassessment team report, SRNL-RP-2014-01198.
[23] U.S.DOE (2014): Accident investigation report, Phase 1, Radiological release event at
the Waste Isolation Pilot Plant, February 14, 2014.
[24] U.S.DOE (2015): Accident investigation report, Phase 2, Radiological release event at
the Waste Isolation Pilot Plant, February 14, 2014.
[25] U.S.DOE (2018): WIPP recovery: Plans and reports,
http://wipp.energy.gov/wipprecovery-plans_reports.asp (accessed 2020-03-24).
[26] U.S.DOE (2016): Safety evaluation report, DOE/WIPP 16-3565.

5-30

6 POST-CLOSURE SAFETY ASSESSMENT
In this chapter, the assessment will be developed based on representative Site Descriptive
Models (SDMs) for host rocks that might result from site selection, as established in Chapter
3, together with tailored repository designs, as developed in Chapter 4. Based on
consideration of the latest developments in science and technology, this assessment forms thebasis for discussion of safety after repository closure. This aims to confirm past genericdemonstrations of the fundamental feasibility of safe geological disposal in Japan and show
that, when candidate sites are identified, the scientific and technological foundation forcarrying out the required safety assessment is in place.
In Section 6.1, as a prerequisite for the basic safety assessment concept described in
Chapter 2, the reference inventory of radioactive waste and the design specifications of therepository are presented. The basic framework of risk-informed safety assessment is thendescribed, utilising a classification of scenarios based on their probability and setting
corresponding dose and risk targets for specified assessment timescales. Section 6.2 explains
the systematic methodology of evaluation developed, based on an internationally recognised
approach to safety assessment. Using this methodology, in Section 6.3, post-closure safety ofrepositories tailored to the SDMs is examined, illustrating the development of representative
scenarios and setting analysis cases to quantify their radiological consequences. In Section 6.4,
the analytical models, calculation codes and data sets used are explained and calculated doses
presented. In Section 6.5, the method of stylisation and the calculated results of dose and risk
are described for the special case of scenarios involving human intrusion. Finally, in Section
6.6, a summary of the assessments is presented and required future efforts to improve suchassessments are identified.

6.1 Basic framework of safety assessment
6.1.1 Safety assessment procedure
The methodology of safety assessment is well established, based on research anddevelopment in advanced national programmes and general guidance provided byinternational organisations [1] [2]. The work flow used here is summarised as follows:Characterisation of the geological environment conditions at a site expected to possess
the required safety functions (Chapter 3); repository design considering suchconditions (Chapter 4); leading to creation of a bounding range of scenarios of future
evolution.

Modelling to quantitatively analyse the behaviour of the repository system (alreadypartly discussed in Chapter 4, but further assessed in this Chapter), based on defined
scenarios and associated parameter sets (these are referred to as analysis cases in thisreport).



Based on the analysis cases, selecting analytical models and datasets to quantify
radionuclide (RN) migration and release to the biosphere. Determination of theresulting post-closure radiological effects (doses, risks and other factors) andcomparison with safety regulations as the basis for assessment of safety.

The timescales for evaluation (see Section 6.1.4), as required for safety assessment, alongwith target dose and risks limits for specific classes of scenario (see Section 6.1.5), provide
6-1

the standards for discussing safety and are generally based on established regulations. As
noted in Chapter 2, safety regulations for geological disposal in Japan will be established in
the future. In this report, therefore, preliminary assumptions are made based onrecommendations by international organisations and safety regulations for similar facilities inJapan.
In the H12 and TRU-2 reports, the geological environments commonly observed in Japanare represented generically by typical examples of crystalline and sedimentary rocks. The
required demonstrations of fundamental feasibility for both HLW and TRU waste are assured
by highly robust engineered barriers and a sufficiently conservative safety evaluation.
However, in this report, the safety assessment methodology is intended to be a template forthat carried out during future site selection (i.e. to be further developed in coming stages). For
this reason, more realistic SDMs for representative host rocks that are widely distributed inJapan were developed in Chapter 3. Site-specific repository designs take specific SDM
features into consideration, supported by an analysis that treats their characteristics asrealistically as practicable, while still maintaining overall conservatism in the safetyassessment. For this reason, based on R&D progress since the H12 and TRU-2 reports,
special consideration is given to the following points:In order to appropriately deal with spatial and temporal uncertainty in the behaviour of
the repository, scenarios have been developed where both probability of theiroccurrence and the magnitude of associated impacts to be assessed are decoupled. The
framework of the safety assessment thus introduces risk-informed thinking as part ofthe evaluation.



In order to ensure that the repository design is properly adapted to the given SDM
characteristics, the engineered barrier system (EBS) components, repository designand layout are tailored to the spatial distribution of safety-relevant features, inparticular faults, fractures and other major structures in the host rock (as described in
Chapter 4). Scenarios, models and datasets to assess groundwater flow and RN
migration behaviour aim to reflect a reasonable description of the behaviour of therepository.



In order to ensure both the transparency and reliability of the safety assessment results,
efforts are made to improve traceability of the scenario development process,facilitating assessment of comprehensiveness of treatment and adequacy of models
and datasets.

6.1.2 Prerequisites for safety assessment
Safety assessment prerequisites, which provide a foundation for all subsequent evaluations,
include the SDMs, associated repository designs and the inventory of radioactive waste.

(1) Geological repositories
In Chapter 3, SDMs are defined for plutonic rocks, together with Neogene and PreNeogene sediments. For each, the SDM illustrates the geological setting along with associateddistributions of thermal, hydraulic, mechanical and geochemical characteristics. In particular,
with regard to groundwater chemistry characteristics that affect both evolution of the EBS andRN release/migration, two types of model waters are defined (hereinafter referred to as low
6-2salinity and high salinity). In Chapter 4, two variants of EBS design for the disposal of highlevel radioactive waste (HLW) are presented, an upgraded option for the verticalemplacement assessed in H12 (“H12V”) and an alternative horizontal emplacement of a prefabricated EBS module (“PEM”). For TRU waste disposal, emplacement vault geometry is
adapted to the rock mechanical conditions of the SDM, with two variants of the disposal
waste package (A and B: see Section 4.4.2 (2) (ii)). These variants, summarised in Table 6.1-1,
are referred to as the plutonic, Neogene and Pre-Neogene repositories.
Table 6.1-1 Potential repository host rocks subject to safety assessment
Geological environment model

Repository design(co-location of HLW and TRU waste)
H12V

Plutonic repository: Low salinity
 High salinity

HLW
PEM
TRU (Gr.1, Gr.2, Gr.3,
Gr.4H, Gr.4L)

Package A
Package B
H12V

Neogene repository: Low salinity
 High salinity

HLW
PEM
TRU (Gr.1, Gr.2, Gr.3,
Gr.4H, Gr.4L)

Package A
Package B
H12V

Pre-Neogene repository: Low salinity
 High salinity

HLW
PEM
TRU (Gr.1, Gr.2, Gr.3,
Gr.4H, Gr.4L)

Package A
Package B

Gr.: Group

Table 6.1-2 shows the components of the repository considered when conducting safetyassessment, although not all are explicitly assessed at the present time.

6-3

Table 6.1-2 Repository components considered in the safety assessment
Component

H12V
DT

HLW: Glass waste matrix
HLW: Overpack
HLW: PEM handling shell

MT

CT

AR AS

✓
✓

DT

PEM

TRU
(Gr.1, Gr.2, Gr.4HD, Gr.4HH)

TRU(Gr.3, Gr.4LD, Gr.4LC)

MT

DT

DT

CT AR

MT

CT

AR

AS

MT

CT

AR AS

✓
＊

✓
＊

✓
＊

✓
✓

✓
✓
✓

✓
✓
✓

✓
✓
✓

✓

✓

✓

✓

✓

✓
✓

✓
✓

✓
✓

✓

✓

✓

✓

✓

✓

✓

✓

✓
✓
✓
✓
✓
✓

✓
✓

TRU: Infill for waste package✓✓

TRU: Infill between waste packages
TRU: Structural framework

✓
＊

✓
＊

✓
＊

✓
✓
✓
＊

HLW/TRU: Buffer
TRU: Waste matrix
TRU: Waste package container✓✓

Tunnel/vault: Grout
Tunnel/vault: Steel support

✓
＊

✓
＊

✓
＊

✓
＊

✓
＊

✓
＊

✓
＊

✓
✓
✓
＊

Tunnel/vault: Shotcrete
Tunnel/vault: Rock bolt
Tunnel/vault: Concrete liner

✓
✓

✓
✓

✓
✓
✓

✓
✓
✓

✓
✓
＊

✓
✓
✓

✓
✓
✓

✓
✓
✓

＊

✓
✓
✓
＊

✓
✓
✓

Tunnel/vault: Invert concrete

✓
✓
✓
＊

✓

✓

✓

✓

✓

✓

✓

Tunnel/vault: Central drain
Tunnel/vault: Drain
Tunnel/vault: Waterproof sheet
Tunnel/vault: Backfill material
Plug: Mechanical plug
Plug: Permeable layer (crushed rock)
Plug: Filter material (glass fibre)
Plug: Drain pipe
Plug: Embedded form (glass fibre reinforced cement)
Plug: Hydraulic plug (mixed bentonite and silica sand)

✓

✓

✓

✓

✓
✓
✓

✓
✓

✓
✓

✓
✓

✓

✓

✓

✓

✓

✓
✓
✓
✓

✓
✓

✓
✓

✓
✓

✓

✓

✓

✓

✓
✓
✓
✓

✓
✓

✓

✓

✓

✓
✓

✓
✓

✓
✓

✓

✓

✓

✓

✓

✓
✓
✓
✓

✓
✓

✓
✓
✓
✓

✓
✓

✓

✓

✓

✓
✓

✓

✓

DT-Disposal tunnel; MT-Main tunnel; CT-Connecting Tunnel; AR-Access ramp; AS-Access shaft, ✓: Constituent elements of repository considered for all host rocks. *: Considered only for
Neogene sediments. Gr.4HD: Gr.4H (drum), Gr.4HH: Gr.4H (container as yet not specified – MHHRW – miscellaneous higher heat reprocessing waste), Gr.4LD: Gr.4L (drum), Gr.4LC: Gr.4L(box container).

6-4

Table 6.1-2 includes components introduced in the construction and operation of therepository, which are also included in the Features, Events and Processes (FEPs) (NUMO
FEP list) described in Section 6.3.2 (1). The three repositories are illustrated in Figures 6.1-1,
6.1-2 and 6.1-3. In this report, explanation is focussed on the plutonic rock repository, withdirect reference to the other rocks in cases where there are clear differences in features related
to the geological environment and associated repository design.

6-5

Figure 6.1-1 Illustration of components of the plutonic rock repository.
Top left HLW (H12V), top right HLW (PEM), bottom left TRU (Gr.2), bottom right TRU (Gr.3)

6-6

Figure 6.1-2 Illustration of components of the repository of Neogene repository.
Top left HLW (H12V), top right HLW (PEM), bottom left TRU (Gr.2), bottom right TRU (Gr.3)

6-7

Figure 6.1-3 Illustration of components of the Pre-Neogene repository.
Top left HLW (H12V), top right HLW (PEM), bottom left TRU (Gr.2), bottom right TRU (Gr.3)

6-8

(2) Radionuclides to be evaluated and radioactivity inventory
The specifications of radioactive waste allocated for geological disposal in Japan are given
in Chapter 2. These include a wide range of fission products and activation product RNs,
many of which do not need to be taken into account when considering long-term radiologicalinfluences due to short half-life and/or low concentration. Here, the process of selecting RNs
for post-closure release and transport analysis will be described. For further details, see
Supporting Report 2-3. Selection of the RNs that are considered in the biosphere evaluation
when calculating dose conversion factors is also described (see Supporting Report 6-1).
(i) Selecting relevant RNs for HLWWith reference to domestic and foreign safety assessments since the H12 report [3] [4] [5][6] [7] [8] [9], there is consensus that post-closure analysis focuses on a limited number ofRNs, having half-lives longer than a certain period, provided that short-lived parent RNs indecay chains are included in the source term and short-lived daughters in the biosphereassessment. The selection of such RNs depends on the safety assessment scenarios involved(see Section 6.1.5 (2)).
Specifically, with reference to selection methods in other countries, RNs having a half-lifeof more than 6 months were initially extracted from the 1,252 RNs listed by ICRP
(International Commission on Radiological Protection) [10]. For scenarios involving RN
transport via groundwater, target RNs are based on those selected both in the H12 report andby implementing agencies of other countries that dispose of HLW and/or spent fuel (SF).
Primary concern in safety assessments focuses on RNs that are highly soluble with low
sorption onto either engineered barriers or geosphere transport paths [11]. Thus, C-14, Cl-36,
and I-129 were newly added to the list of potentially important RNs to be included in theanalysis. Pd-107 and Sm-151 were excluded due to expected low importance, with reference
to the RNs dominating doses assessed in recent studies reported by other countries. However,
it is understood that the selection of RNs is preliminary and might need reassessment incoming studies.
For low probability and human intrusion scenarios that may occur at an early stage (seeSections 6.4.3 and 6.5), the RNs Sr-90 and Cs-137, which have relatively short half-lives butvery high initial inventories, were added. A decision was made to add also RNs of theactinide decay series with high importance in terms of internal exposure.
Since HLW disposal occurs only after a period of 30 to 50 y interim storage after
production, the radioactivity of 40,000 packages (reference inventory) at the time of closureof the repository will represent a distribution of actual ages; however this distribution is
uncertain at present. In this report, as described in Section 4.2.2 (1), the repository is designedon the premise of a 50 y HLW interim storage. Furthermore, conservatively, the time fordecay during the operational period of the repository is not considered. Thus, post-closure
safety assessment assumes an age of 50 y for all waste packages. However, considering that
containment by the overpack is > 1 ky, the inventory of RNs for safety assessment after
overpack failure is relatively insensitive to this value.
Table 6.1-3 shows the radioactivity per glass block for the selected RNs after 50 y interim
storage (see Supporting Report 2-3).
Figure 6.1-4 shows the decay chains for the actinides considered in the release andmigration analysis.
6-9

Table 6.1-3 Selected radionuclides and inventories for safety assessment (HLW)
Nuclide
C-14
Cl-36
Se-79
Sr-90
Zr-93
Nb-93m
Nb-94
Tc-99
Sn-126
I-129
Cs-135
Cs-137
Pb-210Ra-226
Ra-228
Ac-227
Th-228
Th-229Th-230Th-232
Pa-231
U-232
U-233

Half-life (y)3

5.7x103.0x105
3.0x1052.9x101.5x106
1.6x102.0x1042.1x105
2.3x105
1.6x1072.3x106
3.0x102.2x101.6x1035.8
2.2x10
1.9
7.3x1037.5x104
1.4x1010
3.3x104
6.9x101.6x105

Inventory (Bq/unit)
8

1.2x104.8x1083.2x1098.2x1014
7.2x1010
6.4x1010
1.5x1085.2x1011
1.1x1010
3.8x1071.8x1010
1.2x1015
7.6x1021.6x103
3.3
8.8x1045.6x106
1.1x1048.7x104
3.3
1.1x1055.5x106
3.0x106

Nuclide
U-234
U-235
U-236
U-238
Np-236
Np-237
Pu-236
Pu-238
Pu-239Pu-240Pu-241
Pu-242
Pu-244Am-241
Am-242m
Am-243
Cm-243
Cm-244
Cm-245
Cm-246
Cm-247
Cm-248Half-life (y)52.5x107.0x1082.3x1074.5x109
1.5x1052.1x106
2.9
8.8x102.4x1046.6x1031.4x103.8x1058.0x1074.3x102
1.4x1027.4x1032.9x101.8x108.5x1034.8x1031.6x1073.5x105

Inventory (Bq/unit)
9.8x1073.0x106
4.6x1073.9x1072.3x105
1.4x1010
2.9x1045.4x1011
6.8x1010
3.3x1011
2.2x1012
4.2x1081.3x1023.5x1013
2.0x1011
8.1x1011
1.9x1011
1.4x1013
1.7x1010
2.8x109
1.1x104
3.4x104

4n chain

4n+1 chain

4n+2 chain

4n+3 chain
Figure 6.1-4 Relevant actinide decay series for RN migration analysis(ii) Selecting relevant RNs for TRU waste
The process for TRU waste is basically the same as HLW, but complicated by muchgreater uncertainties regarding the characteristics and volumes of future arisings [12] [13].
For this reason, it was decided to select the RNs to be evaluated based on the latest
information on those targeted for the TRU-2 safety assessment [14].
A further complication for TRU waste is its inherent variability in RN concentrations,which is much greater than for vitrified HLW. Therefore, it is also difficult to justify referring
6-10

to the RNs of interest in the safety assessment carried out for long-lived L/ILW in othercountries (which has similarities to Japanese TRU waste), as done in Section 6.1.2 (2) (i)above. Therefore, in this report, in addition to the RNs selected for the TRU-2 groundwater
release scenarios and other perturbation scenarios that involve release via groundwater (e.g.,human intrusion), it was decided to add isotopes of the actinide decay series with highimportance for internal exposure.
Currently, there is no specified requirement for interim storage of TRU waste prior todisposal. Thus, the TRU-2 assumption of 25 y storage after production is taken as a reference,
resulting in an approximately one order of magnitude decrease in heat output, due to decay ofshort-lived isotopes. Specifically, the radioactivity inventory at the time of closure used in thisreport considers radioactive decay for 25 y after production [13] for all waste packages. As
some existing TRU waste has already been stored for over 25 y, actual radioactivity
inventories may be significantly lower. Table 6.1-4 shows the selected RNs1 and estimatedvalues of their inventory for each waste group (see Supporting Report 2-3). Gr.4 is classified
into two groups: low heat (4L) and high heat (4H) based on thermal output; these in turn are
further sub-classified by waste container (see Sections 4.4.2 and 4.5.4, giving 4LD and 4LC,
4HD and 4HH; Table 6.1-2). For RN migration analysis, the radioactivity inventory is
separately specified for each of these TRU waste groups, so that the dose can be calculated
individually. The same actinide decay chains are used as for vitrified HLW (Figure 6.1-4).

1 Pd-107 was excluded from the RNs evaluated for HLW, but TRU is more diverse and, since this isotope is subject tosafety evaluation in the TRU-2 report, it is also considered in this report.

6-11

Table 6.1-4 Selected radionuclides and inventories for safety assessment (TRU waste) (1/2)
Group 2
Group classification

Group 4

Group 1

Hulls and
ends

CSD-C

CSD-B

Abbreviation

Gr.1

Gr.2

Gr.2

Gr.2

Gr.3

Container

200 l drum

Stainless steel
canister

Stainless steel
canister

Stainless steel
canister

1,673
335

27,522
5,342

3,780
734

30
7

Number of units
Volume (m3)
Nuclide
C-14
Cl-36
Co-60
Ni-59
Ni-63
Se-79
Sr-90
Zr-93
Nb-94
Mo-93
Tc-99
Pd-107
Sn-126
I-129
Cs-135
Cs-137
Pb-210Ra-226
Ra-228Group 3Half-life (y)35.7 x 10
3.0 x 105
5.3
1.0 x 105
1.0 x 102
3.0 x 105
2.9 x 10
1.5 x 106
2.0 x 104
4.0 x 103
2.1 x 105
6.5 x 106
2.3 x 105
1.6 x 107
2.3 x 106
3.0 x 102.2 x 10
1.6 x 103
5.8

Low heat

High heat

Gr.4L

Gr.4H

Gr.4LD

Gr.4LC

Gr.4HD

Gr.4HH200 l drum

200 l drum

Box container

200 l drum

MHHRW

27,441
5,492

24,076
4,845

287
884

4,587
921

1,188
458

3.4 x 109
3.6 x 107
3.8 x 1015
5.9 x 1013
7.4 x 1015
1.3 x 1011
1.2 x 1016
7.6 x 1012
3.9 x 107
5.4 x 107
3.4 x 1012
3.1 x 1010
2.3 x 1011
1.1 x 108
3.1 x 1011
3.8 x 1016
3.1 x 105
1.5 x 106
1.1 x 102

2.4 x 109
1.1 x 106
1.5 x 1010
7.2 x 105
8.2 x 107
1.8 x 109
1.6 x 1014
1.6 x 1010
8.0 x 105
2.5 x 106
5.6 x 1010
4.0 x 109
3.3 x 109
1.0 x 107
1.9 x 109
2.4 x 1014
2.8 x 103
1.3 x 104
2.5

5.3 x 1011
2.2 x 104
3.1 x 1012
8.2 x 105
9.5 x 107
3.9 x 1011
3.5 x 1016
1.7 x 1012
1.2 x 108
5.3 x 108
1.2 x 1013
1.0 x 1011
7.0 x 1011
1.5 x 109
4.1 x 1011
5.0 x 1016
2.8 x 105
1.3 x 106
1.8 x 102

2.2 x 1012
1.3 x 107
1.3 x 1013
4.1 x 108
4.8 x 1010
1.6 x 1012
1.5 x 1017
7.3 x 1012
5.2 x 108
9.0 x 107
9.7 x 1013
4.2 x 1011
2.9 x 1012
3.0 x 109
1.7 x 1012
2.1 x 1017
7.2 x 103
3.6 x 104
5.7

Radioactivity (Bq)
0.0
0.0
1.3 x 105
5.5 x 103
7.2 x 105
3.9 x 105
3.6 x 1010
1.8 x 106
3.3 x 106
3.7 x 100.0
1.1 x 105
7.4 x 105
5.9 x 1013
4.3 x 105
5.3 x 1010
2.5 x 10-1
1.1
2.2 x 10-414

4.8 x 109.2 x 1012
5.1 x 1016
7.5 x 1015
9.5 x 1017
2.2 x 1012
1.9 x 1017
3.0 x 1014
2.6 x 1015
5.6 x 1013
6.5 x 1014
5.5 x 1011
3.8 x 1012
1.6 x 1011
2.2 x 1012
2.8 x 1017
1.6 x 105
8.1 x 105
8.6 x 10214

1.1 x 100.0
2.2 x 1016
0.0
1.7 x 1017
1.0 x 1012
1.2 x 1017
6.8 x 1013
0.0
0.0
4.3 x 1013
0.0
0.0
1.0 x 1011
1.3 x 1012
1.4 x 1017
5.1 x 104
3.0 x 105
2.9 x 10-1

7

7.5 x 10
3.0 x 108
6.8 x 109
8.3 x 1080.0
2.8 x 1010
4.2 x 1014
0.0
0.0
0.0
7.2 x 1011
0.0
0.0
3.9 x 107
2.8 x 1010
1.5 x 1015
8.9 x 10
4.4 x 102
3.6 x 10-1

6-12

7.4 x 1012
3.3 x 1011
4.6 x 1011
4.8 x 108
5.6 x 1010
6.7 x 109
6.8 x 1014
3.1 x 1010
1.9 x 106
8.1 x 106
3.9 x 1011
2.5 x 109
1.1 x 1010
7.2 x 1011
7.0 x 109
8.6 x 1014
3.0 x 104
1.4 x 105
2.4 x 10

Table 6.1-4 Selected radionuclides and inventories for safety assessment (TRU waste) (2/2)
Group 2
Group classification
Abbreviation

Ac-227
Th-228
Th-229Th-230Th-232
Pa-231
U-233
U-234
U-235
U-236
U-238
Np-237
Pu-238
Pu-239Pu-240Pu-241
Pu-242
Am-241
Am-242m
Am-243
Cm-244
Cm-245

CSD-C

CSD-B

Gr.1

Gr.2

Gr.2

Gr.2

Stainless steel
canister
27, 522
5,342

Stainless steel
canister
3,780
734

Stainless steel
canister
30
7

1,673
335

Number of units
Volume (m3)
NuclideHulls andends

200 l drum

Container

Group 4

Group 1

Group 3
Gr.3

Half-life(y)2.2 x 10
1.9
7.3 x 103
7.5 x 104
1.4 x 1010
3.3 x 104
1.6 x 105
2.5 x 105
7.0 x 108
2.3 x 107
4.5 x 109
2.1 x 106
8.8 x 10
2.4 x 104
6.6 x 103
1.4 x 10
3.8 x 105
4.3 x 102
1.4 x 102
7.4 x 103
1.8 x 108.5 x 103

Low heat

High heat

Gr.4L

Gr.4H

Gr.4LD

Gr.4LC

Gr.4HD

Gr.4HH200 l drum

200 l drum

Box container

200 l drum

MHHRW

27,441
5,492

24,076
4,845

287
884

4,587
921

1,188
458

5.3 x 106
9.5 x 10
1.2 x 105
2.8 x 108
1.6 x 102
1.7 x 107
8.6 x 107
1.3 x 1012
3.2 x 1010
1.3 x 1011
1.6 x 1011
7.0 x 1011
3.5 x 1015
8.7 x 1014
1.0 x 1015
3.8 x 1016
2.4 x 1012
4.8 x 1015
4.6 x 1013
1.2 x 1013
5.0 x 1014
1.4 x 1011

2.9 x 104
2.1
9.6 x 102
2.5 x 106
3.7
9.2 x 104
6.1 x 105
1.2 x 1010
1.7 x 108
3.0 x 109
2.5 x 109
3.8 x 109
2.9 x 1013
3.0 x 1012
4.8 x 1012
1.6 x 1014
1.9 x 1010
2.1 x 1013
3.3 x 1012
2.2 x 1011
9.7 x 1012
2.7 x 109

3.9 x 106
1.5 x 102
1.0 x 105
2.5 x 108
2.7 x 102
1.3 x 107
7.2 x 107
1.2 x 1012
2.4 x 1010
2.2 x 1011
2.0 x 1011
5.2 x 1011
2.6 x 1015
4.8 x 1014
6.2 x 1014
3.6 x 1016
2.0 x 1012
3.9 x 1015
5.8 x 1014
1.5 x 1013
6.7 x 1014
1.8 x 1011

6.4 x 1044.8
2.7 x 105
7.8 x 106
8.3
2.1 x 105
2.3 x 108
5.4 x 1010
3.9 x 108
6.7 x 109
5.5 x 109
2.1 x 1012
7.7 x 1014
8.6 x 1012
3.6 x 1013
6.3 x 1014
9.4 x 1010
4.7 x 1015
4.4 x 1015
1.1 x 1014
5.0 x 1015
1.4 x 1012

Radioactivity (Bq)
2.5
1.9 x 10-4
5.5 x 10-4
2.2 x 102
3.2 x 10-4
8.1
7.8 x 10-1
1.0 x 106
1.5 x 104
2.6 x 105
2.2 x 105
1.7 x 104
2.3 x 109
2.4 x 108
3.7 x 108
2.9 x 1010
1.6 x 106
3.0 x 109
6.6 x 106
1.8 x 107
8.1 x 1082.2 x 1057

1.0 x 10
7.4 x 102
1.3 x 107
1.8 x 108
1.3 x 103
3.2 x 107
5.4 x 109
1.1 x 1012
6.1 x 1010
1.0 x 1012
8.6 x 1011
1.4 x 1012
8.9 x 1015
9.7 x 1014
1.5 x 1015
1.0 x 1017
6.4 x 1012
1.1 x 1016
2.3 x 1013
6.9 x 1013
3.1 x 1015
8.6 x 1011

2

9.8 x 10
2.3 x 10-1
9.2 x 104
8.1 x 107
5.4 x 10-1
4.4 x 103
7.8 x 107
6.8 x 1011
1.7 x 107
8.7 x 108
2.5 x 104
7.4 x 1011
8.7 x 1015
6.8 x 1014
1.2 x 1015
8.5 x 1016
6.5 x 1012
8.1 x 1015
0.0
1.7 x 1015
2.9 x 1015
0.0

3

9.4 x 10
3.0 x 10-1
5.8 x 104
9.1 x 104
5.2 x 10-1
3.0 x 104
4.9 x 107
5.3 x 108
5.7 x 107
4.2 x 108
1.3 x 109
4.5 x 1011
3.5 x 1012
1.3 x 1012
1.7 x 1012
3.3 x 109
4.5 x 109
2.9 x 1013
0.0
3.0 x 1011
5.1 x 1012
3.5 x 109

6-13

2.9 x 105
2.0 x 109.3 x 103
2.7 x 107
3.5 x 109.2 x 105
6.2 x 106
1.4 x 1011
1.7 x 109
2.8 x 1010
2.4 x 1010
4.7 x 1010
5.1 x 1014
5.7 x 1013
8.9 x 1013
3.1 x 1016
3.7 x 1011
2.4 x 1015
7.1 x 1011
5.5 x 1011
1.3 x 1012
2.1 x 109

(iii) Selecting relevant RNs for biosphere assessment
The biosphere assessment in this report treats RN migration near the surface of the Earthand assesses resultant radiation exposure to humans in the same manner as in the H12 report.
From defined hydraulic characteristics of surface waters and relevant food chains, it ispossible to assess radiation exposure to human beings from a defined release from thegeological environment (at the geosphere biosphere interface - GBI). It is assumed thattransfer of RNs within biosphere exposure pathways is much faster than release and transport
through the geological environment [15].
In terms of handling decay of RNs in the biosphere, previous studies have illustrated
methods for both treating daughter RNs in secular equilibrium with their parents [4] [16] [17][18] [19] or individually, with cut-off half-lives of 1 month [20] or 25 days [3] [14] [21] [22].
Based on these references, and in order to assure that all relevant RNs are considered, in thisreport, RNs with half-lives of 25 days or more were evaluated individually. Thus, in addition
to the RNs mentioned in (i) and (ii) above, decay series daughters are also considered. In thiscase, for daughters with half-lives of 25 days or more, dose conversion factors were
determined individually. For daughters with half-lives of less than 25 days, it is assumed thatthey migrate in secular equilibrium with a longer-lived parent and the effective doseconversion factor of the daughter is added to that of the parent. RNs selected for biosphereevaluation are listed in Table 6.1-5. For details of the selection process, see Supporting Report
6-1.
Table 6.1-5 Selected radionuclides for biosphere assessment
NuclideHalf-life (y)NuclideHalf-life (y)

C-14
Cl-36
Co-60
Ni-59
Ni-63
Se-79
Sr-90
Zr-93
Nb-93m
Nb-94
Mo-93
Tc-99
Pd-107
Sn-126
I-129
Cs-135
Cs-137
Pb-210Po-210Ra-226
Ra-228
Ac-227
Th-228
Th-229Th-230Th-232
Pa-231

5.7x103
3.0x105
5.3
1.0x105
1.0x102
3.0x1052.9x101.5x106
1.6x102.0x104
4.0x1032.1x1056.5x106
2.3x105
1.6x1072.3x106
3.0x102.2x103.8x10-1
1.6x1035.8
2.2x10
1.9
7.3x1037.5x104
1.4x1010
3.3x104

Pa-233
U-232
U-233
U-234
U-235
U-236
U-238
Np-236
Np-237
Pu-236
Pu-238
Pu-239Pu-240Pu-241
Pu-242
Pu-244Am-241
Am-242m
Am-243
Cm-242
Cm-243
Cm-244
Cm-245
Cm-246
Cm-247
Cm-2487.4x10-2
6.9x101.6x1052.5x105
7.0x1082.3x1074.5x109
1.5x1052.1x106
2.9
8.8x102.4x1046.6x1031.4x103.8x1058.0x1074.3x102
1.4x1027.4x1034.5x10-1
2.9x101.8x108.5x1034.8x1031.6x1073.5x105

Dose calculation is based on the geosphere release and migration analysis of RNs selected
in Sections 6.1.2 (2) (i) and (ii) above, which defines input at the GBI at specific times. The
impact of this flux is then assessed by multiplying RN fluxes in groundwater by isotope6-14

specific dose conversion factors, which are defined for relevant biosphere environments, see
Section 6.4.1 (6).

6.1.3 Treatment of spatial scales
The SDMs in Chapter 3 and the repository designs in Chapter 4 consider different spatial
scales. In the safety assessment, several spatial scales are also considered (see Chapter 2,
Section 2.2.4) in a pragmatic and consistent manner. Key scenarios considered whenevaluating the safety of the repository involve RNs in the waste dissolving into groundwaterand moving through the geological environment until they eventually reach the biosphere(See Section 6.3). In order to rigorously calculate potential doses from such scenarios,
simulation of the three-dimensional (3D) spread of RNs from each individual waste packagethrough the engineered barriers and the geological environment to the GBI is required.
The large-scale, 3D RN migration model has to consider characteristics of both theengineered and natural barriers that impact solute transport. Heterogeneities, which exist at all
spatial scales, play a key role in such migration. For groundwater flow, as discussed in
Section 3.2.2, a nested model can be used that allows the impacts of smaller features to beaveraged out at larger scales: these large-scale models provide external boundary conditions,with smaller features considered explicitly only in smaller scale models. For solute transport,
however, such simplification is not valid because the microscale features of the flow pathcompletely dominate quantification of effective retardation. Although computer technology isdeveloping rapidly, detailed simulation of RN migration on a kilometre scale while capturing
retardation-relevant features on a scale of centimetres or millimetres is not yet feasible.
Therefore, as described in Chapter 2, the concept of nested SDMs to cover multiple spatial
scales is taken over for safety assessment and the uncertainties thus introduced discussed
explicitly.
(1) Near field scale
The “near field” scale PA model includes the EBS, the disposal tunnels/vaults and asurrounding 100 m of the host rock. This is tailored to the analytical code Partridge [23],which was developed to simulate the detailed composition, shape and size of the engineered
barriers and emplacement tunnels/vaults. Information on the heterogeneity of the rock would
be relatively detailed at the time of repository implementation due to surveys during
excavation (e.g. spatial distribution and hydraulic conductivity characteristics of waterconducting fractures, as shown in Section 3.3.3 (3), and fine structure of solute transport
pathways, shown in Section 3.3.3 (4)), allowing analysis of RN release and migration in three
dimensions. This makes it possible to identify differences in RN containment performance(for example, as a function of the engineered barrier design), allowing feedback to the designteam as required. However, since the current models are relatively simple, it is recognised thatthese allow only limited feedback on some aspects of disposal concepts. For feedback on
details of the design, more elaborate near field models will certainly be needed in the future
and this is identified as an R&D goal.

6-15

(2) Panel scale
For computational reasons, a larger “panel” scale (several hundred metres by several
hundred metres) is also used with the aim of analysing RN migration from an entire disposal
panel of the HLW repository (as shown in the designs of Chapter 4) or, in the case of TRU
waste, the entire disposal zone. This utilises the output of the near field scale model, which
only covers a few tunnels, to assess influences of larger-scale facility design, such as layout ofdisposal tunnels, major access tunnels and other required infrastructure, together with a rangeof larger dimension host rock characteristics.
(3) Repository scale
At the “repository” scale (several km x several km), RN migration analysis covers theentire repository, i.e. both the TRU and HLW zones (area: ≈10 km2) plus several hundred
metres of surrounding host rock. Thus, the performance of the repository can be evaluated toassess impacts of both design features (e.g. tunnel and vault plugs) and the retardation
capacity of the host rock, over an area greater than that covered by the near field and panel
scale models.
(4) Regional scale
At the “regional” scale (tens of km x tens of km, or more depending on the setting),
releases from the repository scale model are put in context of the entire flow path from therepository to the GBI. This is based on a general methodology [24] for analysing groundwaterflow and solute transfer at such a scale, which makes it possible to estimate site-specificinformation such as RN retardation and dispersion over the entire flow path. This thenprovides evaluations of the safety functions expected for specific geological environments.
However, it is understood that this approach may not fully capture the nature of the flow field
in a heterogeneous host rock and more elaborate tools for capturing flow and RN migration
will be considered in the future.
The basic idea of nested spatial scales for RN migration analysis as described above can be
widely used, regardless of the stage of the project. This will allow investigations at specificsites to be gradually refined. It is also possible to flexibly respond to the quality and quantityof information and the evolving design of the repository, iteratively improving the resulting
safety evaluation.
In view of the current situation, including uncertainties in the SDMs and the limitations ofboth analytical models and computer capacity, a pragmatic approach to release and migration
modelling was established (see Section 6.4).
6.1.4 Treatment of evaluation period
A consideration of uncertainties that increase with time contributes towards setting thetime scale for the safety evaluation [25]. As noted in Chapter 2, safety regulations defining
the period during which quantitative evaluations are required have been established in many
countries: in many of these cases, a time frame of 1 My is adopted (e.g. [26] [27] [28]). In
Japan, safety regulations will be set in the future and, currently, no safety assessment
timescales are defined. Thus, in this report, with an aim of understanding the behaviour ofrepositories in specific SDMs in the same way as the generic H12 and TRU-2 assessments, it
6-16

was decided to conduct evaluations in order to determine the time of appearance of thecalculated maximum dose (see Supporting Report 6-3 for details). However, it must be
understood that the uncertainty in site evolution is very large at times beyond 1 My. The
doses calculated for such long time periods must not be interpreted literally, but more as
illustrations of the containment function of the repository.
6.1.5 Risk-informed safety assessment and evaluation criteria(1) Approach based on probability of occurrence of scenarios
In assessing repository safety, it is necessary to consider the various uncertainties resulting
from extremely long timescales together with the heterogeneity and large spatial scale of thegeological environment. To respond to such uncertainty, scenarios for possible future
repository evolution that are as complete as possible are developed along with an assessmentof their probability of occurrence. This results in a “risk-informed” approach, which isconsistent with that proposed by international agencies [29] and incorporated specifically into
regulations in some countries (for example, [26] [28] [30] [31]). Such a risk-informed
approach forms the basis of the safety assessment in this report.
There are two major approaches to apply risk-informed thinking: the integrated approachand the dose/stochastic approach [29]. The former quantifies the probability of occurrence of
scenarios and derives a total risk by multiplying this by the calculated radiological impact.
The latter disaggregates the probability of scenario occurrence, which allows qualitative
judgment of this parameter, from the resultant calculated radiological impact. Both options
can be used to assure radiological protection, but decoupled dose/probability presentation
provides more information to support system assessment and/or decision making [29].
Additionally, for the long-term scenarios required for radioactive waste safety assessment, it
is sometimes difficult to justify specifying quantitative probability [32].
In this report, all scenarios are quantitatively analysed based on the evaluation period
described in Section 6.1.4, however, assessing associated scenario probability is problematic,
especially given the fact sites are not specified at present. Nevertheless, methods to quantify
the probability of particular scenarios have been developed in some countries [33] [34] [35][36] while, for others, probabilities are defined in regulations [28] [30] [37]. The way inwhich scenario probability is assessed is discussed further below.
(2) Scenario classification and use in safety evaluation
In the chosen disaggregated dose/probability approach, scenarios are classified based on
their probability of occurrence, with radiological impact criteria (dose and/or risk) set for eachclass. These scenarios include “natural processes” and “human intrusion” events [2] [29] [38][39] [40].
The repository design aims to ensure that radiological impact is significantly less than thedose constraint value of 300 μSv/y recommended for the general public, or the risk2 constraint
value of 1 x 10-5/y, and that associated radiation protection optimisation has been carried out[29] [38] [42].

2Radiological risk (of cancer or heritable effects) per year from the repository = yearly dose due to scenario(Sv/y) × ICRP 2007 recommended [41] conversion coefficient of dose to risk of 0.057 (/Sv).

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Two sets of scenarios are included to cover impacts of natural perturbations. The “basescenario” assume that the repository is designed and constructed on the basis of appropriate
site selection and with appropriate consideration of the site geological conditions. Theseshould assure that the expected safety functions can be relied on and reflects the mostprobable evolution of the repository system. The alternative “variant” scenarios deviate fromthe base scenario in order to reflect various inherent uncertainties in the evolution of both thegeological environment conditions and the repository itself. The variant scenarios should havemuch lower probability of occurrence than the base scenario in order to justify assessment
against a higher dose limit. However, such probabilities have not been strictly quantified asyet.
In this report, analyses of base scenario are called base case; these together provide thefoundation for variant scenarios and associated variant analysis cases. The uncertainties
captured in the variant scenarios include those associated with the models and datasets used in
the base case and also alternative descriptions of the repository and its evolution. These arethen reflected in the variant analysis cases.
An overview of the dose and risk constraint standards set in the safety regulations in othercountries (refer to Supporting Report 6-3) shows reasonable consistency for scenarios basedon natural perturbations: with dose standards set between 10 and 300 μSv/y and risk standards
between 10-5 and 10-6/y. A risk of 10-6/y corresponds to a dose of about 20 μSv/y. Therefore,
10 μSv/y is the most severe criterion internationally.
In order to evaluate the robustness of the repository system, it is also important to consider
very low probability natural perturbations that could significantly affect barrier functions, andalso inadvertent human intrusion scenarios [2] [42]. There is, however, less consistency in theway that such scenarios are handled internationally (see Supporting Report 6-3).
As in other international assessments, “what-if scenarios” are also defined, based onassumptions or models that are extremely unlikely or physically impossible, but aim only to
improve understanding of the performance (especially robustness) of the repository [1] [2].
Thus, in this report, system evolutions based on natural processes are classified into “basescenario”, “variant scenarios” and “low probability perturbation scenarios”. In terms of
inadvertent human intrusion3, avoiding areas during site selection where mineral resources
with significant economic value exist [43] and disposal at a depth of more than 300 m as
stipulated in the Final Disposal Act, should ensure that the probability of occurrence is low.
The following expands on the definition of these scenarios and, in the absence of specificJapanese regulations, discusses relevant comparison values for doses calculated from theresultant RN releases, based on the international norms noted above.
(i) Base scenario
The base scenario reflect what is considered to be the most reasonable representation ofevolution of specific geological repositories that are sited, constructed and operated
appropriately. These form the basis for judging whether the repository fulfils the stated goalof minimising risks to the human population at all times in the future. For such scenarios, 10

3

As intentional human intrusion is not assessed in other safety assessments [2] [29] [38] [42], it is also notconsidered in this report (see Section 6.5).

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μSv/y, the most stringent standard applied in the safety regulations of other countries(Supporting Report 6-3), was set as the “target value”4.
(ii) Variant scenarios
Variant scenario extend the safety arguments developed in the base scenario by taking into
consideration safety-relevant uncertainties or alternative assumptions in both the evolution of
the geological repository with time and the models and databases used to simulate such
evolution. The IAEA [38] and ICRP [42] provide guidance on such scenarios, on the basisthat the long-term evolution considers natural perturbations with relatively high probability ofoccurrence. Based on this, 300 μSv/y (the risk constraint value of ≈ 1 x 10-5/y if theprobability of occurrence was one) is recommended as a dose constraint value for this
scenario category [38] [42] and is adopted in this report for discussion of safety.
(iii) Low probability perturbation scenario
Low probability perturbation scenario are defined to cover the inherent uncertainty thatremains in descriptions of very long timescale evolution, even if site selection and facility
design are appropriately carried out. Even if the probability of occurrence of major
perturbations is very low, the repository system aims to be sufficiently robust that, even insuch cases, there is no major radiological impact.
To assess significance of radiological impacts from such scenarios, the thinking for“disruptive natural events” in geological disposal described by ICRP [42] (based on present
existing exposure levels and emergency exposure levels as comparison standards) was
followed. Specifically with regard to dose, the range of the reference levels for emergency
exposure situations immediately after occurrence (20 to 100 mSv for one year) and for longterm exposure (after the second year: 1 to 20 mSv/y) were set as target values. For thesescenarios, risk calculation was also estimated by quantifying the occurrence frequency of theperturbation considered. A target value of 10-5/y was used, which is an internationally
recommended value for the sum of the risks of credible scenarios.
(iv) Human intrusion scenario
Inadvertent human intrusion is defined as a possible event that could lead to a significant
reduction or loss of safety functions of the geological disposal system. The probability of suchoccurrence is considered to be low for a repository deeper than 300 m in an area without
significant mineral resources, especially given efforts to preserve records and establish
warning markers. Therefore, as in the case of the low probability perturbation scenarios,
emphasis is on demonstrating repository robustness by showing that there is no major
radiological impact.
Regarding human intrusion scenario, the ICRP [42] concludes that there are no scientificgrounds for estimating future human activities and their probability of occurrence, and hence
it is not appropriate to conduct quantitative evaluations to compare such scenarios with doses
or risks. However, in the safety regulations of some countries, there are cases where it isnecessary to evaluate the potential impact of future human activities, such as in the UK [37].
4

Even if the doses for the base scenario exceed 10 μSv/y, safety can be assured unless the internationally
recommended dose-constraint of 300 μSv/y or risk constraint of 10 -5/y is exceeded.

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In addition, in the safety case report that SKB produced in 2011 to support the license
application of a geological repository for spent fuel [44], the impact of stylised human
intrusion scenario, including dose to intruders, is calculated but, in accordance with regulation,
not included in the risk summation. According to the US Environmental Protection Agency
regulation standard [45], the US WIPP operation licence [46] includes human intrusion
scenarios as a major focus for safety assessment and quantitatively specifies their probabilityof occurrence. Based on this background, it was decided to calculate doses and radiologicalrisks for illustrative scenarios, with radiological exposure targets the same as for low
probability perturbation scenario.
Table 6.1-6 summarises the scenario classification used in this report.
Table 6.1-6 Safety assessment scenario categories and dose targets
Scenarioclassification

Definition of scenarios and associated dose targets


Scenario considered to be a reasonable representationof repository evolution
In order to minimise radiation exposure, the dose target
is set based on the lowest value used in other countries

10 μSv/yScenario considering uncertainties in the base scenario
A target is set based on dose constraints for the generalpublic recommended by the IAEA [38] and ICRP [42].

300 μSv/yFor repositories constructed on the basis of proper site
selection and design, scenario of natural perturbations
that are considered to be extremely unlikely
For such scenario, to confirm that there is no
significant radiological impact, the approach andreference values are based on recommendations madeby ICRP [42] for unlikely accident scenario, or the risk
constraint value is applied.

Basescenario

Variant
scenario

Low
probability
perturbation
scenario






Human
intrusion
scenario

Dose target

The repository is located and designed so as tominimise the risk of human intrusion. While theprobability of such intrusion is very difficult to assess,
it is considered appropriate to apply the sameradiological exposure target as for low probability
perturbation scenario.

20 - 100 mSv(1st year)
1- 20 mSv/y(thereafter)

20 - 100 mSv(1st year)
1 - 20 mSv/y(thereafter)

In both the H12 and TRU-2 reports, the influence of perturbing phenomena such as
uplift/erosion, initial engineering defects, future human activities, etc. are handled as variant
scenario, and variations of model and data are included in the base scenario. However, thesescenarios were not defined according to a clear risk-informed approach that discussed theprobability of each scenario. Therefore, it should be noted that the definitions of scenario
categories are different for this report, even though scenario terminology is similar.

6.2 Methodology of safety assessment
6.2.1 Basic procedure of safety assessment
The methodology of safety assessment used in this report is in line with internationallyaccepted general practice, as described in Section 6.1.1. Specific supporting tasks include
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consideration of the geological setting at sites selected by appropriate procedures(summarised for the reference host rocks in the SDMs illustrated in Chapter 3) and repository
design tailored to provide the required safety functions (in this case, reference designs foreach SDM, as shown in Chapter 4).
In this section, following a basic framework described in Section 6.1, a series of tasks
proceeding from the construction of scenarios, setting of analysis cases for these scenarios,
RN release and migration analyses, to final dose evaluation and discussion of uncertainties
will be described. The basic procedure of the safety assessment in this report is shown in
Figure 6.2-1. This is basically the same for both HLW and TRU waste disposal.

Figure 6.2-1 Basic procedure for safety assessment in this report. The dotted line indicates thatthe procedure is carried out when required

In order to develop appropriate scenarios, the starting point is usually either a top-levelassessment of the temporal changes of the safety functions for a specific site and repository
design (termed the “Top-Down” approach [1] [2]), or a comprehensive list of relevant
Features, Events and Processes (FEPs) that is used to assemble scenarios (the “Bottom-Up”
approach [1] [2]). Here a variant was developed (the “hybrid” approach) for efficiently
constructing key scenarios related to repository safety functions, while ensuring
comprehensive coverage of all relevant FEPs [47].
In this method, the intended safety functions of a repository that are thought to evolve overthe long period of time after closure based on current scientific and technological knowledge
is described. In order to properly develop such descriptions, it is necessary to assess thethermal, hydrodynamic, mechanical, and chemical (including biological, if relevant) state
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(“THMC state”) of the entire system and how its evolution with time impacts defined safetyfunctions. Therefore, it is necessary to understand THMC state changes of the engineered andgeological barriers over all spatial scales for the models discussed in Section 6.1.3 (regional
scale/repository scale/panel scale/near field scale) over a time scale that extends frominitiation of the repository project and includes recovery of perturbed geological conditions
after construction and operation. “Storyboards” have been introduced as an effective methodof illustrating such evolution in a way that enables an overview of the scenario and to checkthe consistency and completeness of the representation of the repository on all relevant timeand spatial scales [48]. In order to assess changes in the THMC state for each component FEP(termed “process analysis”), a specific FEP list was created to reflect repository designs in thespecific geological environments considered in this report (as described in Chapter 4) basedon the latest scientific and technical knowledge (termed the “NUMO FEP list”: see Section
6.3.2 (1)). The NUMO FEP list was reviewed to assure relevance and completeness and used
to assure adequacy of description of repository evolution in associated storyboards andresulting scenarios.
The process of development of concrete scenarios can be simply summarised as:Clarify physical/chemical quantities (these are referred to as state variables) thatdefine safety functions for each repository componentExtract and correlate factors considered likely to affect the state variables related toeach safety function from each FEP in the NUMO FEP list (referred to as “factor
analysis”)



Based on the factor analysis, integrate relevant FEPs to examine the impact these arelikely to have on the state variables (termed “impact analysis”)

Develop base scenario from the assessed most likely evolution of each repository
design considered

Assess uncertainties and alternative assumptions to develop storyboards for variant
scenarios

Based on understanding of Japanese geology and international recommendations,develop illustrative storyboards for representative low probability perturbation andhuman intrusion scenarios.

For resulting scenarios, quantitative analysis cases are developed by clarifying specificconditions (such as the models and parameters to be applied) that must be considered in orderto quantify radiological impacts. This leads to selection of RN release and migration models
and datasets for each analysis case. Finally, these models and datasets are used to calculate
radiological impacts (dose or risk), which can be compared with reference values according to
the scenario classification, and discussed in order to assess if associated safety targets are met.
This process is covered in more detail in Section 6.3. Subsequently, in Section 6.4, foreach analysis case, the required models, calculation codes and datasets for RN release andmigration analysis are specified. The biosphere model that defines the RN-specific doseconversion factors for releases to the GBI is also presented. Radiation doses calculated for thebase, variation and low probability perturbation scenarios for both HLW and TRU waste ineach of the three geological settings, are then presented and discussed.
For human intrusion, special consideration is required to develop representative stylised
scenarios: these are presented in Section 6.5 and results of dose assessments then discussed.
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6.2.2 Treatment of uncertainty in the safety assessment
In order to carry out the safety assessment according to the methodology described above,it is necessary to comprehensively examine factors influencing safety functions to determine
if these could be potentially compromised. Such examination is based on current scientificand technological knowledge of the constituent elements of the repository system, their longterm evolution and potential external disturbances that may occur. However, there is a limit toavailable scientific knowledge and thus it is necessary to carefully consider handling ofresulting uncertainty. Uncertainty arises firstly in scenario definition, associated with theassumed processes causing evolution of the repository. In particular, it is essential to assurethat all factors that could potentially adversely affect safety functions are captured. FEPs witha positive impact, but judged as uncertain, are identified but not considered further in thecurrent assessment (“reserve” FEPs that may be utilised in the future to strengthen safetyarguments). After that, as discussed in Section 6.1.5, the probability that each scenario will
occur has to be determined in order to assign it to a specific class, with any uncertainties inthis discussed. These uncertainties will also be key input for the planning of further sitecharacterisation, in order to reduce those related to the SDM.
For the base and variant scenario, there are also uncertainties associated with model
assumptions and others arising from constraints of the codes and datasets used. Although theanalysis is aimed to be as realistic as possible, for the base scenario the most probable stateand states considering associated uncertainty cannot be clearly distinguished. As a result, inthe presence of such uncertainties, treatment tends to move towards conservatism (see Section
6.3.3 (2)). Nevertheless, based on future research and development, particularly when sites
are specified and safety regulations are defined, a more realistic scenario analysis will be
carried out, particularly for the base scenario, allowing them to be more clearly distinguishedfrom variant scenario. In the case of stylised representation of low probability perturbationand human intrusion scenario and associated analysis cases, uncertainties are more
fundamental and are emphasised during the discussion of assessment output.
In terms of quantitative estimates of radiological impacts of specific scenarios, total
uncertainty results from propagation of all individual contributions from the analysis models
and databases used, which can be difficult to quantify. To the extent possible, uncertainties
are reduced by rigorous testing (verification and validation) of models and databases, or the
impact of these minimised by introducing robustness into the coupled design and safetyassessment process. However, all uncertainties cannot be handled in this way and remaining
uncertainties need to be described to such an extent that their impact on safety can be assessed.
Such analyses will also be essential input to focus future R&D in order to reduce uncertainties.
More details on the handling of specific uncertainties are provided during the discussion ofsetting scenarios and associated analysis cases in Section 6.3.

6.3 Developing scenarios and establishing analysis cases
In this section, development of safety assessment scenarios and setting the associatedanalysis case for the specific repository designs and SDMs considered in this report isdescribed. Firstly, in Section 6.3.1, the latest knowledge is used to describe the behaviour of
the repository after repository closure. Based on this, Section 6.3.2 describes the stepwise
procedure for scenario development and the process of collecting and analysing information
at each step. Section 6.3.3 describes how analysis cases are set to quantitatively evaluate theradiological impacts of these scenarios.
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6.3.1 Description of system behaviour after repository closure
Based on the geological survey technology described in Chapter 3, it can be assured that
selected sites will not be significantly influenced by volcanoes, active faults, uplift/erosion,etc., for a long time in the future, while selected representative host rock at depths > 300 m
will provide a favourable environment for geological disposal. Based on the characteristics ofthe host rock, the safety functions of the geological environment are complemented with thoseof tailored engineered barriers to assure the safety of the repository system, as described in
Chapter 4. The post-closure safety functions assigned to the repository as a whole werepreviously presented in Tables 4.2-4 and 4.2-5. The expected behaviour of the repository
system after repository closure, together with resultant release and migration of RNs, can be
described and captured in storyboards, as illustrated in Figure 6.3-1 (see Supporting Report 611).
In the description of post-closure behaviour of the repository, it is necessary to start withthe state of the system at the time of closure, specified as a function of time and space, taking
into account the impacts of construction and operation and their possible influence on later
RN containment. A common process is used to handle the evolution of different repository
designs for the representative host rocks and different types of waste. The resulting
storyboards cover specific time steps:T1: period from repository closure until complete re-saturation.
T2: period after re-saturation until releases of RNs occur.
T3: period after releases of RNs occur, during which the characteristics of thegeological environment are considered not significantly changed.
T4: period during which the uncertainty regarding characteristics of the geologicalenvironment increases significantly with time.

For each of these time steps, the state of the repository safety functions (see Chapter 2) are
assessed considering four spatial scales: near field, panel, repository and regional, as
previously described in Section 6.1.3. The periods for which safety function for eachcomponent is expected to perform is shown in Figure 6.3-2.

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Period during which the uncertainty regarding
characteristics of the geological environment increases
significantly (T4)

Period during which the current geological environment is expected to be little altered (T1 to T3)
Period from repository closure until complete re-saturation (T1)

Panel scale

Repository scale

Regional scale

Construction & operation▼

Repository closure

Period after re-saturation until releases of RNs from the EBS (T2)

▼ Re-saturation complete

Period after RN release, when the characteristics of thegeological environment are considered not to havesignificantly changed (T3)

▼Initiation of radionuclide migration

 Adequate separation from volcanoes and large active faults
 Very low risk of new faults developing
 Slow, regional processes such as uplift, erosion and sea-level change that gradually alter temperature, hydraulic, stress and chemical fields

 Increased potential for changes in the geologicalenvironment to affect the safe function of therepository

 Continuous inflow of groundwater into therepository, local flow field towards therepository Water table draw down in the repository
area as groundwater flows into open tunnels
 Air is drawn into the repository through theaccess galleries The temperature in the vicinity of the repository increases dueto the heat generated by the waste, and then decreases in linewith radioactivity decay Groundwater flow towards and into the repository will continue
until completely saturated

 The temperature in the vicinity of the repository gradually decreases andapproaches the initial rock ambient value The hydraulic field in the vicinity of the repository recovers to its preconstruction state and reaches a steady state The initially reducing environment is restored

 The temperature, hydrological, stress and chemical
fields in the vicinity of the repository remain in a steadystate RNs migrate through the geological environment, withadvective flow in faults and fractures as the dominant
pathway to reach the biosphere Changes in the temperature, hydraulic, stress andchemical fields around the repository may occur

 Increased temperature in the HLW area
 Decrease in water pressure in the vicinity oftunnels EDZ formation around tunnels
 Gaps between backfill and tunnel walls
 Oxidising zone near tunnel walls due tooxygen uptake

 The temperature in the vicinity of emplacement tunnels
increases due to the heat generated by the waste, and thendecreases in line with radioactivity decay Groundwater flow towards and into tunnels will continue until
completely saturated
 Tunnel lining is sound and there is no lithostatic pressure on
backfill EDZ is present in the vicinity of tunnels
 Gaps remain between the backfill and the tunnel walls
 Oxidising areas persist near the tunnel walls
 High alkali content leached from any cementitious materials
present and pH of groundwater in the vicinity of the tunnel
increases The temperature in the vicinity of tunnels gradually decreases and approaches theinitial rock ambient value The hydraulic field in the vicinity of tunnels recovers to its pre-construction state
and reaches a steady state The gap between the backfill and the tunnel wall heals due to swelling anddeformation of the backfill
 The oxygen in the vicinity of the tunnel wall is gradually consumed by reducing
agents such as pyrite, and the environment becomes reducing
 Backfill material and hydraulic plugs are gradually altered by the reactions withgroundwater The temperature and hydraulic fields in the vicinity ofthe tunnel are in steady state The tunnel support concrete is altered and degraded,and hydraulic conductivity increases. A continuous,
highly permeable structure may be formed
 Potential flow short-circuits along tunnels are blockedby hydraulic plugs
 The tunnel deforms due to rock movements
 Reducing environment persists
 RNs migrate with groundwater flow by advection anddispersion

 Changes in the temperature, hydraulic, stress andchemical fields may occur

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Near field scale Formation of cracks in the HLW glass
matrix during manufacture Temperature rise in the vicinity of HLW due
to radiogenic heat
 Groundwater inflow from the tunnel wall
 Gaps remain between contained buffer andthe overpack within the PEM handling shell
 EDZ is generated around disposal tunnels
 Oxidising conditions develop arounddisposal tunnels

 The temperature rises due to the heat generated by the HLW
and eventually decreases in line with radioactivity decay
 Water penetrates the backfill
 After handling shell failure, swelling of the buffer to provide
mechanical protection and prevent groundwater flow
 Gaps exist between the buffer rings and between the buffer
rings and the overpack Gradual corrosion of the overpack and PEM handling shellconsumes any remaining oxygen and environment becomes
reducing Hydrogen gas may be produced due to corrosion of iron in the
reducing environment
 Slight alteration of the buffer material by reaction with iron

 The temperature of the near field gradually decreases and approaches the initialrock ambient value The water pressure outside the EBS recovers to its pre-construction value The buffer swells and any gaps disappear The buffer and backfill are gradually altered by the reaction with groundwater Corrosive expansion of the overpack deforms the buffer; corrosion expansion of
the PEM handling shell deforms the buffer and backfill
 Corrosion of the overpack and PEM container changes the porewatercomposition in the vicinity of the HLW and in the buffer
 Corrosion of the iron continues to produce of hydrogen gas.
 The overpack maintains containment for >> 1000 years

 The temperature and hydraulic fields are in steady state The groundwater flow field is affected by highly
permeable structures such as altered and deteriorated
concrete supports and drains left during backfilling HLW glass dissolution
 When the overpack is completely corroded, corrosive
expansion stops
 Over time, the number of overpacks failing increases
 Hydrogen gas is produced due to both corrosion andradiolysis
 Gradual deterioration of buffer, hydraulic plugs andbackfill due to alteration reactions
 RNs leached from HLW into groundwater migrate
through the buffer by diffusion

Figure 6.3-1 Example of a storyboard: Evolution of the HLW repository in plutonic rock (PEM option)

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 Changes in the temperature, hydraulic, stress andchemical fields may occur

Figure 6.3-2 Time scales where safety functions are expected

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(1) T1: period from repository closure until complete re-saturation
During construction, excavation of access tunnels/shafts and other tunnels, together withdisposal tunnels/holes/vaults takes place, along with installation of all infrastructure required
to ensure work safety and efficiency (ventilation, drainage, grouting/linings, etc.). Such
construction, and the subsequent operational period in which waste is emplaced, will result ingroundwater drainage and hydrogeological perturbations that can include lowering of thewater table and local de-saturation around tunnels. Desaturation and ventilation can result in
air penetration into rock walls and development of a locally oxidising environment. After
HLW and TRU waste emplacement, the temperature of the EBS and surrounding host rockrises for HLW and some higher thermal output TRU waste types (Gr.2, 4HD, 4HH), althoughtemperatures will not exceed limits set to assure that the performance of the EBS will not besignificantly impacted (See Section 4.5.2 (3) (ii)).
Infilled disposal tunnels/vaults will be quality assured, so that their hydraulic conductivity
is sufficiently low. Although disposal and access tunnels together with their surrounding
excavation damaged zones (EDZs) are connected to each other, repository layout andinstalled hydraulic plugs limit the potential for short-circuit paths for groundwater (seeSection 4.5.3 (3)). After closure, groundwater intrusion from surrounding rock will gradually
return the repository to a fully saturated state, while any oxygen in the air introduced during
construction and operation will be consumed due to reaction with EBS components or
minerals in the host rock.
Groundwater and rock thus return to their original temperature and reducing conditions, inline with the decrease in thermal output of the waste as a result of radioactive decay (seeSupporting Reports 4-39 to 4-41). The time required for recovery of geosphere conditions
depends on the geological and hydraulic characteristics of the host rock, methods ofconstruction and operation, and the properties of the waste. Although it will vary dependingon location within a repository extending over a few km2, recovery is estimated to takedecades to hundreds of years after closure [49] [50]. However, it is understood that resaturation time will depend on the local hydraulic conductivity and also to what extent gas isformed. In very low permeable media encountered in some other national programmes, resaturation may take several thousand years, at least for some waste package locations. These
aspects should be considered for specific sites to be explored in the future.
Where present, bentonite buffer in the EBS swells when groundwater intrudes into it andany gaps left at the time of construction seal (self-sealing), producing a very low hydraulicconductivity barrier that prevents significant advective water flow. For the specific case of
H12V after the backfill is installed, groundwater flow within the disposal hole is slow during
saturation, so piping erosion of buffer does not occur to a significant extent (see Section 4.5.4(5)). At this time, minerals in the buffer and surrounding host rock keep the chemistry of thewater contacting the waste package in a favourable condition for ensuring RN containment(see Supporting Report 6-15).
For TRU waste, a concrete structural framework is installed to maintain mechanicalstability during the operational period. For Grs.1, 2, and 4H, this is surrounded by a bentonite
buffer to strengthen the containment functions (see Section 4.4.2 (4)). Such concrete
structures, together with mortar infill, or the liners and grouting around HLW disposal tunnels,
will react rapidly with any inflowing groundwater to produce a hyperalkaline fluid with alocal pH that could be as high as 12 to 13. At such pH, alteration of the buffer andsurrounding host rock will occur, but the extent of such alteration is considered to be limited(see Supporting Report 6-8).
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The HLW carbon steel overpack, which has the safety function of preventing contact
between waste and groundwater for an extended period (which, at least for the SDMs
considered, would certainly extend beyond the buffer re-saturation phase), will corrode
gradually and uniformly while consuming residual oxygen (see Supporting Report 4-6). Afterporewater returns to the reducing conditions of previously undisturbed groundwater,
corrosion becomes extremely slow and results in hydrogen generation (see Supporting Report4-7). It is assumed that microbial activity in the vicinity of the overpacks and waste package
containers is extremely limited, due to the low activity of water in compacted buffer, and thus
their effect on corrosion is considered negligible (see Supporting Report 4-9). However, as
noted in Chapter 4, there is uncertainty related to the potential for microbial activity and workon this issue is ongoing.
For the steel TRU waste package container, the high pH of the porewater (caused byreaction with cementitious material present) ensures that the corrosion rate is extremely low
due to the passivation of iron; however pitting corrosion could possibly occur (see Supporting
Report 4-23).
In addition to corrosion of HLW overpacks and TRU waste packages, a number of othersteel containers are present in the EBS (HLW fabrication canisters, PEM handling shells for
the PEM case, TRU waste conditioning canisters and drums); although these will corrode, thisis not considered directly as they are not currently assigned a containment function after
disposal. Nevertheless, such corrosion will contribute towards maintaining the reducing
conditions of porewater.
Hydrogen generated by corrosion dissolves in groundwater or, if the production rate is
greater than loss by migration in solution, formation of a gas phase will occur. H2 will thencreate an overpressure, move through the EBS and disperse though repository structures
and/or the surrounding rock [14] [49]. Steel, which is a constituent of many EBS components
as noted above, may react with and cause deterioration of buffer and backfill, although theextent of this is considered to be negligible even for long periods (see Supporting Report 6-8).
In the weaker Neogene sediments, rock creep can cause progressive deformation of the tunnel
walls. This, in turn, will cause compaction of the buffer and alteration of its mechanicalproperties; however, it is considered that this will not compromise the designed stability of
the EBS (see Section 4.4.1 (3) (iv)).
In addition, the mechanical impact on the EBS due to shaking in the event of an earthquake
occurring during this period is considered minor (Supporting Report 4-17), with no effects on
hydraulic conditions expected (see Section 3.4.2).
During this period, the THMC conditions in the host rock around the engineered barrier,which were disturbed by the construction and operation of the repository, will return to theoriginal characteristics of the geological environment, so that the expected isolation andcontainment safety functions can be assured.
The behaviour of a repository during this period is influenced by the specification of therepository design and the processes of construction, operation and closure. For the purposesof this report, which does not cover a specific site, these issues have been considered only invery general terms. Once sites have been identified and the repository design andimplementation methods have been proposed, their effects on the geological environment andthe engineered barriers will be assessed by monitoring and other means, and the resultant
impacts will be considered accordingly. In addition, if anomalous events such as those
discussed in Chapter 5 occur before closure, the state of the repository will be set in

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consideration of their impacts, including those of response measures, on the geologicalenvironment after closure.
(2) T2: period after re-saturation until releases of RNs from the EBS
For HLW, as previously mentioned, corrosion progresses at a rate depending on porewaterchemistry, local degree of saturation (e.g. impact of hydrogen gas phase), temperature, extent
of radiolysis, etc. In the H12 report, containment by overpacks considers such factors andprovides evidence that failure of containment will not occur for at least a period of 1 ky,
thereby covering the time when radioactivity of the HLW is high and heat output issignificant (see Supporting Reports 4-39, 4-40). In addition, the thickness of the overpack is
set to ensure a large safety margin. This report also takes the 1 ky lifetime specification as areference value for the HLW overpack but, based on more recent corrosion knowledge, forassessed SDM conditions, it is expected that the overpack containment safety function isassured for at least 17 ky after disposal (see Section 4.4.1 (2) (ix)). Furthermore, for the PEM
option, it is expected that initial trapped oxygen in the repository will be consumed bycorrosion of the PEM handling shell before its failure, which results in an additional extensionof the overpack containment period.
For TRU waste, the containment safety function of the waste package container is notconsidered during design, due to uncertainties associated with pitting corrosion. Nevertheless,it can be expected that loss of containment will not occur for about several hundred years,
especially for the package B design thickness of 50 mm (see Section 4.4.2 (2) (v) (b)). For
waste package A, even if its integrity cannot be assured for a set time period, contact between
the waste and porewater will be delayed for a certain period of time and, thereafter, restrictedby the presence of the waste package components. Together with the mortar infill between
waste packages, the EBS will restrict entry of groundwater. Water that gradually penetrates
through the mortar infill and failed waste package containers will interact with primary waste
containers, conditioning material (mortar and asphalt) and then the solid waste. For thesewastes, the safety function “suppression of dissolution of radioactive substances” isconsidered and, even if RNs are dissolved in groundwater, their migration will be retarded bysorption onto EBS materials (significant for most RNs). Thus, the safety function of reducing
releases of RNs can be assigned to the entire waste package [14], with first releases occurring
only after the required time for re-saturation, which will ultimately ensure significant decay ofshorter-lived RNs.
Groundwater reacts further with cementitious EBS components during this period, withfurther chemical changes and alteration of buffer and infill materials, plugs and backfill
materials taking place. Nevertheless, the extent of alteration is limited and therefore will haveno significant impact on their safety functions (see Supporting Report 6-8). As for period T1,
the effects of earthquake shaking on the repository will not be significant during this period.
(3) T3: period after RN release, when the characteristics of the geologicalenvironment are considered not to have significantly changed
In this period, the conditions of the geological environment are effectively constant.
Although there is the possibility that both groundwater chemistry and hydrology may besubject to short-term transients, due to tectonic processes such as earthquakes, conditions are
assumed to recover quickly, thus there will be no significant effect on RN release andmigration processes. However, the current assessment has not looked at the impact of climate
6-30

change, with associated changes of sea-level and precipitation. This would be needed incoming assessments when potential sites are known, since the actual location of the sitewould determine which future impacts would be important to include here.
(a) Release and migration of RNs in the EBS and host rock
For HLW, after overpack failure, water penetrating the buffer will come into contact withthe glass, which then starts to slowly dissolve. As the dissolution of glass proceeds, thedissolved silica concentration in porewater increases and the dissolution rate will gradually
decrease to a long-term value (residual dissolution rate) [51] [52] [53] [54]. RNs (and otherstable components) in the glass dissolve congruently with the borosilicate matrix. As the flowof water in the buffer is extremely slow, the movement of solutes from the glass surface andthrough any alteration layer present, occurs predominantly by diffusion. Additionally, thesolution concentration of most RNs is constrained by their solubility in the porewater present.
Solubilities are specified on an elemental basis: these are thus shared between the isotopes(both stable and radioactive) of the element present, depending on their specific abundance
ratio [3]. RNs present in gaseous form may be entirely dissolved in groundwater due to theirtrace concentrations, but some may also migrate as a gas phase by mixing with hydrogen gas
if this is present (see Supporting Report 6-8).
RN diffusion from the glass will be retarded by interactions with solid phases, such as
sorption onto overpack corrosion products and minerals in the buffer. Migration of colloids
containing RNs is prevented by the filtration function of the buffer, which effectively
immobilises them [55]. During this period, radioactive decay also contributes to limit therelease of RNs from the EBS. The RNs released disperse into groundwater within the EDZ
and then migrate further into the surrounding undisturbed host rock (RN migration via
incorporation into colloids is not considered in this study, as justified in Section 6.3.2 (3)).
Common features of the three representative host rocks are networks of fractures or otherdistinct water-carrying features (see Section 3.3.3 (3) (ii)), within which advective
groundwater flow will occur. In general, the interior of these features is filled with various
minerals generated as a result of the long-term interaction between host rock and groundwaterflowing through channels or pores within them (see Section 3.3.3 (5)). Nevertheless, for thehost rock as a whole, hydraulic conductivity will be low and the flow of the groundwater slow.
RNs penetrating the host rock will move with the flow of groundwater, subject to mechanicaladvection/dispersion mechanisms and chemical interactions (e.g. sorption) with either flow
path infilling minerals or the rock accessed by matrix diffusion from these paths. Thus, themovement of dissolved RNs will be slower than that of groundwater, with concentrations
reduced by radioactive decay and 3D dispersion. For Neogene sediments, the hydraulicconductivity of this host rock type is relatively large compared to the other two rock types andthus significant advective groundwater flow may occur even within the bulk rock (see Section
3.3.3 (5) (b)).
For TRU waste, as described above for period T2, safety functions such as reduction of RN
release are assigned to the waste matrix and waste package barriers. Such safety functions are
not designed to be maintained for as long a time as the HLW EBS and thus, with time, thebarrier components will degrade and RNs will be released from the EBS. At this time, as for
HLW, the concentration of RNs will be constrained by elemental solubility limits withineither the waste package or the infill porewater. Colloids containing RNs may be formed in
the waste package, but they are not stable in porewater with high Ca concentrations that result
from reaction of groundwater in the host rock with cementitious materials in the structural
6-31

framework and between and within the waste packages. In porewater with high Ca
concentrations, colloids are unstable and precipitate by flocculation [14] [56]. In Gr.3 and 4L
waste packages and other vaults without buffer, colloids in the groundwater are not filtered by
the buffer, but also flocculate and precipitate in the waste package porewater. Thus, theprobability of colloidal RN migration from the waste packages is considered to be small.
In addition, although mortar waste package infill is likely to contain cracks, it has sorptionfunctions that limit the rate of RN elution from the EBS of waste Grs.1, 3 and 4L (see Section
4.2.4 (2) (ii)). Similar functions can be expected of mortar infilled between stacked packages
and structural concrete for all TRU groups. TRU waste Grs.2 and 4H have significant heat
output, which could cause deterioration of EBS materials and resultant RN sorption functions.
EBS materials also degrade due to reaction with groundwater and, also here, confidence inmass transfer resistance functions is gradually lost.
Organic substances contained in TRU waste decompose under the influence of radiation,
heat, etc., and degradation products elute into groundwater along with RNs. Isosaccharinicacid is a typical degradation product (from cellulose under high pH conditions), which
complexes strongly with some RNs, increasing their solubility and decreasing their sorption(see Supporting Report 6-17). In addition, nitrate from TRU Gr.3 eluted into groundwater cancomplex with some RNs, whilst increases in ionic strength may impact both the solubility andsorption of some RNs (see Supporting Reports 6-8, 6-20).
RNs released from TRU waste groups without surrounding bentonite, after interaction withEBS components, disperse into groundwater within the EDZ and then migrate further into thesurrounding undisturbed host rock. During this period, radioactive decay also contributes tolimit the release of RNs from the EBS. For TRU waste including a buffer, the flow of thewater inside the EBS is much lower, hence both RN interactions with barrier components andradioactive decay during the longer transport to the host rock are greater, leading tosignificantly more attenuation of releases. In addition, for all waste groups, decomposition
products could be generated from organic materials present in waste or added to ensure theworkability of concrete. If so, these will impact the function of supressing RN migration in
the host rock and overall geological barrier performance.
RNs released from the EBS migrate through the host rock as described above for HLW.
Downstream of TRU waste vaults, however, groundwater chemistry is significantly altered byleachates from cement-based materials, nitrates from Gr.3, soluble organics, etc., resulting inalteration of minerals of the host rock/flow path and hence the safety function of retardingtransport of RNs. Even in this case, TRU waste disposal will not impact the multiple barriersystem of HLW, as the TRU waste emplacement area is always located at a distance on thedownstream side of the total repository footprint (see Section 4.5.4 (3) (iv)). In addition,
degradation products derived from organic materials left in the repository and fromadmixtures containing organic matter used to improve the workability of cementitious
materials may also alter the performance of the host rock in controlling RN migration. Someof the RNs dissolved in groundwater may migrate in gaseous form through engineered
barriers and host rocks by mixing with any hydrogen gas present (see Supporting Report 6-8).
During this period, the buffer, plug and backfill material will gradually degrade due toreactions with groundwater and hence confidence in the performance of the transport
suppression safety function will also decrease. The extent of such alteration depends on thechemistry of groundwater, advective fluxes and the kinetics of the processes involved forspecific materials. In the long term, alteration zones may become greater and the safetyfunction may be degraded or lost in some cases.
6-32

(b) RN migration to the biosphereFor both HLW and TRU waste, as mentioned in Section 6.1, some long lived and mobile
RNs will migrate through the host rock until they reach a pathway to the GBI. In this report,
the migration model is greatly simplified and this pathway is expected to be provided by alayout determining feature (LDF - which is avoided in the layout design (see Section 4.5.1(2))) or the repository scale model boundary. The migration distance from the near field tosuch a boundary is represented in the RN migration model, but the characteristics of the“short-circuits” to the GBI are not defined, with the assumption that their hydraulicconductivity is high and retardation within them would not contribute a significant barrier role.
It is recognised that this may be highly conservative, as evidence of old groundwaters at depth
in relevant environments suggests limited upwards flow in LDFs while their extensivealteration zones suggest a capacity for significant RN retardation. This may be assessed more
realistically for specific sites.
The discharge from such a flow path to the biosphere occurs at the GBI, which is also very
simplistically represented by mixing into a surface/near surface water body. This results indilution, which is greatly dependent in the characteristics of this interface – e.g. local orregional aquifers, rivers, coastal waters, etc. [3]. Again, this is highly site-specific and will
depend on hydrogeological setting, geography, topography/bathymetry, etc. and hence isrepresented in only a stylised manner at the present time.
(c) RN migration and exposure pathways in the biosphere
RNs that reach the biosphere will be distributed throughout the surface water circulation
system, in which water movement rates are generally much faster and fluxes much higher
compared to deep groundwater [3]. As a result, RN concentrations in water within thebiosphere are homogenised and greatly diluted, but may be re-concentrated or captured bysorption onto soils or sediments. Human activities, such as the use of water for drinking andirrigation, plus the natural transfer of RNs through terrestrial, freshwater and marine foodchains result in potential radiation exposure pathways to the population living in the release
area at this future time.
(4) T4: period during which the uncertainty regarding characteristics of thegeological environment increases significantly
The description of evolution of the repository above assumes careful site selection (seeSections 3.1 to 3.2), so that favourable geological conditions are maintained over a long
period of time. Nevertheless, slow regional evolution due to continuous processes, such as
uplift/erosion, and periodic climate change on a global scale are inevitable. Such processes
can result in changes in topography, groundwater flow and chemistry, surface hydrology, etc.
The specific effects will differ from place to place: for example, the effects of glaciation arelimited to mountainous areas whilst, at coastal areas, the effects of uplift/erosion and sea levelchange may interact. As the repository is located within a deep rock formation, THMC
impacts on the EBS and surrounding rock will not be significant for an long time (see Section4.3), thus these are assumed to maintain their safety functions. However, when specific sites
are assessed, long-term THMC evolution would need to be analysed to check the validity ofthis assumption.

6-33

On a timescale > ≈ 100 ky, however, uncertainty concerning the long-term stability of the
deep geological environment will gradually increase. Additionally, although the probability ofoccurrence is still extremely small, major perturbations resulting from volcanism/magmatic
activity and fault activity near the repository cannot be completely precluded (see Section 3.4).
Furthermore, although not analysed quantitatively in this safety case, uplift and erosion may
change the repository environment and could even, after a very long time, lead to exposure of
repository components at the surface.
6.3.2 Developing scenarios
Based on the safety assessment procedure shown in Figure 6.2-1, together with theunderstanding of the behaviour as the repository after closure described in Section 6.3.1, thestepwise development of the scenarios used for safety assessment is described below.
For development of these scenarios, the specific advantages of conventional top-down andbottom-up approaches were used to formulate the adopted hybrid approach, which associates
FEPs and safety functions [1], as described in Section 6.2.1. By using this hybrid approach,
important scenarios directly linked to safety functions can be efficiently extracted, aiming forcompleteness and consistency with current scientific knowledge, based on logical screeningof FEPs from the viewpoint of their relevance to these safety functions.
(1) Creation of the NUMO FEP list
In the bottom-up approach, FEPs related to the repository are extracted based on structured
assessment and material from international initiatives intended to ensure relevance andcompleteness. In particular, the OECD/NEA has developed a general international FEP list[57] that is intended to be used without consideration of specific geological environments or
waste disposal concepts (here termed i-FEPs). Following the i-FEP hierarchy andclassification, a NUMO FEP list was created by adding and deleting FEPs on the basis of theboundary conditions and waste characteristics to be considered for geological disposal in
Japan. Thus, the features of the geological environments shown in Chapter 3 and disposal
concepts in Chapter 4 are considered, together with past studies aimed at creating Japanese
FEP lists in H12 and TRU-2 [58] [59].
The NUMO FEP list includes 284 FEPs, with a structure as outlined in Figure 6.3-3 andmore details given in Supporting Report 6-4. The NUMO FEP list follows the hierarchicalstructure of the i-FEP list. The first level represents five different areas: external factors,
factors related to packaged waste, factors related to the repository, factors related to the hostrock and factors related to the biosphere. Where the same process occurs in different areas of
the first hierarchy, the FEPs are organised appropriately and a distinction is made by noting inbrackets which area the FEP covers.
In the NUMO FEP list, FEPs are systematically selected with emphasis on completenessand hence the same process or event may be handled in multiple FEPs. For example, in the
NUMO FEP list, the process “expansion of overpack caused by corrosion” is included in FEP“F2.3.4.4 corrosion”, which is the originating event, as well as in environmental features
impacting corrosion: “F2.3.4.1 pH conditions”, “F2.3.4.2 Redox conditions”, “F2.3.3.2
Volume changes of materials”. In addition, the NUMO list classifies FEPs in terms of process
mechanisms: these groups are thermal, hydraulic, mechanical and chemical, covering all

6-34

relevant FEPs occurring within the repository system or the surrounding geosphere5.
Furthermore, to assure the completeness of the list, FEPs describing the coupling of theseprocesses are explicitly included as a special group (e.g., Table 1 in Supporting Report 6-4:
“F2.3.1.4 The effect of thermal processes on other processes (waste package)”).

Figure 6.3-3 NUMO FEP list structure

The NUMO FEP list used in this report will be revised as necessary in the future, forexample, in response to general advances in science and technology, expansion of relevant
knowledge through research and development, formulation of safety regulations, andinformation obtained through site surveys.
(2) Development of scenarios
Using storyboards (see Figure 6.3-1) that capture system evolution after repository closure
with a focus on safety functions, as described in Section 6.3.1, and taking into accountuncertainties associated with such evolution, representative scenarios can be described. Using
the NUMO FEP list, factors that may affect the safety function are identified in acomprehensive manner, so that the extent of such effects can be analysed.
(i) Factor analysis of safety function(a) Extraction of state variables
As in the top-down approach, first the THMC properties that define the safety functions ofeach component of the target repository system are extracted and termed “state variables”.
The extraction of state variables involves:

Expanding on mechanisms defining each safety function on the basis of latest
scientific knowledge.

5

i-FEPs are classified into seven types, including THMC plus biological, radiological and gas. In the NUMO
FEP list, biological and radiological are included in chemical FEPs, whilst gas is included in hydraulic ormechanical FEPs.

6-35




In particular when quantitative models are applied to the safety function, defining theparameters included in the model and the physical and chemical variables that affect it.
Structuring to clearly identify state variables directly affecting the safety function.

As an example, Figure 6.3-4 shows some of the extracted state variables for the safetyfunctions of the H12V buffer, which include “inhibition of migration of RNs by advection”,“suppression of colloidal migration” and “sorption of RNs” presented in a “state variable
definition diagram”.

Figure 6.3-4 Example of some state variables that influence safety functions (buffer safetyfunction)

Here, the safety function of inhibiting migration of RNs by advection requires that that the
bentonite swells as it saturates, sealing fabrication gaps between buffer blocks or between thebuffer and other components. However, over a long period of time, buffer may be eroded byflowing groundwater and thus its density will gradually decrease, thereby increasing thehydraulic conductivity and the possibility of advective transport of RNs. Alteration of buffer(e.g. due to interaction with overpacks or cementitious materials) may, depending on the
groundwater composition, also result in hydraulic conductivity increases due to change ofmineral composition, for example accompanying ion exchange in montmorillonite(conversion to Ca-type, Fe-type, etc.). In addition, there is a possibility that the hydraulicconductivity may be increased by thermal deterioration of the buffer; due directly to theincrease in temperature or to associated changes in water chemistry, such as changes in ionicstrength that alter the microstructure of compacted bentonite. In consideration of these aspects,“density”, “mineral composition”, “gap structure”, “thickness”, “temperature” and “porewater
quality” are extracted state variables affecting the safety function of preventing advective
transport through the buffer.
Supporting Report 6-6 summarises the state variables for all safety functions, organised
according to the concept outlined above.
(b) Extraction of FEPs affecting state variables (use of factor analysis diagrams)
Next, the future evolution of the safety functions is analysed. Based on an understanding of
the behaviour of the repository system, as described in Section 6.3.1, FEPs extracted from the
6-36

NUMO FEP list are organised by a bottom-up approach, related to the state variables defining
the safety functions (this is called factor analysis: see Figure 6.3-5).

Figure 6.3-5 Example of a factor analysis diagram relating to the safety functions of the buffer(explanation of Integrated FEPs in text following)This analysis is subject to the following constraints:










As described in Section 2.4.4, risks due to chemical hazards are not handled in thesafety evaluation of this report. Although assessments in some countries are subject tosafety regulations covering such risks (for example, Sweden [60], UK [37], France[31]), these are not included in Japanese safety regulations and, for some TRU waste
types, the specifications are not defined, so it would be difficult to evaluate them.
FEPs on natural processes affecting the safety function of the repository after closure
are decoupled from anthropogenic FEPs relevant to human intrusion scenarios.
Some FEPs in the NUMO FEP list relate to actions associated with the selection ofsites, design, construction, operation and decommissioning, carried out before
repository closure and which define the initial status of particular safety functions.
Uncertainties associated with these should be considered if they affect safetyfunctions (for example, boundary condition FEPs or quality control FEPs, see Figure
6.3-3). With regard to such FEPs, when developing scenarios (see Section 6.3.2 (3)below), the uncertainties involved need to be considered after the factor and impact
analyses described in Section 6.3.2 (2) (ii).
FEPs which relate to migration behaviour of RNs (for example, “F2.4.1.2 dissolution”
or “F2.4.1.3 diffusion”) are not taken into consideration in the factor analysis ofsafety functions or the impact analysis described in Section 6.3.2 (2) (ii) below. In
describing the scenarios, it is necessary to include the behaviour of the RNs in acomplete and consistent manner.
Coupling between processes is not explicitly displayed in the factor analysis diagram,but, in the process of influence analysis (Section 6.3.2 (2) (ii)), such interactions areanalysed along with associated uncertainties.
In the scenarios of RN migration and resultant radiation exposure in the biosphere, atthe present stage without a specific site, a generalised biosphere model based on thatin the H12 report is used.

Based on the latest scientific knowledge on each FEP, this approach allows influences onstate variables to be identified.
As previously noted, FEPs in the NUMO FEP list are systematically selected withemphasis on completeness, so specific processes may be handled in multiple FEPs. Therefore,
6-37

the factor analysis diagram becomes complicated and it is difficult to efficiently create
scenarios while keeping traceability. Therefore, after associating FEPs with each statevariable, related FEPs are merged and integrated into I-FEPs in an integrated FEP list.
Integration of 105 THMC FEPs resulted in 29 I-FEPs.
This simplifies the factor analysis diagram and facilitates traceability, thus enabling anefficient impact analysis to be carried out. In the Swiss safety case [51], the “OPA FEP
Database” consists of 482 FEPs extracted exhaustively from the international FEP list, and“Super FEPs” (effectively integrated FEPs) consists of 38 FEPs extracted top-down based on
the understanding of the disposal system. The inclusion of individual FEPs in the OPA FEP
Database guarantees the comprehensiveness of the Super FEPs. This is a similar approach tothat in this report, where the integrated FEP list and the NUMO FEP list are used together inorder to achieve both efficiency in scenario creation and comprehensiveness of scenarios.
For the I-FEPs, it was decided to label them based on the initiating event that influences
other processes. For example, expansion of overpack due to corrosion covered by “F2.3.3.2
Volume change of material (waste package)” is included within integrated FEP “IF16corrosion”. The classification of all FEPs and the integrated FEP6 development process are
given in Supporting Report 6-5.
Figure 6.3-5 shows an example of integration of FEPs that influence “mineralcomposition”, “porewater quality”, and “density”, extracted as state variables for “sorption ofradionuclides”, which is a safety function of the HLW buffer (see Section 6.3.2 (1)).
Supporting Report 6-7 contains factor analysis diagrams for the other safety functions.
(ii) Analysis of impacts on safety functions and handling in scenarios
The starting point is provided by the I-FEPs, the state variables describing the mostprobable state of safety functions of the repository after closure (described in Section 6.3.1)
and uncertainties in current knowledge. Scenario development then results in base scenario,
variant scenario and low probability perturbation scenario. Impact analysis determines themost probable state of the safety functions and the uncertainties in the FEPs to be consideredin the resultant scenario and hence its classification in terms of probability. In preparing theimpact analysis table, in addition to scientific information from existing literature on
phenomena related to FEPs affecting the safety function as extracted by factor analysis,
additional clarification (termed “process analyses”) was performed (Table 6.3-1).

6

For integrated FEPs, “IF” is attached to the FEP number.

6-38

Table 6.3-1 List of process analyses performed in this assessment
Analysis

Overview

SR

Buffer–overpack chemicalalteration

Evaluate the effect of chemical alteration of buffer due toreaction with iron released by overpack corrosion on its safetyfunctions.

6-8

Buffer–concrete chemicalalteration

Evaluate the effect of chemical alteration of buffer due toreaction with leachate from concrete on its safety functions.

6-8

Concrete–host rock chemicalalteration

Evaluate the effect of chemical alteration of host rock due toreaction with leachate from concrete on its safety functions.

6-8

Structural frameworkchemical alteration

Evaluate the influence of leaching of cementitious materials inthe structural framework on the hydraulic conductivity of thestructural framework.

6-8

NitrateimpactEvaluate spread of, and chemical changes due to, nitratecontained in TRU waste Gr.3.

6-8

Gas migration in buffer

Evaluate effect of porewater displacement by hydrogen on thecontainment function and the effect of gaseous RNs on dose.

6-8

Overpack settling in buffer

Evaluate the long term (after re-saturation) mechanical stabilityof buffer in terms of overpack sinking due to its higher density.

4-17
4-28

Corrosion expansion

Evaluate the long term (after re-saturation) mechanical stabilityof buffer against additional loads accompanying overpack
corrosion expansion (coupled to rock creep)

4-17
4-28

Earthquake response

Evaluate the mechanical stability of the EBS against large scale
earthquake motions

4-17
4-28

Thermal impact

Evaluate the temperature rise in the buffer due to the heatgenerated by the waste.

4-39
4-40
4-41

Factors that are judged to have negligible influence on safety functions during impact
analysis are not taken into consideration during scenario definition. Some factors that act toenhance safety functions are also clarified, but are not currently covered by the models/codesand databases used for quantitative evaluation of scenario evaluation; these are conservatively
ignored. Examples here include the transport resistance of the surface-altered layer formed
during dissolution of HLW glass, sorption of RNs onto iron corrosion products derived fromoverpacks and irreversible uptake of RNs in the buffer and host rock (mineralisation).
For HLW and TRU waste, the impact analysis tables are prepared for the components of
the repository expected to have safety functions (see Table 4.2-4 and Table 4.2-5). Here,components expected to have common safety functions regardless of host rock, like buffers,
are included in one table to avoid duplication. However, if there were differences identified
for specific geological setting, these would be described explicitly (this is not the case for thepresent SDMs). As an example, the impact analysis table summarising the output of the factor
analysis diagram on the safety function of the buffer (Figure 6.3-5) is presented as Table 6.3-2.
The integrated FEP “IF15 water chemistry” includes “porewater chemistry”, the probablestate and uncertainty of which (for both HLW and TRU waste) results from reactions between
groundwater and the buffer and, in particular the degree of progress of Ca ion exchange,
which, in turn, impacts sorption and diffusion of RNs.
The impact analysis tables for FEPs associated with all safety functions related to HLW
and TRU waste repositories, together with associated state variables, are included in
6-39

Supporting Report 6-9. In the example shown in Table 6.3-2, there are no FEPs that would beclassified as variant or low probability perturbation scenario but, if such FEPs existed, theywould be captured in the “Scenario classification” column. For a full description on all
impacts on safety functions considered, see Supporting Report 6-9.

6-40

Table 6.3-2 Examples of impact analysis: influences of state variables describing the safety function of buffer “sorption of RNs”
Safetyfunction/Statevariable

Integrated
FEP




15. Water chemistry

Porewater chemistry

Uncertainty to be considered in scenario
analysis

Probable status







In the case of HLW, groundwater reacts with buffer to form theresultant porewater. Depending on the composition of thegroundwater, alteration to Ca-type montmorillonite may occur (seeSupporting Report 6-15).
Hydroxyl ions, produced by corrosion of the overpack, may increaseporewater pH at the contact with the buffer, but any deterioration islimited to the vicinity of this interface (see Supporting Report 6-8).
For HLW, the influence of cementitious material on the buffer can beignored: for hard rock the amount of concrete is limited, for weakersediments a concrete lining would be needed, but this will not affectthe buffer for the H12V case. For the PEM case, the concrete lining
needs to be considered but the impact is limited since buffer alteration
is suppressed (see Supporting Report 6-8).



For TRU waste also, groundwater reacts with buffer to form theresultant porewater (Supporting Report 6-15). At this time, high pHinteractions between montmorillonite and concrete occur, but thereaction is limited to the interface of the buffer material/concrete. The
assessment carried out suggests that unaltered montmorillonite
predominates in the buffer, and the rise in pH is suppressed byreactions with buffer material (see Supporting Report 6-8).
Na – Ca exchange in remaining montmorillonite affects its sorption
characteristics, depending on the degree of alteration (See Supporting
Reports 6-8, 6-15).
For TRU waste, since no iron containers contact directly with buffer,the effect of iron corrosion is negligible.







6-41

Handling classification in scenario analysis

In alteration analysis of buffer due tocorrosion of the overpack, uncertainty
exists in the dissolution rate ofmontmorillonite, associatedthermodynamic data, the secondary
minerals precipitated and their effects on
RN migration.
As a result of the above, there is
uncertainty in both the porewaterchemistry and alteration ofmontmorillonite, however it is consideredthat alteration is limited in extent(Supporting Report 6-8).



Uncertainty exists in the analysis of buffer
alteration due to concrete associated withthe dissolution rate of montmorillonite,
associated thermodynamic data, thesecondary minerals precipitated and theireffects on RN migration.
As a result of the above, there is
uncertainty in both the porewaterchemistry and alteration ofmontmorillonite, however it is consideredthat alteration is limited in extent(Supporting Report 6-8).









Consider the degree of alteration and impacts
on sorption and diffusion of RNs in HLW
buffer.
In the base/variant scenario, changes in pH
and ionic strength of buffer porewater due tocorrosion of the overpack are not considered.
For HLW, it is considered that the effect on
buffer due to cementitious material isnegligible, this effect is not considered in thebase/variant scenario.

In consideration of sorption and diffusion of
RNs in buffer, consideration is given to thedegree of Ca conversion.
For TRU waste, the influence of cement
interaction is limited to the interface withbuffer, so it is not considered in base/variant
scenario. Also, the influence of pH increase
due to iron corrosion is not considered in any
scenario, since the analyses of iron corrosion
presented in Supporting Report 6-8 suggest
the impacts on pH are too small to affectbentonite stability.

Safetyfunction/Statevariable

Integrated
FEP

Uncertainty to be considered in scenario
analysis

Probable status

Handling classification in scenario analysis

Effects on HLW buffer by high pH plumes derived from TRU waste
are defined by the groundwater flow direction and the separation
distance, allowing these to be ignored for current layouts (see Section
4.5.4).



Since it can be avoided by design, such
scenarios are not considered. It is
understood that this may be more
complicated to show for sites where theflow direction may change in the future,e.g. coastal sites in case of sea-levelchange.



Influence on the HLW repository of the highpH plume from TRU waste is not considered.



The nitrate plume due to TRU waste Gr.3 may affect the buffer
porewater chemistry of other waste, depending on layout relative tothe groundwater flow direction (Section 4.5.4) and the separation
distance between HLW and TRU waste disposal panels.



Since it can be avoided by design, such
scenarios are not considered. It is
understood that this may be more
complicated to show for sites where theflow direction may change in the future,e.g. coastal sites in case of sea-levelchange.



Influence of the nitrate plume on other waste
is not considered.



In HLW and TRU waste repositories, it is assumed that nocomplexants are present at a level to have a significant effect on thebuffer sorption characteristics, so this can be ignored. The potential ofcomplexants from organic degradation is assessed, but judged notimportant.

Since no significant source of complexants
assumed to be present, this is notconsidered.

In the base/variant scenario, the influence ofcomplexants on the buffer porewaterchemistry is not considered.

6-42

(3) Selection of scenarios
Here, based on storyboards (e.g. Figure 6.3-1) showing how the safety functions of thespecific repository systems are expected to perform after closure of the repository, scenarios
are developed in a structured manner. It is noted that there is a subjective aspect to thisapproach at the present time, with expert opinion playing a key role, but the aim is that the
total set of scenarios would be as comprehensive as possible.
By definition, the base scenario represent the expected behaviour of specific repository
systems, as illustrated in a storyboard. Based on this, the other scenarios are developed froman analysis of probability/associated uncertainties of each FEP in the impact analysis tables
described in Section 6.3.2 (2). After organising the safety functions of the repository
components and corresponding RN migration process for each time frame, this results in
descriptions of the base, variant and low probability perturbation scenario. Transparency ofthe safety evaluation is ensured by considering:The uncertainty in the influence of FEPs on safety functions, which can have either apositive or negative impact. For discussing safety of repository options, uncertainty
that has a positive influence on safety functions is conservatively ignored in thevariant scenarios, which focus on uncertainties that could negatively impact safety.
From impact analysis tables for safety functions influenced by integrated FEPs thatwere noted to be treated as low probability perturbations, special scenarios to consider
these were developed.
In cases where multiple uncertainties result from many related factors, uncertainty
propagation is considered: where factors have very low probability it is consideredreasonable to postpone detailed assessment of uncertainties until the repository
system is better defined for specific sites.

In the following, base, variant and low probability perturbation scenario are describedbased on storyboards covering the temporal-spatial scales described in Section 6.3.1.
In describing these scenarios, regardless of the host rock, focus is on the EBS for HLW
options (H12V; PEM) and options for TRU waste groups, since these have many common
components and expected safety functions, which can be presented in a unified manner. In thecases where there is a difference due to the host rock, it is noted explicitly. Here, componentsof the repository include not only those that provide the designed safety functions, but also allothers (Table 6.1-2) that could influence RN release and migration (see Section 6.3.2 (2) (ii)).
In terms of gaseous RN migration, it is considered that the inventory of sources in thewaste is small (Section 6.3.1 (a)) and hence gases will dissolve in groundwater when therepository re-saturates. However, conservative dose assessment assuming that all gaseous
RNs do not dissolve in groundwater and are transported to the biosphere has confirmed thatthe effect would be negligible (see Supporting Report 6-8). For these reasons, gaseous RN
migration scenarios are not currently considered in the safety assessment.
With regard to RN migration via incorporation into colloids, organic matter andmicroorganisms in groundwater, there are some examples of studies [3], [14], [61], [62] thatdeal with such migration. However, the dose assessment results are highly dependent on thetype and concentration of such colloids, organic matter and microorganisms present, and thus
will be treated based on ambient geochemistry after a site has been identified.

6-43

(i) Description of base scenario
The main features of the base scenario are described below, focusing on clarification of thehandling of safety functions in the RN migration analysis. This report covers the disposal of
HLW and TRU waste, but elements of scenarios concerning the geology beyond the near field,
the biosphere and natural perturbations are common to both.
(a) T1: period from repository closure until complete re-saturation(A) Near field scale- EBS
For HLW and TRU waste Grs.2 and 4H, heat generation results from decay of short-lived
RNs contained therein. Locally, temperature rises in the wastes and associated buffer, butwaste loading is such that buffer maximum temperature will not exceed 100 °C (see Section
4.5.2 (3) (ii)). Key buffer safety functions, Reduction of advection of RNs, Suppression ofcolloidal migration, and Sorption of RNs are thus not impaired as a result of any thermal
deterioration. In addition, the temperature of the glass matrix is not high enough to result inthermal deterioration, and the “Low dissolution of the glass matrix” safety function is
maintained. Cracks in the glass matrix that may occur during manufacture remain and are
explicitly accounted for the analysis.
In the case of TRU waste Grs.2 and 4H, it is assumed that an increase in the temperature of
the near field may cause alteration of concrete mineralogy, and hence it is conservatively
assumed that the suppression of dissolution of RNs, function of the waste or waste package(Gr.4H), and sorption of RNs, function of waste package infill (mortar), may change.
When the repository is closed, groundwater invades slowly from the surrounding hostrocks into the buffer, which was in an unsaturated state during the operational period.
Groundwater saturation in the entire repository within about several hundred years after
closure is assumed, although this could be longer in some cases [49] [50]. As the saturation of
the buffer progresses, the thermal output of the waste decreases along with the temperature of
the near field, until it approaches original rock ambient.
As the buffer for HLW and TRU Grs.1, 2 and 4H gradually saturates, it swells andgenerates a swelling pressure that contributes to increasing mechanical stability of the nearfield. This supports buffer safety functions related to restriction of the transport of RNs. In thecase of the PEM option, the handling shell surrounds the buffer, but, after it loses containment,
the buffer saturates resulting in similar safety functions as for the H12V case (Supporting
Report 4-19).
The minerals of the host rock and buffer tend to maintain constant water chemistry(chemical buffering): oxygen introduced in air during construction and operation will bequickly consumed after closure and the porewater becomes reducing like the original,
undisturbed groundwater. Residual oxygen is even more rapidly consumed by any steel
repository components present (see Table 6.1-2: steel supports for disposal in Neogene
sediments, rock bolts commonly used in all repositories, TRU waste packages, HLW
overpacks and PEM handling shells), ensuring that porewater is reducing. As a result, this
supports the safety function, suppression of dissolution of RNs.
The buffer may be partially altered (e.g., Ca–Na ion exchange) due to reaction withgroundwater, but its safety functions will not be impaired by this (see Sections 4.4.1 (3), 4.4.2(4), Supporting Reports 6-8, 6-15). In addition, due to leaching of cement-based components
6-44

of the repository (see Table 6.1-2: grout material common to all repositories, shotcrete, linerand structural concrete, invert concrete, drainage infill mortar, TRU waste package infill andvault backfill), highly alkaline water is produced which can alter nearby buffer (cementbentonite reaction), but this reaction is extremely slow and hence the extent of alteration islimited and the safety functions of the buffer are maintained (see Supporting Report 6-8). The
buffer may also react with dissolved iron (iron-bentonite reaction) supplied from steel
components of the repository as described above, but alteration is slow and similarly limited.
Hydrogen is generated by corrosion of iron under reducing conditions, but the amount
generated from rock bolts and steel rebars outside the buffer is limited: thus it is assumed thatgenerated hydrogen dissolves in groundwater and quickly diffuses into the rock.
Gas conductivity of the saturated buffer is small and hence hydrogen generated inside thebuffer (e.g., from the HLW overpack or TRU waste package) may exceed solubility in thebuffer porewater and accumulate in the gas phase at interior buffer interfaces. As a result,pressure may rise until it results in breakthrough gas flow, potentially forming flow paths
through the buffer. Compacted bentonite is, however, characterised by self-healing and so,any paths formed will close and the buffer will quickly recover and thereafter maintain its
safety functions [14] [49] [63].
As the buffer saturates, HLW overpacks corrode. Shielding by the overpack is sufficient tosuppress any oxidant formation by radiolysis of water (Section 4.4.1 (2) (iii) (c)). The
corrosion rate of steel under reducing conditions is slow and thus the quality-assured
overpack assures the safety function, prevention of contact between waste and groundwater,
throughout this period. This assures no RN release while the buffer experiences hydraulic,chemical and thermal transients.
For TRU waste Grs.3 and 4L without buffer, chemical reductants in the host rock and steel
repository components, will rapidly consume trapped oxygen during corrosion and thereafter
assure reducing conditions. However, it is noted that this assumption may need further
justification under high pH conditions, when corrosion is extremely low. Also, nitrate from
Gr.3 may prevent the environment of this waste becoming completely reducing.
- Backfill and hydraulic plugs
At the near field scale, backfill (a mixture of bentonite and crushed rock) in disposal
tunnels and other access tunnels for HLW and TRU waste, is subject to groundwater intrusion(see to Section 4.5.3 (1)). Swelling of component bentonite assures it exerts the function of
prevention of access routes serving as RN pathways, although its effectiveness gradually
decreases due to reaction with concrete leachate, erosion, etc. Hydraulic plugs are constructedusing bentonite and thus also contribute to assuring this function, especially in terms ofsealing pathways through the EDZ (see Section 4.5.3 (3)).

- Behaviour of repository components without safety functions (see Table 6.3-3)
With a direct focus on release and migration of RNs, many components of the HLW or
TRU waste repository systems do not have specific safety functions. As summarised in Table
6.3-3, the behaviour of these components over this time can be described for the mostprobable evolution scenarios to ensure that they do not impact other EBS safety functions.

6-45

Table 6.3-3 Behaviour and uncertainty of repository components without safety features(Period covered: T1 to T4)
Components

Assumed most likely statePEM handing
shell





Structural
framework





Grout





Steel SupportCondition considering uncertainty

Corrosion and associated volume expansion are complete during time period T2. From period T3
onwards, no further alteration occurs.
The change in the stress field in the near field due to volume expansion is small and is buffered by theswelling of the buffer material (see Supporting Report 4-16). The corrosion products increase masstransfer resistance and may inhibit RN migration.



Taking into account the uncertainty in the corrosion rate,
the time for complete corrosion may be shorter. In thiscase, the most likely impact is an increase in mass transfer
resistance, implying that uncertainty would not negatively
affect the source term.

As the groundwater infiltrates, the concrete component gradually dissolves from the surface inwards.
By the end of period T2, the dissolution of the concrete component has progressed throughout thewhole structure and is then complete. As the dissolution progresses, the hydraulic conductivity of thestructure increases.
In low salinity groundwater, the internal rebars retain their passive film and so the corrosion rate isvery low, so that no new cracks are formed due to corrosion expansion.
For a repository with high salinity groundwater, in Neogene and Pre-Neogene sediments, thecorrosion rate of the rebars is higher and the hydraulic conductivity is increased by cracking due tocorrosion expansion (see Supporting Report 6-16). Corrosion of rebars in high salinity groundwater
progresses through the entire structural framework by time period T2 and no further alteration occurs
from time period T3 onwards.



Uncertainty in the rate of dissolution of the concrete
components may lead to an earlier dissolution of the entire
structural frame.
Uncertainty in the state of the cracks during construction
may lead to higher hydraulic conductivity, even in theinitial state.

The grout dissolution progresses and is complete throughout by period T2. The dissolution of thegrout is complete over the whole area by T2 and no further alteration occurs from T3 onwards.
Such dissolution increases the hydraulic conductivity of the grout-filled area.



Uncertainty in the rate of grout dissolution may lead to an
earlier loss of the concrete component.

Corrosion proceeds by consuming residual oxygen, which contributes to the recovery of thegroundwater to a reducing state. Thereafter, the reducing corrosion progresses slowly, and by period
T2, the corrosion is complete throughout. No further alteration occurs from T3 onwards.
Referring to the results of the assessment of the effect of hydrogen gas generated by corrosion of
HLW overpacks and TRU waste package containers (see Supporting Report 6-8), the generatedhydrogen gas is rapidly transported into the surrounding host rock by dissolution into groundwater. In
addition, volume increases due to corrosion products.
The mechanical effect of the volume increase due to corrosion products on the EBS is very small (seeSupporting Report 4-28). Therefore, the effects on the field conditions for RN migration, such asincreased hydraulic conductivity, are negligible.



Uncertainty in the corrosion rate of steel supports may leadto earlier complete corrosion.

6-46



Components
Shotcrete

Assumed most likely state

By the end of the period T2, the dissolution of the cementitious components has progressed
throughout and is then complete. As the dissolution progresses, the hydraulic conductivity of thestructure increases.



Uncertainty in the rate of dissolution of cementitious
materials may lead to an earlier disappearance of theconcrete component.



Corrosion proceeds by consuming residual oxygen, which contributes to the recovery of thegroundwater to a reducing state. Thereafter, corrosion under reducing conditions progresses slowly,
and by period T2, the corrosion is complete throughout. From T3 onwards, no further alteration
occurs. Any hydrogen gas is dissolved in the groundwater and migrates quickly into the surrounding
host rock. Even if rock bolts are completely corroded, the effect of volume expansion is limited andnegligible.
The dissolution of the rock-bolt anchoring material (cementitious material) progresses, and thedissolution is completed throughout by T2. After the period T3, no further alteration occurs. The
hydraulic conductivity increases with the dissolution of the concrete.



Uncertainty in the rate of corrosion of rock bolts may leadto complete corrosion occurring earlier.
Uncertainty in the rate of dissolution of the cementitious
material used to anchor the rock bolts may lead to an
earlier loss of the concrete component.



By the end of the period T2, the dissolution of the concrete components has progressed throughout
and is then complete. As the dissolution progresses, the hydraulic conductivity of the structure
increases.



Uncertainty in the rate of dissolution of cementitious
materials may lead to an earlier disappearance of theconcrete component.



By the end of the period T2, the dissolution of the concrete components has progressed throughout
and is then complete. As the dissolution progresses, the hydraulic conductivity of the structure
increases.



Uncertainty in the rate of dissolution of cementitious
materials may lead to an earlier disappearance of theconcrete component.



During closure, the filling concrete is dissolved, and by the time of T2 the dissolution is complete
throughout. The dissolution is completed throughout by T2, and no further alteration occurs from T3
onwards. With the dissolution of the concrete, the structure becomes highly permeable soon after
backfilling.



Uncertainty in the rate of dissolution of cementitious
materials may lead to an earlier disappearance of theconcrete component.



During closure, the filling concrete is dissolved, and by the time of T2 the dissolution is complete
throughout. The dissolution is completed throughout by T2 and no further alteration occurs from T3
onwards. With the dissolution of the concrete, the structure becomes highly permeable soon after
backfilling.



In view of the uncertainty regarding the high quality of theconcrete filling in the construction, the hydraulicconductivity of a part may be high immediately after
closure.



The decomposition of the synthetic resin material progresses and is completed during period T2. The
degradation is complete by T2 and no further alteration occurs after T3.
The main degradation products of the synthetic resin material have a low complexing ability with RNs
[64].



In view of the uncertainty regarding the rate ofdecomposition, the time of complete decomposition may
be delayed, but the complexing capacity remains
unchanged.

Rock bolts
Concrete liner
ConcreteinvertCentral drain

Drain pipe

Waterproof
sheet

Condition considering uncertainty



6-47



Components

Assumed most likely stateMechanicalplugPermeable layer (crushed rock)
Filter material(glass fibre)

Drainpipe

Corrosion of the rebar progresses and is complete throughout by period T2. After period T3, no further
alteration occurs. Corrosion expansion leads to increased cracking and hydraulic conductivity of thestructure.
By the end of the period T2, the dissolution of the concrete component has progressed throughout andis then complete. As the dissolution progresses, the hydraulic conductivity of the structure increases.
After backfilling, this exists as a highly permeable structure. This condition persists throughout theentire period T1 to T4.



Uncertainty in the corrosion rate of reinforced concrete
may lead to an earlier increase in hydraulic conductivity
due to cracks in the frame.
Uncertainty in the rate of dissolution of cementitious
materials may lead to an earlier loss of the concrete
component.



－



The dissolution of the glass fibre progresses and by period T2 is complete. After T3, no further
alteration occurs. As the glass fibre dissolves, the hydraulic conductivity increases.



Uncertainty in the dissolution rate of glass fibres may leadto an earlier disappearance.



The decomposition of the synthetic resin material progresses and is complete by period T2. No further
alteration occurs from T3 onwards.
The main decomposition products of the synthetic resin have little complexing ability with RNs [64].

Given the Uncertainty in the rate of decomposition, thetime of complete decomposition may vary, but thecomplexing capacity remains unchanged.

The dissolution of the glass fibres progresses and is complete by T2. No further alteration occurs fromT3 onwards. As the glass fibres dissolve, the hydraulic conductivity increases.
The dissolution of the concrete progresses and is complete by T2. No further alteration occurs from T3
onwards. The hydraulic conductivity increases with the dissolution of the concrete.



Uncertainty in the dissolution rate of glass fibres may leadto an earlier disappearance.
Uncertainty in the dissolution rate of concrete may lead toan earlier disappearance of concrete components.


Buriedformwork(glass-fibre
reinforced
concrete)

Condition considering uncertainty




6-48



- Host rock
During excavation of disposal holes, tunnels and vaults, an EDZ in which permeability
generally increases is created in surrounding rocks, In addition, hydraulically unsaturated
regions may develop during construction and operation, allowing air ingress and developmentof locally oxidising conditions. The HLW overpack assures the safety function prevention ofcontact between waste and groundwater, during this period, so the safety function of host rock
suppression of migration of RNs is redundant during this period.
Also, for TRU waste, there are assessed to be effectively no RN releases over this period
and the situation is the same. However, high pH leachate from concrete components will be
released from the EBS, so that the surrounding host rock may be altered, but the extent of thisis also assessed to be limited and the safety function, suppression of migration of RNs, will
not be significantly impacted (see Supporting Report 6-8). For TRU waste Gr.3, nitrates arealso eluted, which conservatively can be assumed to result in a plume of oxidisinggroundwater gradually developing, in which the sorption of RNs may be affected (seeSupporting Reports 6-8).
(B) Panel scale- Backfill and hydraulic plugs
At this scale, the backfilled tunnels following swelling of bentonite are designed to havehydraulic properties equivalent to those of the surrounding host rocks (see Section 4.5.3 (1)),
providing the function, prevention of access routes serving as RN pathways. In addition, theinstalled plugs contribute to preventing any possible short-circuit flow though repository
access tunnels (see Section 4.5.3 (3)).
- Host rock
An EDZ occurs in the host rock in the vicinity of excavations and permeability of theseareas generally increases. As noted for the near field scale, RNs are contained in both the
HLW and TRU waste EBS, which means that the safety function of suppressing RN
migration is redundant. During operations, inflowing groundwater will be pumped out,
potentially causing a local draw down of the water table; however, after closure,hydrogeological conditions will return to a state essentially the same as that before repository
construction.
(C) Repository scale- Backfill and hydraulic plugs
The fundamental properties of tunnel backfill and plugs at this scale are generally the same
as for the above panel scale.
Exceptions relate to plugs (and other actions such as grouting) emplaced at points where
tunnels penetrate major water-conducting features (layout-determining features) and to
backfill and plugs in ramps and shafts that connect to the surface and hence penetrate a rangeof other rocks with variable mechanical, hydrogeological and chemical properties. In terms of
post-closure performance, the barrier role of such components is unlikely to be significant,although this would need to be confirmed for particular sites.
6-49

- Host rockAt the repository scale, the impacts on the host rock noted at the panel scale also apply.
(D) Regional scale- Geological environment
The geological environment gradually changes by slow processes such as uplift/erosion
and glacial cycles, but safety functions such as “isolation” and “restriction of migration ofRNs” and the features that determine flow paths to the GBI are maintained for very long times.
For both HLW and TRU waste, no releases of radioactivity occur over this time period, so thelatter function is redundant.
- Biosphere
Although the surface environment is vulnerable to changes on a shorter timescale than any
other part of the system, these are inherently unpredictable so current conditions are assumed
unless significant surface modification occurs (e.g. as a result of glacial cycles).
(b) T2: period after re-saturation until releases of RNs from the EBS(A) Near field scale- EBS
While the HLW overpack maintains the function of prevention of contact between waste
and groundwater, there is no significant physical alteration of contained HLW. As complete
containment can be assured for more than 1 ky after closure, relatively short-lived RNs decay(e.g. Cs-137 and Sr-90) and hence heat generation and radiation dose significantly decrease
and EBS temperature becomes effectively the same as the original rock ambient.
As mentioned for period T1, PEM handling shells are not assigned a containment function
and hence buffer re-saturation and overpack failure times are assumed the same for both
H12V and PEM options. As corrosion progresses, the mechanical performance of theoverpack gradually decreases until it eventually fails mechanically, losing the function of
prevention of contact between waste and groundwater. The period until the loss of function isestimated to be about 17 ky (realistic estimation: see Section 4.4.1 (2) (ix)), even when
accounting for the various uncertainties involved. It is, however, conservatively assumed that
this function will be maintained for only 1 ky after closure to allow for consideration ofdesigns with less durable overpacks. Indeed, at 1 ky after closure, containment is assumed to
be lost by all overpacks although, in reality, failures would be spread over an extended periodof time. This very conservative assumption will need to be re-assessed when developing sitespecific overpack designs.
As discussed for T1, alteration of the buffer by dissolved iron occurs at the interface
between the overpack and PEM handling shell, but its extent is limited (see Supporting
Report 6-8) while hydrogen produced by this corrosion can be lost without disruption of
barrier functions [49] (see Supporting Report 6-8).
For TRU waste, release of RNs into groundwater is assumed to occur after re-saturation,conservatively ignoring containment by the waste package container or primary waste
containers (drums or canisters). Such releases are constrained by the functions of suppressing
6-50

elution of RNs from waste or waste packages and reducing migration of RNs due to sorption
in packaging or infill. Such functions gradually deteriorate due to slow leaching/alteration of
concrete components of the EBS. Sorption of RNs onto the infill material is not assumed forGrs.2 and 4H, because of alteration of concrete components by heat.
Leaching of RNs from most TRU waste in contact with groundwater is assumed to occur
rapidly. However, in the case of metal wastes (Gr.2 and some components of other wastes),
the leaching of RNs progresses congruently with metal corrosion. The leached RNs diffuse
through the waste package whilst sorbing onto the cementitious infill and waste package
container. Some RNs, such as C-14, may also be released as gases, which may dissolve ingroundwater or be transported with other gases (e.g., hydrogen from corrosion) [14].
RNs leached from the waste after contact with water will diffuse within the waste package
and sorb onto cementitious materials infilled or around the waste package. Concrete infill
gradually leaches and the solution composition in contact with it eventually evolves from
Region I (pH > ≈ 13) towards Region II (pH ≈ 12.5) (see Supporting Report 6-15).
Especially in the case of TRU waste Gr.2 containing degradable organic substances,solubility of some RNs may be enhanced by complexation and sorption reduced (SeeSupporting Report 6-17). This may also happen for bituminised waste Gr.3, while nitratefrom this waste may act as a complexant or oxidant, having similar impacts (see Supporting
Reports 6-8, 6-20).
Alteration progresses at the interface between buffer and concrete, but the thickness of thislayer is still limited (see Supporting Report 6-8). Conversion of buffer to Ca-type causes anincrease of hydraulic conductivity and a deterioration of the swelling performance, but
specifications are set so that desired functions can be assured even when this occurs (seeSections 4.4.1 (3), 4.4.2 (4)). Water chemistry resulting from reaction of buffer and concrete
is buffered, maintaining RN solubility and sorption performance. For TRU waste Grs.2 and
4H, heat output gradually decreases causing temperature in the EBS to return to rock ambient(see Supporting Report 4-41).
Hydrogen gas continues to be generated due to corrosion of metals, but dissipates into thehost rock without perturbing the safety functions of the buffer.
- Backfill and hydraulic plugs
Performance as during period T1.
- Behaviour of repository components without safety functions
Corrosion and alteration of the components listed in Table 6.3-3 continue as during period
T1.
- Host rock
The EDZ is assumed to persist during this period, with hydrogeology and hydrochemistryelsewhere restored to original undisturbed states. While RNs are contained by the HLW
overpack or within the TRU EBS, the host rock safety function of the “Restriction of RN
migration” can be assured, but is redundant.

6-51

The volume of host rock influenced by the spreading high pH/nitrate plumes resulting fromconcrete leaching and TRU waste Gr.3 will increase, but the range of any alteration will still
be limited and would not impacts its safety functions (see Supporting Report 6-8, 6-22).
(B) Panel scale- Backfill and hydraulic plugs
The backfill and plugs will continue to degrade but their essential function of prevention ofaccess routes serving as RN pathways will be maintained and short circuit flow via suchtunnels can be precluded.
- Host rock
Over this period, the essential functions of the host rock are basically the same as for thenear field scale described above.
(C) Repository scale- Backfill and hydraulic plugs
Over this period, performance at this scale is as during T1.
- Host rock
Performance is as near field and panel scale.
(D) Regional scale- Geological environment
As T1.
- Biosphere
As T1. Although during this period impacts due to climate change are possible (e.g.
altering characteristics of GBI and surface water flows), these are irrelevant as no RN releases
occur at this time.
(c) T3: period after releases of RNs occur during which the characteristics of thegeological environment are considered not to have significantly changed
The period during which the characteristics of the geological environment are considered
not to have significantly changed is assumed to last ≈100 ky.

6-52

(A) Near field scale- EBS
After the HLW overpack loses the function of prevention of contact between waste andgroundwater, the EBS is at rock ambient temperature and the borosilicate glass matrix begins
to dissolve in water with a constant chemistry resulting from the reaction between buffer
minerals, overpack corrosion products and groundwater. RNs are released congruently withthe glass matrix.
Even after loss of containment, the generation of hydrogen gas continues until all steel
components are completely corroded. In addition, the volume of corrosion products continues
to increase, but the mechanical impact is minimised by the plasticity of the buffer. Buffer
deformation due to expansion of overpack corrosion products and sinking of the dense
overpack under gravity modifies the distribution of bentonite density. However, this is taken
into account during design and hence there are no significant impacts on its safety functions(see Section 4.4.1 (3) (iv)).
Chemical alteration of buffer due to reactions with groundwater, iron from overpack
corrosion and interactions with any concrete structures present continues, but the spatial
extent of this is still limited and hence safety functions of low hydraulic conductivity, RN
migration dominated by diffusion and colloidal filtration can all be assured (see Supporting
Report 6-8).
Porewater in the EBS is in equilibrium (or steady state) as a result of groundwaterinteractions with iron corrosion products and bentonite, which buffer any impacts of waterradiolysis and assure pH and redox conditions that lead to low solubility and high sorption ofmost RNs.
In the diffusion-dominated environment within the failed overpack, glass dissolution
releases silica into solution until a limiting concentration is reached, after which point theglass dissolution rate is constant and very low (termed the residual dissolution rate) [53] [54].
Unless transport rates through the bentonite are sufficiently high, RNs released will
precipitate after solubility limits are reached. Some RNs have very high solubilities (e.g. I129, Cl-36 and Cs-135), and hence concentrations are assumed to increase until loss bydiffusion matches supply from glass dissolution.
For TRU waste, after the waste package container fails, RNs are eluted from the waste
matrix involved for the different groups and migrate through the EBS, with release andtransport constrained by solubility limits, gradients driving advective/diffusive flow andassociated material permeability/diffusivity and retardation processes (generally covered bysorption).
Colloids containing RNs may be formed in the waste package, but they are not stable inporewater with high ionic strength (Ca concentrations) that result from reaction ofgroundwater in the host rock with cementitious materials in the structural framework andbetween and within the waste packages. In porewater with high ionic strength, colloids are
unstable and precipitate by flocculation [14] [56]. Therefore, RNs are assumed to migrate as
solutes without forming, and/or sorbing onto, colloids.
In the current assessment, release constraints set by slow waste matrix degradation isconservatively ignored and instant releases are assumed, apart from the metallic waste in Gr.2.
In a more detailed, less conservative, assessment such release processes should be describedin detail and associated uncertainties assessed.
6-53

Cementitious components in TRU waste packages and other wastes, as well as in the infill
material, are further dissolved, with dissolved RN migration constrained by sorption reactions
on solids present. In the case of high heat wastes Gr.2 and 4H, no sorption function isassumed due to possible thermal degradation of the infill material. As previously noted, in thecase of TRU waste Gr.2, solubility of some RNs may be enhanced by organic complexationand sorption reduced (See Supporting Report 6-17) while nitrate from waste Gr.3 may act as acomplexant or oxidant, having similar impacts (see Supporting Reports 6-8, 6-20).
TRU waste Grs.1, 2 and 4H include a buffer, which will continue to interact withcementitious material over this time, but the extent of alteration is limited and key safetyfunctions (preventing advection, filtering colloids, buffering water chemistry, limiting RN
solubility, RN sorption) are assumed to be maintained (see Supporting Report 6-8) for theentire period.
The corrosion of metals (metals in the waste, waste packages, reinforcing bars (rebars) in
the structural framework) and the radiolysis of water in a reducing environment will continue
to produce hydrogen gas, but if it migrates through the buffer either as a solute in porewateror in the gas phase, it will dissolve and dissipate quickly in the groundwater in the host rock,thus maintaining the safety function of the buffer [14].
- Backfill and hydraulic plugs
The backfill and plugs will continue to degrade but their essential function of prevention ofaccess routes serving as RN pathways will be maintained during this period and short circuit
flow via such tunnels is assumed to be precluded.
- Behaviour of repository components without safety functions
Corrosion and alteration of the components listed in Table 6.3-3 continue until complete
degradation occurs.
- Host rock
Conservatively, the EDZ is assumed to persist, although some self-healing may occur insofter rocks. Over this period the host rock generally maintains its original mechanical,hydrological and chemical properties: changes may occur over this time due to processes likeglacial cycles, especially in coastal locations, but these are very site-specific and notconsidered at the present time. For both HLW and TRU waste, the safety function of hostrock restricting releases of radioactive substances, applies for any RNs released from the EBS.
RN transport through the host rock in plutonic rocks and Pre-Neogene sediments
predominantly occurs in a fracture network, with advection/dispersion in the fractures andmatrix diffusion into, and sorption onto the minerals within, the surrounding rock (seeSections 3.3.3 (5) (a) and (c)). Transport through, and reactions with, fracture infill areconservatively neglected at present. In the Neogene sediments, in addition to advection in
fracture networks, RN transport by advection/dispersion may occur in the rock matrix,
associated with sorption onto minerals contacted (see Section 3.3.3 (5) (b)). Since bentonite
colloids are considered unlikely to occur in the model groundwater in this report (seeSupporting Report 6-9), RNs transferred to the host rock will migrate as solutes without
sorption to colloids.

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The volume of host rock influenced by the spreading high pH/nitrate plumes resulting fromconcrete leaching and TRU waste Gr.3 will increase, but it is currently assumed that, for atleast some of the flow path, the host rock RN migration limitation function is maintained forthis period (see Supporting Report 6-8) although sorption of RNs will be affected within theseplumes (see Supporting Reports 6-8, 6-22). It is not expected that the organic material insome of the TRU wastes will result in a significant organic plume in the host rock.
(B) Panel scale- Backfill and hydraulic plugs
As for near field scale.
- Host rock
As for near field scale.
(C) Repository scale- Backfill and hydraulic plugs
As for near field and panel scales.
- Host rockConstant hydrogeological rock properties at this scale are presently assumed, as for thenear field and panel scales. Over this time period, for the sites considered, RNs gradually
migrate along the groundwater flow paths defined by rock structures and hydrogeologicalboundary conditions that are effectively constant, whilst sorbing onto the minerals contacted.
In principle, the flow path length is defined by regional hydrogeology, but assessment
considers short-circuits to the GBI through major faults or undefined features at thedownstream boundary of the repository scale model.
As at smaller scales, the portion of host rock influenced by the spreading high pH/nitrate
plumes resulting from concrete leaching and TRU waste Gr.3 will increase, but its range will
be limited by mass-balance constraints and it is presently assumed that the host rock RN
migration limitation function is maintained to some extent, but with increasing uncertainties(see Supporting Reports 6-8, 6-22).
(D) Regional scale- Geological environment
The geological environment continues to change gradually due to uplift, subsidence,erosion and climate/sea level change, with the thickness of overburden above the host rock
generally decreasing along with depth below surface, but the safety functions “isolation” and“restriction of migration of RNs” are assumed to be maintained over this period (Supporting
Reports 3-36, 6-10). RNs released from the host rock are assumed to travel through shortcircuits to the GBI provided by major water-carrying features, being retarded by sorption on
infilling minerals.
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- Biosphere
During this period, RNs will reach the GBI and then potentially impact populations livingin the surrounding area at that time. Before and during the time of RN release, it is likely that
the GBI will change, along with both surface hydrology and human lifestyles. As suchchanges are inherently unpredictable, distribution of RNs through the surface environment
and food chains and resultant radiation exposures utilise stylised biosphere models based on
biosphere characteristics and human lifestyles typical of Japan at the present time.
(d) T4: period during which the uncertainty in characteristics of the geologicalenvironment increases with time
In general, the geological environment of well selected sites is considered to be stable forat least ≈100 ky, this being the period covered by T3 above. However, as time increases, sodoes the likelihood that geological environment changes will significantly affect the safetyfunctions of the repository.
Uplift and erosion are key natural perturbations that may have a significant influence on
the likely evolution of a Japanese repository. This report assumes that a site with a highprobability of being suitable is selected (see Section 1.4.1). Even with this assumption, it is
not possible to constrain impacts on RN release and migration behaviour on timescales in theorder of 1 My at repository design depths (Section 4.3); although overburden decreases, its
impact is constrained by uncertainties related to uplift/erosion rates. Thus, in the current
assessment, it is simply assumed that there will be no decrease in depth or associated THMC
state changes (see Supporting Report 6-10). However, it is emphasised that models coveringsuch time periods are used to illustrate general trends and output is not to be considered as
fully quantitative. Nevertheless, more realistic assessments of at least some of the time period
beyond 100 ky may be possible for specific sites.
(ii) Description of variant scenario
As described in Section 6.1.5 (2), the variant scenario capture uncertainties in theunderstanding and conceptualisation of the evolution of the repository not captured in thebase scenario, with a focus on impacts on safety functions. For the base scenario describedabove (see Section 6.3.2 (3) (i)), uncertainties in each component of the repository impacting
safety functions are extracted by impact analysis (see Supporting Report 6-9) and thefollowing variant scenario were developed.

(a) Variant scenario for HLW
Scenario in which the rate of glass dissolution could be greater than the base scenario wereconsidered, taking into account the uncertainty due to the following factors influencing glassdissolution.
Changes in the extent of cracking due to overpack corrosion expansion
Although the manufacture of HLW is carried out under strict quality control, cracks do
occur during cooling. After the HLW is encapsulated in an overpack and disposed of in therepository, the stress field changes due to swelling of the buffer during re-saturation andvolume expansion caused by corrosion of the overpack, but this does not affect the initial
6-56

extent of cracking described above (base scenario). However, knowledge of changes inapplied stress and consequent changes in fracturing after disposal is limited and uncertain, so
this is treated in the variant scenario.
Concentration of dissolved silicate in groundwater in contact with HLW
Following overpack corrosion and loss of its integrity, the glass comes into contact withbuffer porewater and begins to dissolve, resulting in a high concentration of dissolved silicate(base scenario). The concentration of dissolved silicate in the solution in contact with glass is
uncertain, as it may decrease due to either sorption onto the corroded overpack or buffer or
co-precipitation. These are taken into account in the variant scenario, as uncertainties
associated with the concentration of dissolved silicate have a significant effect on glassdissolution rate.
(b) Variant scenarios for metal wastes
RNs contained in Gr.2 wastes are leached into groundwater due to congruent corrosion of
metals such as zircaloy (base scenario). The corrosion rate of zircaloy is affected by both thetemperature and extent of hydride formation. Considering the uncertainty of this corrosion
rate due to the temporal variation of repository conditions, this is developed as a variant
scenario.
(c) Variant scenarios covering uncertainty in spread of the nitrate plume
A variant scenario covers uncertainty in the spread of the nitrate plume from Gr.3 waste
into downstream host rock, which would affect the containment safety function of the hostrock. The base scenario assumes this occurs, but only to a limited extent (base scenario).
However, the extent of spread of nitrate from each waste package is uncertain because it
depends on the distribution of hydrological and chemical conditions on a repository scale andtiming of water contact within each waste package. In view of this uncertainty, the variant
scenario assumes a wider plume spread than the base scenario.

(d) Variant scenarios for temperature effects on solubility
The solubility of RNs is affected by temperature, which is assumed to decrease to that of
the ambient host rock temperature by the time the RNs are leached from higher heat emitting
waste, and it is assumed there is no significant temperature dependence of solubility between
room temperature and rock temperature (base scenario). However, the temperaturedependence of solubility is uncertain and would need to be revisited in future safetyassessments. The timing of the onset of RN leaching is particularly uncertain for TRU wastes,
as it depends on the corrosion of the waste package container and the leaching behaviour ofthe cementitious material in the waste package. Due to these uncertainties, the leaching of
RNs may start when the temperature in the repository is higher than rock ambient.
Uncertainties in temperature that increase solubility are treated as variant scenarios.
For the repository components that have not been given a safety function, Table 6.3-3
shows that the most plausible state, described in the base scenario, is that reaction withgroundwater will progress further with time from the state assumed in T1 to the state assumed
in T2 and continues after T3. In the variant scenarios, the most plausible states of the
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alteration process are treated by setting up states that take into account associateduncertainties, which are also shown in Table 6.3-3.

(iii) Description of low probability perturbation scenarios
Low probability perturbation scenarios are defined to cover the inherent uncertainty thatremains in descriptions of very long timescale evolution, even if site selection and facility
design are appropriately carried out. Despite the probability of occurrence of major
perturbations being very low over the time periods considered in Sections 6.3.2 (3) (i) and6.3.2 (3) (ii), the repository system aims to be sufficiently robust that, even if these were tooccur, there is no major radiological impact (see Section 6.1.5 (2)).
Volcanic/magmatic activities and active fault movement are important perturbing natural
events in Japan which, although they have a very small probability of occurrence for relevant
geological environments (Supporting Report 6-9), may have a significant impact on therepository.
As described in Section 3.4.1, in the forearc region of Tohoku and Shikoku,
volcanic/magmatic activity is not credible (negligibly low probability) over the next 1 My or
so. However, in other areas, this cannot be assured for times greater than ≈100 ky. Thus, after
100 ky, such activity is assumed to occur, resulting in a “new volcano scenario”.
In terms of major fault movement directly perturbing a deep repository, as shown in
Section 3.4.2 and discussed in Supporting Report 3-35, it is unlikely that relatively smallscale faults at a repository scale could develop into much larger active faults over relevanttimescales. Nevertheless, this assumption is the basis of the developed “fault perturbation
scenario”. Here, the probability that a fault may directly impact the repository was calculated
to be in the order of 10-7/y (see Supporting Report 3-35).
In this report, the safety of repositories located within the three representative SDMs isdiscussed without considering regional aspects. At present, therefore, the probability ofoccurrence of these such scenarios is taken to be the same for the three host rocks.
With regard to very unlikely natural perturbations (see Table 3.1-3), calculations of doses
involve quantitative evaluations based on data that is currently very uncertain. After sites are
specified, however, better founded evaluations will be made based on the more detailed
information obtained through site investigations.
6.3.3 Setting analysis cases
Based on the scenario descriptions, analysis cases are set in order to quantitatively evaluate
their radiological impacts. In setting these cases, the required models and datasets are
summarised, and then used for the RN migration analyses and associated dose evaluations. It
should also be noted that the migration calculations extend until 10 My, i.e., much beyond thetime when repository and site evolution can be predicted. In this report, with the aim ofcomparing the behaviour of repositories for specific SDMs to the generic H12 and TRU-2
assessments, it was decided to continue the evaluation to determine the time of appearance ofthe maximum dose (see Section 6.1.4).
The base case, which are the analysis case corresponding to the base scenario, are set
according to the most reasonable assessment of safety functions and their evolution in time.
For the variant analysis cases, corresponding to specific variant scenarios, the influence of
6-58

different models and data that capture scenario uncertainties are investigated, whilst
maintaining links to the base case migration models and datasets. As described in Section
6.1.5 (2), alternative evaluation models and data sets are also presented, considering
uncertainties in the base case models and data.
For low probability perturbations scenarios, analysis cases (hereinafter referred to as “lowprobability perturbation cases”) are set by establishing appropriate models and data for eachof the selected scenarios.
The analysis cases corresponding to each of these classes of scenario are described below,
with more details provided in Supporting Report 6-11.
(1) Setting analysis cases for base scenario
As noted above, there is no major difference in the description of the base scenario for the
H12V and PEM options. It is understood that this is a great simplification, e.g., evolution of
the system when the PEM shell ruptures and how this affects the buffer and surrounding
backfill could affect near field migration. However the simplified model used in thisassessment does not consider such impacts. Nevertheless, it is clear that the definition of basescenario will need to be revisited for future, more realistic analyses.
As the goal is RN release and migration analysis, the HLW base scenario focus on time
period T3 (see Section 6.3.2 (3) (i) (c), Table 6.3-4 and Table 6.3-5). Table 6.3-4 summarises
the migration behaviour of RNs in the base scenario and the way of handling this in theassociated quantitative analysis. Table 6.3-5 shows the most probable state assumed forcomponents of the repository listed in Table 6.3-4 and the impact analysis table for safetyfunctions (Supporting Report 6-9). The “most probable states” for the components of therepository that have not been assigned safety features are shown in Table 6.3-3 and, for eachof these components, treatment in the RN migration analysis is described. As described in thebase scenario, no RN migration occurs up to period T2, but the change in repository
conditions prior to period T3, are captured in their specification when RN migration starts. In
period T4, the repository behaves in the same way as in T3.
The process is similar for TRU waste, which also focuses on time period T3 (see 6.3.2 (3)(i) (c)) while Table 6.3-6 and Table 6.3-7 summarise the handling of RN migration analyses
and the state of each component, respectively.
Based on the concepts presented here, the background is set for the RN migration models
and datasets described in Section 6.4.1.

6-59

Table 6.3-4 RN release and transport for HLW base scenario (T3: period after releases of RNs occurs during which the characteristics of thegeological environment are considered not to have significantly changed)
Description of RN migration behaviour in base scenario

Concept of handling in RN migration analysis

Congruent release of RNs with glass dissolution.



The concentration of the RNs in the groundwater on the inner side of the buffer is limited
by solubility; precipitated RNs will remain on the inner side of the buffer (between
overpack and buffer).
There is the possibility that the formation of colloids on the inner side of the buffer will
result in the uptake of RNs, but because of the colloidal filtration function of the buffer,
colloidal RN transport will be inhibited.

RNs dissolved in the porewater migrate through the buffer material by diffusion whilesorbing onto minerals, being released to the EDZ.
Since bentonite colloids are unlikely to be generated in the model groundwater set up in
this report, RNs migrate to the host rock as solutes without sorption to bentonite colloids.



Representation of RN leaching by constant rate.
Set the release rate according to the dissolution rate of glass.

Consider dissolution and precipitation of RNs in the region on the inner side of thebuffer, assuming thermodynamic equilibrium and redox buffering by iron corrosion
products.
Consider migration only of RNs dissolved in porewater.










Volume between glass matrix and buffer modelled as a mixing tank.
Consider RN migration by Fickian diffusion with linear sorption in buffer andempirical correction for charged microporous structure.
Outer boundary modelled as mixing tank in EDZ.
Sorption by glass alteration products, overpack corrosion products, and also transfer
path restriction in fractured glass/overpack is ignored.
The EDZ is treated as a porous media and the RNs mix instantaneously withgroundwater.
Consider migration by advection on the downstream side of EDZ.
Simplify fractures or other discontinuities in which advective flow occurs and modelwith parallel plates.
Consider diffusion from fracture into rock matrix and sorption to mineral surfaces.
Sorption on the fracture surface is conservatively not taken into consideration.

RNs in the EDZ are transferred through the host rock by slow groundwater flow throughthe matrix and/or structural discontinuities, with diffusion into connected non-flowing
porosity and sorption onto mineral surfaces contacted.

Randomly distributed fractures form heterogeneous 3D network, in which RN migration
takes place.



Express the characteristics of RN migration in a 3D heterogeneous medium with onedimensional model.

Radioactive decay during transition.

Consider RN decay and ingrowth.

After reaching a major fault, RNs migrate through it and reaches the biosphere.



Set up a pathway in which RNs migrate through host rocks to major faults and then to
GBI.



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


Table 6.3-5 Analysis of components within the HLW base scenario (T3: period after release of RNs occur during which the characteristics of thegeological environment are considered not to have significantly changed) (1/2)
Component

Safety function

Assumed state




Glass

Reduction of elution
by glass matrixOverpack

Prevention ofcontact between
waste andgroundwater

Prevention of RN
migration by
advection

Buffer

The earliest loss of safety functions after 1 ky.
Volume expansion due to iron corrosion products continues.
As the fabrication canisters are not expected to prevent contact withgroundwater, water will immediately come into contact with the glass after the
overpack loses its safety function.








RN sorption

Groundwater contacts the vitrified waste and the glass dissolves.
Dissolution of the glass decreases with the resulting increase of the dissolved
silicate concentration and then reaches a constant long-term dissolution rate.
There is no significant impact on the concentration of dissolved silicate due toreactions involving iron corrosion products and/or buffer.
There is no noticeable modification of cracks caused during glass production.
The temperature at the time of glass dissolution is ambient rock temperatureand does not change.
The concentration of RNs is limited by their solubility in porewater resulting
from reaction of groundwater, buffer and iron corrosion products of theoverpack. RNs that exceed the solubility limit precipitate.






Prevention ofcolloid migration

Concept of handling in RN migration analysis








Since the uncertainty of the time required for resaturation is large, this time is not set and it is assumed
re-saturation is complete immediately after closing. Resaturation would, in any case, be complete at the time
when the overpack fails.
Apply a glass dissolution model using the long-term
dissolution rate.
Set the glass dissolution rate in consideration of theincrease in surface area due to cracks during fabrication.

In the analysis model, failure assumed after 1 ky.
Overpacks and canisters are not considered in theanalysis of RN migration.

Alteration of buffer due to corrosion of overpacks (and PEM handling shells, if
present) is restricted to the interface and, since most parts are sound, lowhydraulic conductivity is assured and RN migration is dominated by diffusion,
so the safety function is maintained.
The buffer deforms due to corrosive expansion of overpack and overpack
sinking; the distribution of bentonite density changes, but the safety function is
maintained by the design specifications.



Only solute transport by diffusion is handled.

Because deterioration of buffer due to overpack corrosion is limited to theinterface and, since most part is sound, colloid filtration function is maintained.
The buffer deforms due to corrosive expansion of overpack and overpack
sinking; the distribution of bentonite density changes, but the safety function is
maintained by the design specifications.



Do not consider colloidal migration through the buffer.

Consider alteration of bentonite from Na-type to Ca-type
according to the groundwater chemistry when setting the
RN sorption parameters (see Supporting Report 6-15).
Consider changes in the ionic strength of the buffer

The buffer porewater is assumed to be in equilibrium with the bentonite,defining its sorption properties.
Depending on groundwater chemistry, ion exchange of montmorillonite occurs
and the sorption properties change.

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

Component

Safety function

Assumed state

Concept of handling in RN migration analysis
porewater due to reaction with groundwater when setting
the RN sorption parameters.

Protection fromsignificant impactsof natural
perturbations

Reduction of RN
dissolution

Site selection will exclude areas where such risk is significant.



Due to the reaction between buffer and groundwater, the composition of buffer
porewater changes with time.
The concentrations of RNs in porewater are limited according to theirsolubility, set for porewater chemistry with redox set by iron corrosion
products; RNs exceeding the solubility limits precipitate.
The near field temperature is host rock ambient and the solubility limits aredefined for this.

Geologicalenvironment

RN sorption

By definition not considered in these scenario.



Set solubility in consideration of temporal and spatial
variability of buffer porewater.
Consider migration only of RNs dissolved in buffer
porewater.


The influence of the alkaline and nitrate plumes originating from Gr.3 concrete
is limited to the vicinity of the tunnel and sorption performance is maintainedin other areas.
RNs may react with any surfaces encountered, which can generally be
represented as reversible, concentration-independent sorption, although
effectively irreversible mineralisation may also occur.


RN dispersion

Retardation of RN
migration due toslow groundwaterflow velocity








RNs are dispersed by a network of fractures of different sizes, orientations andhydraulic properties.

A slow groundwater flow field is maintained.
Slow advective flow occurs through the matrix and/or structural discontinuities.
An EDZ is generated near tunnels and hydraulic conductivity in this region
increases.
The EDZ is mechanically stable over this time and remains highly permeable,

6-62






Do not consider the effect of high pH in setting the RN
migration parameters of host rock, since only a very
short part of the migration path would be affected.
Model sorption with Kd.
The RN migration analysis is performed assuming thatthe safety function operates only in the host rock area
downstream of the tunnel at the end of a HLW disposal
or of the TRU repository.
Model the flow and migration in multiple realisations of
stochastically generated Discrete Fracture Networks(DFNs).
The average transport characteristics were set based on thestatistically generated DFN.
Apply a stable hydraulic gradient and hydraulicconductivity during the evaluation period.
Consider the influence of discontinuities on the flow ofgroundwater.
Set the EDZ width and hydraulic conductivity.

Component

Hydraulicplug,
Backfillmaterial

Safety function

Prevention oftunnels acting as
short-circuit routes
for RN migration

Assumed stateConcept of handling in RN migration analysis

although it is subject to stresses on the rock mass due to swelling of the buffer
and corrosive expansion of the iron-based components [65].

For RN-transport, the EDZ is conservatively modelled asa mixing tank.

Swelling backfill material seals and gaps between it and the tunnel wall.
When concrete is present, the backfilling material deteriorates due to high pH
leachate and hydraulic conductivity increases.
Alteration of the hydraulic plug is limited and the desired performance ismaintained.

The backfill hydraulic conductivity is set considering theextent of alteration, depending on the amount of concreteused.
The hydraulic plug is treated assuming its designed
function is assured.

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

Table 6.3-5 Analysis of components within the HLW base scenario (T3: period after releases of RNs occur during which the characteristics of thegeological environment are considered not to have significantly changed) (2/2)
Component

Safetyfunction

Assumed most likely statePEM handling
shell

-

Grout

-

Steel support-

Shotcrete-

Concrete liner
Concrete invert

Rock bolts

Central drain-Concept of handling in RN migration analysis

Corrosion and accompanying volume expansion occur. The change in the stressfield of the near field due to volume expansion is small and is buffered by theplasticity of the buffer (see Supporting Report 4-16).
Mass transfer resistance (of solute/gas) increases due to corrosion products andthere is possibility of these suppressing migration of RNs.



Leaching of grout proceeds, possibly increasing hydraulic conductivity locally.



Corrosion under reducing conditions is slow and generated hydrogen gas is
assumed to dissolve in groundwater and migrate into the surrounding host rock.
Mechanically stable, with compaction of the backfill due to volumetric
expansion of the corrosion products.

Complete leaching of the cementitious material and localised hydraulicconductivity increase.

Eventually the rock bolts will completely corrode, fixing material willcompletely leach and hydraulic conductivity will increase.

Is not considered in the analysis model.

Considering the uncertainty in the leaching rate, assume complete
degradation and ignore any effect of reduced hydraulicconductivity of EDZ due to grout.



Alteration is limited and, since the altered part does not become acontinuous structure, it is not included in the model.

Considering the uncertainty of rate of alteration, in the 3D
groundwater flow analysis model, assume complete degradation at
the time of first RN release.



Although the hydraulic conductivity increases, since it does nothave a continuous structure, the influence on the groundwater flow
field is small and is not incorporated into the model.

Because it is a local structure, the influence on the groundwaterflow field is considered to be limited, and therefore is notincorporated into the model.
Since the decomposition products of the drain pipe are distant fromthe RN source, treat as having no influence on the barrier
performance.

-

-

-




Leaching of concrete infill progresses until closure.
After backfilling, it exists as higher hydraulic conductivity feature.

6-64



Component

Waterproof
sheet

Safetyfunction

Assumed most likely state

Concept of handling in RN migration analysis

Complete decomposition of the synthetic resin.
The expected decomposition products of the synthetic resin waterproof sheet
have little complexation with RNs, so their influence is negligible [64].
However, the potential impact of such degradation would need re-assessment
when materials are selected in future designs.

-

Mechanical
plug

-



Corrosion expansion of rebars increases cracks in concrete and increases
hydraulic conductivity.
Complete leaching of the cementitious material and localised hydraulicconductivity increase.

Permeablelayer(crushed rock)-

After backfilling, it exists as a highly permeable feature.

Filter material(glass fibre)-Complete dissolution of the glass fibre.

Complete decomposition of synthetic resin.
The expected decomposition products of the synthetic resin drain pipe havelittle complexation with RNs, so their influence is negligible [64]. However,the potential impact of such degradation would need re-assessment whenmaterials are selected in future designs

Drainpipe-

Buriedformwork(glass fibre
reinforced
concrete)-Complete dissolution of the glass fibre.
Leaching of concrete progresses, and any concrete component disappears.

6-65



Since the decomposition products of the waterproof sheet are
distant from the RN source, treat as having no influence on thebarrier performance.

Because it is a local structure, the influence on the groundwaterflow field is considered to be limited, and therefore is notincorporated into the model.

Because it is a local structure, the influence on the groundwaterflow field is considered to be limited, and therefore is notincorporated into the model.

Because it is a local structure, the influence on the groundwaterflow field is considered to be limited, and therefore is notincorporated into the model.

Consider hydraulic conductivity equivalent to sand.

Because it is a local structure, the influence on the groundwaterflow field is considered to be limited, and therefore is notincorporated into the model.

Table 6.3-6 RN release and transport for TRU waste base scenario (T3: period after releases of RNs occur during which the characteristics of thegeological environment are considered not to have significantly changed)
Description of RN migration behaviour in the base scenarioRNs are eluted from TRU waste and concentration gradients drive releases from thewaste package.
For metal waste contained in Gr.2, RNs contained in the metal leach into porewater asa result of corrosion.
The groundwater in the host rock entering the waste package has reacted with thecementitious and iron-based components and, in Grs.1, 2 and 4H, with the buffermaterial installed.










Some RNs sorb to the cement components of the waste package infill.
Solute migration through the EBS by advection (Grs.3, 4L) or diffusion (Grs.1, 2, 4H).



The concentrations of RNs in the porewater within and around waste packages arelimited by the solubility, and those exceeding solubility limits precipitate.
Colloids are assumed unstable in the high ionic strength of the porewater.

Concept of handling in RN migration analysis








For most TRU waste, elution of RNs into the EBS porewater is treated as being instantaneousafter contact with porewater, with the entire EBS is treated as a mixing tank.
An exception is the metal waste in Gr.2, RNs are handled as being eluted congruently with thecorrosion of the metal, specified by a constant corrosion rate, and then, as described above,
instantly mixed in porewater.
The concentration of RNs is limited by solubility in the porewater within and around waste
packages, set by reactions with cementitious material, iron corrosion products and buffer(Grs.1, 2, 4H).
Consider sorption on waste package infill.
However, in the infill between waste packages in Grs.2 and 4H, there is uncertainty in thesorption by cement hydrates that have been altered by higher temperature, so such sorption isconservatively excluded from the model.
RNs eluted from TRU waste are mixed instantaneously with the porewater within the
structural framework.
Consider dissolution and precipitation of RNs in waste packages, assuming thermodynamic
equilibrium.
Consider migration of RNs dissolved in porewater.
Consider sorption of the RNs in the buffer.
In the buffer, only migration by diffusion is considered.
Outer boundary is defined as a mixing tank in the EDZ (see Supporting Report 6-13).

RNs dissolved in porewater migrate by diffusion, while being sorbed by minerals inthe buffer (Grs.1, 2, 4H).



RNs reaching the EDZ are transferred through the host rock by slow groundwater flow
through the matrix and/or structural discontinuities, with diffusion into connected nonflowing porosity and sorption onto mineral surfaces contacted.
Bentonite colloids are assumed unstable in the model groundwater and only RNs
dissolved in groundwater migrate.

Randomly distributed fractures form a heterogeneous 3D network, in which RN
migration takes place.



Express the characteristics of RN migration in a 3D heterogeneous medium with a onedimensional model.

Radioactive decay during transition.

Consider RN decay and ingrowth.

After reaching a major fault, RNs migrate through it to the biosphere.



Set up a pathway in which RNs migrate through host rocks to major faults and then to GBI.




Consider migration by advection on the downstream side of EDZ.
Simplify fractures or other discontinuities in which advective flow occurs and model withparallel plates.
Consider diffusion from fracture into rock matrix and sorption to mineral surfaces.
Sorption on the fracture surface is conservatively not taken into consideration.

6-66

Table 6.3-7 Analysis of components within the TRU waste base scenario (T3: period after releases of RNs occur during which the characteristics ofthe geological environment are considered not to have significantly changed) (1/2)
Components Safety function

Assumed most likely state

Concept of handling in RN migration analysis

Waste packageReduction of RN
elutionThe waste package container fails (in the case of waste package
B).
Water penetrates into the waste package and RNs are eluted fromthe waste.
For metal waste contained in Gr.2, RNs contained in the metalleach into porewater as a result of corrosion.
The mortar component of infill gradually degrades.
The composition of the solution in contact with the mortar within
the waste package evolves from defined Region I to Region Ⅱ.
The concentrations of RNs in porewater are limited according totheir solubility, set for porewater chemistry by cementitious
materials and buffer; RNs exceeding the solubility limits are
precipitated.
The near field temperature is host rock ambient and the solubility
limits are defined.











Vault
infill

RN sorptionSame as fill within waste packages.

6-67

Since the uncertainty of the time required for re-saturation is large, this time is
not set and complete re-saturation is assumed immediately after closure.
Although the waste package is expected to be resistant to mass transfer during
the period, when unaltered parts of these remain, this resistance is disregarded in
the RN migration analysis because of the large uncertainty about the duration ofthis state of the package.
Elution of RNs begins immediately after groundwater comes in contact with
waste.
For the metal waste in Gr.2, RNs are handled as being eluted congruently withthe corrosion of the metal, specified by a constant corrosion rate.
Assumed the buffer and groundwater are in equilibrium instantly, and consider
the change of porewater chemistry with time and spatial heterogeneity inresponse to groundwater intrusion (Grs.1, 2 and 4H).
Porewater chemistry is further changed by direct contact between the concreteand the groundwater in the host rock, and the solubility of the porewater is set
taking into account the change in solution composition from Region I to Region
II (see "Reduction of radioactive material dissolution" in "Safety features of thegeological environment").
Consider migration only of RNs dissolved in porewater.
See the Table 6.3-6 for the handling of migration in the waste package, infill andstructural framework
For the infill between waste packages in Grs.2 and 4H, there is uncertainty in thesorption by cement hydrates that have been altered by higher temperature, sosuch sorption is conservatively excluded from the model.
With regard to infill in the waste package of Grs.1 and 4L, consider sorption to
cement hydrate and set appropriate distribution coefficient.
For Gr.3, consider the influence of nitrate when setting the distribution
coefficients for cement hydrate.

Same as fill within waste packages.

Components Safety function

Assumed most likely stateBufferPrevention of
RN migration
by advectionPrevention ofcolloid
migration



Geological environmentConcept of handling in RN migration analysis

Alteration of buffer at contacts to concrete is restricted to theinterface and, since most parts are sound, low hydraulicconductivity is assured and RN migration is dominated bydiffusion, so the safety function is maintained.
The buffer deforms due to settling of the vault structure; thedistribution of bentonite density changes, but the safety function
is maintained by the design specifications.



Model only RN migration by diffusion.

Alteration of buffer at contacts to concrete is restricted to theinterface and, since most parts are sound, low hydraulicconductivity is assured and RN migration is dominated bydiffusion, so the safety function is maintained.
The buffer deforms due to settling of the vault structure; thedistribution of bentonite density changes, but the safety function
is maintained by the design specifications.



Migration of colloids is not considered.

Consider alteration of bentonite from Na-type to Ca-type according to thegroundwater chemistry when setting the RN sorption parameters.
Consider changes in the ionic strength of the buffer porewater due to reaction
with groundwater when setting the RN sorption parameters.



By definition not considered in these scenario.



Set solubility in consideration of temporal and spatial variability of EBS
porewater.
The solubility is set taking into consideration the evolutions of the cement
porewater from the Region I to the Region II.
Consider migration only of RNs dissolved in porewater.

RN sorption

The buffer porewater is assumed to be in equilibrium with thebentonite, defining its sorption properties.
Depending on groundwater chemistry, ion exchange ofmontmorillonite occurs and the sorption properties change.

Protection
fromsignificant
impacts ofnaturalperturbations

Site selection will exclude areas where such risk is significant.



Due to the reaction between buffer and groundwater, thecomposition of buffer porewater changes with time.
Depending on the inflow of ground water, cement porewater
evolves from Region I to Region Ⅱ.
The concentrations of RNs in porewater are limited at theirsolubility, set for porewater; RNs exceeding the solubility limits
are precipitated.
The near field temperature is host rock ambient and the solubility
limits are defined for this.


Reduction of RN
dissolution




6-68




Components Safety function

Assumed most likely state

RN sorption


RN dispersion

Retardation of
RN migration due
to slow
groundwater flow
velocityConcept of handling in RN migration analysis

The influence of high alkaline plume originating from concrete islimited to the vicinity of the vault and sorption performance ismaintained in other areas.

Nitrate contained in Gr.3 results in a plume influencing
downstream groundwater chemistry.



It is assumed that nitrate influences only the host rock downstream of Gr.3,
influencing the Kds selected.

Model the flow and migration in multiple realisations of stochastically generated
DFNs. The average transport characteristics were set with statistically generated
DFN.
For the Neogene sediments, flow and migration though the porous rock between
the fractures is also considered.

RNs are dispersed by a network of fractures of different sizes,
orientations and hydraulic properties.

A slow groundwater flow field is maintained.
Slow advective flow occurs through the matrix and/or structural
discontinuities.
An EDZ is generated near tunnels and hydraulic conductivity inthis region is higher.
The EDZ is mechanically stable over this time and remains
highly permeable, although it is subject to stresses on the rockmass due to swelling of the buffer and corrosive expansion of theiron-based components [65].

6-69








Do not consider the effect of high pH in setting the RN migration parameters ofhost rock.
Model sorption with Kd.

Apply a constant hydraulic gradient and hydraulic conductivity.
Set the EDZ width and hydraulic conductivity in the 3D transport model.

Table 6.3-7 Analysis of components within the TRU waste base scenario (T3: period after releases of RNs occur during which the characteristics ofthe geological environment are considered not to have significantly changed) (2/2)
Components

Safety function

Hydraulic plug,
Backfill material

Prevention of tunnels
acting as short-circuit
routes for RN
migration

Assumed most likely stateStructural
framework

Swelling backfill seals gaps between it and the tunnel wall.
When concrete is present, the backfilling material deteriorates due to highpH leachate and hydraulic conductivity increases.
Alteration of the hydraulic plug is limited and the desired performance ismaintained.
In the structural framework, leaching of the concrete occurs gradually as
the groundwater infiltrates.
For low salinity groundwater, the internal rebar has a passive film, so the
corrosion rate is extremely low and new cracks due to corrosion expansion
do not occur to a significant extent during re-saturation.
Corrosion of the rebars is faster in high salinity groundwater, cracks occur
due to corrosion expansion, and the hydraulic conductivity increases.

Mechanical plug-Corrosion expansion of rebars increases cracks in concrete and increases
hydraulic conductivity.
Complete leaching of the cementitious material and localised hydraulicconductivity increase.

Permeable layer(crushed rock)-

After backfilling, it exists as a highly permeable feature.

Filter material(glass fibre)Complete dissolution of the glass fibre.

-Complete decomposition of the synthetic resin.
The expected decomposition products of the synthetic resin drain pipe
have little complexation with RNs, so their influence is negligible [64].
However, the potential impact of such degradation would need reassessment when materials are selected in future designs.

Drain pipe-

Buried formwork(glass fibre
reinforced concrete)-Complete dissolution of the glass fibre.
Leaching of concrete progresses, and any concrete component disappears.

6-70

Concept of handling in RN migration analysis






The backfill hydraulic conductivity is set considering theextent of alteration, depending on the amount of concreteused.
The hydraulic plug is treated assuming its designed
function is assured.

Consider temporal changes of structural framework
hydraulic conductivity.
Set the high hydraulic conductivity in high salinity
groundwater immediately after closing.



Because it is a local structure, the influence on the
groundwater flow field is considered to be limited, and is
not incorporated into the model.

Because it is a local structure, the influence on the
groundwater flow field is considered to be limited, and is
not incorporated into the model.

Because it is a local structure, the influence on the
groundwater flow field is considered to be limited, and is
not incorporated into the model.

Because it is a local structure, the influence on the
groundwater flow field is considered to be limited, and is
not incorporated into the model.
Since the decomposition products of the drain pipe are
distant from the RN source, treat as having no influenceon the barrier performance.

Because it is a local structure, the influence on the
groundwater flow field is considered to be limited, and is
not incorporated into the model.

Shotcrete-

Concrete liner-

Concrete invert-

Steel support(Neogene sediments)Complete leaching of the cementitious material and localised hydraulicconductivity increase.

Corrosion under reducing conditions is slow and generated hydrogen gas
is assumed to dissolve in groundwater and migrate into the surrounding
host rock.
The effect on the EBS as a result of volume increase due to corrosion
products is judged to be negligible.

Rock bolts

-

Grout

-



Eventually the rock bolts will completely corrode, fixing material willcompletely leach and hydraulic conductivity will increase.

Leaching of grout proceeds, possibly increasing hydraulic conductivity
locally.

Complete decomposition of the synthetic resin.
The expected decomposition products of the synthetic resin waterproof
sheet have little complexation with RNs, so their influence is negligible[64]. However, the potential impact of such degradation would need reassessment when materials are selected in future designs.

Considering the uncertainty of rate of alteration, in the
3D groundwater flow analysis model, assume complete
degradation at the time of first RN release.



Alteration is limited and, since the altered part does not
become a continuous structure, it is not included in themodel.



Although the hydraulic conductivity increases, since itdoes not have a continuous structure, the influence on the
groundwater flow field is small and is not incorporated
into the model.

Considering the uncertainty in the leaching rate, assume
complete degradation and ignore any effect of reduced
hydraulic conductivity of EDZ due to grout.



Since the decomposition products of the waterproof sheet
are distant from the RN source, treat as having noinfluence on the barrier performance.

Waterproof sheet-

Central drain pipe-Leaching of concrete infill progresses until closure.
After backfilling, it exists as a highly permeable feature.

Consider hydraulic conductivity equivalent to sand.

Drain

-



After backfilling, it exists as a highly permeable feature.

Consider hydraulic conductivity equivalent to sand.

6-71

(2) Setting of analysis cases for variant scenarios
Following the RN migration approach for constituent repository elements in the base case,these are defined for variant cases in Table 6.3-8 and Table 6.3-9 for HLW and TRU waste,
respectively. It is recognised that the current scenario selection is greatly simplified and willneed to be revised in the future, when more realistic assessment will handle information for
actual sites and consider repository concepts adapted to these sites.
Based on this material, Table 6.3-10 summarises the analysis cases derived for variant
scenarios, along with associated assumptions and simplifications introduced for the RN
migration analysis. In Table 6.3-10, as described in Section 6.1.5 (2), not only variants in theassumed evolution of the repository are presented, but also alternative evaluation models anddatasets to the base case. As described in Section 6.2.2, due to current constraints,
conservative simplifications are introduced when setting both the base and variant scenarios.
Cases that combine variant options are not considered in this report, but will be examined in
the future.

6-72

Table 6.3-8 Handling of analysis cases within the HLW variant scenarios (T3: period after releases of RNs occur during which thecharacteristics of the geological environment are considered not significantly changed) (1/2)
Component

Safety function

Uncertainties or perturbation of the model or data set to be considered-The following uncertainty may increase the dissolution rate of the glass:Uncertainty of initial fracturing
Uncertainty in cracking due to external stress from overpack corrosion
Uncertainty concerning the influence of iron corrosion products andbuffer on glass dissolution.

Glass

Reduction of elution
by glass matrix

Overpack

Prevention of contact
between waste andgroundwaterUncertainty is taken into account in the base scenario.

Prevention of RN
migration by
advection

Consider uncertainties in diffusion coefficient measurement data.

As a result of the impact analysis on the safety function, the buffer
should assure no colloid migration.
However, at least for time scales beyond 100 ky, buffer performance
could degrade (e.g. due to chemical erosion by dilute waters). Such achange is not considered here, but the long-term stability of the buffer
might need further study in the future.

BufferPrevention ofcolloid migration



Study impacts of different glass dissolution rates. This
analysis case is referred to as the “Increase in glassdissolution case (HLW)”.



Included in the base scenario.



Vary buffer effective diffusion coefficients: “Increased
diffusivity in buffer case (HLW) ”.

-

Consider uncertainties in RN sorption and diffusion coefficientmeasurement data.

Protection fromsignificant impactsofnatural perturbations

As a result of the impact analysis on the safety function, performance
should be assured for the first 100 ky. Impacts of likely geosphere
evolution for longer time periods are not yet considered (see Supporting
Report 6-10).

Reduction of RN
dissolutionConsider uncertainty on thermal effects on solubility.
Consideration of the uncertainty in the assumption of thermodynamic
equilibrium and completeness of the thermodynamic database (TDB).



Vary solubility: “Thermal increase in solubility case (HLW)”.
Vary solubility: “Uncertainty in thermodynamic data case
(HLW)”.

Consider uncertainties in measured sorption distribution coefficient anddiffusion coefficient data and other uncertainties in modelling theseprocesses.

Vary distribution coefficients and effective diffusion
coefficients: “ Lower sorption in host rock case (HLW)”,“Increased diffusivity in host rock case” (HLW).

RN sorption

GeologicalenvironmentConcept of handling in RN migration analysis

RN
sorption

6-73



Vary buffer distribution coefficients: “Lower sorption in
buffer case (HLW)”.

-

Component

Safety function

RN
dispersion
Retardation of RN
migration due to slow
groundwater flow
velocity
Prevention of tunnels
Hydraulic plug,
acting as short-circuit
backfillroutes for RN
material
migration

Uncertainties or perturbation of the model or data set to be consideredConsider the uncertainty concerning the nitrate plume from TRU waste
as well as impacts from microbial activity on sorption.

Consider uncertainty concerning the hydraulic conductivity of theprobabilistic fracture network on the degree of dispersion.

Consider uncertainty concerning the hydraulic conductivity of theprobabilistic fracture network on the water velocity.



Even if uncertainty is considered, the desired performance of thehydraulic plug is assumed to be maintained.
When cementitious materials surround backfill, hydraulic conductivity
may increase more rapidly if uncertainty in the reactions between themis taken into consideration.



6-74

Concept of handling in RN migration analysis

Vary distribution coefficients: “Nitrate plume impact case
(HLW)”.

The fracture network realisation with the fastest mass transfer
is used in the analysis case: “Fracture connectivity in hostrock case”.

As above, the fracture network realisation with the fastest
mass transfer is used in the analysis case: “Fracture
connectivity in host rock case”.

Such uncertainty is included in the analysis case for the basescenario, because it is set taking into account suchdegradation.

Table 6.3-8 Handling of analysis cases within the HLW variant scenarios (T3: period after releases of RNs occur during which the characteristics ofthe geological environment are considered not to have significantly changed) (2/2)
Component

Safety function

Uncertainties or perturbation of the model or data set to be considered

Concept of handling in RN migration analysis

Mechanical plug-

The corrosion of rebars and dissolution of cementitious materials is
almost complete during the period T2, which can be regarded as
equivalent to the most probable state (see Table 6.35), taking into
account the uncertainty of corrosion and concrete dissolution rates.

Permeable layer(crushed rock)

-

-



Covered by base scenario analysis cases.

Buried formwork(glass fibre reinforced
concrete)Consider the uncertainty of the dissolution rate of the glass fibre.
Consider the uncertainty of the leach rate of cementitious material andthus the time of its complete loss.

-Even considering such uncertainties, impacts are localised and theinfluence on the groundwater flow field is considered to belimited, so assumed covered by the base scenario.

Filter material(glass fibre)Consider the uncertainty of the dissolution rate of the glass fibre.

-Even considering such uncertainties, impacts are localised and theinfluence on the groundwater flow field is considered to belimited, so assumed covered by the base scenario.

Drain pipe-Consider the uncertainty of its decomposition rate and thus the timeof its complete loss.

Even considering such uncertainties, impacts are considerednegligible, so assumed covered by the base scenario.

ShotcreteConsider the uncertainty of the leach rate of cementitious material andthus the time of its complete loss.
Consider uncertainty in the construction quality, so may be locallyhigh hydraulic conductivity.

Such uncertainty is included in the analysis case corresponding to
the base scenario, because it is set taking into account suchdegradation.

Even considering such uncertainties, impacts are localised and theinfluence on the groundwater flow field is considered to belimited, so assumed covered by the base scenario.

Even considering such uncertainties, impacts are localised and theinfluence on the groundwater flow field is considered to beConcrete liner-

Concrete invert-

Steel support
Neogene sediments)-

Rock bolts

-





Consider the uncertainty of the corrosion rate of the support, and thus
the time of its complete loss.

Consider the uncertainty of the corrosion rate, and thus the time of its
complete loss.

6-75



Even considering such uncertainties, impacts are localised and theinfluence on the groundwater flow field is considered to belimited, so assumed covered by the base scenario.

Component

Safety function

Grout

-

Waterproof sheet

-

Drain

PEM handling shell--

Uncertainties or perturbation of the model or data set to be considered

Concept of handling in RN migration analysis

Consider the uncertainty of the leach rate of cementitious fixing
material and thus the time of its complete loss.

Consider the uncertainty of the leach rate of cementitious material andthus the time of its complete loss.
Consider uncertainty in the construction quality, so may be locallyhigh hydraulic conductivity.

Such uncertainty is included in the analysis case corresponding to
the base scenario, because it is set taking into account suchdegradation.

Consider the uncertainty of the decomposition rate of the waterproof
sheet and thus the time of its complete loss.

Even considering such uncertainties, impacts are considerednegligible, so assumed covered by the base scenario.

Consider the uncertainty of the leach rate of cementitious material andthus the time of its complete loss.
Consider uncertainty in the construction quality, so may be locallyhigh hydraulic conductivity.

Such uncertainty is included in the analysis case corresponding to
the base scenario, because it is set taking into account suchdegradation.

Consider the uncertainty of the corrosion rate of the support, and thus
the time of its complete loss.

Such uncertainties do not impact the safety functions of the buffer
and backfill, so this is covered by the base scenario.






6-76

limited, so assumed covered by the base scenario.

Table 6.3-9 Handling of analysis cases within the TRU variant scenarios (T3: period after releases of RNs occur during which the characteristics ofthe geological environment are considered not significantly changed) (1/2)
Component

Safety function

Uncertainties considered



Waste
package

Reduction of RN
elution

Vault infill

RN sorption

Prevention of RN
migration by
advection

Buffermaterial

Prevention of colloid
migration

RN sorption
Protection from
Geological significant impacts ofenvironment natural perturbations

Consider the uncertainty of steel corrosion rate and the associated time of
failure of containment.
Consider the following uncertainties in the corrosion rate of Gr.2 metal
waste:For Zircaloy, uncertainties concerning the effects of temperatureand hydrogen
For stainless steel, uncertainty regarding the influence oftemperature.
Consider the uncertainty of the concrete leach rate, thermal degradationand the effect by nitrate and its impact on RN sorption.

Consider the uncertainty of the leach rate of concrete and its impact on
RN sorption.

Consider uncertainties in diffusion coefficient measurement data.

As a result of the impact analysis on the safety function, the buffer should
assure no colloid migration.
However, at least for time scales beyond 100 ky, buffer performance
could degrade (e.g., due to chemical erosion by dilute waters). Such achange is not considered here, but the long-term stability of the buffer
might need further study in the future.

Consider uncertainties in RN sorption and diffusion coefficientmeasurement data.

As a result of the impact analysis on the safety function, performance
should be assured for the first 100 ky. Impacts of likely geosphere
evolution for longer time periods are not yet considered (see Supporting
Report 6-10).

6-77

Concept of handling in RN migration analysis



Covered by the base scenario for waste package
A, since the time until elution of the RNs isignored.
Vary corrosion rate of Gr.2 metal waste:“Increased corrosion of TRU Gr.2 case (TRU
waste)”.

Due to uncertainty in the evolution from Region I
to Region II, the lower Kd value for thesebounding cases is already conservatively selected
for the base case.



Vary buffer effective diffusion coefficient:“Increased diffusivity in buffer case (TRU
waste)”.

-

Vary buffer distribution coefficient: “Lower
sorption in buffer case (TRU waste)”.

-Component

Safety function
Reduction of RN
dissolution

Uncertainties consideredConsider uncertainty on thermal effects on solubility.
Consideration of the uncertainty in the assumption of thermodynamic
equilibrium and completeness of the TDB.

Consider uncertainties in measured sorption distribution coefficient anddiffusion coefficient data and other uncertainties in modelling theseprocesses.

Consider the uncertainty concerning the nitrate plume from TRU waste aswell as impacts from microbial activity on sorption.

Consider uncertainty concerning the hydraulic conductivity of theprobabilistic fracture network on the degree of dispersion.

Consider uncertainty concerning the hydraulic conductivity of theprobabilistic fracture network on the water velocity.



Even if uncertainty is considered, the desired performance of thehydraulic plug is assumed to be maintained.
When cementitious materials surround backfill, hydraulic conductivity
may increase more rapidly if uncertainty in the reactions between them istaken into consideration.

RN sorption

RN dispersion

Retardation of RN
migration due to slow
groundwater flow
velocity
Hydraulicplug,
Backfillmaterial

Prevention of tunnels
acting as short-circuit
routes for RN
migration

Structural
framework-

Corrosion of rebars and dissolution of concrete components are almost
complete during T2, except in low salinity groundwater where theseprocesses are considered to be less rapid (see Table 6.3 7).
In the case of low and high salinity groundwater in plutonic rock,degradation due to alteration may progress to a highly permeable state at
an early stage, taking into account uncertainties such as the state ofcracking that occurs during construction and the rate of leaching of
concrete components.

6-78

Concept of handling in RN migration analysis

Vary solubility: “Thermal increase in solubility
case (TRU waste)”.
Vary solubility: “Uncertainty in thermodynamic
data case (TRU waste)”.



Vary distribution coefficients and effective
diffusion coefficients: “Lower sorption in hostrock case (TRU waste)”, “Increased diffusivity inhost rock case (TRU waste)”.



Vary distribution coefficients: “Nitrate plume
impact case (TRU waste)”.



The fracture network realisation with the fastest
mass transfer is used in the analysis case:“Fracture connectivity in host rock case”.

As above, the fracture network realisation withthe fastest mass transfer is used in the analysis
case: “Fracture connectivity in host rock case”.

Such uncertainty is included in the analysis casecorresponding to the base scenario, because it is
set taking into account this degradation.



Assume deterioration of the structural framework
when setting its hydraulic conductivity:“Uncertainty in structural frame degradation case
(TRU waste)”.

Table 6.3-9 Handling of analysis cases within the TRU variant base scenario (T3: period after releases of RNs occur during which thecharacteristics of the geological environment are considered not significantly changed) (2/2)
Component

Safetyfunction

Uncertainties considered

Concept of handling in RN migration analysis

Mechanical plug-Consider the uncertainty in the corrosion rate of the rebars andthus hydraulic conductivity due to resulting cracks.
Consider the uncertainty of the leach rate of cementitious material
and thus the time of its complete loss.

Permeable layer(crushed rock)

-



-



Covered by base scenario analysis cases.

Filter material(glass fibre)Consider the uncertainty of the dissolution rate of the glass fibre.
Consider the uncertainty of the leach rate of cementitious material
and thus the time of its complete loss.

-Even considering such uncertainties, impacts are localised and theinfluence on the groundwater flow field is considered to be limited,so assumed covered by the base scenario.

-

Consider the uncertainty of the decomposition rate and thus thetime of complete loss.

Drain pipe

Even considering such uncertainties, impacts are localised and theinfluence on the groundwater flow field is considered to be limited,so assumed covered by the base scenario.

Buried formwork(glass fibre
reinforced concrete)Consider the uncertainty of its decomposition rate and thus thetime of its complete loss.

-Even considering such uncertainties, impacts are considerednegligible, so assumed covered by the base scenario.

Shotcrete-Consider the uncertainty of the leach rate of cementitious material
and thus the time of its complete loss.
Consider uncertainty in the construction quality, so may be locallyhigh hydraulic conductivity.

Concrete liner

Such uncertainty is included in the analysis case corresponding to
the base scenario, because it is set taking into account suchdegradation.

Concrete invert-

Steel support(Neogene sediments)

-Even considering such uncertainties, impacts are localised and theinfluence on the groundwater flow field is considered to be limited,so assumed covered by the base scenario.

Rock bolts-

Even considering such uncertainties, impacts are localised and theinfluence on the groundwater flow field is considered to be limited,so assumed covered by the base scenario.

Consider the uncertainty of the corrosion rate of the support, andthus the time of its complete loss.

Consider the uncertainty of the corrosion rate, and thus the time ofits complete loss.
Consider the uncertainty of the leach rate of cementitious fixing



6-79



Even considering such uncertainties, impacts are localised and theinfluence on the groundwater flow field is considered to be limited,so assumed covered by the base scenario.

Component

Safetyfunction

Uncertainties considered

Concept of handling in RN migration analysis

material and thus the time of its complete loss.


Grout

-

Waterproof sheet-Consider the uncertainty of the leach rate of cementitious material
and thus the time of its complete loss.
Consider uncertainty in the construction quality, so may be locallyhigh hydraulic conductivity.

Such uncertainty is included in the analysis case corresponding to
the base scenario, because it is set taking into account suchdegradation.

Consider the uncertainty of the decomposition rate of thewaterproof sheet and thus the time of its complete loss.

Even considering such uncertainties, impacts are considerednegligible, so assumed covered by the base scenario.

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Table 6.3-10 Overview of Analysis Cases (1/2)
Analysis casenameHandling in the base and variant cases

Base

Apply a glass dissolution model using the long-term dissolution rate, taking into account theincrease in the area due to cracking during fabrication.

Variant

Consider the following uncertainties to set increases in the dissolution rate of the glass
Uncertainty in cracking due to external stress from overpack corrosion
Uncertainty on the impacts of iron corrosion products and buffer.

Increased corrosion of
TRU Gr.2 case
(TRU)

Base

For Gr.2 metal wastes, the RNs are leached congruently with the corrosion of the metal,
depending on the corrosion rate.

Variant

Consider uncertainties in measurement data, set increased corrosion rates.

Uncertainty instructural framedegradation case
(TRU)

Base

Change the hydraulic conductivity of the structural framework with time.

Variant

Consider uncertainty in the degree of cracking at the time of construction and the leach rate;
set hydraulic conductivity assuming degraded structure from the beginning.

Base

Set host rock RN migration parameters assuming nitrate affects TRU waste Gr.3.

Increase in glassdissolution case
(HLW)

Nitrate plume impact
case (HLW/TRU)

Fracture connectivity
in host rock case
(HLW/TRU)

Variant
Base
Variant

Consider uncertainty in the spread of the nitrate plume and set host rock RN migration
parameters assuming nitrate affects waste located parallel to Gr.3 with respect togroundwater flow direction.
The average migration rate calculated from the probabilistic discrete fracture network(DFN).
Consider uncertainty in the DFN and apply the realisation giving fastest migration (evaluatedusing the time until maximum migration rate of the tracer at the outlet reached).

6-81

Plutonic

Neogene

PreNeogene
Low High

Low

High

Low

High

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

*

✓

*

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

Table 6.3-10 Overview of Analysis Cases (2/2)
Handling in the base case (upper row) andHandling in variant case (lower row)

Analysis case name

Lower sorption in buffer case
(HLW/TRU)

Base
Variant

Increased diffusivity in buffer case
(HLW/TRU)

Base
Variant

Lower sorption in host rock case
(HLW/TRU)

Increased diffusivity in host rockcase
(HLW/TRU)

Thermal increase in solubility case
(HLW/TRU)

Uncertainty in thermodynamic
data case (HLW/TRU)

Base
Variant
Base
Variant
Base
Variant
Base
Variant

Apply the average value of measured apparent diffusion coefficients andeffective distribution coefficient for specified buffer and porewaterchemistry using an empirical formula based on Fickian assumptions
Consider uncertainties in measured data and set lower sorption
coefficients in the database
Apply the average value of the database of effective diffusion
coefficients for specified buffer and porewater chemistry
Consider uncertainties in measured data and set higher diffusion
coefficients in the database
Apply the average value of the database of sorption distribution
coefficients for specified rock and groundwater chemistry
Consider uncertainties in measured data and set lower sorption
coefficients of 95% confidence interval in the database
Apply the average value of measured diffusion coefficients for specified
rock and groundwater chemistry

Plutonic

Neogene

Pre-Neogene

Low

High

Low

High

Low

High

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

Consider uncertainties in measured data and set higher diffusion
coefficients in the database
It is assumed that the influence of the temperature on solubility is
negligible
Taking into consideration the uncertainty of solubility with respect topossible repository temperatures, set a higher solubility
Apply the derived solubility limits calculated from the thermodynamic
database

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

✓

Consider associated uncertainties and set higher solubility limits

✓

✓

✓

✓

✓

✓

High: high salinity groundwater, Low: low salinity groundwater, ✓: application to variant cases
HLW: HLW repository only, TRU: TRU waste repository only, HLW/TRU: applicable to both repositories* Since the salinity in the groundwater is high, corrosion of the steel rebars in the structural concrete progresses during the operation period, and it is considered that the hydraulic conductivity
increases due to cracks associated with such corrosion; hence, even in the base cases, the hydraulic conductivity is set to be high from the beginning of the assessment.

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(3) Setting the analysis cases for low probability perturbation scenarios
Table 6.3-11 outlines the low probability perturbation scenarios and the calculational
approach to dose evaluation in the corresponding analysis cases. This concerns new
volcanism and fault activity.
For the fault perturbation scenarios (see Section 6.3.2 (3) (iii)), the area of the repository
significantly affected is inherently limited in the fault consolidation scenario (Supporting
Reports 3-35). However, the uncertainty associated with the fault extension scenario is large
and it is difficult to reliably establish the area of the repository impacted; thus, in both the
HLW and TRU waste analysis cases, the entire area of the repository is conservatively
assumed to be affected. For dose assessment, only the analysis cases for fault extension
scenarios, in which more waste is affected, are considered.

6-83

Table 6.3-11 Analysis cases for low probability perturbation scenarios
Scenario/analysis
case

Overview

Treatment in analysis cases

New volcano

Fault
consolidation

Consider the possibility
that a new volcano willdevelop and directly
affect the repository

Consider the possibility
relatively small faults
consolidate and impact
the repositoryThis scenario and the fault extension scenario are common in
the sense that the fault extends and affects the repository, andthe evaluation conditions are set so that the analysis casecorresponding to the fault extension scenario includes theanalysis case corresponding to this scenario.

Conservatively define the area affected by the extended fault.
The fault and associated process zone gradually increase as
the fault grows, so it is unlikely that these areas will develop
with one event [66]. However, conservatively assume
immediate formation in a single growth event.
Safety functions of EBS and host rock in the HLW repository
are handled:o The overpack, buffer material and surrounding hostrock located on the fault plane and in the area of thefault zone are damaged by the fault extension and lose
their safety functions. It is assumed that the dissolution
rate of the vitrified waste in the fault zone increases
with the increase in groundwater flow rate and thedissolution rate of the vitrified waste in the fault planeincreases with the increase in groundwater flow rate.
o For engineered barriers located in the process zone,
there is little influence on the safety function. The
safety function of the surrounding host rock is nottaken into account because it is difficult to quantify theincrease in its hydraulic conductivity.
Safety functions of EBS and host rock in the TRU repository
are handled:o For disposal tunnels directly struck by an extensional
fault, RNs from all waste packages will leach instantly
into groundwater and migrate directly to theextensional fault, except for Gr.2.
o In the case of Gr.2, RNs contained in metal parts willleach out and migrate to the fault due following
congruent corrosion.
For the waste affected, all released RNs are transported to the
GBI via the fault.
This fault is treated as a porous medium in which sorption of
RNs is considered.
Set GBI and biosphere as for the base case scenario.



Fault extension

A new volcano occurs and magma directly intrudes into therepository.
RNs in the area directly hit by the volcanic conduit reach thesurface of the Earth with the ejecta.
Assume the ejecta containing radioactive material is depositedon the surface and mixes with existing soil: calculate the doseto local inhabitants.

Consider the possibility
that a fault grows to
impact the repository






6-84

6.4 Radionuclide migration analysis and dose assessment
In this section, the RN migration analysis models and associated datasets are introduced
for each of the scenario analysis cases, plus dose conversion factors to allow releases to thebiosphere to be compared with established targets.
As noted in Section 6.3, RN migration models are applicable to several analysis cases,
while dose conversion factors are generally applicable. Datasets, however, tend to differ
according to the scenario involved. Section 6.4.1 describes the RN migration models, datasetsand dose conversion factors used for the base case scenarios, while differences in the datasets
used for the variant scenarios are summarised in Section 6.4.2. Finally, Section 6.4.3
describes the evaluation of the low probability perturbation scenarios.
6.4.1 RN migration analyses and dose evaluations of base scenario(1) RN migration analyses for different spatial scales
In the H12 report, the focus of the safety assessment was identification of generic researchand development needs and thus emphasis was on the performance of the HLW EBS andbehaviour of the near field, which is relatively easy to study and characterise. This approachallowed demonstration of the fundamental feasibility of geological disposal of HLW in Japan,
without specifying sites. The models used in the H12 report conservatively simplify thegeometry of the EBS, the heterogeneity of RN migration pathways and routes through thegeosphere, and the impacts of different host rocks. In addition, the barrier performance ofoverlying rocks between the host rock and the surface was not evaluated, with considerationof only a short-circuit migration route via a highly permeable fracture zone. Such approaches
and analytical models were also used in the TRU-2 report.
As noted in Section 6.1.1, the intention here is to more realistically reflect thecharacteristics of the geological environment at sites represented by the SDMs, together with
tailored repositories, as shown in Section 6.1.3. The approach to RN migration analysis usedin this report for different spatial scales will now be described.
As indicated in Figure 6.4-1, at the near field scale, a 3D hydrogeological model is usedthat considers the design specifications of the EBS and heterogeneous properties ofsurrounding host rock. Here, since a diverse range of RNs are to be analysed, the 3D model isused to efficiently determine the fundamental characteristics of movement of groundwater,
providing the boundary conditions to allow use of a simplified RN migration model (seeSection 6.4.1 (2) (iv) below).
To model RN migration at the panel scale, both the EBS components and the panel layout
in terms of disposal and access tunnels, together with the characteristics of immediately
surrounding host rock, are taken into account. For complex RN migration analysis, the 3D
near field model is established at the downstream end of disposal panels and assumed also torepresent panels further upstream with longer flow paths. This extremely simple, conservativemodel is applied to calculate the release rate of RNs from each disposal panel into thesurrounding host rock.

6-85

Figure 6.4-1 Concept of RN migration analysis illustrating how calculated RN release rates at
one scale are passed on to the next spatial scale (hydro boundary conditions are defined fromlarge to small scales)

On the repository scale, RNs released from each panel migrate through the host rock as
determined by the local flow of the groundwater to an analysis boundary (usually the nearest
downstream fault as defined by the SDM). Furthermore, it is assumed that this boundary
represents a short circuit to the GBI via a major fault. It is recognised that this is a simplified
conservative assumption that would be revised when specific sites are evaluated. Again, as in
the near field scale, groundwater flow using a 3D model is analysed, which conditions a
simplified RN migration analysis (See Section 6.4.1 (4) (i)).
At a regional scale, extending from the repository scale to the GBI, it is necessary tocapture the characteristics and temporal changes of the topography, geological structure,surface environment, etc., and reflect these in the RN migration and biosphere evaluations. As
already mentioned, this report follows the concepts used in the H12 and TRU-2 safetyassessments, assuming that large-scale faults act as short-circuits to the biosphere. The
biosphere model considers the GBI as the connection point between such large-scale faults
and the surface water flow system. Simplistically, the radiation dose is calculated bymultiplying RN release fluxes by dose conversion factors calculated for stylised biospheres;for the current assessment these are assumed not to change with time.
In the RN migration analysis described above, it is necessary to set the THMC boundary
conditions for each spatial scale. For this, the following approach (see Figure 6.4-1) is
adopted.
Based on the regional scale SDM, the THMC conditions at repository depth is set. These
conditions are assumed to be reasonably uniform and stable on the repository scale.
In the near field and the panel scales, there will be perturbations of such conditions due toimpacts of repository excavation and operation, heat generation from waste, EBS materials
introduced, etc. (see Table 6.1-2). For TRU waste, the loss of containment is assumed to beearly enough that thermal effects need to be considered when selecting migration data, incontrast to the HLW overpack containment which is lost only after the entire repository is at
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rock ambient temperature. The data set used for RN migration analysis also takes into
consideration changes of groundwater composition and material properties over time.
Hydrogeological conditions at the repository scale boundaries are set based on the spatial
distribution of hydraulic head obtained by regional scale groundwater flow analysis. Then, by
analysing groundwater flow at the repository scale, boundary conditions are set for analysis of
RN migration at panel and near field scales. This approach, which is commonly applied insuch assessments in other countries, aims for hydraulic consistency between scales, withpractical simplifications applied according to the scale of model (see Section 3.3.3).
Based on this approach, and the repository designs for each SDM presented in Chapter 4,
the spatial domains for RN migration analysis considered for dose assessment are defined.
Figure 6.4-2 shows an example based on a HLW repository layout option for plutonic rocks,
assuming co-disposal of TRU waste.

Figure 6.4-2 Setting spatial scales for RN migration analysis (HLW repository in plutonic rock
(H12V) with co-located TRU repository).

It should be noted that in this figure the TRU repository is located downstream (for thecurrent hydrogeological state) of HLW panels (thus avoiding any negative effects from the
TRU repository on the barrier performance of the HLW repository), but this is potentially not
optimised in relation to location of faults. In this Figure, the area surrounded by a dotted green
line defines the area of EBS/host rock covered at the panel scale, which includes within it, thenear field scale model within dashed brown lines, as explained in Section 6.4.1 (2) (ii) below.
6-87

As described in Section 3.3.3, analysis of groundwater flow at the repository scale (in theplane of the host rock covering an area of 5 km x 5 km) considers impacts of the distributionof major water-carrying faults on groundwater flow velocities and transport times, based onboundary conditions from the regional scale model. This extends beyond the area analysed on
the panel scale, and is based on the SDMs, including an additional 500 m perimeter to allow
transport routes in peripheral host rock to be captured in cases where the disposal footprintlies near a repository-scale model boundary. RN migration analysis at the repository scale isdescribed in Section 6.4.1 (4).
As mentioned in Section 4.5.4, a planar area extending beyond the repository scale was set
for each host rock (see Section 3.3.3 (3) (ii) (c): 10 km x 10 km for plutonic rocks and PreNeogene sediments; 7.5 km x 7.5 km in the case of Neogene sediments). This is used todefine relative groundwater migration times from different disposal areas (see Supporting
Report 4-49). The regional distribution of hydraulic head obtained from this groundwaterflow analysis was used for migration modelling at repository, panel and near field scales andparticle tracking analysis at repository and near field scales (6.4.1 (2) (iv) (b) and (4) (i)).
Using the analysis system described above, a more realistic RN migration assessment canbe conducted, considering the EBS design and layout while maintaining practicality. An
outline of the analysis codes used is given in Supporting Report 6-12. However, it should also
be noted that the current model scales and their representations will be revisited and made lessstylised when assessing real sites in the future.
(2) Near field scale model and dataset(i) Basic model concept
3D groundwater flow and particle tracking analyses were conducted for the purpose ofevaluating RN transfer paths, transport distances, flow rates, etc. in the near field scale region,
based on the features of the EBS and surrounding host rock.
Chapter 4 provides detailed repository system designs in terms of 3D shapes and layouts,
together with material specifications of tunnels and engineered barriers. These are specificallyestablished on the basis of key SDM characteristics, such as spatial distributions of fractures,rock mechanics and hydraulic gradient, which influence RN migration behaviour (see Section
6.1.2 (1)).
For near field host rock, common features of the different SDMs are networks of distinct
water-carrying features which can be simulated by discrete fracture network (DFN) models(see Section 3.3.3). Such models were constructed for all host rocks although, for Neogene
sediments, an integrated model accounting for flow in both fracture networks and the porous
matrix was developed.
The host rock modelling concept described above was implemented in the versatile
calculation code “Partridge” [23] (see Section 6.4.1 (2) (iii) below). Partridge can simulatecharacteristics of both the EBS and near field host rocks in some detail, allowing particle
track analysis by a 3D random walk method to quantify the mass transfer processes occurring.
However, this requires a long calculation time and, therefore, a simplified RN migration
model was developed, which allows more practical implementation of a large number ofcalculations targeting a wide variety of RNs for various scenarios. Figure 6.4-3 shows theprocedure for near field modelling of RN migration analysis.

6-88

Figure 6.4-3 Procedure for analysis of RN migration in the near field(ii) Analysis approaches and components of the models
For the HLW repository layout option for H12V in plutonic rocks (or Pre-Neogene
sediments) and co-disposal of TRU shown in Figure 6.4-2, the structures to be assessed at anear field scale are shown in Figure 6.4-4. Note that the access tunnels are not included in thenear field model, since the flow direction is perpendicular to the disposal tunnels/vaults andparallel to the access tunnels.

6-89Figure 6.4-4 Repositories in plutonic rock or Pre-Neogene sediments: illustration ofcomponents for three-dimensional groundwater flow/ particle track analysis at the near field
scale (a): HLW–H12V, (b): TRU waste Gr.2, (c): TRU waste Gr.3.

For the H12V option (Figure 6.4-4 (a)), the model to calculate the release rate of RNs
includes 10 disposal holes, along a tunnel length of 44.4 m, and a volume of rock extending
100 m downstream. Here, the 100 m length of downstream host rock was set from theviewpoint of practicality, based on the capabilities of the Partridge code used.
The shape and size of the tunnel cross-sections and component materials are specified in
Section 4.5.2. In the assessment model, the EDZ is modelled as a mixed tank defined by its
volume and porosity. The EDZ for the H12V option components is set on the basis of the
excavation methods assumed (see Supporting Report 4-44, 4-67), resulting in thecharacteristics and thicknesses used for the models (see Supporting Report 6-14).
For the HLW PEM option, the size of the analysed area is the same, but the length ofdisposal tunnel considered is 45.6 m, which contains 13 PEMs. Shapes and dimensions of the
6-90disposal tunnel are based on design specifications given in Section 4.5.2. EDZ dimensions
and characteristics are defined in a similar manner to the H12V case.
Figures 6.4-4 (b), (c) illustrate the analysed systems for TRU waste. As described in
Section 6.4.1 (2) (iv), for RN migration analysis, a simplified one-dimensional model isconstructed based on the 3D particle tracking results as for the HLW case. For 3D particle
tracking, a model reflecting the shape and characteristics of the repository components isused; however, for TRU waste, a range of different disposal vault cross-sections are definedfor each host rock (see Supporting Report 4-36). Conditioned by the 3D particle tracking
analysis, RN migration is approximated by a one-dimensional, multi-channel model (Section
6.4.1 (2) (iv). In order to capture key differences, representative disposal vaults for cases withand without buffer are modelled, selecting for these Grs.2 and 3, respectively, and defining
associated EDZ characteristics as previously described for HLW. The analysed area is thesame as the HLW case, with the vault length including three emplacement pits (Supporting
Report 4-42).
For repositories in Neogene sediments (Figure 6.4-5), the representation is somewhat
different, reflecting the fact that disposal tunnels/vaults are excavated parallel to thegroundwater flow direction (see Section 4.5.4). RN migration paths intercept connecting
tunnels at the downstream end of the disposal panel and thus these need to be handled
explicitly. For HLW, five disposal tunnels were modelled, including the area of downstream
host rocks extending 100 m from the last disposal hole. For the TRU waste disposal vaults, asimilar procedure is adopted and the access tunnels downstream are included in the model.
Again, representative disposal vaults were modelled for cases with and without buffer,selecting for these, Grs.2 and 3, respectively, and defining associated EDZ characteristics aspreviously described for HLW. The analysed area is the same as the HLW case, with the vault
length including three emplacement pits (Supporting Report 4-42).
EDZ characteristics are defined as for plutonic rock, although it is understood thedevelopment and evolution of an EDZ in Neogene sediments may differ. Such differences
will need to be explored further once specific sites are assessed. For Pre-Neogene sediments,
the system components are the same as for plutonic rocks (Section 6.1.2, Figure 6.4-4) andhence the models and system representations within them are also the same as describedabove.
In all of the repositories for the host rock considered, components other than engineered
barriers, such as backfill, shotcrete and other cementitious materials, were given
conservatively high hydraulic conductivity values so that their influence on RN migration in
the near field could be clearly taken into account, reflecting their alteration with time (seeTable 6.3-3).
The 3D groundwater flow analyses are described in more detail in Supporting Report 6-14.

6-91

Figure 6.4-5 Repositories in Neogene sediments: illustration of components for the threedimensional groundwater flow/particle track analysis at the near field scale, (a): HLW–H12V,(b): TRU waste Gr.2, (c): TRU waste Gr.3

6-92

(iii) Application of the 3D RN migration code Partridge
Partridge is a code developed for particle tracking analysis, reflecting features such as
repository design and characteristics of a fractured host rock [23], which is applied to nearfield concepts described in Section 6.1.1. Its features are as follows:Rocks containing major heterogeneities, in particular highly permeable structures, canbe deterministically expressed, although this option was not used. Instead, the large
number of faults and fractures are stochastically represented, while engineered
structures expressed as homogeneous continuums, to allow the target area to be
represented as a 3D hydraulic conductivity tensor field.
For the constructed tensor field, by imposing representative boundary conditions,saturated groundwater flow analysis is used to calculate the water head distribution
and Darcy velocity vector field.
For the calculated 3D groundwater velocity vector field and using a random walkmethod, advection and dispersion in fractures, matrix diffusion and solute dispersion
and sorption could be simulated. However, in this application, sorption was notconsidered.

Here, information on fracture networks from SDMs (Section 3.3.3) is used to stochasticallygenerate hydraulic conductivity tensor field realisations (using statistical data on distributionsof fracture orientations, length distributions, 3D fracture densities, etc.). The hydraulicgradient required for groundwater flow analysis at the near field scale is derived from thehead distribution from panel scale analysis, which in turn is based on results of thegroundwater flow analysis on the repository scale. This repository scale analysis is the sameas that carried out in the design study for the disposal panel layout (see Section 4.5.4), basedon rock-specific SDMs to establish an equivalent continuum model that deterministicallyaccounts for major fault zones.
Below, the transport processes specifically considered in Partridge as applied in this report
are described.
(a) EBSFor particle tracking, the source term is defined as a unit mass represented by a large
number of particles. These particles are considered to diffuse through the EBS, which isrepresented as a porous medium.
(b) Host rock
As noted in Section 3.3.3, advection in faults and fractures dominates groundwater flow inplutonic and Pre-Neogene sediments, allowing a DFN model to be applied. Since advection in
Neogene sediments can also occur in the rock matrix, a model expressing features of both
fracture and porous matrix flow is applied.
The DFN model particle tracking analysis includes solute transport by advection anddispersion in the fractures, with diffusion into the rock matrix. For Neogene sediments,
however, the DFN model has to be extended to include advection within the rock matrix,
treated as a porous medium, as shown in Figure 6.4-6.

6-93

Figure 6.4-6 Conceptual illustration of migration model for fractures and matrix of Neogene
sediments (Example of HLW (H12V))(iv) Development of the simplified RN migration model
Based on the modelling concepts described in 6.4.1 (2) (i) and (ii), the process ofconstructing a simplified RN migration model for the near field scale using Partridge isillustrated in Figure 6.4-7. The model is similar to the one used for the H12 study, apart fromthe way channels are identified by particle tracks in the 3D DFN model. As noted earlier, thismodelling approach will be further developed in coming assessments.

Figure 6.4-7 Construction of a simplified RN migration model for the near field

After generating a DFN model from each SDM, Partridge is used to carry out groundwaterflow analysis and particle tracking in the calculated hydraulic field. The groundwater flow
paths in the host rock are then simulated by a set of one-dimensional parallel plate fractures ina simplified model (multi-channel model) to allow simulation by the analysis code Goldsim ©
[67]. The time-dependent migration of particles released from the entire near field is obtainedby 3D particle tracking analysis. This release curve is simulated by a set of one-dimensional
channels of different transmissivity (Ti) and the weighting value (Wi). These parameters are
selected such that the weighted sum of the breakthrough of a unit release for each channel
sufficiently well match the calculated release curve, see Figure 6.4-12. The Ti of each channelof the multi-channel model is based on the distribution of the hydraulic conductivity with
6-94

respect to host rock fractures, where the upper and lower limits given to the channel are used
as optimisation variables. Solute release from the EBS is captured by a mixing tank model ofthe EDZ, with weighting of distribution to each channel given by Wi values (Section 6.4.1 (2)(iv) (e)).
In H12 [3], the Wi channel weighting was defined on the basis of hydraulic conductivity
distributions of fractures in the host rock (log-normal distribution), which is anoversimplification in terms of determining the migration rate of RNs. To more realistically
represent the solute transport characteristics of the DFN in this report, the Wi was derived
instead from the 3D Partridge simulations.
As mentioned in Section 6.4.1 (2) (iii) (b), in addition to advective flow in DFNs forplutonic or Pre-Neogene sediments, advection-dispersion processes in the rock matrix (Figure
6.4-5) have to be considered for Neogene sediments. For this case the advection-dispersion
processes in the rock matrix option were used in the 3D Partridge simulations.
The simplified RN migration analysis model for the host rock and EBS at a near field scale,depending on the host rock involved, is described further below.
(a) The DFN model
Since the spatial distribution of fractures is statistically treated, the computed distributionof fractures in each realisation does not produce the same flow field. As an example, Figure
6.4-8 shows the finite elements of the groundwater flow model generated by Partridge forplutonic rock (see Section 3.3.3), with the two different realisations showing the assigned
hydraulic conductivities.

(a)

(b)

Figure 6.4-8 Examples of hydraulic conductivity distributions of finite elements for two
different realisations of groundwater flow models

Comparing realisations (a) and (b), the impact of large fractures with high connectivity andhydraulic conductivity lying parallel to the main groundwater flow direction is highlighted in
the latter. It can be expected that such features will have a large influence on RN migrationand hence, when conducting the safety assessment, it is necessary to pay attention to how todetermine representative groundwater flow and RN transfer behaviour, taking into
consideration the statistical range involved.
6-95One way to deal with statistical variability is to create as many realisations as possible andconduct groundwater flow and particle track analysis for each of them. In general, however, ittakes a long computation time for particle tracking analysis in such a 3D DFN model, so thepractical number of such analyses is limited.
Therefore, in this report, groundwater flow analysis, which has shorter calculation times
than particle tracking analysis, is performed for many realisations. Based on groundwater
travel time and migration path length to the downstream boundary, decision on the DFN
models to be used for particle tracking analysis to capture the diversity of the generated fields
is made, using the obtained hydraulic conductivity as an indicator. Specifically, 100 DFN
realisations were created for groundwater flow analysis and the average hydraulicconductivity of the modelled rock volume with respect to the main flow direction ofgroundwater was evaluated. Then, these 100 realisations were broken down into 5 groups,
each covering the results of 10 realisations, in order to select the realisations used for particle
tracking analysis (see Figure 6.4-9). For each group, 2 realisations were sampled.

Figure 6.4-9 Stratified sampling based on 100 realisations

These results are multiplied by a weighting for each group, obtained from the frequency
distribution, to obtain the average value. This allows particle tracking analysis to generate
solute transport characteristics with a relatively small number of realisations. Details of thisapproach and a discussion of its validity are given in Supporting Report 6-14.
Even for actual sites, initial fracture information is obtained by relatively limited borehole
surveys and hence fracture networks of the entire repository will be statistically generated andevaluated. Thus, taking into consideration practical computing restrictions, it is necessary to
determine what would be a sufficient number of realisations to properly evaluate a site in a
stepwise manner.
Even if the number of realisations as described above is reduced, for each realisation, using
the functions of the Partridge code (see 6.4.1 (2) (iii)) for both the EBS and the 3D analysis oftransport processes in the host rock still requires a great deal of computation time. For thisreason, according to the method described in (b) to (e) below, the calculation load for eachrealisation was reduced.
(b) Particle tracking analysis using Partridge
First of all, for the analysed system of the HLW EBS (H12V) mentioned in Section 6.4.1(2) (ii) above, a 3D groundwater flow analysis was carried out assuming a homogeneous
equivalent porous medium model. The information on the pressure head at each node of thefinite elements obtained here was taken over to the DFN model and the actual flow velocity in
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the fracture and rock matrix was calculated. Then, based on the diffusion coefficients of the
EBS and host rock, non-sorbing particle tracking analysis in three dimensions, for an instantrelease of “tracer” from waste and without considering decay was performed. Hereparameters used for 3D analysis of the groundwater flow are, as discussed in Section 6.4.1 (1),for repository layout studies on a 10 km x 10 km scale (details are summarised in Supporting
Report 6-14).
In the particle tracking analysis, the tracer breakthrough to the downstream side of theregion depends on the shape/transport characteristics of the EBS and the heterogeneous
fracture network of the surrounding host rock, and on the location of the downstream
boundary. The analysis results in a “basic solution” on the transport rate of solute, without
taking into consideration elution behaviour, solubility, sorption and decay. These processes,
which would depend on the source term and different migration properties of different RNs,
could then be modelled for the set of 1D channels obtained from the 3D particle tracking
analysis.
For H12V, 2 realisations were randomly extracted from each hydraulic conductivity group(S1 to S5) shown in Figure 6.4-9, to give a total of 10 simulations of tracer migration to themodel downstream exit. A weighted average, also shown in the figure (solid line) was
obtained by applying the weights of the permeability group representing each individual
breakthrough curve. The tracer release curves from the near field, normalised by the initial
number of particles released at time zero, is shown in Figure 6.4-10.

Figure 6.4-10 Calculated average tracer migration (solid line) for HLW (H12V) based on particle
tracking analysis for 10 realisations (see Supporting Report 6-14 for details on eachrealisation)

The figure shows dispersion effects due to advection in the heterogeneous fissure network,with the results used to construct a one-dimensional multi-channel model for the host rock, as
described in (d) below. A similar procedure was applied for the other rock types and for the
TRU repository.

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(c) Simplified RN migration model for the EBS
To model the EBS for HLW and TRU waste, simplifications are introduced to the basescenario to capture specific designs without considering the associated host rock.
(A) HLW
Figure 6.4-11 shows the simplified one-dimensional coupled box model of EBS RN
release and migration for HLW. In the migration processes shown in the figure, radioactive
decay/ingrowth is taken into consideration.

Figure 6.4-11 Concept of simplified RN release and transport model for the HLW EBS

After the overpack loses the function of containment, the vitrified HLW contacts
groundwater and dissolves at a defined rate. RNs are uniformly distributed within the glass
and are considered to dissolve congruently with the glass matrix. Although, initially, the glassdissolution rate is related to the local dissolved silica concentration, this reaches saturation
after which the dissolution rate is constant. RN concentrations in pore fluid within the
overpack may be limited by elemental solubility (assuming thermodynamic equilibrium); if
this is exceeded, precipitation (and, possibly, co-precipitation) occurs. Since colloids arefiltered out by the buffer, as long as its safety functions are upheld, RN partitioning between
water and colloids need not be modelled. For RN migration analysis, solubility limits are notspecified for elements that are expected to be highly soluble (e.g., Cl, Cs, I). For elementswith multiple isotopes, solubility concentration limits are shared between them, according totheir relative abundance. Here, unlike H12, the impact of stable isotopes was conservatively
ignored due to uncertainty in the concentration of these, although this may be revisited in the
future. Radioactive decay and ingrowth are also modelled in the mixing tank.
Dissolved RNs in the overpack mixing tank are assumed to diffuse through the buffermaterial, with a concentration gradient set by the concentration in the source and EDZ mixing
tanks, while being retarded by sorption and attenuated by radioactive decay. Dissolution andprecipitation are not modelled in the buffer, since RN concentrations are higher at the source
term end. It is conservatively assumed that the RNs released from the buffer mix
instantaneously with groundwater within the EDZ, in turn forming the source term fortransport through the host rock. The EDZ mixing tank releases RNs to multiple onedimensional channels representing the host rock, based on the defined Wi (see (e) below). The
migration path was simplified, while maintaining conservatism, by not considering thebackfilled part of the emplacement tunnel and the access tunnels. However, for the
hypothetical case that the function of the backfill was lost, this simplification means that the

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current assessment would not allow the analysis to distinguish between different EBS designsof the tunnel backfill.
For the PEM, no safety function is assigned to the steel handling shell and thus the releasemodel from the EBS is the same as for H12V. This implies that the difference in flowgeometry, where the EDZ for the PEM is around the deposition tunnel and where there isdiffusion from buffer through the backfill, is ignored. Furthermore, it is assumed that thecorrosion of the handling shell does not have any impact on the buffer. While these
simplifications can be conservatively justified, they also mean that the current assessment
does not allow the analysis to distinguish between H12V and the PEM options, which will bea goal for future model improvements.
(B) TRU waste
Figure 6.4-12 shows the concept of a simplified one-dimensional coupled box model of
EBS RN release and migration for TRU waste. As for HLW, in the migration processes
shown in the figure, radioactive decay/ingrowth is taken into consideration. For waste
package A, the delay in the release due to the saturation time and containment time of thewaste containers are conservatively ignored, whereas the containment time for waste package
B is included in the analysis.

Figure 6.4-12 Concept of simplified RN release and transport models for TRU EBS

RNs present in the metal waste contained in Gr.2 are released congruently with metal
corrosion: the other Gr.2 materials and waste in other groups are assumed to release contained
RNs instantly into surrounding porewater.
Released RNs are assumed to be sorbed on the EBS waste packaging and infill materials,
which are treated together as a mixing tank. However, as discussed in Section 6.3, for Grs.2
and 4H, which are heat-emitting, it was conservatively assumed that cement hydrate will bethermally altered and sorption onto it is not considered. Furthermore, as also discussed in
Section 6.3, Gr.3 includes a large amount of nitrate which also is assumed to affect sorption.
In all cases, porewater RN concentrations are assumed to be constrained by elemental
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solubilities – treated in exactly the same way as for HLW, while considering the impact of theambient hyperalkaline conditions.
In the case of Grs.1, 2 and 4H, RN transport within the EBS is predominantly by diffusion,
while additional sorption and radioactive decay occurs within the buffer. For Grs.3 and 4L,
RNs from the EBS mixing tank pass directly to the EDZ by advective flow. The effect of
decay/ingrowth during transport processes is taken into account and the treatment of releases
from the EDZ mixing tank into the multiple channels of the host rock is the same as for the
HLW case.
In modelling the RN migration from the EBS, in both HLW and TRU waste cases,
supporting engineered structures that are not assigned safety functions are not considered.
(d) RN migration model for a simplified host rock
As mentioned in (b) above, tracer migration rates obtained by 3D transport analysis areused to condition a model in which the host rock is represented by one-dimensional parallel
plates with different hydraulic conductivities. This greatly increases the efficiency of
calculation of the various analysis cases for safety assessment scenarios. The validity of such
simplification is discussed in a NUMO technical report [23].
Section 6.4.1 (2) (iii) (b) noted that, for plutonic rocks and Pre-Neogene sediments, amulti-channel model of advection-dispersion in one-dimensional parallel plate fractures cansimulate RN migration processes including decay chains and diffusion into and sorption onto
the rock matrix. The situation is generally similar for Neogene sediments but, in addition toadvection in fractures, advection-dispersion in the rock matrix, considered as a porous
medium, has also to be handled as discussed in section 6.4.1 (2) (iv).
(e) Approximation of Partridge results by the simplified model
A simplified near field model is constructed in which the one-dimensional radial migration
analysis model of the EBS mentioned in (c) and the multi-channel model of host rockdescribed in (d) are combined, with the EDZ handled by a mixing tank model. The hydraulicconductivity of each channel (Ti) is used to derive a normalised particle flux (φi(t)) from the
EDZ, which is then integrated using the Wi to derive a total release flux (φ3D(t)) as shown in
Figure 6.4-13 for the example of H12V in plutonic rock.

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Figure 6.4-13 Approximation using the one-dimensional multi-channel model to match the
combination of a set of channels (red line) with the average tracer migration rate (the solid line
in Figure 6.4-10, here represented by blue dots) obtained by 3D particle tracking analysis (H12V
plutonic rock case)

The figure shows how the calculated 3D release curve was simulated by the multi-channel
model, as explained in section 6.4.1 (2) (iv) and Figure 6.4-7 above. The length of eachchannel is 100 m. At first the fitting was made with 41 channels, but it was found that it waspossible to adequately simulate tracer migration using only 13 channels. Given the allocated
Wi of the RN flowing into each channel, RN migration processes described in (c) and (d)above were analysed using the code GoldSim ©.
Neogene sediments were simulated by combining the DFN model with a porous medium
model and assessing flow using particle tracking. In the multi-channel model, porous flow
was represented by one of the channels, but it was concluded that this did not contributesignificantly as a RN migration pathway (Supporting Report 6-13).
(v) Datasets for near field scale migration
Based on the above simplified RN release and migration model, required parameters wereset for the associated analysis cases as described below.
(a) THMC data
(A) Thermal data
As discussed in Section 6.3.2 (3) (i), for both HLW and TRU waste, the repository takes
time to re-saturate, after which contact between waste and water is restricted due tocontainment by the overpacks or waste packages for a certain period of time: only thereafter
does elution of RNs commence. Based on the results of thermal analysis for buffer in the
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different repository designs (see Supporting Reports 4-39, 4-40, 4-41), the entire near field
has returned to the original rock ambient temperature when RNs are released – so thistemperature is used to set relevant RN migration parameters, apart from conservatively
discarding sorption in the potentially degraded cement hydrate for TRU Grs.2 and 4H, asdescribed previously
The rock ambient temperature was set at 45 °C for plutonic rocks and Pre-Neogene
sediments (disposal depth 1,000 m) and 30 °C for Neogene sediments (disposal depth 500 m)(see Section 3.3.3 (6) (ii)).
(B) Mechanical properties
After the repository is closed, the density of the buffer may change, due to overpack
corrosion, structural framework settlement, etc., but it is designed so that the required density
is kept for a long time after closure, maintaining the swelling pressure developed after resaturation. Thus, the influence of any change of mechanical properties when setting RN
migration parameters was not considered.
(C) Hydraulic properties
For all repositories and host rocks considered, backfilling material and cementitious
support structures that are not part of the EBS will alter with time after closure and hence areconservatively assigned high hydraulic conductivity values.

- Hydraulic gradient
Basically, it is assumed that the initial hydraulic gradient of the SDM recovers quickly
after any perturbations due to construction and operation and is then maintained constant for
the entire calculation period. Long-term changes of the geological environment are notconsidered, as explained in Section 6.3.
From the results of the groundwater flow analysis (Supporting Report 4-49) at a repository
scale, the gradient is set as 0.05 in the case of plutonic rocks and Pre-Neogene sediments; for
Neogene sediments it is set as 0.06 (Supporting Report 6-14).
- Hydraulic conductivity of buffer
In the setting of the buffer hydraulic conductivity, potential impacts of long-term
degeneration as a result of interaction with any concrete present were taken into consideration.
Degradation due to future changes in groundwater chemistry was not considered.
For H12V, no cementitious material is assumed to contact the buffer for any host rock,therefore there is no change in hydraulic conductivity due to alteration and the design
hydraulic conductivity specification (1.0 x 10-12 m/s) is assumed for the entire assessment
period (see Supporting Report 4-15). For the PEM variant, a concrete pedestal is assumed tointeract with the buffer (although backfilling material is arranged around the handling shell
containing it), so it is assumed that the hydraulic conductivity increases by two orders ofmagnitude to 1.0 x 10-10 m/s for the entire assessment period. However, this value is still
sufficiently low for migration to be diffusion dominated in the buffer.

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For TRU waste, a range of concrete structures are laid in contact with any buffer present,so alteration was investigated taking this into consideration. According to the result of thesafety evaluations so far, non-sorbing I-129 dominates the dose from TRU waste, but isreleased from the EBS over a relatively short time period [3][14]. From this point of view, theperiod during which the buffer material function of suppressing RN migration is important isshorter than the HLW case, while the degree of change in hydraulic conductivity due toalteration would be relatively small. Thus, for host rocks in which the amount of cementitious
material in contact with the buffer is relatively small (plutonic rocks and Pre-Neogene
sediments), the design hydraulic conductivity specification of 1.0 x 10-12 m/s is assumed. For
the Neogene sediments, although the quantity of cementitious materials in contact with thebuffer is larger, the period of alteration is shorter than that for HLW and thus the hydraulicconductivity was assumed to increase by one order of magnitude to 1.0 x 10-11 m/s. However,again, these values are still sufficiently low for migration to be diffusion dominated in thebuffer.
- Hydraulic conductivity of cementitious materials
TRU wastes, other than Gr.2, are conditioned with cementitious materials. In addition,
packaging and infilling of all waste groups involves cementitious material and the structural
frameworks are predominantly constructed from concrete. Buffer layers surround the disposal
tunnels of Grs.1, 2, 4H and inside these solute transport through cementitious materials is
dominated by diffusion. This will be the case even though cracking of infill is considered
likely to occur and integrity of structural frameworks cannot be assured (see Supporting
Report 4-35). As it has little impact on performance, the hydraulic conductivity of the entire
EBS region within the bentonite is conservatively set equivalent to that of sand (1.0 x 10-5m/s) [68].
For Grs.3 and 4L, without buffer, advective flow occurs, which is dependent on thehydraulic conductivity of the EBS. As above, it is impossible to preclude cracking after
construction. Based on core data from old dams [69], hydraulic conductivity of the structural
framework is set at 1.0 x 10-7 m/s for 200 y after closure, and 1.0 x 10-5 m/s, equivalent to thatof sand, thereafter; the same as for waste package and vault infill (Supporting Report 6-16).
- EDZ hydraulic conditions
In the mixing cell model of the EDZ, to calculate the RN concentrations therein, thegroundwater flow rate through it and its thickness is required. Porosity is conservatively set to
1. For the HLW repository, based on based on expert knowledge, the H12V disposal tunnel isassumed to be excavated by the drill and blast method, producing an EDZ with a of thicknessof 1.0 m and a hydraulic conductivity of 100 times that of the host rock. For the PEM, an
EDZ of thickness of 0.5 m resulting from excavation by TBM is assumed, with hydraulicconductivity again 100 times that of the host rock (see Supporting Reports 4-44, 4-67). The
TRU waste disposal vault EDZ was set similar to that of the disposal tunnel for HLW.
Groundwater flow rate through the EDZ used for RN migration analysis was calculated
from the hydraulic conductivity of the EDZ and permeabilities of undisturbed rock and theEBS (taken as that of buffer, except for TRU waste Grs.3 and 4L where the value for backfill
was used). The hydraulic conductivity and hydraulic gradient of the rock were determinedusing the theoretical solution of groundwater flow presented in the TRU-2 report [14], as
described in more detail in Supporting Report 6-13.
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- Hydraulic conductivity of host rock
The hydraulic characteristics of the one-dimensional multi-channel model were derived by
the methods already described in Section 6.4.1 (2) (iv). The hydraulic conductivity of the nearfield rock could thus be represented by the modelled set of fracture channels, with theirrespective transmissivity values (Ti.) The width of each channel was set in a similar way tothat in the H12 report, and was calculated from Ti using empirical rules. For Neogene
sediments, the channel representing the porous medium between fractures was given thehydraulic conductivity of the matrix provided in Section 3.3.3 (4) (ii) (b).
(D) Chemistry- Groundwater chemistry in host rocks
The groundwater chemistries of host rocks, as specified in Chapter 3 (Table 3.3-16), areassumed to be constant over the entire assessment period. Long-term changes of thegeological environment are not considered in this assessment, as explained in Section 6.3.
- Water chemistry in the EBS
For HLW, it is necessary to model the chemistry of the buffer porewater in order to derive
sorption distribution coefficients and effective diffusion coefficients in this material. In
addition, in order to determine the solubility of RNs within the overpack, it is necessary to set
the water chemistry therein. Solubilities of RNs are derived from thermodynamic calculations,for porewater chemistry in the HLW buffer and the TRU mortar infilling material. Although
the strict applicability of this approach to a microporous medium like compacted bentonite isdebatable [20], this is used here to assess ion exchange reactions of montmorillonite, acidbase reactions of surface hydroxyl groups, dissolution and precipitation reactions ofaccompanying minerals and reactions of iron corrosion products. Based on mass balance
arguments, the Na-bentonite will persist for a long time, assuring relatively constant waterchemistry in contact with the failed overpack, where it can be considered that elemental
solubilities are set (colloids may be released from corroding glass and migrate through thefractured overpack, but are completely immobile in compacted bentonite).
It should be noted that the backfill includes the same bentonite material as the buffer,
mixed with excavated rock, so no significant additional changes of groundwater composition
are assumed to occur (Section 4.5.3 (1)) and hence this is not considered in setting the watercomposition of the EBS.
The simplified TRU waste EBS models in Section 6.4.1 (2) (iv) (c) distinguish between
waste groups with (1, 2 and 4H) and without (3 and 4L) external buffer and these are assigned
different porewater chemistries. For the former case, TRU waste buffer porewater chemistry
was taken to be the same as in the HLW case. As explained in Supporting Report 6-15, thealteration of bentonite is negligible and the specification of the buffer for HLW and TRU arealmost the same, while the thickness of the TRU buffer is larger than that for HLW.
Porewater chemistry of the infill (“infill porewater”) inside this buffer was modelled byconsidering reaction of buffer porewater with this cementitious material, assuming
thermodynamic equilibrium with evolving solid phases. Initially porewater in cementitious
material is characterised as “Region I” with pH = 13 or more due to elution of sodium andpotassium hydroxides. Later, in “Region II”, equilibrium with portlandite buffers porewater
pH to about 12.5. After that, in Region III, a range of calcium silicate hydrate phases control
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pH, which drops below 12.5 [70]. Coupled reaction/solute transport modelling of
cement/buffer interaction suggests that portlandite in the infill remains over the modelled
timescale (Supporting Report 6-8). Thus, infill porewater chemistry was set for Regions I andII, using a method developed by JAEA [71]. It is noted that this model approach includes
many simplifications, since many key solids are only metastable and the assumption of amixing tank would not properly describe the much slower diffusive transport inside the EBS.
These simplifications need to be revisited in future safety cases.
For TRU waste without buffer, groundwater reacts directly with cementitious infill, which
has a relatively high hydraulic conductivity (Section 6.3.2 (3) (i)) and, together with concrete
structures, provides a large inventory of reactive material (Supporting Report 6-15). Again,
for this case, coupled reaction/solute transport modelling of cement/groundwater interaction
suggests that infill porewater remains in Regions I and II (Supporting Report 6-8) over themodelled timescale.
Geochemical calculation code PHREEQC ver.3 [72] and thermodynamic database JAEA
β-TDB v 1.07 [73] were used for the analysis of equilibrium reactions.
For TRU waste Gr.3 (without buffer), porewater chemistry in the EBS is also impacted bydissolution of the nitrate contained in the waste, which could yield high ionic strengths of up
to about 8 mol/l (see Supporting Report 6-8), since it is conservatively assumed that thenitrate contained in the bitumen is immediately dissolved, regardless of the speed ofdegradation of the bitumen. However, the thermodynamic models/databases used do notallow activity corrections required for such high ionic strength and hence this impact onporewater chemistry was not included directly, but its influence on RN migration parameters
is taken into account, as described in (b) below.
Appendix Tables 5 to 8 present the model EBS porewater chemistries, which are describedin more detail in Supporting Report 6-15.
(b) RN migration parameters(A) Parameters related to waste- RN release time and RN inventory
As noted previously, it is conservatively assumed that the HLW overpack will maintain its
containment function for at least 1 ky, even if the containment might actually be at least 10
times longer. However, as explained in Section 6.3.2 (3) (i), it is also very conservatively
assumed in the analysis that all 40,000 overpacks fail simultaneously 1 ky after repository
closure and RN release begins at this time. The radioactivity inventory of HLW at time of
emplacement is given in Table 6.1-3, so the impacts of decay between then and the failure
time of the overpack is calculated.
The TRU waste package A is not assigned a containment function (see Section 6.3.2 (3)(i)), since there is no upper lid on these waste packages. However, as discussed in Section
4.4.2 (2) (v) (a), it may still take some time before releases from the waste would start. In
order to address the uncertainty associated with estimating this time, it is, as in the TRU-2
report, conservatively assumed that release of RNs starts immediately after repository closure.
Waste package B is designed to contribute to the safety function “prevention of contact
between waste and groundwater” for a limited time. As described in Section 4.4.2 (2) (v) (b),even considering the possibility of local corrosion and the influence of radiolysis, theassumed corrosion depth is less than 10 mm in 300 y. Since the waste package container is 50
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mm thick, it is considered corrosive failure will not occur within this time. As further
discussed in Section 4.4.2 (2) (v) (b), the structural integrity of the waste package is also
considered due to the impact of the hydrostatic pressure after re-saturation and the reduced
wall thickness. Even so, structural integrity is considered to be maintained. From this, it isassumed likely to prevent contact between groundwater and waste for at least 300 y, and thisis set as the initiation time for releases in the assessment.
The radioactivity inventory of TRU waste at the time of emplacement is given in Table
6.1-4, which is the assumed start of elution of RNs in the case of waste package A. In the caseof the waste package B, these inventories have to be corrected for 300 y of decay (seeSupporting Report 2-3).
- RN dissolution rate
For HLW, RNs are released congruently with glass dissolution, which proceeds at a longterm rate (after silica saturation) as given in the H12 report. Here, corrections were made to
the glass dissolution rates obtained in long-term leaching tests to account for the temperatureof the repository and the surface area increase due to cracking during manufacture of the glass.
The uncertainties in the glass dissolution rate are briefly discussed in Supporting Report 6-9.
The period required for total dissolution of the glass is conservatively estimated to be about
70 ky, and this value is assumed to apply in this report.
For TRU waste, as in TRU-2, conditioning matrices for Grs.1, 3 and 4 are conservatively
assumed to be instantaneously soluble in EBS porewater (Section 6.3.3 (1)), with the RN
inventory being well mixed within this volume, disregarding the potential diffusive migration
within the degrading waste form. For Gr.2, RNs adhering to the surface of hulls and end
pieces are treated similarly, while those incorporated into metals are released congruently
with their corrosion. In line with TRU-2, corrosion lifetimes for the metals involved are
specified (zircaloy: 11.4 ky, stainless steel: 8.5 ky), allowing elution rates of RNs to be set
assuming that the material corrodes at a constant rate.
(B) Parameters related to the EBS- Solubility
Nuclide solubilities are considered as constraints for HLW and TRU waste Grs.1, 2, 4L
and 4H (Section 6.4.1 (2) (iv) (c)).
For HLW, thermodynamic equilibrium calculations, using the EBS porewater chemistry
specified in Section 6.4.1 (2) (v) (a) and data on minerals with the potential to precipitate,
forms the basis of selecting assumed solubility-limiting solid phases [4] for each RN. Next,
solubility is calculated using thermodynamic calculations assuming that porewater and thesolubility-limited solid phase are in equilibrium. In cases of uncertainty in the stability ofsolid phases, it was decided to select simple solids that are known to exist under relevant
geological environmental conditions, see Supporting Report 6-17.
For TRU waste Grs.1, 2, 4L and 4H, basically the same procedure is adopted; but, as
discussed in Section 6.4.1(2) (v) (a), EBS porewater chemistry evolves with time following
reaction between the buffer and the cementitious material. The solubility was calculated forthe different porewater chemistries involved and the highest value was selected for the
calculational database. The effect of iso-saccharine acid, originating from degradation oforganic matter in TRU Waste Gr.2, is included in the thermodynamic equilibrium calculation.
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For the calculation of the solubility, the calculation code PHREEQC Ver.3 [72] anddatabase JAEA-TDB 140331S0.Tdb [74] were used, with partial modifications of data for U
and Zr (see Supporting Report 6-17). Although the set value of the ambient rock temperature
is 30 °C (Neogene sediments) or 45 °C (plutonic rocks and Pre-Neogene sediments),
solubility is calculated at 25 °C, for which most thermodynamic data are specified. Although
the influence of small temperature changes on solubility were shown not to be significant in
the H12 report or for SR-Site [44], due to the uncertainty [75] involved, increased solubility
due to temperature was studied in variant scenarios.
For TRU waste Gr.3, thermodynamic equilibrium calculations cannot be applied as
porewater chemistry is not set (see 6.4.1 (2) (v) (a) 4) and hence solubility limits areconservatively not specified (see Supporting Report 6-17). Therefore, from the viewpoint of
ensuring traceability in safety evaluation, it is assumed that the entire RN inventory is
instantaneously dissolved in the EBS porewater.
Appendix Tables 9 to 11 summarise solubilities used for HLW and TRU waste RN
migration analysis, with more information on the solubility setting process in Supporting
Report 6-17.
- Effective diffusion coefficient
Effective diffusion coefficients of elements in buffer are selected based on a method
developed during joint research between NUMO and JAEA [76], as shown in Figure 6.4-14.
The analysis considers the effects of porewater ionic strength and the dry density of the buffer.
This treatment is judged sufficient at the present time, although it is understood that Fickian
diffusion is not strictly applicable to compacted bentonite, considering its microporous, semipermeable membrane structure.

Figure 6.4-14 Procedure for setting effective diffusion coefficients for buffer and cementitious
material

From the groundwater chemistry defined for the three representative host rocks, it isexpected that bentonite in the buffer will remain Na-type for a long period of time, while
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reaction between groundwater and leachate from cementitious material causing alteration to
Ca-type is considered (see Supporting Report 6-15): for each of these effective diffusion
coefficients are given.
With respect to the effective diffusion coefficients of elements in Na-type bentonite, the
JAEA-DDB [77] was used to select effective diffusion coefficients for Cs+, Sr2+, I-/ Cl- for the
given specifications of buffer materials (dry density, quartz sand mixing ratio, etc.), together
with the ionic strength dependency of diffusivity (Supporting Report 6-19) as determinedfrom effective diffusion coefficients of tritium. The speciation of relevant elements in buffer
porewater and their solubility limits were calculated assuming thermodynamic equilibrium at
25 °C (See Section 6.4.1 (2) (v) (b) 2.), using the thermodynamic database JAEA-TDB
140331S0.Tdb [74] with data partially modified for U and Zr (see Supporting Report 6-17).
Effective diffusion coefficients are set based on the charge of the predominant aqueous
species (Supporting Report 6-18), using the greater abundance of data for Cs+and Sr2+; I-/ Cland HTO to define reference values for cations, anions and neutral species respectively.
Furthermore, temperature corrections for the reference elements, and corrections based on the
diffusion coefficients in free water for actual and reference species, are applied. Here, thereliability of thermodynamic data for Pa is noted to be inferior to other elements, giving alarge uncertainty in the calculated chemical species in porewater and hence, for this element,
the largest among the effective diffusion coefficients of other elements was used. The sameprocess was also applied to data for Ca-type bentonite.
Effective diffusion coefficients of each element were determined for both Na-type and Catype bentonite as above, but uncertainties concerning the alteration from Na-type to Ca-type
are large and it is difficult to determine the temporal evolution process quantitatively. For the
RN migration analysis, therefore, it was conservatively decided to apply the larger values forthe effective diffusion coefficient for each element between the two bentonite forms.
Regarding the effective diffusion coefficients of RNs in TRU waste infill or in concrete
structural materials, a single value was set for all elements, based on a limited data set for
concrete equivalent to vault infill, which assessed both impacts of porosity and temperature[78]. For the reference porosity of 21%, effective diffusivities were set for 30 and 45°C, fordisposal at 500 and 1000 m, respectively.
Based on chemical and material properties presented in Supporting Reports 6-15 to 6-17,
the process of deriving effective diffusion coefficients of the RNs of interest for HLW andTRU waste are given in Supporting Report 6-19.

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- Sorption distribution coefficients
NUMO and JAEA also collaborated in determining elemental sorption data for buffer [76],using the procedure summarised in Figure 6.4-15. Linear, reversible sorption is assumed andtaken to be conservative, with nonlinearity handled by selecting sufficiently small sorption
coefficients. In terms of sorption distribution coefficients for buffer, sorption data acquired
using compressed bentonite of the form emplaced in the repository is limited [79][80], withmost experimental work carried out using bentonite suspensions. In terms of such sorption
data, there are conflicting views on its inapplicability to compacted bentonite (e.g. [79], [80]).
Thus, in this report, values of the sorption distribution coefficient are set based on diffusion
data determined for compacted bentonite.

Figure 6.4-15 Procedure for setting elemental sorption distribution coefficients for buffer

Targeting the specified buffer material for HLW and TRU waste (Section 6.4.1 (2) (v) (a)),
apparent diffusion coefficients for compacted Na-bentonite measured under test conditions
compatible with expected porewater chemistry were extracted from the JAEA-DDB [77].
Combining such data with the effective diffusion coefficients defined above, applicability of
Fick's law was assumed to calculate distribution coefficients [81]. For RNs with no reported
apparent diffusion coefficients for relevant porewater chemistry, data were derived fromelements considered analogous in terms of aqueous speciation.
Due to the complexity of the speciation of Pa and lack of clear analogues in the diffusion
coefficients, in this case the distribution coefficients were directly measured using bentonite
suspensions.
For elements (e.g. Sr, Cs and Ra) which are considered here to sorb by ion exchange
reactions with bentonite interlayer ions, higher sensitivity to ionic strength was expected thanfor other elements (Supporting Report 6-20). For the high salinity groundwater of Neogene
and Pre-Neogene sediments, no relevant apparent diffusion coefficients were available.
Therefore, empirical relationships were used to extrapolate measured sorption distribution
coefficients to higher salinity – again as described in Supporting Report 6-20. For Ca6-109

bentonite, this approach was limited by the lack of data for Ra, so Sr was used as an analogue.
Due to uncertainties in the degree and extent of Ca conversion as a function of time, as for
effective diffusivities, distribution coefficients for Na- and Ca-bentonite were derived and thesmaller value chosen for the safety assessment calculations.
TRU waste infill specification and porewater chemistry provide the basis for selecting
distribution coefficient data: based on uncertainties these are taken from the conservativevalues [71] in the JAEA-SDB [82], regardless if they represent region I or region II. For Gr.3
nitrate-containing wastes, sorption data are selected conservatively, choosing a value foroxidising conditions if this was lower than for reducing conditions.
Based on chemical and material properties presented in Appendix Tables 15 to 17, theprocess of deriving distribution coefficients of the elements of interest for HLW and TRU
waste are given in Supporting Report 6-20.
(C) Parameters for near field host rock
For the models used, the main parameters influencing RN migration in host rock for all
three representative host rock types are classified as those related to solute transport within
the channels represented by the one-dimensional multi-channel model and those related tosolute transport in the matrix. These were set as follows.
- Parameters related to the channels in near field host rock
• RN migration distance and dispersion length
As described in Section 6.4.1 (2) (iv) (e), the simplified model represents a network of
fractures by a set of channels: the migration distance of RNs in each channel is defined to be
100 m. Each channel is characterised by a dispersion length representing the mixing occurring
within the channel: the value selected follows the concept in the H12 report, where thedispersion length is set as one tenth of the channel length (100 m), i.e. 10 m.
• Matrix diffusion area
Because of channelling in fractures, not all of the surface of a planar fracture is accessiblefor diffusion into the surrounding rock matrix (flow-wetted surface), as discussed further forplutonic rocks in Supporting Report 3-29. As a percentage of the total fracture area, flowwetted surface area has been measured as 40 ~ 50% at the Kamaishi mine [83] and reported
by Pyrak-Nolte et al. [84] to lie in the range of 58 to 92%. On this basis, a matrix diffusionarea ratio of 50% is selected. However, no account is taken for the fact that, in aheterogeneous fracture, some of the flow will have access by diffusion to low flow or evenstagnant areas within the fracture. This may need to be considered in future assessments.
For Neogene sediments, in addition to fracture flow, advection may occur within thealtered rock matrix (Supporting Report 3-30) and hence the available matrix area is set to100%.
For Pre-Neogene sediments, like plutonic rocks, it is considered that the main migration
pathways of the groundwater are limited to fractures (see Supporting Report 3-31). Survey
results in the Shimanto Belt [85] indicate diffusion into the matrix from 60 to 80% of the

6-110fracture surface, as determined from the abundance of openings and fracture infill. The flow
wetted surface is conservatively set to 60% for this model.
- Rock matrix parameters• Matrix diffusion depth
The matrix diffusion depth is set based on the continuity of porosity extending from thefracture into the rock matrix. For plutonic rocks, alteration zones around fractures indicate
such connected porosity has a typical depth in the range of 0.01 to 0.1 m and diffusion into
the unaltered zone could also be considered (see Supporting Report 3-29). Here it was
decided to use a diffusion depth of 0.1 m, as in the H12 report.
For the Neogene sediments, there is no limitation on the range in which diffusion occurs,due to the connected porosity of the entire matrix. Specifically, Supporting Report 3-27
indicates an average fracture separation of 1.6 m, so that half the distance between thefractures is about 0.8 (see Supporting Report 3-30.
For the Pre-Neogene sediments, it is assumed that 10% of fractures are water carrying,corresponding to an average separation of 1 m, so that half the distance between the fractures
is about 0.5 m (see Supporting Report 3-28).
• Effective porosity and rock density
From Section 3.3.3 (7) the effective porosity for plutonic rocks, Neogene sediments and
Pre-Neogene sediments were set as 0.8%, 25% and 3.5%, respectively. In terms of rockdensity [86], a value of 2,700 kg/m3 was set for all host rocks.
• Effective RN diffusion coefficients
Here, the method used in H12 [3] [87] [88] was applied to take into consideration thedependence of the effective diffusion coefficient on rock porosity. Specifically, for each rock
type, JAEA-DDB [77], includes the most recent diffusion data. For any host rockgroundwater chemistry (see Table 3.3-16), the effective diffusion coefficient can be obtainedusing an empirical approximation for porosity correction (Supporting Report 6-21).
For the effective diffusion coefficient in granite, the JAEA empirical relationship [22]applies to all RNs irrespective of speciation and thus, for a defined porosity, the same value
was set for all elements.
The Neogene and Pre-Neogene sediments contain a range of clay minerals that may
influence the effective diffusion coefficient depending on the charge of chemical species, as
was the case for the buffer. The JAEA-DDB [77] effective diffusion data were analysed todevelop empirical functions relating these to species charge and rock porosity. The speciationof relevant elements was calculated as before, with effective diffusion coefficients set basedon the charge of the predominant aqueous species (see Supporting Report 6-18), corrected forambient rock temperature.
With regard to TRU waste Gr.3, assuming that nitrate flows into the host rock and the
distribution of the chemical species may change, in order to be conservative, all elements are
assigned an effective diffusion coefficient equal to the smallest among the calculated values.

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The effective diffusion coefficients set and further supporting details are provided in
Supporting Report 6-21.
• Sorption distribution coefficients
Sorption data for the different host rocks and specified groundwater chemistries wereextracted from the JAEA-SDB [44] [82] [89]. In cases where data were lacking, chemical
analogy was used to complete the database. For TRU waste Gr.3, the influences of increase inionic strength due to the nitrate and possible complex formation with ammonia generated by
reduction of nitrate are assessed. Since Cs, Sr, Ra and Pb do not form complexes, theirsorption is affected by the ionic strength, whereas Co, Ni and Pd are assumed to form
complexes. Sorption is conservatively not considered for RN migration analysis in the EDZ,
since its thickness is limited and its assumed porosity is high.
The selected distribution coefficients are listed in Supporting Report 6-22. Further
parameters used for the one-dimensional RN transport analysis are summarised in Table 6.4-1.

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Table 6.4-1 THMC conditions set for the near field (1/2)
Set THMC conditions
T

H

Temperature (°C)

Plutonic rocks, Pre-Neogene sediments: 45
Neogene sediments: 30

Hydraulic gradient

Plutonic rocks, Pre-Neogene sediments: 0.05
Neogene sediments: 0.06

Hydraulic conductivity of
buffer (m/s)

1.0 x 10-12 (H12V plutonic rock/Pre-Neogene sediments, TRU waste)
1.0 x 10-10 (PEM)-11
1.0 x 10 (Neogene sediments, TRU waste)

Hydraulic conductivity ofstructural framework (m/s)

1.0 x 10-7 (up to 200 y after closure)
1.0 x 10-5 (after 200 y after closure)

Hydraulic conductivity ofvault infill (m/s)
Hydraulic conductivity ofEDZ

1.0 x 10-5

EDZ flow rate

Appendix Table 4

Number of channels in model

41

Multi-channeltransmissivity (m2/s)

The transmissivity (Ti) of each channel obtained by fitting to the particle
tracking analysis result from Partridge (See 6.4.1 (2) (iv) (e))

Open width of fracture (m)Empirical relationship based on hydraulic conductivity of each channel
given by Ti above
The influences of changes of the stress field on RN migration parameters
are not consideredMC

100 times that of the host rock

Porewater chemistry of
bufferAppendix Table 5

Glass
Overpack
Porewater chemistry

Appendix Table 6

Cementitious infill/backfill
Porewater chemistry

Appendix Tables 7, 8

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Table 6.4-1 THMC conditions set for the near field (2/2)
Nuclide migration parameters

Nuclide dissolution rate (1/y)- HLW: 1/65,270 (congruent with glass)- TRU waste:Gr.1: instantaneous dissolution

Gr.2: 1/11,400 (Zircaloy), 1/8,500 (stainlesssteel/inconel), instantaneous dissolution (other)
Gr.3: instantaneous dissolution

Gr.4: instantaneous dissolution

Solubility

Appendix Tables 9 (Plutonic rocks), 10 (Neogene
sediments), 11 (Pre-Neogene sediments)

Effective diffusion coefficient

Appendix Tables 12 (Plutonic rocks), 13 (Neogene
sediments), 14 (Pre-Neogene sediments)

Sorption distribution coefficient

Appendix Tables 15 (Plutonic rocks), 16 (Neogene
sediments), 17 (Pre-Neogene sediments)

Substrate

Near field host rock

Fracture

EBS

Waste

Nuclide elution start time andRadioactive inventory at that time

- HLW
Elution start time: 1 ky after closureRadioactive Inventory: See SR 2-3- TRU wasteWaste package A Elution start time: Immediately after closure Radioactive Inventory: Table 6.1-4

Waste package B
 Elution start time: 300 y after closure Radioactive Inventory: See SR 2-3

Transfer distance (m)100Dispersion length (m)10Matrix diffusion
flow wetted surface (%)

Plutonic rocks: 50, Neogene sediments: 100, Pre-Neogene
sediments: 60

Matrix diffusion
depth (m)Plutonic rocks: 0.1, Neogene sediments: 0.8, Pre-Neogene
sediments: 0.5

Effective porosity (%)

Plutonic rocks: 0.8, Neogene sediments: 25, Pre-Neogene
sediments: 3.5

Rock density (kg/m 3)

2,700

Effective diffusion
coefficient

Appendix Tables 18 (Plutonic rocks), 19 (Neogene
sediments), 20 (Pre-Neogene sediments)

Sorption distribution
coefficient

Appendix Tables 21 (Plutonic rocks), 22 (Neogene
sediments), 23 (Pre-Neogene sediments)(3) Panel scale models and datasets
As described in Section 6.4.1 (2) (ii), at the near field scale, the models and datasets aredeveloped for downflow waste packages in each panel and 100 m of host rock flow path. At
the panel scale, the migration of RNs calculated in the near field scale was expanded by thefollowing simple method and converted into a release rate from each disposal panel.

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For HLW disposal in plutonic rocks, RN migration analysis at the panel scale is the samefor both the H12V and PEM options. The near field model for calculated RN migration froma number (ni) of representative waste packages at the downflow end of disposal tunnels(φN(HLW), i (t)), for each disposal panel i7. This is converted to a release rate from the entire
panel (φP(HLW), i (t)) by multiplying by the total number of waste packages in it. Here,conservatively, the impact of the up-flow waste packages suppressing the release of RNs fromthe EBS downstream is ignored (as in H12). This assumption is further justified byconsideration that the flow in the rock takes place in a discrete fracture network, reducing theprobability that the groundwater passing a waste package would also pass other waste
packages. Also, as described in Section 4.5.4, a main access tunnel is located downstream at
the edge of the disposal panel (separation distance 30 or 60 m from the nearest disposal
tunnel), but such tunnels are not considered in the analysis (see Figure 6.4-4). The influenceof such access tunnels on RN migration analysis was examined and confirmed that it is not
significant (see Supporting Report 6-14).
For TRU waste in plutonic rocks, release rates of RNs on the panel scale are calculated foreach waste group. Specifically, the calculated RN releases at near field scale for group j areused to derive a release rate (φN (TRU), j8 (t)) from one disposal vault, which is thenmultiplied by the number of vaults, mj, for that group in order to assess the panel scalereleases (φP (TRU), j (t)). As noted in Section 4.5.4, a main tunnel exists at a distance ofabout 30 m downstream of the disposal tunnel, within the area analysed at the near field scale,but, as for HLW, this is not considered (see Figure 6.4-5). Also, here the influence of such atunnel on RN migration was examined separately and confirmed that it is not significant (seeSupporting Report 6-14).
In the cases of both HLW and TRU waste in Neogene sediments, the near field RN
migration model explicitly considers potential short circuits to access tunnels downstream(See Figure 6.4-4). Thus, the panel scale RN migration analysis considers the influence of theaccess tunnel at the downstream end of the vault on RN migration. As for HLW, thisapproach was also conservative in its treatment of the upstream vault.
The analysis models and approach for both HLW and TRU waste in Pre-Neogene
sediments is the same as described for plutonic rock. The thermal and mechanical datasets
used for RN migration analysis on the panel scale are the same as those for the near field.
Regarding the geochemical conditions, the groundwater chemistry of the host rock may bechanged by leachate from the disposal tunnels and other materials in the access tunnels, but
the extent of change is assumed limited and its influence on mobility of RNs was ignored.
This assumption is judged adequate for the HLW repository, since the source impact on
groundwater chemistry is limited, but may need further analysis for the TRU repository. At
the same time, such reactions also alter the backfill material in tunnels: the hydraulicconductivity of the backfilling material is set taking such alteration into account. Since thehydraulic plug is designed to maintain the same hydraulic conductivity as the host rock for along period of time, it is not treated as a preferential migration route for RN migration
analysis. Mechanical plugs are treated as having the same hydraulic characteristics as the
backfill material. These hydraulic conditions are as shown in Table 6.4-1.

7

Repository in plutonic rocks and Pre-Neogene sediments: i = 1 ... 6; Repository in Neogene sediments: i = 1 ...
8.
8
j = 1, 2, 3, 4HD, 4HH, 4LC, 4LD

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(4) Repository scale models and datasets(i) Simplified RN migration model
Figure 6.4-16 shows the concept of the simplified model used for RN migration analysis
on the repository scale.

Figure 6.4-16 Overview of the simplified RN migration model - repository scale (HLW andTRU waste)

For HLW, the RN releases from the panel scale analysis described above, φP(HLW), i(t),are used for input to a multi-channel model, as described in Section 6.4.1 (2) (iv) (shown in
Figure 6.4-7) to calculated transfer rate through repository scale host rock. More specifically,for each panel i with release rate φP(HLW), i(t), a multi-channel model of host rock with a
distribution of hydraulic characteristics pk, for each channel k (k = 1,…, 41) is developed in amanner described in more detail below. This allows RN migration calculations using the
GoldSim © code to determine releases at a repository scale, φR(HLW), i(t). Then, by
summing releases from each panel, the release from the entire repository φR(HLW)(t) is
calculated.
Similarly, for TRU waste, RN migration at the panel scale for each group j, φP(TRU), j(t),is used as input to the repository scale multi-channel model and summed to derive the total
release from all groups, φR(TRU)(t). In the case of co-located disposal, the total release fromthe repository, φR(HLW + TRU)(t), is the sum of the releases from the two components (seeSupporting Report 6-14).
Since the hydraulic conductivity of each channel within the multi-channel model of therepository host rock depends only on its large scale hydraulic and mass transfer characteristics,
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the same values are used for HLW and TRU waste. Hydraulic conductivities of each channel
are derived from particle tracking analysis of the 3D DFN (analysis region 100 m x 100 m x
150 m), for non-sorbing tracers configured to approximate the actual groundwater flow, as
discussed previously.
The channel lengths in the repository scale multi-channel depend on the location of thedisposal panel, in the case of HLW, and the specific group disposal vaults, in the case of TRU
waste. The repository layout at the disposal depth is set by selecting areas where thegroundwater travel time is relatively long and the Darcy flow velocity is small, based onregional scale groundwater flow analysis. However, how this can be assured in practice at agiven site will need more consideration. Although it is affected by the spatial distribution ofanisotropic faults and by changes in local topography, it is assumed here that groundwaterflows at repository depth roughly in a horizontal direction (see Supporting Reports 3-20, 3-21,3-22). Therefore, for HLW, channel length from the down flow edge of each panel isdetermined by distance in the flow direction to a boundary (either a fault with length of 1 km
or more9, a TRU waste disposal area or a modelling area boundary), as shown in Figures 6.417, 6.4-18 and 6.4-19. However, such an assumption will need careful site specificassessment in the future.
Specifically for Pre-Neogene sediments, hydraulic characteristics may differ greatly in theregion of a thrust fault and thus, when such areas lie on the line defining the channel length,
only the distance within sound rock is considered. In the case where another disposal panellies on a straight line connecting it to a downstream boundary, the migration through thedownstream panel is conservatively excluded, i.e. by not adding the distance through thepanel to the channel length. When the channel length is 100 m or less, it is treated as 0 m.
The procedure for selecting channel lengths for TRU waste was similar: as the footprint of
this facility is relatively small, the separation of the entire repository from a boundary wasused for all waste groups for both plutonic and Pre-Neogene host rocks, but distances fromtunnels of the different waste groups are considered separately for the Neogene sediments due
to layout adaptions, resulting in placing the TRU repository close to a water conducting fault.
It is noted that the resulting very short channel lengths for the Neogene case could possibly
have been increased by considering another layout.
For more details of the multi-channel model used for RN migration analysis on therepository scale, see Supporting Report 6-13.
(ii) Dataset for RN migration analysis at the repository scale
The RN migration parameters required for the multi-channel model at the repository scale
are the same as those for the near field scale analysis, described in 6.4.1 (2) (v) (b) 3 andgiven in Table 6.4-1. The multi-channel hydraulic conductivities and channel lengths are
given in Tables 6.4-2 to 6.4-4, which are set as described previously based on the reference
repository layouts for the different host rocks shown in Figures 6.4-17, 6.4-18 and 6.4-19.
More details about setting channel lengths for the multi-channel model at the repository
scale are provided in Supporting Report 6-13.

9

Faults avoided when developing repository layout, as described in Chapter 4.

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Figure 6.4-17 Set distances from the repository to model boundaries for reference layouts inplutonic rock (Upper: H12V, Lower: PEM, Gr.: Group)

6-118

Figure 6.4-18 Set distances from the repository to model boundaries for reference layouts in
Neogene sediments (Upper: H12V, Lower: PEM, Gr.: Group)

6-119

Figure 6.4-19 Set distances from the repository to model boundaries for reference layouts in
Pre-Neogene sediments (Upper: H12V, Lower: PEM, Gr.: Group)

6-120

Table 6.4-2 Channel lengths for multi-channel model of plutonic rock (repository scale)
HLW disposal
optionWaste disposal
panel

Multi-channel model
boundary

Model channel
length (m)1, 2, 3, 4

Fault > 1 km

220

5Fault > 1 km

170

6

TRU waste disposal vault

140

All Groups

Modelling area boundary

2,260

1, 2

Fault > 1 km

0

3, 4

Fault > 1 km

360

5, 6

Fault > 1 km

0

All Groups

Modelling area boundary

2,260

H12V

PEM

Table 6.4-3 Channel lengths for multi-channel model of Neogene sediments (repository scale)
HLW disposal
option

H12V

PEM

Waste disposal
panel

Multi-channel model
boundary

Model channel
length (m)1Modelling area boundary580

2

Modelling area boundary

770

3

TRU waste disposal vault

160

4

TRU waste disposal vault

1305

Fault > 1 km

410

6Fault > 1 km

660

Gr.1, Gr.2, Gr.4LD

Modelling area boundary

780

Gr.3

Modelling area boundary

440

Gr.4HH, Gr.4HD,
Gr.4LC

Gr.3 disposal tunnel

01,2Modelling area boundary

800

3,4

Fault > 1 km

590

5Fault > 1 km

360

6Fault > 1 km

700

Gr.1, Gr.2, Gr.4LD

Modelling area boundary

780

Gr.3

Modelling area boundary

440

Gr.4HH, Gr.4HD,
Gr.4LC

Gr.3 disposal tunnel

0

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Table 6.4-4 Channel lengths for multi-channel model of Pre-Neogene sediments (repository
scale)
HLW disposal
optionWaste disposal panel

Multi-channel model
boundary

Model channel length(m)

1, 3, 5, 7

Thrust0

2

TRU waste disposal vault

480

4, 6, 8Fault > 1 km

560

All Groups

Thrust310

1, 3, 5, 7

Thrust380

2

TRU waste disposal vault

760

4, 6

Fault > 1 km

900

8Fault > 1 km

530

All Groups

Thrust

310

H12V

PEM

(5) Regional model and dataset
When conducting RN migration analysis on a regional scale, it is necessary to consider
impacts of likely changes in the migration path due to topographic changes resulting from
uplift/erosion, climate and sea level change, etc. Since such changes depend largely on thecharacteristics of the actual sites (as yet unspecified), as described in Section 6.4.1 (1), outside
the repository scale, it was decided to apply a model in which all RNs are transferred to thebiosphere via flow within large-scale faults. These large-scale faults are represented by anequivalent porous medium model, and parameters related to hydrogeology and RN transfer
are set based on current understanding, as shown in Table 6.4-5 and with further backgroundgiven in Supporting Report 6-23.
Table 6.4-5 Parameters for large faults
Parameter

Set value

Transport distance

700 m (plutonic rocks and Pre-Neogene sediments),
200 m (Neogene sediments)

Dispersion length

70 m (plutonic rocks and Pre-Neogene sediments),
20 m (Neogene sediments)

Hydraulic conductivity

1.0 x 10-5 m/s

Hydraulic gradient infault

Plutonic rocks, Pre-Neogene sediments: 0.05
Neogene sediments: 0.06
0.1

Effective porosity

2,700 kg/m3

Density
Effective diffusion
coefficient

Appendix Tables 18 (Plutonic rocks), 19 (Neogene sediments),
20 (Pre-Neogene sediments)

Sorption distribution
coefficient

Appendix Tables 21 (Plutonic rocks), 22 (Neogene sediments),
23 (Pre-Neogene sediments)

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Since this fault model is very conservative, the dose results finally calculated using thesemodels and parameters are almost the same as those obtained by directly inputting the flux of
RNs output from the repository scale model directly into the biosphere model, showing thatthese do not have a significant impact (see Supporting Report 6-23). Therefore, the doses
shown in the following (base cases Section 6.4.1 (7), variant cases Section 6.4.2) arecalculated with the conservative assumption of repository scale releases directly into the GBI,
the latter being modelled as a mixing tank with parameters representing the surface
environment and not considering how it is actually connected to the flow paths. In addition, it
is expected that the depth of the repository will change over the long-term due to
uplift/erosion, which will change the fault length in the RN migration model for the regional
scale and thus this approach is robust against such uncertainties. Other aspects of
uplift/erosion are not considered in the current safety case, as previously discussed in Section
6.3.

(6) Biosphere model and dataset
In the safety assessment, for RNs released from the geosphere into the surface environment,
the resultant consequences were modelled based on both migration processes in the biosphereand human lifestyle in the area impacted. In terms of biosphere evaluation, a general
assessment method is illustrated in IAEA's international collaborative project BIOMASS [90],but it is noted, as highlighted in the H12 report, that the specific lifestyle of each country has
to be taken into consideration. It is inherently difficult to assess human lifestyles and surface
environmental conditions for long times into the future; thus, the future lifestyle (stylised) is
generally assumed to be the same as at present, to form a basis for analysis of RN transfers inthe biosphere and resultant doses (e.g., [25] [29]). In this report, the biosphere evaluation was
conducted based on such an international approach.
The International Commission on Radiological Protection (ICRP) focuses on public
protection and the determination of compliance with dose constraints, assessing doses togroups with relatively high exposure in order to define a “representative person” [91]. For thedose evaluation in this report, the radiation dose for such a representative individual was
calculated and is compared to the dose constraints in Table 6.1-6. At present, since therepository sites are not decided, the biosphere will be evaluated by modelling potentially
exposed groups as representative individuals, as described below. For details of the biosphereevaluation, see Supporting Report 6-1.
(i) Modelling of biosphere RN migration and resulting doses
Modelling migration processes and resultant radiation exposure in the surface environmentis based on the latest developments in Japan and abroad [90] and proceeds as follows.
Firstly, the characteristics of the surface environment that influence behaviour of RNs
released into the biosphere need to be considered; including the GBI, climate,geography/topography, surface hydrology, soil type, biota together with human activities andlifestyles. In this description, long-term surface evolution by climate change, uplift anderosion, etc. is also taken into consideration. Current anthropogenic climatic change impacts
can also be considered in such scenarios, although this was not done in the present analysis.
Next, individual environmental compartments (such as river water or soil) are identified andrepresented as boxes of fixed volume. While it is understood that the surface environmentchanges with time, these boxes are assumed to represent typical biosphere compartments also
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in the future. It is recognised that there are approaches for more explicitly consideringchanges to the surface environment that may be used in coming assessments. Processes arethen identified that transport RNs between compartments and a box model is developed toquantify these, while assuring mass balance and accounting for impacts of RN decay andingrowth.
Without a site being identified at present, conditions necessary for describing the biospheresystem are derived from the previous H12 and TRU-2 studies. The Nationwide Map
published recently [92] identified not only excluded areas (e.g., within a radius of 15 km fromthe centre of Quaternary volcanoes), but also scientifically favourable areas, which have an
average altitude in the order of 100 m above sea level (Supporting Report 6-10). Such terrain
is classified (according to [93]) as relatively low-lying plains. If the repository is assumed to
be built in such a location, expected groundwater flow systems will result in releases to thesurface being located downstream from the repository footprint, so it was consideredreasonable for rivers in such plains to act as the GBI. The release into the river can then bepassed to other compartments, such as irrigated farmland or the sea. A major river represents
the most typical GBI in Japan, but, in a more site-specific assessment, it is understood that the
GBI selection must be more representative of the site and could also consider other objects
such as small aquifers and smaller rivers.
In addition, it is important to consider the temperate climate assumed in the biosphereassessment model, because the majority of areas in Japan today are in such a zone, andglaciated areas are limited to a few areas, even if long-term climate change is considered [94].
For the exposure process, the groups most likely to be exposed by RNs that reach thebiosphere are identified, while taking into account the group lifestyle, food sources, drinking
water, etc., in order to calculate internal doses by RN ingestion, respiration and external
exposure.
Representative persons are selected from groups assumed to have relatively high radiation
doses. In the H12 and TRU-2 reports, such groups included “farmers”, “freshwater fishermen”
and “marine fishermen”, which are the focus also in this report. Dose conversion factors (see6.4.1 (6) (iii) below) are calculated in the related compartments based on the assumed lifestyle
for adults of these three groups.
An example of the exposure assessment for a representative person in the farmers group,with RN migration and uptake processes that assume a warm climate and a river water GBI, isshown in Figure 6.4-20.

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Figure 6.4-20 Example of RN migration and radiation exposure processes in the biosphere(River water GBI, temperate climate, farmer exposure group)This model is the same as that used in the TRU-2 report. Each box in the figure represents
a compartment in the biosphere: by solving simultaneous equations for mass balance during
RN transfers between these compartments, the distribution of each RN can be calculated ateach time step.

(ii) Dataset
As above, the required dataset, based on that used in both the H12 and TRU-2 reports, hasbeen updated to reflect average and median values calculated from the latest statistical data on
consumption, river flow, irrigation water, food production, etc. Thus, the latest knowledge
about processes influencing RN transfer parameters in the surface were reflected in theeffective dose conversion factors, which mirror the method of setting data in the latest
Japanese biosphere evaluation [22]. Regarding the river flow rate, a logarithmic mean value
was derived from the statistical data, assuming that it reflects a lognormal distribution. In the
H12 and TRU-2 reports, the river flow was set as a conservative value based on thenationwide data on class A rivers10. The value of river flow set in this report uses the average
10Rivers that are part of river systems considered to be particularly important for the maintenance of the land or
national economy. These rivers are designated by the Minster of Land, Infrastructure and Transport. N.B. This
footnote is not included in the Japanese version of the report.

6-125value of such data, since this is judged more representative before sites are selected. This isabout 10 times the value used in the H12 and TRU-2 reports (same as the applied value in[22]). Details on the selection of the dataset are given in Supporting Report 6-1.
(iii) Setting the dose conversion factor for a representative individual
The dose conversion factor relating exposure dose to the input flux of RNs at steady stateis directly derived by the biosphere model used in the H12 and TRU-2 studies. RN
redistribution times in the biosphere are shorter than the transit time from the repository to the
GBI, so use of such steady-state values in the safety assessment is justified and doses are
simply obtaining by multiplying the input rate of the RNs flowing into the biosphere by thesefactors. In this report, this method was applied and dose conversion factors calculated using
various biosphere models and corresponding datasets. For the representative persons, anexample of dose conversion factors is shown in the Table 6.4-6, including modelled doses toagricultural workers as shown in Figure 6.4-20. It may be noted that, despite a much higher
dilution in the sea water compartment, the dose conversion factors to the marine fishery groupare higher for C-14 compared to the other groups and not much less for other RNs. This is dueto the modelled bio-accumulation of C-14 in the marine ecosystem and the eating habits
assumed for the different representative groups.
Table 6.4-6 Dose conversion factors for representative persons(river water GBI, temperate climate) (1/2)(Sv/y)/(Bq/y)Nuclide
C-14
Cl-36
Co-60
Ni-59
Ni-63
Se-79
Sr-90
Zr-93
Nb-93m
Nb-94
Mo-93
Tc-99

Farmer group-18

1.6 x 10
6.3 x 10-18
3.9 x 10-17
3.6 x 10-17
1.5 x 10-17
8.4 x 10-16
3.8 x 10-16
2.9 x 10-18
1.4 x 10-19
1.8 x 10-15
1.9 x 10-17
4.6 x 10-18

Freshwater fishery group-18

5.8 x 10
7.4 x 10-19
6.9 x 10-17
4.4 x 10-20
9.6 x 10-20
1.9 x 10-18
2.0 x 10-17
8.2 x 10-19
8.1 x 10-20
1.2 x 10-16
2.0 x 10-18
3.9 x 10-19

6-126

Marine fishery group
2.0 x 10-17
1.2 x 10-22
1.3 x 10-17
1.4 x 10-19
3.2 x 10-19
2.8 x 10-17
2.1 x 10-19
2.7 x 10-18
7.5 x 10-20
2.7 x 10-17
1.8 x 10-19
2.6 x 10-18

Table 6.4-6 Dose conversion factors for representative persons(river water GBI, temperate climate) (2/2)(Sv/y)/(Bq/y)NuclideFarmer group

Freshwater fishery group

Marine fishery group

Pd-107
Sn-126
I-129
Cs-135
Cs-137
Cm-248
Pu-244
Cm-244
Pu-240
U-236
Th-232
Ra-228
Th-228
Np-236
Pu-236
U-232
Cm-245
Pu-241
Am-241
Np-237
Pa-233
U-233
Th-229
Cm-246
Pu-242
U-238
Am-242m
Pu-238
U-234
Th-230Ra-226
Pb-210Po-210Cm-247
Cm-243Am-243
Pu-239
U-235
Pa-231
Ac-227

1.0 x 10-18
2.9 x 10-15
2.6 x 10-16
7.2 x 10-17
1.1 x 10-16
6.5 x 10-15
1.4 x 10-15
1.4 x 10-16
9.1 x 10-16
1.8 x 10-16
2.4 x 10-14
8.7 x 10-16
1.5 x 10-16
6.8 x 10-16
8.3 x 10-17
7.8 x 10-16
2.0 x 10-15
1.6 x 10-17
5.6 x 10-16
1.9 x 10-15
9.3 x 10-19
2.9 x 10-16
3.3 x 10-15
1.1 x 10-15
9.5 x 10-16
1.9 x 10-16
4.4 x 10-16
2.7 x 10-16
2.3 x 10-16
8.4 x 10-15
1.6 x 10-14
1.3 x 10-15
1.3 x 10-15
1.9 x 10-15
1.9 x 10-16
1.5 x 10-15
9.7 x 10-16
4.0 x 10-16
1.7 x 10-14
1.6 x 10-15

2.3 x 10-20
1.4 x 10-16
7.0 x 10-17
4.9 x 10-18
1.1 x 10-16
2.3 x 10-15
2.4 x 10-16
1.0 x 10-16
1.6 x 10-16
2.9 x 10-17
3.4 x 10-16
7.1 x 10-16
1.5 x 10-15
1.1 x 10-17
5.7 x 10-17
2.1 x 10-16
3.1 x 10-16
3.1 x 10-18
1.6 x 10-16
7.0 x 10-17
6.2 x 10-18
3.1 x 10-17
8.5 x 10-16
1.8 x 10-16
1.6 x 10-16
3.0 x 10-17
1.6 x 10-16
1.5 x 10-16
3.0 x 10-17
1.5 x 10-16
5.7 x 10-16
4.7 x 10-16
1.8 x 10-15
6.8 x 10-16
3.3 x 10-16
3.3 x 10-16
1.6 x 10-16
3.2 x 10-17
4.7 x 10-16
9.2 x 10-16

3.1 x 10-20
4.4 x 10-15
4.1 x 10-17
2.6 x 10-19
7.2 x 10-18
1.0 x 10-14
4.1 x 10-16
1.6 x 10-15
4.1 x 10-16
8.2 x 10-19
1.0 x 10-15
7.1 x 10-16
2.1 x 10-16
9.2 x 10-18
1.4 x 10-16
8.5 x 10-18
2.8 x 10-15
8.0 x 10-18
1.8 x 10-15
5.9 x 10-17
2.8 x 10-19
9.0 x 10-19
1.0 x 10-15
2.7 x 10-15
3.9 x 10-16
9.2 x 10-19
1.7 x 10-15
3.8 x 10-16
8.6 x 10-19
3.4 x 10-16
4.0 x 10-16
8.0 x 10-16
2.8 x 10-14
2.6 x 10-15
2.0 x 10-15
1.8 x 10-15
4.1 x 10-16
1.4 x 10-18
2.6 x 10-16
1.0 x 10-15

As described in Section 6.4.1 (6) (i), assuming a river or marine GBI in a lowland plain area,
the doses are calculated for the above three groups in the corresponding biosphere systems.
Comparing the conversion coefficients for the different representative groups shows that, theconversion factor for farmers, with the river GBI, almost always is higher than the conversion
factors for the marine fishermen due to the much higher dilution in the sea compartment.

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Considering the possible impacts on lifestyle due to climatic change, it is considered thatassuming current temperate conditions is conservative.
For the calculations of dose in this report, the dose conversion factors to farmers living in atemperate climate with a river GBI was applied to all analysis cases.
Dose conversion factors for farmers for a representative person after the updating thedataset result in changes relative to the H12 and TRU-2 studies, as illustrated in Figure 6.4-21.

Figure 6.4-21 Dose conversion factor for farmers(GBI river water, temperate climate)

In comparison with the dose conversion factors in the TRU-2 report, values for Ni-59, Ni63, Nb-94, Sn-126 and Np-237 have increased, despite the increased dilution resulting fromthe increased river flow. This is due to a major revision of the data base concerning theamount of food ingestion, distribution coefficients (Kd values) and corrections of half-lives
for these RNs, where the Kd has increased two orders of magnitude for Np and about oneorder of magnitude for Nb and Sn compared to the TRU-2 study. The assumed half-life of Ni59 has increased by about 25% since the time of the TRU-2 report publication. Data for other
RNs have also changed. Furthermore, the impact of these changes depends on the
combination of data, which for some RNs makes the irrigated agricultural land thedominating pathway instead of the drinking water pathway. Values for all other RNs havedecreased, generally by about an order of magnitude due to the larger river flow now assumed
resulting in an order of magnitude more dilution and reduction of dose to the farmer by thesame factor [22].
(7) Results for the base cases
This report focuses on sites with co-location of disposal facilities for HLW and TRU waste,with the options shown in Table 6.1-1 taken into consideration. Based on the models andparameters presented in Sections 6.4.1 (2)-(6) above, the dose assessments of the base cases
for these options are now presented. For more details on these calculations, see Supporting
Reports 6-24, 6-25 and 6-26 for plutonic rocks, Neogene sediments and Pre-Neogene
sediments, respectively.

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In the following, figures showing the results of assessments for the base scenario (SeeTable 6.1-6) present annual effective doses to the reference person and also show that due tonatural radiation in Japan (2,100 μSv/y), which helps put the results in context.
(i) Dose from the entire repository
Safety assessment of the repository as a whole was carried out by summing contributions
from both HLW and TRU waste. As mentioned in Section 6.4.1 (2) (v) (a) 4., uncertainty ingroundwater chemistry is captured by two model waters (with low and high salinity). HLW
assessments include the H12V and PEM options while, for TRU waste, two kinds ofcontainers are considered: A (loss of containment immediately after closure) and B (loss ofcontainment 300 y after closure). By combining these, a total of 8 base case variants for eachhost rock will be considered.
Figure 6.4-22 shows the calculated results. In all cases, the maximum dose is less than 10
μSv/y, set as the target dose for the base scenario. As discussed in Section 6.1.5 (2) (i), thisprovides a good argument for safety.
For Figure 6.4-22 and similar figures thereafter, the following assumptions are made:For all waste packages, the timing of contact between the waste package andgroundwater and the start of RN leaching:o for vitrified waste: 1,000 years after closure of the repositoryo for waste package A: immediately after closure of the repositoryo for waste package B: 300 years after repository closureDose calculation using RN migration rates from repository scale host rock as direct
input to the biosphere6-129

Plutonic

Neogene

Pre-Neogene

Legend:Figure 6.4-22 Doses from the entire facility for each of the disposal options (Base case)

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In terms of groundwater chemistry variants, in plutonic rock there is little difference indose curves for the two model waters but, for both sedimentary rocks, dose peaks are
significantly higher for the high salinity cases. This will be further discussed below. For thetwo HLW disposal options, the calculated doses were effectively identical in all host rocks;
this reflects the simplifications adopted in the near field transport model, lack of optimisationof repository layout and, in particular, the dominant contribution of TRU waste to total dose,so should not be over-interpreted.
In terms of TRU waste variants, in the case of waste package A, the conservative
assumption is made that elution of RNs begins immediately after closure, hence initial
breakthrough occurs earlier, but the dose maximum is effectively the same as that for waste
package B. Below, the results of dose evaluation will be discussed in more detail.
(ii) RNs that dominate dose
As described above, there are almost no differences in the dose evaluation results between
the H12V and PEM variants with regard to calculated doses, due to the simplifications
adopted in the near field transport model, Therefore, the following discussion will focus on
H12V. It is noted, however, that a more detailed near field model may have resulted indifferences and also that there are several differences between the concepts with regard tohandling and ease of installation, as discussed in Chapter 4.
(a) Plutonic rocks (see Supporting Report 6-24)
Figure 6.4-23 shows the separate doses resulting from HLW and TRU waste in plutonic
rocks and the main RNs contributing to them. By comparing Figure 6.4-22 and Figure 6.4-23,
it is clear that in all cases, RNs which dominate dose up to 100 ky are derived from TRU
waste, in particular I-129, giving a maximum dose of about 2 μSv/y after about 3 ky This
report conservatively sets the amounts of I-129 and Cl-36 in the HLW inventory and so,
unlike in the H12 study, these are listed as the target RNs (Section 6.1.3). However,contribution to the maximum dose of the entire repository is mostly I-129 derived from TRU
waste.
I-129 is dominant due to use of a conservative TRU Gr.1 model with a mixed tank source
term after instantaneous release from the waste matrix. This is combined with modelled
speciation as the stable iodide anion, with no solubility limit, zero sorption and lack ofsignificant radioactive decay due to its very long half-life (1.6 x 107 y). Cl-36 has similar
chemical characteristics and a long half-life (3.0 x 105 y), so its contribution to the dose issignificant, even if I-129 dominates by a factor of 10 for HLW and very much more for TRU
waste.
Even though models for such timescales must be treated with great caution, after 100 ky
dose is dominated by Se-79 from HLW. This is because Se, though very mobile, reacts withiron derived from overpacks, buffer, pyrite in host rock, etc. and precipitates as irondiselenide. Nevertheless, the model predicts anionic speciation, and hence very low sorption
in buffer and host rocks combines with a long half-life (3.0 x 105 y) to give significant
releases.

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HLW
Low salinity

High salinity

TRU Waste
High salinity

Waste package B

Waste package A

Low salinity

Figure 6.4-23 Doses and dominant RNs for plutonic repository (base case)

Regarding the cases with early calculated doses, combinations of conservatisms lead toresults that should be considered more as “What if?” cases, rather than any kind ofquantitative assessment of credible system evolution. This is particularly evident for TRU
waste, where the mixing tank model of release from the EBS ignores the time required forrepository re-saturation and failure of waste containers, constraints set by leaching of thewaste matrix and slow RN diffusion through the infilled vault and the surrounding structuralconcrete.
As it is assumed that elution of RNs begins immediately after closure for TRU waste
package A, the combination of very conservative transport properties (see Supporting Report
6-21 and Supporting Report 6-22) result in very short-lived Sr-90 (30-y half-life) from Gr.3
giving an early dose breakthrough, even though its value is extremely small. It is certain that
this Sr-90 peak will disappear if more realistic modelling was done, which will be necessary
to allow future assessment of varying EBS concepts for TRU waste. As such, features like

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breakthrough of short lived RNs and associated relatively high dose maxima are again
artefacts resulting from the model and should not be over-interpreted.
Because the two model waters in plutonic rocks are similar, the values of RN transfer
parameters such as solubility, distribution coefficient and diffusion coefficient are similar, so
the calculated doses are almost the same.
(b) Neogene sediments (see Supporting Report 6-25)
Figure 6.4-24 shows the doses resulting from HLW and TRU waste in Neogene sediments
and the main RNs contributing to them.
HLW
Low salinity

High salinity

TRU Waste
High salinity

Waste package B

Waste package A

Low salinity

Figure 6.4-24 Doses and dominant RNs for Neogene repository (base case)

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For low salinity water, by comparing Figures 6.4-22 and 6.4-24, it is clear that the dose isdominated by TRU for the entire evaluation period. The dominant RN in the TRU waste is
initially C-14 from Gr.4, but then I-129 from Gr.1, giving a maximum dose of about 0.1
μSv/y (after 35 ky). The reason why C-14 peak arrives earlier than I-129 is an artefact of thereference layout (Figure 6.4-17) and the simplified model assumptions. Although all Gr.4 isthe same distance from the model boundary, that upstream of Gr.3 (Gr.4H, the main source ofC-14) is treated as if the migration distance was zero (as the distance to the nearest TRU vault
is < 100 m). Again, this emphasises the need for a more realistic model to assess TRU waste
disposal variants. The large impact of I-129 is explained as for the plutonic rock case.
In the case of high salinity groundwater, up to about 40 ky after closure, the situation issimilar, with TRU waste dominating: initially C-14 from Gr.4 and then I-129 from Gr.1.
Thereafter, however, the situation changes completely with the maximum dose from the entire
repository after about 120 ky (0.7 μSv/y) predominantly due to U-233 from HLW.
Such unusual mobility of U in this reducing environment is due to the carbonate
concentration of the high salinity groundwater set for Neogene sediments (about 40 mM,
compared to about 2 mM for low salinity Neogene groundwater and about 0.2 mM for thehigh salinity plutonic rock). From calculations using the latest thermodynamic data, U forms
stable anionic carbonate complexes in both buffer porewater and groundwater in this case,thus stabilising the VI valent oxidation state (particularly UO2(CO3)34- and UO2(CO3)22-). No
sorption data on such species in a reducing environment were found, but experimental studies
of sorption of this kind of anionic complex in an oxidising environment indicate that sorption
partition coefficients on clay minerals are extremely small [95]. Based on these factors, U is
assigned a high solubility and an extremely small sorption distribution coefficient on hostrock, causing U-233 to become the dominant RN. However, the approach may be overly
conservative as neither sorption of such complexes onto carbonate minerals nor microbial
utilisation of U(VI) under reducing conditions has been assessed.
The reason why U-233, with a half-life of 1.6 x 105 y, dominates the dose over such a long
period is due to continuous ingrowth from its parent Np-237 (with a half-life of 2.1 x 106 y),which will remain in the buffer and rock for long times due to its low solubility and highsorption. The difference between peaks from U-233 originating from HLW and TRU isbasically due to the fact that the inventory of Np-237 is higher in HLW than in TRU waste.
It is also noticeable that, after about 100 ky after closure, there is a major contribution of
Pb-210 (half-life 22 years), derived mainly from HLW but also from TRU waste Gr.4. At thistime, Pb-210 and its parent Ra-226 (half-life 1600 y) would normally be expected to be insecular equilibrium with Th-230 (half-life 7.5 x 104 y). The dose calculation considers the
behaviour of relevant daughters in the decay chains and hence the inventory of Th-230
increases due to ingrowth from U-234, itself supported by ingrowth from Pu-242 (with a halflife of 3.8 x 105 y) which, as for Np-237, is effectively immobile due to its low solubility andhigh sorption. For Pb-210, it is considered that it can form a stable carbonate complex ingroundwater; however, there are no sorption data for the carbonate concentrations andchemistry of the high salinity groundwater and hence the distribution coefficient on the hostrock is very conservatively set to zero. The Pb-210 doses are thus determined by the
migration of Ra-226 from HLW and the different TRU waste sources and its rapid transport
to the biosphere when it is generated by Ra-226 decay.
It is clear that, unlike the plutonic case, there is little difference in the resultant doses from
TRU waste packages A and B, apart from an earlier breakthrough of low levels of C-14 and I129. In particular, this is due to distribution coefficients for Sr-90 set one or more orders of
6-134

magnitude larger, allowing effectively complete decay of this short-lived RN (half-life 29
years).
The fact that C-14 does not contribute to the initial doses of plutonic rocks is due to much
longer modelled migration distance for the repository scale, allowing for greater retardation ofthis RN by matrix diffusion. In terms of dose from I-129, the Neogene sediments maximum isabout an order of magnitude less than that for plutonic rocks. This is because, as described in
Section 6.4.1 (2) (v) (b) 3., for the Neogene sediments, the multi-channel model of the hostrock assumes double the flow wetted surface and eight times the matrix diffusion depth of theplutonic rock case (see Appendix Figures 1 and 2), causing much greater dispersion of therelease plume.
For plutonic rock, the dose of Se-79 dominates for times > 100 ky. For Neogene sediments,
the dose contribution from Se-79 is negligibly small for HLW and of second order importancefor the TRU waste. This can be explained by the sorption coefficient of Se-79 for the
Neogene sediments, which is about 40 times higher for the low salinity case for plutonic
rocks and about 50 times higher for the high salinity case. For TRU waste, similar sorption
distribution coefficients are set, but modelled migration distances from the disposal vaults areconservatively set to be much shorter than from HLW panels, as explained previously, whichmeans that Se-79 from the TRU waste is not as well retarded as that originating from HLW.
In the case of the plutonic rocks, Sr-90 in Gr.3 and other waste packages is the dominant
RN in the initial very small dose to the whole repository for any model water chemistry in
Waste Package A. For the Neogene sediments, the host rock Kd of Sr-90 is significantly
higher than the plutonic case (25 times higher in low salinity groundwater and 8.75 times
higher in high salinity groundwater) and its half-life is relatively short (29 years), so it decays
to insignificance during migration. The reason why C-14 does not contribute to the initial
dose of the entire repository for plutonic rocks is that the RN migration distance at therepository-scale is long enough to allow significant retardation caused by matrix diffusion.
(c) Pre-Neogene sediments (see Supporting Report 6-26)
Figure 6.4-25 shows the doses resulting from HLW and TRU waste in Pre-Neogene
sediments and the main RNs contributing to them.
For low salinity water, by comparing Figures 6.4-22 and 6.4-25, it is clear that again thedose is dominated by TRU for the entire evaluation period. The dominant RN in the TRU
waste is I-129 derived from Gr.1, giving a maximum dose of about 0.2 μSv/y (after 40 ky).
The large impact of I-129 is again due to use of a conservative model of instantaneous release
unconstrained by solubility limits after package failure and, as a stable anion, assigned zero
sorption and lack of significant radioactive decay due to its very long half-life.
For HLW, the breakthrough is different from the plutonic rock case but very similar to the
Neogene case apart from a significantly longer release tail. This is because, as described in
Section 6.4.1 (2) (v) (b) 3., for Neogene sediments, the multi-channel model of the host rockassumes greater flow wetted surface and matrix diffusion depth, causing much greater
dispersion of the release plume.

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HLW
Low salinity

High salinity

TRU Waste
High salinity

Waste package B

Waste package A

Low salinity

Figure 6.4-25 Doses and dominant RNs for Pre-Neogene repository (base case)

For the high salinity case, as for the Neogene sediments, up to about 40 ky after closure,
TRU waste dominates releases. However, unlike the Neogene case, there is no early
breakthrough of C-14 from Gr.4 and so I-129 from Gr.1 dominates throughout this time, since
the modelled repository scale migration distance is much longer in the Pre-Neogene case.
Thereafter, however, the situation is very similar to the Neogene case, apart from a longer
release tail as explained above, with the maximum dose from the entire repository after about
130 ky (0.9 μSv/y), predominantly due to U-233 from HLW. This is again due to the highcarbonate concentration of this groundwater (47 mM). The explanation of the roles of decay
chain daughters is also the same here.
As for the Neogene sediments, the dose contribution from Se-79 has become negligibly
small for HLW. This can be explained by the sorption coefficient of Se-79 for the Pre6-136

Neogene sediments which is about 40 times higher than plutonic rock for the low salinity caseand about 50 times higher for the high salinity case. In contrast to the Neogene sediments Se79, is also negligible for the TRU waste since the modelled migration distance is much longer.
The RNs that control the initial breakthrough for TRU waste package A are Sr-90 and C14 for plutonic rock and Neogene sediments, respectively. However, neither of them issignificant for the Pre-Neogene sediments; only slightly earlier breakthrough of I-129. The
fact that C-14 does not dominate for the Pre-Neogene sediments is due to the much longer
modelled migration distance, allowing for decay of this RN. The insignificant contribution of
Sr-90 is mainly due to a combination of pathlength and a sorption distribution coefficient set
to be 25 times larger than that of plutonic rocks.
(iii) Summary of the base case results
In all base cases, the maximum value of the calculated radiation doses occurs within about
100 ky and lies below the target of 10 μSv/y (see Supporting Reports 6-24, 6-25, 6-26).
As noted, the models used are highly conservative in terms of their treatment of migration
distances through the geosphere and the selection of key parameters impacting RN releaseand migration (solubility limits and sorption distribution coefficients). For TRU waste,assumed instantaneous release and the mixing tank model of the EBS introduces further
conservatisms. The high salinity water chemistry assumed for the Neogene and Pre-Neogene
sediments includes very high carbonate concentrations that are reflected in the modelled
extremely high mobility of U.
For HLW, there is almost no difference between calculated doses for the H12V and PEM
options, with dose maximum in the low salinity case < 0.1 μSv/y for plutonic rock and < 0.01
μSv/y for the sediments; these are significantly less than doses from TRU waste. Doses weregreater for higher salinity water in the sediments, giving a maximam ≈ 1 μSv/y due to higher
U mobility; in these cases, exceeding the TRU waste doses.
The RN with greatest contribution to the maximum dose in plutonic rocks is I-129,
predominantly from TRU waste. Overall repository performance is little impacted bygroundwater salinity or TRU waste package type, although there is a noticeable early
breakthrough of low levels of Sr-90 and I-129 for waste package A. Although this is mostly
the result of overly conservative assumptions as noted previously.
For the Neogene and Pre-Neogene sediments, I-129 from TRU waste is also the dominant
RN for the low salinity groundwater. In the case of high salinity groundwater, U-233(produced from decay of Np-237) from HLW dominates because of assumed greater mobility
as a result of the high carbonate concentration. This RN peaks significantly above I-129 from
TRU waste, which is insensitive to salinity. In addition, since Pb-210 was assigned zero
sorption, it also has a relatively large contribution to the dose from HLW. Although again
maximum doses are not altered by TRU waste package type, minor early releases from waste
package A are noticeable – predominantly C-14 in the Neogene case and I-129 for PreNeogene sediments.
It is clear that, in order to meet the goal of more realistic modelling to support comparisonof different site/repository design options, major improvement is needed to reduce model
simplifications and conservative/over-conservative data used to allow for system uncertainties.
This is discussed further in Chapter 7.

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6.4.2 Dose evaluation for variant scenarios(1) Radionuclide migration model and dataset
Here, the analysis conditions and the corresponding models and databases for evaluating
the dose based on the RN migration analysis will be described for the variant scenarios. The
specification of these scenarios for the three host rocks are as presented in Tables 6.3-8 and6.3-9.
For the variant scenarios, the dose constraint is 300 μSv/y, as described in Section 6.1.5 (2).
However, a careful assessment of whether the likelihood of these scenarios are sufficiently
low to warrant use of this constraint has not yet been done. The dose evaluation results ofthese cases are discussed by extracting the maximum value and time of occurrence from thecalculated dose curve for each, in order to show differences from the appropriate base case. In
the figures, analysis cases for low and high salinity groundwater are marked with the letters L
and H, respectively. In addition, in order to associate each analysis case with the results in the
figure, they are numbered as shown in Table 6.4-7. Background for setting conditions of RN
release and migration modelling and selection of required parameters for each analysis case issummarised in Table 6.3-10.
As mentioned in Section 6.4.1 (5) for the base case, doses for the variant scenarios arecalculated from use of the same simple dose conversion factors. Furthermore, for HLW, there
was no significant difference in the handling of H12V and the PEM analyses, even taking into
account differences in the geometry, since the calculation results are almost the same
(Supporting Reports 6-24, 6-25, see 6-26).
For analysis case No. 2, the variant case is considered only for HLW, and for analysis
cases Nos. 3 to 5, the variant case is considered only for TRU waste. In calculating the dose
due to the whole repository in these analysis cases, the dose in the variant case is summed
with the dose in the base case, where the variant case is not taken into account. For example,
in the case of analysis No. 2, the dose for the whole repository is calculated using the dose in
the variant case for the HLW repository and the dose in the base case for the TRU repository.

6-138

Table 6.4-7 Analysis cases and corresponding numbers
Name of analysis case(see Table 6.3-10)No.
1

Base case

Analysis conditions (model and dataset)
Section 6.4.1 (2) to (6).

Variant case2

Increase in glass dissolution case3

Increased corrosion of TRU Gr.2 case4

Uncertainty in structural frame degradation case

Ten times the glass dissolution rate of the base case (see SR 6-13).
Increased corrosion rate relative to the base case for zircaloy hulls (x9) and stainless-steel end pieces (x 14) (SR 613).
The hydraulic conductivity of the structural framework is set that of sand after repository closure.

5

Nitrate plume impact case

The impact of nitrate from TRU Gr.3 is defined asPlutonic rocks: For the TRU waste repository, the nitrate plume impacts RN migration at near field scales
and repository scales.

Neogene sediments: For TRU waste Grs.1, 2 and 4LD and HLW, it is assumed that the nitrate plume
affects the host rock at the repository scale.
Pre-Neogene sediments: For the TRU waste repository, the nitrate plume impacts RN migration at nearfield scales and repository scales.

6

Fracture connectivity in host rock case

Assuming higher facture network connectivity (see Appendix Figures 4 to 7).

7

Lower sorption in buffer caseSet sorption distribution coefficient values lower than the average value (see SRs 6-19 and 6-20, Appendix Tables
15 to 17).

8

Increased diffusivity in buffer caseSet effective diffusion coefficient values higher than the average value (see SR 6-19 and Appendix Tables 12 to 14).

9Lower sorption in host rock caseSet sorption distribution coefficient values lower than the average value (see SR 6-22 and Appendix Table 21 to 23).

Increased diffusivity in host rock caseSet effective diffusion coefficient, values lower than the average value (see SR 6-20 and Appendix Tables 18 to 20)

10
11

Thermal increase in solubility case

12

Uncertainty in thermodynamic data caseSet solubility limit 100 times greater than the base case (see SR 6-17 and Appendix Tables 9 to 11)
Set higher solubility limits based on thermodynamic calculations considering other limiting solid phases (see SR617 and Appendix Tables 9 to 11)

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(2) Plutonic rock (see Supporting Report 6-24)
The results of calculated total repository dose maxima for the variant cases listed in Table
6.4-7 for plutonic rock are summarised in Figure 6.4-26.
High salinity

Waste package B

Waste package A

Low salinity

Figure 6.4-26 Total repository maximum dose and time of occurrence calculated for plutonic
rocks (Note that case No. 4 is not applicable for package B)

These are presented separately for the two groundwaters and TRU waste package options
considered.
It is clear that, in all cases, dose maximum is lower than the constraint value of 300 μSv/y.
Regardless of differences in groundwater chemistry or TRU waste packaging, the maximum
dose results from I-129 from TRU waste Gr.1 with no significant difference from the basecase (1), with the single exception of case No. 6 (Fracture connectivity in host rock).
For case No. 6 and TRU waste package A, the maximum dose appears 100 y after closure
and its value is about 7 μSv/y, predominantly due to Sr-90 derived from TRU waste Gr.3.
In this case, the transmissivities in multi-channel model are increased by an order ofmagnitude compared to the base case, resulting in higher groundwater flow velocity. For Sr90, nitrate contained in Gr.3 results in a low sorption distribution coefficient for the host rockthat, when combined with a very high initial inventory and the high groundwater flow rate
gives the high and early maximum dose. As described in Section 6.4.1 (7) (ii) (a), in the RN
migration analysis for Gr.3, an excessively conservative RN migration analysis model, in
which the RN is instantaneously eluted from the waste, is used. A more realistic assessmentof when waste could realistically contact groundwater and accounting for slow leaching of thebitumen matrix would certainly remove this artefact.
In the case of waste package B, for case No. 6, the maximum dose of 5 μSv/y is reachedabout 1 ky after closure, with I-129 from Gr.1 the dominant RN. The peak appears earlier
than the base case and is slightly higher due to fast transport and less dispersion of this non-

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sorbing RN. Here, there is no significant contribution from Sr-90, due to decay of 10 halflives (a factor of over 1000) before loss of containment at 300 y.
Figure 6.4-27 shows the maximum value of the dose and the timing of its occurrence
separately for HLW and TRU waste. Variant cases related to HLW are No. 2 and Nos. 6 to 12,
as listed in the Table 6.4-7. For the case of TRU waste, analysis cases Nos. 3 to 12 are
relevant.
Although the maximum dose of the HLW variant cases has little impact on the total dosefrom the entire repository, these can be compared with the base case to show that that there
are impacts in cases No. 2 “glass dissolution rate uncertainty”, No. 6 “Fracture connectivity inhost rock” and No. 9 “Lower sorption in host rock”. Although case No. 2 gives an earlier peakarrival time, increases in dose maximum are larger for cases Nos. 6 and 9 but, at most, only
about an order of magnitude.
The RN dominating the maximum dose in the base case was Se-79 (see Section 6.4.1 (7)(ii) (a)), but the maximum dose for case No. 2 results from I-129. This is because the glassdissolution rate is set to 10 times the base case and I-129 is treated without a solubility limit.
In contrast, since Se-79 concentration in the source term is controlled by solubility, the release
rate from the EBS is limited even when the glass dissolution rate increases.
Analysis cases No. 6 and No. 9 both show profiles of RN release similar to the base case
(Figure 6.4-22), with Se-79 dominating the dose peak.

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HLW
Low salinity

High salinity

TRU Waste
High salinity

Waste package B

Waste package A

Low salinity

Figure 6.4-27 Separated HLW and TRU waste maximum dose and time of occurrence calculated
for plutonic rocks (Note that case No. 4 is not applicable for package B)

For analysis case No. 6, the hydraulic conductivity of the host rock multi-channel model isincreased by an order of magnitude compared to the base case (see Table 6.3-10, Appendix
Figures 1 and 4), which implies that the transfer rate of RNs to the biosphere increases. As aresult, the dose is higher than in the base case, but this effect works the same way for all the
RNs. Therefore Se-79 will continue to dominate the maximum dose. Also, in the variant caseNo. 9, as shown in Appendix Table 20, the sorption distribution coefficient of Se-79 on hostrock is reduced, and Se-79 becomes even more dominant in terms of the maximum dose.
Based on the above analysis, the only uncertainties that impact overall performance for
HLW are associated with the long-term dissolution rate of the glass and sorption/fracture
network characteristics. Even though the variant case is very pessimistic, bettercharacterisation of glass dissolution will increase safety case robustness as this directly scaleswith the release from the EBS for the most mobile, highly soluble RNs that dominate dose.

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Host rock characteristics will be very site specific but, given the key role of RNs such as Se79, it is important to continue to improve the knowledge base on the release and transport
properties of RNs that are determined to significantly contribute to assessed doses.
For TRU waste, the maximum dose and the timing of its occurrence for both waters andpackage types are little affected by any of the variants except No. 6, as discussed above forthe total repository dose. Also discussed above are the impacts of the 2 different waste
packages for this variant.
For the TRU waste, the variant cases demonstrate the importance of RN release andmigration model assumptions and parameters for the EBS and geosphere, together with theimpacts of perturbations – especially that from the nitrate plume. There is thus a clear need tomake the repository migration model more realistic in order to assess what issues identified(e.g. role of longer-lived waste packages on containment of high inventory/short half-lifeRNs) are real (and should be a focus for future R&D) and which are artefacts of current model
simplifications.
(3) Neogene sediments (see Supporting Report 6-25)
The results of calculated total repository dose maxima for the variant cases listed in Table
6.4-7 for Neogene sediments are summarised in Figure 6.4-28. These are presented separately
for the two waters and TRU waste package options considered. It is clear that, in all cases,
dose maximum is lower than the constraint value of 300 μSv/y.

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High salinity

Waste package B

Waste package A

Low salinity

Figure 6.4-28 Total repository maximum dose and time of occurrence calculated for Neogene
sediments (Note that case No. 4 is not applicable for package B)

For both groundwaters, the maximum dose and its appearance time for the two TRU waste
packages are almost identical.
In the case of low salinity groundwater, as for plutonic rock, only variant case No. 6 causesa significant divergence from the base case, resulting in a higher dose peak (by about an orderof magnitude) at an earlier time (about 10 ky). Again, this is derived from I-129 in Gr.1. As
for the plutonic case, No. 6 increases the transmissivities in the multi-channel model by anorder of magnitude compared to the base case (Table 6.3-10 and Appendix Tables 2, 5 and 7),
resulting in a higher groundwater flow velocity and thus earlier breakthrough of I-129 withless dispersion.
In the case of high salinity groundwater, variant cases No. 6 and No. 10 (increased
diffusivity in host rock) both cause slight deviations from the base case. It can be noted thatthe effect of the higher hydraulic conductivity of the host rock is smaller than the low salinity
case, where I-129 is the dominant RN.
For variant No. 6, the maximum dose is derived from the HLW repository, as for the basecase (see Section 6.4.1 (7) (ii) (b)). This is also due to U-233, which has a maximum doseabout 2 times higher than that for the base case. As explained previously, the reason why U233 dominates the dose over long times for the base case is explained by continuous ingrowthfrom its parent Np-237, which will remain in the buffer and rock for long timescales due to its
low solubility and high sorption.
For variant case No. 10, Pb-210 derived from TRU Gr.4 is the dominating RN. The
reasons for this RN having such a large contribution were discussed previously, with thelower rock diffusivity assumed for variant case No. 10 resulting in a dose contribution about

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two times higher than for the base case, which then slightly exceeds that of U-234. This is dueto the importance of matrix diffusion for retarding non-sorbing RNs, which is reduced in thisvariant.
Figure 6.4-29 shows the maximum value of the dose and the timing of its occurrence
separately for HLW and TRU waste. Again, variant cases related to HLW are No. 2 and Nos.
6 to 12 as listed in the Table 6.4-7. For the case of TRU waste, analysis cases Nos. 3 to 12 are
relevant.
For HLW, as for plutonic rock, cases No. 2 and No. 6 show some divergence from the basecase in low salinity water, although here the impacts on both arrival time and maximum doseare less. The RN that dominates the maximum dose is I-129, as in the base case, and the
explanation is as for plutonic rock. For high salinity groundwater, only variant case No. 6
deviates from the base case, but only to a small extent, with both the dominant RN (U-233)
and its explanation being the same. Plutonic rock comments on uncertainty reduction thus
also apply here.
For TRU waste, in low salinity groundwater, the variants do not deviate from the base case
except for analysis case No. 6. In terms of the dominant species, this is I-129 in all cases. The
impact of increased conductivity on migration of this non-sorbing RN is as discussed above.
For high salinity conditions, variants No. 6 and 10 show significant differences in maximum
dose and time of occurrence, with very minor increases in maximum does also for variants No.
11 “Thermal increase in solubility case” and No. 12 “Uncertainty in thermodynamic datacase”. For variant No 6, this is directly equivalent to the low salinity case. For variant case No.
10, the peak is due to Pb-210 results from Gr.4 as discussed for the entire repository above.

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HLW
Low salinity

High salinity

TRU Waste
High salinity

Waste package B

Waste package A

Low salinity

Figure 6.4-29 Separated HLW and TRU waste maximum dose and time of occurrence calculated
for Neogene sediments (Note that case 4 is not applicable for package B)

Regarding variants No. 10 and 11, the dominant RNs are I-129 and Pb-210, with thecontribution of the latter increased sufficiently to give a notably higher total dose. The
increased dose from Pb-210 results from increased migration rate of its parent Th-230, due tohigher solubility and decreased sorption.
In terms of reducing impacts of uncertainties, in addition to the points noted fromassessment of the TRU variants for plutonic rock, the unexpected domination of releases of
Pb-210 from TRU would merit R&D to support models and databases required for moredetailed assessment of the treatment of migration of the actinide decay chains, especially forwaters with high carbonate concentrations.

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(4) Pre-Neogene sediments (see Supporting Report 6-26)
The results of calculated total repository dose maxima for the variant cases listed in Table
6.4-7 for Pre-Neogene sediments are summarised in Figure 6.4-30. These are presented
separately for the two waters and TRU waste package options considered. It is clear that, inall cases, dose maximum is lower than the constraint value of 300 μSv/y.
High salinity

Waste package B

Waste package A

Low salinity

Figure 6.4-30 Total repository maximum dose and time of occurrence calculated for PreNeogene sediments (Note that case 4 is not applicable for package B)

Basically, it can be seen that the pattern is very similar to the Neogene host rock, withvariant case No. 6 causing the only significant reduction of barrier performance for bothwaters and TRU package options for the same reasons. The extent of deviation is, however,less for the Pre-Neogene sediments and, for high salinity groundwater, the impact of variantcase No. 10 is negligible.
Figure 6.4-31 shows the maximum value of the dose and the timing of its occurrence
separately for HLW and TRU waste. Again, variant cases related to HLW are No. 2 and Nos.
6 to 12 as listed in the Table 6.4-7. For the case of TRU waste, analysis cases Nos. 3 to 12 are
relevant.
For HLW, only analysis case No. 2 shows a significant impact for low salinity water,
decreasing the peak arrival time but having little influence on peak dose. Case No 6 shows an
influence for both waters, but it is extremely small. In all cases, the dominant RNs are as forthe base case: I-129 in low salinity water and U-233 for high salinities with the same
justification as discussed for Neogene sediments above and the same issues for reducing
uncertainties for HLW as identified for the other two rocks.

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HLW
Low salinity

High salinity

TRU
High salinity

Waste package B

Waste package A

Low salinity

Figure 6.4-31 Separated HLW and TRU waste maximum dose and time of occurrence calculatedfor Pre-Neogene sediments (Note that case No. 4 is not applicable for package B)

For both ground waters, the maximum dose and its appearance time for the two TRU waste
packages are identical. As for the Neogene sediments, case No. 6 gives rise to earlier
breakthrough of a higher dose maximum, but the degree of change is significantly less. The
transmissivity distribution is not as wide compared with other rock types, which also meansthat the transmissivity increase for case No. 6 was not as large. Unlike the Neogene sediments,
however, variant case No. 10 has no significant impact and hence it is more similar to theplutonic case. Issues for reducing uncertainties for TRU waste are basically as for plutonic
rock.

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(5) Summary of dose evaluation results for variant cases
For all variant cases studied, the maximum doses calculated for the total repository withboth TRU waste package options and both groundwaters lie below the 300 μSv/y target.
Indeed, the value of the maximum dose for variant cases did not change significantly from thevalue for the relevant base case (Section 6.4.1 (7) (ii)).
This lack of sensitivity to more conservative model assumptions and data used certainlyreflects a combination of both the very simplified models that are used for the base case andthe generally conservative approach to data selection there. Of particular note here are themodels of instantaneous release into a mixing tank for the TRU EBS and selection of nosolubility limits and zero sorption for RNs in the base case, where there are considered to belarge gaps in laboratory data and/or supporting mechanistic models.
Despite such inherent constraints, evaluation of the variant cases enable to identify someareas where further work can be focused to reduce uncertainties and add robustness to thesafety case, such as improving the knowledge base to support low dissolution rates of HLW
glass; extending studies of releases from TRU waste matrices and retardation by EBS
components to support more realistic source term models; and improving understanding of
RN geochemistry, especially in the case of evolving perturbations from high pH cement
leachate or nitrate, or extreme groundwaters with high carbonate concentrations. It is clearthat the geological barrier performance is sensitive to the characteristics of the flow system, as
highlighted for the cases of increased DFN connectivity or reduced diffusivity into the rockmatrix. Apart from appropriate site-specific characterisation, the capabilities of RN transportmodels will need to be improved in order to capture the key features of real rocks, such as
spatial variability. Finally, it was noted that the output of models of transport of actinide
decay chains are difficult to interpret and hence improvement here should include not only
better representation of the processes involved, but also improved transparency of theirimpacts to allow these to be tested more rigorously. These issues are discussed further in
Chapter 7.
6.4.3 Evaluation of low probability perturbation scenarios
In this section, the concepts supporting evaluation models and parameters and the dose
evaluation results for analysis cases for the low probability perturbation scenarios noted inTable 6.3-11 are outlined. A large uncertainty is involved in setting the state of the repository
for such scenarios and hence it is stylised, with a focus on clarity of evaluation and assured
conservatism - without consideration of how physically realistic it might be. This enables
investigation of the robustness of the repository system and the dose obtained by the analysis
performed may be drastically over-conservative. Therefore, results of the analysis cases
should be compared to dose targets for such scenarios only with great care.

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(1) New volcanism scenarios(i) Concept of evaluation model
As described in Section 6.3.2 (3) (iii), in this analysis case, new volcanoes are assumed todevelop on the back side of the volcanic front around 100 ky in the future. The stylised model
assumes associated magma directly impacts the repository, as illustrated in Figure 6.4-32. It is
understood that there could be other consequences from volcanoes, such as impacts due togeothermal water entering the repository, but these have not been analysed in this safety case.

Figure 6.4-32 Conceptual model of new volcanism caseIn this model, the following three points are assumed:For an appropriate site, 100 ky after closure is the minimum time for development of
a new volcanic centre (for a site selected on the basis that no indicators of such a risk
were seen): at this time, it is assumed that a conduit transporting magma from a deep
chamber to a vent penetrates the repository footprint.
RNs present in the region of this conduit (conduit area fraction) are taken up into themagma and released to the surface as ejecta (such as volcanic ash), which is thendeposited in the vicinity.
The amount of radioactivity incorporated into magma is defined by the inventory of
RNs in the repository at that time, taking into account decay/ingrowth, butconservatively ignoring mobilisation by groundwater flow within the intervening 100
ky period.

Based on these assumptions, dose is evaluated as follows:

Farmers are set as representative persons living in the fallout area; they cultivate soilthat is homogeneously mixed with radioactivity falling out with volcanic ash. Dose is

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

evaluated considering external exposure from soil, inhalation exposure of dust fromsoil, and oral ingestion exposure from eating local crops. Here, it is considered thatmainly root vegetables are cultivated in the volcanic ash soil.
For this evaluation model, differences in repository layout as a function of the hostrock (as used in the evaluation of the base and variant scenarios) are taken into
account.

(ii) Setting evaluation parameters
The following describes the setting of the required parameters for the model outlined
above.
(a) Radioactive inventory in the repository
Assuming the occurrence of new volcano at 100 ky after closure, changes to the inventoryfor HLW and TRU waste due only to radioactive decay/ingrowth are considered, as given in
Table 6.4-8.

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Table 6.4-8 Radioactivity inventory used for new volcanism cases
NuclideHLW (Bq)

NuclideHLW (Bq)

NuclideTRU waste
(Bq)

C-14
Cl-36
Se-79
Sr-90
Zr-93
Nb-93m
Nb-94
Tc-99
Sn-126
I-129
Cs-135
Cs-137
Pb-210Ra-226
Ra-228
Ac-227
Th-228
Th-229Th-230Th-232
Pa-231
U-232
U-233
U-234
U-235
U-236
U-238
Np-236
Np-237
Pu-236
Pu-238
Pu-239Pu-240Pu-241
Pu-242
Pu-244Am-241
Am-242m
Am-243

2.6 x 107
1.5 x 1013
1.0 x 1014
0.0
2.8 x 1015
2.7 x 1015
2.0 x 1011
1.5 x 1016
3.4 x 1014
1.5 x 1012
7.1 x 1014
0.0
8.0 x 1012
8.0 x 1012
2.8 x 107
3.9 x 1011
7.8 x 108
2.7 x 1014
8.1 x 1012
2.8 x 107
3.9 x 1011
7.5 x 108
3.0 x 1014
1.2 x 1013
5.2 x 1011
5.9 x 1012
1.6 x 1012
5.9 x 109
8.3 x 1014
7.5 x 108
0.0
9.6 x 1014
3.8 x 1011
1.9 x 1011
1.6 x 1013
6.4 x 106
2.0 x 1011
0.0
2.7 x 1012

Cm-243
Cm-244
Cm-245
Cm-246
Cm-247
Cm-2480.0
0.0
1.9 x 1011
5.2 x 107
4.3 x 108
1.1 x 109

C-14
Cl-36
Co-60
Ni-59
Ni-63
Se-79
Sr-90
Zr-93
Nb-94
Mo-93
Tc-99
Pd-107
Sn-126
I-129
Cs-135
Cs-137
Pb-210Ra-226
Ra-228
Ac-227
Th-228
Th-229Th-230Th-232
Pa-231
U-233
U-234
U-236
U-238
Np-237
Pu-239Pu-240Pu-241
Pu-242
Am-241
Am-243
Cm-244
Cm-245
Cm-246

3.7 x 109
7.5 x 1012
0.0
3.8 x 1015
0.0
4.3 x 1012
0.0
3.7 x 1014
8.8 x 1013
2.3 x 106
5.8 x 1014
1.1 x 1012
5.7 x 1012
5.9 x 1013
5.8 x 1012
0.0
8.3 x 1012
8.3 x 1012
1.3 x 107
1.9 x 1011
1.3 x 107
4.7 x 1012
8.4 x 1012
1.3 x 107
1.9 x 1011
5.2 x 1012
1.2 x 1013
2.7 x 1012
1.3 x 1012
1.4 x 1013
2.2 x 1014
1.3 x 1011
8.0 x 108
1.5 x 1013
8.4 x 108
1.7 x 1011
0.0
8.0 x 1080.0(b) Spatial considerations
Based on the radioactivity inventory shown in Table 6.4-8, the dose was calculated for the
H12V repository in plutonic rocks, as this is relatively compact. The release of RNs into thebiosphere is calculated from the inventory in the area of the HLW repository or intersected by
the conduit divided by the volume of ejecta. This gives the radioactivity concentration of the
RN in the ash that reaches the surface. To set the area of the volcanic conduit and the volume

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of the volcanic eruption, reference eruptions selected in Supporting Report 3-34 were used(see Table 6.4-9).
Table 6.4-9 Conduit areas and ejecta volumes used for evaluation
Name
Mikugawa caldera
Komochi-yama
Kamitakara
Unzen-dakeConduit area
(km2)
0.2
7 x 10-2
4
2 x 10-2

Ejecta volume(km3)
1 x 10210400.2(c) Biosphere evaluation parameters
External exposure, inhalation exposure, and oral ingestion exposure are assessed as thereference farmer`s exposure routes. Stylised biological and lifestyle characteristics [96] wereused to calculate dose due to external exposure and inhalation of dust. Furthermore, assumingthat root vegetables were cultivated in this soil, doses from internal exposure resulting fromconsumption of local crops [97] were calculated (see Supporting Report 6-27 for details).
(iii) Results and discussion
The evaluation of the worst case stylised new volcano scenario resulted in a dose of 0.09
mSv/y which is well below the dose target (20-100 mSv) for the first year and also the dosetarget thereafter (1-20 mS/y), as specified for low probability scenarios.
According to the ITM-TOPAZ method [33] [34] [35] [36], as described in Chapter 3, theprobability of such a new volcano generation can be derived based on expert judgment.
According to this approach, the probability of occurrence greatly depends on regional features,but if sites within 15 km radius from Quaternary volcano centres are avoided, in the region of
Tohoku the probability is calculated to be ~ 10-6/y/100 km2 (Supporting Report 3-34). Using
this, the associated risk for the above dose (see Section 6.1.5 (2)) and the repository footprint
is about 1 x 10-12/y, which is far below the internationally recommended risk constraint value
10-5/y.
(2) Fault extension scenarios
In this analysis case, it is assumed that, over a long period of time, a major fault graduallyextends until it impacts the repository and forms a short circuit from the near field to thebiosphere. Since the impact of such a migration path depends on the layout of the repository
and the geological environment, the analysis considers both model waters defined for plutonic
rocks, Neogene and Pre-Neogene sediments. In addition, the H12V option is assumed for the
HLW repository, with the conservative assumption that the entire RN inventory remains in
the EBS, as for the new volcanism case. See Supporting Report 6-28 for details of models andparameters.

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(i) Conceptual model

Figure 6.4-33 Concept of fault extension scenario

The scenario concept to be captured by the evaluation model is shown in Figure 6.4-33.
In this model, the following assumptions are made:









Faulting starts at great depth (6 to 20 km) and then gradually extends to the surface.
An extension of this major fault directly hits a disposal panel at a certain time after
repository closure.
The fault zone and associated process zone gradually increase in size as the fault
repeatedly activates (see Supporting Report 3-35); so these do not develop in one fault
movement event [66]. In the evaluation, however, an extreme condition in which thefault zone and the process zone result from a single event is assumed.
For parts of the repository located in the fault plane/fracture zone, all safety functionsof the EBS and the geological environment are lost. For parts located in the process
zone, the groundwater flow velocity of the rock rises, so the safety function of
reducing RN migration is degraded.
All RNs eluted in groundwater from waste within the fault zone move to the surface
environment via flow in the fault.
The RN migration path to biosphere ends with releases to the GBI equivalent to thatfor the base case scenarios (see Section 6.4.1 (6)). Assuming that a breccia zone withstructures typical of existing faults is present (see Supporting Report 6-28), a RN
migration model with faults handled as a porous medium in which sorption of RNs
occurs is utilised, as for the regional scale base case (see Section 6.4.1 (5)).

(ii) Setting parameters(a) When the event occurs
It is expected to take a very long time until a fault extension directly hits the repository forthe given siting criteria, but it is difficult to specify the minimum time needed. Considering
the great uncertainty regarding the timing when such a scenario occurs, as for H12 possible
times 1 ky, 10 ky and 100 ky after closure are considered.

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(b) Characteristics of the fault extension
With regard to the size of the fault that reaches the repository and affects its safetyfunctions, the depth of the fault is set to 6 to 20 km as noted above. Assuming that an
earthquake occurs from this seismogenic layer, it was decided to target a 20 km scale fault
that could conservatively generate earthquakes up to around M7.0.
In accordance with Supporting Report 3-35, a fault zone having a width of 100 m on eachside of the fault plane and a process zone with a width of 220 m outside the fault zone was set
for the calculations.
Depending on the location and shape of the fault, the number of waste packages impacted
will be different. Since it is not possible to predict how this fault hits the repository withoutsite-specific information, the resultant dose from HLW and TRU waste is calculated byextremely conservative, stylised scenarios as shown in Table 6.4-10.
Table 6.4-10 Stylised fault representation
Waste type

HLW

TRU

Assumptions


It is assumed that a linear fault occurs at a position where it intersects the largest
number of waste packages for each panel of the reference repository (seeSupporting Report 6-28). Table 6.4-11 shows the number of packages impacted
by the fault plane, fault zone and process zone.

The dose is first calculated for intersection with one panel and then dose fromthe entire HLW repository is obtained by multiplying this dose by the total
number of panels. For the Pre-Neogene repository, since the number of tunnels isdifferent for each panel, the dose is calculated for the panel with the greatest
number of waste packages, and this dose is multiplied by the number of panels.

Because it is difficult to define how the fault hits the repository, it isconservatively estimated that all TRU waste vaults are directly hit by the fault.

Table 6.4-11 Impact of defined fault on HLW disposal layouts
Host rock

Impact per panel (number of waste
packages)
Fault plane

Fault zone

Process zone

Number ofpanelsPlutonic

139

2780

3750

6

Neogene

131

2100

4450

6

Pre-Neogene

250

3000

1750

8

(c) Characteristics of RN migration pathways
As outlined in section (i), and further detailed in Supporting Report 6-28, the modelling of
the EBS is different for different parts of the fault, as indicated in Figure 6.4-33. For waste
packages located in the fault plane, all safety functions of the EBS are lost and all RNs in thewaste are assumed to be instantaneously released into the fault. Waste packages located in thefault zone are assumed to lose the buffer safety functions but, for HLW, slow release from theglass matrix would remain. The EBS is assumed to be intact in the fault process zone, but the
groundwater flow velocity of the EDZ increases.

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The parameters set for the characteristics of the stylised fault RN migration path are shownin Table 6.4-12 for the three representative rock types.
Table 6.4-12 Parameters for RN migration analysis in the fault
Parameters
Hydraulic gradient

Plutonic rocks
0.05

Neogene sediments
0.06

Pre-Neogene sediments
0.05

1.0 x 10-5

Hydraulic conductivity (m/s)
0.1

Porosity

0.3

Rock density (kg/m3)

0.5

2700

Path length (m)

700

200

700

Dispersion length (m)

70

20

70

Appendix Table 21

Appendix Table 22

Appendix Table 23

Sorption distribution coefficient(m3/kg)

The migration distance is based on the repository depth, hydraulic gradients are the sameas the base case and fault hydraulic conductivity is conservatively set to the equivalent ofsand.
(iii) Results and discussion
The results of the dose evaluation are shown in Table 6.4-13 for each model water, thethree representative host rocks and the different assumed times at which faulting occurs.
Despite the very conservative assumptions described above, the sum of the doses from HLW
and TRU waste is a maximum of 14 mSv/y for the Neogene sediments (for both groundwatertypes).
Table 6.4-13 Maximum doses calculated for HLW and TRU waste fault extension analysis cases(for different occurrence times)

Host rock

Model
water
(salinity)

Maximum dose (mSv/y) and dominant RNs
HLW

TRU waste

3

10 y

4

10 y

5

10 y

3

10 y

104 y

105 y

Low

0.3
Pu-2390.2
Pu-2390.1
Np-237

4
I-129

4
I-129

4
I-129

High

0.3
Pu-2390.2
Pu-2390.1
Np-237

4
I-129

4
I-129

4
I-129

Low

0.1
Tc-99

0.1
Tc-99

0.08
Tc-99

14
I-129

14
I-129

14
I-129

High

0.4
Tc-99

0.4
Tc-99

0.7
U-233

14
I-129

14
I-129

14
I-129

Low

0.1
Tc-99

0.1
Tc-99

0.07
Tc-99

4
I-129

4
I-129

4
I-129

High

0.9
Tc-99

1.0
Tc-99

2
U-233

4
I-129

4
I-129

4
I-129

Plutonic

Neogene

Pre-Neogene

In all cases the estimated dose is within the target range of 1 to 20 mSv/y for the low
probability perturbation scenarios from the second year after the occurrence. For all three hostrocks, the TRU waste dose is higher than that from HLW.

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For HLW, the only EBS function is congruent dissolution from the glass matrix for
packages located on the fault plane. With this release, the dominating RN changes for thedifferent rocks and groundwater chemistry due to the differences in sorption distribution
coefficients for the fault and the timing of event occurrence (due to radioactive decay andingrowth). For TRU waste, the doses are dominated in all cases by I-129, for the same reasons
as for the base case.
These results reflect the assumption that all RNs are confined in the waste until the fault
directly hits, with elution of RNs commencing immediately thereafter, which is probably
realistic for HLW up to at least 10 ky, but conservative for TRU waste. The RN migration
model and parameters used are as for the base case, with the same limitations anduncertainties as previously discussed.
For this scenario, a 20 km scale fault is assumed to have occurred in the area of therepository; the probability of such an event is ≈ 2.2 x 10-7/y (see Supporting report 3-35).
Converting the maximum dose of 14 mSv/y (for the Neogene sediments) into a radiologicalrisk (see Section 6.1.5 (2)) and multiplying this by the probability of occurrence, the total risk
is calculated to be 2 x10-10/y, far below the risk constraint recommended by ICRP (10-5/y).
The dose calculation is considered to be extremely conservative and thus, even for such astylised scenario, supports the conclusion that the repository concept is robust in the case ofsuch a perturbation.
(3) Summary of results for low probability perturbation scenarios
For the low probability perturbations analysed, development of a new volcano or an
extended fault that directly impacts the repository, even very pessimistic scenarios indicatethat doses would not be unacceptably high and, when coupled to an assessment of likelihoodof occurrence, risks are extremely small. Thus, it is concluded that, by selecting sites anddesigning the repository appropriately, long term safety of the repository would not be losteven if such low probability events should occur

6.5 Evaluation of inadvertent human intrusion scenarios
6.5.1 Development of human intrusion scenarios
In assessment of the safety of the repository after closure, future human activities that
could impact barrier performance must be considered. The focus is entirely on actions that
inadvertently or accidentally impact the repository, as opposed to any intentional operation
specifically aimed at intruding into, or influencing the behaviour of, the repository,
corresponding to the international consensus and guidelines on such treatment[2][29][38][42][98].
In Japan, as elsewhere, the risk of unintentionally impacting a repository is reduced bylocating it below the depth at which human activities are common and siting it to avoid areas
where there are underground resources with economic value that could be a motive for humanactivities (for example, [98][99]). Further, restrictive measures with respect to activities suchas drilling will apply to the repository during the institutional control period after closure (e.g.,[27] [42]), while records will be maintained to assure that such restrictions continue

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indefinitely11 (e.g., [27][37]). Finally, there are arguments ([51][100]) that physical resistanceto intrusion of the repository itself can be expected. For example, in the case of strong
overpacks, these may resist intrusion by drilling until they are significantly degraded.
As also discussed in Section 6.1.5 (2) (iv), it may be argued that the probability that thesafety functions of the repository will be significantly impaired by human intrusion may beconsidered to be small, but also extremely difficult to assess in any rigorous manner.
Therefore, human intrusion scenario evaluations are considered in a similar manner to theevaluation of low probability perturbation scenarios, as discussed in Section 6.4.3: given
inherent uncertainties, stylised scenarios aim to illustrating the robustness of the repositoryand show that no significant radiological impacts are to be expected.
In terms of handling human intrusion scenarios, various arguments have been made ininternational organisations and regulatory and implementing agencies (e.g., [101][102][103]).
Through these discussions, it is clear that there is no strict scientific basis to predict humanactions and the likelihood of future human intrusion. Therefore, the approach of developing
stylised scenarios to confirm the robustness of the safety functions of the repository is widely
accepted.
In this report, the specific methodology for developing human intrusion scenarios was also
considered, with reference to the latest trends in Japan and abroad. Although human intrusion
scenarios depend on site conditions, since such scenarios are considered for repositories inmany countries, these are referred to directly to the extent possible.
(1) Setting requirements for building human intrusion scenarios (see Supporting
Report 6-29)
The IAEA [2][38], ICRP [29][42], OECD/NEA [103] and the Nuclear Safety Commission[40] [104] discuss requirements relating to the development of human intrusion scenarios as
follows:
1. Target actions that give direct disturbance to the repository.
2. Target actions that are expected to occur during the period after the repository is
closed.
3. Focus on inadvertent human intrusion (do not consider intentional intrusion).
4. Measures against inadvertent human intrusion12 should reduce its probability ofoccurrence.
5. Envisaged scenarios should be credible (plausible) and stylised based on current
techniques and procedures that could lead to intrusion.
(2) Selection of relevant human intrusion actions (see Supporting Report 6-30)
In view of the above requirements, the actions to be considered in human intrusion
scenarios are extracted from the NUMO FEP list (Supporting Report 6-4). As a result, “F1.4.5
Drilling activities” was selected whereas “F1.4.9 Water management (groundwater, surface

11

In Japan, it is prescribed that records are permanently preserved by the Minister of Economy, Trade andIndustry (Article 18, paragraph 2 of the Final Disposal Act).
12
Drilling restriction measures and record preservation, location signs, depth of construction more than 300 m,
avoiding economically important resources, etc.

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water)”13 and “F1.4.6 Mining and other underground activities” were discarded, as explainedin Supporting Report 6-30.
According to a report that collated information on drilling in Japan [105], objectives areclassified into five major areas: hot spring development; well development; academic
research; resource exploration; and construction work. Most of the drilling carried out to
depths of 300 m or more (minimum repository depth required by the Final Disposal Act) is
aimed at hot spring development and resource exploration.
According to the Hot Springs Act in Japan, hot springs are defined as “hot water from theground, mineral water, steam and other gases” which implies that springs containing a certain
amount or more of dissolved substances, regardless of temperature, are treated as hot springs.
Therefore, hot spring development can occur regardless of the regional setting. In terms ofresource exploration, as maps of economically valuable mineral resources are used to exclude
areas for consideration as potential sites, the likelihood of occurrence is considered to besmall compared to drilling for hot spring development, while the degree of disturbance can be
expected to be equivalent. From the above requirements 4 and 5, it was decided to evaluate
the drilling for the purpose of hot spring development as a plausible human intrusion scenario
in Japan.
(3) Stylised human intrusion considerations
In stylising the act of human intrusion, it is necessary to consider the timing of occurrenceof the event, the radioactive inventory in the repository at that time, the location of intrusion,and the route of transfer of RNs from the EBS to the biosphere. These are set so as to satisfy
the above requirements 1 to 5.
(i) Setting time of occurrence (see Supporting Report 6-31)
As in other countries, it is assumed that human intrusion does not occur during the period
when institutional control is effective [31][44][51][99]. Thus, times of 200 y [99], 300 y [44]or 500 y [31] after closure have been taken as possible times when the institutional
management period ends, which is taken as the earliest time for a human intrusion scenario.
In this report, the time when human intrusion occurs based on similar considerations forJapanese boundary conditions is also considered. In the safety regulations for medium depth
disposal currently under development [106], an institutional control period of 300-400 y isassumed, allowing human intrusion scenarios to be considered immediately after this period.
Similar assumptions for deep disposal were made. It is noted, however, that Japan has anextensive record of preservation of historical documents (e.g., the “Shōsōin documents”
dating from AD 702 [107]). In light of such experience, it was decided to assume that human
intrusion will occur 1 ky after closure, but consequences of human intrusion at 300 y after
closure are also analysed. These times are not significantly different from those set in othercountries.
In addition, it may be argued that there are cases where the timing of occurrence is being
considered based on the idea that human intrusion does not occur during the period when the
13 Water management (groundwater, surface water), including intake, reservoirs, dams and river management.

N.B. This footnote is not included in the Japanese version of the report.

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facility can be expected to have physical resistance. In Switzerland, for example, the reference
disposal concept for both HLW and SF has a steel overpack with an expected lifetime of 10
ky (although Cu is an alternative option, with a minimum lifetime of 100 ky and expected
lifetime of 1 My). The arguments for intrusion scenarios [51] depend on whether a borehole
directly hits an overpack or not. If it does, no penetration of the overpack is assumed until it isextensively corroded. If drilling is between the overpacks, then releases occur after loss ofcontainment (10 ky).
Based on this example, expected waste package resistance to drilling for a hot spring wascompared. This allows setting the time of occurrence of the scenario based on when these areconsidered to have lost such physical resistance.
The time of loss of physical resistance of the HLW overpack is taken to be 35 ky after
closure, i.e. much longer than the assumed time for loss of containment, since even then mostof the steel would remain. The loss of physical resistance is assumed to be 18 ky in the case of
a TRU waste package B (see Supporting Report 6-31). For TRU waste package A, since it has
no top cover, no physical resistance can be expected. Incidentally, for the estimation of theamount of steel corrosion, a more realistic corrosion rate was used (overpack Section 4.4.1 (2)(iii) (a), TRU waste package Section 4.4.2 (2) (v) (b)).

(ii) Setting the intrusion location
For the designed repository layout, hitting a waste package with a borehole and subsequent
penetration without this event being noticed by drillers is considered to have a very low
probability. Based on the requirements described in Section 6.5.1 (1), however, the borehole
is assumed to be drilled from the surface directly above the waste emplacement positions for
either HLW or TRU waste.
(iii) Migration path and exposure routes (see Supporting Report 6-32)
For the human intrusion scenario, there is a possibility of exposure to both the intruder andthe general public around the repository. In accordance with requirement 5 from Section 6.5.1(1), with reference to the current use of boreholes for the purpose of hot spring development
in Japan, a situation in which intruders and the general public are exposed was set as
described below [105] [108] [109]. It is noted that some safety regulations in other countries
do not require evaluation of dose to intruders [27].

(a) Driller exposure (worker exposure case)
Workers are exposed while drilling through the waste when material is transported to thesurface together with the excavated spoil or by exposure from the core. Currently, withboreholes aiming at hot spring development in Japan, it is common to periodically observe the
excavated spoil in order to understand the setting, with observation of bore core sometimes
performed instead. In consideration of this, it is assumed that one of the workers is
responsible for such observation of the spoil and/or core.

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(b) General public exposure (borehole pathway case)
For the general public radiation exposure, the borehole was assumed to be backfilled after
excavation, but provides a short circuit path between the repository and the biosphere. RNs
thus released to a river via a near surface aquifer can result in doses from the utilisation ofsuch water.
(4) Models and datasets used for dose evaluation (see Supporting Report 6-33)(i) Radionuclide inventory
To evaluate doses for human intrusion, it is important to specify the inventory of RNs
within the EBS at the time of occurrence. When the time of initiation of RN release fromwaste is shorter than the occurrence time, a part of the RN inventory is lost from the EBS; thisacts to reduce the dose calculated. Here, from the viewpoint of conservative evaluation, it isassumed that the original RN inventory will remain within the waste matrix, with correction
only for the effects of radioactive decay/ingrowth.
In addition, for the borehole pathway described in Section 6.5.1 (4) (iii), it is assumed thatall RNs released in aqueous solution will migrate to the surface through the borehole. If theborehole penetrates the waste package, it is conservatively assumed that the waste matrix iscompletely destroyed and all RNs are instantaneously released into the borehole.
(ii) Worker exposure caseIn the dose evaluation model of this case, one worker performs drilling work around theborehole. The drilling spoil tank is located at a fixed distance from the borehole, as is the
temporary store of drill cores. The worker may be externally exposed to radiation either fromobservation of the drilling spoil or by observation of the drill core. Inhalation of dust andinternal exposure by oral intake is also considered. Only exposure to the driller is consideredin this scenario, potential doses that may occur later, e.g., when analysing and mapping thecore or from subsequent use of the drilling spoil, are not considered.
The RN inventory contained in spoil or drill core is that of the waste matrix through which
the borehole penetrated. For HLW, the H12V option was considered. For TRU waste, a rangeof different waste groups and package emplacement designs are included. In this case, basedon its highest potential radiotoxicity, Gr.4H is conservatively selected for assessment.
In the case where representative information is available, such as the specifications of theequipment used in drilling for the purpose of hot spring development (e.g. the diameter of theborehole etc.), this is used in the dose evaluation model. For parameters that are highly
dependent on the working environment, such as dust concentration and core observation time,values are taken over from IAEA reference documents (e.g. [110]).
(iii) Borehole pathway specification
After hitting the waste, it is assumed the hole is backfilled with a material such as
cementitious grout; to model RN migration, the backfilled borehole is represented as a porous
medium. Abandoned open boreholes were not considered, since backfilling boreholes is thenormal practice in Japan.

6-161For HLW, dose assessment assumes one package penetrated and, for TRU waste, one
disposal tunnel for each waste group.
With respect to the RN transfer parameters in the borehole, due to uncertainty,
conservative values for the infilled hole were selected. For example, a high hydraulicconductivity (1 x 10-5 m/s) similar to sand is set together with an upward hydraulic gradient.
Since the gradient in the borehole depends on the specific conditions of the geologicalenvironment and surface topography, it is difficult to set in a generic manner, so values
equivalent to that of the representative host rocks14 are set (see Section 6.4.1 (2) (v) (a) 3).
With regard to the sorption of RNs on borehole infill, it is assumed that leaching of it does notproceed to a significant extent for the timescale considered (Supporting Report 6-16), andhence sorption distribution coefficients were chosen to be the same as those used in the basecase for the cementitious vault infill.
The length of the borehole migration path was defined by the repository depth minus the
thickness of an assumed permeable surface cover. Dose was calculated by multiplying therelease rate of RNs by the dose conversion factors used in the base case (see 6.4.1 (6) (iii)).
6.5.2 Results and discussion
For the worker exposure case, assuming that the work end or someone notice anabnormality within one year after the start of drilling the applied dose target is 20 to 100 mSv
over a 1- year period (see Section 6.1-6). However, in the case of borehole releases impacting
the general public, since there is a possibility that these will continue to be affected for a longtime after intrusion occurs, in the first year the dose target is 20 to 100 mSv. From the second
year onwards, comparison was made with a target of 1 to 20 mSv/y.
Table 6.5-1 summarises results for the two analysis cases (see Section 6.5.1 (4)), showing
the maximum value of the calculated radiation dose and its time of appearance.
In the worker exposure case, the exposure doses are highest for all rocks when the
intrusion time is 300 y from the closing of the repository (50 to 60 mSv). Nevertheless, thedose falls within the target range of 20 to 100 mSv. In this analysis case, it is assumed thatradiation comes directly from the waste, so there is no difference in results between rocktypes for core exposure and only a slight difference in dose from spoil, (due to a difference inrock density). In view of the timing of the human intrusion scenario, going from the period
considered for effective institutional control (300 to 1 ky after closure) to that defined byphysical resistance of the waste package (3.5 and 18 ky for HLW and TRU waste package B
respectively), decreases the calculated dose by one to two orders of magnitude due toradioactive decay of the inventory.

14 Plutonic rocks and Pre-Neogene sediments: 0.05, Neogene sediments: 0.06.

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Table 6.5-1 Maximum values of dose (mSv) for human intrusion scenarios

Analysis caseSpoilanalysis
WorkerExposure caseTarget dose:
20 to 100mSv
Core
analysis

Effective institutional
management
time (y)Physical resistancetime (y)

300

1,000

18,000(TRU)

35,000(HLW)

Plutonic

50
(HLW)

20
(HLW/TRU)6

0.6

Neogene

60
(HLW)

20
(HLW/TRU)

6

0.7

Pre-Neogene

50
(HLW)

20
(HLW/TRU)6

0.6

Plutonic40(HLW/TRU)

20
(TRU)

6

0.4

Neogene

40
(HLW)

20
(TRU)

4

0.4

Pre-Neogene

40(HLW/TRU)

20
(TRU)

6

0.4

Plutonic

3(TRU Gr.2)

0.7(TRU Gr.1)

0.7(TRU Gr.1)

4 x 10-5

Neogene7(TRU Gr.2)

2(TRU Gr.1)

2(TRU Gr.1)

9 x 10-5

Pre-Neogene

2(TRU Gr.2)

0.6(TRU Gr.1)

0.6(TRU Gr.1)

4 x 10-5

Host rock

Borehole pathway caseTarget dose:
1 to 20 mSv/yNotation used() type of waste that contributes to the maximum dose
HLW: High level radioactive waste
TRU: TRU waste
Gr.: GroupThe maximum value of the exposure dose to the general public from the borehole pathway
case is due to TRU waste when the scenario occurs 300 y after closure: for plutonic rocks, 3
mSv/y (77 y after drilling); for Neogene sediments, 7 mSv/y (36 y after drilling); and for PreNeogene sediments 2 mSv/y (84 y after drilling). In any case, these fall within the target rangeof 1 to 20 mSv. The largest value is for Neogene sediments, due primarily to shallower
repository depth and hence a shorter migration distance. For the 300-y occurrence, Cs-137(half-life: 30 y) from TRU waste Gr.2 dominates releases from the borehole pathway but, due
to decay, this is replaced at 1 ky by I-129 of Gr.1 (half-life: 1.6 x 107 y). The half-life of I-129
is so long that it is not attenuated even if the duration of physical resistance for TRU waste isconsidered.
As mentioned in Section 6.5.1 (3) (ii), the probability that the drilling for hot springdevelopment penetrates HLW or TRU waste is thought to be low, the probability ofoccurrence to calculate risk using the same method as the low probability perturbation
scenarios discussed in Section 6.4.3 is investigated.
The frequency of drilling for hot spring development [105] within the range of 300 to
1,000 m depth in Japan was noted to be 1 x 10-9 to 6 x 10-10/m2/y, regardless of terrain.
Conservatively assuming 8 x 10-9/m2/y as the upper limit of this range. As described above,

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considering both the worker exposure case and borehole pathway cases, the maximum doses
result from institutional control for 300 y for Neogene sediments: 60 mSv and 7 mSv/y,respectively. These values can be put into context by considering the area occupied by 40,000
HLW packages (about 6 x 103 m2) for H12V disposal. Thus, using the dose to risk conversion
factor of 5.7 x 10-2/Sv (see Section 6.1.5 (2)), the integrated risk for the worker exposure caseis ≈ 2 x 10-7/y and for the borehole pathway case ≈ 2 x 10-8/y, are far below the internationally
recommended risk limit of 10-5/y (see Supporting Report 6-34 for details).
Thus, even if only considering a minimum period of institutional control as limiting theearliest time that human intrusion occurs, this conservative analysis indicates that doses
would not be unacceptably high and, when coupled to an assessment of likelihood of
occurrence, risks are extremely small. Considering the physical resistance of waste packages,
the doses further decrease, suggesting that the reference repository design has somerobustness to limit impacts of human intrusion scenarios.

6.6 Summary and Future Perspective
6.6.1 Summary of safety evaluations
In this chapter, the three representative SDMs developed in Chapter 3 were assessed, basedon repositories tailored to them as described in Chapter 4. In Japan, safety regulations forgeological disposal will be defined only in the future, so the framework for conducting safetyassessment is not clear at this time, including radiation protection standards, required
scenarios for post-closure evaluation and the time scales to be considered. Therefore, for thisreport, the fundamental concept of safety assessment and the associated regulatory
requirements are based on those recommended by international organisations or used in othercountries, as was the case for the H12 and TRU-2 studies.
In the safety assessment, a risk-informed approach to assessing radiological impacts is
adopted, based on the dose calculated for specific scenarios and the probability of such
scenarios occurring. For this disaggregated dose/probability approach, results are compared totargets defined for different categories of scenario (base, variant, low probability perturbation,and human intrusion scenarios). In addition, the basic approach to safety assessment was
discussed, along with the assessment period, the RNs to be evaluated and the analyticalmodels and datasets used to calculate doses.
To initiate this approach, systematic methods for developing scenarios belonging to eachcategory were used. After formally describing the behaviour of the repository after closure,facilitated by the use of storyboards, a top-down method of considering the safety functions
expected for the components of the repository system was combined with a bottom-up
approach based on assessment of relevant FEPs. Such FEPs are compiled from the latest
internationally developed databases and the FEP lists used in previous Japanese safetyassessment to ensure that the resulting list (NUMO FEP list) is comprehensive.
In developing the models and setting up the data sets for scenario consequence analysis,
consideration was given to realistically reflect the characteristics of both the SDM andrepository to the extent possible. Specifically, a 3D model is utilised to explicitly evaluate thearrangement of components of the repository with respect to the network of features in which
groundwater flows and RNs migrate through a limited portion of the surrounding host rock.

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However, it is admitted that the current approach is simplistic and does not fully take into
account the 3D aspects of the entire repository and its setting within the SDM.
The model databases include characteristics of repository components derived from theSDMs and assessment of the evolution of the EBS structures situated within them (e.g.
hydraulic conductivities reflecting the occurrence of cracks in the structural framework of
TRU waste vaults). Here, again, there are many simplifications which constrain the extent towhich all characteristics of the repository systems specified can be captured. Databases
including, or used to derive, RN migration parameters, such as thermodynamic databases used
for setting chemical speciation and elemental solubility in evolving EBS porewater, sorption
distribution coefficients and diffusion coefficients are specified, even if all uncertainties
associated with their application are not fully assessed.
The analyses of the base scenario indicate that the proposed dose targets can be met for all
host rocks/repository design options. The variant scenarios assess impacts of uncertainties in
the models and databases used for the base scenario, but most of these have little impact on
doses, and all meet the specified dose targets.
The results of dose assessment for low probability perturbation scenarios and human
intrusion scenarios also fall within the range of target doses set according to internationalguidelines. When the assessed probability of occurrence for such scenarios is included, thetotal risk calculated was also shown to be much lower than the risk constraint values
suggested by international organisations.
In conclusion, the assessment indicates that NUMO is well situated to assess the postclosure safety of repositories tailored to the sites likely to result from the volunteering process.
Furthermore, no fundamental issues were identified that would call previous assessments ofthe fundamental feasibility of safe geological disposal in Japan into question. However,further development of the safety assessment methodology, enhancing realism and avoiding
over-conservative assumptions, would be needed in later stages of the repository programme.

6.6.2 Extension of assessment to coastal sites
By considering constraints on site acceptability, such as it being favourable fortransportation of waste and the high probability of confirmation of favourable characteristics,
scientifically preferable areas have been identified [92]. These particularly include coastal
areas and small islands.
A study group on “Technical issues for geological disposal under the coastal seafloor”
clarified the present state-of-the-art and identified R&D to develop the technology necessary
for the repository to be located below the sea, but accessed from land. This included safety
considerations of coastal sub-seabed repositories, applicable safety evaluation methodology,
examples of safety assessments, and future technical issues to be clarified [111].
In coastal areas, seawater often penetrates under fresh water due to its higher density,
forming a saltwater intrusion wedge (see Section 3.1.3 (1)). The freshwater/saltwater interface
moves in response to shifts of coastline position due to sea level changes accompanying longterm glacial cycles (although it can also be impacted by human activities, such as water
extraction from aquifers). As a result, the direction, flow velocity and chemical properties ofgroundwater may periodically change, depending on the location of the repository and local
topography/bathymetry. Depending on whether the initial location of the repository is in a

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saltwater or freshwater environment, required EBS material properties and the evolution of
the repository system will be different, as will parameters related to RN release and transport(solubility, sorption, diffusion behaviour, etc.). This becomes more complicated if therepository environment cyclically changes between fresh and marine conditions. Also, unlike
a repository located far from the coast, in order to calculate radiation doses caused by RNs
released from a repository located under the coastal seafloor, it is necessary to consider RN
migration processes to a relevant GBI coupled to an associated oceanic biosphere model.
Although not included in this safety case, a safety assessment method for coastal areas has
been developed and performed a preliminary evaluation of coastal submarine disposal [112].
For this, key characteristics of relevant geological environments are firstly defined, including
uplift/erosion, climate cycles and associated sea level changes. These allow development of
evolution scenarios, which can be assessed using models developed to analyse groundwaterflow driven by density, variations in salt concentration in groundwater as a result of flow andmixing, etc. These form the basis for conducting RN migration analysis to an identified GBI,
using appropriately modified RN release and transport parameters.
In this evaluation, RN migration analysis is carried out for the THMC characteristics
expected around the repository, assessed in terms of impacts due to sea level change
combined with uplift/erosion. For repositories located under the coastal seabed, the hydraulicgradient is usually extremely small, so very favourable conditions for ensuring safety may be
expected. However, this analysis needs to be extended in the future to ensure that it will bemature enough to apply to volunteer sites.
6.6.3 Potential to improve future assessments
In order to further improve the technical basis for safety evaluation, NUMO cooperates
with both national and international R&D organisations on the following initiatives.








In the future, a goal is to make assessments more realistic and to account for allcomponents of the repository system in their specific geological settings. In particular,
more realistic RN release and migration models for both the EBS and geosphereshould better reflect their 3D characteristics for specific sites and the repository
concepts tailored to them, in line with the stepwise improvement of the knowledge
base. In addition, more emphasis on model testing, verification and validation will beneeded in later stages of this process.
In order to improve traceability, the methodology involved in the work flow fromdevelopment of scenarios to setting of analysis cases based on the understanding of
the behaviour the repository system will be developed in a more systematic way,utilising advanced knowledge management tools.
In terms of such technology development, focus will be on the actual volunteer sites
in order to ensure that more realistic assessment of safety functions of repositories
tailored to them can effectively contribute to the comparison between sites and the
optimisation of repository designs for these.
Continual expansion of the RN release and migration database for relevant geologicalenvironments and associated evolution of the EBS for the wastes considered.

Table 6.6-1 shows key issues and technical development items related to these topics.

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Table 6.6-1 Technical development tasks to improve post-closure safety assessment
Classification of tasks

Major technical development items

1. Upgrading disposal systemmodels
2. Improvement of scenario
development for risk-informed
assessment

3. Advanced modelling of RN
migration to capture systemevolution
4. Advanced modelling of RN
migration to capture facility
design5. RN-specific databasedevelopment



Ensure a less simplified description of the 3D characteristics ofboth the EBS and the geological environment
Improvement of RN release models for all waste types
Improvement of models of evolution/interaction of near fieldcomponents
Further development and assessment of models of gases derived
from waste
Development of evaluation methods for colloids, organics andmicroorganisms
Testing of model of nitrate impacts
Development of models for, and analyses of, repository
component evolution and its impact on safety function,
considering couplings between different processes in an
integrated fashion
Structured approach to model testing, verification and validation
Management tool to support RN migration analysis case setting
within the scenario development process
Improved sophistication of storyboards
Further studies of human intrusion and low probability
perturbation scenariosModelling of the near field
Modelling of the far field
Improved biosphere modelling to reflect evolution of specific
sites



Construction of advanced RN migration models capturing key
aspects of facility designExpansion of the database supporting setting of RN-specificparameters for conditions relevant to repositories
Expansion of data to support evaluations of relevant biospheres
Development of a methodology for defining RN release andmigration parameters during stepwise site characterisation(1) Improving models of the repository(i) Improvement of RN release models for all waste types
For vitrified HLW, processes that influence glass dissolution rate have already been
identified. In the future, long-term immersion tests will be conducted to better quantify theeffects of overpack derived iron and protection by alteration layers, aimed to confirm thevalidity of existing RN release models. Also, based on the knowledge acquired through suchtesting, improvement of the associated RN dissolution model within the limited volume insidea failed overpack may be considered.
For Gr.2 TRU waste, RN dissolution tests on actual waste, including compacted hulls andend pieces will be conducted, to support development of a more realistic RN release model.

(ii) Improvement of models of evolution/interaction of near field components
To evaluate evolution of near field consisting of several different constituent materials, test
data on alteration of buffer by both overpack and cementitious material has been acquired and6-167

supporting analytical models have been constructed. In the future, a wider emphasis on the
evolving groundwater environment will be developed, explicitly including issues for coastal
areas. This will obtain data on thermal, hydraulic, mechanical and chemical aspects ofoverpack and buffer alteration, in particular, to validate the current models. More generally,
when important evolution processes are clarified, the goal will be to capture these inquantitative models. For example, studies on use of cementitious grout during construction of
the repository are conducted and a reactive transport model developed to assess the impacts ofa resultant alkaline plume on the host rock.
Perturbations related to buffer erosion in the transient period after closure of the repository
have also been identified as a concern; thus, acquisition of relevant data for Japaneseboundary conditions and development of an analytical model will be carried out, along withassociated validity assessment.
(iii) Study of models of waste package derived gases
Gas migration tests of buffer and cementitious materials have been carried out to evaluate
the influence of gases generated within a repository on key safety functions. In addition toacquiring relevant data, construction of a model capable of coupled two-phase analysis isbeing developed. In the future, the validity of this model will be tested by small scale gas
migration experiments. After confirming its validity, it will be used to examine the effect ofgas on release and transport of RNs.
(iv) Development of models for colloids, organics and microorganisms
To date, the development of a model to quantify the impact of bentonite colloids on RN
migration is ongoing. In the future, the production and stability of bentonite colloids will be
captured in this model, the validity of which will be tested in various geological environments.
With regard to natural colloids present in groundwater, construction of impact assessment
models for RN migration are under development for relatively large colloids (> 0.2 μm), and
it is expected that, in the future, the model will be extended to smaller colloids.
In terms of the effect of organic solutes on release and migration of RNs, assessment has
been limited to the influence of isosaccharinic acid (ISA) derived from TRU waste on thesolubility of actinide elements, based on a chemical thermodynamic complexation model. To
extend this, data on impacts of ISA on the solubility and sorption of transition elements will
be obtained, and RN partitioning data in the ternary system of natural organic matter-RN-rockwill be obtained and analysed in order to develop a model of its impacts on repository
performance.
To assess the effects of microbial activity on RN migration, relevant data on uptake of RNs
have been acquired for some specific microorganisms. In the future, this work will be
expanded to develop a better overview of the potential issues involved.

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(v) Testing of nitrate impact models
Existing approaches for evaluating the influence of nitrate contained in TRU waste Gr.3 on
the safety functions of the repository will be extended to better capture direct interactions
between nitrate and RNs and thus impacts on RN release and transport. Models will be tested
using data acquired by long-term immersion tests and analogue cases of nitrate contaminationof groundwater. In addition to direct complexation by nitrate ions, consideration of its
potential role as an oxidant will be extended and the impacts of reduction products, such as
ammonia, on the mobility of RNs will be assessed.
(vi) Construction of a 4D evolution model
In order to understand temporal and spatial evolution of the repository system,
development of a platform that enables analysis by integrating various component alteration
models, with a focus on the near field, has been developed. In the future, this will be extended,with the goal of expanding the platform to enable more advanced coupled analysis of all
relevant processes. This will aid identification of important issues to further improve therealism of the safety assessment.
(2) Improvement of risk-informed scenario development(i) Management tool concept
As shown in this report, a methodology for developing scenarios based on their probability
has been developed, focused on the evolution of the safety functions of specific repository
components. It is planned that management tools will be implemented to ensure traceability ofthis process, systematically capturing all tacit assumptions and supporting arguments.
In addition, the FEP list required for the scenario development has been updated, throughparticipation in the OECD/NEA FEP Database Project and associated work documented intechnical reports. Supporting information will continue to be gathered within national andinternational R&D projects, updating the FEP list if and when appropriate.
(ii) Improving storyboards
In order to contribute to the completeness of scenarios, storyboards are being developed tocapture the latest scientific knowledge from experts in relevant fields and set this in thecontext of a Japanese repository. In the future, in order to contribute to a consistent
representation of the evolution of the repository, including the period before closure, themethodology for construction of storyboards that can capture, synthesise and illustrate a wider
range of knowledge will be improved. For example, utilisation of animation can better
represent the long-term evolution of repository system and its component parts. When thisincludes a link function to supporting databases, the general overview can be combined withmore detailed, technical information. In order to contribute to development of scenarios
compatible to a risk-informed approach, the aim is to incorporate functionality that canvisually express different possible evolutions and their associated probabilities.

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(iii) Further study of human intrusion and low probability perturbation
scenarios
A wider range of credible human intrusion and low probability perturbation scenarios will
be developed, with reference to work by relevant Japanese and other national/internationalorganisations. In addition, supporting information (such as assessments of the effectiveness ofpreservation of records to reduce risks of human intrusion) will continue to be collected withthe aim of being able to tailor such scenarios to the boundary conditions of actual sites.
(3) Advanced RN migration models that capture system evolution(i) Modelling of the near field
For the near field, where processes such as reaction of concrete and buffer are coupled andproceed in a microporous charged environment where normal thermodynamic models arelimited or completely inapplicable, acquisition of basic data to support model development
will continue. Thus, for example, further data relating to the sorption and diffusion of RNs
will be measured in buffer converted to Fe-type, due to corrosion of the overpack, or Ca-type,by cementitious leachate from material. In the future, a more realistic near field evolution
model will be developed, based on data on the behaviour of RNs in cementitious buffermaterials as alteration progresses. Advanced R&D aims both to establish RN migration
models applicable to such an evolving system and test their validity to the extent possible.

(ii) Modelling of the far field
To advance models of the migration of RNs within fractured rock, it is intended to go
beyond the simple treatment of regular channels in parallel plate fractures to allow more
realistic representation of the micro-scale models of water conducting fractures in crystalline
rock (e.g. granite) and also other relevant features in sedimentary rock (e.g. sand channels in
mudstone). Such improved representation of the RN migration paths will be complemented
by expanded databases of interactions that result in retardation, including diffusion into, andsorption onto, relevant phases resulting from rock-water interaction as described in the SDMs.
Model development will proceed in parallel to establishing test cases to validate bothhydrological and solute transport models over larger scales (extending from several tens to ahundred metres or so, and times of many years).
(iii) Improved biosphere modelling to reflect geological evolution
Based on evaluation of Japan as a whole from the viewpoint of climate and topography, asimple generic biosphere compartment model has been developed that allows assessment ofdoses to specified critical groups for certain stylised RN release scenarios. In the future, inorder to express capture of the surface environment more realistically, this model will be
refined by improving the compartment representations in order to reflect changes with time
and the spatial extent of the GBI, based on a long-term evolution model of the geologicalenvironment. It is intended that this will focus on actual volunteer sites as soon as they areavailable, also taking local lifestyle into account, especially for coastal communities.

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(4) Advanced RN migration models that capture repository design
In the future, in addition to improving the computational capability to support more
extensive and detailed 3D solute transport modelling, it is necessary to demonstrate model
consistency between the spatial scales considered (regional to repository scale, panel scaleand near field scale) to allow differences between different repository layouts and EBS
designs to be assessed. This will require models that can realistically quantify RN migration
for the geometry, layout and physicochemical characteristics of the constituent components of
the repository, but also capture their temporal evolution, e.g. as a result of changes of thegroundwater flow field and groundwater chemistry due to climate/sea level change. In orderto perform such analysis more efficiently, high-speed processing methods will be utilised,such as parallel computation.
(5) RN-specific database development(i) Expansion of data required for RN release and transport models
The thermodynamic database (TDB), sorption database (SDB) and diffusion database(DDB) utilised for performance assessment are continuously expanded through laboratory
research and review of the technical literature. In the future, particular emphasis will beplaced on accumulating relevant data for rocks and geochemical conditions relevant to thecoastal submarine environment.
In particular, as shown in this report, the measurement database under high carbonate
concentrations is poor and hence RN migration parameters were set in an extremely
conservative manner and thus yield unrealistically high doses (even if these are below
specified targets). For this reason, data acquisition to determine more realistic RN migration
parameters for high carbonate groundwater under reducing environments is a future focus.
Acquisition of data in such systems is experimentally difficult and, to interpret such data,
development of a mechanistic sorption model more suited to such an environment will also be
considered.
In addition, associated with model development noted in (1) and (3) above, databases forsetting RN release and migration parameters taking into consideration long-term systemevolution due to interaction between the EBS components and the surrounding geosphere will
be developed (e.g., sorption data of RNs under high-concentrations of nitrate, sorption data of
RNs on Ca-type bentonite, etc.).
(ii) Expansion of biosphere data
Acquisition of data such as sorption distribution coefficients of relevant RNs onto surface
soil and research on carbon cycling mechanisms in the biosphere have so far been carried out
for Japan as a whole. In the future, there will be a focus on areas relevant to specific sites,
including the coastal ocean floor, where data necessary for biosphere assessment is not
sufficiently developed. In addition, data determining the behaviour of iodine and actinides(together with their daughters), which were identified as particularly relevant in this report,
will be a focus.

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(iii) Refinement of RN migration parameters following site surveys
So far, data acquired in the laboratory under test conditions was applied, focusing on
generic geological environments in order to set parameters such as sorption distribution
coefficients and diffusion coefficients. From now on, based on geological informationacquired during investigations of actual sites, focused data on such parameters will beobtained, as preparation for preliminary investigation area (PIA) and detailed investigation
area (DIA) selection.

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Supporting Reports (SRs)SR 6-1 Biosphere assessment models and parameters
SR 6-2 Time period for safety assessment
SR 6-3 International approach to scenario section and setting performance goals
SR 6-4 NUMO FEP list
SR 6-5 FEP screening and integration
SR 6-6 Safety function diagrams used for scenario construction
SR 6-7 Factor analysis diagrams
SR 6-8 Process analysis used for impact assessment
SR 6-9 Impact analysis tables for safety functions
SR 6-10 Examination of the effects of uplift / erosion
SR 6-11 Storyboards
SR 6-12 Summary and verification of analysis codes used for post-closure safety evaluation
SR 6-13 Radionuclide release and migration models
SR 6-14 Near-field 3D groundwater flow and particle tracking models
SR 6-15 Setting porewater chemistry in the EBS
SR 6-16 Permeability of cementitious materials
SR 6-17 Setting solubilities in EBS porewater
SR 6-18 Setting speciation in groundwater and buffer porewater
SR 6-19 Setting diffusion coefficients in the EBS
SR 6-20 Setting EBS sorption distribution coefficients
SR 6-21 Setting diffusion coefficients in host rockSR 6-22 Setting sorption distribution coefficients in the host rockSR 6-23 Representation of fault zones in the safety assessment (based on H12)
SR 6-24 Results of the safety assessment for plutonic rocks (base and variant cases)
SR 6-25 Results of the safety assessment for Neogene sedimentary rocks (base and variant
cases)
SR 6-26 Results of the safety assessment for Pre-Neogene sedimentary rocks (base andvariant cases)
SR 6-27 New volcano scenarios
SR 6-28 Fault extension scenarios
SR 6-29 Requirements for development of human intrusion scenarios
SR 6-30 Human intrusion scenarios
SR 6-31 Human intrusion scenario constraints
SR 6-32 Human intrusion scenario migration paths
SR 6-33 Analysis models and data for human intrusion scenarios
SR 6-34 Safety assessment results for human intrusion scenarios
SR 6-35 Study of the effects of high carbonate groundwater

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6-183

Appendix
Table 1 NUMO FEP ListNo.

FEP No.

FEP name

1 F1.1.1

Quality assurance and control

2 F1.1.2

Site investigations

3 F1.1.3

Repository design

4 F1.1.4

Schedule and planning

5 F1.1.5

Construction

6 F1.1.6

Operation

7 F1.1.7

Closure8 F1.1.8

Accidents and unplanned events

9 F1.1.9

No.

FEP No.

30 F1.3.7
31 F1.3.8
32 F1.3.9
33 F1.3.10

FEP name
Hydrological response to climatechangeEcological response to climatechangeHuman response to climatechangeGeomorphological response to
climate changes

34 F1.4.1

Human influences on climate

Repository administrative control

35 F1.4.2

Social and institutional
developments

10 F1.1.10

Monitoring

36 F1.4.3

Technological developments

11 F1.1.11

Repository markers

12 F1.2.1

Tectonic movement

37 F1.4.4

Knowledge and motivational
issues (repository)

13 F1.2.2

Orogeny

38 F1.4.5

Drilling activities

14 F1.2.3

Deformation (elastic, plastic, orbrittle)39 F1.4.6

Mining and other underground
activities

15 F1.2.4

Seismicity

40 F1.4.7

Unintrusive site investigation

16 F1.2.5

Volcanic and magmatic activity

41 F1.4.8

Surface Environment

17 F1.2.6

Metamorphism

42 F1.4.9

18 F1.2.7

Hydrothermal activity

Water management (groundwaterand surface water)

43 F1.4.10

Explosions and crashes

19 F1.2.8

Regional erosion andsedimentation

44 F1.4.11

Remedial Actions

20 F1.2.9

Diagenesis

45 F1.4.12

Deliberate human intrusion

21 F1.2.10

Pedogenesis

46 F1.5.1

Meteorites and human spacedebris47 F1.5.2

Evolution of biota

48 F2.1.1.1

Solid

49 F2.1.2.1

Metallic wastes

22 F1.2.12
23 F1.2.13

Hydrological/Hydrogeologicalresponse to geological changes
Geomorphological response togeological changes

24 F1.3.1

Global climate change

50 F2.1.2.2

Organic wastes

25 F1.3.2

Regional and local climatechange51 F2.1.2.3

Non-metallic, inorganic wastes

26 F1.3.3

Sea level change

52 F2.1.3

Waste conditioning matrix

27 F1.3.4

Periglacial effects

53 F2.1.4.1

Radionuclide content

28 F1.3.5

Local glacial and ice-sheet effects

54 F2.1.4.2

Chemical content

Warm climate effects (tropicaland desert)

55 F2.1.5

Waste form properties

29 F1.3.6

56 F2.2.1

Containers

6-A-1

No.

FEP No.

57 F2.2.2
58 F2.3.1.1
59 F2.3.1.2
60 F2.3.1.3
61 F2.3.1.4
62 F2.3.2.1

FEP name

No.

Overpacks

FEP No.

84 F2.3.5.1

Radiogenic heat production andtransfer
Chemical heat production andtransfer
Biological heat production andtransfer
Impact of thermal processes onother processes (waste package)
Resaturation/desaturation (waste
package)

85 F2.3.5.2
86 F2.3.5.3
87 F2.3.6.1

FEP name
Microbial growth and poisoning(waste package)
Microbially/biologically mediatedprocesses (waste package)
Impact of biological processes onother processes (waste package)
Radioactive decay and ingrowth(waste package)

88 F2.3.6.2

Radiolysis (waste package)

89 F2.3.6.3

Helium production

90 F2.3.6.4

Radiation attenuation (waste
package)
Radiation damage (waste
package)
Impact of radiological processes
on other processes (waste
package)

63 F2.3.2.2

Thermal effects (waste package)

64 F2.3.2.3

Gas effects (waste package)

65 F2.3.2.4

Impact of hydraulic processes onother processes (waste package)

66 F2.3.3.1

Package deformation

92 F2.3.6.6

67 F2.3.3.2

Material volume changes (waste
package)

93 F2.3.7.1

68 F2.3.3.3

Package movement

94 F2.3.7.2

69 F2.3.3.4

Stress corrosion cracking

70 F2.3.3.5

Gas explosion (waste package)

95 F2.3.7.3

Impact of mechanical processes
on other processes (waste
package)

96 F2.3.7.4

Radiolysis (waste package)

71 F2.3.3.6

97 F2.3.7.5

Volatilisation (waste package)

72 F2.3.4.1

pH conditions (waste package)

98 F2.3.7.6

Gas dissolution (waste package)73 F2.3.4.2

Redox conditions (waste package)

99 F2.3.7.7

Gas-induced failure

74 F2.3.4.3

Perturbing species' concentrations(waste package)

100 F2.3.7.8

Impact of gas generation on otherprocesses (waste package)

75 F2.3.4.4

Corrosion (waste package)

101 F2.4.1.2

Dissolution (waste form)

Polymer degradation (waste
package)

102 F2.4.1.3

Diffusion (waste form)

76 F2.3.4.5
77 F2.3.4.6

Dissolution (waste package)78 F2.3.4.7

Mineralisation (waste package)

79 F2.3.4.8
80 F2.3.4.9
81 F2.3.4.10
82 F2.3.4.11
83 F2.3.4.12

Precipitation reactions (waste
package)
Chelating agent effects (waste
package)
Colloid formation (waste
package)
Chemical concentration gradients(waste package)
Impact of chemical processes onother processes (waste package)

91 F2.3.6.5

103 F2.4.1.4
104 F2.4.1.5

Metal corrosion (waste package)
Organic degradation (waste
package)Radon production (waste
package)Speciation and solubility (waste
form)
Sorption and desorption (waste
form)

105 F2.4.1.6

Complexation (waste form)

106 F2.4.1.7

Colloids

107 F2.4.2.1

Gaseous wastes

108 F2.4.2.2

Radon production (waste form)

109 F2.4.2.3

Volatilisation (waste form)

110 F2.4.2.4

Radiolysis (waste form)

111 F2.4.3

Solid-mediated release

112 F2.4.4

Human-action-mediated release

6-A-2

No.

FEP No.

FEP name

113 F2.5.1

Transport pathways (waste
package)

114 F2.5.2.1

Advection (waste package)

115 F2.5.2.2

Dispersion (waste package)

116 F2.5.2.3
117 F2.5.2.4
118 F2.5.2.5
119 F2.5.2.6

Molecular diffusion (waste
package)
Dissolution, precipitation, andmineralisation (waste package)Speciation and solubility (waste
package)
Sorption and desorption (waste
package)No.

FEP No.

FEP name

142 F3.2.4.5

Dissolution (repository)

143 F3.2.4.6

Mineralisation (repository)

144 F3.2.4.7
145 F3.2.4.8
146 F3.2.4.9
147 F3.2.4.10
148 F3.2.4.11
149 F3.2.4.12

Precipitation reactions(repository)
Chelating agent effects(repository)
Colloid formation (repository)
Chemical concentration gradients(repository)
Impact of chemical processes onother processes (repository)
Polymer degradation (repository)

120 F2.5.2.7

Complexation (waste package)

121 F2.5.2.8

Colloid transport (waste package)

122 F2.5.3

Gas-mediated transport (waste
package)

123 F3.1.1

Design124 F3.1.2

Buffer/backfill

125 F3.1.3

Room/tunnel seals

153 F3.2.6.1

126 F3.1.4

Shaft/ramp seals

154 F3.2.6.2

Radiolysis (repository)

127 F3.1.5

Other engineered features

155 F3.2.6.3

Radiation attenuation (repository)

156 F3.2.6.4

Radiation damage (repository)

157 F3.2.6.5

Criticality

158 F3.2.6.6

Impact of radiological processes
on other processes (repository)

159 F3.2.7.1

Metal corrosion (repository)

160 F3.2.7.2

Organic degradation (repository)

161 F3.2.7.3

Radon production (repository)

162 F3.2.7.4

Radiolysis (repository)

163 F3.2.7.5

Volatilisation (repository)

164 F3.2.7.6

Gas dissolution (repository)

165 F3.2.7.7

Gas-induced dilation (repository)

166 F3.2.7.8

Impact of gas generation on otherprocesses (repository)

128 F3.1.6
129 F3.2.1.1
130 F3.2.1.2
131 F3.2.2.1
132 F3.2.2.2
133 F3.2.2.3
134 F3.2.3.1

Excavation damaged anddisturbed zones
Thermal conduction andconvection
Impact of thermal processes onother processes (repository)
Resaturation/desaturation(repository)
Piping/hydraulic erosion
Impact of hydraulic processes onother processes (repository)
Material volume changes(repository)

150 F3.2.5.1
151 F3.2.5.2
152 F3.2.5.3

Microbial growth and poisoning(repository)
Microbially/biologically mediatedprocesses (repository)
Impact of biological processes onother processes (repository)
Radioactive decay and ingrowth(repository)

135 F3.2.3.2

Creep

136 F3.2.3.4

Gas explosion (repository)
Impact of mechanical process onother processes (repository)

167 F3.3.1

Transport pathways (repository)

137 F3.2.3.5

168 F3.3.2.1

Advection (repository)

138 F3.2.4.1

pH conditions (repository)

169 F3.3.2.2

Dispersion (repository)

139 F3.2.4.2

Redox conditions (repository)

170 F3.3.2.3

Molecular diffusion (repository)

140 F3.2.4.3

Perturbing species' concentrations(repository)

171 F3.3.2.4

Dissolution, precipitation, andmineralisation (repository)

141 F3.2.4.4

Corrosion (repository)

6-A-3

No.

FEP No.

172 F3.3.2.5
173 F3.3.2.6

FEP name
Speciation and solubility(repository)
Sorption and desorption(repository)

174 F3.3.2.7

Complexation (repository)

175 F3.3.2.8

Colloid transport (repository)

176 F3.3.3
177 F3.3.4
178 F3.3.5

Gas-mediated transport(repository)
Solid-mediated transport(repository)
Human-action-mediated transport(repository)No.

FEP No.

199 F4.2.4.1
200 F4.2.4.2

201 F4.2.4.3

202 F4.2.5.1
203 F4.2.5.2
204 F4.2.5.3

FEP name
Geochemical effects of repository(geosphere)
Geochemical effects of climatechange (geosphere)
Other processes affecting futuregeochemical conditions in
geosphere
Biological effects of repository(geosphere)
Biological effects of climatechange (geosphere)
Other processes affecting future
biological conditions in geosphereRadiological processes(geosphere)

179 F4.1.1

Stratigraphy

180 F4.1.2

Host rock lithology

205 F4.2.6

181 F4.1.3

Large-scale discontinuities

206 F4.2.7.1

Gas sources (geosphere)

182 F4.1.4

Geological resources

207 F4.2.7.2

Radon production (geosphere)

183 F4.1.5

Undetected features

208 F4.2.7.3

Volatilisation (geosphere)

184 F4.1.6

Current geothermal state

209 F4.2.7.4

Gas dissolution (geosphere)

185 F4.1.7

Current hydraulic state

210 F4.2.7.5

Gas-induced dilation (geosphere)

186 F4.1.8

Current stress state

211 F4.3.1

Transport pathways (geosphere)

187 F4.1.9

Current geochemical state

212 F4.3.2.1

Advection (geosphere)

188 F4.1.10

Current biological state

213 F4.3.2.2

Dispersion (geosphere)

189 F4.1.11

Current gas state

214 F4.3.2.3

Molecular diffusion (geosphere)

190 F4.2.1.1

Thermal effects of repository(geosphere)

215 F4.3.2.4

Matrix diffusion

216 F4.3.2.5

191 F4.2.1.2

Thermal effects of climate change(geosphere)

192 F4.2.1.3
193 F4.2.2.1
194 F4.2.2.2

195 F4.2.2.3

196 F4.2.3.1
197 F4.2.3.2
198 F4.2.3.3

Other processes affecting future
thermal conditions in geosphere
Hydraulic effects of repository(geosphere)
Hydraulic effects of climatechange (geosphere)
Other processes affecting future
hydraulic conditions in thegeosphere
Mechanical effects of repository(geosphere)
Mechanical effects of climatechange (geosphere)
Other processes affecting future
stress conditions in geosphere

217 F4.3.2.6
218 F4.3.2.7

Dissolution, precipitation, andmineralisation (geosphere)
Speciation and solubility(geosphere)
Sorption and desorption(geosphere)

219 F4.3.2.8

Complexation (geosphere)

220 F4.3.2.9

Colloid transport (geosphere)

221 F4.3.3
222 F4.3.4
223 F4.3.5

Gas-mediated transport(geosphere)
Solid-mediated transport(geosphere)
Human-action-mediated transport(geosphere)

224 F5.1.1

Topography and morphology

225 F5.1.2

Biomes226 F5.1.3.1

Surface soils

227 F5.1.3.2

Overburden

6-A-4

No.

FEP No.

FEP nameNo.

FEP No.

FEP name
Transport associated with surface
water bodies
Dissolution and precipitation(biosphere)
Speciation and solubility(biosphere)
Sorption and desorption(biosphere)

228 F5.1.3.3

Aquatic sediments

259 F5.3.1.3

229 F5.1.4

Near-surface aquifers and waterbearing features

260 F5.3.1.4

230 F5.1.5.1

Wetlands

231 F5.1.5.2

Lakes and rivers

232 F5.1.5.3

Spring and discharge zones

233 F5.1.6

Coastal features

234 F5.1.7

Marine features

263 F5.3.1.7

Complexation (biosphere)

235 F5.1.8

Atmosphere

264 F5.3.1.8

Colloid transport (biosphere)

236 F5.1.9

Vegetation

265 F5.3.2.1

Gas discharge to biosphere

237 F5.1.10

Animals

266 F5.3.2.2

Radon production (biosphere)

238 F5.1.11

Climate and weather

267 F5.3.2.3

Volatilisation from soil/water

239 F5.1.12

Hydrological regime and waterbalance (near-surface)

268 F5.3.3

240 F5.1.13

Erosion and deposition

269 F5.3.4

241 F5.1.14
242 F5.2.1

Ecological/biological/microbial
systems
Human characteristics(physiology, metabolism)

261 F5.3.1.5
262 F5.3.1.6

270 F5.3.5
271 F5.3.6

Solid-mediated transport(biosphere)
Human-action-mediated transport(biosphere)
Atmospheric transport anddeposition
Biologically-mediated transport
Foodchains and uptake ofcontaminantsContaminated drinking water andfood

243 F5.2.2

Age, gender, and ethnicity

244 F5.2.3.1

Farming diet

245 F5.2.3.2

Hunter/gatherer diet

273 F5.4.1

246 F5.2.3.3

Other diets

274 F5.4.2

Contaminated non-food products

247 F5.2.4

Habits (excluding diet)

248 F5.2.5.1

Community type

275 F5.4.3

Other contaminatedenvironmental media

249 F5.2.5.2

Community location

276 F5.4.4.1

Exposure of humans

250 F5.2.5.3

Water source

277 F5.4.4.2

251 F5.2.6

Food preparation and water
processing

278 F5.4.5.1

252 F5.2.7

Dwellings

253 F5.2.8
254 F5.2.9
255 F5.2.10
256 F5.2.11
257 F5.3.1.1
258 F5.3.1.2

Natural/semi-natural land andwater use
Rural/agricultural land and wateruse
Urban/industrial land and wateruse
Leisure and other uses of theenvironment
Groundwater discharge tobiosphere
Transport associated with surface
soil and overburden

272 F5.3.7

279 F5.4.5.2
280 F5.4.6.1
281 F5.4.6.2
282 F5.4.7.1
283 F5.4.7.2
284 F5.4.8

6-A-5

Exposure of biota other thanhumans
Dosimetry and biokinetics forhumans
Dosimetry and biokinetics forbiota other than humans
Radiological toxicity/effects forhumansRadiological toxicity/effects forbiota other than humans
Chemical toxicity/effects forhumansChemical toxicity/effects for biota
other than humans
Radon and radon daughter
exposureTable 2 NUMO FEP list classification scheme (level 1 and level 2)
Level 11
Level 2F1.1 Repository issues
F1.2 Geological factors
F1 EXTERNAL FACTORS
F1.3 Climatic factors
F1.4 Future human actions
F1.5 Other external factors
F2.1 Waste form characteristics and properties
F2.2 Waste packaging characteristics and properties
F2 WASTE PACKAGE
F2.3 Waste package processes2
FACTORSF2.4 Contaminant release (from waste form)
F2.5 Contaminant transport (waste package)F3.1 Repository characteristics and properties
F3 REPOSITORY FACTORS
F3.2 Repository processes2
F3.3 Contaminant transport (repository)
F4.1 Geosphere characteristics and properties
F4 GEOSPHERE FACTORS
F4.2 Geosphere processes2
F4.3 Contaminant transport (geosphere)
F5.1 Surface environment
F5.2 Human behaviour
F5 BIOSPHERE FACTORS
F5.3 Contaminant transport (biosphere)
F5.4 Exposure factors
1

The NUMO FEP consists of a maximum of four hierarchical levels as indicated in the FEP number.
Table 6-2 shows the classification of levels 1 and 2. F1 External factors, F2.2 Characteristics and properties of
packaged waste, F3.1 Characteristics and properties of the repository, and F4.1 Characteristics and properties of
the host rock are up to level 3, and the others are up to level 4.
2

F2.3, F3.2 and F4.2 have a common structure. Level 3 is divided into thermal processes, hydraulic processes,
mechanical processes, chemical processes, biological processes, radiological processes, and processes related togases, which are listed as FEPs representing specific phenomena in level 4.

6-A-6

Table 3 Integrated FEP list
FEP No.
IF1
IF2
IF3
IF4
IF5
IF6
IF7
IF8
IF9
IF10
IF11
IF12
IF13
IF14
IF15
IF16
IF17
IF18
IF19
IF20
IF21
IF22
IF23
IF24
IF25
IF26
IF27
IF28
IF29

FEP name
Orogeny
Deformation of geosphere
Seismicity
Volcanic and magmatic activity
Hydrothermal activity
Regional erosion and sedimentation
Diagenesis
Climate changeRadiolysis
Radiation damage
Thermal processes
Resaturation/desaturation
Package deformation
Package movement
Water chemistry
Corrosion
Polymer degradation
Dissolution
Precipitation reactions
Chemical alteration
Microbially/biologically mediated processes
Gas phase formation
Water transport
Piping/hydraulic erosion
Material volume changes
Creep
Undetected geosphere features
Hydraulic effects of repository
Mechanical effects of repository

6-A-7

3

Table 4

Excavation damaged zone (EDZ) flow rate (m /y)Rock type
Plutonic
Neogene
Pre-Neogene-1-2
HLW (H12V)
8×10
7×10
6×10-2-1-2
HLW (PEM)
8×10
7×10
6×10-2TRU Gr.1
4×10
4
3
2
2
TRU Gr.2
8×10
1×10
6×10
Repository TRU Gr.3
4×102
5×10
3×10
2
TRU Gr.4H
Gr.4HD
9×10
1×10
7
2
Gr.4HH
2×10
2×10
1×10
TRU Gr.4L
Gr.4LC
4×10
3
3
2
Gr.4LD
4×10
4×10
3×10
* Gr.4HD: Group 4H (Drum), Gr.4HH: Group 4H (MHHRW), Gr.4LD: Group 4L (Drum), Gr.4LC: Group 4H(Box container)
In Tables 6-5 to 6-23, high level radioactive waste is abbreviated as HLW, TRU waste is abbreviated as TRU,
and groups of TRU are abbreviated as Gr.

6-A-8

Table 5 Buffer porewater chemistry
Rock type
Plutonic
Neogene sediments
Groundwater
Low salinity
High salinity
Low salinity
High salinitypH
7.2~9.6
7.1~8.4
7.2 ~9.8
6.3~6.5222
22
2
Eh (mV)-4.1×10 ~ -2.0×10-3.2×10 ~ -2.0×10-3.9×10 ~ -1.9×10-1.7×102 ~ -1.4×102
-3
-1
-2
-1
-3
-1
Ionic strength
4.2×10 ~ 2.1×10
5.1×10 ~ 2.5×10
3.6×10 ~ 2.1×10
2.4×10-1 ~ 4.0×10-1
Na (total)
3.1×10-3~ 1.6×10-1
1.7×10-2~ 2.0×10-1
2.8×10-3~1.6×10-1
2.2×10-1~3.4×10-1
-5
-3
-4
-2
-5
-3
Ca (total)
1.5×10 ~ 4.3×10
4.7×10 ~ 1.6×10
1.4×10 ~ 4.2×10
3.5×10-3 ~ 1.5×10-2
K (total)
1.6×10-5 ~ 8.3×10-4
1.0×10-4 ~ 1.0×10-3
3.0×10-5 ~8.3×10-4
1.9×10-3 ~ 3.2×10-3
-7
-4
-5
-4
-7
-4
Mg (total)
4.6×10 ~ 3.7×10
1.7×10 ~ 5.3×10
4.0×10 ~3.0×10
1.4×10-3 ~ 4.9×10-3
Fe (total)
7.0×10-7 ~ 1.1×10-6
7.6×10-9 ~ 5.0×10-7
8.4×10-7 ~ 8.6×10-7
1.7×10-11 ~ 3.3×10-5-7-7-8Al (total)
7.9×102.8×102.2×10
1.3×10-9
Si (total)
4.8×10-4 ~ 1.4×10-3
4.7×10-4 ~ 5.8×10-4
3.3×10-4 ~ 9.4×10-4
3.1×10-4 ~3.2×10-4
-6
-2
-5
-2
-4
-2
S (total)
7.2×10 ~ 7.8×10
2.0×10 ~ 7.8×10
1.2×10 ~ 7.8×10
4.1×10-6 ~ 7.7×10-2
TIC
9.6×10-4 ~ 7.9×10-3
1.2×10-4 ~ 4.6×10-3
1.6×10-3 ~ 9.1×10-3
3.1×10-2 ~ 4.0×10-2
-3
-3
-2
-2
-3
-3
Cl (total)
2.3×10 ~ 4.5×10
4.9×10 ~ 5.1×10
1.1×10 ~ 3.3×10
2.1×10-1
F (total)
5.7×10-4
1.3×10-4
1.9×10-4
6.5×10-6-6-4-6
B (total)
4.6×10
2.7×10
4.6×10
1.0×10-2
P (total)
6.5×10-7
5.3×10-6
5.6×10-6
5.9×10-6-5-5-6N (total)
2.0×102.8×10
2.5×10
1.0×10-2
Br (total)
4.3×10-6
3.9×10-5
4.4×10-6
8.0×10-4-6-6-6
I (total)
7.9×10
5.5×10
3.9×10
1.8×10-4*Solute concentrations and ionic strength in mol/l

6-A-9

Pre-Neogene sediments
Low salinity
High salinity
7.1~9.6
6.1~6.3
22-4.0×10 ~ -2.0×10-1.7×102 ~ -1.4×102-3-1
3.6×10 ~ 2.0×10
2.4×10-1 ~ 4.0×10-1
2.8×10-3~1.6×10-1
2.2×10-1 ~ 3.4×10-1-5-3
1.4×10 ~ 4.3×10
3.5×10-3 ~ 1.5×10-2
3.0×10-5 ~ 8.2×10-4
1.9×10-3 ~ 3.2×10-3-7-4
4.1×10 ~ 3.6×10
1.6×10-3 ~ 4.9×10-3
8.3×10-7 ~ 9.0×10-7
1.7×10-11 ~ 3.3×10-5-8
4.3×10
2.4×10-9
4.8×10-4 ~ 1.3×10-3
4.4×10-4 ~ 4.6×10-4-4-2
1.2×10 ~ 7.8×10
4.1×10-6 ~ 7.7×10-2
1.7×10-3 ~ 8.6×10-3
3.6×10-2 ~ 4.7×10-2-3-3
1.1×10 ~ 3.3×10
2.1×10-1
1.9×10-4
6.5×10-6-6
4.6×10
1.0×10-2
5.6×10-6
5.9×10-6-6
2.5×10
1.0×10-2
4.4×10-6
8.0×10-4-6
3.9×10
1.8×10-4

Table 6 Porewater chemistry inside HLW overpack
Rock type
Plutonic
Neogene sediments
Groundwater
Low salinity
High salinity
Low salinity
High salinitypH
7.2 ~ 9.6
7.1 ~ 8.4
7.2 ~ 9.8
6.4 ~ 6.622
2
2
2
Eh (mV)-4.1×10 ~ -1.5×10-3.2×10 ~ -1.7×10-3.9×10 ~ -8.9×10-1.7×102~ -1.3×102
Ionic
4.2×10-3 ~2.1×10-1
5.1×10-2 ~ 2.5×10-1
3.6×10-3 ~ 2.1×10-1
2.4×10-1 ~ 4.0×10-1
strength
Na (total)
3.1×10-3 ~ 1.6×10-1
1.7×10-2 ~ 2.0×10-1
2.8×10-3 ~ 1.6×10-1
2.2×10-1 ~ 3.4×10-1
Ca (total)
1.5×10-5 ~ 4.3×10-3
4.7×10-4 ~ 1.6×10-2
1.4×10-5 ~ 4.2×10-3
3.5×10-3 ~ 1.5×10-2
-5
-4
-4
-3
-5
-4
K (total)
1.6×10 ~ 8.3×10
1.0×10 ~ 1.0×10
3.0×10 ~ 8.3×10
1.9×10-3 ~ 3.2×10-3
Mg (total)
4.6×10-7 ~ 3.7×10-4
1.7×10-5 ~ 5.3×10-4
4.0×10-7 ~ 3.0×10-4
1.4×10-3 ~ 4.9×10-3
-9
-6
-8
-6
-9
-6
Fe (total)
2.1×10 ~ 1.6×10
5.1×10 ~ 4.2×10
1.3×10 ~ 1.2×10
5.4×10-4 ~ 8.9×10-4
Al (total)
7.9×10-7
2.8×10-7
2.2×10-8
1.3×10-9
-4
-3
-4
-4
-4
-4
Si (total)
4.8×10 ~ 1.4×10
4.7×10 ~ 5.8×10
3.3×10 ~ 9.4×10
3.1×10-4 ~ 3.2×10-4
S (total)
7.2×10-6 ~ 7.8×10-2
2.0×10-5 ~ 7.8×10-2
1.2×10-4 ~ 7.8×10-2
4.1×10-6 ~ 7.7×10-2
-4
-3
-4
-3
-3
-3
TIC
9.6×10 ~ 7.9×10
1.2×10 ~ 4.6×10
1.6×10 ~ 9.1×10
3.1×10-2 ~ 4.0×10-2
Cl (total)
2.3×10-3 ~ 4.5×10-3
4.9×10-2 ~ 5.1×10-2
1.1×10-3 ~ 3.3×10-3
2.1×10-1-4-4-4
F (total)
5.7×10
1.3×10
1.9×10
6.5×10-6-6-4-6
B (total)
4.6×10
2.7×10
4.6×10
1.0×10-2
P (total)
6.5×10-7
5.3×10-6
5.6×10-6
5.9×10-6-5-5-6N (total)
2.0×102.8×10
2.5×10
1.0×10-2
Br (total)
4.3×10-6
3.9×10-5
4.4×10-6
8.0×10-4-6-6-6
I (total)
7.9×10
5.5×10
3.9×10
1.8×10-4*Solute concentrations and ionic strength in mol/l

6-A-10Pre-Neogene sediments
Low salinity
High salinity
7.1 ~ 9.6
6.2 ~ 6.422
-4.0×10 ~ -1.3×10-1.7×102 ~ -1.4×102
3.6×10-3 ~ 2.0×10-1

2.4×10-1 ~ 4.0×10-1

2.8×10-3 ~ 1.6×10-1
1.4×10-5 ~ 4.3×10-3
3.0×10-5 ~ 8.2×10-4
4.1×10-7 ~ 3.6×10-4
1.7×10-9 ~ 1.2×10-6
4.3×10-8
4.8×10-4 ~ 1.3×10-3
1.2×10-4 ~ 7.8×10-2
1.7×10-3 ~ 8.6×10-3
1.1×10-3 ~ 3.3×10-3
1.9×10-4
4.6×10-6
5.6×10-6
2.5×10-6
4.4×10-6
3.9×10-6

2.2×10-1 ~ 3.4×10-1
3.5×10-3 ~ 1.5×10-2
1.9×10-3 ~ 3.2×10-3
1.6×10-3 ~ 4.9×10-3
4.1×10-4 ~ 7.5×10-4
2.4×10-9
4.4×10-4 ~ 4.6×10-4
4.1×10-6 ~ 7.7×10-2
3.6×10-2 ~ 4.7×10-2
2.1×10-1
6.5×10-6
1.0×10-2
5.9×10-6
1.0×10-2
8.0×10-4
1.8×10-4

Table 7 Porewater chemistry within and between waste packages (Gr.1, 2, 4H) (1/3)
Rock type
Plutonic
Groundwater
Low salinity
High salinity
Porewater chemistry
Region Ⅰ
Region Ⅱ
Region Ⅰ
Region Ⅱ
pH
12.7
11.8 ~ 12.4
12.6 ~ 12.7
11.7 ~ 12.4
Eh (mV)-6.0×102 ~ -5.7×102-5.6×102 ~ -5.4×102-6.0×102 ~ -5.9×102-5.7×102 ~ -5.4×102
Ionic strength
3.7×10-1 ~ 5.3×10-1
4.6×10-2 ~ 1.7×10-1
3.8×10-1 ~ 5.6×10-1
9.8×10-2 ~ 2.0×10-1
-1
-1
-3
-1
-1
-1
Na (total)
1.4×10 ~ 3.0×10
3.1×10 ~ 1.6×10
1.6×10 ~ 3.4×10
1.7×10-2 ~ 2.0×10-1
Ca (total)
1.1×10-3 ~ 1.3×10-3
2.5×10-3 ~ 1.8×10-2
1.1×10-3 ~ 1.5×10-3
2.9×10-3 ~ 3.2×10-2
-1
-5
-4
-1
K (total)
2.2×10
1.6×10 ~ 8.3×10
2.2×10
1.0×10-4 ~ 1.0×10-3
Mg (total)
3.8×10-7~4.4×10-7
7.4×10-7 ~ 3.1×10-6
3.9×10-7 ~ 4.8×10-7
8.1×10-7 ~ 4.5×10-6
-6
-6
-7
-7
-6
-6
Fe (total)
1.3×10 ~3.5×10
3.8×10 ~ 9.0×10
2.3×10 ~ 3.5×10
4.1×10-7 ~ 1.2×10-6
Al (total)
4.7×10-5~6.0×10-5
5.4×10-6 ~ 2.5×10-5
4.4×10-5 ~ 5.8×10-5
4.5×10-6 ~ 2.3×10-5
-5
-5
-5
-5
-5
-5
Si (total)
5.2×10 ~9.2×10
1.0×10 ~ 3.2×10
5.0×10 ~ 9.6×10
1.0×10-5 ~ 3.3×10-5
S (total)
4.4×10-2~8.4×10-2
3.1×10-4 ~ 8.2×10-3
3.8×10-2 ~ 8.4×10-2
2.8×10-4 ~ 8.1×10-3
-4
-4
-6
-5
-4
-4
TIC
1.5×10 ~3.1×10
8.5×10 ~ 4.7×10
1.3×10 ~ 3.2×10
8.3×10-6 ~ 4.7×10-5
Cl (total)
2.3×10-3 ~ 4.5×10-3
2.3×10-3 ~ 4.6×10-3
4.9×10-2 ~ 5.1×10-2
4.9×10-2 ~ 5.2×10-2-4-4-4
F (total)
5.7×10
5.7×10
1.3×10
1.3×10-4
B (total)
4.6×10-6
4.6×10-6 ~ 4.7×10-6
2.7×10-4
2.7×10-4 ~ 2.8×10-4
-7
-7
-7
-6
-6
P (total)
6.4×10 ~ 6.5×10
6.5×105.2×10 ~ 5.3×10
5.3×10-6
N (total)
2.0×10-5
2.0×10-5 ~ 2.1×10-5
2.8×10-5
2.8×10-5
-6
-6
-6
-5
Br (total)
4.3×10
4.3×10 ~ 4.4×10
3.9×10
3.9×10-5
I (total)
7.8×10-6~ 7.9×10-6
7.9×10-6 ~ 8.0×10-6
5.5×10-6
5.5×10-6 ~ 5.6×10-6*Solute concentrations and ionic strength in mol/l

6-A-11

Table 7 Porewater chemistry within and between waste packages (Gr.1, 2, 4H) (2/3)
Rock type
Neogene sediments
Groundwater
Low salinity
High salinity
Porewater chemistry
Region Ⅰ
Region Ⅱ
Region Ⅰ
Region ⅡpH
13.2 ~ 13.3
12.3 ~ 12.9
13.2 ~ 13.3
12.3 ~ 12.8
Eh (mV)-5.5×102 ~ -5.4×102-5.4×102 ~ -5.3×102-6.2×102-5.9×102 ~ -5.7×102
Ionic strength
3.6×10-1 ~ 5.2×10-1
5.1×10-2 ~ 1.6×10-1
5.7×10-1 ~ 6.9×10-1
2.8×10-1~ 3.5×10-1
-1
-1
-3
-1
-1
-1
Na (total)
1.4×10 ~ 3.0×10
2.8×10 ~ 1.6×10
3.6×10 ~ 4.9×10
2.2×10-1 ~ 3.5×10-1
Ca (total)
7.5×10-4 ~ 1.0×10-3
2.5×10-3 ~ 1.9×10-2
8.8×10-4 ~ 1.2×10-3
4.1×10-3 ~ 2.4×10-2
-1
-5
-4
-1
K (total)
2.2×10
3.0×10 ~ 8.4×10
2.2×10
1.9×10-3 ~ 3.2×10-3
Mg (total)
2.1×10-7 ~ 2.7×10-7
5.2×10-7 ~ 2.1×10-6
2.4×10-7 ~ 3.0×10-7
7.1×10-7 ~ 2.3×10-6
-6
-6
-7
-6
-5
-5
Fe (total)
1.2×10 ~ 1.5×10
7.5×10 ~ 1.1×10
2.4×10 ~ 2.6×10
3.5×10-6 ~ 8.4×10-6
Al (total)
3.5×10-5 ~ 4.8×10-5
3.7×10-6 ~ 1.6×10-5
3.4×10-5 ~ 4.5×10-5
4.8×10-6 ~ 1.4×10-5
-5
-4
-5
-5
-5
-4
Si (total)
6.5×10 ~ 1.3×10
1.0×10 ~ 3.2×10
9.4×10 ~ 1.5×10
1.6×10-5 ~ 3.9×10-5
S (total)
1.2×10-2 ~ 2.7×10-2
7.2×10-5 ~ 1.7×10-3
1.5×10-2 ~ 2.9×10-2
2.0×10-4 ~ 1.9×10-3
-4
-4
-6
-5
-4
-4
TIC
1.9×10 ~ 4.6×10
8.2×10 ~ 4.6×10
2.6×10 ~ 5.1×10
1.1×10-5 ~ 4.9×10-5
Cl (total)
1.1×10-3 ~ 3.3×10-3
1.1×10-3 ~ 3.3×10-3
2.1×10-1
2.1×10-1
-4
-4
-6
-6
F (total)
1.9×10
1.9×10
6.5×10 ~ 6.6×10
6.5×10-6 ~ 6.6×10-6
B (total)
4.6×10-6~ 4.7×10-6
4.6×10-6~ 4.7×10-6
1.0×10-2
1.0×10-2
-6
-6
-6
-6
-6
-6
P (total)
5.6×10 ~ 5.7×10
5.6×10 ~ 5.7×10
5.9×10 ~ 6.0×10
5.9×10-6 ~ 6.0×10-6
N (total)
2.5×10-6 ~ 2.6×10-6
2.5×10-6 ~ 2.6×10-6
1.0×10-2
1.0×10-2
-6
-6
-6
-6
-4
-4
Br (total)
4.4×10 ~ 4.5×10
4.4×10 ~ 4.5×10
8.0×10 ~ 8.1×10
8.0×10-4 ~ 8.1×10-4
I (total)
3.9×10-6 ~ 4.0×10-6
3.9×10-6 ~ 4.0×10-6
1.8×10-4
1.8×10-4 ~ 1.9×10-4*Solute concentrations and ionic strength in mol/l

6-A-12

Table 7 Porewater chemistry within and between waste packages (Gr.1, 2, 4H) (3/3)
Rock type
Pre-Neogene sediments
Groundwater
Low salinity
High salinity
Porewater chemistry
Region Ⅰ
Region Ⅱ
Region Ⅰ
Region Ⅱ
pH
12.7
11.8 ~ 12.4
12.6 ~ 12.7
11.8 ~ 12.3
Eh (mV)-5.8×102 ~ -5.7×102-5.7×102 ~ -5.4×102-6.2×102-6.0×102 ~ -5.7×102
Ionic strength
3.7×10-1 ~ 5.3×10-1
4.5×10-2 ~ 1.7×10-1
5.7×10-1 ~ 7.0×10-1
2.7×10-1 ~ 3.5×10-1
-1
-1
-3
-1
-1
-1
Na (total)
1.4×10 ~ 3.0×10
2.8×10 ~ 1.6×10
3.6×10 ~ 4.8×10
2.2×10-1 ~ 3.4×10-1
Ca (total)
1.1×10-3 ~ 1.3×10-3
2.5×10-3 ~ 1.7×10-2
1.3×10-3 ~ 1.6×10-3
4.0×10-3 ~ 2.2×10-2
-1
-5
-4
-1
K (total)
2.2×10
3.0×10 ~ 8.3×10
2.2×10
1.9×10-3 ~ 3.2×10-3
Mg (total)
3.8×10-7 ~ 4.4×10-7
7.3×10-7 ~ 3.0×10-6
4.3×10-7 ~ 5.1×10-7
1.0×10-6 ~ 3.3×10-6
-6
-6
-7
-7
-6
-6
Fe (total)
1.2×10 ~ 1.9×10
3.5×10 ~ 9.5×10
7.1×10 ~ 7.5×10
1.2×10-6 ~ 3.0×10-6
Al (total)
4.7×10-5 ~ 6.0×10-5
5.4×10-6 ~ 2.5×10-5
4.5×10-5 ~ 5.5×10-5
7.1×10-6 ~ 2.1×10-5
-5
-5
-5
-5
-5
-4
Si (total)
5.2×10 ~ 9.2×10
1.0×10 ~ 3.2×10
7.5×10 ~ 1.1×10
1.6×10-5 ~ 3.9×10-5
S (total)
4.4×10-2 ~ 8.4×10-2
3.2×10-4 ~ 8.3×10-3
5.1×10-2 ~ 8.7×10-2
9.1×10-4 ~ 8.6×10-3
-4
-4
-6
-5
-4
-4
TIC
1.5×10 ~ 3.1×10
8.5×10 ~ 4.7×10
2.0×10 ~ 3.4×10
1.1×10-5 ~ 4.9×10-5
Cl (total)
1.1×10-3 ~ 3.3×10-3
1.1×10-3 ~ 3.3×10-3
2.1×10-1
2.1×10-1-4-4-6
F (total)
1.9×10
1.9×10
6.5×10
6.5×10-6 ~ 6.6×10-6
B (total)
4.6×10-6
4.6×10-6 ~ 4.7×10-6
1.0×10-2
1.0×10-2
-6
-6
-6
-6
P (total)
5.6×10
5.6×10 ~ 5.7×10
5.9×10
5.9×10-6 ~ 6.0×10-6
N (total)
2.5×10-6
2.5×10-6 ~ 2.6×10-6
1.0×10-2
1.0×10-2
-6
-6
-6
-4
-4
Br (total)
4.4×10
4.4×10 ~ 4.5×10
7.9×10 ~ 8.0×10
8.0×10-4 ~ 8.1×10-4
I (total)
3.9×10-6
3.9×10-6 ~ 4.0×10-6
1.8×10-4
1.8×10-4*Solute concentrations and ionic strength in mol/l

6-A-13

Table 8 Porewater chemistry within and between waste packages (Gr.4L) (1/2)
Rock type
Plutonic
Neogene sediments
Groundwater
Low salinity
High salinity
Low salinity
High salinity
Porewater chemistry
Region Ⅰ
Region Ⅱ
Region Ⅰ
Region Ⅱ
Region Ⅰ
Region Ⅱ
Region Ⅰ
Region Ⅱ
pH
12.7
11.8
12.6
11.7
13.2
12.3
13.2
12.3
Eh (mV)-5.9×102-5.4×102-6.0×102-5.5×102-5.4×102-5.4×102-6.2×102-5.7×102
Ionic strength
3.7×10-1
4.7×10-2
3.8×10-1
1.0×10-1
3.6×10-1
5.1×10-2
5.7×10-1
2.8×10-1
-1
-3
-1
-2
-1
-3
-1
Na (total)
1.4×10
3.1×10
1.6×10
1.7×10
1.4×102.8×10
3.6×102.2×10-1
Ca (total)
1.3×10-3
1.8×10-2
1.5×10-3
3.2×10-2
1.0×10-3
1.9×10-2
1.2×10-3
2.4×10-2
-1
-5
-1
-4
-1
-5
-1
K (total)
2.2×10
1.6×102.2×10
1.0×102.2×10
3.0×102.2×10
3.2×10-3
Mg (total)
4.4×10-7
3.1×10-6
4.8×10-7
4.5×10-6
2.7×10-7
2.1×10-6
3.0×10-7
2.3×10-6
-6
-7
-6
-7
-6
-7
-5
Fe (total)
2.6×10
4.0×10
3.5×10
4.3×10
1.3×10
7.8×10
2.4×10
3.5×10-6
Al (total)
4.7×10-5
5.4×10-6
4.4×10-5
4.5×10-6
3.5×10-5
3.7×10-6
3.4×10-5
4.8×10-6
-5
-5
-5
-5
-5
-5
-5
Si (total)
5.2×10
1.0×10
5.0×10
1.0×10
6.5×10
1.0×10
9.4×10
1.6×10-5
S (total)
4.4×10-2
3.1×10-4
3.8×10-2
2.8×10-4
1.2×10-2
7.2×10-5
1.5×10-2
2.0×10-4
-4
-6
-4
-6
-4
-6
-4
TIC
1.5×108.5×10
1.3×10
8.3×10
1.9×108.2×10
2.6×101.1×10-5
Cl (total)
2.3×10-3
2.3×10-3
4.9×10-2
4.9×10-2
1.1×10-3
1.1×10-3
2.1×10-1
2.1×10-1
-4
-4
-4
-4
-4
-4
-6
F (total)
5.7×10
5.7×10
1.3×10
1.3×10
1.9×10
1.9×10
6.5×10
6.5×10-6
B (total)
4.6×10-6
4.6×10-6
2.7×10-4
2.7×10-4
4.6×10-6
4.6×10-6
1.0×10-2
1.0×10-2
-7
-7
-6
-6
-6
-6
-6
P (total)
6.4×10
6.5×105.2×10
5.3×10
5.6×10
5.6×10
5.9×10
5.9×10-6
N (total)
2.0×10-5
2.0×10-5
2.8×10-5
2.8×10-5
2.5×10-6
2.5×10-6
1.0×10-2
1.0×10-2
-6
-6
-5
-5
-6
-6
-4
Br (total)
4.3×10
4.3×10
3.9×10
3.9×10
4.4×10
4.4×108.0×108.0×10-4
I (total)
7.8×10-6
7.9×10-6
5.5×10-6
5.5×10-6
3.9×10-6
3.9×10-6
1.8×10-4
1.8×10-4*Solute concentrations and ionic strength in mol/l

6-A-14

Table 8 Porewater chemistry within and between waste interpackage fillers (Gr.4L) (2/2)
Rock type
Pre-Neogene sediments
Groundwater
Low salinity
High salinity
Porewater chemistry
Region Ⅰ
Region Ⅱ
Region Ⅰ
Region Ⅱ
pH
12.7
11.8
12.6
11.8
Eh (mV)-5.7×102-5.4×102-6.2×102-5.7×102
Ionic strength
3.7×10-1
4.5×10-2
5.7×10-1
2.7×10-1-1-3-1
Na (total)
1.4×102.8×10
3.6×102.2×10-1
Ca (total)
1.3×10-3
1.7×10-2
1.6×10-3
2.2×10-2-1-5-1
K (total)
2.2×10
3.0×102.2×10
3.2×10-3
Mg (total)
4.4×10-7
3.0×10-6
5.1×10-7
3.3×10-6-6-7-6
Fe (total)
1.3×10
3.6×10
7.1×101.2×10-6
Al (total)
4.7×10-5
5.4×10-6
4.5×10-5
7.1×10-6-5-5-5Si (total)
5.2×10
1.0×10
7.5×10
1.6×10-5
S (total)
4.4×10-2
3.2×10-4
5.1×10-2
9.1×10-4
-4
-6
-4
TIC
1.5×108.5×10
2.0×101.1×10-5
Cl (total)
1.1×10-3
1.1×10-3
2.1×10-1
2.1×10-1-4-4-6
F (total)
1.9×10
1.9×10
6.5×10
6.5×10-6
B (total)
4.6×10-6
4.6×10-6
1.0×10-2
1.0×10-2-6-6-6
P (total)
5.6×10
5.6×10
5.9×10
5.9×10-6
N (total)
2.5×10-6
2.5×10-6
1.0×10-2
1.0×10-2-6-6-4
Br (total)
4.4×10
4.4×10
7.9×108.0×10-4
I (total)
3.9×10-6
3.9×10-6
1.8×10-4
1.8×10-4*Solute concentrations and ionic strength in mol/l

6-A-15

Table 9
Analysis case
Repository
GroundwaterC (inorganic)C (organic)
Cl
Co
Ni
Se
Sr
Zr
Nb
Mo
Tc
Pd
Sn
I
Cs
Pb
Ra
Ac
Th
Pa
U
Np
Pu
Am
Cm

Solubility and solubility limiting solid phases (Plutonic repository) (1/3)

HLW
Solubility limited solid phase
Solubility (mol/l)
Low salinity
High salinity
CaCO3 (calcite)
8×10-3
5×10-3(Not assessed)―
Not setSolubleSoluble(Not assessed)―(Not assessed)―
FeSe2(cr)
3×10-8
1×10-8
SrCO3(strontianite)
4×10-4
4×10-4Zr(OH)4(am,fresh)
1×10-8
1×10-8-5
Nb2O5(s)2×10
1×10-6(Not assessed)―
5×10-95×10-9TcO2・1.6H2O(s)(Not assessed)―
SnO2(am)
6×10-6
7×10-7
Not setSolubleSolubleNot setSolubleSolublePbCO3(cerussite)
2×10-6
4×10-6-4
RaCO3(cr)
8×10
8×10-4-5
AcCO3OH(am)
2×10
3×10-5
ThO2(am,aged)
4×10-82×10-8-9
Pa2O5(s)2×10
2×10-9UO2(am)
2×10-6
6×10-7
NpO2(am)
2×10-8
8×10-9PuO2(am)
2×10-8
3×10-8-7
2×10
4×10-7
AmCO3OH･0.5H2O(cr)2×10-7
4×10-7
CmCO3OH･0.5H2O(cr)Base case
TRU
Gr.1, 4H
Solubility limited solid phase
Solubility (mol/l)
Low salinity
High salinity(Not assessed)―
Not setSolubleSolubleNot setSolubleSolublebeta-Co(OH)2
4×10-6
4×10-6beta-Ni(OH)2
6×10-6
6×10-6
FeSe2(cr)
4×10-6
5×10-6
SrCO3(strontianite)
9×10-5
1×10-4Zr(OH)4(am,fresh)
1×10-72×10-7-2
Nb2O5(s)3×10
3×10-2
CaMoO4(cr)4×10-4
4×10-4-7
4×10
4×10-7
TcO2・1.6H2O(s)Pd(OH)2(am)9×10-7
9×10-7-2
SnO2(am)
4×10
4×10-2
Not setSolubleSolubleNot setSolubleSolublePb(OH)2(am)6×10-2
6×10-2-4
RaCO3(cr)
5×10
5×10-4-10
Ac(OH)3(am)7×10
7×10-10
ThO2(am,aged)
2×10-92×10-9-9
Pa2O5(s)1×10
1×10-9UO2(am)
5×10-7
5×10-7
NpO2(am)
1×10-9
1×10-9PuO2(am)
2×10-11
2×10-11
Am(OH)3(cr)4×10-11
4×10-11
Cm(OH)3(cr)4×10-11
4×10-11

6-A-16

Gr.2
Solubility (mol/l)
Low salinity
High salinity―
SolubleSolubleSolubleSoluble4×10-6
4×10-6
6×10-6
6×10-6
4×10-6
5×10-6
9×10-5
1×10-4
1×10-72×10-7-2
3×10
3×10-2
4×10-4
4×10-4-7
4×10
4×10-7
9×10-7
9×10-7-2
4×10
4×10-2
SolubleSolubleSolubleSoluble6×10-2
6×10-2-4
5×10
5×10-4-10
9×10
9×10-102×10-92×10-9-9
1×10
1×10-95×10-7
5×10-7
5×10-95×10-92×10-92×10-95×10-11
5×10-11
5×10-11
5×10-11

Gr.4L
Solubility (mol/l)
Low salinity
High salinity―
SolubleSolubleSolubleSoluble4×10-6
4×10-6
5×10-6
5×10-6
2×10-6
5×10-6
9×10-5
1×10-48×10-82×10-7-2
2×10
2×10-2
2×10-42×10-4-7
4×10
3×10-7
7×10-7
7×10-7-2
3×10
3×10-2
SolubleSolubleSolubleSoluble5×10-2
5×10-2-4
4×10
4×10-4-10
7×10
7×10-102×10-92×10-9-9
1×10
1×10-93×10-7
1×10-7
1×10-9
1×10-92×10-11
2×10-11
4×10-11
4×10-11
4×10-11
4×10-11

Table 9
Analysis case
Repository
GroundwaterC (inorganic)C (organic)
Cl
Co
Ni
Se
Sr
Zr
Nb
Mo
Tc
Pd
Sn
I
Cs
Pb
Ra
Ac
Th
Pa
U
Np
Pu
Am
Cm

Solubility and solubility limiting solid phases (Plutonic repository) (2/3)
Thermal increase in solubility – Variant case

HLW
Solubility (mol/l)
Low salinity High salinity
8×10-1
5×10-1(Not assessed)
SolubleSoluble(Not assessed)(Not assessed)
3×10-6
1×10-6-2
4×10
4×10-2-6
1×10
1×10-6-3
2×10
1×10-4(Not assessed)
5×10-7
5×10-7(Not assessed)
6×10-4
7×10-5
SolubleSolubleSolubleSoluble2×10-4
4×10-4-2
8×10
8×10-2-3
2×10
3×10-3-6
4×10
2×10-6-7
2×10
2×10-7-4
2×10
6×10-5-6
2×10
8×10-7-6
2×10
3×10-6-5
2×10
4×10-5-5
2×10
4×10-5

Gr.1, 4H
Solubility (mol/l)
Low salinity High salinity(Not assessed)
SolubleSolubleSolubleSoluble4×10-4
4×10-4-4
6×10
6×10-4-4
4×10
5×10-4-3
9×10
1×10-2-5
1×10
2×10-53
3
4×10-2
4×10-2
4×10-5
4×10-5-5
9×10
9×10-5
4
4
SolubleSolubleSolubleSoluble6
6
5×10-2
5×10-2
7×10-8
7×10-8-7
2×10
2×10-7-7
1×10
1×10-7-5
5×10
5×10-5-7
1×10
1×10-7-9
2×10
2×10-9-9
4×10
4×10-9-9
4×10
4×10-96-A-17

Gr.2
Solubility (mol/l)
Low salinity High salinity(Not assessed)
SolubleSolubleSolubleSoluble4×10-4
4×10-4-4
6×10
6×10-4-4
4×10
5×10-4-3
9×10
1×10-2-5
1×10
2×10-53
3
4×10-2
4×10-2
4×10-5
4×10-5-5
9×10
9×10-5
4
4
SolubleSolubleSolubleSoluble6
6
5×10-2
5×10-2
9×10-8
9×10-8-7
2×10
2×10-7-7
1×10
1×10-7-5
5×10
5×10-5-7
5×10
5×10-7-7
2×10
2×10-7-9
5×10
5×10-9-9
5×10
5×10-9

Gr.4L
Solubility (mol/l)
Low salinity High salinity(Not assessed)
SolubleSolubleSolubleSoluble4×10-4
4×10-4-4
5×10
5×10-4-4
2×10
5×10-4-3
9×10
1×10-2-6
8×10
2×10-52
2
2×10-2
2×10-2
4×10-5
3×10-5-5
7×10
7×10-53
3
SolubleSolubleSolubleSoluble5
5
4×10-2
4×10-2
7×10-8
7×10-8-7
2×10
2×10-7-7
1×10
1×10-7-5
3×10
1×10-5-7
1×10
1×10-7-9
2×10
2×10-9-9
4×10
4×10-9-9
4×10
4×10-9

Table 9

Analysis case
Repository
GroundwaterC (inorganic)C (organic)
Cl
Co
Ni
Se
Sr
Zr
Nb
Mo
Tc
Pd
Sn
I
Cs
Pb
Ra
Ac
Th
Pa
U
Np
Pu
Am
Cm

Solubility and solubility limiting solid phases (Plutonic repository) (3/3)
Uncertainty in thermodynamic data - Variant case
HLW
Gr.1, 4H
Gr.2
Gr.4L
Solubility (mol/l)
Solubility (mol/l)
Solubility (mol/l)
Solubility (mol/l)
Low salinity
High salinity Low salinity
High salinity Low salinity
High salinity
Low salinity
High salinity
8×10-3
5×10-3(Not assessed)(Not assessed)(Not assessed)(Not assessed)
SolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleSoluble(Not assessed)
8×10-6
8×10-6
8×10-6
8×10-6
8×10-6
8×10-6
-5
-5
-5
-5
-5
(Not assessed)
2×10
2×10
2×10
2×10
2×10
2×10-5
-6
-6
-5
-5
-5
-5
-5
3×10
2×10
3×10
3×10
3×10
3×10
2×10
3×10-5
-4
-4
-4
-4
-4
-4
-5
4×10
4×10
1×10
1×10
1×10
1×10
9×10
1×10-4
-8
-8
-7
-7
-7
-7
-7
2×10
2×10
2×10
2×10
2×10
2×10
2×10
2×10-7
-5
-6
-2
-2
-2
-2
-2
5×10
4×10
5×10
5×10
5×10
5×10
4×10
4×10-2
-4
-4
-4
-4
-4
(Not assessed)
4×10
4×10
4×10
4×10
3×10
3×10-4
-8
-8
-6
-6
-6
-6
-7
2×10
1×10
1×10
1×10
1×10
1×10
8×10
7×10-7
-6
-6
-6
-6
-6
(Not assessed)
5×10
5×10
5×10
5×10
4×10
4×10-6
-5
-6
-2
-2
-2
-2
-2
4×10
6×10
4×10
4×10
4×10
4×10
3×10
3×10-2
SolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleSoluble2×10-6
4×10-6
6×10-2
6×10-2
6×10-2
6×10-2
5×10-2
5×10-2
-3
-3
-4
-4
-4
-4
-4
2×10
2×10
8×10
8×10
8×10
8×10
6×10
6×10-4
SolubleSolubleSolubleSolubleSolubleSolubleSolubleSoluble3×10-6
9×10-7
8×10-8
8×10-8
8×10-8
8×10-8
8×10-8
8×10-8
-9
-9
-9
-9
-9
-9
-9
2×10
2×10
1×10
1×10
1×10
1×10
1×10
1×10-9
-5
-5
-5
-5
-5
-5
-5
4×10
3×10
2×10
2×10
2×10
2×10
1×10
9×10-6
-6
-7
-7
-7
-7
-7
-7
2×10
8×10
1×10
1×10
5×10
5×10
1×10
1×10-7
-7
-7
-10
-10
-8
-8
-10
2×10
3×10
2×10
2×10
3×10
3×10
2×10
2×10-10
-6
-6
-10
-10
-10
-10
-10
2×10
4×10
5×10
6×10
7×10
7×10
5×10
6×10-10
-6
-6
-10
-10
-10
-10
-10
2×10
4×10
5×10
6×10
7×10
7×10
5×10
6×10-10

6-A-18

Table 10
Analysis case
Repository

Base case
HLW
Solubility limited solid phase

GroundwaterC (inorganic)C (organic)
Cl
Co
Ni
Se
Sr
Zr
Nb
Mo
Tc
Pd
Sn
I
Cs
Pb
Ra
Ac
Th
Pa
U
Np
Pu
Am
Cm

Solubility and solubility limiting solid phases (Neogene repository) (1/3)

CaCO3(calcite)(Not assessed)
Not set(Not assessed)(Not assessed)
FeSe2(cr)SrCO3(strontianite)
Zr(OH)4(am,fresh)
Nb2O5(s)(Not assessed)
TcO2・1.6H2O(s)(Not assessed)
SnO2(am)
Not setNot setPbCO3(cerussite)
RaCO3(cr)AcCO3OH(am)ThO2(am,aged)
Pa2O5(s)UO2(am)NpO2(am)PuO2(am)
AmCO3OH･0.5H2O(cr)CmCO3OH･0.5H2O(cr)Solubility (mol/l)
Low salinity
1×10-2High salinity
4×10-2―

SolubleSoluble――3×10-8
4×10-4
1×10-8
3×10-5

5×10-11
2×10-38×10-8
4×10-8―

5×10-93×10-8―

9×10-6
SolubleSoluble3×10-6
8×10-42×10-5
5×10-82×10-93×10-6
3×10-82×10-8
8×10-8
8×10-8

4×10-7
SolubleSoluble3×10-6
6×10-3
9×10-5
7×10-7
8×10-93×10-5
4×10-7
4×10-7
6×10-7
6×10-7

TRU
Solubility limited solid
phase

Gr.1, 4H

Gr.2

Gr.4L

Solubility (mol/l)

Solubility (mol/l)

Solubility (mol/l)

Low salinity(Not assessed)
Not setNot setbeta-Co(OH)2
beta-Ni(OH)2
FeSe2(cr)SrCO3(strontianite)
Zr(OH)4(am,fresh)
Nb2O5(s)CaMoO4(cr)TcO2・1.6H2O(s)Pd(OH)2(am)SnO2(am)
Not setNot setPb(OH)2(am)RaCO3(cr)
Ac(OH)3(am)ThO2(am,aged)
Pa2O5(s)UO2(am)NpO2(am)PuO2(am)
Am(OH)3(cr)Cm(OH)3(cr)6-A-19High salinity

Low salinity―
SolubleSoluble6×10-6
2×10-52×10-6
1×10-42×10-6
8×10-2
5×10-42×10-6
3×10-6
2×10-1
SolubleSoluble2×10-1
3×10-4
6×10-102×10-9
1×10-92×10-6
1×10-92×10-11
3×10-11
3×10-11High salinity

Low salinity―
SolubleSoluble1×10-52×10-5
3×10-52×10-42×10-6
8×10-2
5×10-42×10-6
6×10-42×10-1
SolubleSoluble2×10-1
4×10-4
6×10-102×10-9
1×10-9
1×10-6
1×10-92×10-11
3×10-11
3×10-11

SolubleSoluble6×10-6
2×10-52×10-6
1×10-42×10-6
8×10-2
5×10-42×10-6
3×10-6
2×10-1
SolubleSoluble2×10-1
3×10-4
9×10-102×10-9
1×10-92×10-6
5×10-92×10-95×10-11
5×10-11High salinity―

SolubleSoluble1×10-52×10-5
3×10-52×10-42×10-6
8×10-2
5×10-42×10-6
6×10-42×10-1
SolubleSoluble2×10-1
4×10-4
9×10-102×10-9
1×10-9
1×10-6
5×10-92×10-94×10-11
4×10-11

SolubleSoluble5×10-6
2×10-5
7×10-7
8×10-5
6×10-7
6×10-2
3×10-42×10-6
3×10-6
1×10-1
SolubleSoluble2×10-1
3×10-4
6×10-102×10-9
1×10-92×10-6
1×10-92×10-11
3×10-11
3×10-11

SolubleSoluble1×10-52×10-5
3×10-52×10-4
9×10-7
6×10-2
4×10-42×10-6
6×10-4
1×10-1
SolubleSoluble2×10-1
4×10-4
6×10-102×10-9
1×10-94×10-7
1×10-92×10-11
3×10-11
3×10-11

Table 10

Solubility and solubility limiting solid phases (Neogene repository) (2/3)
Thermal increase in solubility - Variant case
HLW
Gr.1, 4H
Gr.2
Gr.4L
Solubility (mol/l)
Solubility (mol/l)
Solubility (mol/l)
Solubility (mol/l)Groundwater Low salinity High salinity Low salinity High salinity Low salinity High salinity Low salinity High salinity
C (inorganic)
14(Not assessed)(Not assessed)(Not assessed)
C (organic)(Not assessed)
SolubleSolubleSolubleSolubleSolubleSolubleCl
SolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleCo(Not assessed)
6×10-4
1×10-3
6×10-4
1×10-3
5×10-4
1×10-3
-3
-3
-3
-3
-3
Ni
(Not assessed)
2×10
2×10
2×10
2×10
2×10
2×10-3
-6
-9
-4
-3
-4
-3
-5
Se
3×10
5×10
2×10
3×10
2×10
3×10
7×10
3×10-3
-2
-1
-2
-2
-2
-2
-3
Sr
4×10
2×10
1×10
2×10
1×10
2×10
8×10
2×10-2
-6
-6
-4
-4
-4
-4
-5
Zr
1×10
8×10
2×10
2×10
2×10
2×10
6×10
9×10-5
-3
-6
Nb
3×10
4×10
8
8
8
8
6
6
Mo
(Not assessed)
5×10-2
5×10-2
5×10-2
5×10-2
3×10-2
4×10-2Tc
5×10-7
3×10-6
2×10-42×10-42×10-42×10-42×10-42×10-4
-4
-2
-4
-2
-4
Pd
(Not assessed)
3×10
6×10
3×10
6×10
3×10
6×10-2
-4
-5
Sn
9×10
4×10
2×10
2×10
2×10
2×10
1×10
1×10
I
SolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleCs
SolubleSolubleSolubleSolubleSolubleSolubleSolubleSolublePb
3×10-4
3×10-4
2×10
2×10
2×10
2×10
2×10
2×10
Ra
8×10-2
6×10-1
3×10-2
4×10-2
3×10-2
4×10-2
3×10-2
4×10-2Ac
2×10-3
9×10-3
6×10-8
6×10-8
9×10-8
9×10-8
6×10-8
6×10-8
-6
-5
-7
-7
-7
-7
-7
Th
5×10
7×10
2×10
2×10
2×10
2×10
2×10
2×10-7
-7
-7
-7
-7
-7
-7
-7
Pa
2×10
8×10
1×10
1×10
1×10
1×10
1×10
1×10-7
-4
-3
-4
-4
-4
-4
-4
U
3×10
3×10
2×10
1×10
2×10
1×10
2×10
4×10-5
-6
-5
-7
-7
-7
-7
-7
Np
3×10
4×10
1×10
1×10
5×10
5×10
1×10
1×10-7
-6
-5
-9
-9
-7
-7
-9
Pu
2×10
4×10
2×10
2×10
2×10
2×10
2×10
2×10-9
-6
-5
-9
-9
-9
-9
-9
Am
8×10
6×10
3×10
3×10
5×10
4×10
3×10
3×10-9
-6
-5
-9
-9
-9
-9
-9
Cm
8×10
6×10
3×10
3×10
5×10
4×10
3×10
3×10-9Analysis case
Repository

6-A-20Table 10

Solubility and solubility limiting solid phases (Neogene repository) (3/3)
Uncertainty in thermodynamic data - Variant case
HLW
Gr.1, 4H
Gr.2
Gr.4L
Solubility (mol/l)
Solubility (mol/l)
Solubility (mol/l)
Solubility (mol/l)Groundwater Low salinity High salinity Low salinity High salinity Low salinity High salinity Low salinity High salinity
C (inorganic)
1×10-2
4×10-2(Not assessed)(Not assessed)(Not assessed)
C (organic)(Not assessed)
SolubleSolubleSolubleSolubleSolubleSolubleCl
SolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleCo(Not assessed)
2×10-52×10-52×10-52×10-5
1×10-52×10-5
-5
-5
-5
-5
-5
Ni
(Not assessed)
8×10
8×10
8×10
8×10
6×10
6×10-5
-6
-8
-4
-4
-4
-4
-5
Se
1×10
3×10
2×10
3×10
2×10
3×10
8×10
3×10-4
-4
-3
-4
-4
-4
-4
-5
Sr
4×10
2×10
1×10
2×10
1×10
2×10
9×10
2×10-4
-8
-7
-6
-6
-6
-6
-6
Zr
2×10
2×10
2×10
2×10
2×10
2×10
1×10
2×10-6
-5
-7
-1
-1
-1
-1
-1
Nb
8×10
2×10
2×10
2×10
2×10
2×10
2×10
2×10-1
-4
-4
-4
-4
-4
Mo
(Not assessed)
6×10
6×10
6×10
6×10
4×10
4×10-4
-8
-8
-6
-6
-6
-6
-6
Tc
2×10
5×10
4×10
4×10
4×10
4×10
3×10
3×10-6
-5
-3
-5
-3
-5
Pd
(Not assessed)
2×10
1×10
2×10
1×10
2×10
1×10-3
-5
-6
-1
-1
-1
-1
-1
Sn
7×10
3×10
2×10
2×10
2×10
2×10
1×10
1×10-1I
SolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleCs
SolubleSolubleSolubleSolubleSolubleSolubleSolubleSolublePb
3×10-6
3×10-6
2×10-1
2×10-1
2×10-1
2×10-1
2×10-1
2×10-1
-3
-3
-4
-4
-4
-4
-4
Ra
2×10
8×10
6×10
8×10
6×10
8×10
5×10
7×10-4
Ac
SolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleTh
3×10-6
5×10-58×10-8
8×10-8
8×10-8
8×10-8
8×10-8
8×10-8
-9
-9
-9
-9
-9
-9
-9
Pa
2×10
8×10
1×10
1×10
1×10
1×10
1×10
1×10-9
-5
-4
-5
-5
-5
-5
-5
U
8×10
7×10
8×10
7×10
8×10
7×10
4×10
4×10-5
-6
-5
-7
-7
-7
-7
-7
Np
3×10
4×10
1×10
1×10
5×10
5×10
1×10
1×10-7
-7
-6
-10
-10
-8
-8
-10
Pu
2×10
3×10
2×10
2×10
3×10
3×10
2×10
2×10-10
-7
-6
-10
-10
-10
-10
-10
Am
8×10
6×10
5×10
5×10
7×10
7×10
5×10
5×10-10
-7
-6
-10
-10
-10
-10
-10
Cm
8×10
6×10
5×10
5×10
7×10
7×10
5×10
5×10-10Analysis case
Repository

6-A-21

Table 11

Analysis case
Repository

GroundwaterC (inorganic)C (organic)
Cl
Co
Ni
Se
Sr
Zr
Nb
Mo
Tc
Pd
Sn
I
Cs
Pb
Ra
Ac
Th
Pa
U
Np
Pu
Am
Cm

Solubility and solubility limiting solid phases (Pre-Neogene repository) (1/3)
Base case
HLW
TRU
Gr.1, 4H
Gr.2
Solubility limited solid
Solubility limited
Solubility (mol/l)
Solubility (mol/l)
Solubility (mol/l)
phase
solid phase
Low
Low
High
Low
High
High salinity
salinitysalinitysalinitysalinitysalinity
CaCO3(calcite)
9×10-3
5×10-2(Not assessed)――(Not assessed)―
Not setSolubleSolubleSolubleSolubleNot setSolubleSolubleNot setSolubleSolubleSolubleSoluble(Not assessed)―
beta-Co(OH)2
4×10-6
3×10-4
4×10-6
3×10-4-6-5-6(Not assessed)―
beta-Ni(OH)2
6×10
4×10
6×10
4×10-5
-8
-11
-6
-6
-6
FeSe2(cr)
3×10
8×10
FeSe2(cr)
2×10
8×10
2×10
8×10-6
SrCO3(strontianite
SrCO3(strontianite)
4×10-4
3×10-3
9×10-52×10-4
9×10-52×10-4)Zr(OH)4(am,fresh
Zr(OH)4(am,fresh)
1×10-8
8×10-8
1×10-7
1×10-7
1×10-7
1×10-7)
Nb2O5(s)2×10-5
4×10-8
Nb2O5(s)3×10-2
2×10-2
3×10-2
2×10-2-4-4-4(Not assessed)―
CaMoO4(cr)
4×10
4×10
4×10
4×10-4
-9
-8
-7
-7
-7
5×10
4×10
4×10
4×10
4×10
4×10-7
TcO2・1.6H2O(s)TcO2・1.6H2O(s)(Not assessed)―
Pd(OH)2(am)9×10-72×10-3
9×10-72×10-3
-6
-7
-2
-2
-2
SnO2(am)
5×10
4×10
SnO2(am)
4×10
3×10
4×10
3×10-2
Not setSolubleSolubleNot setSolubleSolubleSolubleSolubleNot setSolubleSolubleNot setSolubleSolubleSolubleSolublePbCO3(cerussite)
2×10-6
4×10-6
Pb(OH)2(am)6×10-2
5×10-2
6×10-2
5×10-2
-4
-3
-4
-4
-4
RaCO3(cr)
9×10
8×10
RaCO3(cr)
5×10
6×10
5×10
6×10-4
-5
-4
-10
-10
-10
AcCO3OH(am)
2×10
2×10
Ac(OH)3(am)
7×10
7×10
9×10
9×10-10
-8
-7
-9
-9
-9
ThO2(am,aged)
4×10
7×10
ThO2(am,aged)
2×10
2×10
2×10
2×10-9
-9
-8
-9
-9
-9
Pa2O5(s)2×10
2×10
Pa2O5(s)
1×10
1×10
1×10
1×10-9
-6
-5
-7
-8
-7
UO2(am)
2×10
2×10
UO2(am)
5×10
7×10
5×10
7×10-8
-8
-7
-9
-9
-9
NpO2(am)
2×10
4×10
NpO2(am)
1×10
1×10
5×10
5×10-9
-8
-5
-11
-11
-9
PuO2(am)
2×10
3×10
PuO2(am)
2×10
2×10
2×10
2×10-9
-7
-6
-11
-11
-11
2×10
2×10
Am(OH)3(cr)
4×10
4×10
5×10
5×10-11
AmCO3OH･0.5H2O(cr)CmCO3OH・
2×10-72×10-6
Cm(OH)3(cr)4×10-11
4×10-11
5×10-11
5×10-11
0.5H2O(cr)6-A-22

Gr.4L
Solubility (mol/l)LowsalinityHighsalinity―

SolubleSoluble4×10-6
5×10-6
9×10-7

SolubleSoluble3×10-4
4×10-5
7×10-6

9×10-52×10-48×10-8

1×10-72×10-2
2×10-4
4×10-7
7×10-7
3×10-2
SolubleSoluble5×10-2
4×10-4
7×10-102×10-9
1×10-93×10-7
1×10-92×10-11
4×10-11

2×10-2
3×10-4
4×10-72×10-3
3×10-2
SolubleSoluble5×10-2
5×10-4
7×10-102×10-9
1×10-94×10-8
1×10-92×10-11
4×10-11

4×10-11

4×10-11

Table 11

Analysis case
Repository
GroundwaterC (inorganic)C (organic)
Cl
Co
Ni
Se
Sr
Zr
Nb
Mo
Tc
Pd
Sn
I
Cs
Pb
Ra
Ac
Th
Pa
U
Np
Pu
Am
Cm

Solubility and solubility limiting solid phases (Pre-Neogene repository) (2/3)
Thermal increase in solubility - Variant case
HLW
Gr.1, 4H
Gr.2
Gr.4L
Solubility (mol/l)
Solubility (mol/l)
Solubility (mol/l)
Solubility (mol/l)
Low salinity High salinity Low salinity High salinity Low salinity High salinity Low salinity High salinity
9×10-15(Not assessed)(Not assessed)(Not assessed)(Not assessed)
SolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleSoluble(Not assessed)
4×10-4
3×10-2
4×10-4
3×10-2
4×10-4
3×10-2
-4
-3
-4
-3
-4
(Not assessed)
6×10
4×10
6×10
4×10
5×10
4×10-3
-6
-9
-4
-4
-4
-4
-5
3×10
8×10
2×10
8×10
2×10
8×10
9×10
7×10-4
-2
-1
-3
-2
-3
-2
-3
4×10
3×10
9×10
2×10
9×10
2×10
9×10
2×10-2
-6
-6
-5
-5
-5
-5
-6
1×10
8×10
1×10
1×10
1×10
1×10
8×10
1×10-5
-3
-6
2×10
4×10
3
2
3
2
2
2
(Not assessed)
4×10-2
4×10-2
4×10-2
4×10-2
2×10-2
3×10-2
5×10-7
4×10-6
4×10-5
4×10-5
4×10-5
4×10-5
4×10-5
4×10-5
-5
-1
-5
-1
-5
(Not assessed)
9×10
2×10
9×10
2×10
7×10
2×10-1
-4
-5
5×10
4×10
4
3
4
3
3
3
SolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleSoluble2×10-4
4×10-465
6
555
9×10-2
8×10-1
5×10-2
6×10-2
5×10-2
6×10-2
4×10-2
5×10-2
2×10-3
2×10-2
7×10-8
7×10-8
9×10-8
9×10-8
7×10-8
7×10-8
-6
-5
-7
-7
-7
-7
-7
4×10
7×10
2×10
2×10
2×10
2×10
2×10
2×10-7
-7
-6
-7
-7
-7
-7
-7
2×10
2×10
1×10
1×10
1×10
1×10
1×10
1×10-7
-4
-3
-5
-6
-5
-6
-5
2×10
2×10
5×10
7×10
5×10
7×10
3×10
4×10-6
-6
-5
-7
-7
-7
-7
-7
2×10
4×10
1×10
1×10
5×10
5×10
1×10
1×10-7
-6
-3
-9
-9
-7
-7
-9
2×10
3×10
2×10
2×10
2×10
2×10
2×10
2×10-9
-5
-4
-9
-9
-9
-9
-9
2×10
2×10
4×10
4×10
5×10
5×10
4×10
4×10-9
-5
-4
-9
-9
-9
-9
-9
2×10
2×10
4×10
4×10
5×10
5×10
4×10
4×10-96-A-23

Table 10

Solubility and solubility limiting solid phases (Pre-Neogene repository) (3/3)
Analysis case
Uncertainty in thermodynamic data - Variant case
Repository
HLW
Gr.1, 4H
Gr.2
Gr.4L
Solubility (mol/l)
Solubility (mol/l)
Solubility (mol/l)
Solubility (mol/l)Groundwater Low salinity High salinity Low salinity High salinity Low salinity High salinity Low salinity High salinity
C (inorganic)
9×10-3
5×10-2(Not assessed)(Not assessed)(Not assessed)
C (organic)(Not assessed)
SolubleSolubleSolubleSolubleSolubleSolubleCl
SolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleCo(Not assessed)
8×10-6
3×10-48×10-6
3×10-48×10-6
3×10-4
-5
-4
-5
-4
-5
Ni
(Not assessed)
2×10
1×10
2×10
1×10
2×10
1×10-4
-6
-8
-5
-4
-5
-4
-5
Se
2×10
5×10
3×10
7×10
3×10
7×10
2×10
7×10-4
-4
-3
-4
-4
-4
-4
-5
Sr
4×10
3×10
1×10
2×10
1×10
2×10
9×10
2×10-4
-8
-7
-7
-7
-7
-7
-7
Zr
2×10
2×10
2×10
2×10
2×10
2×10
2×10
2×10-7
-5
-7
-2
-2
-2
-2
-2
Nb
5×10
1×10
5×10
4×10
5×10
4×10
4×10
4×10-2
-4
-4
-4
-4
-4
Mo
(Not assessed)
4×10
4×10
4×10
4×10
3×10
3×10-4
-8
-8
-6
-7
-6
-7
-7
Tc
2×10
8×10
1×10
9×10
1×10
9×10
8×10
7×10-7
-6
-3
-6
-3
-6
Pd
(Not assessed)
5×10
2×10
5×10
2×10
4×10
2×10-3
-5
-6
-2
-2
-2
-2
-2
Sn
4×10
3×10
4×10
4×10
4×10
4×10
3×10
3×10-2I
SolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleCs
SolubleSolubleSolubleSolubleSolubleSolubleSolubleSolublePb
2×10-6
4×10-6
6×10-2
5×10-2
6×10-2
5×10-2
5×10-2
5×10-2
-3
-2
-4
-4
-4
-4
-4
Ra
2×10
1×10
8×10
9×10
8×10
9×10
6×10
8×10-4
Ac
SolubleSolubleSolubleSolubleSolubleSolubleSolubleSolubleTh
3×10-6
5×10-58×10-8
8×10-8
8×10-8
8×10-8
8×10-8
8×10-8
-9
-8
-9
-9
-9
-9
-9
Pa
2×10
2×10
1×10
1×10
1×10
1×10
1×10
1×10-9
-5
-4
-5
-6
-5
-6
-5
U
5×10
5×10
2×10
7×10
2×10
7×10
1×10
4×10-6
-6
-5
-7
-7
-7
-7
-7
Np
2×10
4×10
1×10
1×10
5×10
5×10
1×10
1×10-7
-7
-5
-10
-10
-8
-8
-10
Pu
2×10
9×10
2×10
2×10
3×10
3×10
2×10
2×10-10
-6
-5
-10
-10
-10
-10
-10
Am
2×10
2×10
5×10
6×10
7×10
8×10
5×10
6×10-10
-6
-5
-10
-10
-10
-10
-10
Cm
2×10
2×10
5×10
6×10
7×10
8×10
5×10
6×10-10

6-A-24

2

Table 12 Effective diffusion coefficients of RNs in EBS materials (m /s) (Plutonic repository)
Components
BufferInfillAnalysis caseBase case
Variant caseBase case
Repository
HLW, TRU
HLW, TRU
TRU
Groundwater Low salinity High salinity
Low salinity
High salinity
Average salinity
C (inorganic)
3×10-11
3×10-11
8×10-11
8×10-11(Not assessed)
-10
-10
-10
C (organic)
3×10
3×10
7×10
7×10-108×10-10
-11
-11
-10
-10
Cl
5×10
5×10
2×10
2×10
8×10-10
-8
-10
-7
-10
Co
5×10
1×10
2×10
3×10
8×10-10Ni
5×10-8
1×10-102×10-7
3×10-108×10-10
-11
-11
-10
-10
Se
4×10
5×10
1×10
1×10
8×10-10Sr
4×10-8
3×10-92×10-7
9×10-98×10-10
-10
-10
-10
-10
Zr
1×10
1×10
3×10
3×10
8×10-10Nb
1×10-10
1×10-10
3×10-10
3×10-108×10-10
-11
-11
-11
-11
Mo
3×10
3×10
6×10
6×10
8×10-10Tc
1×10-10
1×10-10
3×10-10
3×10-108×10-10
-10
-10
-10
-10
Pd
1×10
1×10
3×10
3×10
8×10-10Sn
1×10-10
1×10-10
3×10-10
3×10-108×10-10
-11
-11
-10
-10
I
5×10
5×10
2×10
2×10
8×10-10Cs
8×10-92×10-92×10-8
5×10-98×10-10
-10
-9
-10
-9
Pb
1×10
1×10
3×10
3×10
8×10-10Ra
5×10-8
3×10-92×10-7
1×10-8
8×10-10
-9
-9
-9
-9
Ac
4×10
1×10
9×10
3×10
8×10-10Th
1×10-10
1×10-10
3×10-10
3×10-108×10-10
-8
-9
-7
-8
Pa
5×10
3×10
2×10
1×10
8×10-10U
1×10-10
1×10-10
3×10-10
3×10-108×10-10
-10
-10
-10
-10
Np
1×10
1×10
3×10
3×10
8×10-10Pu
4×10-9
1×10-99×10-93×10-98×10-10
-9
-9
-9
-9
Am
4×10
1×10
9×10
3×10
8×10-10Cm
4×10-9
1×10-99×10-93×10-98×10-10

6-A-25

2

Table 13 Effective diffusion coefficients of RNs in EBS materials (m /s) (Neogene repository)
Components
BufferInfillAnalysis caseBase case
Variant caseBase case
Repository
HLW, TRU
HLW, TRU
TRU
Groundwater
Low salinity
High salinity
Low salinity
High salinity
Average salinity
C (inorganic)
2×10-11
8×10-11
6×10-11
3×10-10(Not assessed)
-10
-10
-10
C (organic)
2×10
2×10
5×10
5×10-10
6×10-10Cl
4×10-11
5×10-11
9×10-11
8×10-11
6×10-10
-8
-10
-7
-9
Co
5×10
5×10
2×10
2×10
6×10-10Ni
5×10-8
5×10-102×10-72×10-9
6×10-10
-11
-11
-11
-11
Se
3×10
4×10
8×10
7×10
6×10-10Sr
4×10-8
4×10-102×10-72×10-9
6×10-10
-11
-11
-10
-10
Zr
8×10
7×10
3×10
3×10
6×10-10Nb
2×10-11
7×10-11
5×10-11
3×10-10
6×10-10
-11
-11
-11
-11
Mo
2×10
3×10
5×10
4×10
6×10-10Tc
8×10-11
7×10-11
3×10-10
3×10-10
6×10-10
-11
-10
-10
-10
Pd
8×10
3×10
3×10
7×10
6×10-10Sn
8×10-11
7×10-11
3×10-10
3×10-10
6×10-10
-11
-11
-11
-11
I
4×10
5×10
9×10
8×10
6×10-10Cs
7×10-97×10-102×10-82×10-9
6×10-10
-11
-11
-10
-10
Pb
8×10
7×10
3×10
3×10
6×10-10Ra
4×10-8
4×10-102×10-72×10-9
6×10-10
-9
-10
-9
-10
Ac
3×10
3×10
7×10
7×10
6×10-10Th
8×10-11
3×10-11
3×10-10
4×10-11
6×10-10
-8
-10
-7
-9
Pa
5×10
7×10
2×10
2×10
6×10-10U
8×10-11
3×10-11
3×10-10
4×10-11
6×10-10
-11
-11
-10
-11
Np
8×10
3×10
3×10
4×10
6×10-10Pu
3×10-93×10-10
7×10-97×10-10
6×10-10
-9
-10
-9
-10
Am
3×10
3×10
7×10
7×10
6×10-10Cm
3×10-93×10-10
7×10-97×10-10
6×10-10

6-A-26

2

Table 14 Effective diffusion coefficients of RNs in EBS materials (m /s) (Pre-Neogene repository)
Components
BufferInfillAnalysis caseBase case
Variant caseBase case
Repository
HLW, TRU
HLW, TRU
TRU
Groundwater Low salinity High salinity
Low salinity
High salinity
Average salinity
C (inorganic)
3×10-11
2×10-108×10-11
4×10-10(Not assessed)
-10
-10
-10
C (organic)
3×10
2×10
7×10
7×10-108×10-10Cl
5×10-11
7×10-11
2×10-10
1×10-108×10-10
-8
-10
-7
-9
Co
6×10
6×10
2×10
2×10
8×10-10Ni
6×10-8
6×10-102×10-72×10-98×10-10
-11
-11
-10
-11
Se
4×10
6×10
1×10
9×10
8×10-10Sr
5×10-8
5×10-102×10-72×10-98×10-10
-10
-11
-10
-10
Zr
1×10
9×10
3×10
3×10
8×10-10Nb
1×10-10
9×10-11
3×10-10
3×10-108×10-10
-11
-11
-11
-11
Mo
3×10
4×10
6×10
5×10
8×10-10Tc
1×10-10
9×10-11
3×10-10
3×10-108×10-10
-10
-10
-10
-10
Pd
1×10
5×10
3×10
9×10
8×10-10Sn
1×10-10
9×10-11
3×10-10
3×10-108×10-10
-11
-11
-10
-10
I
5×10
7×10
2×10
1×10
8×10-10Cs
9×10-99×10-102×10-82×10-98×10-10
-10
-10
-10
-10
Pb
1×10
5×10
3×10
9×10
8×10-10Ra
6×10-8
6×10-102×10-72×10-98×10-10
-9
-10
-9
-10
Ac
5×10
5×10
9×10
9×10
8×10-10Th
1×10-10
4×10-11
3×10-10
5×10-11
8×10-10
-8
-10
-7
-9
Pa
6×10
9×10
2×10
2×10
8×10-10U
1×10-10
4×10-11
3×10-10
5×10-11
8×10-10
-10
-11
-10
-11
Np
1×10
4×10
3×10
5×10
8×10-10Pu
5×10-95×10-10
9×10-99×10-108×10-10
-9
-10
-9
-10
Am
5×10
5×10
9×10
9×10
8×10-10Cm
5×10-95×10-10
9×10-99×10-108×10-10

6-A-27

3

Table 15
Components
Analysis case
Repository
GroundwaterC (inorganic)C (organic)
Cl
Co
Ni
Se
Sr
Zr
Nb
Mo
Tc
Pd
Sn
I
Cs
Pb
Ra
Ac
Th
Pa
U
Np
Pu
Am
Cm

Sorption distribution coefficients for engineered barrier materials (m /kg) (Plutonic repository)
BufferInfillBase case
Variant caseBase case
HLW, TRU
HLW, TRU
Gr.1, 4L
Gr.3
Low salinity High salinity
Low salinity
High salinity
Low salinity High salinity
Low salinity
High salinity
0000(Not assessed)(Not assessed)
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
3
3
7×10-1
7×10-1
1×10-3
1×10-3
1×10-3
1×10-33
3
7×10-1
7×10-1
1×10-3
1×10-3
1×10-3
1×10-3
-4
-4
-2
-2
-2
4×10
4×10
0
0
1×10
1×10
1×10
1×10-2
-3
-3
-4
-4
-3
-3
-3
2×10
2×10
9×10
9×10
1×10
1×10
1×10
1×10-3
-1
-1
-1
-1
-1
3
3
7×10
7×10
1×10
1×10
1×10
1×10-1
-1
-1
-2
-2
-2
-2
-2
2×10
2×10
5×10
5×10
2×10
2×10
2×10
2×10-2
-3
-3
-3
0
0
0
0
1×10
1×10
1×10
1×10-3
-1
-1
-5
5
5
1
1
1×10
1×10
2×10
2×10-5
-2
-2
-2
-2
-3
-3
-3
5×10
5×10
1×10
1×10
1×10
1×10
1×10
1×10-3
-1
-1
-1
-1
-1
3
3
7×10
7×10
1×10
1×10
1×10
1×10-1
-4
-4
-4
0
0
0
0
1×10
1×10
1×10
1×10-4
-2
-2
-2
-2
-4
-4
-4
6×10
6×10
3×10
3×10
2×10
2×10
2×10
2×10-4
-3
-3
-3
4×10
4×10
1×10
1×10
1×10
1×10
1×10
1×10-3
-3
-3
-4
-4
-3
-3
-3
2×10
2×10
9×10
9×10
1×10
1×10
1×10
1×10-3
-1
-1
-1
-1
-1
-1
-1
8×10
8×10
2×10
2×10
2×10
2×10
2×10
2×10-1
-1
-1
-1
-1
-1
-1
-1
8×10
8×10
2×10
2×10
2×10
2×10
2×10
2×10-1
-1
-1
-1
2×10
2×10
8
8
1×10
1×10
1×10
1×10-1
-3
-3
-4
-4
-2
-2
-2
2×10
2×10
4×10
4×10
2×10
2×10
2×10
2×10-2
-1
-1
-1
-1
-1
1
1
3×10
3×10
2×10
2×10
2×10
2×10-1
-1
-1
-1
1×10
1×10
3
3
2×10
2×10
2×10
2×10-1
-1
-1
-1
-1
-1
-1
-1
8×10
8×10
2×10
2×10
2×10
2×10
2×10
2×10-1
-1
-1
-1
-1
-1
-1
-1
8×10
8×10
2×10
2×10
2×10
2×10
2×10
2×10-1

6-A-28

3

Table 16
Components
Analysis case
Repository
GroundwaterC (inorganic)C (organic)
Cl
Co
Ni
Se
Sr
Zr
Nb
Mo
Tc
Pd
Sn
I
Cs
Pb
Ra
Ac
Th
Pa
U
Np
Pu
Am
Cm

Sorption distribution coefficients for engineered barrier materials (m /kg) (Neogene repository)
BufferInfillBase case
Variant caseBase case
HLW, TRU
HLW, TRU
Gr.1, 4L
Gr.3
Low salinity
High salinity
Low salinity
High salinity
Low salinity High salinity Low salinity High salinity
0000(Not assessed)(Not assessed)
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
3
3
7×10-1
7×10-1
1×10-3
1×10-3
1×10-3
1×10-33
3
7×10-1
7×10-1
1×10-3
1×10-3
1×10-3
1×10-3
-4
-4
-2
-2
-2
4×10
4×10
0
0
1×10
1×10
1×10
1×10-2
-3
-3
-4
-4
-3
-3
-3
2×10
6×10
9×10
5×10
1×10
1×10
1×10
1×10-3
-1
-1
-1
-1
-1
3
3
7×10
7×10
1×10
1×10
1×10
1×10-1
-1
-1
-2
-2
-2
-2
-2
2×10
2×10
5×10
5×10
2×10
2×10
2×10
2×10-2
-3
-3
-3
0
0
0
0
1×10
1×10
1×10
1×10-3
-1
-1
-5
5
5
1
1
1×10
1×10
2×10
2×10-5
-2
-2
-2
-2
-3
-3
-3
5×10
5×10
1×10
1×10
1×10
1×10
1×10
1×10-3
-1
-1
-1
-1
-1
3
3
7×10
7×10
1×10
1×10
1×10
1×10-1
-4
-4
-4
0
0
0
0
1×10
1×10
1×10
1×10-4
-2
-2
-2
-3
-4
-4
-4
6×10
4×10
3×10
7×10
2×10
2×10
2×10
2×10-4
-3
-3
-3
4×10
4×10
1×10
1×10
1×10
1×10
1×10
1×10-3
-3
-2
-4
-3
-3
-3
-3
2×10
2×10
9×10
4×10
1×10
1×10
1×10
1×10-3
-1
-1
-1
-1
-1
-1
-1
8×10
8×10
2×10
2×10
2×10
2×10
2×10
2×10-1
-1
-1
-1
-1
-1
-1
-1
8×10
8×10
2×10
2×10
2×10
2×10
2×10
2×10-1
-1
-1
-1
2×10
2×10
8
8
1×10
1×10
1×10
1×10-1
-3
-3
-4
-4
-2
-2
-2
2×10
2×10
4×10
4×10
2×10
2×10
2×10
2×10-2
-1
-1
-1
-1
-1
1
1
3×10
3×10
2×10
2×10
2×10
2×10-1
-1
-1
-1
1×10
1×10
3
3
2×10
2×10
2×10
2×10-1
-1
-1
-1
-1
-1
-1
-1
8×10
8×10
2×10
2×10
2×10
2×10
2×10
2×10-1
-1
-1
-1
-1
-1
-1
-1
8×10
8×10
2×10
2×10
2×10
2×10
2×10
2×10-1

6-A-29

3

Table 17

Components
Analysis case
Repository
GroundwaterC (inorganic)C (organic)
Cl
Co
Ni
Se
Sr
Zr
Nb
Mo
Tc
Pd
Sn
I
Cs
Pb
Ra
Ac
Th
Pa
U
Np
Pu
Am
Cm

Sorption distribution coefficients for engineered barrier materials (m /kg) (Pre-Neogene repository)
BufferInfillBase case
Variant caseBase case
HLW, TRU
HLW, TRU
Gr.1, 4L
Gr.3
Low salinity
High salinity
Low salinity
High salinity
Low salinity High salinity Low salinity
High salinity
0000(Not assessed)(Not assessed)
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
3
3
7×10-1
7×10-1
1×10-3
1×10-3
1×10-3
1×10-33
3
7×10-1
7×10-1
1×10-3
1×10-3
1×10-3
1×10-3
-4
-4
-2
-2
-2
4×10
4×10
0
0
1×10
1×10
1×10
1×10-2
-3
-3
-4
-4
-3
-3
-3
2×10
6×10
9×10
5×10
1×10
1×10
1×10
1×10-3
-1
-1
-1
-1
-1
3
3
7×10
7×10
1×10
1×10
1×10
1×10-1
-1
-1
-2
-2
-2
-2
-2
2×10
2×10
5×10
5×10
2×10
2×10
2×10
2×10-2
-3
-3
-3
0
0
0
0
1×10
1×10
1×10
1×10-3
-1
-1
-5
5
5
1
1
1×10
1×10
2×10
2×10-5
-2
-2
-2
-2
-3
-3
-3
5×10
5×10
1×10
1×10
1×10
1×10
1×10
1×10-3
-1
-1
-1
-1
-1
3
3
7×10
7×10
1×10
1×10
1×10
1×10-1
-4
-4
-4
0
0
0
0
1×10
1×10
1×10
1×10-4
-2
-2
-2
-3
-4
-4
-4
6×10
4×10
3×10
7×10
2×10
2×10
2×10
2×10-4
-3
-3
-3
4×10
4×10
1×10
1×10
1×10
1×10
1×10
1×10-3
-3
-2
-4
-3
-3
-3
-3
2×10
2×10
9×10
4×10
1×10
1×10
1×10
1×10-3
-1
-1
-1
-1
-1
-1
-1
8×10
8×10
2×10
2×10
2×10
2×10
2×10
2×10-1
-1
-1
-1
-1
-1
-1
-1
8×10
8×10
2×10
2×10
2×10
2×10
2×10
2×10-1
-1
-1
-1
2×10
2×10
8
8
1×10
1×10
1×10
1×10-1
-3
-3
-4
-4
-2
-2
-2
2×10
2×10
4×10
4×10
2×10
2×10
2×10
2×10-2
-1
-1
-1
-1
-1
1
1
3×10
3×10
2×10
2×10
2×10
2×10-1
-1
-1
-1
1×10
1×10
3
3
2×10
2×10
2×10
2×10-1
-1
-1
-1
-1
-1
-1
-1
8×10
8×10
2×10
2×10
2×10
2×10
2×10
2×10-1
-1
-1
-1
-1
-1
-1
-1
8×10
8×10
2×10
2×10
2×10
2×10
2×10
2×10-1

6-A-30Table 182

Effective diffusion coefficients of RNs in host rock (m /s) (Plutonic repository)
Analysis caseBase case
Variant case
Repository
HLW, TRU
HLW, TRU-13
All RNs
9×10
5×10-14

6-A-31

2

Table 19

Analysis case
Repository
GroundwaterC (inorganic)C (organic)
Cl
Co
Ni
Se
Sr
Zr
Nb
Mo
Tc
Pd
Sn
I
Cs
Pb
Ra
Ac
Th
Pa
U
Np
Pu
Am
Cm

Effective diffusion coefficients of RNs in host rock (m /s) (Neogene repository)
Base case
Variant case
HLW, Gr.1, 2, 4
Gr.3
HLW, Gr.1, 2, 4
Gr.3
Low salinity
High salinity
Low salinity
High salinity Low salinity High salinity
Low salinity
High salinity
1×10-11
1×10-11(Not assessed)
2×10-12
2×10-12(Not assessed)
6×10-11
6×10-11
1×10-11
1×10-11
3×10-12
3×10-12
2×10-12
2×10-12
-11
-11
-11
-11
-12
-12
-12
1×10
1×10
1×10
1×10
2×10
2×10
2×10
2×10-12
-11
-10
-11
-11
-12
-12
-12
6×10
1×10
1×10
1×10
3×10
3×10
2×10
2×10-12
-11
-10
-11
-11
-12
-12
-12
6×10
1×10
1×10
1×10
3×10
3×10
2×10
2×10-12
-11
-11
-11
-11
-12
-12
-12
1×10
1×10
1×10
1×10
2×10
2×10
2×10
2×10-12
-10
-10
-11
-11
-12
-12
-12
1×10
1×10
1×10
1×10
3×10
3×10
2×10
2×10-12
-11
-11
-11
-11
-12
-12
-12
6×10
1×10
1×10
1×10
3×10
2×10
2×10
2×10-12
-11
-11
-11
-11
-12
-12
-12
1×10
1×10
1×10
1×10
2×10
2×10
2×10
2×10-12
-11
-11
-11
-11
-12
-12
-12
1×10
1×10
1×10
1×10
2×10
2×10
2×10
2×10-12
-11
-11
-11
-11
-12
-12
-12
6×10
6×10
1×10
1×10
3×10
3×10
2×10
2×10-12
-11
-10
-11
-11
-12
-12
-12
6×10
1×10
1×10
1×10
3×10
3×10
2×10
2×10-12
-11
-11
-11
-11
-12
-12
-12
1×10
6×10
1×10
1×10
2×10
3×10
2×10
2×10-12
-11
-11
-11
-11
-12
-12
-12
1×10
1×10
1×10
1×10
2×10
2×10
2×10
2×10-12
-10
-10
-11
-11
-12
-12
-12
1×10
1×10
1×10
1×10
3×10
3×10
2×10
2×10-12
-11
-11
-11
-11
-12
-12
-12
6×10
6×10
1×10
1×10
3×10
3×10
2×10
2×10-12
-10
-10
-11
-11
-12
-12
-12
1×10
1×10
1×10
1×10
3×10
3×10
2×10
2×10-12
-11
-10
-11
-11
-12
-12
-12
1×10
1×10
1×10
1×10
2×10
3×10
2×10
2×10-12
-11
-11
-11
-11
-12
-12
-12
1×10
1×10
1×10
1×10
2×10
2×10
2×10
2×10-12
-11
-11
-11
-11
-12
-12
-12
1×10
1×10
1×10
1×10
2×10
2×10
2×10
2×10-12
-11
-11
-11
-11
-12
-12
-12
1×10
1×10
1×10
1×10
2×10
2×10
2×10
2×10-12
-11
-11
-11
-11
-12
-12
-12
1×10
1×10
1×10
1×10
2×10
2×10
2×10
2×10-12
-11
-10
-11
-11
-12
-12
-12
1×10
1×10
1×10
1×10
2×10
3×10
2×10
2×10-12
-11
-10
-11
-11
-12
-12
-12
1×10
1×10
1×10
1×10
2×10
3×10
2×10
2×10-12
-11
-10
-11
-11
-12
-12
-12
1×10
1×10
1×10
1×10
2×10
3×10
2×10
2×10-126-A-32

2

Table 20

Analysis case
Repository
GroundwaterC (inorganic)C (organic)
Cl
Co
Ni
Se
Sr
Zr
Nb
Mo
Tc
Pd
Sn
I
Cs
Pb
Ra
Ac
Th
Pa
U
Np
Pu
Am
Cm

Effective diffusion coefficients of RNs in host rock (m /s) (Pre-Neogene repository)
Base case
Variant case
HLW, Gr.1, 2, 4
Gr.3
HLW, Gr.1, 2, 4
Gr.3
Low salinity High salinity Low salinity High salinity
Low salinity
High salinity Low salinity High salinity
3×10-13
3×10-13(Not assessed)
4×10-14
4×10-14(Not assessed)
-12
-12
-13
-13
-14
3×10
3×10
3×10
3×10
9×10
9×10-14
4×10-14
4×10-14
-13
-13
-13
-13
-14
-14
-14
3×10
3×10
3×10
3×10
4×10
4×10
4×10
4×10-14
-12
-12
-13
-13
-14
-14
-14
3×10
2×10
3×10
3×10
9×10
9×10
4×10
4×10-14
-12
-12
-13
-13
-14
-14
-14
3×10
2×10
3×10
3×10
9×10
9×10
4×10
4×10-14
-13
-13
-13
-13
-14
-14
-14
3×10
3×10
3×10
3×10
4×10
4×10
4×10
4×10-14
-12
-12
-13
-13
-14
-14
-14
2×10
2×10
3×10
3×10
9×10
9×10
4×10
4×10-14
-12
-13
-13
-13
-14
-14
-14
3×10
3×10
3×10
3×10
9×10
4×10
4×10
4×10-14
-13
-13
-13
-13
-14
-14
-14
3×10
3×10
3×10
3×10
4×10
4×10
4×10
4×10-14
-13
-13
-13
-13
-14
-14
-14
3×10
3×10
3×10
3×10
4×10
4×10
4×10
4×10-14
-12
-12
-13
-13
-14
-14
-14
3×10
3×10
3×10
3×10
9×10
9×10
4×10
4×10-14
-12
-12
-13
-13
-14
-14
-14
3×10
2×10
3×10
3×10
9×10
9×10
4×10
4×10-14
-13
-12
-13
-13
-14
-14
-14
3×10
3×10
3×10
3×10
4×10
9×10
4×10
4×10-14
-13
-13
-13
-13
-14
-14
-14
3×10
3×10
3×10
3×10
4×10
4×10
4×10
4×10-14
-12
-12
-13
-13
-14
-14
-14
2×10
2×10
3×10
3×10
9×10
9×10
4×10
4×10-14
-12
-12
-13
-13
-14
-14
-14
3×10
3×10
3×10
3×10
9×10
9×10
4×10
4×10-14
-12
-12
-13
-13
-14
-14
-14
2×10
2×10
3×10
3×10
9×10
9×10
4×10
4×10-14
-12
-12
-13
-13
-14
-14
-14
2×10
2×10
3×10
3×10
9×10
9×10
4×10
4×10-14
-13
-13
-13
-13
-14
-14
-14
3×10
3×10
3×10
3×10
4×10
4×10
4×10
4×10-14
-13
-13
-13
-13
-14
-14
-14
3×10
3×10
3×10
3×10
4×10
4×10
4×10
4×10-14
-13
-13
-13
-13
-14
-14
-14
3×10
3×10
3×10
3×10
4×10
4×10
4×10
4×10-14
-13
-13
-13
-13
-14
-14
-14
3×10
3×10
3×10
3×10
4×10
4×10
4×10
4×10-14
-13
-12
-13
-13
-14
-14
-14
3×10
2×10
3×10
3×10
4×10
9×10
4×10
4×10-14
-12
-12
-13
-13
-14
-14
-14
2×10
2×10
3×10
3×10
9×10
9×10
4×10
4×10-14
-12
-12
-13
-13
-14
-14
-14
2×10
2×10
3×10
3×10
9×10
9×10
4×10
4×10-14

6-A-33

3

Table 21 Distribution coefficients for sorption on host rock (m /kg) (Plutonic repository) (1/2)
Analysis caseBase case
Repository
HLW, Gr.1, 2, 4
Gr.3 (< 1,000 y after repository closure)
Gr.3 (> 1,000 y after repository closure
Groundwater Low salinity High salinity
Low salinity
High salinity
Low salinity
High salinity
C (inorganic)
00(Not assessed)(Not assessed)
C (organic)
0
0
0
0
0
0
Cl
0
0
0
0
0
0
Co
1×10-1
1×10-1
1×10-5
1×10-5
1×10-5
1×10-5Ni
1×10-1
1×10-1
1×10-5
1×10-5
1×10-5
1×10-5
-3
-3
-3
-3
-3
Se
1×10
1×10
1×10
1×10
1×10
1×10-3
-3
-3
-6
-6
-5
Sr
8×10
8×10
8×10
8×10
8×10
8×10-5
-1
-1
-1
-1
-1
Zr
8×10
8×10
8×10
8×10
8×10
8×10-1
-1
-1
-1
-1
-1
Nb
4×10
4×10
4×10
4×10
4×10
4×10-1
Mo
0
0
0
0
0
0
Tc
8
8
8
8
8
8
Pd
6×10-1
6×10-1
6×10-3
6×10-3
6×10-3
6×10-3
Sn
1×10
1×10
1×10
1×10
1×10
1×10
I
0
0
0
0
0
0
Cs
4×10-2
4×10-2
4×10-5
4×10-5
4×10-4
4×10-4Pb
1
1
1×10-3
1×10-3
1×10-3
1×10-3
-1
-1
-4
-4
-3
Ra
1×10
1×10
1×10
1×10
1×10
1×10-3
Ac
1
1
1
1
1
1
Th
3×10-1
3×10-1
3×10-1
3×10-1
3×10-1
3×10-1
Pa
2
2
2
2
2
2
U
3×10-1
3×10-1
3×10-1
3×10-1
3×10-1
3×10-1Np
3×10-1
3×10-1
3×10-1
3×10-1
3×10-1
3×10-1
-1
-1
-1
-1
-1
Pu
3×10
3×10
3×10
3×10
3×10
3×10-1
Am
1
1
1
1
1
1
Cm
1
1
1
1
1
1

6-A-34

3

Table 21 Distribution coefficients for sorption on host rock (m /kg) (Plutonic repository) (2/2)
Analysis case
Variant case
Repository
HLW, Gr.1, 2, 4
Gr.3 (< 1,000 y after repository closure)
Gr.3 (> 1,000 y after repository closure)
Groundwater
Low salinity High salinity
Low salinity
High salinity
Low salinity
High salinity
C (inorganic)
00(Not assessed)(Not assessed)
C (organic)
0
0
0
0
0
0
Cl
0
0
0
0
0
0
Co
6×10-3
6×10-3
1×10-5
1×10-5
1×10-5
1×10-5Ni
6×10-3
6×10-3
1×10-5
1×10-5
1×10-5
1×10-5
-4
-4
-4
-4
-4
Se
3×10
3×10
3×10
3×10
3×10
3×10-4
-4
-4
-6
-6
-5
Sr
5×10
5×10
8×10
8×10
8×10
8×10-5
-1
-1
-1
-1
-1
Zr
1×10
1×10
1×10
1×10
1×10
1×10-1
-2
-2
-2
-2
-2
Nb
5×10
5×10
5×10
5×10
5×10
5×10-2
Mo
0
0
0
0
0
0
Tc
3×10-1
3×10-1
3×10-1
3×10-1
3×10-1
3×10-1Pd
4×10-2
4×10-2
6×10-3
6×10-3
6×10-3
6×10-3
-1
-1
-1
-1
-1
Sn
5×10
5×10
5×10
5×10
5×10
5×10-1
I
0
0
0
0
0
0
Cs
1×10-3
1×10-3
4×10-5
4×10-5
4×10-4
4×10-4Pb
2×10-1
2×10-1
1×10-3
1×10-3
1×10-3
1×10-3
-2
-2
-4
-4
-3
Ra
1×10
1×10
1×10
1×10
1×10
1×10-3
-2
-2
-2
-2
-2
Ac
7×10
7×10
7×10
7×10
7×10
7×10-2
-3
-3
-3
-3
-3
Th
8×10
8×10
8×10
8×10
8×10
8×10-3
-3
-3
-3
-3
-3
Pa
9×10
9×10
9×10
9×10
9×10
9×10-3
-3
-3
-3
-3
-3
U
8×10
8×10
8×10
8×10
8×10
8×10-3
-3
-3
-3
-3
-3
Np
8×10
8×10
8×10
8×10
8×10
8×10-3
-3
-3
-3
-3
-3
Pu
8×10
8×10
8×10
8×10
8×10
8×10-3
-2
-2
-2
-2
-2
Am
7×10
7×10
7×10
7×10
7×10
7×10-2
-2
-2
-2
-2
-2
Cm
7×10
7×10
7×10
7×10
7×10
7×10-2

6-A-35

3

Table 22 Distribution coefficients for sorption on host rock (m /kg) (Neogene repository) (1/2)
Analysis caseBase case
Repository
HLW, Gr.1, 2, 4
Gr.3 (< 1,000 y)
Gr.3 (> 1,000 y)
Groundwater
Low salinity High salinity
Low salinity
High salinity
Low salinity
High salinity
C (inorganic)
00(Not assessed)(Not assessed)
C (organic)
0
0
0
0
0
0
Cl
0
0
0
0
0
0
Co
3
5×10-1
3×10-4
5×10-5
3×10-4
5×10-5Ni
3
5×10-1
3×10-4
5×10-5
3×10-4
5×10-5
-2
-2
-2
-2
-2
Se
4×10
5×10
4×10
5×10
4×10
5×10-2
-1
-2
-4
-5
-3
Sr
2×10
7×10
2×10
7×10
2×10
7×10-4
Zr
2
0
2
0
2
0
Nb
6
6
6
6
6
6
Mo
0
0
0
0
0
0
Tc
1×10-20
1×10-20
1×10-20Pd
1
3
1×10-2
3×10-2
1×10-2
3×10-2
Sn
1×102
1×102
1×102
1×102
1×102
1×102
I
0
0
0
0
0
0
Cs
1
1×10-1
1×10-3
1×10-4
1×10-2
1×10-3Pb
2
0
2×10-30
2×10-3
0
-1
-2
-4
-5
Ra
3×10
3×10
3×10
3×10
3×10-3
3×10-4
Ac
2×102
2×10
2×102
2×10
2×102
2×10
Th
3×10
5
3×10
5
3×10
5
Pa
2
2
2
2
2
2
U
6
2×10-6
6
2×10-6
6
2×10-6
Np
3×10
5
3×10
5
3×10
5
Pu
3×10
5
3×10
5
3×10
5
Am
2×102
2×10
2×102
2×10
2×102
2×10
Cm
2×102
2×10
2×102
2×10
2×102
2×10

6-A-36

3

Table 22
Analysis case
Repository
GroundwaterC (inorganic)C (organic)
Cl
Co
Ni
Se
Sr
Zr
Nb
Mo
Tc
Pd
Sn
I
Cs
Pb
Ra
Ac
Th
Pa
U
Np
Pu
Am
Cm

Distribution coefficients for sorption on host rock (m /kg) (Neogene repository) (2/2)
Variant case
HLW, Gr.1, 2, 4
Gr.3 (< 1,000 y)
Gr.3 (> 1,000 y)
Low salinity
High salinity
Low salinity
High salinity
Low salinity
High salinity
00(Not assessed)(Not assessed)
0
0
0
0
0
0
0
0
0
0
0
0
8×10-1
4×10-1
3×10-4
5×10-5
3×10-4
5×10-58×10-1
4×10-1
3×10-4
5×10-5
3×10-4
5×10-5
-2
-2
-2
-2
-2
2×10
5×10
2×10
5×10
2×10
5×10-2
-2
-2
-4
-5
-3
1×10
2×10
2×10
7×10
2×10
7×10-4
2
0
2
0
2
0
6
6
6
6
6
6
0
0
0
0
0
0
1×10-20
1×10-20
1×10-20
7×10-2
7×10-1
1×10-2
3×10-2
1×10-2
3×10-2
6×10
6×10
6×10
6×10
6×10
6×10
0
0
0
0
0
0
4×10-2
5×10-3
1×10-3
1×10-4
1×10-2
1×10-31
0
2×10-30
2×10-3
0
-1
-2
-4
-5
1×10
1×10
3×10
3×10
3×10-3
3×10-4
1×102
1×10
1×102
1×10
1×102
1×10
2×10
5
2×10
5
2×10
5
2
2
2
2
2
2
4
8×10-84
8×10-84
8×10-8
2×10
5
2×10
5
2×10
5
2×10
5
2×10
5
2×10
5
1×102
1×10
1×102
1×10
1×102
1×10
1×102
1×10
1×102
1×10
1×102
1×10

6-A-37

3

Table 23 Distribution coefficients for sorption on host rock (m /kg) (Pre-Neogene repository) (1/2)
Analysis caseBase case
Repository
HLW, Gr.1, 2, 4
Gr.3 (< 1,000 y)
Gr.3 (> 1,000 y)
Groundwater
Low salinity High salinity
Low salinity
High salinity Low salinity High salinity
C (inorganic)
00(Not assessed)(Not assessed)
C (organic)
0
0
0
0
0
0
Cl
0
0
0
0
0
0
Co
3
5×10-1
3×10-4
5×10-5
3×10-4
5×10-5Ni
3
5×10-1
3×10-4
5×10-5
3×10-4
5×10-5
-2
-2
-2
-2
-2
Se
4×10
5×10
4×10
5×10
4×10
5×10-2
-1
-2
-4
-5
-3
Sr
2×10
7×10
2×10
7×10
2×10
7×10-4
Zr
2
0
2
0
2
0
Nb
6
6
6
6
6
6
Mo
0
0
0
0
0
0
Tc
1×10-20
1×10-20
1×10-20Pd
1
3
1×10-2
3×10-2
1×10-2
3×10-2
Sn
1×102
1×102
1×102
1×102
1×102
1×102
I
0
0
0
0
0
0
Cs
1
1×10-1
1×10-3
1×10-4
1×10-2
1×10-3Pb
2
0
2×10-30
2×10-3
0
-1
-2
-4
-5
Ra
3×10
3×10
3×10
3×10
3×10-3
3×10-4
Ac
2×102
2×10
2×102
2×10
2×102
2×10
Th
3×10
5
3×10
5
3×10
5
Pa
2
2
2
2
2
2
U
1×10
2×10-6
1×10
2×10-6
1×10
2×10-6
Np
3×10
5
3×10
5
3×10
5
Pu
3×10
5
3×10
5
3×10
5
Am
2×102
2×10
2×102
2×10
2×102
2×10
Cm
2×102
2×10
2×102
2×10
2×102
2×10

6-A-38

3

Table 23 Distribution coefficients for sorption on host rock (m /kg) (Pre-Neogene repository) (2/2)
Analysis case
Variant case
Repository
HLW, Gr.1, 2, 4
Gr.3 (< 1,000 y)
Gr.3 (> 1,000 y)
Groundwater
Low salinity
High salinity
Low salinity
High salinity
Low salinity
High salinity
C (inorganic)
00(Not assessed)(Not assessed)
C (organic)
0
0
0
0
0
0
Cl
0
0
0
0
0
0
Co
8×10-1
4×10-1
3×10-4
5×10-5
3×10-4
5×10-5Ni
8×10-1
4×10-1
3×10-4
5×10-5
3×10-4
5×10-5
-2
-2
-2
-2
-2
Se
2×10
5×10
2×10
5×10
2×10
5×10-2
-2
-2
-4
-5
-4
Sr
1×10
2×10
2×10
7×10
2×10
7×10-4
Zr
2
0
2
0
2
0
Nb
6
6
6
6
6
6
Mo
0
0
0
0
0
0
Tc
1×10-20
1×10-20
1×10-20
Pd
7×10-2
7×10-1
1×10-2
3×10-2
1×10-2
3×10-2
Sn
6×10
6×10
6×10
6×10
6×10
6×10
I
0
0
0
0
0
0
Cs
4×10-2
5×10-3
1×10-3
1×10-4
1×10-2
1×10-3Pb
1
0
2×10-30
2×10-3
0
-1
-2
-4
-5
Ra
1×10
1×10
3×10
3×10
3×10-3
3×10-4
Ac
1×102
1×10
1×102
1×10
1×102
1×10
Th
2×10
5
2×10
5
2×10
5
Pa
2
2
2
2
2
2
U
8
8×10-88
8×10-88
8×10-8
Np
2×10
5
2×10
5
2×10
5
Pu
2×10
5
2×10
5
2×10
5
Am
1×102
1×10
1×102
1×10
1×102
1×10
Cm
1×102
1×10
1×102
1×10
1×102
1×10

6-A-39

HLW repository (H12V)

TRU repository (Gr.1, Gr.2, Gr.4HD, Gr.4HH)

HLW repository (PEM)

TRU repository (Gr.3, Gr.4LD, Gr.4LC)
Leaching of RNs starts immediately after repository
closureTRU repository (Gr.3, Gr.4LD, Gr.4LC)
Leaching of RNs begins 300 y after repository
closureFigure 1

Transmissivity distribution of near field scale channels for the Base case
(Plutonic)

6-A-40HLW repository (H12V)

TRU repository (Gr.1, Gr.2, Gr.4HD, Gr.4HH)

TRU repository (Gr.3, Gr.4LD, Gr.4LC)
HLW repository (PEM)

Leaching of RNs starts immediately after repository
closureTRU repository (Gr.3, Gr.4LD, Gr.4LC)
Leaching of RNs begins 300 y after repository
closureFigure 2

Transmissivity distribution of near field scale channels for the Base case
(Neogene sediments)

6-A-41

HLW repository (H12V)

TRU repository (Gr.1, Gr.2, Gr.4HD, Gr.4HH)

TRU repository (Gr.3, Gr.4LD, Gr.4LC)
HLW repository (PEM)

Leaching of RNs starts immediately after repository
closureTRU repository (Gr.3, Gr.4LD, Gr.4LC)
Leaching of RNs begins 300 y after repository
closureFigure 3

Transmissivity distribution of near field scale channels for the Base case(Pre-Neogene sediments)

6-A-42

HLW repository (H12V)

TRU repository (Gr.1, Gr.2, Gr.4HD, Gr.4HH)

TRU repository (Gr.3, Gr.4LD, Gr.4LC)
HLW repository (PEM)

Leaching of RNs starts immediately after repository
closureTRU repository (Gr.3, Gr.4LD, Gr.4LC)
Leaching of RNs begins 300 y after repository
closureFigure 4

Variant case – near field scale host rock fracture connectivity
Transmissivity distribution of channels (Plutonic)

6-A-43

HLW repository (H12V)

TRU repository (Gr.1, Gr.2, Gr.4HD, Gr.4HH)

TRU repository (Gr.3, Gr.4LD, Gr.4LC)
HLW repository (PEM)

Leaching of RNs starts immediately after repository
closureTRU repository (Gr.3, Gr.4LD, Gr.4LC)
Leaching of RNs begins 300 y after repository
closureFigure 5 Variant case – near field scale host rock fracture connectivity
Transmissivity distribution of channels (Neogene sediments)

6-A-44

HLW repository (H12V)

TRU repository (Gr.1, Gr.2, Gr.4HD, Gr.4HH)

TRU repository (Gr.3, Gr.4LD, Gr.4LC)
HLW repository (PEM)

Leaching of RNs starts immediately after repository
closureTRU repository (Gr.3, Gr.4LD, Gr.4LC)
Leaching of RNs begins 300 y after repository
closureFigure 6 Variant case – near field scale host rock fracture connectivity
Transmissivity distribution of channels (Pre-Neogene sediments)

6-A-45

Plutonic
Base case

Variant case

Neogene sediments
Base case

Variant case

Pre-Neogene sediments
Base case

Variant case

Figure 7 Base and Variant cases - host rock fracture connectivity
Distribution of transmissivities in repository-scale channels

6-A-46

7 SAFETY CASE SYNTHESIS
7.1 Objectives of this chapter
As noted in Chapter 1, this report presents a safety case that does not target a specific site,
but uses the latest scientific knowledge and technological developments to assess thefeasibility and safety of geological disposal in representative Japanese geologicalenvironments. The aim is therefore to illustrate, as summarised in this chapter, a safety caseframework and the associated technical basis that can be applied in the future to any identified
site. A related intention is that this chapter should function as a standalone description of the
Safety Case for readers who have not read the other chapters. For this reason, Chapter 7
summarises some of the key findings in earlier chapters and adds the additional elements
needed for a safety case.
The basic approach involved is discussed in Chapter 2 and applied in a stepwise manner
through Chapters 3-6.






In addition to showing a method for identifying geological environments suitable for
hosting a repository, representative rock types were selected that are widelydistributed in Japan in order to illustrate how nationwide geological information canbe integrated into representative site descriptive models (SDMs) (Chapter 3).
A method for repository design based on the required safety functions was
illustrated, which specifically shows how specifications of a repository can be
tailored to these SDMs (Chapter 4).
Such tailored designs are assessed in terms of safety, both before and after closure(Chapter 5, Chapter 6).
Through these studies, research and development efforts were identified that will
provide required further safety case improvement in the future (each chapter).

In this chapter, based on this input, multiple lines of evidence are developed to build asafety case [1] [2] that is appropriate to the current pre-siting stage of our programme.
Therefore, for the goals and boundary conditions set for this report (Chapter 2), arguments
include consideration of the following points:(1) The established approach for building a repository with the required safety features in
relevant Japanese geological environments (the basic components of the safety caseas shown in Figure 1.4-2) that is consistent with the current scientific andtechnological state-of-the-art. This provides the knowledge base required in order toproceed with siting after the Literature Survey (LS) commences in volunteer
communities.
(2) The established safety assessment methodology, tools and data that form a soundbasis for evaluation of the safety of geological disposal concepts tailored toappropriate geological environments.
(3) The efforts made to increase confidence in the safety case in terms of supporting
arguments in which:The quality of established procedures leading from the creation of SDMs to thedesign of the repository and the associated safety assessment is shown to be
assured by an appropriately structured, logical methodology.

7-1

(4)



Appropriate procedures to deal with the range of uncertainties that need to beconsidered are clarified, together with measures to manage these in the future.



Developments in the technical knowledge base since the H12 [3] and TRU-2[4] reports are analysed.

To ensure that lessons learnt from the different stages in the programme will be
addressed, and thereby further improve confidence, creation of a safety case templatethat can incorporate the results of step-by-step site surveys and technologicaldevelopment in updates that are produced as required at programme milestones.

The points are considered further in the following sections.

7.2 Integration of arguments in a safety caseAs described in Chapter 2, the basic idea of implementing a deep geological repository
with the required safety functions includes:(1)

Avoiding natural perturbing phenomena that may have a significant impact on therepository by selecting a geological environment that maintains favourable
characteristics for a long period of time.

(2) For such a geological environment, assurance of radiological safety, general
occupational safety and minimal impacts on the surrounding environment during
construction, operation, and after closure. Design of the repository thus includes
engineering measures so that the safety functions relating both to the operational andpost-closure phases of the multi-barrier system last for a sufficient period of time.
(3) For all credible future evolution scenarios, assessed potential radiological impacts are
assured to lie below targets appropriate to their likelihood of occurrence, providing
confidence that the required safety functions are robust, even in the light of inherent
uncertainties.
In the following, the three fundamental technologies that form the basis for the evaluation
required to provide assurance of safety are summarised, based on the material alreadypresented in Chapters 3-6:
Technology for characterising suitable geological environments to support site
selection.

Technology for designing and implementing a repository with the required safetyfunctions for relevant geological environments.

Technology to evaluate the safety of such a repository.

7.2.1 Summary of the basis for safety evaluation(1) Basis for selection of suitable geological environments
In contrast to siting approaches in countries involving nomination of favoured sites, theJapanese approach of calling for volunteers, supported by a map that identifies scientificallyfavourable locations, means that the site selection process will involve an initial LS inapplicant regions that may contain various possible host rocks. To accommodate such a rangeof potential geological settings, assessment techniques are required that can be tailored to sites
as necessary during the different investigation phases. This requires:
7-2

i.

Clarification of the general requirements and criteria for determining site eligibility
in a stepwise site selection process.

ii.

Systematic development of characterisation technology needed to acquire
geological information necessary for judging site eligibility in light of specificrequirements/criteria, together with that for integrating such information into a site
descriptive model (SDM).

iii.

Checks of the applicability of such technologies on the basis of evolving
understanding of specific sites.

Regarding these, the relevant material from Chapter 3 is summarised in the following.
(i) Requirements/criteria for site selection and their application in stepwise
surveys
The Geological Disposal Technology WG re-examined the requirements and criteria for
selection of sites suitable for geological disposal [5], based on the scientific knowledge
accumulated since the H12 report. In addition to post-closure safety of the repository and thepracticality and safety of construction, operation and transportation were considered. Basedon this, areas with scientific features relevant to geological disposal were presented in the
Nationwide Map [6].
In this report, technical considerations associated with statutory terminology requirements
and criteria to be considered during stepwise site selection were summarised in Section 2.1.3(1). Furthermore, Section 3.1 discussed the requirements and criteria for determining theeligibility of such sites, with emphasis on safety during construction, operation and closure of
the repository and assurance of isolation and containment safety functions for specificgeological environments. In particular, areas at risk from perturbing natural phenomena, now
or in the future, will be excluded on the basis of the planned stepwise investigations. For any
potentially eligible site, the selection process takes into account expected long-term evolutionof the site geological environment, for which a properly selected EBS can provide sufficient
safety functions even accounting for such evolution processes.
As established in the Basic Policy on Final Disposal, the Nuclear Regulatory Authority
will specify factors to be considered during the selection of Preliminary Investigation Areas(PIAs), following Literature Survey (LS) of volunteer sites. These factors should then be
taken into account when formulating siting requirements/criteria. The factors specified in the
future for stepwise selecting Detailed Investigation Areas (DIAs) and, finally, the repository
site, should also be reflected at these project milestones.
(ii) Development of required investigation/assessment technology
The required technology for characterising the geological environment and assessing thepotential impacts of natural perturbing phenomena during stepwise site investigations was
summarised in Section 3.2, with emphasis on technological developments since the H12report. At the Preliminary Investigation (PI) stage, non-invasive surface mapping andgeophysical surveys lead to a programme of characterisation boreholes, using approaches andtechnology established in areas such as natural resource exploration and undergroundconstruction (summarised in Supporting Report 3-11). A combination of these individual
technologies can provide the geological knowledge base necessary for demonstrating that7-3

perturbing natural phenomena are avoided and to serve for the design and safety assessmentof the repository. This knowledge is based on an approach developed and demonstrated at the
JAEA underground research laboratories (URLs, see Figure 3.2-3 in Section 3.2.2), and isincorporated in a work flow specified for the LS and PI stages. Additionally, a planning
manual [8] and documents related to quality management (see Supporting Report 3-12) havebeen prepared to facilitate these investigations. The applicability of the concepts andtechnology for the PI stage has also been established, and practical experience gained,
particularly in collaborative NUMO/CRIEPI studies of borehole investigation technology [9].
With a particular focus on understanding evolution of the geological environment and thepotential impacts of natural perturbing phenomena, NUMO is reviewing the state-of-the-art
technologies systematically to assure best available technology is applied at LS and PI stages
for specific sites. Also R&D for developing technologies for Detailed Investigation (DI) is
ongoing in JAEA URLs, aiming to provide further input for selecting best technology for thatstage (for example, [10] [11]).
Although siting is intended to avoid the risk of natural perturbations that could causesignificant impacts on the repository (for example, volcanism), for particular assessment
timescales beyond 100 ky, extending to ≈ 1 My, any assurance inevitably involves greater
uncertainty due to the limits of current knowledge. Thus, NUMO has investigated aprobabilistic method to quantitatively assess risks (ITM-TOPAZ technique) [12], which forms
a basis to increase confidence in the associated safety assessment.
(iii) Information integration into a SDM
JAEA 's URL research [13] [14] iteratively evaluates the relationship between the type,
quality and quantity of information and the degree of understanding of the geologicalenvironment as site investigation progresses. By feeding back such understanding into theplanning of the next investigation step, this approach gradually reduces uncertainty inunderstanding the spatially heterogeneous characteristics of the geological environment. It has
also been demonstrated that the SDM developed by integrating such investigation results canbe used to effectively support design and construction of underground facilities.
In Section 3.3, based on geological information gathered on a nationwide scale, plus thatfrom URLs, the distribution of different rocks at relevant depths was assessed. By focusing on
key attributes required for a repository (e.g. practicality of safe construction, appropriate
barrier properties in terms of RN containment), three representative host rocks with different
characteristics were selected: plutonic rocks, Neogene sediments, and Pre-Neogene sediments.
In order to develop SDMs, in addition to representative geological settings, the length, density,and orientation of faults and fractures as well as other structural features, together with theirassociated hydrogeological characteristics, were set based on observations throughout Japan.
SDMs at regional scale (tens of km × tens of km), repository scale (5 km × 5 km), and panel
scale (800 m × 800 m) were developed, with the level of detail increasing as dimensions
decrease (see Figure 7.2-1). For each host rock, appropriate thermal and mechanicalproperties were established along with model groundwater chemistries and small-scale
descriptions of groundwater flow paths.
As emphasis is on the deep environment, the descriptions of topography andsurface/shallow geological features are illustrative only, without specified quantitative
information. In any case, these characteristics are extremely site specific and could vary
greatly even for a specified host rock. Thus, the characteristics of, and evolution scenarios for,7-4

the surface environment are not explicitly discussed in this report. The evolution of thegeological environment of the host rock is also not discussed explicitly, as it is assumed to bestable for long a time and its evolution might depend on that of the surface environment. In
coming assessments these aspects will be given more attention.

Figure 7.2-1 SDM illustration at different scales (example of plutonic rocks)(Integrating Figures 3.3-3, 3.3-9, 3.3-28, and 3.3-44). Note that the description of topography isillustrative only. In the future, when SDMs are developed for actual sites, local scale variations
in topography that may impact groundwater flow will be captured.

Nevertheless, the surface characteristics are accepted to be key attributes of sites, withquantitative information being provided by applying the investigation technology described in(ii) above.
As described in Section 2.2.5, the present report assumes that underground facilities will belocated at sufficient depth within a stable geological setting (which is likely to be much
deeper that than the legally required minimum depth 300 m for placement of the repository),so that significant impacts from uplift and erosion would not occur over relevant timescales.
However, as described in Section 3.2.2, due to the limited information that can be obtained
during the LS stage, temporal evolution of the SDM is represented at a conceptual level only.
A more quantitative SDM, capturing evolution with time (a “4D” SDM), will be developed
only during the PI stage, although NUMO is already assessing the information and technology
required to develop such models.
(iv) Current status and future needs
As described in (i) to (iii) above, NUMO will utilise BAT (term expanded as either “bestavailable techniques” or “best available technology”) to ensure any site selected meets therequirements and criteria for determining its eligibility. The availability of required
technology has been established or it is currently under development. The applicability of

7-5

such technologies for NUMO’s specific needs is confirmed by demonstration tests at researchsites and in URLs. Thus, it can be said that the technical basis for investigations after LS has
been established.
As discussed in Section 3.5.2, in order to better assure the suitability of specific geologicalsettings to host a repository, NUMO will continue a focused programme of R&D, including:Improving the understanding of natural perturbations that may affect the safetyfunctions of a repository (volcanic/igneous activity, volcanic hydrothermal/deep fluid
movement, earthquake/fault activity, uplift/erosion). This will include better
determination of probability of occurrence, together with associated impacts, which
will allow improved representation of perturbation scenarios and more complete
capture of included uncertainties, to form the basis for focused design countermeasures to reduce impacts and more realistic assessment of safety margins.

Improving investigation technology with a particular focus on the requirements todevelop and refine the SDMs and to also cover the site evolution (4D SDM) during
the PI and DI stages, leading up to selection of a site for repository implementation.
This will include characterisation of the features of sites that determine the release ofradionuclides (RNs) from the engineered barrier system (EBS) and their transport to awell-defined geosphere-biosphere interface (GBI). Again, output will be focused by
the requirements of evolving repository designs and associated safety assessments.

Investigation plans and technology used will be tailored to the conditions of specific sites(geology, geography/topography, socio-political constraints, etc.). As previously noted, thiswill allow focused determination of the potential significance of uplift and erosion (e.g.
Supporting Report 6-10) and allow models to be developed that assess consequences fordifferent repository designs, taking into account the timescale for safety assessment.
(2) Basis for repository design and implementation
To construct a repository that meets the required pre- and post-closure safety functions,key considerations include:

The characteristics and quantity of waste to be disposed of, leading to design
requirements set to assure safety functions expected for each component of therepository, including engineered barriers and associated surface and undergroundfacilities. The methodology allows tailoring the specifications of each component of
the repository, in order to flexibly satisfy the design requirements for constraints set
by siting environments (see Section 2.1.2 (4)).



Utilisation of engineering technology to construct, operate, and close such a repositorythat has been demonstrated to be applicable or that can be reasonably be expectedbased on future technological development.

For the HLW repository, as in the H12 report, NUMO will incorporate a system ofmultiple engineered barriers, with robust design of performance and applicability to a wide
range of stable geological environments. Thus, the design requirements for vitrified waste,
overpacks, and buffer materials as engineered barriers and for supporting undergroundinfrastructure have been identified. These are specified for each of the representative
geological formations defined in the H12 report. Furthermore, through demonstration of
production of overpacks and construction tests using buffer materials, it was shown that7-6

implementation can be carried out based on H12 report technology or that considered to befeasible in the near future.
For TRU waste, as in the TRU-2 report, the design of engineered barriers and undergroundfacilities considers the characteristics of the four groups of such waste. These designs again
lead to specifications for each geological formations considered. Also here, engineering tests
support that the required technology for implementation is available, or is reasonably
expected to be so in the near future.
Extending the basis established in the H12 and TRU-2 reports by taking into account the
updated scientific and engineering knowledge base, Chapter 4 developed repository designs
and assessed the required engineering technology, as summarised in the following.
(i) Repository design method
The design requirements for the repository considered in the H12 report focused
predominantly on post-closure safety and engineering feasibility. Since H12, NUMO has
introduced a comprehensive set of design requirements for individual components of therepository to meet design factors including: post-closure safety, operational safety,
engineering practicality, retrievability, environmental impact and socio-economic aspects. By
applying these requirements to corresponding components, detailed specifications are derived(materials, shape, dimensions, etc.). However, in this study, factors that are considered to bevery site-specific, such as environmental impact and socio-economic aspects, are notconsidered in the design process, but will be as the siting programme proceeds.
As a starting point for developing repository concepts, those presented in the H12 andTRU-2 reports were re-assessed in the light of recent technology development. From thedesign requirements identified for individual design factors of each component of the HLW
and TRU waste repositories, basic specifications that meet the design requirements aredeveloped. In order to introduce more flexibility, variants of the disposal concept andassociated technology for EBS implementation are considered. This will form a basis for
future optimisation of the design and tailoring of concepts to the siting environment.
(ii) Reference repository designs
Designs of repositories for HLW and TRU waste are developed for the three reference
SDMs assuming that these facilities will be co-located at a single site.
(a) Setting the disposal depth
A key parameter for the design of underground facilities is the implementation depth. This
will be set on a site-specific basis, depending on the distribution of suitable host rock(s),subject to the fundamental constraint set by law that this should be 300 m or deeper. A
balance has to be found between rock stability, practicality of drainage and ventilation during
construction, which are generally more challenging at greater depths, and post-closure safety,
which might be expected to improve with depth (especially with regard to future uplift anderosion). As previously noted (and discussed in Section 2.2.5), constraints set by uplift anderosion are not assessed prior to site-specific analyses. Pragmatically, therefore, a reference
depth can be set as the greatest depth for which design can ensure reasonable stability of

7-7

underground openings (tunnels, vaults) during construction and operation, while available
equipment can provide required services, such as ventilation and drainage.
For the stronger plutonic rocks and Pre-Neogene sediments, the limiting practicalconstraint on depth is the construction safety requirement to keep temperatures in the tunnels
below 37 °C by realistic ventilation equipment, leading to setting 1,000 m as the installation
depth. For the weaker Neogene sediments, the requirement for cavity stability leads to setting
the installation depth as 500 m. However, it should be noted that these depths are only
indicative and are based on preliminary simplifying assumptions. As the designs aredeveloped here on the deep side of the range of acceptable depths, this will indicate flexibility
in tailoring to the depths of suitable formations at actual sites.
(b) EBS design
For the HLW repository, the specifications of the overpack and buffer derived in the H12report [3] are taken over as they meet all the design requirements for the three SDMs. In thisprocess, however, current knowledge indicates an overpack lifetime of more than 10 ky, far
beyond the design requirement to prevent groundwater contacting the HLW for 1 ky (seeSection 4.4.1). To assess constraints set by practicality of emplacement, an H12 in-hole
option (H12V) is compared to a horizontal, in tunnel option with a prefabricated EBS module(PEM). For TRU waste, the TRU-2 waste package [4] (in the present report called waste
package A) was reassessed along with a variant waste package that includes a lid andimproved robustness against dropping (referred to as waste package B), and which facilitates
operations by allowing lifting by a crane. The assessment shows that both options satisfy thedesign requirements for each waste group (see Section 4.4.2).
The designed EBS is set within a layout specified for each SDM (discussed further in (c)below), for example, as illustrated for the PEM and TRU waste in plutonic rock in Figure 4.418 and Figure 4.4-25 respectively, which are integrated in Figure 7.2-2.
The EBS design specifications meet current design requirements with sufficient margins
for all three representative rocks, despite inherent uncertainties at the present time. These
provide a sound basis for assessing operational and post-closure safety (described in Chapters
5 and 6). However, from a post closure perspective, not all aspects have been fully assessed,
as already indicated in previous chapters. It is also important to recognise that these designs
may be over-conservative and are certainly not optimised (see Section 7.2.1 (3)), which is an
issue that will be considered when developing an EBS tailored to the geological environmentand other boundary conditions for specific sites.

7-8

Figure 7.2-2 Example of EBS design and underground facility layout(Example of plutonic rocks, PEM dead-end emplacement tunnel)

7-9(c) Underground facility design
For underground facilities, geological structures and hydrogeological conditions wereconsidered in order to determine the shape and placement of tunnels and disposal panels inorder to meet the design requirements (Section 4.5.4). In particular, concerns about potential
water inflow during construction and/or operation and preferential paths for transport of any
released RNs led to definition of layout determining features (LDFs), such as larger fault
zones, that should be avoided during design. Together with orientation of the hydraulicgradient, such LDFs are the main constraints on disposal panel layout in plutonic rock, taking
into account also inherent uncertainty in their specified locations – taken to be approximately
1% of the large fault length which is more than 1 km - Section 3.1.3 (1). For the sedimentaryrocks, fold structures and permeability variations due to different lithologies present require
additional considerations during development of layouts. Further, for the weaker Neogene
sediments, tunnel orientation in relation to the ambient stress field has also to be taken into
account.
On a smaller scale, emplacement determining features (EDFs) that result in localised water
inflow need to be considered, as these impact the ability to excavate tunnels and disposal
holes and assure that the EBS can be emplaced to required quality levels. For the defined
distribution of such features in each SDM, the percentage of disposal holes or tunnels that canbe used is quantified and, if required, the area of disposal panels increased.
By taking into account LDFs and EDFs, required design assessments can be conducted toensure constructability and repository containment functions. For example, the size anddistribution of faults, fractures and other features in which advective flow occurs is aparticularly important characteristic in terms of ensuring the repository containment function.
Design requirements are formulated and the design is adapted to such features andcharacteristics. Such adaption can also be further developed during future layout optimisation.
Additional issues to be considered when defining the layout of disposal panels involve thepracticality of construction and operation, leading to specification and placement of access/connecting tunnels and shafts, ventilation shafts, effective work flow lines for transporting
waste and material to/from construction zones, and ventilation/drainage routes in the tunnels.
The procedure for assessing such issues focuses on worker safety (e.g., provision of multiple
escape routes, counter-measures against accidents, etc) and work efficiency (especially given
the relatively high reference emplacement rate for HLW). As an example, the resulting
layouts for HLW PEMs and TRU waste in plutonic rocks are also shown in Figure 7.2-2.
Together with the EBS design, the layout is taken into account during the assessment of
post-closure safety for the three different SDMs, as described in Chapter 6. This
demonstration of requirements-driven design provides a foundation for tailoring layout toevolving SDMs produced during stepwise site investigations. In the future, however, feedback
from post-closure assessment of RN migration will allow this to be directly considered alongwith operational safety and work efficiency in an optimisation process. For this, it isimportant to have a more realistic assessment of post-closure safety, as considered further in
Section 7.2.1 (3).
(d) Design of surface facilities
For “active” surface facilities handling waste, the concepts and requirements alreadyestablished for related nuclear facilities that handle, inspect, and package HLW and TRU
waste can be applied to ensure radiation protection during operations. The resulting designs
7-10thus ensure radiation protection of both the general public and workers (see Section 4.6.2).
For surface facilities (see Section 4.2.3 (4)) that do not handle radioactive material,conventional design approaches can be applied. The transportation of waste from interim
storage elsewhere to the surface facility can utilise extensive experience in both land andmarine transportation in Japan and other countries, including that of spent fuel and HLW. In
particular, the specifications for transport containers and ships that return waste after overseas
reprocessing are considered applicable to geological disposal operations (Section 4.6.1 (1)).
The approach illustrated in this report allows design of surface facilities that satisfy therequired safety functions for the waste reception, inspection, and encapsulation processes thatare most important from a radiation safety perspective. After sites are identified, NUMO willproceed with more detailed design of both surface facilities and transportation infrastructure,
based on local environmental conditions. By constructing the surface facility directly above
the disposal footprint, access tunnels can be shortened, which is advantageous in terms ofsafety, environmental protection, work efficiency, economics, etc. However, depending onsite-specific constraints, spatial separation may be required in terms of optimisation, alongwith associated tailoring of concepts during stepwise characterisation to assure stakeholder
acceptance, ease of construction of the underground facility and safety both before and after
closure.
(iii) Technology required for repository construction, operation and closure
Basic research and development since the H12 report has led to improvements intechnology related to practicality and quality assurance of fabrication of EBS components andtheir emplacement underground (see Section 4.4.3). For example, welding overpacks with athickness of 19 cm and non-destructive detection of welding defects as small as 2 ~ 3 mm has
been demonstrated [14][15]. For current overpack specifications, the analysis shows that, fora wide range of conditions, the safety function of suppressing contact between groundwaterand HLW for at least 1 ky is greatly exceeded on the basis of well-established fabrication
technology.
In addition, manufacture of full-scale buffer blocks with the required density for the H12V
concept has been demonstrated [16], along with a vacuum suction technology allowing suchblocks be gripped and placed in the disposal holes by remote control (on the basis of both
domestic and overseas verification tests [17] [18]). PEM manufacture and transport is underdevelopment, with initial full-scale technology demonstrations ongoing [19]. As the PEM
concept is similar to the horizontal Swedish KBS-3H option, demonstration tests of thismethod conducted at the Äspö URL are directly relevant [20]. For TRU waste, full-scaledemonstration tests have included the construction of surrounding buffer and the filling of
gaps between waste packages [21] [22]. These technological developments, for practicalapplication under realistic deep underground conditions, provide confidence in the EBS
construction techniques assumed in the illustrative designs (see Section 4.4.3).
In terms of general construction of underground facilities, the technology proven in
existing facilities, such as the construction of underground power plants, can be applied. In
addition, with regard to the tunnel backfill and plug construction technology required forrepository closure, full-scale equipment development and demonstration tests in URLs are
already underway in Japan and overseas (see Section 4.5.7). For the construction andoperation of active surface facilities, technology that has been proven in existing nuclear
facilities can be applied.

7-11

When considering ease of retrieval, it is necessary to prevent this function from adversely
affecting operational or post-closure safety, e.g., due to delaying closure of the repository orparts of it containing waste. For example, to assure that ventilation air does not impact thecorrosion resistance of the overpack, and in order to reduce erosion of buffer, disposal tunnels
should be closed as soon as possible after waste is emplaced. Even after backfilling andemplacement of a mechanical plug, retrieval was shown to be technically feasible (see Section
4.7.1). In particular, development of a method of removing buffer around the H12V overpackwith salt water is ongoing [23]. Such a technique can also be applied to the PEM backfill,which is considered to be easier to retrieve. Thus, at least before final closure, retrieval isconsidered to be sufficiently well demonstrated (see Section 4.7.2).
After sites are identified, technology will be refined, in line with evolution of therepository design and associated construction, operation and closure plan, based on
information obtained from stepwise characterisation. This will combine demonstrations ofengineering technology required with practical aspects of ensuring reliability and meeting
quality levels to ensure designed safety functions during full scale implementation.
(iv) Current status and future needs
As shown above, based on information provided by the three SDMs, repositories can be
designed that meet a range of requirements to realise expected safety functions and can be
implemented based on existing technology. Nevertheless, the technology required to construct,
operate and close such repositories will be continually re-assessed, based on progress in otherrelevant civil engineering areas, with the aim of assurance that application to geologicaldisposal remains state-of-the-art and can be tailored to the evolving understanding of
requirements for specific sites.
Depending on regulatory or socio-political requirements impacting the disposal plan,
NUMO will move forward with more concrete designs incorporating optimisation,
considering both operational and post-closure safety. For example, alternative materials and
EBS designs can be assessed with the aim of improving engineering practicality, operational
efficiency, flexibility to tailor to site conditions, socio-economic aspects, etc. This will require
improved assessment technology for designs and construction/operation plans, such as better
quantifying long-term behaviour of engineered barriers, assessing impacts of water inflow
during tunnel excavation and responding to feedback from safety assessment.
Specifically, for the technology of waste retrieval, NUMO will further promotedemonstration tests to confirm its feasibility. More generally, NUMO will continue to utilise
the outcome from JAEA URLs and also plans increased participation in technology
development and demonstration in international URLs, covering aspects of repository
construction, EBS emplacement, monitoring and closure to required quality levels.
(3) Basis for safety assessment
The repository concepts developed for the representative SDMs, together with definedprocesses for construction, operation and closure, form the basis for safety assessment, with acurrent focus of radiological impacts both before and post-closure. At the present time, in theabsence of specific national regulations for geological disposal, guidelines from internationalorganisations and safety regulations in other countries are used to provide the technicalfoundation for a safety assessment based on the latest scientific knowledge. In terms of the
7-12period before the repository is closed, it is also necessary to ensure general occupationalsafety for workers based on safety measures and safety management applied in similarindustries, in accordance with the Industrial Safety and Health Act. However, such safety is
not considered in the present report.
Section 2.4 described the internationally agreed methodology for assessing radiologicalconsequences[24] [25], which is adopted in this report. Safety assessment includes thefollowing tasks:
Repository features, events and processes (FEPs) impacting safety functions are
systematically evaluated to establish a comprehensive set of scenarios.



Quantitative analysis of such scenarios to determine possible radiological effects of
the repository pre- and post-closure.

Assessment of whether these radiological effects meet the given safety criteria andtargets.

This safety assessment is based on that used for the H12 and TRU-2 reports, updated toreflect advances resulting in R&D since then. Chapter 5 presented the assessment ofoperational safety, whilst Chapter 6 covered post-closure safety.
(i) Operational safety
For safety assessment covering operation of the repository, regulatory standards and safetyassessment methods for other nuclear facilities can be referred to. Radiation protection ofworkers and the surrounding public is assured by the design of equipment and procedures forhandling radioactive material, incorporation of sufficient shielding and a comprehensive
radiation monitoring system. Safety assessment calculates radiation exposure for normal
operations and also for perturbed operational scenarios that are in line with the defence in
depth principle, associated with all waste handling operations. Such assessment should
confirm that the required level of safety can be ensured based on regulatory standards oridentify aspects of the design of the repository that need to be reviewed. This also provides
guidance for implementing appropriate radiation control and monitoring measures.
As noted in Chapter 5, with reference to IAEA safety guidelines [26] and new regulatory
standards for nuclear facilities in Japan (promulgated after the Fukushima Daiichi Nuclear
Power Plant accident), both planned operations at the repository (normal operations), andimpacts of perturbing events (abnormal operations) are evaluated in this report. Under normal
operating conditions, radioactive waste stores and handling equipment prevents leakage ofradioactivity, with maintenance of a negative pressure preventing external loss of any surface
contamination that might be present. Such rigorous containment prevents any exposure ofworkers or the surrounding public to radioactivity, which is confirmed by continuous
monitoring. A combination of remote operation of parts of the facility for handling waste,
which are generally located at basement level, together with additional shielding by walls,
effectively ensures direct radiation doses not only meet standards, but are as low asreasonably achievable (ALARA). The assessment of the normal state includes all operations
involving waste acceptance, inspection, encapsulation, handling, interim storage,transportation and disposal (See Section 5.3).
To assess abnormal operations, perturbations that could lead to leakage of radioactivity
from HLW or TRU waste are assessed, using event trees that include both natural phenomena
and human events that could cause sequential loss of the multiple layers of protection to
7-13

prevent such events. Where possible, failsafe counter-measures are introduced to prevent suchperturbations and, in the event of failure, mitigation of the consequences (see Section 5.4.1).
For abnormal operation scenarios (e.g., fire or waste package drops) conservative numericalanalysis is used to determine if this could credibly lead to leakage of radioactivity (seeSection 5.4.2). A conclusion from the scenarios examined to date is that loss of containment isextremely unlikely in all analysed cases. This is further addressed in Section 7.2.2, in terms of
the results of the safety assessment.
When sites become available, the operational safety assessment will be specifically tailoredto them, with special consideration of common mode failures, which have been little
examined so far. Additionally, feedback from the safety assessment will guide refinement ofdesign and associated safety measures, to assure meeting the legal requirements for both
radiological and general occupational safety.
(ii) Post-closure safety
Unlike the case before closure, post-closure safety of the repository relies on a passive
system of multiple engineered and geological barriers. Even through post-closure monitoring
may be implemented; this is not required to assure safety. A key to assuring safety is siting ina stable geological setting, where required safety functions can be assured for a sufficiently
long period of time. In such a site, the repository is designed including an EBS with safetyfunctions to complement those of the local geology. Furthermore, even if specific perturbing
events and processes may occur that degrade future barrier performance, their probability andradiological consequences should be sufficiently low that regulatory standards are met. If theassessment shows safety cannot be assured with confidence, the design of the repository may
be modified or, if this is insufficient, the site will be rejected. As regulatory standards havenot been specified, the methodology and technical basis developed in this report is prepared toallow adjustment when such standards are available.
As described in Chapter 6, based on the system understanding incorporated in the SDMs
and the associated repository designs, scenarios that reflect potential future evolutions aredeveloped. Base scenarios that describe expected evolution are complemented by variant
scenarios, which capture inherent temporal and spatial variability of evolution of therepository in its geological setting and also account for uncertainties in system understanding
and in associated models and databases. Variant scenarios are considered to be less probablethan the base scenarios, but their likelihood is not quantified at the present time. However, formajor perturbations, the very low probability involved is estimated to the extent possible.
When indicative probabilities to these major perturbation scenarios can be quantified, this,
together with the calculated dose, can be compared with the risk target for very low
probability scenarios (see Section 6.1.5).
RN release and transport models account for the composition and geometry of the EBS, thelayout of the disposal panels and the spatial heterogeneity of the geological setting in terms ofhydrogeology and solute transport characteristics. A three-dimensional groundwaterflow/solute transport model has been used, which can better evaluate the performance of thenear-field. This is intended to more realistically evaluate the effects on post-closure
performance resulting in differences in the repository design and its geological setting forspecific sites. Such realistic assessment also provides important feedback for planned
optimisation of the designs. Since the detailed analyses to improve realism may involve large
computational loads, 3D models currently cover only a limited area of the near field, but
expansions are ongoing, in line with increased power of parallel computers, with the aim of
7-14

better assessment of release and transport on the scale of entire sites. Depending on theanalysis objectives, processes to be considered, boundary conditions, configurations, etc., it isimportant to apply codes that do not require as much computer power.
Current models/databases introduce conservative simplifications, especially with regard totreatment of uncertainty, to ensure the robustness of the safety assessment. In particular, thechemical-thermodynamic models, used to derive porewater chemistry in the EBS, elemental
solubility limits and speciation (that support selection of RN migration parameters such as
sorption and diffusion coefficients), are inherently unrealistic in that they assume equilibriumthat is rarely found in nature. Nevertheless, a description of RN release and transport can be
developed that is consistent with that in other safety assessments and compatible withobservations in relevant analogue systems.
The RN release and transport models are adapted for use in the consequence analyses ofbase, variant and unlikely perturbation scenarios, quantifying the RN releases used for dose
assessment (see Section 6.4). Stylised inadvertent human intrusion scenarios are also
developed, in line with recommendations of international organisations and those used inother national programmes, with scoping RN release and transport models used to estimate
potential dose impacts (see Section 6.5).
(iii) Current status and future needs
As described above, the H12 safety assessment methodology has been further developed inlight of evolution of international discussions on regulations for geological disposal as well as
those for other nuclear facilities. Additionally, considerations both pre- and post-closure of
repositories tailored to representative SDMs are now included. This will form the basis forassessment of specific sites after the LS stage is initiated. More realistic, site-specific safetyassessment will provide a basis for comparing repository design options and, possibly,comparing different potential host rocks at a specific site or assessing pros and cons ofdifferent sites.
Specifically for the assessment of operational safety, the hazard database will be updated in
the light of experience in Japan and overseas, expanding current knowledge about potential
perturbations to normal operation and representation of these in abnormal operation scenarios.
In particular, assessment of scenarios involving common mode failure will be expanded,widening the range of scenarios for abnormal events (such as waste drops and fire) and
improve consequence analysis, with associated verification and validation of the models used.
In addition to more realistic 3D models of RN release and transport over larger scales, as
noted above, a future aim is to capture post-closure evolution of the repository system withtime (“4D modelling”). This will require a model that can simulate the evolving distributionsof thermal, hydrogeological, mechanical, and chemical (THMC) conditions in detail, alongwith the impacts of these on the EBS evolution and the RN migration characteristics of theflow path from the EBS to the GBI. This allows consistent assessment from construction,
operation and closure of the repository to the period after closure, so that the results of thesafety assessment can be appropriately fed back to the evolving design, taking into accountadvances in technology over this period. Parameters to define RN transport must reflect longterm changes in the surface and deep environment over time, which are not considered in thisreport and depend on input from characterisation of specific sites. As a starting point, 4D
SDMs that take into account climate and sea-level changes, which are already underdevelopment, will be used, as described in Section 7.2.1 (1) (iv). In particular, site-specific 4D
7-15

SDMs will more explicitly reflect changes in the GBI, which has a large impact on
assessment of the doses resulting from most release scenarios. This requires better definitionof the flow paths that lead from the EBS to the surface environment, along with their RN
transport characteristics, which will require both focused laboratory studies and in-situ tests in
URLs. In addition, site-specific biosphere models will be developed that are tailored to boththe present and likely future physical setting and lifestyle of local populations. A special focus
in all such work will be extension of the work included in this report, implicitly assuming adisposal footprint under land, to the very different hydrogeological and geochemical
conditions that may be encountered for disposal panels offshore for a likely coastal repository
setting.
Finally, with the aim of facilitating future assessments, a goal will be to utilise advanced
technology for managing knowledge, information, and data related to scenario construction,model development and data setting, together with visualising these within storyboards andanimations. This will require close coordination of geological characterisation, repository
design, and safety assessment, increasing traceability and facilitating uncertainty/sensitivity
analyses and quality assurance, all leading to improving the reliability of future safety cases.
7.2.2 Evaluating the safety of geological disposal
As mentioned in Section 7.1, the assessment basis developed in this report, under current
boundary conditions, does not yield a safety case for specific sites, but rather illustrates thefundamental feasibility of demonstrating safety for designs tailored to representative siting
environments. The safety assessment thus highlights issues that need to be considered todetermine if any volunteer site would be suitable, and also could lead to a specific safety casethat meets regulatory standards to be specified in the future, while identifying open issues thatneed to be clarified by future R&D. Here, the results of the safety assessment will be
discussed, with the aim of identifying such issues.
(1) Safety assessment results(i) Operational safety(a) Normal operation
Chapter 5 assessed operations within a repository specified in Chapter 4, with a focus onradiological protection of the general public. Under normal operating conditions, the radiationdose at the boundary of the site was significantly lower than the dose target value (50 μ Sv/y)for general public stipulated in the Business License Standards Regulations (see Section 5.3).
If it is difficult to ensure a sufficient distance to the site boundary for specific sites, this can bereadily managed by increasing the thickness of shielding walls and/or re-arrangement of
surface facility layout. Extensive experience in the nuclear industry will assure that all
operations and equipment avoid any risk of loss of containment during any of theperturbations that must be considered during normal operations, which is particularly
facilitated by the robustness of the waste packages considered.
So far, the influence of topography and the surface environment on surface facilities is notconsidered, with the reference case layout on a large flat area above the disposal footprint. For
specific sites, the available area for the surface facility may be small, it may not be flat, andmay be located at a distance from the optimal disposal footprint, further constrained byeconomic and social aspects. Many of these issues are common to the design of other nuclear
7-16

facilities, and it is considered possible to flexibly deal with them by combining system
understanding with accumulated experience in Japan and overseas.
(b) Abnormal operations
In assessment of the radiological effects of anomalous conditions in surface andunderground facilities that handle waste, constraints set by countermeasures that minimise
likelihood or possible impacts, are taken into account. For example, interlocks to reduce therisk of dropping wastes and physical lifting height limits that define the worst case drop for all
relevant perturbations (see Section 4.6.2 (6)). Assuming that safety measures to prevent
accidents fail, illustrative scenarios were developed for both HLW and TRU repositories andtheir consequences evaluated. In all of the cases examined, it has been confirmed that there is
no predicted release of radioactivity (see Sections 5.4.2 and 5.4.3). All designs andoperational processes assessed are commonly applied for the three SDMs. For specific sites,
this assessment will be extended and tailored to the repository and operating processes
involved, taking account of advances to allow more realistic scenario specification,
considering also potential common mode failures.
(ii) Post-closure safety
In this report, a risk-informed strategy is adopted, with distinction between the probabilityof the base, variant and low probability perturbation scenarios, which are complemented bystylised human intrusion scenarios, with target dose assigned to each (see Section 6.1.5).
(a) Assessment of base and variant scenarios
The base scenario represents the expected evolution of a well-sited and designed repository,
assuming all safety functions perform as required, while variant scenarios scope the range ofscientifically reasonable uncertainties in the associated assumptions, models and data (seeSection 6.3.3 (2)).
Quantitative consequence analyses of such scenarios, for all SDMs and corresponding
repository designs, show that the maximum total dose occurs within 100 ky and is well below
the set target (10 µSv/y) for the base case (see Section 6.4.1). In addition, no variant scenarios
are near the set dose target (300 µSv/y), which is internationally recommended as an indicatorfor repository safety (see Section 6.4.2). The results also show that, for all three SDMs andtwo model groundwaters, the HLW options H12V and PEM system result in effectively thesame maximum dose. For TRU waste, the maximum dose in the package A case was slightly
higher than that of Package B. Thus, Figure 7.2-3 for H12V co-disposal with TRU waste
package A, shows the maximum doses calculated.
Figure 7.2-3 shows differences in maximum dose compared to the base case for many ofthe variant cases, predominantly due to scenarios in which I-129 and U-233 are assumed
present as anions with high solubility and extremely small sorption onto both buffer and hostrock, in which resultant doses are little impacted by other varied parameters.

7-17

ScenariosBase
scenario

Analysis case

Maximum dose (µSv/y)
NUMO Target
10 (μSv/y)Base case

Increase in glassdissolution

Guideline
300 (μSv/y)

Increased corrosionof TRU Gr.2
Plutonic rocks
low salinity GW

Uncertainty instructural framedegradation

Plutonic rocks
high salinity GW
Neogene rocks
low salinity GW

Nitrate plume
impactNeogene rocks
high salinity GW
Pre-Neogene rocks
low salinity GW

Fractureconnectivity in hostrock
Variant
scenario

Pre-Neogene rocks
High salinity GW

Lower sorption in
bufferIncreased diffusivity
in buffer

Lower sorption inhost rock

Increased diffusivity
in host rock

Thermal increase insolubility

Uncertainty inthermodynamic data

Figure 7.2-3 Base and variant scenario maximum doses for the analysis cases of co-disposal of
HLW (H12V) and TRU waste (package A)(b) Assessment of human intrusion and low probability perturbation scenarios
Although the likelihood of a major perturbing event is very low given proper site selection
and facility design, in Japan these cannot be precluded for the very long timescales underconsideration. Hence, assessment is carried out to determine their potential radiologicalsignificance and assess robustness despite associated uncertainties. In particular, volcanic and7-18

seismic activities were targeted, developing very conservative scenarios of a new volcano
occurring directly beneath the repository and a major fault growing until it intercepts therepository at some point in the future. Calculated maximum doses for these scenarios areshown in Figure 7.2-4, compared to the defined targets (in the first year: < 20 - 100 mSv,
thereafter: < 1 - 20 mSv/y). In addition, the probability of occurrence of such events wasestimated and total risk calculated by converting dose into the probability of occurrence of
lethal cancer or serious genetic effects and the result compared with a risk target of < 10-5/ywhich corresponds to the internationally recommended dose constraint (300 µSv/y). In nocase were the dose or risk limits reached (see Section 6.4.3).
Scenarios

Analysis caseNew volcanism(dose assessment
results for a repository
in plutonic rock – highdensity disposal)

Maximum dose (µSv/y)Plutonic

Target dose from the second
year on after the event
1～ 20 mSv/yLow
probability
scerarios

Fault extension

Borehole worker
exposure(observation of
excavated spoil)

Human
intrusion
scenarios

Target dose in the firstyear after the event20～ 100 mSv

Plutonic
low salinity GW

Neogenelow salinity GW

Pre-Neogene
low salinity GW

Plutonic
high salinity GW

Neogenehigh salinity GW

Pre-Neogene
High salinity GW

Plutonic

Neogene

Pre-Neogene

Plutonic

Neogene

Pre-Neogene

Plutonic

Neogene

Pre-Neogene

Borehole worker
exposure(observation of core)

Migration of nuclides
in borehole

Figure 7.2-4 Maximum doses for human intrusion and low probability scenarios(Co-disposal of HLW (H12V) and TRU waste (package A))

The probability that future human actions will affect the safety function of the repository is
judged to be very low due to disposal at depths greater than 300 m and avoiding locations
containing economically valuable mineral resources during site selection. It is expected thatthe probability will be further reduced by institutional control measures, such as keeping
records and prohibiting some activities on the site – e.g., at least by the time of final
decommissioning and release of the site from institutional control. To evaluate possible
radiological effects, stylised human intrusion scenarios were assessed, based on borehole
drilling as it occurs in Japan today. Both the calculated maximum doses (Figure 7.2-4) andassessed total risks lie well below the targets defined for low probability perturbation
scenarios above (see Section 6.5).
In the future, such analysis will be refined for site conditions, accounting for expected
geological evolution and the possibility of several perturbations occurring together.
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(c) Conclusions from safety assessment
The three SDMs and the repositories tailored to them are expected to be representative of
suitable sites in Japan that could result from the site selection process. The robust arguments
developed above, adopting conservatism to account for uncertainties, form a good basis forassuming that operational safety can be assured for both normal and perturbed states. After
closure, uncertainties generally increase with time, but there is reasonable confidence thatthese will not preclude demonstration of safety for suitable sites, especially with emphasis on
the period during which the toxicity of the waste is of most concern (≈ 100 ky). Nevertheless,there are limitations to the post-closure assessment that have been identified and will need tobe addressed in order to meet the requirements of a safety case to support selection of sites
after the LS stage is complete.
One important limitation is the assessment of scenarios including uplift and erosion. By
appropriate site selection, sufficient uplift and erosion to impact a deep repository in the first
100 ky should be excluded. However, in Japan, this cannot be precluded over much longer
timescales. The impacts of such scenarios are highly site-specific, depending not only on
uplift/erosion rates, but also geological structure, topography/bathymetry, impacts of glacial
cycles and the slow evolution of the repository system as depth of overburden gradually
decreases or increases (e.g., as may occur under the sea). On the basis of information derived
from the stepwise siting process after the LS, appropriate scenarios will be developed, alongwith the tools and databases required to quantify them.
Representation of the processes resulting in RN release and transport, in both the base andvariant scenarios, has been constrained by both lack of sufficiently detailed geologicalinformation for the different spatial scales considered and also the capabilities of current
models, codes and computers. Using site-specific information and next generation models,significant improvements in realism can be introduced. For example, assessing RN migration
on a repository scale (several km x several km) would be a great advance compared toassuming unrealistic LDF short circuits to the GBI (see Section 6.4.1 (5)). When combinedwith more detailed 3D (or 4D) solute transport models at panel/repository scales, underdevelopment by extending the present model for near-field scale, the degree of conservatismshould be significantly reduced, which should be reflected in better performance of both theengineered and natural barriers. The degree of improvement of models depends also on
general advances in computing technology, which will be continuously monitored.
As volunteer sites are likely to be located in coastal settings, further major improvements
required involve the development of scenarios and assessment tools for repositories with
waste disposal footprints partially or completely below sea rather than land, together with 4D
SDMs explicitly representing the impacts of sea level change. In general, it would be
expected that performance of a subsea repository would be greatly improved by negligible
hydraulic gradients, but variations in hydrogeology and geochemistry for disposal zones thatvary between being under land or sea may complicate analysis (but would still allow
development of a rigorous safety case, as shown in Sweden [27] and Finland [28], where
current sites may be under the sea in the future).
Finally, stylisation that takes into account site conditions is essential for scenarios related
to low probability perturbation and future human actions, which may also need refinement toreflect guidelines set forth in the safety regulations that will be formulated in the future.

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(2) Supporting arguments for post-closure safety
The formal safety assessment and comparison of calculated doses (or risks) with a target isa necessary, but not sufficient, argument of the safety of a particular repository. Quantitative
assessment inevitably involves much simplification of the real system and hence a safety casemust include additional arguments that support conclusions reached, as also presented in the
H12 report. Here, such supporting arguments are developed for both the assessment basis andsafety assessment presented in this report.
(i) Safety arguments based on complementary indicators
In the safety assessment described in Chapter 6, dose is used as the main index to assess
safety, estimating the radiological effects on humans by evaluating this parameter ascalculated by release rates of RNs to a highly stylised GBI/biosphere. Since the uncertainty
related to future human lifestyles and the characteristics of the surface environment are
extremely large, it is important to show the performance of the geological disposal systemusing other indicators that do not depend on these factors. Such alternatives to dose aretermed complementary indicators (see Supporting Report 7-1) and have been used both inJapan and other countries to support safety cases. Here examples of these are presented tosupport arguments for the safety of the repositories specified in Chapter 4.
(a) Arguments to support repository containment performance
The basic safety feature of RN containment within repositories tailored to the SDMs can bewell illustrated by the base case RN migration analysis described in Section 6.4.1, focusing on
the distribution of total radioactivity within the repository system (described in detail in
Supporting Report 7-2). Figure 7.2-5 shows an example, distinguishing between the activity
located:Inside the panel scale region (EBS and immediately surrounding rock).

Outside the panel scale but within the repository scale geosphere.

Outside the repository scale region.

The assessment is for the base case for co-disposal of HLW (H12V – which is little
different to the PEM: Section 6.4.1 (7)) and the more conservative (lower performance) TRU
waste package A, distinguishing between the 2 model groundwater variants.
Figure 7.2-5 presents result up to 10 My, despite the fact that the models are extremely
uncertain (or inapplicable) beyond 1 My and, by this time the total radioactivity has been
greatly reduced by radioactive decay. For the plutonic case, ≈ 99% of the overall radioactivity
remains within the panel scale region after 1 My, regardless of the model water chemistry.
For the Neogene and Pre-Neogene sediments, performance is similar for the lower salinity
groundwater. For the high salinity groundwater, both sediments show radioactivity releasefrom the panel scale is about 1% after ≈ 100 ky, with about 30% loss for the Neogene
sediments and about 25% for Pre-Neogene sediments after 1 My. Even then, the release fromthe repository scale at 1 My is only about 20% for both the Neogene and the Pre-Neogene
sediments.
This clearly shows that the repository maintains containment in terms of high retention forthe period of concern. In the future, however, it will be important to confirm such

7-21

performance using more realistic models that take into account evolution of the repository
system.

7-22

(b) Neogene sediments

High salinity groundwater

Low salinity groundwater(a) Plutonic rock

Figure 7.2-5 Distribution of radioactivity in different repository regions(Base case for co-disposal of HLW (H12V) and TRU waste (package A))

7-23(c) Pre-Neogene sediments

(b) Arguments to support radiotoxicity reduction with time
As shown above, almost all radioactivity is contained within the panel scale region for 100
ky for all base case calculations. To put this in context, potential radiotoxicity [29] iscalculated as an alternative indicator of radiological hazard. This is derived by multiplying thequantity of each radionuclide present by the dose conversion factor for oral ingestion.
Although it is clearly impossible for such material to be orally ingested, this is often used as ayardstick for potential radiological hazard
For each of the cases shown in Figure 7.2-5, calculated potential radiotoxicity of RNs
remaining in the panel scale region is presented in Table 7.2-1. This involved calculation ofpotential radiotoxicity of the entire panel scale region (Sv) for both the HLW and TRU waste
repositories, with radionuclide content derived from the migration analysis and doseconversion factors for oral ingestion taken from ICRP [30].
Table 7.2-1 Average potential radiotoxicity within the panel scale area at 100 ky after closure(base cases)
Unit: Sv/kg
Host rock

Plutonic rocks

Neogene sediments

Pre-Neogene sediments

Water chemistryLowsalinityHighsalinityLowsalinityHighsalinityLowsalinityHighsalinity

HLW repository

1.2 x 10-2

1.2 x 10-2

9.5 x 10-3

6.5 x 10-3

8.3 x 10-3

7.3 x 10-3

TRU repository

4.2 x 10-2

4.2 x 10-2

2.7 x 10-2

2.7 x 10-2

4.3 x 10-2

4.3 x 10-2

Natural uranium ore
Cigar Lake [31]2.5
Ningyo-toge [32]2.4 x 10-2

While the panel scale areas defined in Section 3.3 extend 100 m downstream with 50 m
vertical extent above and below this area, the volume to calculate potential radiotoxicity isconservatively set by neglecting downstream extension. The average potential radiotoxicity
per kg (Sv/kg) was calculated, assuming this entire volume is rock, which dominates over thevolume of the EBS and using appropriate densities (plutonic rocks: 2.69 Mg/m3, Neogene
sediments: 2.28 Mg/m3, Pre-Neogene sediments: 2.64 Mg/m3) as described in Supporting
Report 7-3. Table 7.2-1 additionally includes for comparison, potential radiotoxicity
calculated for ores from the Canadian Cigar Lake [31] and Japanese Ningyo-toge [32]deposits, with uranium contents of about 8 % and 0.05 %, respectively.
From the table, it is clear that the potential radiotoxicity calculated in this way, is sensitive
to the repository layout, which results in values in the smaller TRU facility greater than the
more extensive panels for HLW, and in Neogene sediments less than for the other rocks. In
any case, the potential radiotoxicity of the repositories is far below that of Cigar Lakeuranium ore, being similar to the uranium ore of Ningyo-toge.
Taking (a) and (b) together, at about 100 ky after repository closure, most radioactivity
remains in and around the EBS and, by then, its potential radiotoxicity is similar to lower
grade uranium ores.

7-24

(c) Arguments on potential radiotoxicity of releases into the biosphere
To complement the arguments on the effectiveness of the containment safety function, aperspective on the significance of any releases to the surface can be provided by indicatorsthat do not depend on very uncertain assumptions about the food chains and lifestyle that are
required for biosphere dose calculations. As an example, the impact of such releases can beevaluated by assessing the case of these occurring directly into a river.
Specifically, calculated maximum base and variant case RN releases from the repository
scale regions (Sections 6.4.1 and 6.4.2), given in Bq/y, are assumed to go directly into a
typical Japanese class A river, with the same volumetric flow as assumed for the Chapter 6
biosphere model (1 × 109 m3/y, see Supporting Report 6-1) [33]. The resulting average RN
concentrations [Bq/m3], are assessed using the ICRP dose conversion factor for internal
exposure by ingestion [30] and summed to derive potential radiotoxicity (Sv/m3). This
calculation process is described in detail in Supporting Report 7-4.
The results are presented in Figure 7.2-6 for co-disposal of HLW and TRU waste, which iscompared to the World Health Organisation “Guidelines for Drinking-water Quality” [34],taking the U-238 guideline values for a concentration limit (<10 Bq/l) and converting this into
a potential radiotoxicity. From this figure, it is clear that, in all cases, the potential
radiotoxicity from repository releases is very small compared to the limit for natural RNs indrinking water.

7-25

(b) Neogene sediments(c) Pre-Neogene sediments

High salinity groundwater

Low salinity groundwater(a) Plutonic rock

Figure 7.2-6 Potential maximum radiotoxicity in rivers based on base and variant case concentrations(co-disposal of HLW (H12V) and TRU (package A))(L: low salinity groundwater, H: high salinity groundwater, number: variant case numbers given in Table 6.4-6 of Section 6.4.2)

7-26

(ii) Supporting arguments from natural analogues
The extent to which repository safety functions can be validated by short timescalelaboratory studies of subsystems or in-situ experiments in URLs is inherently limited. Suchexperiments are used to the extent possible to test the assumptions, models, and data sets used
to assess safety. However, these need to be extended using natural analogue test cases thatcover appropriate timescales and the complexity of the natural environment, as discussed
further in the following and in Supporting Report 7-5.
(a) Natural analogue arguments supporting long-term geological stability
In some national programmes, natural analogues have been used to demonstrate thefundamental feasibility of geological disposal based on the ultra-long-term (Gy) stability ofuranium ore bodies in particular settings (e.g., Oklo, Cigar Lake [31] ). Although no rocks ofanything like this age are found in Japan, the fundamental principle of stability ofgeochemical anomalies in deep, chemically-reducing environments is applicable. The
requirement is thus to illustrate that stability is sufficient to support safety arguments over My
timescales, when site selection excludes areas vulnerable to volcanic and/or tectonic
perturbations and uplift and erosion, as discussed in Chapter 3. Such natural analogue
arguments are based predominantly on paleo-hydrological and geochemical studies.
Unlike the cases in countries where the host rock extends from repository depth to thesurface (e.g., in Canada or Fennoscandia), in Japan – as captured in the SDMs – lowpermeability host rocks are generally overlain by a range of more permeable sediments. Evenin areas with significant topographic relief, such a situation tends to decouple very slow
groundwater flow at depth from more rapid flow in overlying aquifers. For example, studies
in the Horonobe area demonstrated that deep Neogene sediments contain very old, stagnant
saline water, with a calculated age of about a few My, where mass transport is dominated bydiffusion, despite changes in chemical conditions that are clearly observed at shallower depths.
The low impact of moving saline-fresh water boundaries associated with long-termclimate/sea-level changes in this location [35][36][37] indicate that, for coastal sites, theimpacts of such changes may be minor in the repository host formation.
This conclusion is consistent with observations at the Tono uranium mine, where uranium
ore has been preserved in Neogene sediments for about 10 My. This is explained by a stablehydrogeological setting in which pH and reducing conditions are buffered by interactions
between water, minerals and microbes [38]. There is no isotopic evidence of uranium
mobilisation as a result of natural perturbations, such as uplift and erosion, or sea-levelchanges [39]. Indeed, despite evidence of movement of faults that penetrate ore bodies, there
are no indications of either uranium or its more mobile daughters being lost.
It can be noted that such analogue arguments also imply that the scenarios examined todetermine impacts of future fault movement, or even impacts of human intrusion, are overpessimistic. In the event of a short circuit flow path being formed, typical geological settings
in Japan would not result in any significant flow either upwards or downwards through arepository and, even if there was a short-term geochemical perturbation, this would be
buffered back to original conditions by biogeochemical reactions. There is further evidence tosupport such a conclusion, such as analytical study of calcite crystallisation at the surface ofwater conducting fractures. These indicate that the chemical conditions of groundwaterremain virtually unchanged on a My timescale at depths of 320 m – 610 m in the granite
bedrock of Tono, despite the presence of a major fault zone [40].
7-27Over safety-assessment timescales, therefore, there are good arguments to support
expectations that suitable host rock settings should provide adequate protection of the EBS.
One caveat here, however, is that artesian conditions could occur in some sedimentary hostrocks that are underlain by more permeable formations. These would generally be identified
during site characterisation, but even this case would not necessarily preclude sufficient hostrock performance for a low permeability sediment – as demonstrated, for example, by the
Swiss Opalinus Clay in which solute transport occurs by diffusion despite a very high verticalhydraulic gradient and the presence of fractures [41].
In addition to its prime role of protecting the EBS, there is also analogue evidence
suggesting that the natural barrier performance in terms of limiting RN mobility may begreater than currently assumed. A good example here is based on careful studies of thelimiting concentrations of U and Th in the old groundwaters found at Horonobe andMizunami [42]. Although not clearly explainable based on chemical thermodynamic models,measured concentrations in true solution (using filtration to remove colloids) is ≈ 10-12 mol/lfor U and ≈ 10-13 mol/l for Th – vastly below the solubility limits set in Section 6.4.1 (≈ 10-6mol/l for U and ≈ 10-8 mol/l for Th). Even though it is unclear if such observed concentrations
represent equilibrium solubility set by an undefined mineral phase or a steady state
representing slow rock-water interactions, such values indicate that some input parameters
used in the safety assessment may be highly over-conservative [43]. Nevertheless, care must
be taken extrapolating such information to a specific site as, especially for Th, elemental
concentrations in groundwater are predominantly in the form of suspended particles or
colloids, indicating that potential transport of such phases would also need to be explicitly
considered.
Even though a geological setting may be stable, it also has to be shown that the presence ofthe repository does not degrade favourable properties. In the case of argillaceous sedimentary
host rocks, a particular concern is alteration of minerals along flow paths due to the high pHof leachate from concrete – particularly for the TRU disposal area. Several natural analogues
allowing assessment of long-term interaction of hyper-alkaline plumes with sedimentaryrocks present a consistent picture, supporting arguments that alteration reactions are slow andthe products of these are generally as expected from models, although secondary mineral
assemblages differ in detail. For example, at the Maqarin site in Jordan, hyper-alkaline water(up to pH 12, similar to Portland cement porewater) has been in contact with natural clays forat least 80 ~ 100 ky, but alteration is seen to be restricted. In particular, water conducting
fractures are observed to seal due to precipitation of secondary minerals, which limits
alteration of host rock to within 3 - 4 mm of such features [44][45]. Migration of plumes overlonger distances occurs only if there is a mechanism for reactivation of sealed fractures, which
can be expected to be limited at repository depth in a stable tectonic setting.
(b) Natural analogue arguments supporting EBS barrier roles
As noted already in the H12 report, the EBS developed for HLW is very robust and assureshigh containment of RNs for any reasonably stable geological setting. Such high performanceis assured by mechanistic understanding, empirical laboratory data and supporting analogue
arguments.
Laboratory studies show dissolution of HLW glass in contact with water is slow, withkinetics influenced by dissolved silica concentration. Additionally, dissolution tends todecrease with time as protective alteration layers build up (involving minerals such as
palagonite) – which can be particularly important for larger glass blocks. Such processes can
7-28

be seen in analogues and, for the particularly relevant case of glass within a Japaneseargillaceous rock similar to compacted bentonite, studies of shards of 1 My old glass incontact with saline groundwater (pH ~ 8, Eh ~ -50 mV) have been studied. Here, despite
being so small that alteration layers are not significant, persistence of the glass can be related
to the very low water flow rate and resultant high silica concentration (equivalent toamorphous SiO2 saturation) [46][47][48].
Although conclusions must be drawn with care due to chemical differences between
analogues and HLW glass – and also the absence of significant RN concentrations in theformer – the large empirical database suggests that the lifetime of HLW glass assumed forsafety assessment based on laboratory experiments (~ 70 ky, see Section 6.4.1 (2) (v) (b)) isvery conservative. Indeed, for a fractured monolithic block within protective Fe corrosion
products and a bentonite diffusion barrier, a lifetime of at least an order of magnitude greaterthan this value might be more realistic.
The carbon steel of the overpack is also thermodynamically unstable and could reducewater in the absence of any stronger oxidants. In relevant geochemical environments,
however, Fe corrosion is kinetically slow and may decrease to negligibly low values asprotective oxide layers build up. This is supported by a large number of natural analogues
based on archaeological artefacts. Indeed, even in more corrosive environments, protective
coatings may assure the longevity of large steel objects. This is well illustrated by a hoard of
Roman nails of low carbon steel discovered in Inchtuthil, Scotland, where nails with very low
corrosion rates, or even survival without corrosion for almost 2 ky, were observed in central
parts of the hoard. Despite an environment with weakly acid, oxidising water, corrosion ofouter nails produced an alteration layer with which chemically reducing/low flow rate
conditions protect those within [49].
Such qualitative observations are consistent with more quantitative analyses of thecorrosion of Japanese archaeological iron artefacts buried near surface for a few hundred toabout a thousand years. Corrosion depths lie in the range of between 0.2 mm and 5 mm,
which is interpreted as due to formation of protective corrosion alteration layers [50][51].
Under repository conditions, where any oxygen introduced during the construction andoperation period will be rapidly consumed, the long-term corrosion rate of the massive
overpack is thus expected to be very low, giving containment lifetimes much longer than thetimes in excess of 10 ky calculated and presented in Chapter 4.
The clay minerals in bentonite, the main constituent of buffer and backfill, are generally
stable under the conditions found in Japanese host rocks and, indeed, are often components ofthe sedimentary host rocks or found as alteration infill in fractures in plutonic rocks. Althoughthese may degrade at higher temperatures, an extensive analogue database supports argumentsthat such alteration occurs only in the presence of water, is generally constrained by supply of
K and, even when this element is abundant, is very slow at temperatures below about 120 –150 °C. Hence the key roles of the HLW buffer can be assured over assessment timescales of> 100 ky as long as the groundwater water composition is not significantly altered and theemplacement quality can be assured (which can be strongly supported for the PEM, inparticular).
The longevity of bentonite is less assured in the presence of significant quantities of
cementitious materials (e.g. for buffer around TRU waste), as the component clay minerals
are thermodynamically unstable at high pH, just as discussed above for sedimentary hostrocks. Several natural analogues allowing assessment of long-term interaction of hyperalkaline plumes with natural bentonites present a consistent picture, further supporting
7-29

arguments that alteration reactions are slow and the products of these are generally as
expected from models. For example, bentonite mines in the Mangatarem region of thePhilippines, allow study of long-term interactions with highly alkaline and high Ca content
water with a pH of between 9.3 and 11.6 (similar to that resulting from low-pH cement
interactions). Alteration zones in bentonite are quite limited (generally within 5 mm),predominantly due to the formation of a protective low-permeability, iron-rich layer (3 mm)[52]. Although details of thermodynamic alteration models are not confirmed, these analogues
provide good support for assuming bentonite retains its barrier roles for timescales much
longer than ≈ 100 ky.
Concrete and other cementitious materials used as TRU waste immobilisation grout,emplacement cell infill, backfill and to form structural components, are also
thermodynamically unstable and will alter by interaction with groundwater and othermaterials present – e.g. bentonite (as considered above), steel, organic materials, etc. Again,
an extensive archaeological analogue database supports arguments that such alteration is slow
and key properties (e.g., high pH, low permeability) will persist for timescales from thousands
to hundreds of thousands of years, depending on the local hydrogeological environment.
Despite the extensive analogue support for assumed high performance of the HLW EBS,
the situation for TRU waste is less complete – which is reflected in the very conservativemodels used to represent RN release and transport. For example, release from bituminised
waste (Gr.3) is assumed to be instantaneous despite analogue evidence of similararchaeological and natural materials which show very low degradation over ky timescales.
Here it has to be acknowledged that examples under hyperalkaline conditions are limited, butthere are certainly arguments to support very low solute release rates from this material,
which could considerably constrain concerns about the impacts of resultant high nitrate
plumes.
(c) Synthesis
Overall, natural analogues support the main conclusions of the high isolation potential of arobust system of engineered barriers located in a suitable deep geological setting in Japan.
However, these analogues highlight limitations in the models used – in particular in terms ofoverly simplistic hydrogeology and assumption of thermodynamic equilibrium. While themodels tend to be very conservative, lack of realism will limit their application to compare
different sites and associated repository concepts, as will be needed as the NUMO programmeadvances. Development of better models will also require appropriate analogues to test them,which is noted as a future goal – particularly for coastal/sub-sea siting options where theanalogue knowledge base is particularly limited.
(iii) Supporting arguments by comparison with other safety cases
The work involved in post-closure safety assessment is particularly complex, making it
difficult to ensure that all model development and implementation is performed properly andhence that the resulting calculated doses are reliable. One useful check of credibility of output
involves comparing different safety assessments with each other [53] [54]. This method
provides a holistic overview, providing confidence when any differences in the safetyassessment results compared can be explained by variations in the approach, design andgeological environment assumed, together with the scenarios assessed and the models/data
sets used. However, since the objectives and other conditions vary significantly between
7-30

safety cases, a strict comparison is not trivial and efforts on such work has only been initiatedrecently.
Regarding the safety assessment results of this report, examination has begun bycomparing with relevant safety cases produced in other countries [55]. Detailed output fromthe assessments of the three representative host rocks (Supporting Report 6-24, 6-25, 6-26)
have also been compared to those of the earlier H12 and TRU-2 studies. For example, theplutonic rock H12 EBS specifications are similar to H12V in this report but, for the H12reference case, which is equivalent to the base scenario here, the nuclide that contributes the
maximum dose is Cs-135 [3] while this report calculates doses dominated by Se-79 (Figure
6.4-23). The reason for this difference is the higher Se-79 solubility in this report while
sorption of Cs-135 on buffer material and host rock is larger, giving significant delay anddispersion over the long transport path assumed (see Supporting Report 6-24). Further, thebiosphere assessment in this report, assumes a river flow rate 10 times higher than that in the
H12 report (based on the latest statistical data, see Section 6.4.1 (6) (ii)), which reduces all
doses, although the impact on maximum dose is limited compared to the factors previously
noted.
(iv) Synthesis of additional arguments in support of the safety caseIn addition to the results of safety assessment, there are several additional supporting
arguments that can, or will eventually, support the safety of geological disposal in Japan.
Among the approaches to such arguments so far proposed, the safety case developed in thisreport has considered supplementary safety/ performance indicators, natural analogues andinter-comparison with other safety cases. The purpose, conditions, and applicability of these
multiple lines of arguments are briefly summarised in the following.
Different supplementary safety/performance indicators have been applied based on thebase scenario and associated models and data, which generally demonstrate the large
containment safety function provided by the engineered barriers and near-field host rock forthe three SDM cases. Assessing the distribution of total radioactivity with time within therepository system illustrates the great conservatism which has been included in these models
and data. The robustness of the containment functions should be further evaluated by applying
more realistic modelling and dataset. Also, assessing the distribution of radiotoxicity and thepotential radiotoxicity of releases into the biosphere is considered to be useful, but should be
carefully applied by taking into account much greater uncertainties associated with thebiosphere characteristics and evolution. More realistic biosphere modelling can be developed
at a specific site, which can then increase reliability of the application of such an indicator.
While the application of natural analogues strongly depends on the geologicalenvironments and repository design, it is useful even for the safety case at this generic stage.
However, a careful discussion of the applicability and identifying limitations is needed toclarify what message can be transferred. For example, the analogues applied in this report forgeological stability and favourable conditions in Japan only support the claim that appropriate
sites could be selected, not that the selected sites can be assured to be favourable. Natural
analogues for the engineered system are useful to promote understanding of system behaviour
and relevant processes and indicate that present assessment models and data could be overly
conservative, which can help prioritisation of further development of more realistic models
and datasets.

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Inter-comparison of different safety cases is also a very useful tool to increase confidence
in the reliability of each element of the assessment basis and the process of integrating these
into a safety case, based on analysing the reasons for the similarities and differences between
then, see e.g. [54]. Application of the approach for the safety case in this report is limited
mainly to the H12 and TRU-2 studies, but illustrates the QA process to check all safetyassessment elements and their relationships. Further efforts here will be made in the future, as
outlined in Section 7.3.3 (4).
Multiple lines of argument are very useful to support the safety assessment results andstrengthen the demonstration of safety, but their applicability and limitations need to becarefully acknowledged. The arguments applied in this report support safety discussions in a
generic sense. How these arguments can be used to support the claim of safety for a sitespecific repository system is not yet rigorously analysed. At the site-specific stages, however,
the application of the supporting arguments will be refined, taking siting environments andtailored repository designs into account.
7.2.3 Constraints on, and scope of, the safety case presented in this report
At the present stage of NUMO`s programme, the safety case developed is inherently
limited by the lack of a site-specific context, uncertain boundary conditions and the simplified
assumptions, models and databases used for quantitative assessment. This section describes
how the current safety assessment allows NUMO to focus and prioritise efforts to develop therefined safety cases needed for future project milestones.
(1) Assuring safety and practicality of geological disposal in JapanNUMO has established a methodology for selecting favourable geological environments
and developing site descriptive models (SDMs) for a range of potential host rocks, anapproach for designing a repository with the required safety features for such SDMs that is
consistent with the current technological state-of-the-art, and methodology, tools and data thatform a sound basis for assessment of the safety of such repository designs. These have been
integrated into the current safety case, which provides the knowledge base required in order toproceed with siting after the LS commences in volunteer communities.
The safety case developed in this report cannot be applied to specific sites, but illustrates
the fundamental feasibility of demonstrating safety for designs tailored to representative siting
environments. In order to move further, specific developments that have been prioritised
include:

Development of a more realistic inventory database that covers all important waste
characteristics, based on a bounding range of waste production scenarios. Thisrequires interaction with all those responsible for waste production, conditioning,
packaging, storage and transport.



For TRU waste, the safety case has shown that this can dominate total releases from a
combined repository. This justifies reassessment of the EBS and developing more
realistic models/databases to quantify its performance in order to increase safetymargins. This will require further accumulation of knowledge on identified issues forthis waste, such as the effects of nitrates, organic substances, gas generation, andhighly alkaline solutions on the safety functions of both engineered and naturalbarriers.
7-32Expansion of the assessment base to allow better assessment of coastal/subsea
disposal options. A priority is investigation technology to determine geologicalstability, such as quantifying uplift/erosion and determining the existence of magma
chambers and deep hydrothermal fluids. This will be complemented by improvementof the knowledge base supporting development of base and variant scenarios
quantifying long-term impacts, including evolution of the EBS and RN release andtransport in a saline or mixed/varying freshwater/saline environment.



As Pre-Neogene sediments represent a poorly studied geological setting, these will bea focus for systematic improvement of the knowledge base, with emphasis on features
noted to impact the safety case – such as the consequences of high carbonate waters,
which could also be of relevance for other sedimentary host rocks.

The approach and methodology for a more consistent and coherent safety case that
integrates pre- and post- closure phases, taking into account both the perturbation from
repository construction and operation activities on the original geologicalenvironment, and also THMC evolution of the EBS and near-field host rock during theoperational period. This will provide more realistic initial conditions for the postclosure evolution of the repository system required for safety assessment. In addition,
it will be important to develop an associated monitoring system/strategy in order toestablish baseline conditions before any site perturbations occur.

(2) Applicability as a safety case template
The three SDMs developed are likely to reasonably represent volunteer sites, subject to theconstraints noted above. Thus, much of the structure and technical content developed may bedirectly taken over for the next safety case, after LS to support PI site selection and planning.
Nevertheless, problems encountered during the production of this safety case provide
guidelines for essential improvements that should be implemented to facilitate updating andrestructuring for specific sites. In particular, as noted in the JAEA H17 project, integrationand synthesis of the large, multidisciplinary knowledge base that supports the safety case ischallenging, with conventional approaches, such as those used in the H12 and TRU-2 reports,
now being more difficult, and hence utilisation of more advanced KM tools is identified as a
key requirement – possibly more like that illustrated in JAEA’s H22 and H26 projects(discussed further in Section 7.3.3 below).
A related challenge was technical QA, which was complicated by parallel work being
carried out by different teams on different chapters and their associated SRs. Even thoughboth internal and external technical reviews were carried out, these resulted in many different
versions of text components that were difficult to integrate – especially as the documentation
software used was very limited in its capacity to support change management. This is again anarea where improvement could be built into the safety case template (see also Section 7.3.2
below).
Following the discussion in Section 7.3 on confidence building for the safety case in thisreport, future perspectives for the application of the safety case as a template are presented in
Section 7.4.

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7.3 Assuring the safety case is “fit for purpose” (confidencebuilding)
In addition to the issues noted above, development of the safety case after sites are
identified will have to incorporate flexibility to respond to site-specific environmental
conditions and adapt to both future advances in science and technology and changes in social
conditions over the extremely long implementation period (in the order of a century). These
are almost unique boundary conditions for any technological project, but adoption andmodification of the management framework utilised for this safety case (shown in Figure 1.42) can provide a structure for addressing such issues. The goal is thus to assure that, for all
programme milestones, decisions can be supported by a safety case that is fit for purpose,contributing to building confidence in the credibility of the repository project.
Thus, Section 7.3.1 assesses the requirements for development of the safety assessment
basis, Section 7.3.2 discusses properly handling limits of scientific and technologicalknowledge and other uncertainties associated with long term predictions, while Section 7.3.3
considers the management aspect of actually implementing such an ambitious goal. Based on
this, Section 7.3.4 summarises the confidence that such goals can be met, based on theexperience of production of the safety case documented in this report.
7.3.1 Technical basis for Assessment(1) Ensuring technical quality
In the preparation of this report, a quality management system established within NUMO
was applied, which complies to the ISO 9001 standard. It was, recognised, however, that sucha general management standard is necessary but not sufficient due to the special challenge ofassuring technical quality of the complex, multidisciplinary work involved, especially in thelight of the particular problems of documentation under time pressure with limited manpower
resources (discussed in Section 7.2.3 above). Here advanced tools to facilitate strict review
and associated issues resolution will be implemented in the future.
Clearly, the reliability of the assessment basis, for all components of site characterisation,
repository design and safety assessment, depends on the application of scientificunderstanding based on the latest knowledge. Here, technical quality is assured byconventional approaches: including peer review of concepts, data and assumptions; testing of
implementation structures and processes; and verification/validation of models and databases.
The Supporting Reports for Chapters 3 to 6 present the detailed scientific and technicalknowledge supporting the safety case, in many cases with information to support its quality such as publications in peer-reviewed scientific literature. For the mathematical models andcalculation codes, in particular, track records of applications and documented verification andvalidation are reported. For the integrated processes included in the safety assessment, input
data for each analysis and resultant output results are stored together in a structured database,
together with the concepts and assumptions used to set analysis conditions, to ensure thetraceability and transparency of the process. Additionally, as noted in Section 7.2, credibilityof output is checked by comparisons with similar cases included in the H12 and TRU-2
reports.
At a more strategic level, the technical basis of this report was reviewed during production
by domestic and international experts in the Technical Advisory Committee (TAC)
established by NUMO. In addition, objective assessment by the Technology Development
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Evaluation Committee, which consists of outside experts, also plays an important role inensuring technical quality and continuous improvement. The history of such external support
is summarised in Supporting Report 7-6.
As described further below, contributions to general quality management result fromcollaborations with both Japanese and international partners, particularly involving field workand projects in URLs (Section 7.3.3 (3)), where practical experience is gained that will be
invaluable when work progresses to the PI stage and beyond. Nevertheless, it is recognisedthat technical QA is one of the biggest challenges for implementation of a safe repository
project and the need for continuous improvement, using BAT, is a primary goal.
(2) Major advances since the H12 and TRU-2 projects
The H12 and TRU-2 projects, completed more than a decade ago, demonstrated that safe
disposal of HLW and TRU waste is feasible in Japan. However, when NUMO embarked on
the volunteer approach for site selection after the publication of these studies, it was realisedthat this would require a capacity to characterise and assess sites with quite different levels ofgeological complexity and also that novel repository concepts might be needed for some ofthe potential sites. Furthermore, there has been a need to shift focus from research on
feasibility to preparing for actual site characterisation, site selection, repository design andimplementation. Meeting these challenges has been a focus of NUMO’s programme
supported by, and closely linked to, R&D carried out by other organisations in Japan, such as
JAEA and CRIEPI. Thus, a synthesis of all key developments in Japan which can provide
support for increasing the confidence of the safety case is presented below.
(i) Geological knowledge baseSince the start of NUMO’s programme, there has been a significant expansion of thegeological knowledge base, in preparation for the coming steps in the siting process. It is
acknowledged that the siting process requires NUMO to be prepared for very different siting
environments and this has been a key aspect of such activities.
A deep understanding of the detailed characteristics of some geological formations that
could be potential host rocks has been developed through the work carried out at the two
URLs in Japan and through collaborative work in overseas facilities. Crystalline andsedimentary systems constitute a large portion of Japan’s geological environment and havebeen studied at the Mizunami and Honorobe URLs, respectively. The research at these URLs
has allowed studies of these rock types at potential repository depth, demonstrated techniques
for detailed characterisation and allowed for various tests and experiments, such as studying
the evolution of the EBS and other properties over relatively long times, in addition toaccumulating more general knowledge on how to construct and develop a repository.
Furthermore, NUMO and associated research organisations in Japan have conductedcollaborative projects in overseas facilities, thereby expanding knowledge on differences
between, not only rock types, but also associated underground conditions that may be
encountered at sites.
A nationwide-scale geological database has been developed and will form the basis for theupcoming LS, which will gradually be expanded as more volunteer sites come forward. Using
this database, existing data, representing the three different geological settings in Japan thatmight be suitable for a repository, have been assessed and synthesised into SDMs. In turn
7-35

these SDMs will be used as input for developing site specific repository designs andassociated safety assessment (see Chapter 3).
The geological environmental characteristics used to construct these SDMs were derived
from this nationwide scale geodatabase and do not specify a region, with the exception of dataobtained from the JAEA URLs. Nevertheless, the development of the SDMs demonstrates
how geological information should be handled in a quality-controlled manner and be reflectedin the SDM. In particular, the expansion of information on the spatial distribution of rock
structural features and groundwater composition, which are important for repository
performance, contributes to development of more practical repository designs and more
realistic safety assessment. In addition, it is illustrated how sediments of different ages
(Neogene and Pre-Neogene) can significantly differ in key mechanical characteristics, which
are important for the design of repository, and the structure of flow paths, which are importantfor safety. This is captured in the SDMs developed for these potential host rocks. It was notedthat information on the geological environment for Pre-Neogene sediments are relatively
limited compared to the other two potential host rocks examined.
Apart from advancing the specific knowledge of these geological settings, the SDM
methodology will be a key tool for synthesising data from other geological settings that may
result from the siting process. Furthermore, NUMO also continues developing the SDM
methodology, including how to assess and incorporate time variation, such as sea-levelchange, uplift and erosion or tectonic evolution.
Finally, an approach for handling the tectonic hazard, which could potentially be an issue
at some sites, has been developed. NUMO has also conducted an international project [56] on
how to assess, and mitigate, risks from the tectonic perturbations that potentially would be anissue at some sites.
(ii) Engineering knowledge base
A requirements-driven design process, outlined in Chapter 4, has been established. This
process will meet the needs of an implementing organisation to consider a wide range ofrequirements, including practicality of construction and operational safety, to adapt andoptimise designs during the different steps of the siting process and to develop repository
concepts for novel siting environments.
To date, assessment and further development of the H12 and TRU-2 concepts have been afocus, in particular for consideration of practical construction and operational safety. This led
to an alternative H12 design that utilises a PEM, in order to meet potential challenges during
waste package emplacement. Alternatives for overpack design, material and welding havealso been assessed. Regarding TRU waste, a waste package design that will be more robust
during the operational period has been studied, as well as alternatives for emplacement andbackfilling the waste. This report also focuses on the co-disposal option, taking into accountinteractions between the HLW and TRU repositories. Finally, it has been demonstrated how
to adapt repository layouts for these concepts for the three different siting environments
assessed.
As discussed in Chapter 2, it is inevitable that large uncertainties will exist at early stagesof the stepwise implementation programme and hence there is a need for the programme to be
sufficiently flexible, i.e. to have the capacity to deal with changed conditions and to evolve
towards implementation and eventual final closure accepted by all stakeholders [57]. This canbe achieved by identifying the most significant uncertainties and focusing R&D to either
7-36reduce uncertainty directly, or the impacts of this on repository performance. At the current
stage of the siting process, there is a need for developing robust repository concepts to reduce
impacts of safety case uncertainties. Furthermore, and especially since sites with properties
deviating from the three studied environments could emerge, there might also be a need todevelop novel repository concepts for such environments, as further discussed in Section 7.4.
A process for adapting repository design and layout to the specific geological conditions
has been demonstrated. In addition to the basic requirement of excluding active faults fromthe repository area, a more elaborate assessment is presented that assesses what to avoid,
depending on the scale of any faults or similar structural features present. Criteria on wherewaste and buffer materials can and cannot be placed (see Section 4.5.4) are presented. For
example, it is assumed that, for faults with a length of 1 km or more, the width of thedamaged zone surrounding the fault is about 1% of the fault length [5] and consequently
emplacement within this zone should be avoided. However, as there is likely to be uncertainty
in the position of the fault, emplacement close to the damaged zone boundary should also be
avoided. Furthermore, waste emplacement locations need to consider operational factors suchas waste transportation, appropriate drainage and ventilation systems, etc.
In this way, repository layout is adapted to characteristics such as the three-dimensional
distribution of faults, fractures and permeability represented in the SDM and this can bereflected in the RN migration analysis model for safety assessment. This process will thereby
allow feedback from the safety assessment, both to site characterisation and formulation of
criteria used in design work.
Engineered barrier designs have been improved and judged more reliable, reflecting new
test results on overpack corrosion and the swelling of buffer material (see Section 4.4). In
addition, as preparation for flexibly responding to various site environmental conditions,development of different design options was carried out (see Section 4.5.4). For example, for
HLW repositories, the introduction of a PEM for horizontal emplacement in a deposition
tunnel is judged to simplify the emplacement operation and quality control of the buffer, andto make it much less sensitive to ingress of groundwater during operation and re-saturationcompared to the vertical emplacement option using unprotected bentonite buffer blocks.
Layout, workability, operability and compatibility with the geological environment for deadend versus open-ended deposition tunnels have also been assessed.
Regarding TRU waste, an alternative waste package container with a lid allowing liftingfrom the top has been introduced to improve both handling and safety during normal
operations and also reduce vulnerability to abnormal states, such as dropping of waste.
Engineering technology has also developed in areas such as overpack welding, nondestructive testing and buffer manufacturing / installation (block method, in-situ compaction
method, pellet filling method, etc.). The applicability of some of these technologies has been
confirmed by empirical studies at full-scale (see Sections 4.4.1 and 4.4.2). Several full-scaledemonstration tests on waste deposition technology, grouting technology, tunnel plugconstruction technology, etc. have been conducted (in domestic and overseas URLs),
demonstrating their technical reliability. Furthermore, basic R&D has shown the basic
feasibility of the technology that supports retrievability of already emplaced waste, as
specified in the Final Disposal Policy.
Finally, focus on actual project implementation is a key aspect of the design work. For
example, regarding operational safety, safety measures for both above and below groundfacilities that refer to the regulatory standards for nuclear facilities (reviewed after the
Fukushima Daiichi Nuclear Power Station accident) were specifically examined. This means
7-37

that, for the operational phase, there is a need to look at low probability events if they arelikely to have large consequences. For underground facilities, in addition to nuclear safety,
key issues concern construction, ventilation, drainage and worker safety.
(iii) Safety assessment knowledge base
The safety case presented in this report demonstrates the use of several approaches andmethodologies that will allow it to identify further development needs and eventually be thebasis for site selection and licensing of a repository. More specifically, the following can benoted:


A fault tree approach is included for scenario development during the operational
phase.
A formal process for identifying and developing post-closure scenarios using astoryboard approach is developed and applied.
There has been a move towards increased realism and less conservatism, but, as
noted in Section 7.3.2, further developments in this direction are needed.
A more systematic and well-documented process for setting key parameters of RNs
migration is adopted.

These points are further elaborated in the following subsections.
(a) Safety measures and safety assessment before closure
Regarding safety before the closure of the repository, the regulatory standards for nuclear
facilities were reviewed after the accident at the Fukushima Daiichi Nuclear Power Stationand the implications for how to assess operational safety has also considered experience inother countries where repository projects are in, or nearing, the implementation stage. This
forms the basis for developing the more systematic and quantitative methodology for
operational safety assessment presented in this report.
An event tree approach is established to identify and evaluate scenarios that describe thetransition to abnormal or potentially detrimental conditions that could occur in the repository.
In addition, by creating and continuously expanding a database of such events in relevant
industries, it has become possible to ensure the traceability and completeness of the scenario
development process.
Using this technique, several potentially detrimental scenarios have been identified andassessed, including waste drop, fire, and loss of power. The probability of these events andthe ability of the waste and waste packages to retain their containment properties have beenassessed. Also examined were countermeasures in the unlikely event of an accident. It is
concluded that, with the right procedures and countermeasure in place, the occurrence andconsequences of assessed events are low and radiological impacts on workers or the public
are very unlikely(b) Assessment methodology for post closure safety
Currently, a risk-informed approach to assessing long-term radiological impacts is adopted,
based on the dose calculated for specific scenarios and the probability of these occurring. To
initiate this approach, systematic methods for developing scenarios belonging to each
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category were used. After formally describing the behaviour of the geological disposal systemafter closure, facilitated by the use of storyboards, a top-down method of considering thesafety functions expected for the components of the repository system was combined with abottom-up approach based on assessment of relevant FEPs. Such FEPs are compiled from thelatest internationally developed databases and the FEP lists used in previous Japanese safetyassessment to ensure that the resulting list (NUMO FEP list) is comprehensive.
In developing the models and setting up the data sets for scenario consequence analysis,
consideration was given to realistically reflecting the characteristics of both the SDM andrepository to the extent possible. Specifically, a three-dimensional model was utilised to
explicitly evaluate the arrangement of components of the repository with respect to features inwhich groundwater flows and RNs migrate within a limited portion of the surrounding hostrock. However, it is admitted that the scenario selection will need to be updated in Safety
Cases for specific sites and that the current groundwater and RN modelling approach issimplistic and does not fully take into account the three-dimensional aspects of the entire
repository in its setting within the SDM. As is elaborated upon in Section 7.3.2, further
developments in this direction is needed.
(iv) Progress in knowledge integration
Compiling and integrating geology, engineering and safety assessment knowledge within acoherent safety case is a key challenge. This has forced geoscientists, designers and safetyassessors in the programme to interact and provide feedback on development needs across a
range of different disciplines. Developing a comprehensive SDM as demonstrated in Chapter
3, using this for developing site-adapted designs (Chapter 4) and deriving associated safetyassessment calculation cases (Chapters 5 and 6) are critical steps in this process. In addition,
this work also provides the framework for the feedback loop where safety assessment should
guide further site characterisation and repository design work.
Assessing co-disposal of HLW and TRU waste is another area of knowledge integration.
This forces consideration of how evolutionary processes for TRU waste, such as the effect ofnitrate contained in Gr.3, would affect migration from the HLW part of the repository.
Using storyboards to visualise the temporal evolution of repository safety functions, as
demonstrated in Section 6.3.1, facilitates understanding of the behaviour of the repository
system and improves dialogue between experts in different fields. This approach will also
ensure consistency and traceability from scenario creation to formulating the corresponding
analysis cases, based on the understanding of the SDM and the related repository design.
However, it is recognised that storyboards alone may not suffice to develop a comprehensive
set of scenarios, especially since the different evolutions expressed may not be independent.
Further developments to improve this situation are thus planned.
7.3.2 Dealing with uncertainty in the safety case
When creating a safety case, it is necessary to take careful measures to deal withuncertainties associated with the limits of scientific and technical knowledge and the resulting
constraints on predictions of future evolution. Uncertainties originate from lack of system
understanding as reflected in the assessment basis, i.e. the site descriptive model, repository
design and approaches for safety assessment. Key uncertainties that are due to lack ofknowledge and those due to inherent variability need to be distinguished and the impacts ofthese understood.

7-39

As will be further outlined in the following subsections, this safety case has highlighted
many important uncertainties. This will form the basis for uncertainty management, such that
their impacts can be reduced in a structured manner as siting proceeds. However, as is also
pointed out in several places throughout the report, progress here has been limited. Instead,
uncertainties are often handled by simplifying descriptions, making conservative assumptions
and introducing stylisations. This approach is justified at present, since data from specificsites are not yet available and detailed designs for these remain to be developed. However,this also means that, at this stage, only limited conclusions can be drawn on key issues, likedifferences between sites, waste packages or layouts. In coming steps, it will be essential tointroduce much more realism in all components of the safety case and to quantify the
associated uncertainties. This will be a key goal for the future.
(1) Dealing with uncertainties related to the assessment basis(i) Uncertainty in the site descriptive model
As mentioned in Section 2.5.1, uncertainties related to the investigation and evaluation of
the temporal and spatial characteristics of geological environment include measurement errors,
measurement constraints (spatial resolution, measurement density, etc.) and lack of data.
There are also uncertainties in the characterisation data due inherent heterogeneities,
limitation of associated interpretation and the resulting conceptualisation of the geologicalenvironment. Since these uncertainties depend on the geological conditions of the site and themeasurement methods applied, it is impossible to quantify all of these at this stage. When
constructing the SDMs presented in this report, the distribution and trends of numerical datasuch as mechanical, hydraulic, and thermal characteristics are based on data collected on anationwide scale. Representative characteristic values are set for each host rock to be
examined, but uncertainty is not directly treated. These uncertainties will be considered whendeveloping the SDMs during the actual site characterisation stages, taking into consideration
the technology applied and the geological setting of the actual site.
In addition, as mentioned above (Section 7.2.1), the current SDMs do not reflect thetemporal changes in topography or characteristics of geological structures. In the future, as
described in Section 3.5.2, the construction of a 4D SDM will capture temporal evolution
with its associated uncertainties.
The characteristics of faults and fractures have a great influence on the extent of RN
migration. The inherent variability in parameters such as the hydraulic conductivity of therock matrix and preferential flow structures such as faults and fractures are represented asprobability distributions. Currently, large faults are represented by an average permeability. In
the coming characterisation of actual sites, the positions and orientations of large-scale faults
and active faults with lengths of >≈10 km will be described deterministically, since thesefeatures need to be excluded from the repository area. Smaller faults and fractures will betreated stochastically, based on assessed distributions of their characteristics. In safetyassessment, associated uncertainty will be handled by generating and analysing different
realisations based on the distributions, thus capturing the impacts of heterogeneity of thesefeatures. These uncertainties will, however, be reduced as the characterisation database
expands.

7-40

(ii) Uncertainties in the repository design and engineering
In the design of the repository, the uncertainty included in the SDM is taken into
consideration to some extent. Specifications have been set with margins for engineered
barriers and underground facilities to ensure that expected safety functions are notcompromised for a wide range of host rock properties. For example, in the design ofengineered barriers (see Section 4.4), the specification of the overpack is set to ensure that it
will remain tight for at least 1 ky after repository closure for all three host rocks examined.
Indeed, this safety margin results in a design that would likely maintain the safety function of
suppressing contact between vitrified waste and groundwater well in excess of 10 ky based ona more realistic assessment. In addition, the mechanical effects of various phenomena, such as
repeated seismic motions , corrosion expansion of metals, chemical alteration of buffer, andcreep deformation of bedrock and buffer materials, are currently evaluated withconservatively set parameters that may have greatly exaggerated potential effects on thesafety functions. Therefore, depending on the conditions of the geological environment, thecurrently applied margins and criteria should be revisited.
In addition, before sites have been specified, it is necessary to prepare for uncertainties inboth site environmental conditions and socio-economic requirements. For this reason, as
described in Section 4.8.3, consideration of tailoring is based on multiple requirements,
including safety and engineering practicality for different geological and environmental
conditions, ease of material procurement, and economics / environmental impacts. At thesame time, alternative materials for engineered barriers and further design options are being
developed. Advancing the capabilities of technology for evaluating the long-term behaviourof the EBS and assessing the uncertainties related to quality during manufacturing andconstruction of engineered barriers, will be important in order to reduce uncertainty related toquantifying long-term EBS performance. Further work on confirming the applicability ofengineering technology through demonstration tests will be carried out in order to examine
measures to reduce current safety margins based on reduction of some uncertainties
associated with repository design.
(iii) Uncertainty associated with safety assessment
In the pre-closure safety assessment, abnormal state scenarios in which safety measures do
not function normally are developed. Efforts are being made to improve the evaluations ofsuch scenarios within an international project on the operational safety of geological disposal[58], which may reduce associated uncertainties.
The range of knowledge and data available on post-closure behaviour of the repository is
considered sufficient for the current stage of assessment. In consideration of known
uncertainties, the evaluation is divided into base scenarios, variant scenarios and low
probability perturbation scenarios. At present, safety evaluation is performed to confirm
feasibility of safety demonstration, taking a conservative approach and intentionally using
models and data that overestimate consequences. This applies not only to the variant scenarios,
but also to the analysis cases for the base scenarios. Additionally, not only a dose constraint
value (300 μSv/y) consistent with recommendations by international organisations is
considered, but also the strictest limit from safety regulations of other countries is set as atarget value for the base scenario (10 μSv/y). Further, within safety assessment scenarios,
FEPs in the NUMO list considered to have a favourable effect on safety are not included if
there are no reliable data or models to quantify them (often termed reserve FEPs, see Section
6.3.2 (3)).
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In the modelling, the focus was on the containment performance of the repository, and it isassumed that RNs that migrate from the repository scale area (several km x several km)
directly enter the biosphere for dose calculation purposes (see Section 6.4.1 (5)). As a result,the evaluation is conservative with regard to the safety functions of RN sorption andmigration in the large volume of rock between the repository and the biosphere.
For the RN migration parameters, as described in Section 6.4.2, conservative values forsolubility and sorption coefficients are selected. In addition, a sensitivity analysis isperformed by considering a wider range of migration parameter values. Apart from cases withunrealistic combinations of parameter values, none of the results exceeded 300 μSv/y.
Low probability perturbation scenarios and human intrusion scenarios are, in accordancewith international practices, handled by a stylised approach. Some parameter values used arealso conservatively set so that the effects of these phenomena are overestimated (see Sections
6.4.3 and 6.5). However, even if such a stylised approach implies that detailed site-specificanalyses may not be needed, current assumptions are possibly oversimplified and will berevised in coming assessments.
In both the process evaluations and RN migration analyses carried out in this report,
limitations in computational capacity restricted models of some of the phenomena occurringin the disposal system and on the scale of RN migration considered. In addition, the migration
data is constrained by the limited range of information in the existing database. For example,
U-233 migration data in groundwater with a high carbonate concentration does not exist atpresent. Such limitations are handled by a conservative approach, i.e. by discarding sorption
both in the near- and far-fields in cases where there are no reliable data.
As further discussed in Section 6.6.2, several actions are planned to further improve thetechnical basis for safety assessment and carry out sensitivity analyses to identify and reduce
key uncertainties. This includes the following initiatives:
1.

Make assessments more realistic and to account for all components of the repository
system in their specific geological settings. In particular, more realistic RN release andmigration models for both the EBS and geosphere should better reflect the threedimensional characteristics of both for specific sites and the repository concepts
tailored to them, in line with the stepwise improvement of the knowledge base. In
addition, more emphasis on model testing, verification and validation will be needed
in later stages of this process. However, when uncertainties are large and difficult toquantify, conservative assumptions will also be needed in the future.

2.

In order to improve traceability, the components within the workflow, fromdevelopment of scenarios to setting of analysis cases based on the understanding of
the behaviour the repository system, will be formulated in a more systematic way,utilising advanced knowledge management tools.

3.

In terms of such technology development, focus will be on the actual volunteer sites inorder to ensure that more reliable assessment of safety functions of repositories
tailored to them can effectively contribute to the comparison between sites and the
optimisation of repository designs for these.

4.

Continual expansion of the RN release and migration database for relevant geologicalenvironments and associated evolution of the EBS for the wastes considered.

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(2) Confidence in dealing with uncertainty
In the post-closure safety evaluation of this report, as mentioned above, it is essential toensure confidence in the process of integrating individual evaluations based on the knowledge
available at this stage. Thus, safety is also assessed using complementary indicators (otherthan dose) that are not affected by uncertainties regarding future human lifestyles, surface
environment conditions or the long-term geological stability assumed in the main analysis. In
this way, credibility of conclusions is built even in the light of such uncertainties.
Information on uncertainties related to site characterisation and modelling should be
integrated into the SDM and reflected in repository design and safety evaluation. Users thenprovide feedback on how to focus further characterisation work. Based on this safety case,some factors related to the geological environment were identified that could have asignificant impact on engineering practicality and safety of the repository, and uncertainties
associated with these could be reduced when setting targets and their priorities for coming sitecharacterisation stages will be clarified. Using safety sensitivity analyses to focus sitecharacterisation and design work will continue and be further developed as the programmemoves forward.
Once characterisation efforts are started for a specific site, the understanding of thegeological environment will be enhanced and factors that are highly relevant to uncertainty(for example, spatial distribution of faults and hydrological characteristics) will be assessedand better described in the SDM, thereby reducing uncertainty. At the same time, cooperationand iterations between the site characterisation, repository design and safety assessment teams
should be more focused in the future.
In order to facilitate integration and cooperation, changes in geology and various otherconditions that may be expected over the long project period should be considered whencreating a safety case. Thus, Section 7.3.3 presents a management framework and method of
responding to uncertainty related to development of the SDM, the design of the repository,
and the safety assessment which will increase confidence in the overall project.
7.3.3 Structured management of safety case development(1) R&D management
In order to constantly improve safety cases, it is necessary to clarify key open issues andestablish a R&D framework to resolve these. In Japan, this framework has already been
established and will undergo continuous review to assure that it meets NUMO’s (and otherstakeholders’) requirements.
As mentioned in Section 2.5.5, the Coordination Council on R&D of Geological Disposal,
consisting of METI, NUMO and related research institutes, has been established to provide
the framework for tackling R&D issues related to geological disposal. Goals and priorities areagreed and set out in an Overall R&D Plan [59], which is reviewed as appropriate in order toreflect the progress of science and technology, changes in requirements, and the R&D results
obtained.
In line with the overall plan, NUMO compiles a medium-term technology development
plan every five years, with a clear division of roles between basic R&D institutions. Thisexplicitly identifies progress required for the safety cases needed for future milestones, with acurrent focus on planning PI and integrating output as a basis for site selection [60]. At astrategic level, key components of this plan are regularly reviewed by Japanese and7-43international experts in NUMO’s Technical Advisory Committee (TAC). This role waspreviously covered by separate ITAC (International Technical Advisory Committee) andDTAC (Domestic Technical Advisory Committee), but integration was implemented to
improve both efficiency and development of common domestic/international perspectives.
Assurance that the output of R&D is of sufficient quality and is representative of the stateof-the art is provided by regular publication in scientific journals and presentations at national
and international conferences, together with focused national/international review workshops
on key topics.
(2) Promotion of R&D cooperation
In order to effectively promote technological development, NUMO participates in joint
research with related organisations in Japan and in collaborative projects with international
partners. This facilitates capture of both tacit and explicit knowledge and access to
infrastructure that NUMO does not possess, particularly conventional laboratories, test
facilities and URLs.
Such research cooperation is internationally recognised to contribute greatly to thedevelopment and implementation of geological disposal projects (e.g., OECD/NEA [61]),
resulting in a wide range of initiatives coordinated and/or funded by the IAEA, NEA, EU andsimilar organisations. NUMO and supporting R&D organisations actively participate in manysuch joint international projects, but also collaborate with partner implementers in bilateral or
multilateral R&D projects [62]. In addition to input received, NUMO will actively contributeby sharing the results of technological development in Japan and also the experience gained increating safety cases at various international cooperation forums.
Research cooperation is also beneficial for human resource development (see (5) below),
and mid-level or higher-level staff are involved in overseas projects in order to maintain andextend the international level of technology. At the same time as solving problems, NUMO istrying to improve technical capabilities by participating in an international network of experts.
A cooperative framework for young staff members has been established, so that they canexperience relevant work from test planning to test implementation and data acquisition, andthus develop as the key scientists and engineers who will carry NUMO forward in the future.
In the future, topics that may be ideal for collaboration due to common interests with someinternational partners will be examined – such as designing and assessing safety of coastal
repositories, repository design optimisation, and developing new analogues to test key areasof safety cases (e.g., RN release and transport). Here, NUMO would expect to play more of alead role than has been the case in the past.
(3) Structured response to changing boundary conditions
It is understood that the boundary conditions for the safety case will change with time andas the siting programme proceeds. Such changes need to be managed in a structured way.
As the nuclear power programme in Japan evolves, this will affect both the volumes andtypes of nuclear waste that will eventually need to be disposed of. This may primarily affectthe siting programme, as a larger disposal area may be needed, and the repository design work,
since new designs may be needed to handle other types of waste. However, the general

7-44

approach in developing SDMs and the requirements-driven design process is judged
sufficiently flexible to handle such potential developments.
Nuclear safety regulations may also change, e.g. in the approach to handling risk or to what
extent they will be more, or less, prescriptive on specific requirements for sites or engineered
barriers. The safety case methodology must be sufficiently flexible to handle such changes. It
is judged that NUMO’s safety case methodology has this flexibility, since it builds oninternationally accepted principles and practices applicable in a very wide range of regulatoryenvironments.
Properties and other conditions at sites that will come forward and then be investigated arelargely unknown at present. Surprises may also come later, when detailed underground data
become available during the PI and DI stages. Repository designs may need to be revised andthe means of assessing the safety relevance of these surprises must be available. The
comprehensive SDM methodology and the requirements-driven design process developed andapplied are judged capable of providing sufficient resilience, i.e., the capacity to deal withchanged conditions and meet associated challenges. Again, the fact that NUMO’s safety casemethodology is in line with principles and practices applicable in a very wide range of siting
environments worldwide adds to this confidence.
As elaborated in Section 4.1, repository design will be optimised in steps starting from a
conceptual design, followed by basic design and finally, detailed design that will be tailoredto the developing understanding of the siting environment. The word “optimisation” has a
special meaning in radiation protection, but needs special consideration for geologicalrepositories1. Thus, optimisation in NUMO’s programme goes beyond radiological safety andalso includes searching for more practical and efficient repository designs, while ensuring theradiological and other environmental impacts are kept sufficiently low. Furthermore,
optimisation needs to balancing different, often conflicting, requirements. For example, thedevelopment of technologies that support retrievability shown in Section 4.7 is response touncertain social factors, such as changes in requirements from stakeholders. A holistic
approach to optimisation allows parallel consideration of a range of relevant issues, including
efficiency, worker safety during repository construction, operational radiological safety, postclosure safety, environmental impact and efficient use of natural resources.
Optimisation is related to the concept of BAT, but is not the same thing. In the European
Union, BAT means the most effective and advanced stage in the development of activities
and their methods of operation. This “indicates the practical suitability of particulartechniques for providing the basis for emission limit values and other permit conditions
designed to prevent and, where that is not practicable, to reduce emissions and the impact on
the environment as a whole” (IED (2010/75/EU) [64]).
Application of BAT, in the context of geological disposal, means that the technology used
for siting, design, construction, operation and closure of the repository should be suitable forthe overall goal to prevent, limit and delay releases from both engineered and geologicalbarriers as far as is practically possible. Since final designs are not yet selected, BAT can also
be seen as an awareness of the current state-of-the-art, to ensure that development is not
1

For these cases the ICRP [63] states that the elements guiding or directing the optimisation process should be
those that directly or indirectly determine the quality of the components of the facility as built, operated, andclosed, where quality refers to the capacity of the components to fulfil the safety functions of containment andisolation in a robust manner. The assessment and judgement of the quality of system components essentially
includes the site characteristics, elements of Best Available Technique, as well as the concepts of good practice,sound engineering, and managerial principles.

7-45

frozen and technology is further developed and modifications to concepts and designs
introduced as and when appropriate. Once an optimised concept is decided upon, therefore,
the best available technology is chosen for its implementation.
Science and technology will advance during the course of the repository project andtechnological breakthroughs may allow for novel or more optimal designs. To meet thesechallenges, it is essential that the programme retains and further develops a sufficientscientific and technological competence. Ensuring that there is a comprehensive
understanding of the structure and content of the NUMO SC is essential for this. In meetingthese challenges NUMO has also started to build a system for systematically managing
knowledge and human resources (Sections (4) and (5) below).
(4) Requirements and knowledge management
A safety case requires the compilation, integration and synthesis of a huge volume ofdiverse, multidisciplinary knowledge in a structured and quality-assured manner. Furthermore,such a knowledge base is expanding at an exponential rate, which can be expected to continue
over the duration of the repository project (in the order of a century – or longer if institutional
control is considered). It is clear that the traditional knowledge management (KM) approaches
used at the time of the H12 and TRU-2 reports are no longer suitable for this task – as wellrecognised in other fields that handle “big data”. JAEA, as an R&D organisation supporting
both geological disposal implementers and regulators, established a project in 2005 to develop
improved KM approaches and tools (e.g., [65] [66]). In the development of such an advanced
knowledge management system (KMS), both explicit and tacit knowledge were included,with application of cutting-edge information and knowledge engineering tools. The resulting
KMS was focused on safety case development and its effectiveness demonstrated in the H22
project, made available on the “CoolRep” web-based communication platform (e.g., [67][68]).
International guidance on what to include in a safety assessment, e.g. as provided by the
OECD/NEA [69], to a large extent originates from intercomparisons between safetyassessments developed for a wide range of geological environments, waste inventories andstages of development. Since NUMO’s safety case aims to be consistent with such advice(Section 1.4.2), this provides further confidence in the aim for completeness of the work andis an essential source for enhancing the knowledge base.
However, the information, data and knowledge to be incorporated in a safety assessment isdramatically increasing and hence the intercomparison of different safety assessments is very
challenging and requires participation of experts in all the different disciplines contributing todevelop a safety case. The inter-comparison of safety assessment is a key area forinternational collaboration and an advanced knowledge management system can also be of agreat help here, providing a common structured framework for all participants.
As mentioned in Sections 2.5.3 and 2.5.4, NUMO is developing systems for both
requirements and knowledge management, with emphasis on its roles as a repository
implementer [70]. In developing safety cases that evolve but preserve traceability over thelong project implementation period, ensuring that decisions are made in an open andtransparent manner, it is internationally agreed good practice to introduce a requirements
management system (RMS). This integrates constraints set by environmental protection,societal acceptance and economics with the fundamental principles, regulations andguidelines for geological disposal, as suggested by the IAEA [26]. As a result, the RMS
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facilitates developing solutions that satisfying occasionally conflicting requirements, withappropriate balances and trade-offs explicitly recognised.
There is a clear relationship between the RMS, which defines the decisions to be made,and the KMS, which provides the knowledge to allow decisions to be soundly based – as wasrecognised during development of the JAEA KMS (e.g., [65] [66]). The NUMO concept for
combining requirements and knowledge management within an integrated system is shown in
Figure 7.3-1.
The development of such a combined requirements and knowledge management system
for geological disposal is included in both the most recent Overall R&D Plan and the NUMO
medium-term technology development plan mentioned above. Knowledge management inradioactive waste management is increasingly recognised an important issue, leading toinitiatives like the OECD/NEA international project WP-IDKM (Working Party On
Information, Data And Knowledge Management [71]) and NUMO will actively participate insuch activities to ensure that our developed system is and remains at state-of-the-art levels.

Figure 7.3-1 NUMO requirements management concept and its relationship to knowledge
management(5) Human resource management
In order to make geological disposal sustainable for its very long implementation period,
required human resources need to be available, with critical tacit knowledge preserved andcommunicated/transferred across generations. This is highlighted in both the Overall R&D
Plan and NUMO 's medium-term technology development plan. Traditionally, an effective
approach was for experienced specialists to pass on their experience and know-how to young
staff through on-the-job training (OJT).
Such OJT forms a key component of both general R&D and collaborative projects, whichis recognised by NUMO. Indeed, this was an explicit aim of the current report, which was
produced predominantly by a young team rather than the senior experts responsible for the
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H12 and TRU-2 reports. Even if this is not the most efficient way to produce a safety case, theclose interaction between the production team and both national and international expert
reviewers provided an invaluable opportunity for knowledge transfer.
It is important that human resource development efforts are not limited to NUMO, which is
the implementer, but are widely promoted by regulatory agencies, related research institutes,
and industry, which are major producers of new knowledge. In response to the Overall R&D
Plan, the Ministry of Economy, Trade and Industry (METI), NUMO, and related research
institutes are promoting efforts to jointly hold training courses for young staff from bothindustry and universities.
Tacit knowledge capture is a particularly urgent concern in Japan, due to aging of highly
experienced specialists and generalists who have been involved in past geological disposal
planning, with a clear risk of loss of institutional knowledge and know-how. It is difficult tocapture such experience and know-how and convert it into explicit knowledge, and OJT is atime-consuming process, so this is a priority area for application of tools to support tacitknowledge preservation – for example based on expert systems.
(6) Communication of safety cases to stakeholders
Safety cases provide important background for stakeholders, facilitating theirunderstanding of the key issues associated with geological disposal and encouraging dialogue
to ensure their contribution to decision making. In order to perform this role, a safety casemust be communicated at technical levels appropriate to the different stakeholder groups.
Stakeholder communication is also recognised as a challenge at the international level, forexample in the long-running OECD/NEA Forum on Stakeholder Confidence (FSC;
established in 2000) [72]. NUMO actively participates in such international activities, with
communication of the safety case being promoted as a special area of concern.
This particular report was developed at a technical professional level, aimed in particular at
experts in the geological disposal field, and hence is not easy for non-experts to follow. Thus,to respond to the interests of stakeholders and promote communication, it is complemented bya pamphlet explaining the main points that this report aims to present (“The concept ofensuring the safety of geological disposal” [73], in Japanese only). A wider perspective,
including more international background, is provided in the report “Why Geological Disposal”[74] (also in Japanese only).
To facilitate accessibility, NUMO has also developed a web system for viewing the entire
safety case documentation (currently in Japanese only), making it easy to read or download
the report as a whole or as summaries (e.g. Main Report chapters and Supporting Reports).
Specifically, the Supporting Reports here are linked directly to chapters that reference them,making it easier to browse specific topics. In addition, wherever possible, references cited inthis report are also linked to the main text and can be directly accessed as pdfs (including
NUMO technical reports, JAEA research and development reports, and reports of some othersupporting organisations). To widen availability, such access is possible not only frompersonal computers but also from other devices such as tablets. In addition to documentation,
more direct presentation of the safety case to various stakeholders by qualified experts is also
promoted [75].
Through nationwide communication with stakeholders via interactive briefings [76] andestablishment of open and transparent dialogue in volunteers areas where the initial literaturesearch is accepted, NUMO will continue to address any stakeholder concerns related to the
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safety case and reflect them in future planning. When the programme advances to specificsites, efforts to enhance and widen communication with all stakeholders will progress toensure that different stakeholders can also be involved in relevant parts of future safety caseproduction.
7.3.4 Acceptance of safety case content
The efforts described above to develop and improve components of the safety case arenecessary but not sufficient. In order to assure that a safety case is fit for a specified purpose,
it is extremely important to conduct independent reviews of the safety case as a whole. As
mentioned in Chapter 1, when preparing this report, NUMO published a Japanese language
review version and asked the Atomic Energy Society of Japan (AESJ) to organise such a
review by qualified experts who were not directly involved in its preparation. This technicalreview was then used to make revisions reflected in the present text – which is then assumed
as accepted to be fit for its role at this pre-siting stage. NUMO will confirm this by reviews ofthis English version, in the first instance by our TAC and then by an international organisationsuch as NEA.
Also, in the future, NUMO will continue to carry out independent safety case reviews by
national and international organisations, to ensure that these reflect the progress of scienceand technology, trends in safety regulations, etc., and thus that these are fit for the specified
purposes of the milestones in our evolving programme.

7.4 Outlook for future safety cases
The previous sections outline the benefits and limitations of the present safety case to serve
as a technical base for stepwise site investigations, leading to site selection and, eventually,
repository implementation. It provides direct input to guide literature searches at volunteer
sites, as have recently been initiated in Hokkaido. However extensive development is required
to develop the site-specific safety cases that utilise LS information to support selection of
PIAs and planning the focused investigations that will be required at each site. This, in turn,will lead to further development needed for future programme milestones, as discussed in
more detail below.
7.4.1 Building an evolving safety case(1) Development of site-specific safety cases
As noted in Section 2.2.3, geological information obtained at each stage of stepwise site
investigation provides an evolving basis for repository design and safety evaluation,
contributing towards determining the eligibility of a site (or specific areas within it). The
safety case and the developed RMS serve a key integration role, assuring safety of
construction, operation and after closure, which are the highest level requirements for apotential project, even if these may not be entirely without conflicts that require careful
resolution. Site selection will, however, require considerations in addition to safety but, evenafter a site is selected, the safety case will play a key role in moving towards and eventually
applying for construction and operation licenses. Indeed, even after construction is initiated,
geological and engineering information obtained will result in continuous updating of theknowledge base supporting the project and the safety case will facilitate optimisation of theimplementation plan to take advantage of this. It is also expected that regulations will require7-49regular safety re-evaluations during the operational period, up to a final safety case to support
licensing closure of the repository. This is summarised in Figure 7.4-1, which also indicatesthe changing boundary conditions and constraints over this period.

Figure 7.4-1 Concept of stepwise safety case evolution(2) Tailoring the safety case to stepwise site selection
In the following outline, an idealised process for moving through the siting process ispresented, with a focus on the role of the safety case within it. It emphasises assuringflexibility, but it is recognised, based on international experience, that surprises andperturbations may arise that disrupt this process – not only due to the findings during site
investigations but also changes of external boundary conditions. Nevertheless, this provides agood basis for planning and will, in the future, be complemented by more detailed risk
assessment to determine potential perturbations and strategies to respond to them.
(i) Literature survey stage
At this stage, it is first required to confirm that the legal requirements for an acceptable site
are met. From the viewpoint of creating a safety case, the geological information that can beobtained is limited and it is likely that input will include only the extent and location ofpotential host rock(s) and very general properties of associated geological setting(s). However,in order to prepare for the next siting stage, additional work is needed during LS.
While fulfilling the legal requirements, consent of the local municipality will be required togo forward to the PI stage. It also clear that only sites that fulfil the legal requirements andhave reasonable prospects of hosting the repository should be considered for further study.
This will be especially important if many different LS sites arise that meet this basic
qualification. Planning the site investigations for the PI stage is based on what is obtainablefrom the LS and identifying the repository concept development needs, with the dual
objectives of determining the prospect of the site being suitable and forming the basis forplanning effort in case the site is selected to go forward.
Therefore, regarding the production of the safety case, technical improvements will beneeded, as discussed in Sections 7.3.3 (1) and (2). In addition to study of the legal
requirements, the following will be assessed:

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

Arrangement of underground conditions such as distribution of strata, rock formations,faults, etc.

Examination of which strata are considered to be more preferable.

Examination of land use restrictions, etc.

Based on these considerations, the scope and uncertainties of the safety cases presented inthis report will be re-assessed, reflecting available site-specific information and considering
how key uncertainties will be handled in case the site goes forward to PI. It can also be noted
that, in the case of a coastal site, the seabed will be considered in the LS as a potential
footprint for the repository.
(ii) Preliminary investigation stage
In the PI, a wider range of information will be obtained through field work. The geologicalinformation obtained will be integrated into a 4D SDM, which will be the basis of tailoring
potential designs of HLW/TRU waste repositories and conducting associated safetyassessments.
In creating the safety case at the PI stage, it is considered that the following would bespecific examples of such studies:Construction of a more detailed repository scale SDM, based on geologicalenvironment information obtained from measured values, such as the threedimensional distribution of potential host rock(s) and the mechanical and hydrauliccharacteristics of these and surrounding formations. It is accepted that bothcharacterisation and development of the SDM might be more complicated if thegeological setting of the sites differs greatly from those assessed in this report.
Practicalities of the actual investigations will also differ, especially if potential hostrocks are located under the seabed.

It will be important to capture past, ongoing and future changes of the sites, such as
sea-level change and uplift/erosion. Methods for 4D site descriptive modelling alreadytested will be applied and further developed. This would be especially important for acoastal location and it is essential that the required assessment tools and databases will
be available for this assessment.



Repository design(s) and layout(s), including installation location and depth, adapted
to site conditions as captured in the SDM, would be developed for further assessmentof safety before and after repository closure. In case siting environments strongly
deviate from those studied in the current safety case, this might imply a needdevelopment of new repository concepts suitable for these sites.



Confirmation of implementation practicality, based on engineering technology
verification tests related to construction, operation, and closure of the designed
repository.

Measures to ensure safety of the repository before closure and to protect thesurrounding environment will be further developed. There will also be special focus
on the safety of construction and an integrated assessment of interactions between
construction and operation will be needed.

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

Application of RN migration model and parameters that more realistically reflect thecharacteristics of the designed repository, the host rock and groundwater chemistry.
Also modelling of the biosphere that takes into account the surface environment andhuman lifestyle of the site.



As identified in Section 7.3.2 a key objective would be a much more comprehensive
safety assessment, determining the importance of remaining uncertainties, based on amore realistic abstraction of site and design information and using less conservative
assumptions in the consequence analysis. Thereby, the safety assessment could
provide key feedback on what may need to be explored during the DI stage.

Based on the findings from a more comprehensive uncertainty assessment, thepotential for optimisation of repository designs will be assessed.

When selecting a DI site, it may be necessary to compare repository designs for each of the
PI sites. It is important that such designs consider site-specific environmental conditions andinclude optimisation based on design factors. Here the safety case can be useful to support
comparisons, improving transparency in the choice of DI. The uncertainty analysis will bevery important in this respect, as the effort required to reduce uncertainties may be very site
specific. Limitations on direct accessibility to the suitable host formations could potentially behandled by locating the surface and underground facilities far apart, which then need to beconnected by an underground access tunnel. This could, for example, be the case if thesuitable host rock is located offshore. Different options have to be compared in terms ofpros/cons in a structured manner. Furthermore, the situation may arise that there will be nosingle suitable formation at a potentially site suitable for both HLW and TRU, as well as thelocations of surface and underground facilities(iii) Detailed investigation stage
The key aim of the DI-stage is to develop an optimal site-specific repository design,
proven to provide both operational and post-closure safety through a comprehensive safetycase. Detailed planning of the DI stage would have to wait until the PI stage is underway,
since the relative importance of different issues will strongly depend on the knowledge gained
during the PI stage.
Following further surface-based investigations, the DI stage involves constructing a sitespecific underground investigation facility (UIF), allowing both underground investigations
and demonstration tests of the engineering technology [77]. An updated SDM and repository
design / layout adapted to this model will be developed. Based on this, and other scientificknowledge obtained, the safety case will be refined. This will determine if the site conditions,
technology and other information necessary for applying for a repository construction license
are in place.
Since the DI stage could lead to a decision to start construction of a repository at the site, itis essential that the underground activities in the UIF are planned such that they would notcause unnecessary perturbations to the host formation that might impair its suitability. Testsand data from generic URLs should thus be utilised to the extent possible. Development of
repository designs and installation techniques would not necessarily require access to theactual host formation and several research issues could, at least initially, be studied elsewhere.

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7.4.2 Fundamental safety case improvements
In addition to tailoring the safety case to the siting process, there are other fundamental,
site-independent improvements that are needed to keep a safety case fit for purpose over a
period in the order of a century, as discussed further below.
(1) Continuous improvement of safety cases
NUMO will improve the assessment basis for safety cases by continuing to tackle theissues identified in the R&D plan, as described in Section 7.3.3 (1). Such improvements
concern the scientific and technical knowledge base, modelling technology and approaches toverification and validation. These issues are common to all sites, with continuous
improvement of the confidence of safety cases essential when assuring the general feasibilityof safe geological disposal in Japan. This is important from the viewpoint of encouraging
more communities to accept a LS, since safety is a critical issue for discussions to support
gaining understanding and establishing credibility.
When a site is identified, NUMO will proceed with production of a site-specific safety casethat reflects the characteristics of that location. For this also, the assessment basis needs to becontinuously improved, with this safety case developed while site investigations ongoing. In
this way, issues specific to site conditions will be clarified, with plans flexibly reviewed sothat any R&D necessary can be carried out for these issues.
There are restrictions on the human resources and budgets available for producing safetycases and associated R&D. Thus a balance is needed between efforts on more generic issues
and those focused on site-specific safety cases. As the site selection progresses, and narrows
down to a few locations, efforts to improve the confidence of the site-specific safety cases
will be the main priority. Nevertheless, even in this case, it is necessary to maintain the R&D
framework to provide a wider perspective of key knowledge areas included within theassessment basis.
(2) Maintaining state-of-the-art technology and knowledge baseIn addition to the considerations above, as part of general confidence building it isimportant to advance research and development according to R&D plan mentioned in Section
7.3.3. To promote this effectively, NUMO takes part in the Coordination Council on R&D of
Geological Disposal and also conducts joint research with overseas organisations throughparticipation in international joint projects (see Section 7.3.3 (2)). Thereby, NUMO willcontinue to maintain state-of-the-art knowledge related to safety cases and geo-environmental
characterisation technology while conducting planned work during the LS, PI and DI stages.
This will also include a accumulating explicit and tacit knowledge required to conduct
demonstration tests as required at different stages of the siting programme.
The development of both the requirements and knowledge management systems describedin Section 7.3.3 (4) can be expected to be an effective tool for efficiently and appropriately
managing the different R&D issues that will need to addressed after sites are identified.
Considering the long period during which the disposal programme will evolve, thesemanagement tools will also continue to be maintained at the state -of-the-art.

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(3) Optimisation of repository design and implementation
As emphasised previously, safety is an absolute requirement, but the more holistic
overviews provided by advanced safety cases may allow this to be assured while optimisingother aspects of the overall repository implementation programme, such as social acceptance,
cost, sustainability, environmental impact and spin-off benefits (e.g. utilisation of repository
heat, spoil, etc.).
The design process described in Chapter 4 considers different design factors that need to beaddressed in the project implementation, such as operational safety, post-closure safety,
engineering feasibility, retrievability, environmental impact, and socio-economical aspects.
These design factors lead to requirements on repository components and their specifications(materials, configurations, dimensions, etc.), and it needs to be ensured that all requirements
are sufficiently met, with explicit considerations of any trade-offs that may be required.
Therefore, development of a clear methodology for performing overall optimisation is an
important issue to be addressed in the future (see Section 7.2.1 (2) (iv)). Here, the design
factors and the requirements related to them are defined in association with the upper-level
requirements systematically managed by the RMS described in Section 7.3.3 (4).
Starting from the generic safety case presented in this report, Figure 7.4-2 displays how
site-specific safety cases lead to repository design optimised in light of design factors,
assuring that it conforms to the site environmental and geological conditions that aregradually refined by the site survey, and thus leads to selection of a preferred site/concept forlicense application. Within this process, repository design considers requirements that go
beyond purely technical issues, but the safety case can act as a basis for soliciting stakeholder
input to relevant decision-making [78].

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Figure 7.4-2 Repository design development on the basis of site-specific safety cases

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7.5 Summary
This chapter demonstrates that NUMO has established methodology for selecting
favourable geological environments and developing SDMs for representative host rocks,
established an approach for designing a repository for such SDMs with the required safety
features that is consistent with the current state-of-the-art, and established methodology, tools
and data that form a sound basis for assessment of the safety of these disposal concepts insuch geological environments. These have been integrated into the current safety case, which
provides the knowledge base required in order to proceed with siting after the Literature
Survey commences in volunteer communities. This chapter also functions as a standalone
description of the Safety Case for readers who have not read the other chapters. Thus, some of
the key findings in earlier chapters are summarised, but the chapter also adds additional
elements needed for a safety case. The latter includes supporting arguments for post-closure
safety in terms of complementary indicators, supporting arguments from natural analogues
and comparison with other safety cases, as well as an identification of limitations andconstraints on the current safety case.
In order to move further in developing safety cases to support the siting programme,
further development is needed. For this reason, an outline of the strengths and limitations of
the present safety case serves to identify and prioritise the technical R&D needed to build theknowledge base to support literature searches at volunteer sites, stepwise site investigations
and design tailoring leading to site/design selection and, eventually, repository
implementation. Additionally extensive development aims to support production ofincreasingly detailed site-specific safety cases, which involves both development ofmethodology to manage the production process (e.g. advanced requirements and knowledge
management tools) and assuring availability of essential resources – in particular experienced
staff. Finally, the wider role of the safety case in decision-making during stepwise siting isoutlines, with special emphasis on establishing dialogue with non-technical stakeholders andbringing them into the decision-making process.

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Supporting Reports (SRs)
SR 7-1
SR 7-2
SR 7-3SR 7-4
SR 7-5
SR 7-6The concept of complementary performance indicators
Evolution of radioactivity in components of the geological disposal system
Evolution of radiotoxicity in components of the geological disposal system
Alternative indicator of release: potential radiotoxicity in river water
Natural analogues
History of external review

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8 SUMMARY AND CONCLUSIONS
The aim of this report is to assess the basis for ensuring safe geological disposal of HLW
and TRU waste for the geological environments found in Japan, from NUMO’s perspective asa repository implementer. Based on the latest scientific knowledge and the results oftechnological development to date, this report presents a preliminary safety case forrepository concepts tailored to representative geological environments. As such it will provide
a basis for both initiation of site investigations, and, to guide further activities in research,development and design work. In accordance with the IAEA glossary [1] a safety case in thisearly stage of development should acknowledge the existence of any unresolved issues andshould provide guidance for work to resolve these issues at future development stages.
In this report, the following elements are presented:•

The methodology for selecting appropriate Japanese geological environments to host arepository for disposal of HLW and TRU wastes is illustrated. This leads to capture ofunderstanding in three representative site descriptive models (SDMs) to serve as thebasis for the design and safety assessment of a repository. The process of SDM
development also leads to identification of the approaches and technology required for
future site-specific investigations.

•

The methodology for repository design of to meet the requirements for safety andengineering feasibility is developed and examples of design specifications are given
for each SDM. The development and demonstration of the individual technologies
required to construct, operate and close a repository based on the design, is shown tobe progressing steadily and priorities for future technological development are
identified.

•A series of methods and analytical techniques have been developed to assess the safetyof the repository, before and after closure, reflecting the characteristics of the SDMs
and the repository designs tailored to them. These form the basis for safetyassessments that illustrate how specified safety functions of the repository system canbe evaluated and assessed against refined performance targets in terms of radiologicalimpacts.

•

In order to further improve the reliability of the techniques for investigation andevaluation of the geological environment, repository design and safety assessment,
and to increase their applicability to expected volunteer sites, this report identifies
issues that need to be addressed in the future. In this way, the direction oftechnological development for the implementation of this long-term project isindicated. In addition, NUMO is laying the groundwork for quality and knowledge
management and human resource development to support the project.

•

The safety case presented in this report will be updated and extended to build
confidence in the safety and practicality of geological disposal projects in a stepwisemanner as the investigations leading to site selection and eventual repository licensing
proceed.

In accordance with the goals and boundary conditions set for this report in Chapter 2 it isdemonstrated that NUMO has:

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•

Established an approach for designing a repository with the required safety features in
relevant Japanese geological environments that is consistent with the current scientificand technological state-of-the-art,•

Established a safety assessment methodology, tools and data that form a sound basis
for evaluation of the safety of geological disposal concepts tailored to appropriate
geological environments, and•Made efforts to build confidence in the safety case in terms of supporting arguments inwhich the quality of established procedures leading from the development of SDMs tothe design of the repository and the associated safety assessment is shown to be
assured by an appropriately structured, logical methodology, and•Identified future efforts to further improve the confidence of the safety case bycreating a safety case template that can incorporate the results from step-by-step site
surveys and technological development in updates that are produced as required at
programme milestones.

This provides the knowledge base required in order to proceed with siting after theliterature survey (LS) commences in volunteer communities.
Based on geological information gathered on a nationwide scale, plus that from URLs, thedistribution of different rocks at relevant depths was assessed. By focusing on key attributes
required for a repository, three representative host rocks with different characteristics wereselected and SDMs developed.
Repository designs for HLW and TRU waste are developed for the three reference SDMs.
A comprehensive set of design requirements are introduced for individual components of therepository. By applying these requirements, detailed specifications are derived. This process
will meet the needs of an implementing organisation to consider a wide range of requirements
including practicality of construction and operational safety, to adapt and optimise designs
during the different steps of the siting process and to develop repository concepts to novel
siting environments.
Basic research and development (R&D) have led to improvements in technology related topracticality and quality assurance of fabrication of engineered barrier system (EBS)components and their emplacement underground. For the three siting environments assessed,
implementation can be based on existing technology. Nevertheless, the technology required toconstruct, operate and close such repositories will be continually re-assessed, based onprogress in other relevant civil engineering areas, with the aim of assurance that application togeological disposal remains state-of-the-art and can be tailored to the evolving understandingof requirements for specific sites.
The safety case developed in this report illustrates the fundamental feasibility ofdemonstrating safety for designs tailored to representative siting environments. It thus
highlights issues that need to be considered to determine if any volunteer site would besuitable and could lead to a specific safety case that meets regulatory standards, while
identifying open issues that need to be clarified by future R&D. The robust arguments
developed adopting conservatism to account for uncertainties, form a good basis for assumingthat operational safety can be assured for both normal and perturbed states. After closure,
uncertainties generally increase with time but there is reasonable confidence that these will
not preclude demonstration of safety for suitable sites. Nevertheless, the limitations that have

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been identified will need to be addressed in order to meet the requirements of a safety case tosupport selection of sites after the LS stage is complete.
The safety assessment methodology has evolved. Developing a comprehensive SDM as
demonstrated, using this for developing site adapted designs followed by assessment ofoperational and post-closure safety of these designs has been a key step forward and is key
progress in knowledge integration, forcing geoscientists, designers and safety assessors in the
programme to interact and eventually to provide feedback on development needs across thedifferent disciplines. This will allow the safety case to be a key vehicle for identifying further
development needs and eventually be the basis for site selection and a license to construct therepository. In the future, more realistic, site-specific safety assessment will provide a basis forcomparing repository design options and, possibly, comparing different potential host rocks at
a specific site or assessing pros and cons of different sites. This will allow consistentassessment from construction, operation and closure of the repository to the period after
closure, so that the results of the safety assessment can be appropriately fed back to theevolving design, taking into account advances in technology over this period.
In order to constantly improve safety cases, it will be necessary to clarify key open issues
and establish an R&D framework to resolve these. The safety case has highlighted many
important uncertainties and in the future these uncertainties will be reduced in a structured
manner as siting proceeds. In coming steps, it will be essential to introduce much more
realism in the abstraction of data and processes from the sites and in the design descriptions
and to quantify the implications of the uncertainties.
It is understood that the boundary conditions for the safety case will change with time andas the siting programme proceeds. Such changes need to be managed in structured way.
Technological breakthroughs may allow for novel or more optimal designs, whereas theimpact of new scientific findings could be both positive and potentially negative for the
programme. Development of the safety case after sites are identified will have to incorporate
flexibility to respond to site-specific environmental conditions and adapt to both futureadvances in science and technology and changes in social conditions over an extremely long
period. The goal is thus to assure that, for all programme milestones, decisions can besupported by a safety case that is fit for purpose, contributing to building confidence in thecredibility of the repository project.
In order to make geological disposal sustainable for its very long implementation period, itneeds to be assured that required human resources are available and that critical tacitknowledge is preserved and communicated across generations. To meet these challenges, it isessential that the programme retains and further develops a sufficient scientific andtechnological competence. Ensuring that there is a comprehensive understanding of thestructure and content of the NUMO safety case is essential for this. In meeting thesechallenges NUMO has also started to build a system for systematically managing knowledge
and human resources.
Safety cases also provide important background for stakeholders, facilitating theirunderstanding of the key issues associated with geological disposal and facilitating dialogue
to support their contribution to decision making. In order to perform this role, a safety casemust be communicated at technical levels appropriate to the different stakeholder groups and
NUMO has taken several actions in order to enhance communication of the safety caseoutside the group of technical professionals in the geological disposal field. NUMO willcontinue to address stakeholder concerns related to the safety case and reflect them in futureplanning.
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In summary, it is concluded that this report illustrates the preparations made to select asuitable site for the implementation of a repository based on the characteristics of areas whereliterature reviews are accepted, and to proceed with the project in stages whilst
simultaneously developing more reliable technology to achieve safe disposal.

8-4References[1] IAEA (2019): IAEA safety glossary: Terminology used in nuclear safety and radiation
protection, 2018 edition.

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