Technical Strategic Plan 2017
for Decommissioning of the Fukushima Daiichi
Nuclear Power Station ofTokyo Electric Power Company Holdings, Inc.

August 31, 2017

Nuclear Damage Compensation andDecommissioning Facilitation Corporation

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the Copyright Act for any purpose whatsoever without the prior written permission of Nuclear
Damage Compensation and Decommissioning Facilitation Corporation.

Table of Contents
1. Introduction ..............................................................................................................................1-1
2. Strategic Plan ..........................................................................................................................2-1
2.1 Purpose of the Strategic Plan ............................................................................................ 2-1
2.2 Basic concept of the Strategic Plan ................................................................................... 2-1
2.2.1 Fundamental policy ..................................................................................................... 2-1
2.2.2 Five guiding principles ................................................................................................. 2-1
2.3 Positioning of Strategic Plan 2017 and its overall structure ............................................... 2-6
3. Strategy for reducing risks posed by radioactive materials ......................................................3-1
3.1 Progress in Decommissioning of Fukushima Daiichi Nuclear Power Station ..................... 3-1
3.2 Basic thought in reducing risk of radioactive materials ...................................................... 3-4
3.2.1 Risk source identification ............................................................................................. 3-6
3.2.2 Risk estimation ............................................................................................................ 3-7
3.2.2.1 Hazard Potential ..................................................................................................3-8
3.2.2.2 Safety Management ..........................................................................................3-10
3.2.3 Risk evaluation .......................................................................................................... 3-11
3.2.4 Current status of risk reduction measures ................................................................. 3-13
3.2.4.1 Category I risk reduction measures ...................................................................3-13
3.2.4.2 Category II risk reduction measures ..................................................................3-15
3.2.4.3 Category III risk reduction measures .................................................................3-15
3.2.5 Challenges during risk reduction ............................................................................... 3-16
3.3 Implementing risk reduction strategies............................................................................. 3-18
4. Strategic Plan for fuel debris retrieval ......................................................................................4-1
4.1 Study plan for fuel debris retrieval (risk reduction) ............................................................. 4-1
4.1.1 Positioning of a strategic proposal towards determining a fuel debris retrieval policy . 4-3
4.1.2 Flow of studies on a strategic proposal towards determining a fuel debris retrieval policy ....4-5
4.2 Basic concept for ensuring safety in the fuel debris retrieval operation ............................. 4-7
4.3 Current states of each unit ............................................................................................... 4-10
4.3.1 Maintenance and management of a stable state at each unit ................................... 4-10
4.3.2 Understanding and estimating reactor conditions for each unit ................................. 4-12
4.3.3 Conditions of reactor buildings .................................................................................. 4-25
4.3.4 Understanding the condition in the PCV by actual investigation ............................... 4-27
4.4 Effects of risk reduction by fuel debris retrieval ................................................................ 4-28
4.4.1 Potential risk of fuel debris ........................................................................................ 4-28
4.4.2 Risk level of fuel debris ............................................................................................. 4-29
4.4.3 Discussions on reduction of risk of fuel debris .......................................................... 4-30
4.5 Feasibility of fuel debris retrieval method ......................................................................... 4-32

4.5.1 Characteristics of fuel debris retrieval methods ......................................................... 4-32
4.5.2 Technical requirements for the fuel debris retrieval methods .................................... 4-34
4.5.2.1 Ensuring containment functions.........................................................................4-34
4.5.2.2 Maintaining cooling functions.............................................................................4-47
4.5.2.3 Criticality control ................................................................................................4-49
4.5.2.4 Ensuring

the structural Integrity (aseismicity) of thePCV and reactor building 4-56

4.5.2.5 Reducing occupational radiation exposure during on-site works .......................4-62
4.5.2.6 Ensuring occupational safety .............................................................................4-71
4.5.2.7 Establishment of access routes to fuel debris....................................................4-75
4.5.2.8 Development of the fuel debris retrieval equipment and devices.......................4-86
4.5.2.9 Establishment of system equipment and working areas ....................................4-92
4.5.3

Technical requirements for storage of fuel debris in safe and stable condition ...... 4-98

4.5.3.1

Fuel debris handling (collecting, transferring and storing) ............................4-98

4.5.3.2

Handling of waste generated during fuel debris retrieval work ...................4-103

4.5.3.3

Safeguards .................................................................................................4-104

4.5.4

Evaluation of feasibility of the methods ................................................................ 4-106

4.5.4.1

Technical requirements for ensuring safety while retrieving fuel debris ......4-106

4.5.4.2

Technical requirements for method ............................................................. 4-111

4.5.4.3

Evaluation summary of method feasibility ................................................... 4-112

4.6 Comprehensive evaluation based on Five Guiding Principles ....................................... 4-114
4.6.1

Evaluation Method ............................................................................................... 4-114

4.6.2

Evaluation result................................................................................................... 4-115

4.7 Proposal for deciding fuel debris retrieval policies and efforts after decision (strategic
proposal)......................................................................................................................... 4-118
4.7.1

Proposal for deciding fuel debris retrieval policies ............................................... 4-118

4.7.2

Efforts after deciding fuel debris retrieval policies ................................................ 4-120

5. Strategic Plan for waste management .....................................................................................5-1
5.1 Study policy on the Strategic Plan for waste management ................................................ 5-1
5.2 International safety principles on radioactive waste management ..................................... 5-2
5.2.1 Basic principles for ensuring the safety of radioactive waste ...................................... 5-2
5.2.2 Principles for ensuring the safety concerning predisposal management of radioactive
waste ..................................................................................................................... 5-3
5.2.3 Principles for ensuring the safety of radioactive waste disposal .................................. 5-5
5.3 Current Status of Solid Waste Management ...................................................................... 5-6
5.3.1 Storage ........................................................................................................................ 5-6
5.3.2 Waste characterization .............................................................................................. 5-10
5.3.3 Processing and disposal ........................................................................................... 5-13

5.4 Proposal for compilation of the basic concept of solid waste management (strategic
proposal)...........................................................................................................................5-15
5.4.1 Characteristics of solid waste .................................................................................... 5-15
5.4.2 Solid waste management policies ............................................................................. 5-15
5.4.3 Pesent efforts and R&D based on solid waste management policies........................ 5-17
6. R&D initiatives .........................................................................................................................6-1
6.1 Basic policy for R&D .......................................................................................................... 6-1
6.1.1 Basic policy ................................................................................................................. 6-1
6.1.2 Entire perspective of R&D ........................................................................................... 6-1
6.2 Promotion of R&D aimed at utilization for the decommissioning process .......................... 6-4
6.2.1 R&D activities performed by TEPCO ........................................................................... 6-4
6.2.2 Project of decommissioning and contaminated water management............................ 6-4
6.3 Reinforcing cooperation in R&D ......................................................................................... 6-5
6.3.1 R&D priority agenda based on needs and its strategic promotion ............................... 6-6
6.3.2 Building R&D Infrastructure Based on Mid-and-Long–Term vision .............................. 6-7
6.3.3 Developing and securing human resource .................................................................. 6-8
7. Enhancement of international cooperation ..............................................................................7-1
7.1

Integrating and utilizing wisdom .................................................................................... 7-1

7.2

Active information dissemination to international communities ..................................... 7-6

7.3

Close cooperation between relevant organizations ....................................................... 7-6

8. Future actions ..........................................................................................................................8-1
8.1 Enhancing project management capabilities...................................................................... 8-1
8.2 Engagement with stakeholders .......................................................................................... 8-2
8.3 Consideration for project sustainability .............................................................................. 8-3

＜Appendix＞
Appendix 1: Japanese Government's Organizational Framework for Decommissioning the
Fukushima Daiichi NPS ....................................................................................... A-1
Appendix 2: Information on the Strategic Plan ........................................................................ A-2
Appendix 3: Overview of SED indicator .................................................................................. A-5
Appendix 4.1: Periodic measurement of plant data ................................................................ A-8
Appendix 4.2: Results of PCV internal survey ...................................................................... A-16
Appendix 4.3: Survey results of S/C and torus room ............................................................ A-23
Appendix 4.4: Results of muon detection.............................................................................. A-25
Appendix 4.5: BSAF project outline and results so far .......................................................... A-28
Appendix 4.6: Characteristics of MAAP code and SAMPSON code and analysis results..... A-30
Appendix 4.7: Summary of heat balance method and estimation result ............................... A-32
Appendix 4.8: Location of fuel debris estimated from plant parameter trend ........................ A-36
Appendix 4.9: List of properties of fuel debris ....................................................................... A-40
Appendix 4.10: Status of the comprehensive analysis and evaluation of the conditions inside
the reactor in each unit ...................................................................................... A-43
Appendix 4.11: Studies on containment function .................................................................. A-45
Appendix 4.12: Study of water level at the bottom of PCV when retrieving fuel debris ......... A-49
Appendix 4.13: Preliminary Seismic Margin Evaluation of RPV and PCV. ............................ A-52
Appendix 4.14: Development of fuel debris retrieval equipment and devices ....................... A-60
Appendix 4.15: Risks accompanied by fuel debris retrieval tasks......................................... A-71
Appendix 5: Status of Solid Wastes Management, and Their Storage Management Plan .... A-74
Appendix 6.1: Interim Report by Research Partnership Task Force...................................... A-76
Appendix 6.2: Changes in elective subjects of secondary examination for professional
engineers (category of atomic energy and radiation) ........................................ A-81
Appendix 7: Conditions at TMI-2 and Chernobyl Unit 4 ........................................................ A-82

Abbreviations and short forms
Abbreviations andshort forms
AC
CAMS
CEA
CRD
CS
CST
D/W
DHC
DOE
DSP
FDW
FP
GPUN
IAEA
ICRP
IRID
JAEA
MCCI
NDA
NDF
NDF Act
Revised NDF Act
NRC
NSC
OECD/NEA
PCV
RPV
S/C
SED
TMI-2
Implementation plan
Underwater ROVStrategic plan
Matters to beaddressed
Mid- and Long-term
Roadmap
TEPCO
Fukushima Daiichi
NPS

Definitions and Officinal Names
Atmospheric control system
Containment Atmospheric Monitoring System
Commissariat à l’énergie atomique et aux énergies alternatives (France)
Control rod drive mechanism
Core spray system
Condensate storage tank
Drywell
Drywell humidity control system
United States Department of Energy
Dryer separator pool
Feed water system
Fission products
General Public Utilities Nuclear Corporation (owner of TMI-2)
International Atomic Energy Agency
International Commission on Radiological Protection
International Research Institute for Nuclear Decommissioning
Japan Atomic Energy Agency
Molten core–concrete interaction
Nuclear Decommissioning Authority (United Kingdom)
Nuclear Damage Compensation and Decommissioning Facilitation
Corporation
Nuclear Damage Compensation Facilitation Corporation Act
A law that partially revises the Nuclear Damage Compensation
Facilitation Corporation Act
Nuclear Regulatory Commission (United States)
New Safe Confinement, arched-shaped, built to cover the remains of
the No. 4 unit at the Chernobyl Nuclear Power Plant
Nuclear Energy Agency, Organization for Economic Co-operation andDevelopment
Primary containment vessel
Reactor pressure vessel
Suppression chamber
Safety and Environmental Detriment, the method developed by the
NDA to express risk levels
Three Mile Island Nuclear Power Plant Unit 2, United States
Implementation Plan of the Measures for the Specified Reactor Facilities
at Fukushima Daiichi Nuclear Power Station
Underwater remotely operated vehicle
Technical Strategic Plan 2017 for the Decommissioning of the Fukushima
Daiichi Nuclear Power Station of Tokyo Electric Power Company
Holdings, Inc.
Matters which are requested to be taken by Tokyo Electric Power
Company, Inc. when the Fukushima Daiichi NPS was designated asSpecified Nuclear Facility
Mid- and Long-term Roadmap towards the Decommissioning of TEPCO’s
Fukushima Daiichi Nuclear Power Station Units 1–4
Tokyo Electric Power Company Holdings, Inc.
Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power
Company Holdings, Inc.

Nomenclature
Term
CRD housing
MAAP

SAMPSON
Reactor well shieldplug
Submersion method
Partial submersion
methodClearance

Grating
Actual debris
Heavy nuclide
Defense in depth
SludgeSlurryFuel debris
Pedestal
Muon-based fueldebris detection
technology
Simulated debris
Mock-up
Preparatory
engineering
Loose debris
Robustness

Description
Housing that stores the control rod drive mechanism
The severe accident simulation code improved and controlled by the U.S.
Electric Power Research Institute (short for Modular Accident Analysis
Program)
The simulation code improved and controlled by the Institute of Applied
Energy, consisting of the models and modules formulated based on themulti-dimensional equations and theoretical formulae that precisely
express physical phenomena (short for Severe Accident analysis codewith Mechanistic, Parallelized Simulations Oriented toward Nuclear Fields)
A concrete lid on top of PCV installed for radiation shielding (a part of thefloor of the uppermost floor of the R/B during operation)
A retrieval method by filling to the top of the PCV with water to submerge
the fuel debris
The debris retrieval method that does not fill the PCV fully with water. Fueldebris is partially exposed to the air when removed.
Clearance system refers to a system under which the government
confirms that the concentration of radioisotopes of materials used in a
nuclear facility is below the "Clearance Levels" (the level at which theimpact is negligible to human health).
The materials confirmed by the government are removed from theregulations on nuclear reactors, and will be subject to regulations underlaws on wastes and recycle as conventional industrial wastes or valuables
Steel grating footings, used as side-ditch coverings and scaffolds.
Actual fuel debris retrieved from the reactor vessels as opposed tosimulated debris
Actinide nuclides such as uranium and plutonium
To protect humans from safety threats, several barriers (levels ofprotection) are provided, each having a specific aim. If protection has
failed on a given level, the barrier of the next level provides protection.
This system is a preparation for uncertainty, and the concept of thissystem is essential to ensure nuclear safety.
Muddy material including radioactive material
Liquid-state material in which minerals and sludge are mixed
Nuclear fuels molten and mixed with parts of reactor internals due to lossof reactor coolant and resulted in a re-solidified state
The cylindrical base that supports the reactor body. Its inside is filled withconcrete.
Technology that determines the positions and forms of fuel, using thecharacteristics of muonic atoms (muons) that arrive from the cosmos andatmospheric air. The number of muons and their trajectories change
depending on the density of the material through which they pass through.
Artificial objects manufactured by estimating the chemical composition andforms based on the examples of TMI-2 accident
Model created to be almost the same as the real object
Engineering work conducted to preliminarily determine the feasibility of theplanned construction work prior to the basic design conducted at the
beginning of ordinary construction
Small-size debris found in the TMI-2 molten pool, including fractured
pellets, control rods, and resolidified fuel debris
The capability to maintain the robust function even when the condition ischanged to a certain extent from what is expected

1. Introduction
The overall approach to the decommissioning of the Fukushima Daiichi Nuclear Power Station ofTokyo Electric Power Company Holdings, Inc. (Fukushima Daiichi NPS) started based on the “Midand Long-Term Roadmap towards the Decommissioning of TEPCO’s Fukushima Daiichi Nuclear
Power Station Units 1–4” (Mid- and Long-Term Roadmap), released by the Japanese Government
in December 2011. The handling of urgent issues, such as contaminated-water management andfuel removal from spent-fuel pools, has been placed on top priority; however, to fully complete
reactor decommissioning, long-term measures are required such as fuel-debris retrieval, and it isessential to prepare a mid- and long-term decommissioning strategy.
On August 18, 2014, the former Nuclear Damage Compensation Facilitation Corporation was
reorganized into the Nuclear Damage Compensation and Decommissioning Facilitation
Corporation (NDF) as an organization responsible for technical studies to proceed with reactor
decommissioning properly and steadily from a mid- and long-term perspective. Based on the
Nuclear Damage Compensation Facilitation Corporation Act (NDF Act), the NDF has beendeveloping the Technical Strategic Plan for the Decommissioning of the Fukushima Daiichi Nuclear
Power Station of Tokyo Electric Power Company (Strategic Plan) annually since 2015, as part ofits statutory obligations: the provision of advice, guidance and recommendations for ensuring anappropriate and steady performance of the decommissioning, and the research and developmentof decommissioning technologies. For this year in particular, the Mid- and Long-Term Roadmap(revised June 2015) requires that decisions be made on fuel debris retrieval policies for each unit,
and that the basic concept of solid waste management be compiled. Based on discussions with
relevant parties, NDF will be presenting strategic proposals regarding these two matters in thestrategic plan.
Six years has passed since the Fukushima Daiichi NPS accident, and short-term planning has
become possible to a certain extent: on the accident site, contaminated-water management madeprogress thanks primarily to the land-side impermeable wall, fuel removal from spent-fuel poolsalso moved ahead, and the outdoor work environment has improved. For the mid- and long-term,
although fuel-debris retrieval, reactor interior investigation and R&D have made progress, it is
becoming known that the work environment inside the buildings is severe, suggesting the difficultyof the work there in the future.
While the decommissioning project is shifting to the phase where mid- and long-term issues should
be addressed specifically, the partially revised Nuclear Damage Compensation Facilitation
Corporation Act (Revised NDF Act) which was established in May 2017 to implement measures
such as obligating TEPCO, the operator, to reserve funds at NDF that are required fordecommissioning in order to execute decommissioning in a more reliable manner. With this Act ineffect, the duties of NDF now include operations for the management of these decommissioningfund reserves: (1) TEPCO will reserve an amount of funds in each fiscal year as determined by the
NDF and approved by the competent minister; and (2) TEPCO will proceed with decommissioning
while withdrawing the reserved funds based on the withdrawal plan prepared jointly by the NDF
and TEPCO and approved by the competent minister (“withdrawal plan”). Additionally, as stated in
the Revised Comprehensive Special Business Plan (The Third Plan) published in the same month,
the decommissioning of the Fukushima Daiichi NPS is a fundamental precondition of the recoveryof Fukushima, and is expected to be carried out correctly and steadily.
Under this new system, the NDF, the major supervisor and administrator of the decommissioning
by TEPCO, is expected to play a greater role and assume greater responsibilities than before,
primarily in terms of project management: It will assume such additional duties as (1) the
appropriate management of funds for decommissioning, (2) the supervision of an appropriate
system for executing the decommissioning, and (3) steady work management based on the

1-1decommissioning reserve fund system. Specifically, through the process of preparing thewithdrawal schedule jointly with TEPCO, NDF will provide support for enabling the correct andsteady execution of decommissioning, such as conducting assessments of the validity of actions
taken by TEPCO from the perspective of project execution and proposing tasks, among othermatters, that should be included in the withdrawal schedule.
Under these circumstances, the roles shared among the organizations associated with the
Fukushima Daiichi NPS decommissioning, which are the Japanese government, NDF, TEPCO, theInternational Research Institute for Nuclear Decommissioning (IRID) that specializes in researchand development, and Japan Atomic Energy Agency (JAEA), are shown in Figure 1-1.
The government is creating a Mid- and Long-Term Roadmap that defines the directions fordecommissioning and contaminated water countermeasures, and conducts progress management
for the various countermeasures based on these. Additionally, the government also provides
support for R&D areas that pose high levels of technical difficulties and require government
leadership.
Based on critical issues and other matters indicated in the Mid- and Long-Term Roadmap, NDF will
prepare a strategic plan for aiding in the reliable execution of the roadmap and in reviews
considering revisions to the roadmap. With regards to TEPCO, NDF will provide advice andinstructions from a technical standpoint to ensure the steady advancement of the decommissioning
process. Moving forward, it will manage and monitor the execution of decommissioning based on
the decommissioning reserve fund program. Additionally, it is engaged in planning and progress
management for R&D by working closely with R&D organizations and sharing with them the statusof progress and issues in order to proceed with R&D smoothly.
TEPCO bears the responsibilities of the operator in the decommissioning of the Fukushima Daiichi
NPS. As such, it is conducting engineering operations including designs and work planning, has
submitted the “Implementation Plan of the Measures for the Specified Reactor Facilities at
Fukushima Daiichi Nuclear Power Station” (Implementation plan) to the Nuclear Regulation
Authority, and is proceeding with approved decommissioning actions. Specifically, it will gear up
actions regarding fuel debris retrieval in addition to implementing actions including the retrieval offuel from spent fuel pools and countermeasures for contaminated water.
R&D organizations are engaged in R&D efforts based on the Mid- and Long-Term Roadmap,
bringing together knowledge from around the world to efficiently and effectively carry out R&D thatwill be needed for decommissioning.
For the system of the government, refer to Appendix 1.

1-2

Figure 1-1 Roles Shared by the Organizations Associated with the Fukushima Daiichi NPS
Decommissioning

1-3

2. Strategic Plan
2.1 Purpose of the Strategic Plan
The aim of the Strategic Plan is to contribute to the correct and steady implementation of the Midand Long-Term Roadmap, and to evaluations for revisions to the Roadmap, as well as to provide
reliable technological grounds to the Roadmap in order to execute the decommissioning of the
Fukushima Daiichi NPS correctly and steadily.
The Strategic Plan is prepared based on discussions at Expert Committees, whose aim is toacquire opinions regarding specific issues from experts having specialized knowledge and from therepresentatives of organizations concerned, and discussions at the Decommissioning Strategy
Board, which is a function of review by experts of various fields. In addition, international experts
are invited to the Decommissioning Strategy Board as International Special Advisors, and various
technical gatherings are held, to acquire experience and knowledge concerning decommissioning.

2.2 Basic concept of the Strategic Plan
2.2.1 Fundamental policy
The Fukushima Daiichi NPS, which has been designated a Specified Nuclear Power Facility, iscurrently maintained and administered to a given level of stability, and safety measures are being
taken as specified by the Nuclear Regulation Authority (NRA), according to “the matters for whichmeasures should be taken” (“Matters to be Addressed”).
However, unlike normal nuclear power plants, the status of the station is not sufficiently grasped,because of the damages of the buildings, fuel debris, spent fuels, generation of contaminated water
including radioactive materials, and various radioactive wastes. To proceed with decommissioning
in the future, the possibility is undeniable that the risks attributable to radioactive materials become
apparent. If no action is conducted and the current conditions are left as they are, the conditions
where the risks of radioactive materials exist will continue. The decay of radioactivity will gradually
decrease the risks, while the risks may increase in reverse due to the aging deterioration of facilities
and other factors. The risks may not always simply decrease with time.
Accordingly, the fundamental policy for decommissioning of the Fukushima Daiichi NPS is "tocontinuously and promptly reduce the risks associated with the radioactive materials generated by
the accidents".
2.2.2 Five guiding principles(1)Basic attitude towards completing the decommissioning
The decommissioning of Fukushima Daiichi NPS is a project in which considerable uncertainty
inheres. As a result of the accident, the reactor interior and surroundings are in a radioactive
environment and workers are unable to have access there easily. Accordingly, the properties ofradioactive materials and the damages of on-site equipment and structures left there remain
unknown, posing uncertainty.
It is desirable if all the information that is currently difficult to confirm could be collected toeliminate such uncertainty before starting decommissioning; however, many resources,
especially a considerably long time, are required to do so.

2-1

To realize prompt decommissioning, it is necessary to assume the attitude of taking flexible andprompt approach, based on the directions determined with previously obtained experience andknowledge and with experiment- and analysis-based simulation, placing safety at the top priority,even though a certain extent of uncertainty remains.
To proceed with decommissioning flexibly and promptly, the attitude of optimizing the entire
project from the long-term, comprehensive viewpoint, and the attitude of making preparations forunexpected cases, are also important.
(2)Five guiding principles
Based on the basic attitudes, this section presents the Five Guiding Principles of risk reduction
to promote the fundamental policy of the Fukushima Daiichi NPS decommissioning.
Principle 1: Safe - Reduction of the risks posed by radioactive materials to ensure labor safety
Principle 2: Proven - Highly reliable and flexible technologies
Principle 3: Efficient - Effective utilization of resources (e.g., human, physical, financial andspace)
Principle 4: Timely - Awareness of time axis
Principle 5: Field-oriented - Thorough application of the “three actuals” (the actual field, the actual
things and the actual situation)
A. Principle 1: Safe - Reduction of the risks*1 posed by radioactive materials to ensure labor
safety*1) Environmental impacts and workers’ exposure

In proceeding with the decommissioning of the Fukushima Daiich NPS, it is needless to saythat safety is the first priority. The International Atomic Energy Agency (IAEA), like otherorganizations, specifies the safety principle as protecting humans and environments against
risks from radioactive materials.
Since safety standards for operational NPSs cannot be applied to damaged reactors,
decommissioning should proceed by ensuring safety in accordance with site conditions.
Accordingly, the viewpoint of giving priority to the risks that need to be reduced is necessary,that is, risks must be reduced expeditiously to achieve a safe and stable condition, based on
the awareness that damaged reactors pose high risks. The attitude of proceeding with thedecommissioning while ensuring safety effectively is important, with an awareness of total
risk*2 reduction with time. With regard to works such as fuel debris retrieval, for which we haveno experience, it will be important to specifically develop concept for ensuring safety fordamaged-reactor decommissioning and to start discussion with the NRA in an early stage, withreference to the precedent case at Three Mile Island Nuclear Power Plant Unit 2, United States(“TMI-2”), where the operator, regulatory authorities and other parties concerned committed
themselves as a single team to deal with the accident.
To reduce environmental impact, such as public exposure to radiation and environmental
contamination risks, it is critical to improve the conditions of controlling the radioactive
substances that exist in various forms in the premises. And while steps have been taken to
improve working environments to reduce worker exposure to radiation, safety measures,
including work-hour management, shield placement, and wearing of protective equipment,

2-2

must be taken comprehensively with regards to work that is performed in highly radioactive
environments, such as in reactor buildings. Furthermore, sufficient attention should be paid tolabor safety as well, to prevent accidents and injuries, because personnel will need to work in
the places that are narrow and not easily accessible.
*2) Total risk in consideration of the continuation and change of existing risks over time as well as the trade-off between
the temporary risk changes due to work performed to reduce risks and resulting risk reduction

B. Principle 2: Proven - Highly reliable and flexible technologies
The decommissioning of the Fukushima Daiichi NPS is an unprecedented project, which istechnically highly difficult and involves many newly developed technologies, because it needs
to be carried out under highly radioactive conditions.
For the measures to be accomplished in a relatively short period of time, new technology
developments should be limited to as few as possible, to minimize the risk of failingdevelopment and make progress steadily.
In this regard, feasible Japanese and international technologies, namely, technologies andintelligences with high technology readiness levels (TRL), should be adopted and applied
where possible. They should be provided with necessary improvements, such assystematization, to suit the site conditions of the Fukushima Daiichi NPS, and their operationneeds to be verified and demonstrated in advance to confirm workability under the severe site
conditions.
Considering a high degree of uncertainty in the site conditions, robust technologies should beselected to enable flexible response to unexpected situations or changes in the situation. Onsite work should be carried out step by step and the course of actions should be adjusted
where necessary. In case that selected technologies do not work, it is important to prepare
counterplans in advance, such as alternative measures.
On the other hand, as decommissioning work proceeds, the development of an entirely newtechnology may be critical in some cases. For the mid- and long-term issues that require sucha technical development, research and development, including fundamentals and baseline
studies, should be conducted with matters concerned specified, including needs, objectives,
the roles of organizations concerned (e.g., universities, public research institutions and privateorganizations). Especially the contribution of remote-control technology is highly expected, foruse under the environments where radioactivity is unlikely to turn for the better because ofdifficulty in decontamination.

C. Principle 3: Efficient - Effective utilization of resources (e.g., human, physical, financial andspace)The decommissioning of the Fukushima Daiichi NPS involves a large amount of complicatedtasks and developments over a long period of time. Therefore, resources, including human,
physical, financial and spatial ones, may function as a restraint. Their reasonable and effective
use will be a key factor for successful decommissioning.
As for human resources, who will need to work in a highly radioactive environment, it isnecessary to plan and manage the total exposure doses of every construction worker during
the construction in order to ensure sufficient personnel supply over a long period of time. In

2-3addition, because the project involves many technical studies concerning R&D and on-site
construction, impracticable and unnecessary work should be avoided, aiming to streamlined
work. It is also important to secure the human resources necessary for successful
decommissioning such as researchers, engineers and workers, and to be committed tohuman-resource development and technology inheritance.
As for physical resources, any facilities and goods brought into the Fukushima Daiichi NPS
are highly likely to be treated as radioactive wastes. It is reasonable to use facilities and goods
efficiently to minimize the generation of wastes, by avoiding bringing in unnecessary goods,
by making the best use of the goods that have been brought in, and by keeping the “3R rule”(reduce, reuse and recycle) in mind.
As for financial resources, since a large amount of tasks and developments are involved overa long period of time, keeping overall cost reduction in mind is important, in addition toevaluating the cost–benefit performance of individual tasks, which is associated with theeffective use of human resources, and the investment–benefit performance of technologicaldevelopments and devices.
As for spatial resources, although the premises of the Fukushima Daiichi NPS are relatively
large compared to those of other nuclear power stations in Japan, it is important that we take
into account the possibility that additional facilities may encroach on working spaces, and makeeffective and efficient use of space including the development and securing of transport routes
for equipment and other items.
To efficiently use the resources (human, physical, financial and spatial), it is important toevaluate individual tasks and developments, but it is also important to prioritize the measuresfrom the viewpoint of overall optimization from the long-term perspective, in consideration of
impact on subsequent tasks.

D. Principle 4: Timely - Awareness of time axis
Because spending an unnecessarily long period of time on the Fukushima Daiichi NPS
decommissioning entails a continuation of risk conditions caused by radioactive materials, it isimportant to be aware of promptly reducing the risks. “Timeliness” and the Principle of “Proven”
may be in a trade-off relationship, but it would be preposterous if decisions are postponed andthe risk situations are left as they are. It is necessary to proceed with the work carefully inconsideration of the risks, while making optimal decisions whenever appropriate.
In order to be aware of “timeliness,” it is important to set a certain target time for each of the“actions to be taken as soon as practicable,” “actions that require steady implementation” and“actions to be carried out over a long time.”
In addition, multilayered preventive measures against project risks are also important to avoid
time losses and rework. When preventive measures are planned, it is important as well todecide on what risks to address, to what degree to take preventive measures, and to what
degree to multilayer the measures. Clarifying the details and levels of safety assessment inadvance is also important to prevent time losses and rework.
Meanwhile, issues such as waste management and station site closure involve unprecedented
situations that are created by a damaged nuclear power station and waste materials produced
from the accident. As such, new regulatory systems and standards may be required for these2-4

issues. Dealing with these issues may take a correspondingly long period of time, and should
be examined with lead time in mind.

E. Principle 5: Field-oriented - Thorough application of the “three actuals” (the actual field, theactual things and the actual situation)
The decommissioning of the Fukushima Daiichi NPS is risk reduction activities associated withradioactive materials on site; therefore, it is important to carry out the tasks in accordance withthe “Three Actuals” concept thoroughly and in a site-oriented manner.
The Three Actuals means to understand the precise needs based on the actual site conditions,actual facilities, and what is actually happening at the site, and to choose technologies focusing
on on-site applicability. Especially there are risks that the understanding of a technology isdifferent between those who developed it and those who actually apply the achievement onsite, or that the understanding by the design sector or management sector and that by on-site
workers have gaps. Special attention needs to be paid to these risks and both sides should be
encouraged to have common understanding.
The “on-site applicability” is to assess whether a technology under feasibility study (FS) is
actually applicable in the site conditions and environment of the Fukushima Daiichi NPS.
The assessment of on-site applicability should be judged based on the following factors:








Environmental resistance (e.g., radiation, temperature, humidity and lightintensity)
Accessibility and transportability (e.g., narrow routes, obstacles such as rubble,
lifting devices and dose rates)
Work space (e.g., inside the buildings and yards)
Infrastructures (e.g., electricity, air, communication and water)
Processing and disposal loads from generated waste materials
Maintainability and failure recoverability
On-site operability

Understanding site conditions may provide knowledge that helps enhance the safety of existing
light-water reactors. Keeping such a perspective in mind is preferable, even though it is outside
the scope of the Fukushima Daiichi NPS decommissioning.
No matter whether for the Three Actuals or for the safety enhancement of light-water reactors,
understanding the on-site conditions involves great difficulties and radioactive exposure under
the severe on-site environment of the Fukushima Daiichi NPS. There is a trade-off that iswhether spending a long time to investigate thoroughly is acceptable from the viewpoint of
total risk reduction; therefore, it may be necessary to map out plans based on presumptions toa certain extent. In such case, multilayered measures should be prepared to deal with
unexpected situations.
While individual work areas should be examined according to these Five Guiding Principles, it
is extremely important to be constantly aware of the interrelationship among all these work
areas and their positions within the entire project, from the viewpoint of total optimization.

2-5

2.3 Positioning of Strategic Plan 2017 and its overall structure(1)Positioning of Strategic Plan 2017
The Mid- and Long-Term Roadmap (revised in June 2015) states, as immediate milestones tofuel-debris retrieval, that “fuel debris retrieval policies for each unit” should be determined aroundsummer 2017, and that the basic concept of solid waste management” should be compiled in FY
2017. The Roadmap also states that the feasibility of the fuel-debris retrieval methods should be
evaluated as part of the Strategic Plan and that fuel-debris retrieval policy should be determinedbased on the evaluation results.
Over the past year, taking into account the current conditions in the primary containment vessels(PCV) that were learned through inspections, NDF has been conducting reviews of technicalissues of the three focus construction methods specified in Strategic Plan 2016, including
feasibility evaluations based on knowledge acquired domestically and internationally. NDF will
be preparing a strategic proposal that will aid in determining the fuel debris retrieval policies. As
for fuel debris retrieval, the cabinet decision of Dec. 20. 2016 "Basic guidelines for accelerating
the recovery of Fukushima from the nuclear disaster," also prescribes that NDF should play acentral role in gathering knowledge domestically and from abroad, accelerate technical reviews
relating to effective policies and processes, conduct feasibility analyses of methods for fuel debris
retrieval, present strategic proposals, and ensure that R&D efforts that will be required moving
forward can be carried out expeditiously.
With regard to waste management, we will summarize international guiding principles for
ensuring safety in implementing countermeasures for radioactive waste and, based on thecurrent situation in waste countermeasure approaches, make strategic proposals for compilationof the basic concept of solid waste management.
The two strategic proposals in the present Strategic Plan 2017 are the main points summarizedbased on the basic attitude toward decommissioning involving considerable uncertainty, as
described in Subsection 2.2.2 (1).
(2)Overall structure of Strategic Plan 2017
Strategic Plan 2017 consists of eight chapters.
It was stated in Chapter 1 that the duties of the NDF newly include operations for the
management of these decommissioning-fund reserves in addition to its existing responsibilitiesof strategy formulation and technical support. As such, its scope of responsibilities is expected
to expand.
Chapter 2 stated that, as the purpose and positioning of Strategic Plan 2017, this Strategic Plan
presents strategic proposals for determination of fuel debris retrieval policies for each unit andestablishment of the basic concept of solid waste management, among the milestones stated in
the Mid- and Long-Term Roadmap. Chapter 2 also presented five principles to achieve the basic
policy for decommissioning of the Fukushima Daiichi NPS is to reduce continually and quickly
the risks associated with the radioactive materials that resulted from the accident and do notexist in normal nuclear power plants.
Chapter 3 mentions strategies for reducing the risks attributable to radioactive materials. It
specifies the various radioactive materials that exist in Fukushima Daiichi NPS as risk sources,
profiles their characteristics, analyzes and evaluates them, determines priorities, and describes
the measures and challenges for risk reduction. It also discusses the necessity of establishing

2-6

the basic safety policy to safely and promptly proceed with risk reduction strategies and suggests
how to establish it.
In the area of fuel debris retrieval, Chapter 4 firstly summarizes the basic concept for ensuringsafety in the retrieval of fuel debris, and secondly summarizes the most up to date status of plants
on a per unit basis including estimates of fuel debris distribution and other parameters for the
purpose of formulating strategic proposal for determination of the fuel debris retrieval policies.
Based on these discussions, the chapter examines the risks of fuel debris and reduction thereof.
Concerning technical requirements for realizing fuel-debris retrieval methods, the chapter
introduces the updates of technological development, along with challenges, thus evaluating thefeasibility of the methods. Based on these findings, the chapter comprehensively evaluates themethods in reference to the Five Principles, and further discusses strategic proposals fordetermination of fuel-debris retrieval policies and for future efforts based on the evaluation.
Chapter 5, which discusses waste handling, introduces the internationally agreed basic policies
intended to ensure safety against general radioactive wastes, and referring to them, summarizes
the points of attention concerning the management of solid wastes resulting from the Fukushima
Daiichi NPS accident. The chapter also evaluates the current conditions of efforts concerningsolid-waste management based on the Mid- and Long-Term Roadmap, and identifies thechallenges that may influence future actions. Based on these discussions and evaluations, thechapter presents the proposals that will contribute to compilation of the basic concept of solid
waste management.
Chapter 6 discusses research and development efforts. When fuel-debris retrieval policy isdetermined, research and development will move forward to a new stage, and specificdecommissioning processes will be identified. The chapter accordingly indicates the necessity offlexibly reviewing the research and development efforts according to the roles shared between
the government and operator. The chapter also discusses, from the mid- and long-term point ofview, the establishment of a research base and the construction of research infrastructure, as
well as the importance of fundamental research and development.
The topic of Chapter 7 is strengthening international cooperation. It discusses the collection of
wisdom concerning decommissioning and related topics from other countries, proactive
information transmission to international communities for more positive communication, andclose cooperation with domestic organizations concerned for effective international
communication.
Chapter 8 discusses how to proceed with the decommissioning project in the future. It mainly
discusses strengthening project management, responding to project risks, social relations, all ofwhich are important to carry out the decommissioning project steadily.
The following are the site description (Fig. 2-1), interior structural drawing of the reactor building,and interior structural drawing of the reactor pressure vessel (Figs. 2-2, 2-3) at the Fukushima
Daiichi NPS.

2-7

a.

b.

Overall View of the Premises

Fukushima Daiichi NPS Premises

Fig. 2-1 Fukushima Daiichi NPS site description

2-8(Provided by TEPCO)

Figure 2-2 Structures/Components inside Reactor Building(Provided by IRID)

Figure 2-3 Structures/Components inside Reactor Pressure Vessel (Provided by IRID)

2-9

3. Strategy for reducing risks posed by radioactive materials
This chapter discusses strategy for reducing risks posed by radioactive materials with the goal of
fulfilling the fundamental policy for the decommissioning of the Fukushima Daiichi Nuclear Power
Station, shown in Section 2.2.1. For this purpose, the major risk sources will be analyzed, evaluated,
and then, after prioritizing them, actions to be taken to reduce these risks and associatedchallenges will be discussed.
The chapter further describes the fundamental safety principles, which are being examined inaccordance with the safety policy set by international organizations and the matters to beaddressed set by the Nuclear Regulation Authority. Establishing the fundamental safety principles
and sharing them with those concerned is beneficial in implementing the risk reduction strategy
safely and promptly.

3.1 Progress in Decommissioning of Fukushima Daiichi Nuclear Power Station
Over the last year, risk reduction activities in the decommissioning of the Fukushima Daiichi Nuclear
Power Station (NPS) made progress as described below in the areas of concrete medium-andlong-term measures defined in the Mid- and Long-Term Roadmap.
(1)Contaminated water management
Contaminated water, which is generated when groundwater flowing into the buildings mixes withthe water cooling the fuel debris, is managed in accordance with the three principles (removing
the contamination source, isolating groundwater from the contamination source, and preventing
leakage of contaminated water) (Figure 3-1).
In removing the contamination source, they are treated with the advanced multi-nuclide removal
system.
In isolating groundwater from the contamination source, the freezing operation of the land-side
impermeable wall began in March 2016 on the sea side and part of the mountain side, and inJune 2016, a process to freeze 95% of the mountain side of the wall began. In October 2016,
the freezing operation was completed on the sea side, with the freezing of four out of theremaining five unfrozen sections underway since March 3, 2017. As of the end of March 2017,
the freezing process has progressed to the point that only one section remains unfrozen on themountain side (Fig.3-2).
After the sea-side area of the wall had been completely frozen, the water volume pumped at thearea 4 m above the sea level was reduced to about a third (The volume averaged about 118
m3/day in March 2017.) In addition, along with the treatment of the water stagnant in the buildings,
sub-drain water pumping was performed to reduce the groundwater level around the buildings.
As a result, groundwater inflows into the buildings reduced to around 120 m3/day on average inMarch 2017.
In preventing leakage of contaminated water, the concentration of the radioactive materials inthe surrounding sea areas is constantly low.
In regards to the stagnant water in the buildings, there was progress in reducing the water levelin the Unit 1 turbine building until March 24, 2017, when it was confirmed that the water level
reduced to the point that the surface of the bottom floor was clear of water.
In addition, a plan has been presented for completion of the treatment of the stagnant water inthe buildings by 2020.

3-1

(2)Removal of spent fuel from the spent fuel pool
In Unit 1, the removal of the cover wall panels for the Reactor Building was completed. An
investigation on the operating floor was started and information gathering on the rubble and thefuel handling machine that is required to develop a rubble removal plan is in progress.
In Unit 2, for fuel removal, a yard around the Reactor Building was prepared and a working
platform for access to the operating floor was constructed.
In Unit 3, measures to reduce the dose rates on the operating floor (decontamination andshielding) were completed. Installation of a cover for fuel removal was started in January 2017.
(3)Fuel debris retrieval
In Unit 1, in a robotic investigation of the conditions in PCV from March 18 to 22, 2017, adosimeter and a camera were hung down from the first floor of PCV to investigate the current
conditions in the basement outside the pedestal and in the vicinity of the access opening of thepedestal.
In Unit 2, a PCV internal survey was conducted by using a robot from January 26 to February
16, 2017. After sharpening the images captured inside the pedestal, the survey team confirmed
the detachment, deformation and other conditions of the gratings, which serves as the scaffolding
for replacement of control rods.
In Unit 3, a PCV internal survey using a remotely operated underwater survey device (hereinafterreferred to as underwater ROV) was conducted from July 19 to 22, 2017, which detected whatlooked like some molten substances solidified in the pedestal, as well as several fallen/deposited
objects, including gratings etc. Additionally, in the latest evaluation based on a core survey witha muon measurement system underway since May 2017, it was found that, while part of the fueldebris might still remain in the reactor pressure vessel (hereinafter “RPV”), no large-sized highdensity material was confirmed.
(4)Waste management
The waste reduction measures are continued to be in place. An operation to reduce waste
protective clothing with an incinerator was started. The projected solid waste generation wasrevised and the solid waste storage and management plan was updated1. Sampling and analysis
is underway in order to characterize the solid waste and construction of facility for analysis isalso underway.
(5)Other specific measures
Since the rubble removal and facing in the area 4m above sea level intended for improving site
environment have resulted in the reduced risks of contamination, the classification for the protective
equipment has been changed to “ordinary clothing area” where the workers are allowed to enter
and work in general workwear or in on-site safety workwear with disposal dust-proof mask.

1

Tokyo Electric Power Company Holdings Inc., “The Solid Waste Storage Management Plan for
TEPCO Fukushima Daiichi Nuclear Power Station,” June 2017 Version, June 29, 2017.

3-2

Figure 3-1. Contaminated water management at Fukushima Daiichi NPS (Source: TEPCO)

Figure 3-2. Partial closure of the land-side impermeable wall (mountain side)(Source: TEPCO)

3-3

3.2 Basic thought in reducing risk of radioactive materials
This section describes the principle of reducing the risk caused by the radioactive material in the
Fukushima Daiichi NPS, and the measures to reduce the risk of radioactive material in reference
to the risk reduction process shown in ISO2.
(1)Definition of the terms
Terms concerning risks may often be used in a variety of ways. In this section, in considering theconcept of reducing the risk caused by radioactive material, the terms used hereinafter shall bedefined as shown in Table 3-1 in reference to the definition of terms shown in ISO.
Table 3-1 Terns and definition
Term

Risk source

Consequence

Likelihood of
occurrence

Definition of the term used in considering
the concept of reducing risks caused byradioactive material
Material containing radioactive material
that has the potential to have a negative
influence on human health and theenvironment.
Severity of the negative impact of radiation
exposure to the public in terms of thehealth of people and the environment,
caused by the release of radioactive
material
Possibility of the occurrence of thenegative impact of radiation exposure to
the public in terms of the health of people
and the environment, caused by therelease of radioactive material

Definition in ISO
A potential cause that has anegative influence on humanhealth and the environment

―

―

Risk

A combination of “consequence” and“likelihood” posed by radioactive material

A combination of the “significanceof the negative influence onhumanhealthandtheenvironment” and the “possibilityof negative influence” caused by
the risk source

Risk level

Magnitude of the risk
In Chapter 3, the product of “consequence”
and “likelihood” expresses the “risk level”.

―

Risk reduction

Elimination of radioactive material or
reduction of “consequence” and/or“likelihood” posed by radioactive material―

Risk reduction
measures

Activity or means to reduce the risk

Activity or means to eliminate therisk source or reduce the risk

Meanwhile, influences caused by radioactive material include:





2

Influence on the environment
Radiation exposure to the public (external exposure, internal exposure)Environmental contamination, wide area diffusion
Radiation exposure of workers (external exposure, internal exposure) etc.

ISO/IEC Guide 51:2014; Safety aspects – Guidelines for their inclusion in standards

3-4This section describes the concept of reducing the risk for the purpose of restraining the radiation
exposure to the public as a representative factor of influence on the environment, and radiation
exposure of workers, which is important when considering risk reduction.
(2)Risk reduction principles
The risk reduction principle for the risk sources residing in the Fukushima Daiichi NPS is toappropriately combine activities or means to reduce the significance of negative influences onhuman health and the environment caused by risk sources and activities or means to reduce thepossibility of negative influences exerted on human health and the environment caused by suchrisk sources, promptly reduce the magnitude of the risk posed by each risk source, and to achieve
and maintain the risk level sufficiently low.
Meanwhile, the total risk of the facilities in the Fukushima Daiichi NPS is expressed by thesummation of risks posed by each risk source. Therefore, the risk reduction principle for the total
risk of the facilities in the Fukushima Daiichi NPSs is to achieve a state in which the level of total
risk of the facilities is sufficiently low by reducing the risk level of each risk source stepwisely andpromptly.
(3)Concept of risk reduction measures
In considering risk reduction measures based on the above principle, it is required to examinevarious alternatives (options) available for risk reduction measures and optimize the risk reduction
measure. When optimizing risk reduction measure, it is critical to make considerations on the basisof the 5 guiding principles described in Section 2.2.2, such as realizing the sufficient risk level
reduction (safe), ensuring workers safety in relation to risk reduction measures (safe and fieldoriented), being technically feasible (proven), ensuring the procurement of resources even withallocation of limited resources (efficient), and being promptly executable (prompt) etc.
In considering the risk sources to be addressed the risk reduction measures and plans on risk
reduction measures etc., the risk reduction procedure shown in ISO shall be followed. As shown in
Figure 3-3, major risk reduction process is performed through four steps: risk source identification,risk estimation, risk evaluation, and risk reduction.

Figure 3-3. Major risk reduction process

3-5

In the following sections, risk sources are identified (Section 3.2.1), and the risks posed by theserisk sources are estimated and evaluated (Sections 3.2.2 and 3.2.3). Then, Section 3.2.4 describes
the present status of risk reduction measures, followed by Section 3.2.5, which shows challenges
for activities to implement risk reduction measures.
3.2.1 Risk source identification
The identification of risk sources is the action of specifying the subject to which risk reduction
measures should be applied. The risk sources identified at this stage undergo subsequent risk
estimation and evaluation, and become the subject of risk reduction measures. Therefore, risk
source identification actions should be applied widely and comprehensively against the substances
existing in Fukushima Daiichi.
The risk sources at the Fukushima Daiichi NPS are substances containing radioactive materials
that may affect people and the environment, such as nuclides like uranium and plutonium, etc.
(hereinafter referred to as “heavy nuclides”), fission products (FPs) such as cesium, and activatedmaterials produced through reactor operation (cobalt-60, etc.). Among these, the risk sources listed
in the following table are identified as “major risk sources” which are currently having a relatively
high risk level.
There are risk sources that are currently assumed to have comparatively low concentration ofradioactive material than the major risk sources and need development of risk reduction measuresover the longer term. Among these risk sources are such as contaminated soil and sediments inthe port area etc. In order to conduct risk estimation and evaluation for these risk sources, collectionof information should continuingly be done regarding the state of contamination and the nuclides
contained in the contamination sources, and the analysis of migration paths of radioactive materials.
Table 3-2. Major risk sources
Fuel debris

Fuel debris in the RPVs/PCVs in Units 1-3
Fuel assemblies stored in the spent fuel pools (SFPs) in
Fuel in SFPs
Units 1-3
Fuel in the common pool
Fuel assemblies stored in the common pool
Fuel in dry casks
Fuel assemblies stored in dry casks
Contaminated water accumulated in the reactor buildings,
Stagnant water in the buildings
turbine buildings, main process buildings, and hightemperature incinerator building in Units 1-4.
Concentrated liquid waste etc. stored in tanks, including
Concentrated liquid waste, etc.
strontium-treated water that has been treated by any facility
other than the multi-nuclide removal equipment
Secondary Waste adsorption Storing adsorbent used in the cesium adsorption apparatus,
the second cesium adsorption apparatus, etc.
waste from columns
Waste sludge
Precipitation from the decontamination instrumentswatertreatment
Slurry produced during the treatment by the multi-nuclide
HIC slurrysystemsSolidremoval facilities stored in high integrity containers (HIC)
waste
Rubble(in the
Rubble with high-dose (30mSv/h and above) stored in thestorage facility)
solid waste storage facility.
RubbleRubble and felled tree etc. stored in Soil covered temporary
etc.
Rubble etc.
storage facility, Temporary storage test, Outdoor container(placed outdoors)
storage, Outdoor seat covered storage, Outdoor storage
Structures, pipes, components, and other items inside theContaminated structures, etc. in the reactor buildings, PCVs or RPVs that are contaminated withbuildings
radioactive materials dispersed due to the accident; andactivated materials from operation before the accident

3-6

In the Strategic Plan, risk estimation and evaluation were performed for the major risk sources
identified above, using such factors as radioactivity, form and containment state. The results areshown in Sections 3.2.2 and 3.2.3.
Six years have passed since the accident, and the radioactivity and decay heat of the risk sources
existing in the Fukushima Daiichi NPS are much lower than the levels immediately after theaccident. With decommissioning work expected ahead, effects of further radioactive decay ofradioactive materials should also be considered in the next stages. Figure 3-4 shows theradioactivity and decay heat of the fuel debris in each unit. The values in the graphs represent
relative figures to the values at the time of the accident and these do not consider radioactive
materials released into the environment. The current radioactivity is less than 1% of the level at thetime of the accident, and the current decay heat, less than 0.1%, both indicating a significant
reduction.

Figure 3-4. Evaluations of radioactivity (left) and decay heat (right) of the fuel debris in each Unit
(Source: JAEA-Data/Code 2012-018)

3.2.2 Risk estimation
Risk estimation is performed by analyzing the risk level of the risk sources. The method used here
to measure the risk level is the Safety and Environmental Detriment (SED) score, which wasdeveloped by the Nuclear Decommissioning Authority (NDA). SED scores quantify, as a major risk
to the public, the potential consequence and likelihood of internal exposure of people through intakeof radioactive materials by oral injection or inhalation. Although SED scores (values) have nophysical unit, this method is effective for analyzing risk sources on which available information
widely varies in reliability and accuracy to determine their risk levels and the effects of work to
reduce the risk level through mutual comparison. The risk level expressed in SED scores can beobtained by the following formula. The formula for the calculation of SED scores is described inAppendix 3.
SED-based risk level = (Hazard Potential) x (Safety Management)

3-7

To express the risk level in SED scores, the “consequence measure” is “Hazard Potential” andthe “likelihood measure” is “Safety Management”. The first term of the formula, “Hazard Potential”
considers the quantity of radioactive material contained in risk sources, i.e. gaseous, liquid or solidetc. state from the viewpoint of diffusion properties and the ease of being migrated into humans
and the environment, and the time margin through recovery when the safety function controllingthe instability specific to risk sources is lost. The second term of the formula, “Safety Management”,
represents the containment status and the state of the risk sources, and is composed of factors
ranking risk sources using a combination of elements including facility integrity and containment
function etc. and factors with combined elements including the state change in risk sources as wellas the packaged/monitored state etc. These factors are evaluated according to classification, where
each classification is assigned a specified score. Therefore, for determining the risk level of the
Fukushima Daiichi NPS's risk sources based on SED scores, the criteria for classifying the factorsof Safety Management were improved to make them more suitable for the risk sources of the
Fukushima Daiichi NPS.

3.2.2.1 Hazard Potential
Hazard Potential can be obtained by the following formula, based on the three factors: Inventory,
Form Factor (FF) and Control Factor (CF):Hazard Potential = (Total Radioactive Toxicity) x (Form Factor) / (Control Factor) = Inventory x FF / CF(1)Inventory
Inventory is a factor representing the toxicity caused by the radioactivity of a risk source.
Inventory used in SED developed by NDA is expressed by the product of the amount ofradioactivity of the risk source (Unit: Bq) and Specific Toxic Potential shown in Table 3-3, and is
the basic amount representing the degree of impact of radioactive substance on human body
(Refer to Appendix 3).
Table 3-3 Major nuclides and their potential toxicity (STP)
NuclideHalf-life3
STP (m3/TBq)4
Pu-238
87.7
years
66,000,000,000
Pu-23972,000,000,000
2.41×104 years
Pu-24072,000,000,000
6.54×103 years
Pu-24114.4
years
1-380,000,000
Am-241
57,600,000,000
4.32×102 years
Cm-24418.1
years
34,200,000,000
Sr-90
29.1
years
96,000,000
Cs1-342.06
years
12,000,000
Cs1-3730.0
years
23,400,000
For fuel debris, fuel in SFPs, stagnant water in the buildings, concentrated liquid waste, etc.,
secondary waste from water treatment systems (waste adsorption columns, waste sludge, HIC
slurry) and rubble (in the storage facility) as well as rubble etc. (placed outdoors), the amount ofradioactivity (Unit: Bq) was estimated from publicly available data. For fuel in the common pooland fuel in dry casks, radioactivity was estimated from the figures obtained for the fuel in SFPs.
3
4

ICRP Publication 72
NDA, Instruction for the calculation of the radiological hazard potential, EGPR02-WI01 Rev 3,
equivalent influence of adsorption

3-8

For contaminated structures, etc. in the buildings, the amount of activated materials wasestimated from publicly available data on ordinary reactors, and the amount of contamination
with FPs volatilized and dispersed at the time of the core melt accident, which was assumed toexist as deposits on the surface, was estimated based on the amount of surface deposits
obtained by using the analysis value from the severe accident analysis code. Regarding stagnant
water in the buildings, concentrated liquid waste, etc., secondary waste from water treatment
systems, and rubble (in the storage facility) as well as rubble etc. (placed outdoors), theradioactive materials contained in these risk sources and their radioactivity were estimated fromvalues obtained by sample analysis.
Meanwhile, as the Inventory was obtained through estimation based on currently available data,when new knowledge or data is obtained in the future, the amount of radioactivity will need to berevised or re-analyzed based on such new data used to review the Inventory for each risk source,
as needed.
(2)Form Factor (FF)FF is the factor representing the form of a risk source and measures a risk source's ease of
spreading and ease of entry into the human body and the environment. Therefore, radioactive
materials in form of gas, liquid and aerosol etc. that are likely to be diffused and migrated into
humans and the environment have higher scores. Concentrated liquid waste etc. and secondary
waste from water treatment systems etc. are directly observed in respect of the form, and FF canbe set comparatively easily. It must be noted that, in contrast to this case, FF involves a greater
degree of uncertainty for fuel debris, which presumably varies widely in form and whose
information as to which form of fuel debris accounts for how much of the total is not based on
direct observation.
To set FF, main forms of fuel debris were estimated based on the calculation analysis results ofthe severe accident analysis code. FF for risk sources other than fuel debris was set from direct
observation and/or analysis. Spent fuel was considered a discontinuous solid because it is an
assembly of fuel elements (pellets), and a part of the volatile FPs contained in spent fuel wereassumed to have accumulated in the plenum and were treated as powder. Concentrated liquid
waste, etc., and stagnant water in the buildings were considered liquid, and secondary wastefrom water treatment systems was treated as sludge or powder. Meanwhile, no score is specified
for FF setting values for the treatment of the product of reaction/adsorption although radioactive
materials are chemically reacted/adsorbed in adsorbents in the waste adsorption columns etc.
(Refer to Appendix 3). For this reason, to be conservative, score for powder materials was used
for the adsorbent in the waste adsorption columns as it was considered to be a surface
contaminants caused by powder radioactive materials. Assuming that the majority of rubble (inthe storage facility) as well as rubble etc. (placed outdoors) is contaminated with radioactive
materials adhered to the surface of them, such rubble, etc. were treated as powder, as it isassumed a surface contamination. Contaminated structures, etc. in the buildings, for which
activated materials are considered solid, contamination with FPs was treated as powder, as it isassumed a surface contamination.
(3)Control Factor (CF)CF is the factor that scores the time margin before the manifestation of an event thatcompromises the containment status due to instability inherent in a risk source under the normal

3-9control state5. For this document, the risk source with the shortest time margin for controllability,or the least controllable risk source, was selected conservatively and CF was set for it. Note thatthe robustness and reliability of the containment status against typical natural disasters (e.g.,
earthquake) are not considered for CF, because these are included in the criteria for modified
FD, which is discussed later.
The value of CF for fuel debris was estimated from the time margin before the cooling function
is lost or the time margin before the hydrogen concentration exceeds the explosion limit after thenitrogen supply function is lost, whichever the shorter. However, it should be noted that the settingof CF values involve considerable uncertainty because fuel debris cooling status and the rate ofchange in the hydrogen concentration etc. are highly uncertain. For fuel in SFPs and fuel in thecommon pool, CF was set as a time margin before a major loss of cooling function begins after
the pool's water level lowers to the top of the fuel due to a loss of the cooling system. Meanwhile,
since there is no need to constantly control the subcriticality by means of dynamic equipment etc.
under the normal control state, considering the current assessment that the state of subcriticality
is maintained in fuel debris and fuels in the pool, controllability against criticality was determined
to be sufficiently high under the normal control state.
As fuels in dry casks, concentrated liquid waste, etc., and rubble etc. that do not require control
by dynamic equipment, such as cooling to deal with decay heat because they can dissipate heatby natural convection, they were determined to be highly controllable. For contaminatedstructures, etc. in the buildings, CF considered the possibility of release of FP deposits on thesurface after a temperature rise due to the loss of cooling function. No need for cooling wasconsidered for secondary waste from water treatment systems. For HIC slurry, which is undercontinuous monitoring of the effects of hydrogen generation, consideration was given to the effectof mitigating hydrogen generation through measures for the removal of supernatant fluid etc.
For waste sludge, the time margin before the loss of agitation function to prevent solidification
was estimated.

3.2.2.2 Safety Management
Safety Management can be obtained by the following formula, using the modified FD and modified
WUD factors:
Safety Management = ((Facility Descriptor) x (Waste Uncertainty Descriptor)) 4= (Modified FD x Modified WUD) 4(1)Modified FD (Facility Descriptor )Modified FD is a factor that describes the current state of containment for a risk source. The risk
sources were classified and scored according to the reliability of their state of containment, basedon such determination criteria as the redundancy of the container etc., whether the state ofcontainment can be ensured until the risk source achieves a sufficiently stable controlled state,
and whether the containment status can withstand typical natural disasters such as earthquakes
or tsunamis, and so on.
5

CF mainly considers the following inherent instability as subject to control: heating (decay heat),corrosiveness (corrosion of the subject), ignition, corrosion (corroding the container, etc.), byproduct(generation of gas), reaction with water (leaching, radiolysis of water under radiation), reaction withgas, and criticality.

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Regarding fuel debris, fuel in SFPs, and stagnant water in the buildings, it was judged that areasonable state of containment is secured because pressure and other key parameters are
generally stable for them, although reactor buildings and other structures that constitute thecontainment function for these risk sources are affected by the core melt accident and ensuinghydrogen explosions, as well as tsunamis. For fuel in SFPs, consideration was given to theeffects of large pieces of rubble found on operating floors in Units 1 and 3, which had probably
fell due to the hydrogen explosion. Among them, large pieces of rubble in the Unit 3 fuel pool
were removed by 2015, and the effects of the removal of the large rubble were taken into accountas well. For stagnant water in the buildings, the containment state was determined to have been
improved because enhancement of sub-drain systems and start of partial operation of the landside impermeable wall have strengthened the water level adjustment function to groundwater.
Consideration was given to the fact that waste sludge produced early on after the accident isstored in the granulated solid waste storage tank located at the basement floor of the mainprocess building and is subject to leakage monitoring and hydrogen ventilation.
Rubble etc., concentrated liquid waste, etc., and secondary waste from water treatment systems
are stored in separate facilities built after the accident, and the common pool and dry casks are
assumed to be unaffected by the accident, and were evaluated based on their respective storage
conditions.
(2)Modified WUD (Waste Uncertainty Descriptor )Modified WUD is a factor that represents the degree of influence that instability inherent in a risk
source and its controlled state have on stable storage of the risk source. Modified WUD classifies
and scores the characteristics of instability in a risk source, and the adequacy of its controlled
state based on the criteria such as the presence/absence of instability (e.g., degradation,
corrosion) in the risk source, whether the risk source is packaged, and whether the risk sourceis subject to appropriate control and monitoring. This means that a greater modified WUD scoreis assigned to a risk source that has large inherent uncertainty and/or that is difficult to handle.
Some other considerations that are not included in the original SED score were also given, suchas whether a risk source affects the risk reduction measures for other risk sources, and whether
a risk source requires early implementation of risk reduction measures.
Risk sources such as fuel in the common pool and fuel in dry casks were assigned a relatively
small modified WUD because, although consisting of fuel elements containing heavy nuclides,
they are well-understood in properties, based on measured values and analysis, and are underappropriate control and monitoring.

3.2.3 Risk evaluation
Figure 3-5 is an example of a two-dimensional chart showing the risk levels of the major risk
sources, based on a risk estimation using information available as of March 2017. The vertical axis
represents “Hazard Potential”, and the horizontal axis, “Safety Management”. In the Figure, for the
purpose of scoring Hazard Potential and Safety Management, the influence of variation in theconcept of determination elements as well as the uncertainty of data on each factor is shown by
the extent. For Hazard Potential, the uncertainty in estimation in respect of the concentration,amount, and the form of radioactive materials as well as time margin until restoration wasconsidered. As for contaminated structures, etc. in the buildings and fuel debris, uncertainty of the

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Hazard Potential (logarithmic scale)form was set to a large value. For Safety Management, in consideration of the fact that two factors
were obtained by quantifying originally qualitative information, uncertainty was used for the widthof the score. Various storage states were considered for rubble etc. (placed outdoors).

Fuels incommonpool

Fuels in dry
casks

Category II
Category I

FueldebrisHIC
slurryWaste
absorption
columns Contaminatedstructures inbldg

Fuels in
SFPs

Concentrated
Waste
liquid waste
sludgeRubble(storage facility)

Stagnant
water in bldgs

Category III
Rubble, etc.
(outdoors)

Region of sufficiently
stable management

Safety Management (logarithmic scale)Figure 3-5. Example of risk levels assigned to the risk sources at the Fukushima Daiichi NPS
Based on the results of relative comparison of risk sources by the risk level, it was deemed
appropriate that risk reduction measures should be developed according to the followingclassification.
[Category I] Risk source to be addressed as soon as practicable

Fuel in SFPs
Stagnant water in the buildings[Category II] Risk source to be addressed safely, effectively and carefully with thoroughpreparations and technologies to realize a more stable condition


Fuel debris[Category III] Risk source that requires actions to be taken for a more stable condition






Concentrated liquid waste, etc.
Waste sludge
HIC slurry
Rubble etc. (placed outdoors)
Contaminated structures, etc. in the buildings

The risk sources other than those listed above, i.e. fuel in the common pool, fuel in dry casks, andrubble (in the storage facility) are stored in facilities unaffected by the accident (from the viewpoint

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of controllability of cooling function etc. and structural integrity) that were designed before theaccident and have been in safe use since before the accident. The waste adsorption columns etc.,
which were designed after the accident, are not subjected to an environment that would propagate
corrosion of the container, as demonstrated by inspection for the presence of residual water andanalysis of residual water. These risk sources can be kept in the sufficiently stable controlled statethrough continuous management.
It is conceivable that these risk sources are under stable management that is similar to that forequipment in other safely operated nuclear power stations. Therefore, it is our immediate objective
to bring the risk sources evaluated fall under Categories I, II and III into the region of sufficiently
stable management.

3.2.4 Current status of risk reduction measures
Among the major risk reduction measures taken to date, fuels in the pool of Unit 4 were removed,
and then transferred to the common pool located in the region of sufficiently stable management(completed in 2014). Processing of concentrated saltwater was completed (in 2015). Highconcentration contaminated water stagnant in the seawater pipe trenches of Units 2 to 4 wasprocessed, and the trenches were blocked (completed in 2015) etc.
Moreover, in order to address the stagnant water in the buildings, the inflow of groundwater was
reduced through the enhanced sub-drain function and the partially started operation of the landside impermeable wall, mitigating the increase in the amount of stagnant water in the buildings.
Additionally, reduction in the total amount of stagnant water in the buildings is pursued by ensuring
the difference between the stagnant water level in the buildings and the groundwater level aroundthe buildings. Meanwhile, as for solid waste, a storage management plan has been formulated toreduce the risks.

3.2.4.1 Category I risk reduction measures
Category I consists of risk sources that should be addressed as soon as practicable, and (1) fuelin SFPs and (2) stagnant water in the buildings fall under this category.
(1)Fuel in SFPs
The fuels in SFPs in Units 1-3 are placed in an earthquake-resistant rack to be kept away fromadjacent ones. In addition, there is a relatively long time margin before fuel assemblies begin to
be exposed to the air due to evaporation of the pool water even if the coolant supply stops. In
addition, the fuel mainly keeps the form of fuel rods contained in cladding tubes, which are lesslikely to spread.
Meanwhile, in Units 1 and 3, due to the hydrogen explosion associated with the core melt
accident, the reactor building was damaged. In addition, as the cooling equipment etc. of Units
1, 2 and 3 were damaged in the accident, functions equivalent to the conventional containmentand controlling functions available before the accident are not ensured at present. Therefore, thefuel in SFPs is planned to be removed from the current storage and transferred to the common
pool, so it will be possible to bring it into the region of sufficiently stable management.
In Unit 3, in preparation for unloading fuels, the removal operation of large rubble in the poolstarted was completed in November 2015. Removal of rubble from the pool and operating floors

3-13eliminated the risk of fallen rubble etc. damaging the fuels in the pool and/or the rackaccommodating fuel assemblies. Currently, in preparation for unloading fuels, work was initiatedto install a spent fuel removing cover etc. in January 2017.
In preparation to start removal of the fuel in the SFP of Unit 1, removal of the rubble on theoperating floor will be required. By means of the past surveys, a high dose rate was confirmedon the operating floor, and the south side of the reactor building, where the SFP is located, is still
covered by the ceiling crane and the collapsed roof. Therefore, the removal method etc. are now
being studied and examined.
In Unit 2, the reactor building wasn’t significantly damaged, and the containment status is better
than that of the fuels in the pools of Units 1 and 3. In preparation for fuel removal, an accessgantry was installed on the operating floor in February 2017.
(2)Stagnant water in the buildings
Actions have been taken to remove the water stagnating in the condenser in the Unit 1 turbine
building and the stagnant water on the bottom floor of the Unit 1 turbine building (removal
completed in March 2017). This helped reduce the overall amount of radioactive materials in thestagnant water in the buildings, as shown in Figure 3-6. In addition, containment function has
tended to be improved through the enhanced sub-drain function and the partially started
operation of the land-side impermeable wall in a reinforced water level control for the purpose ofmaintaining the level of stagnant water lower than the groundwater. Efforts are still in progress
to lower the amount of radioactive materials by draining the water with a high concentration ofradioactive substance stagnating in the condensers in the Unit 2 and 3 turbine buildings.
In the process of pumping the stagnant water in the buildings, sedimentation of sludge containing
radioactive material has been found. To prevent this sediment from being left as a risk source
after the stagnant water in the buildings has been pumped out, it is being collected while being
stirred in the stagnant water.

Figure 3-6. Reduction in the radioactivity of stagnant water in the buildings

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3.2.4.2 Category II risk reduction measures
Risk sources in Category II should be addressed safely, effectively and carefully with adequate
preparations and technologies to achieve a more stable state, and fuel debris falls under thiscategory.
Although fuel debris is assumed to exist in a certain level of containment state, considering thevarious forms of its existence at various places and the possibility of changes in mid- and long-term
chemical properties as well as in physical property, technical studies are now in progress so that itshould be taken out as soon as it is practicable, and placed in an appropriate storage state.
The internal investigations are gradually revealing the contamination state inside the reactor
buildings and the situation inside the PCVs. However, the actual state of fuel debris isn’t fully
grasped, and, though estimations have been made based on analyses and experiment, etc., it mustbe said that there is still a high level of uncertainty. For this reason, investigations are now under
way to gather and accumulate information on the form and distribution etc. of fuel debris.

3.2.4.3 Category III risk reduction measures
Category III are the risk sources required actions to be taken to reduce the risk level and to achievea sufficiently stable condition, and these risk sources consist of concentrated liquid waste, etc.,
waste sludge, HIC slurry, rubble etc. (placed outdoors), and contaminated structures, etc. in thebuildings.








Further reduction of risk level for concentrated liquid waste, etc. is pursued by purifying
strontium-treated water by the multi-nuclide removal facilities.
Options for applicable risk reduction measures are now being studied for waste sludge.
HIC slurry is contained in high integrity containers (HIC) and it is in a comparatively
stable condition. Options for risk reduction measures are now being studied for thefurther reduction of risk level.
In regards to rubble etc. (placed outdoors), measures are under way to store parts ofused protective clothing and combustibles etc. with comparatively low radiation dose insolid waste storage facilities after they are incinerated in miscellaneous solid waste
incineration facility and the incinerated ash accommodated in containers. Through therisk reduction measures, rubble etc. are now being brought into the region of sufficiently
stabile management.
Contaminated structures, etc. in the buildings will be removed when the buildings etc.
are dismantled and removed. However, when these become obstacles to access thefuel debris classified under Category II and so on, consideration should be given to
remove them etc., as needed.

The risk sources classified in Category III are generally assigned lower risk levels than those in
Categories I and II, and they are assumed to be able to retain certain risk levels through continuous
maintenance and management. However, some of these risk sources include those contaminatedby radioactive material attached on the surface, those in a sludge-like form, and those having
reactivity causing generation of hydrogen etc. Therefore, the risk sources in Category III requirecontinued deliberations on risk reduction measures and systematic implementation of risk reduction
measures to bring them into the region of sufficiently stabile management.

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3.2.5 Challenges during risk reduction(1)Risks during risk reduction
In the implementation of risk reduction measures described in Section 3.2.4, temporary changes
in the risk level and exposure of workers must be considered. In the course of implementation, arisk reduction measure may increase or reduce the risk level as it will affect the state of risk
sources and containment functions. Workers will be exposed to radiation as they install, repair
or operate facilities. This section summarizes the concept for addressing these issues.
If no risk reduction measure is taken, existing risk sources may rise in the risk level over time as
facilities degrade and the state of risk sources changes, and may eventually reach an
unacceptable level after a long period of time. Even if facilities and risk sources undergo nochange, risks will become more likely to manifest themselves over time (e.g., an earthquake may
occur and affect the facilities). Such a situation must be avoided by implementing risk reduction
measures within a reasonable timeframe to reduce risks to a broadly acceptable level.
If the implementation of any of such risk reduction measures may temporarily increase the risk
level, minimizing the increase is crucial. However, if too much attention is paid to minimizing theincrease, it would prolong the time needed for preparations and work, and the existing risks would
remain for a longer period on the contrary. The prolonged work period may also increase theexposure of workers. Considering these factors, a solution to the case involving the possibility ofa temporary increase in the risk level should be carefully developed with well-balanced
consideration of promptly mitigating the existing risks and reducing the exposure of workers.
The way the risk level changes over time varies by risk source, and therefore it is essential totime an action appropriately to the characteristics of the risk source and to prepare for it carefully.
Given that this project must be carried out with various uncertainties, the project team must be
flexible enough to stop and review the plan every time an uncertainty is resolved.
(2)Examples of risks and actions on themThis section shows examples of risks during risk reduction and examples of associated actions,particularly for fuel in SFPs and fuel debris, which are classified into Categories I and II,
respectively, and have a large inventory of radioactive materials. Before implementing any risk
reduction measure, it is crucial to perform an in-depth consideration and take adequate actions
to reduce the risk level.
Fuel in SFPs will be removed from the SFPs, transported in the local transport container to thecommon pool for storage.6 Both the SFPs and the common pool provide sufficient subcriticality
and cooling capacity. The local transport container is also designed to provide an adequate
margin of criticality and cooling safety. For containment, fuel pellets are contained in cladding,and the local transport container has sufficient strength. However, considering the possibility that
fuel may be damaged by a fallen heavy object or dropping during transport, impact evaluation
and anti-drop measures are implemented in advance. To prevent scattering of radioactive dust
and to restrict exposure of workers during rubble removal, decontamination and shielding
activities are in progress with extensive preparations.
Fuel debris will be cut into pieces of specified size and put in canisters, which will then betransported to the storage area for storage. To prepare for fuel debris retrieval, measures to
6Special issues, Efforts toward Spent Fuel Removal from Spent Fuel Pools of Fukushima Daiichi
Nuclear Power Station, Journal of the Atomic Energy Society of Japan, Vol. 59, No. 1 (2017).

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address recriticality due to the rise in the water level or other events, development of a circulation
loop for cooling water, and design of cans for accommodating fuel debris etc. are now being
pursued. In addition, studies are underway to enhance the containment function for particulate
fuel debris that will be produced in the cutting process, and studies for decontamination andshielding are underway to reduce exposure of workers in the contaminated reactor buildings.
These issues will be discussed in detail in Chapter 4.

Column: Change in Risk over TimeThe figure below shows the concept of risk management in the U.K. Even if the current risk levelof a risk source is in the white region, it does not mean that the state will be tolerable forever; itwill become intolerable (reach the yellow region) sometime in the future. As time passes, the risk
level may further increase due to the degradation of the facilities or the risk source (dotted line).
By contrast, if risk reduction measures are taken, the risk level can be prevented from reaching
the unacceptable region through careful preparation and thorough management, although the risk
level may temporarily increase. The goal should be sufficiently reducing the risk level while keeping
it from entering the unacceptable or intolerable regions (solid line).

References: V. Roberts, G. Jonsson and P. Hallington, Collaborative Working Is Driving Progress inHazard and Risk Reduction Delivery at Sellafield, 16387, WM2016 Conference, March
6-10, 2016.
M. Weightman, The Regulation of Decommissioning and Associated Waste
Management, 1st International Forum on the Decommissioning of the Fukushima
Daiichi Nuclear Power Station (April 2016).

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3.3 Implementing risk reduction strategies
Among the primary risk sources, fuel in SFPs and stagnant water in the building in Category I havealready been undergoing measures by the government and TEPCO. The stagnant water in thebuildings has been significantly reduced, with the treatment expected to finish in 2020. However,since this treatment will have to continue beyond that year, a medium-term strategy should be
developed.
A concrete risk reduction strategy for fuel debris, in Category II, and that for secondary waste fromwater treatment systems and rubble etc., in Category III, is described in Chapters 4 and 5,
respectively (Figure 3-7).

Figure 3-7. Deployment of risk reduction strategy
In implementing the risk reduction strategy, ensuring safety is listed at the head of 5 basic principles,
which is considered in Chapters 4 and 5. Here the concept common to both chapters is described.
Meanwhile, the term “safety” as an objective here is to protect people and the environment fromharmful effects of radiation generated from facilities and various activities.
The starting point of the basic concept for ensuring safety is the fundamental safety objective andfundamental safety principles presented in the IAEA's SF-1 7 document, which states that toachieve the fundamental safety objective, safety requirements should be established in accordancewith the fundamental safety principles, and safety measures be taken. The document declares thatthe fundamental safety objective is to "protect people and the environment from harmful effects ofionizing radiation" and that measures should be taken to control the radiation exposure of people
and the release of radioactive material to the environment, to restrict the likelihood of events, andto mitigate the consequences of events.

7

Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1 (2006).

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Although SF-1 sets forth 10 fundamental safety principles, the Strategic Plan considers thefollowing principles in particular, except for those on responsibility and roles etc., because it isintended for technical deliberations.
“Justification of facilities and activities” (Principle 4), “optimization of protection” (Principle 5) and“limitation of risks to individuals” (Principle 6) correspond to the International Commission on
Radiological Protection (hereafter referred to as “ICRP”) three fundamental principles ofradiological protection (justification, optimization of protection, and dose limitation).
“Protection of present and future generations” (Principle 7) requires the radiological impact,
including environmental effects, to be considered across a wide geographical and temporal range
and is even relevant to the management of radioactive waste. “Prevention of accidents” (Principle8) refers to defense in depth for preventing accidents and mitigating the consequences of accidents.
One of the situations considered in “protective actions to reduce existing or unregulated radiation
risks” (Principle 10) is that remedial actions are taken after an uncontrolled release of radioactive
material to the environment. The decommissioning of the Fukushima Daiichi NPS is regarded aspart of this situation.
When these principles are applied to the Fukushima Daiichi NPS, the actual situation should befully considered to allow flexibility in rationally implementing measures to reduce exposure andimpacts on the environment and decommissioning measures.
The safety requirements that should be established in accordance with the above fundamental
safety objective and fundamental safety principles will differ depending on the phase of the longterm decommissioning. In defining the safety requirements at each phase of decommissioning, theclassification of the post-accident decommissioning process described in the IAEA's NW-T-2.78
document can serve as a guide. The document divides post-accident activities into the phases of
emergency response, stabilization, post-accident cleanup, safe enclosure, and active
decommissioning & site remediation. Since the Fukushima Daiichi NPS is past the emergency
response phase, this section focuses on the subsequent phases.
A particular requirement in early phases, such as stabilization and post-accident cleanup, is limiting
the risks arising from the activities to protect people and the environment from radiological impact.
Such safety requirement is also important in making well-balanced decisions to respond to
temporary changes in the risk level resulting from the implementation of risk reduction measures,
as mentioned in Section 3.2.5. In later phases, such as safe enclosure and active decommissioning& site remediation, even more emphasis should be placed on protecting people and theenvironment from the radiological impact of risk sources in the containment than on ensuring safety
in associated work.
Among the risk sources present at the Fukushima Daiichi NPS, Category I and II risk sources arein the phase of stabilization or post-accident cleanup. As for stagnant water in the buildings andfuel in SFPs, which are Category I risk sources, the already-planned risk reduction measures arein progress or becoming ready, and associated safety measures are also underway in accordancewith “Matters to be addressed”.
The retrieval of fuel debris, which is a Category II risk source, will also be carried out in accordancewith “Matters to be addressed” in general. Since the activities for this have yet to be specified indetail, safety requirements should be developed as appropriate to them. Considering the8

Experiences and Lessons Learned Worldwide in the Cleanup and Decommissioning of Nuclear
Facilities in the Aftermath of Accidents, IAEA Nuclear Energy Series No. NW-T-2.7 (2014).

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characteristics of fuel debris and its retrieval work, Chapter 4 defines the basic concept for ensuringsafety for fuel debris retrieval.
Solid waste classified in Category III varies widely in characteristics, including solid waste
generated from Category I and II risk reduction measures. Risk reduction in respect to solid waste
shall be implemented basically in accordance with “Matters to be addressed”. Chapter 5 below
summarizes principles for ensuring the safety concerning waste storage/management andtreatment/disposition on the basis of examinations and studies on safety policies for waste
treatment/disposition proposed by international organizations (e.g., IAEA, ICRP) by taking into
account its various characteristics different from those occurring from wastes in ordinary nuclear
power plants.
In any of these cases, to develop an effective basic concept for ensuring safety, it is essential to beflexible in doing so considering the situation at the Fukushima Daiichi NPS and sharing issues withrelated organizations.

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4. Strategic Plan for fuel debris retrieval
4.1 Study plan for fuel debris retrieval (risk reduction)(1)Stable state at present
Fuel debris is characterized by “containing nuclear fuel materials, not being contained in cladding
tubes, and existing in a mixed state with other materials”, naturally involving the risk for criticality,
decay heat, containment, and high radiation, as well as risk-related factors such as generationof hydrogen and deterioration of the integrity of support structure. When implementing themanagement of these risks, we face difficulties such as “uncertainty” caused by the fact that thesituation in the reactor is not sufficiently understood, “instability” caused by the fact that fuels
melted and facilities were damaged during the accident, and “inadequate management” causedby the fact that access is limited due to the severe radiation environment.
On the other hand, during the six years after the accident occurred, the risk level associated withthe fuel debris in Units 1-3 has been significantly improved. Through the decay of short half-life
nuclides with the passage of time after the accident, the amount of radioactivity (Bq) in the reactor
was reduced to about a several-hundredth, and the decay heat was considerably reduced to no
more than a thousandth (Refer to section 3.2 for details.). Moreover, by means of the emergency
response measures after the accident and subsequent stabilization work by TEPCO, the plantparameters showing the state of criticality, cooling, and containment etc. have maintained adegree of a stable state (Refer to section 4.3 for details.).
(2)Significance of fuel debris retrieval
The currently maintained stable state differs from the state where nuclear materials are controlledwith sufficient engineering measures in ordinary nuclear power stations, and the degree of thestable state is maintained by a temporary measure for the damaged reactor building and themelted reactor. Considering the above-mentioned difficulties in risk management, retrieval of fueldebris is a means to fundamentally improve such a situation and to bring it to a more stable andsafe condition, which requires strategies from the following two standpoints depending on therisks involved; reduction of mid-term risks and long-term risks.
Risks seen from a mid-term standpoint include having negative impacts on external objects
through deviation from “a degree of a stable state” currently maintained for fuel debris, examplesof which include; the possibility of occurrence of recriticality and a cooling problem, deteriorationof the internal structure of the reactor, releaking of radioactive materials etc. (Refer to 4.4 fordetails.) Though the likelihood of these unforeseen events is expected to be low as long as thecurrent degree of the stable state is maintained through appropriate control, it is desirable to takemeasures to better understand the situation and eliminate risk sources as soon as possible inresponse to the situation in the nuclear reactor where no direct control is established. We expectthat nuclear reactors can be controlled in a more stable manner by retrieving highly unstable fueldebris (fuel debris accompanying instability in existence and physical/chemical state) andbringing it into a stable stored state, as well as by taking adequate measures in a timely manner
after confirming the situation of fuel debris in the reactor and the internal structure.
The risk seen from a long-term standpoint is environmental contamination caused in the future
by highly toxic nuclear fuel materials leaking into the environment in association with thedeterioration of buildings. In Japan, it has conventionally been a basic policy to ensure the very
long-term safety of spent nuclear fuels by isolating high-level radioactive waste from the humanenvironment (geological disposal) after separating and stabilizing them through reprocessing.

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Leaving the fuel debris of Units 1-3, equivalent to 270 tons of spent fuels, in the reactor buildings
affected by the accident and causing concerns over the containment function for a long time
doesn’t comply with this policy. This is because the durability of the damaged reactor buildings
is limited, and the containment function can’t be guaranteed to be maintained for a long time.
Therefore, it is our basic policy to retrieve the fuel debris within the time period when thecontainment function in the reactor buildings can be maintained (about several decades), bringthem into a stable stored state under sufficient control, and ultimately reduce the risk to almostthat of the backend project.
The above standpoints would generate our concerns over the efforts to address the accident in
Chernobyl Nuclear Power Station Unit 4.Leaving the nuclear fuel materials as they are may
ensure the containment function for the time being, but such a measure without a definite
perspective for retrieval is nothing but an easy way of postponing the solution to future
generations as it is difficult to control them safely in the long run. At present, as the fuel debris is
unexpectedly aging together with deteriorated claddings etc., efforts are now being continued
through installation of additional containment structures etc. under the policy of ultimately
removing the debris and so forth.
Therefore, in the decommissioning of the Fukushima Daiichi NPS, measures like the above toleave the nuclear fuel materials as they are should not be adopted, but the ones to retrieve thefuel debris shall be pursued as shown below.
While reduction of both mid-term risks and long-term risks are important in the retrieval of fueldebris, the former requires prompt responses and the effectiveness of stabilization in the reactor,
while the latter, taking some time to achieve, is expected to lead to a high rate of fuel debris
retrieval. For this reason, during the early operations for removing fuel debris, it will be necessary
to attach importance to the reduction of mid-term risks, and at the same time to choose a method
for retrieving fuel debris as effectively as practicable. It may be our objective for the time being
to use this method to reduce mid-term risks by removing a certain amount of fuel debris, and tosecure “a risk level as low as widely permitted by society”. Then afterwards, we will aim toeliminate risks from a longer point of view (elimination and isolation of nuclear fuel materials)
through efforts like subsequent further retrieval of fuel debris and dismantlement of facilities etc.
Therefore, for the moment, it will be required to retrieve fuel debris in preparation for the reductionof mid-term risks.
(3)Controlling the risks at the time of retrieving fuel debris
Operation to retrieve fuel debris is the act of modifying the current stable status, and bringing
about a change in the status through access to the fuel debris. However, the risks accompanying
such operation (leak of radioactive materials and the radiation exposure of workers caused byany defect etc. during the work) must not be allowed to get higher than the permitted range. In
addition, resources such as manpower and time etc. that can be invested to decommissioning
are not infinite. Therefore, retrieval of fuel debris needs to be carried out by a safe, reliable andrealistic means that can contain the risks associated with such retrieval to a permissible range.
That is, on the basis of the “attitude not prioritizing the process but focused on risks”, as shownin the Mid- and Long-Term Roadmap (revised on June 12, 2015), it is required to strike anappropriate balance between immediate risk reduction through the retrieval of debris andcontainment of risks during the retrieval work. For this purpose, in considering the retrieval of fueldebris, according to international safety principles of IAEA/ICRP etc., the retrieval work will need
to be pursued in a flexible manner in accordance with the careful evaluation of risks.

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4.1.1 Positioning of a strategic proposal towards determining a fuel debris retrieval policy
In the Mid- and Long-Term Roadmap (revised on June 12, 2015) stipulates that the realization
evaluation and technical comparative verification be conducted to determine the feasibility ofmultiple construction methods envisaged by combining the PCV water level and access routes tofuel debris etc. and on the basis of the results, the policy for retrieving fuel debris from each unit beestablished by the summer of 2017. In addition, drawn is a process where, based on the establishedpolicy, R&D in consideration of the characteristics of each unit and engineering work for applying
such R&D results to actual units will be carried out, and then, initiatives should be defined for
deciding fuel debris retrieving method for the first unit and towards actually starting fuel debris
retrieval.
So far, from the viewpoint of element technologies required for fuel debris retrieval and of the entire
system combining such technologies, feasibility of the three priority fuel debris retrieval methods(1) method to flood the PCV and access the debris from above, 2) method to access the debris
from above in the air, and 3) method to access the debris from the side in the air) were intensively
examined. By means of the current studies, requirements for the realization of each engineering
method (Refer to Figure 4.1-1.) and the characteristics as well as the advantages/disadvantagesof access methods were clarified, and the internal situation in respect of the distribution andproperties etc. of fuel debris could be grasped to a certain degree. Additionally, certain results wereobtained in terms of the development of element technologies.
A strategic proposal towards determining a fuel debris retrieval policy, intended to aim for the“determination of a policy for fuel debris retrieval from each unit” by the government, shows thetechnical grounds for the fuel debris retrieval policy on the basis of the information obtained up tonow, and the concept of retrieving fuel debris that is conceivable at the current stage, as well as
the approach to the initiation of fuel debris retrieval which is considered most appropriate at present.

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4-4

Figure 4.1-1 A logic tree established toward the reduction of risks posed by fuel debris

4.1.2 Flow of studies on a strategic proposal towards determining a fuel debris retrieval
policyA study on a strategic proposal was conducted following the flow shown in Figure 4.1-2. In thisStrategic Plan, the detailed study content is presented in Section 4.2 through Section 4.7 as
described below.
In Section 4.2, the concept of ensuring safety that needs to be observed in fuel debris retrieval issummarized, taking into account the characteristics of the Fukushima Daiichi NPS. The concept ofensuring safety forms the basis of safety requirements and functional requirements for any fueldebris retrieval method.
In Section 4.3, the current status of Fukushima Daiichi NPS is summarized from the viewpoint ofsite conditions represented by the distribution of fuel debris, conditions of reactor internals, and theradiation environment, as it becomes the starting point of subsequent studies.
To determine a policy of fuel debris retrieval, it is important to gain general understanding on the
distribution of fuel debris. Therefore, currently most probable distribution of fuel debris will be
deduced from any available information obtained from analysis, indirect measurement and direct
measurement. Additionally, results of investigation on reactor internals and radiation doses willserve as the basis for deciding access routes.
In Section 4.4, taking into account the distribution of fuel debris summarized in Section 4.3, its
morphology, properties, control status such as cooling, and containment status, the potential risk
of fuel debris to cause radiation exposure is evaluated. For units 1, 2 and 3 of the Fukushima Daiichi
NPS, albeit different in quantity, fuel debris is considered to exist both inside RPV and at the bottomof PCV, and the cooling conditions and properties of fuel debris may vary depending on the location.
Comparative risk assessment of fuel debris in different states will be conducted to provide
approximate indications of attainable risk reduction effects.
In Section 4.5, for the three priority fuel debris retrieval methods, feasibility of their technicalrequirements will be assessed. Here, technical requirements consist of the nine requirements inthe logic tree for reducing the risks of fuel debris as shown in Figure 4.1-1.
In Section 4.6, taking into account technological assessment described from Section 4.3 throughSection 4.5, currently imaginable approaches for starting fuel debris retrieval will be studied andsummarized as comprehensive assessment from the viewpoints of five basic concepts (safe,
certain, rational, swift, and field-oriented).
In Section 4.7, based on the results of comprehensive assessment, recommendations towards
determining a fuel debris retrieval policy and initiatives after determining a retrieval policy are given.
Initiatives after determining a fuel debris retrieval policy are quite important in “determining a fueldebris retrieval method for the first unit” and further in putting subsequent studies towards starting
fuel debris retrieval into concrete shapes.

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4-6
Figure 4.1-2

A flow of considering a strategic proposal

4.2 Basic concept for ensuring safety in the fuel debris retrieval operation
The basic concept of ensuring the safety for fuel debris retrieval is essentially different from that forother nuclear facilities as in for it being an action to reduce already existing potential risks. That is,taking into account time-dependent change in the risk level as described in Section 3.2.5, realizing
early transition to a storage state with high management levels is required. However, starting
retrieval prematurely with insufficient preparations and safety measures may affect the safety insideand outside the site. Therefore, it is important for all the relevant institutions to join forces in realizing
both early commencement of fuel debris retrieval operation and ensuring the safety inside andoutside the site.
In this Section, safety functions pertaining to retrieval of fuel debris and the concept for ensuring
the safety will be summarized based on the basic safety principles of international institutions andthe characteristics of the Fukushima Daiichi NPS in terms of ensuring the safety.
(1)Studies based on IAEA Fundamental Safety Principles (SF-1)
A. Safety objective
The basic safety objective of SF-1 (protecting human life and health and the environment fromharmful effects of radiations) applies to all facilities and activities pertaining to nuclear power,
and measures have to be taken:
To control the radiation exposure of people and the release of radioactive material to theenvironment;
To restrict the likelihood of events that might lead to a loss of control over a nuclear reactor
core, nuclear chain reaction, radioactive source or any other source of radiation;
To mitigate the consequences of such events if they were to occur.

-

B. Safety principles
For the safety principles described in 4, 5, 6, and 10 in Section 3.3, their applicability to fueldebris retrieval operation has been investigated.
Realize risk reduction effects that outweigh an increase of exposure of persons
associated with the operation (Justification)
 Conduct safely and promptly taking into account the points below (Optimization:
principle of ALARA/ALARP) Suppress exposure of the public and workers as much as practically possible
 Avoid extension or an increase in lasting risks of fuel debris associated with extensionof preparations and works
 Reduce collective doses as much as practically possible
 Observe standards on exposure of the public and workers (Limitation of risks toindividuals)
Additionally, from the safety principle 8, “to ensure that the likelihood of an accident having
harmful consequences is extremely low” will be needed. To that end, taking into accountaccidents of radiation sources caused by an external or internal event or progress of an event
associated with equipment failure or operational mistake, measures will be implemented toprevent accidents and to mitigate their effects.
(2)Concept of safety functions in fuel debris retrieval operation
A. Characteristics of the Fukushima Daiichi NPS in ensuring safety
Characteristics of the Fukushima Daiichi NPS in ensuring safety are described below. Notethat risk reduction measures are currently in place for the Fukushima Daiichi NPS, and acertain stable state is maintained (see Section 4.3).
While the radioactivity inventory is reducing mainly due to decay of nuclides with short

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half-lives, nuclides with long half-lives exist in a state not contained in cladding or otherstructures. Although currently a certain level of containment functions is ensured by
PCV, attention needs to be paid to the static deterioration of structures due to aging
from the viewpoint of long-term safety ensuring.
 The decay heat has fallen as time went by to a level below 1/1000 of that immediately
after stopping the reactor. Additionally, in order to suppress escalation of chemical
reactions (e.g., oxidation) of fuel debris, elusion and spreading of radioactive materials,
and thermal effects to structures, cooling functions are maintained and alternative water
injection methods are prepared in case of their loss. The reduction in the decay heatgives enough time to switch to an alternative water injection method, presenting
adequate reliability of cooling function.
 Currently, the state of sub-criticality is maintained. The possibility of reaching recriticality is considered low because there was a change in the form of fuel assemblies
that were likely to cause criticality during the core melt down with impurities expected
to be mixed in core structures and because it is estimated that fuel debris is not retainedin the reactor core part but instead is scattered widely. Additionally, since information on
the state inside PCV and fuel debris has a great level of uncertainty, measures to detectand stop criticality are prepared for the chance of re-criticality.
B. Concept of safety functions
In addition to A. above, giving thought to the characteristics, assumed risks, etc., of fuel debris
retrieval operation, the concept of safety functions was summarized as below. In the future,
activities to reflect information obtained by internal investigations and discussions with relevant
institutions will be required.
Features to ensure the containment function (release prevention and management) andmeasures to prevent loss of the features are necessary for limiting the release ofradioactive materials.
Decay heat removal and criticality control, as well as prevention of pulverization andspread of fuel debris to the extent possible, should be considered to prevent an increasein the concentration of radioactive materials or the dose rates in the PCV.

Additionally, considerations need to be paid to the items below during the stage of planningspecific actions.
It is necessary to observe the dose limits for workers under normal operating conditions,
to reduce the exposure of workers to the extent possible, and to observe the emergency
dose limits for workers engaged in emergency response activities.
 Preparedness for normal operations as well as for anticipated external events, such as
earthquakes and tsunamis, and internal events, such as failures and operator errors, isrequired.
(3)Concept of realizing safety functionStudies towards fuel debris retrieval shall include (a) considerations on a method to materialize
safety functions, (b) establishment of a system (including operation management, not just facility
and equipment) based on the basic concept of the selected retrieval method, and (c) safetyassessment for confirming the appropriateness of design. Practically, (a) to (c) above need to berepeated taking into account limitations in available personal protective measures due to worksite
restrictions, and initiatives to raise the overall level of safety must be implemented by reviewing
the retrieval method based on information as obtained by carrying out retrieval work.
<Worksite restrictions>

It is difficult to check the actual situation of the site, and a great level of uncertainty
remains on the situation inside PCV or effects of core melt accidents on the deteriorationof internals. Therefore, taking all actions based on conservative assumptions may leadto extension of work time and installation of excessive unnecessary equipment.
The radiation dose at the site is high, and repairing, installing, or maintaining equipment

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may be difficult, or exposure amounts of workers may become large.
A. Policy of applying the concept of ‘defense in depth’
The Principle 8 of IAEA Fundamental Safety Principles specifies ‘defense in depth’ as a means
to prevent and mitigate effects of accidents, and its objective is to ensure that:- no single technical, human or organizational failure could lead to harmful effects; and,- the combinations of failures that could give rise to significant harmful effects are of very
low probability.
Additionally, the “Basic Concept of Nuclear Safety (Volume I. Annex. Concept of Defense in
Depth)” (Standards Committee, Atomic Energy Society of Japan) describes that:- the concept of ‘defense in depth’ is preparation for uncertainties, and essential in taking
advance measures to ensure nuclear safety, under the notion that unexpected events still
exist; and,- while specific measures may vary for each nuclear facility, the concept of ‘defense in depth’is universal for ensuring nuclear safety.
Therefore, the basic idea is to adopt the concept of ‘defense in depth’ as a means to prevent
serious harmful effects. However, in order to realize fuel debris retrieval, a policy of applying
the concept needs to be established in accordance with risks specific to the Fukushima Daiichi
NPS and workplace restrictions.
Layers of protection level and the necessity for independence, in particular, need to be studied
carefully from the viewpoint of early realization of fuel debris retrieval and ensuring safety
during work. For example, taking into account the progression speed of each event, scale ofevent, or workplace restrictions, it is possible that prioritizing anomaly spread prevention (e.g.,use of alternative means, restoration of normal systems) over anomaly occurrence prevention(e.g., conservative designing, equipment quality assurance activities) is more effective in
reducing overall risks.
B. Concept of initiatives from studies on functional requirements to system establishment
Studies on system establishment must incorporate flexible concepts like below, aiming at
achieving both early realization of fuel debris retrieval and materializing safety functions aswell as reducing exposure of workers.
Considering that the reactors after the accident are decommissioned, it is important toeffectively use existing equipment, to provide the necessary equipment and to consider
the combination of equipment and work management.
 In order to reduce the exposure of workers engaged in preparatory work andmaintenance work, it is important to ensure the containment function by maintaining the
PCV at a negative pressure with the combination of passive structures and active
equipment and to consider agile action using both permanent and mobile equipment.
 It is important to recognize the presence of uncertainty in the conditions in the PCV andto be able to change plans in a flexible manner if differences from the assumptions
made in the planning stage are found.
C. Concept of safety assessment
It becomes necessary to assess the safety for normal work and accidents depending on the
progress of studies on system establishment, check whether safety functions can bematerialized, and to feed the results back to studies on system establishment. In so doing, it
is important to specify realistic control targets and assessment conditions, such as setting therepresentative person for dose evaluation taking into account the conditions of the environment
surrounding the site.

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4.3 Current states of each unit
On studying fuel debris methods, it is important to understand the condition of inside of the PCV
including fuel debris, radiation environment of reactor buildings, and the damage level of buildings.
In this Section, the current stable state of is summarized for each unit. Subsequently, the results ofinvestigation, analysis and evaluation conducted to now are summarized on the situation inside
PCV (including fuel debris) and then on the radiation environment of reactor buildings and damagelevel of building.
4.3.1 Maintenance and management of a stable state at each unit
Technical developments required for practical application of the fuel debris retrieval methods, suchas the development of equipment and facilities for remote decontamination, investigation, and work
are being advanced. It is, therefore, important to maintain, control and monitor the condition of plant,
fuel debris and FP such as Cs in ensuring safety until fuel debris retrieval starts.
The current conditions of Units 1-3 can be estimated to maintain the stable cold shutdown statebased on the plant data of the inside of the PCV, such as level of radiation, the temperature,hydrogen concentration, pressure and concentration of radioactive materials recorded since theaccident. Appendix 4.1 shows its details.
The situation of information is required to be understood from the perspective of maintaining andmanagement of safety state. The status of the information obtained to date is as follows:(1)Criticality control
A. Monitoring of signs of criticality by monitoring FP with short half-lives
Xe1-35 concentration, which is a short half-life FP, is continuously monitored by the gas
radiation line monitor installed in the PCV gas control system of each Unit. While the criticality
criterion has been set to 1 Bq/cm3 9, and no sign of criticality has been shown.
B. Boric acid solution injection equipment
Considering the a large uncertainty although the possibility of reaching criticality is evaluatedlow, a boric acid water injection system has been installed in order to make it subcritical or
prevent criticality for the case when the fuel debris reaches re-criticality or there is a possibilityof re-criticality. Two tanks of boric acid water has been installed (one for spare), and alkalescentsodium pentaborate solution which has little effect on structures will be injected through thereactor coolant injection system. This system can achieve the boron concentration of 510ppm,
which is equivalent to the reactivity of more than 5%.k. In addition, when boric acid is
exhausted up, the sea water having the reactivity reduction effect of approx. 3%.k will be
injected. The time required from the occurrence of criticality to the completion of injection takes
usually 6 hours and 22 hours at the longest even assuming the equipment is damaged. Other
measures taken include installation of a temporary pool, installation of heaters to prevent
deposition of boric acid crystals, and changing from pressure hoses to polyethylene pipes10.
C. Assessment of effects during re-criticality99TEPCO, Implementation Plan of the Measures for the Specified Reactor Facilities at Fukushima
Daiichi Nuclear Power Station, December 201210
TEPCO, Implementation Plan on Reliability Improvement Measures at Fukushima Daiichi Nuclear
Power Station, May 2012

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The assessment of the impact during re-criticality was performed conservatively to evaluate
the radiation dose at the boundary of the site assuming that the critical state with the output
level, equivalent to 100 times the criticality criterion for Xe-135 concentration continues for one
day. It indicated that the evaluated exposure dose at the site boundary was 2.4 x 10-2 mSv andit does not have significant impact.
(2)Cooling
A. Cooling by circulated water injection
The TEPCO is carrying out the maintenance and management of the equipment for coolingfuel debris as well as continuous monitoring of the parameters including the temperatures of
the reactors. The buffer tanks which are the water source of the circulating injection coolingfacility that cools down the fuel debris was replaced with condensate storage tank (hereinafterreferred to as "CST") on July 201311 and the operation of the CST reactor coolant injection
system has been started. By using CST around buildings, the water circulation loop was
shortened, amount of water stored at source was increased, and the seismic resistance ofwater sources was improved. Additionally, reliability improving measures are taken, such as
replacement from pressure hoses to polyethylene pipes and attaching insulation materials toprevent freezing10,12. Further, to reduce the risk of contaminated water being released outside
the circulation loop system, the desalination system (RO), which is one of the circulating
injection cooling facilities, was installed and the construction for smaller circulation loop in theT/B of Unit 4 was conducted. The circulation loop (outdoor transfer pipes) shortened fromapprox. 3 km to approx. 0.8km (approx. 2.1km if transfer line for stagnant water included) has
been operated since October 201613.
B. Reduction in the amount of injected water and temperature change
The parameters including the temperatures of the reactors indicated stable value which was
lower than that immediately after the accident. Also the temperatures in the RPV and PCV
have been declined due to the continuous cooling and decrease in decay heat, a stable cold
shutdown state are maintained for the reactors. Additionally, to secure surplus capacity ofcontaminated water treatment facilities, since December 2016, a step-wise reduction of thevolume of injected water from 4.5 m3/h to 3.0 m3/h was conducted at each Unit. No remarkableincrease in the temperature was observed for each Unit, and no anomaly was found in thecooling conditions14.
C. Risk assessment of anomaly in the reactor water injection system
according to the evaluation of the reactor coolant injection system in the event of abnormality,
even in the case where the amount of radioactive materials of three plants are assumed to bereleased due to an event equivalent to a severe accident greatly exceeding an expected level(water injection shutdown for 12 hours), the exposure doses per year are approx. 6.3 x 10-5
mSv at the site boundary, approx. 1.1 x 10-5 mSv at a 5 km point from the Specified Nuclear
Facility, and approx. 3.6 x 10-6 mSv at a 10 km point from the Specified Nuclear Facility and
11

TEPCO, Starting Operation of CST Rector Water Injection Systems in Reactor Water injection
Systems, June 27, 201312
TEPCO, Initiatives for Improving the Reliability of Fukushima Daiichi Nuclear Power Station, April 25,
2013
13 TEPCO Holdings, Inc., FY2016 2nd Quarter Nuclear Safety Reform Plan <Including Progress of
Safety Measures at Each Power Station>, November 2, 201614
TEPCO Holding, Inc., Reducing Reactor Injection Water Amount for Fukushima Daiichi Nuclear
Power Station Units 1-3, May 22, 20174-11

therefore, it is considered that there will not be a significant risk of exposure to the generalpublic in the vicinity9.
(3)Containment
A. Prevention of leakage of radioactive materials from the PCV gas phase partThe radioactive materials released to the environment are reduced by extracting and filtering
the gas in the PCV of Units 1-3 using the PCV gas control systems and by monitoring theconcentration and amount of radioactive materials using radiation monitoring and control
equipment. Also, since the pressures in the PCV of Units 1-3 is slightly positive, 0.2-6.0 kPa
above the atmospheric pressure (1 atm = 101-3 kPa), and no serious damage is considered
to have been caused in the gas phase of the PCV.
B. Prevention of leakage of contaminated water (liquid phase) from the R/B
The contaminated water leaking from the PCV of each Unit is accumulated such as in the R/B.
The water level gauges are installed to monitor the conditions of contaminated water
accumulated in the building and other facilities so that the contaminated water does not leakand the level of the accumulated water is controlled so as to be kept lower than theunderground water level. In addition, the underground water level is being checked by thewater level gauges installed in appropriate sub-drains in the vicinity of the building. To improve
the reliability of sub-drains, suspended matters are removed from sub-drain pits, and new pits
are installed for places where restoration of existing pits is difficult12.
C. Hydrogen explosion prevention
The nitrogen filling to RPV/PCV is continued. The amount of injecting nitrogen gas is controlled
when injecting nitrogen so that the hydrogen concentration in the PCV of each Unit does notexceed the burning limit concentration (4%) and the hydrogen concentration are monitored.
Hydrogen concentration in the PCV indicates that a certain value and its concentration is being
controlled at levels lower than 0.4%.Additionally, multiple nitrogen gas separating apparatuses
are prepared, and installation of mutual supply lines, switching to spare apparatus, andpreparation of consumables are conducted in case of unforeseeable events10.
With respect to the fuel debris, the circulation cooling system has been installed, the measures
are also taken to improve reliability, such as installation of multiple of equipment and the stablecooling condition is being maintained. It is important to continuously maintain and manage thestable condition from the safety perspective.
Also, it should be noted that maintaining the stable condition until the commencement of fueldebris retrieval affects ensuring safety during the fuel debris retrieval and has continuity.
Therefore, maintaining and improving the functions to control and manage the radioactive
material release, to cool a reactor, to prevent criticality and to prevent hydrogen explosion, thecooling of fuel debris and the concentration and amount of radioactive materials are being
monitored. In addition, it is important for the monitoring to create databases of plant data so as
to effectively manage the plant information, such as temperatures of RPV/PCV and injection flowrate of cooling water.
4.3.2 Understanding and estimating reactor conditions for each unit
(1)How to proceed with understanding each of the reactor conditions
The internal situations below need to be grasped to conduct studies towards fuel debris retrieval.

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



Information on the distribution of fuel debris (approximate position and quantity on
studying the policy of fuel debris retrieval)
Information for confirming that it is possible to access fuel debris, and information on
judging the practicality of removing obstacles of accessing, if needed
Information for confirming that it is possible to safely carry out retrieval work without
risks of surrounding structures or fuel debris falling, and information on judging thepracticality of conducting preliminary operations for ensuring safety, if needed

It is ideal to obtain such information in detail and for a broad range through internal investigation,
but it is technologically and temporally difficult to obtain all information from internal investigationconsidering the harsh work environment.
Therefore, it is crucial to comprehensively analyze and evaluate necessary information so that themost likely information can be obtained through full utilization of internal investigation as well as
evaluation based on results of severe accident progression analysis and plant parameters, withsetting priorities taking into account the time to need the information, accuracy, and importance. A
schematic description of how to proceed with comprehensive analysis and assessment of reactor
conditions is shown in Figure 4.3-1.
The basic policy of the comprehensive analysis and evaluation above is shown below.
Gathering, analyzing and evaluating information for understanding the reactors
conditions are to be conducted taking into account the necessity of information and withsetting priority.
In comprehensively analyzing and assessing reactor conditions, efforts are to be madeon effectively utilizing obtained information and obtaining most likely results.
Taking into account the balance between “labor, time, exposure, and cost” required for
obtaining information and “safety measures, retrieval device and equipment designing,and cost” required for the retrieval method, efforts are to be made on obtaining
information that is as valuable as possible, with paying attention to the amount of timeand cost allowed to be spent.

Figure 4.3-1 How to proceed with comprehensive analysis and assessment of reactor conditions

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As studies necessary for understanding the reactor conditions, comprehensive analysis andevaluation are to be conducted through (1) estimate by internal investigation, (2) estimate byanalysis, and (3) estimate by knowledge and experiments. These information obtaining methods
have their own characteristics, and information gathering and analysis and assessment shall be
conducted with fully utilizing their characteristics. The characteristics are shown in Table 4.3-1.
Table 4.3-1 Characteristics of information obtaining methods pertaining to understanding thereactor conditions(1)
Estimate byinternal
investigation(2)
Estimate byanalysis(3)
Estimate byknowledge
andexperiments

Method (expected information)
Characteristics
Visual checking and measuring Since actual measurements are conducted, theby PCV internal investigation
reliability of information for the obtained section(Check distribution of fuel debris,
is higher than other estimate methods.
structures, etc.)
 Scope of information obtained differs inaccordance with the scale of the investigation.
Visual checking and measuring
 Establishment of access routes to fuel debris
by RPV internal investigation
inside RPV and development of investigation(Same as above)
devices are highly difficult, and technology
development is in progress, and it takes time
until starting the actual work.
Muon measurement
 Possible to grasp approximate situation of fuel(Check presence of fuel debris
debris inside RPV.
inside RPV)
 Impossible to judge presence of fuel debris thatis smaller than the resolution.
 Requires several months of measuring period.
Analysis by severe accident Possible to estimate the condition of fuel debris
progression analysis codes, andin a relatively short time.
reduction of uncertainty of Suitable for estimating the unknown internal
reactor conditions by sensitivity
behavior that could not be measures at theanalysis and inverse calculation
time of accident, and for understanding the(Position, quantity, PF
general picture.
distribution of fuel debris, and Difficult to estimate local and special
reduce their uncertainties)
conditions.
 Contains uncertainty, and may cause
differences in results depending on theconditions and the model.
Creation of simulated debris, and  Since actual measurements are conducted, thepast findings
reliability of obtained information is high.
(Mechanical properties and Depends on the simulated debris conditions
chemical properties of fuelspecified.
debris)
Evaluation of heat source (fuel Bases on measured values, and has little
debris) by the heat balance
reliance on models etc. Suitable formethod, deduction from plantunderstanding the general picture.
parameters Though Low in quantitativity, allows judgement(Checking trend of fuel debrisof the presence of fuel debris.
distribution)(2)Current survey status
For the three information obtaining methods above, the state of investigation, analysis andassessment are summarized below.
A. State of investigation by internal investigation (see Appendix 4.2-4.4 for details)
1) PCV internal investigation
a. Unit 1 PCV internal investigation

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In April 2015, an investigation device was inserted from the PCV X-100B penetration, andinvestigation on the 1st floor grating using a CCD camera and dose measurement wereconducted. Additionally, in March 2017, a self-propelled investigation device was inserted
from the PCV X-100B penetration, and investigation for the situation of PCV bottom wasconducted by hanging a CCD camera and a dosimeter from the 1st floor grating outside
the pedestal.







No large scale damage on the existing facilities was observed (e.g., PLR pump, wallinside the PCV, ventilation unit (HVH) etc.).
It was confirmed that the PLR piping shielding units have fallen
Sediments were observed at the bottom of PCV, piping, etc. The height of the sedimentsurface differed depending on the measuring point.
Images were taken with approaching the sediments, but no whirling up of sediments
was observed.
The dose rate on the grating was about 4-12 Sv/h.
It was confirmed at each measuring point that the underwater dose rate increased
gradually as the dosimeter was placed closer to sediment surface on the bottom of PCV.

b. Unit 2 PCV internal investigation
In January and February 2017, investigation was conducted by inserting a guide pipe witha CCD camera attached from the X-6 penetration and by inserting a self-propelled
investigation device from the X-6 penetration to the pedestal via the control rod drive
mechanism (hereinafter “CRD”) exchange rail.










For the gratings inside the pedestal, those detached and fallen and those deformed tothe extent where the squares looked irregular were observed on the left side of the slot
opening, but none was detached on the right side. On the gratings, a large amount of
sediments was observed in addition to fallen cables and fallen objects that looked like
TIP guide pipes (see Figure 4.3-2).
No large scale damage was observed for the CRD housing support near the pedestal
entrance.
Some form of substances adhered to CRD exchanger or surrounding traversing in-coreprobes (TIP) guide tube support were observed.
No anomaly such as cracking was observed for the inner wall of pedestal at the pedestal
platform section.
It was confirmed that a steam was rising from the lower part of grating.
It was confirmed that existing structures were almost all placed at assumed locations at
the upper part of the CRD exchange rail, and that there was no considerable damage.
The dose rate at the CRD exchange rail was 70 Sv/h (a place on the rail about 3 m fromthe inner surface of pedestal).

c. Unit 3 PCV internal investigation
In October 2015, an investigation device was inserted from the PCV X-53 penetration,and measurement of dose rate, investigation of PCV internals and PCV bottom using apan-tilt camera and a CCD camera, and sampling of accumulated water were conducted.
Main information obtained by the investigation is as below.


Sediments were observed on the CRD exchange rail and the 1st floor gratings outside
the pedestal.
The dose rate at the gas phase inside PCV was about 0.8-1 Sv/h.
From the results of quality analysis of accumulated water, the corrosiveness inside PCV
was low.

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Additionally, investigation of pedestal internals was conducted in July 2017 using aswimming ROV equipped with a CCD camera, inserted through the penetration of X-53,
and the following information was obtained.
In the pedestal, likely melted materials that are consolidated were confirmed.
It was confirmed that the pedestal internals were accessible through the pedestal
opening.
Inner PCV investigation found some structural damages and CRD housing bracket
fallen off in the pedestal. And there were no grating on the platform but some found at
the bottom of inside the pedestal. There were also some fallen objects and depositedmaterials at the bottom of inside the pedestal.

Figure 4.3-2 Result of PCV internal investigation for Unit 2[Source: TEPCO “Unit 2 Primary Containment Vessel Internal Investigation ~Additional
Investigation by Image Analysis~”]

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4-17

2) Muon measurementa. Unit 1 muon measurement (2015)
As a result of evaluating measurement data, it was judged that no water or fuel sized
greater than 1 m, that is the minimum resolution of transmission muon measurement used,exists at the original location core.
b. Unit 2 muon measurement (2016)
As a result of evaluating measurement data, those below were confirmed.
High-density substance exists at the RPV bottom, which is likely fuel debris.
From quantitative evaluation through comparison with simulation, a large majority offuel debris is estimated to exist at the RPV bottom. It was also indicated that a small
amount of high-density substance may exist below and around the core, which is likely
fuel.
c. Unit 3 muon measurement (2017)

Muon measurement was started in May 2017, to be conducted for several months. Current
evaluation of muon measurement data indicates any large-sized high density materials
were not identified in RPV including both reactor area and the bottom, though there might
be some fuel debris remaining there.
B. Estimate by analysis
The severe accident progression analysis is highly depending on the computational model
employed and estimation scenario and calculated results contain the uncertainties. However,quantitative information, such as the amount and composition of fuel debris and FP distribution
at several locations in the PCVs will be able to be obtained and it will be effective method tounderstand a whole situation of the severe accident. Also, estimating the temperature history
inside the reactors during the progression of the severe accident through the severe accident
progression analysis, and the results are utilized for estimating the state of main internals andequipment.
The Organization for Economic Co-operation and Development/Nuclear Energy Agency(OECD/NEA) BSAF (Benchmark Study of the Accident at the Fukushima Daiichi Nuclear
Power Station) project has been carried out as an international joint research. Eleven countries
participate in the project, and research institutions are conducting estimate of internal
conditions through severe accident progression analysis at each country. The outline andachievement of the BSAF project are shown in Appendix 4.5.
1) Analysis of the location and quantity of fuel debris
As estimation by analysis, understanding the full picture in the PCV, e.g., quantity of fueldebris at each location, is conducted using the MAAP and SAMPSON codes for severe
accident progression analysis. Addition and improvement of physical phenomenon models
deemed necessary for evaluating the behavior of fuel debris and FP at the Fukushima
Daiichi NPS have been completed by FY201515, and analysis was conducted (see Appendix
4.6 for details). In FY2016, issues of the physical phenomenon models were identified fromanalysis and comparison of analysis results with the actual accident measurement data andsite information obtained from PCV internal investigation, and analysis of severe accident
progression scenarios was conducted taking into account the characteristics of analysis
15

IRID, Completion report (Enhancing the understanding of conditions within reactors by accident
progression analysis and data from actual equipment), March 20164-18

codes. In FY2017, advancement of analysis models is to be aimed for the identified issues,and quantitative evaluation of fuel debris by analysis is planned.
2) Analysis of FP distribution
The current MAAP and SAMPSON codes have differences in FP distribution, such as
different compounds are considered for some FP nuclides. Therefore, the analysis is being
made on potentially existing chemical morphologies in the process of evaporation, relocation,and condensation of the main γ source cesium, and establishment of relevant analysis
models is conducted. Additionally, the accuracy of FP distribution assessment is improvedby expanding knowledge on cesium.
3) Estimation for the state of major structures and equipment
In order to estimate the present state of main PCV internals and equipment, estimation ofinternal equipment is conducted in FY2015 based on the results of temperature evaluation
using severe accident progression analysis codes. This estimation was performed byreferencing not only the results of analysis but also the on-site situation.I 15. The hightemperature deformation, creep rupture and corrosion degradation were considered asdegradation events induced in the structures and equipment in the scope of the evaluation.
For the steam dryer, steam separator, top guide and core support plate, the results
suggested potential creep deformation for all units that takes place when a certain level ofload is applied under a high temperature condition. However, to use these analysis results,
it must be taken into account that analysis results by severe accident progression analysis
codes contain uncertainties.
C. Estimation based on knowledge and experiments
Estimation from the knowledge and experiments can be categorized into three cases, which
are the estimation based on the past accidents and researches, engineering estimation onplant data and experiments conducted using simulated fuel debris.
1) Estimation based on knowledge from past accidents and research
The examples of core meltdown accident include the accident at TMI-2 and the Chernobyl
Nuclear Power Plant Unit 4. The findings to be obtained will be utilized in the estimation of
the behavior of fuel debris inside the RPV and estimation by Molten Core ConcreteInteraction (MCCI). The researches performed in the past include the FP tests on migration
behavior performed at the Phebus reactor in France and MCCI test at the U.S. Argonne
National Laboratory. Those results were reflected to the models of the severe accident
analysis code.
2) Engineering estimation based on the plant dataa. Estimation by heat balance method
The ratio of fuel debris in the RPV and PCV was estimated based on the heat balance
assuming that the temperature of the cooling water injected to the RPV is elevated totemperature of the accumulated water by the heat source (fuel debris) inside the RPV/PCV.
That is the balance of the heat input (heat capacity of injected water and decay heat) andheat emittance (heat emittance from the PCV wall surface to the building or to theatmosphere and temperature elevation of cooling water by the fuel debris). As a result, for
Unit 1, the validity of severe accident progression analysis results was confirmed. For
Units 2 and 3, it was indicated that a certain amount of fuel debris may exit inside RPV.
The overview of the heat balance method and the estimate results are shown in Appendix
4.7.

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b. Estimation based on the trend of plant parameters
The heat source (fuel debris) in the RPV was estimated based on the trend of thetemperature around the RPV in the post-accident condition, water temperature of
suppression chamber (S/C), amount of injected water via feedwater (FDW) system andreactor core spray (CS) system. The results of this estimate indicated few heat sources in
the RPV for Unit 1 and a certain amount of heat source in the RPV for Units 2 and 3. The
method of estimate and results are shown in Appendix 4.8.

3) Estimate of fuel debris properties based on experiments using simulated debris, etc.
Studies on the methods of fuel debris retrieval, housing, and storage require data on theproperties of fuel debris existing inside reactors. However, currently, it is difficult to sample
them. Therefore, the morphology (appearance, form) and characteristics (e.g., mechanicalproperties, thermal properties, chemical properties, physical properties) of fuel debris are
estimated based on the results of experiments and analysis using simulated fuel debris likebelow, in addition to the knowledge that have been obtained to date.
a. Characterization of the metal phase of fuel debris
Measurement is conducted on the mechanical properties, etc., of substances indicated to
be included in the metal phase of fuel debris, such as Zr(O) where oxygen is dissolved
into zirconium.
b. Characterization of products of reactions specific to the Fukushima Daiichi NPS accident
Data are obtained on the formation layers, mechanical properties, etc. of simulated fueldebris into which fuel, stainless steel oxides, FP elements, and seawater salt components
are dissolved.
c. Understanding the fuel debris characteristics that affect housing and storage
Using the composition, internal structure and drying conditions of fuel debris asparameters, analysis is conducted on the drying behavior of fuel debris and its powderingand oxidation behavior during a drying process.
d. Characterization pertaining to unevenness of properties
In order to assess the unevenness of large lumps, property evaluation experiments on themechanical properties of MCCI products is conducted at the Alternative Energies andAtomic Energy Commission (CEA) in France. Additionally, at the National Nuclear Centerof Kazakhstan, property evaluation experiments are conducted for large solidified molten
metal ceramics, and property data are obtained for the granularity, density, structure, etc.,
of powder fuel debris and its aggregated solids.
The results of property evaluations above using simulated debris and the knowledge obtained todate are being summarized into a list of fuel debris properties. Specifically, for each location offuel debris estimated to exist inside RPV/PCV, its macroscopic properties (e.g., compression
strength, uranium content) and microscopic properties (e.g., mechanical properties, thermal
properties) are estimated. An example of the current property list is shown in Appendix 4.9. The
property list will be continually updated associated with initiatives for comprehensive analysis
and assessment of reactor conditions, in accordance with needs for information on the propertiesof fuel debris that are necessary for studying the method of fuel debris retrieval.

4-20Additionally, along with the estimate of properties of fuel debris based on experiments using
simulated debris etc., various studies are conducted in preparation of sampling of fuel debris,
including selection of analysis items and flow of analysis, development of analysis technologies
such as fuel debris dissolution method and chemical morphology analysis method, and selectionof transport containers necessary for transporting high dose samples. An international joint study
on analysis of fuel debris in preparation of sampling, OECD/NEA PreADES (Preparatory Studyon Analysis of Fuel Debris) project is under preparation towards implementation, and overseas
knowledge and findings like those summarized by the PreADES project will be utilized as needed.
(3)Comprehensive analysis and assessment of PCV inside conditions
In additional to actual measurement values such as plant parameters obtained after the accident,
information on reactor conditions has been accumulated by severe accident progression analysis,
PCV internal investigation and muon measurement, and scientific findings are expanding fromtests and experiments. The results of comprehensive analysis and assessment on the state offuel debris based on such information are shown in Table 4.3-2. Although the results are relatively
qualitative since the current activities focus on addressing the issues of severe accident
progression analysis that have been identified through analysis and comparison of the results ofsevere accident progression analysis with the site information etc. obtained from PCV internal
investigation, information necessary for studying fuel debris retrieval policies is covered. The
reactor situations (e.g., fuel debris distribution) estimated for each unit summarized by thecomprehensive analysis and assessment are shown in Appendix 4.10.
Information on the distribution of fuel debris as well as information on access route to fuel debris
and conditions of surrounding structures that is necessary for studying fuel debris retrieval
policies are shown in Table 4.3-3.
Additionally, comprehensive analysis and assessment of the distribution of FP and radiation doseare conducted for each unit based on site information obtained by PCV internal investigation etc.
and knowledge and findings obtained to date.
Information on fuel debris etc. is summarized below as the premise of studies on policy of fueldebris retrieval method.
A. Unit 1
 Distribution of fuel debris As shown in Table 4.3-2, a small amount of fuel debris is expected to exist at the RPV
bottom, but a large majority of it is considered to be located at the PCV bottom.
According to some analyses, some of it might have remained there but some might
have spread to outside the pedestal through the access entrance for workers.
 According to tests and analysis using simulated fuel debris, some of fuel debris
located at the PCV bottom may have reacted with concrete and formed MCCI
products.
 Access to fuel debris for retrieval
 By an investigation with a small robot on gratings outside the pedestal conducted in2015 using an investigation device, it was confirmed that the PCV bottom isaccessible from the upper side of gratings.
 Conditions of surrounding structures
 By a PCV internal investigation conducted in 2015, no major damage was observed
for the walls outside the pedestal above gratings, as far as the obtained images
indicate.
 In a PCV internal investigation conducted in March 2017, no major damage was

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observed for structures, as far as the obtained images indicate.
B. Unit 2
 Distribution of fuel debris As shown in Table 4.3-2, a large amount of fuel debris is expected to exist at the RPV
bottom, and a small amount exists at the PCV bottom. According to the results ofmuon measurements, most fuel debris is located inside RPV.
 Access to fuel debris for retrieval
 Through a PCV internal investigation with a small robot conducted in 2017, while
sediments were found on the CRD exchange rail, it was confirmed that access could
be made up to the place near the opening of the pedestal.
 Conditions of surrounding structures As a result of a PCV internal investigation conducted in 2017, though a part of thegratings had fallen, no falling of large structures such as CRD housing was observed
inside the pedestal, as far as the obtained images indicate. Additionally, in thisinvestigation, no anomaly such as cracking was observed on the internal walls of thepedestal at the pedestal platform part.
C. Unit 3
 Distribution of fuel debris As shown in Table 4.3-2, while parts of the fuel debris are likely to remain at the RPV
bottom, more fuel debris are estimated to exist at the PCV bottom when comparedwith those in Unit 2.
 The investigation on the PCV internals conducted in July 2017 confirmed likely meltedmaterials that are consolidated in the pedestal. Further, evaluation of muon
measurement hasn’t confirmed any existence of large-sized high-density object in the
RPV.
 Access to fuel debris for retrieval
 By means of the investigation with a small robot on the PCV internals conducted inJuly 2017, it was confirmed that the pedestal internals were accessible through thepedestal opening.
 Conditions of surrounding structures
 The investigation on the PCV internals conducted in July 2017 confirmed that there
were some structural damages and CRD housing bracket fallen off in the pedestal.
And there were no grating on the platform but some found at the bottom of inside thepedestal. There were also some fallen objects and deposited materials at the bottom
of inside the pedestal.

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4-23

Table 4.3-2

Estimation of the distribution of fuel debris in Units 1 - 3

4-24

Table 4.3-3 Access Route to Fuel Debris and Information on Conditions of Surrounding Structures for Units 1, 2 and 3

4.3.3 Conditions of reactor buildings
Among site conditions of the Fukushima Daiichi NPS, the radiation environment and the damaged
conditions of buildings that form the premise of considering technical requirements pertaining tofuel debris retrieval methods are summarized. The results are shown in Table 4.3-4.
(1) Unit 1
With respect to radiation environment, reduction of radiation dose was progressed bydecontamination etc. mainly for the 1st floor, but some parts still have high radiation doses.
Reduction of radiation dose will be conducted in the future as necessary. The area around X-6
penetration utilized for an internal survey of Unit 2 has a high radiation dose due to venting
operation in the accident, serving as an obstacle for conducting similar investigations. The walls
and roof at the upper section of the operation floor of reactor building are destroyed by hydrogen
explosion. Also, the PCV is leaking cooling water, but the location of apertures has not beenclarified yet. The containment function of PCV gas phase portion is judged to be reduced to someextent, but the micro-positive pressure is maintained by nitrogen injection.
(2) Unit 2With respect to radiation environment: The air dose rate at the 1st floor was reduced to an
average of 5 mSv/h by decontamination etc. Damage of reactor building by hydrogen explosion
was avoided, yet inside the building is contaminated. Also, leaking of cooling water is suspected
for some places at the PCV, but the state of cracks or location of apertures have not been clarified
yet. The containment function of the PCV gas phase portion is judged to be reduced to someextent, but the micro-positive pressure is maintained by nitrogen injection.
(3) Unit 3
With respect to radiation environment: Reduction of radiation dose was progressed bydecontamination etc. mainly for the 1st floor, but the overall radiation dose is still high. Reductionof radiation dose will be conducted in the future. Damage of reactor building was caused to thewalls and roof of the upper section of the operation floor as well as to the lower section of theoperation floor due to hydrogen explosion. While the containment function of the PCV gas phase
portion is judged to have been reduced compared to Unit 1 and Unit 2, the micro-positive
pressure is maintained by nitrogen injection.

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(1) Dose rate at
and aroundreactorbuilding(Air dose rate
inside reactor
buildings areshown inFigure 4.5-10.)(2) Damages(Damage toreactorbuilding, PCV,etc.)

Table 4.3-4 Conditions of Reactor Buildings
Unit 1
Unit 2
 Dose rates were Dose rates werereduced to an average
reduced to an averageof approx. 5 mSv/h forof approx. 2 mSv/h forthe 1st floor.
the 1st floor northwest
 Some high dose areas
and west areas.
remain for higher parts
 Dose reduction is notof the 1st floor and forprogressed for the 1st
the 2nd floor andfloor south area due toabove.
significant effects by Dose rates are high at
high dose of AC piping
stack and SGTS
and DHC equipment.
piping around the Some high dose areas
reactor building (northremain for higher partsside).
of the 1st floor and forthe 2nd floor and above.  A place with a dose
rate of approx. 900
 Dose rate is high at
mSV/h was measured
stack and SGTS piping
for the operation floor.
around the reactor
building (south side).
 As data while removing
rubbles at the operation
floor, a dose rate of
approx. 500 mSv/h wasmeasured.
 No hydrogen There is no upper part
explosion occurred,of reactor building
and the reactor
operation floor due tobuilding is relativelyhydrogen explosion.
wholesome.
 Rubbles are scattered Positions of cooling(removal in progress).
water leakage from Positions of coolingthe lower part of thewater leakage from thePCV to the torus room
lower part of the PCV to- S/C (or connecting
the torus room
piping)- Sand cushion drainpipe- Vacuum break line
bellows
 PCV water level: PCV water level:
Approx. 0.3 m
Approx. 2 m
 S/C water level: Nearby  S/C water level: Upper
end of downcomerfull PCV pressure (nitrogen  PCV pressure(nitrogen injected):
injected): Approx. 1
Approx. 5 kPag
kPag

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Unit 3
 Air dose rates werereduced to an averageof approx. 9 mSv/h forthe 1st floor northwest
and west areas, andto approx. 7 mSv/h forthe southeast area.
 Dose rates are high at
the 1st floor southwest
area with an averageof approx. 19 mSv/h.
 Some high dose areas
remain for higher partsof the 1st floor and forthe 2nd floor andabove.
 Does rates of theoperation floor werereduced to an averageof approx. 2 mSv/h bydecontamination andshielding.
 There is no upper partof reactor building
operation floor due tohydrogen explosion,and parts of floors
have broken andfallen (northwest part).
 Rubbles have been
removed.
 Positions of coolingwater leakage fromthe lower part of the
PCV to the torus room
- MS piping D bellows
 PCV water level:
Approx. 6 m
 S/C water level: Full PCV pressure(nitrogen injected):
Approx. 0.2-0.3 kPag

4.3.4 Understanding the condition in the PCV by actual investigation
It is important to continuously conduct actual investigations and to obtain detailed information on
PCV inside to perform retrieval of fuel debris. It is also important to carry out fuel debris samplingat appropriate times, in order to confirm the properties etc., and to contribute to the retrieval of fueldebris.
Information that should be acquired through the detailed investigation of PCV internals includes thefollowing:Distribution of fuel debris inside and outside of the pedestal
Information for confirming the access route to fuel debris located on the bottom of PCVInformation to contribute to the judgment of safety of fuel debris retrieval work

In addition, for Units where a large amount of fuel debris may exist on the bottom of PCV,
information to understand the degree of contact between fuel debris and the pedestal is also
important.
RPV internal investigation is required to be conducted after examining the appropriate time basedon the evaluation of the detailed investigation results on the PCV inside. For example, it is estimatedthat the reactor internals of Units 1 and 3 are considerably damaged, and therefore it may benecessary to confirm the situation in RPV in order to determine the safety of fuel debris retrieval
work.
The following are the points to note when a detailed investigation of the PCV internals is conducted:
The PCV inside information that are acquired by actual investigation should be
immediately reflected in the development plan of investigation devices and investigationplans with cost rationality consideration.
Arrangement with the surrounding field work schedule should be done by appropriate
coordination to resolve any interference in order to carry out the establishment of accessroute and actual investigation.
In Unit 3, as the water level will need to be lowered for the detailed investigation on the
PCV inside, the feasibility and the time for lowering the water level in PCV need to be
examined.

Similarly, for investigation on the RPV internals:


In both cases of investigation through the top hole drilling method and side hole drilling
method, the investigation time should be coordinated appropriately in order to resolve
any interference with the fieldwork involved in unloading the fuels from inside the pools
around the site in question.

In actual investigations, it is important to contribute to the rationalization of designing retrieval
devices and retrieval plans through continuous acquisition of detailed information on the distribution
and forms of fuel debris (three dimensional information on the distribution of fuel debris, informationon the form of fuel debris, i.e. whether the debris is in granular or massive form etc.) and theenvironmental data in the reactor (temperature, dose rate, neutron flux etc.). Additionally, purchaseof general-purpose items and technology development should be forwarded as needed for thedevices for such investigations.

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4.4 Effects of risk reduction by fuel debris retrieval
Fuel debris presents substantially large uncertainties in its behavior, due to various reasons suchas the difficulties in confirming its form and containment status and potential change over time.
Additionally, as described in Section 3.2, the risk level of fuel debris continues to be relatively highamong the risk sources residing in the Fukushima Daiichi NPS. Therefore, fuel debris is planned
to be dealt safely, assuredly and carefully, with ensuring the safety of operations through wellthought-out preparations and use of advanced technologies. Also, fuel debris is planned to be
retrieved from the current location as early as practicable to realize a mid- to long-term sufficiently
stable controlled state.
In this Section, instabilities specific to fuel debris are described, followed by considerations on
effectively reducing the risk level of fuel debris.

4.4.1 Potential risk of fuel debris
Fuel debris is considered to exist in various characteristics and forms, including large monolithic
solid where fuel elements and structure materials solidified after melting together, pellets fragments
fallen from broken fuel pins, rock lumps fallen and collapsed during melting, granules and powders
due to some cooling conditions during the melting process, those leaked to the bottom part to PCV,and those melted together with concrete after leaking to PCV (MCCI products). Additionally, it
potentially exists at various locations in accordance with the sequence of accident, including thoseleft inside RPV and those leaked to PCV. Further, the current control condition of decay heat, i.e.
the cooling condition, of fuel debris may vary widely, including those onto which cooling water isconstantly poured like fuel debris inside RPV and those that are submerged under the water levelof circulated cooling water like fuel debris leaked to PCV.
Like these, fuel debris existing inside RPV and PCV is considered to be in various characteristics
and forms, and all of them may have different properties. Therefore, each risk level may beconsidered different. On top of that, the instability of fuel debris and the soundness of containment
functions may change over time. Therefore, considerations need to be made on the possibility of
the risk level of fuel debris changing, affected by its diffusability, controllability (e.g., cooling), andcontainment functions, arising from these factors
Fuel debris has its own instabilities in term of criticality, radioactivity, decay heat, chemical
properties, and geometric shape. Currently, it is considered that fuel debris is controlled in a stablecondition since the sub-critical state is confirmed, decay heat is controlled (cooled) by the circulatedcooling system, hydrogen concentration is controlled by injection of nitrogen, and the measurement
values of pressure and temperature remain stable. Additionally, as time goes by after the accident,
the decay heat has substantially decreased and will continue to decrease in the future, and the risk
is decreasing as for fuel debris re-melting due to decay heat, associated re-release of residual FP,
and increased stress to RPV, PCV and other structures due to a heat and increased internalpressure.
However, in a mid- to long-term, the state of fuel debris may change over time, such as radioactive
materials leaching from fuel debris as colloid or ions, and becoming granular or fragmented due tooxidation or collapsing. If the volume of fuel debris in highly mobile forms increases due to leaching
and/or granulation, the risk may rise for such form of fuel debris to flow into a circulated coolingsystem along with coolants, or even released into the environment along with the flow of gas or

4-28

coolant if a major loss of containment functions occurs. Additionally, if granular and fragmented fueldebris with high mobility had high concentration of nuclear fuel materials, there will be an
undeniable possibility of it accumulated at one place and causing local criticality16. On top of that,radioactivity of fuel debris is the cause of hydrogen generation by radiolysis of water. When ahydrogen concentration reaches a certain value, it gives rise to the possibility of hydrogen explosion.
While currently the hydrogen concentration is constantly controlled by injection of nitrogen, in a
mid- and long-term, the risk of losing control increases due to various factors such as deteriorationand failure of equipment.
Fuel debris remaining inside RPV is considered mainly solid solutions of uranium and zirconiumoxides where fuel elements were melted together with cladding and/or with structure material, andtheir instability arising from their chemical properties (e.g., change in composition) is considered
low17. When it comes to MCCI products where fuel debris was leaked to PCV and reacted withconcrete, meanwhile, attention needs to be paid on the fact that the uncertainties of theircharacteristics and chemical properties are relatively high since the number of case where MCCI
products was generated, number of simulated samples, and results of observations and analysis
are limited worldwide.
Reactor buildings and vessels such as RPV and PCV have been affected by the core melt accident
and hydrogen explosion, affected by salinity due to seawater injection, and exposed to hightemperature and high-pressure conditions at the time of the core melt accident. Therefore,
corrosion or deterioration of vessels may progress over time. This means that there remains apossibility of the reliability of the containment functions gradually reducing over time. Additionally,
while RPV/PCV are made of carbon steel and allows a certain level of heat dissipation, the PCV
bottom is concrete, and the route of heat dissipation for fuel debris fallen to the PCV bottom iscoolant retained at the PCV bottom or the lower part of PCV container, and heat is likely
accumulated there. On top of that, concrete of the pedestal part or the PCV bottom may deteriorate
due to long-term exposure to decay heat, and the uncertainty of containment functions for the PCV
bottom is considered relatively large.
Like above, the state and instability of fuel debris have not been fully investigated and still havesome high uncertainties. Additionally, over time, the diffusability of fuel debris may increase due toits instability and the reliability of containment functions may reduce, and the risk level of fuel debris
may increase with time accordingly.

4.4.2 Risk level of fuel debris
As described in the previous section, fuel debris is in various forms, and under various coolingconditions and containment state. In this section, the form and abundance ratio of fuel debris are
estimated 18 , and the risk level of fuel debris was assessed using SED developed by NDA as
described in Section 3.2, taking into account cooling conditions and the state of containment.

16

If fuel debris has high concentration of nuclear fuel materials, has low concentration of neutron
absorbers, and if the mixed ratio with reflector substances or coolants that serve as moderators
accidentally reaches to a system that causes criticality, local and transient criticality events may occur.
17‘FY2014 supplementary budget “Subsidy for Decommissioning and Contaminated Water Treatment
Project, Study on the property of fuel debris (progress report)”’ IRID, April 201618
On assessing risk levels of fuel debris, values derived from results of severe accident progression
analysis, etc., are used for the distribution of fuel debris, form of fuel debris, and abundance ratio of
the forms.

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As a result, as shown in Table 4.4-1, it was found that the risk level of fuel debris largely depend on
the abundance ratio of forms that are highly mobile and easily taken into human body, such as
particulates, slugs, and aerosols, and also affected by the controllability of its specific instability(e.g., decay heat, criticality).
Additionally, the assessment indicated that uncertainty of instability is larger for fuel debris that fellto PCV (considered to be MCCI products mostly) than fuel debris inside RPV, because of being
under less redundant containment functions and limited data on properties of MCCI products, etc.
Table 4.4-1 Impact to the Risk Level of Fuel Debris
SED factor

Impact torisk level

Inventory

Small

Form factor (FF)

Large

Control factor (CF)Moderateto large

Facility descriptor(FD)

Moderate

Waste uncertainty
descriptor (WUD)

Moderate

Description (see Appendix 3)
Risk level changes linearly to the change in inventory*, andthe impact is small.
* Assuming the composition of fuel debris is homogenous.
Among properties that define FF, powders have 104-105
times higher FF than that of solids, for their high mobility
and diffusability, and high likelihood of being taken into
human body and settling in the lungs or air ducts or being
absorbed into blood. Therefore, the impact of abundance
ratio of powder fuel debris on risk level is large*.
* The risk level greatly increases even when only several %is in powder forms.
Since CF is specified in tenfold multiplications, and its
impact on the risk level will be moderate to large.
While FD relies on the state of containment, its change iscategorized in stages. Therefore, its impact on risk level ismoderate.
While WUD relies on the impact of instability specific to fueldebris on the stable storage, its change is categorized instages. Therefore, its impact on risk level is moderate.

4.4.3 Discussions on reduction of risk of fuel debris
According to the abovementioned assessment of risk level of fuel debris, discussions like below
can be derived on the reduction of risk by retrieving fuel debris.
Among fuel debris in various forms, removing and stably controlling fuel debris in highly diffusible
and mobile forms such as particulates and fragments as priority would reduce the risk level
relatively early and have high risk reduction effects.
Additionally, considering the high uncertainty in the properties and stability of MCCI products, low
redundancy in the containment functions for fuel debris that fell into PCV, and high uncertainty on
time-dependent change in the decay heat control state (cooling conditions) and containment
functions, the risk level of fuel debris that fell into PCV (MCCI products in particular) is consideredhigher than debris remaining inside RPV. Therefore, investigating fuel debris at the PCV bottom
and understanding its properties, instability, control state, and containment conditions may havereasonable significance.
For fuel debris retrieval work, attention needs to be paid on the possibility of the risk level of fueldebris changing due to external action such as suction, cutting, and crushing, as it may cause
relocation of highly mobile fuel debris, powdering or fragmenting of fuel debris, or collapsing andfalling of fuel debris. Therefore, on performing fuel debris retrieval, it is important to conduct

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necessary investigations and execute appropriate and systematic operations in accordance withthe state of fuel debris, so that its risk level does not change drastically.
In addition, as stated in Section 3.2, on optimizing the risk reduction measure (fuel debris retrievalwork), it is important to give considerations not only to risk reduction effects but also to otherfactors such as the safety of operation, feasibility of operation, time needed to realize theretrieval, and allocation of resources.

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4.5 Feasibility of fuel debris retrieval method
This section considers feasibility of the methods to safely retrieve fuel debris. First, we summarize
characteristics of three methods that should be considered intensively as indicated in the Strategicplan 2016 from the viewpoint of PCV water level and access routes to fuel debris.
Then, we compile the current status of each method to address evaluation results, issues and next
steps of the nine technical requirements defined in the logic tree in Figure 4.1-1. Finally, thefeasibility of fuel debris retrieval methods will be evaluated based on the above results.

4.5.1 Characteristics of fuel debris retrieval methods
It is assumed that fuel debris at Fukushima Daiichi NPS exists not only inside RPV but also at lower
PCV. As a method to safely retrieve such fuel debris, examining the feasibility of fully submerging
PCV (full submersion method) began shortly after the accident. This method is similar to the one
employed for the core meltdown accident at TMI-2, which occurred in the United States in 1979.
In the case of TMI-2, however, one of the primary system pumps was re-activated shortly after thereactor scram, and heat removal from melted fuel was resumed19. The integrity of the reactor vessel(RV20) was therefore maintained21, and RV could be fully submerged easily. (In fact, this condition
was maintained until fuel debris retrieval was completed.) On the other hand, the RPV itself wasdamaged in the Fukushima Daiichi NPS and accumulating water was not possible. Instead, an idea
to fill PCV with water was considered.
First of all, storing irradiated nuclear fuels in a fully water-filled condition (submerged) is thecommon means of shielding γ rays. Thus, handling fuel debris by the same method seems to berational. Underwater operations also work effectively to prevent dispersing of radioactive dust thatwould occur when cutting fuel debris.
PCV, however, is not designed to fill water like the RPV and the PCV itself was affected by heatand pressure during the Fukushima Daiichi NPS accidents, where cooling of the RPV was notavailable for the long term. In addition, because water tightness of PCV penetration was a concern,
R&D on repairing PCV has been in progress for water sealing. As technical difficulties for repairing
PCV, due to a large number of penetrations requiring repair, a possibility of vast radiation exposureof workers engaged in repair, and durability after repair, and, have been identified, we have decided
to consider methods other than the full submersion method.
(1)PCV water level
The PCV water level is one of the important parameters to determine the fuel debris retrieval
method. This is because this parameter serves as the target (how deep we fill or can fill the water
during fuel debris retrieval) and determines the method to handle fuel debris in a fully or partiallysubmerged condition.
Depending on the water level, the Strategic plan 2016 defines “a full submersion method” as amethod that fills water to just below the operation floor, and “a submersion method” as a methodthat covers fuel debris inside RPV or PCV fully with water (The leftmost and the second from the19

Joy L. Rempe, Darrel L. Knudsen, INL/CON-13-28099, ANIMMA 2013, June 2013.
A term for PWR. It is equivalent to RPV for BWR.
21
(According to NRC staff and utility’s engineers engaged in TMI-2 cleanup), the integrity of RV was
not concerned at all because there was no increase of contaminated water in the basement floor of thereactor building due to the possible overflow of tank water.
20

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left in Figure 4.5-1 show the conceptual diagram of each method). There is no qualitative differencein either of the two in terms of the radiation shielding of fuel debris and anti-dispersing/cooling ofradioactive dust. When setting the water level lower than that for full submersion, a part of upper
structures remains exposed to the air, thereby reducing the effectiveness of radiation shielding.

Figure 4.5-1: Classification of methods according to PCV water level
On the other hand, there is a partial submersion method, in contrast to the submersion method(The second image from the right in Figure 4.5-1 shows a conceptual diagram). If retrieving fueldebris based on the current water level of three damaged reactors, the fuel debris inside the RPV
would be cut in the air, while fuel debris at the PCV bottom would be cut in a partially submerged
condition. Even if we cut fuel debris while showering water into the fuel debris, we cannot expect
the level of effectiveness in radiation shielding and prevention of dispersing radioactive materials
equivalent to the case where fuel debris is fully submerged. In this manner, we must pay more
attention to handling fuel debris by the partial submersion method as compared to the submersion
method. However, the partial submersion method has potential for initiating fuel debris retrieval
without significantly changing the current condition of damaged reactors. A “Dry method” (Therightmost image in Figure 4.5-1 shows a conceptual diagram) is to maintain the whole area of fueldebris in a dry condition without water cooling nor spraying. Cooling of fuel debris is performed
through air cooling. Although this method has the advantage of no increase in contaminated water,
it must be verified that the cooling of fuel debris can be fully maintained for application. Therefore,
it is an option that may be chosen after retrieving fuel debris in another method, and coolability inthe air can be expected with less remaining debris inside. That is, working on cooling evaluations
on fuel debris and continuously considering employing this method are of significance.
(2)Three methods to consider intensively
We have examined three methods selected in the Strategic plan 2016 for intensive considerations
as shown in Figure 4.5-2 below. These are: submersion - upper access, partial submersion - upper
access, and partial submersion - side access. As there is no significant difference between full
submersion and submersion methods in terms of retrieving fuel debris in a submerged condition,
we treat both methods as the same submersion – upper access method.
As a method to access fuel debris, access from the upper PCV (upper access) and that from PCVside (side access) may be possible with consideration for the location of fuel debris.

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If fuel debris exists in the core/RPV bottom, access from above is suitable for retrieval. This routeis similar to refueling during periodic inspections and transportation of spent fuels from theoperation floor at nuclear power stations under the regular outages, and it has an advantage in thesystem evaluation. In order to set the actual access route, however, it is necessary to remove(partially or fully) a number of structures including the well shield plug, PCV upper lid, RPV upper
lid and steam dryer, etc. Depending on the damage to in-core structures, it may affect access andwe need to see the condition via internal surveillance, etc. in advance.
If fuel debris exists in PCV bottom (inside/outside the pedestal), access from the side is suitable for
retrieval. Although there has been a track record of access to the PCV bottom during internalsurveillance, considering the method to expand opening is required to insert the debris retrieval
device, etc. Access to the core/RPV bottom from the bottom is likely to be difficult, thus, in-depth
examinations are needed if application is considered. Similar to upper access, bottom access may
be affected by damage condition of internal structures and we need to see the condition via internalsurveillance, etc. in advance, depending on the level of damage to the PCV bottom.

Figure 4.5-2: Three methods to consider intensively
4.5.2 Technical requirements for the fuel debris retrieval methods
As for three methods to consider intensively, we evaluate issues and review the technicalrequirements defined in the logic tree shown in Figure 4.1-1.
At present, “the value of safety requirements” and “acceptable risk level”, which serve as a basis
to evaluate methods and technologies, have not yet been defined. As a result, all we can performnow is confined to the scope of qualitative evaluation. In the future, we need to substantiate
methods and system equipment, as details on the on-site condition are further identified throughinvestigations, and the like.
4.5.2.1 Ensuring containment functions
This section describes the concept of ensuring containment functions, issues in establishing thecontainment system for the liquid/air phases, the current status of developing PCV repairing
technologies, approaches to PCV water level, establishing water level controllability, and next step
actions.
4.5.2.1.1 Concept of ensuring containment functions(1)Objectives

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To protect the residents and environment from the impact by confining the radioactive materials
including alpha nuclide generated during the fuel debris retrieval work for a long period of time(for normal operation and accident condition) and controlling and preventing its dispersion.
(2)Major requirements
A. Define concept for ensuring containment functions during fuel debris retrieval work.
Assuming the current dispersion control of the radioactive materials at the Fukushima Daiichi
NPS where containment functions have been deteriorated by the accident, it is necessary toprovide an appropriate securement of the containment functions and dispersion control ofradioactive materials under the environment which is more severe condition where particulate
radioactive material including alpha nuclides are generated by the retrieval work. In this regard,
it is important to consider the dispersion and migration pathway of radioactive materials. There
are two major pathways, which is via groundwater from the liquid phase and via atmospherefrom the gas phase.
Moreover, dispersion control/management methods of radioactive material need to be
discussed for normal operating condition as well as standards for radiation exposure doseneed to be defined for the case of an accident. It is important that a concept of containment
function is be presented and shared based on the study results on the realistically achievable
containment functions including consideration of the positioning of the Fukushima Daiichi NPS
as a Specified Nuclear Facility.
B. Establish a containment system for the liquid phase
As mentioned above, it is necessary to establish a specific containment system for the liquid
phase, including reduction and control of radioactive material dispersion. In this regard, it isimportant to aim at a feasible system under a high radiation environment. If necessary, thestudy on containment system is required to be conducted again referring the fundamental
concept of containment functions. Also as the major requirements to be satisfied for thecontainment system for the liquid phase, the impact of the radioactive materials contained in
the liquid phase during the fuel debris retrieval work on the outside is required to be controlled
sufficiently, the impact on outside must be limited even at the accident condition such as alarge amount of radioactive material dispersion or at seismic event.
C. Establish a containment system for the air phase
A study is necessary on the containment system for the gas phase including the dispersion
control of radioactive materials based on the concept for ensuring containment functions
described above. In this regard, it is important to aim at a system which is feasible under thesevere radiation environment. If necessary, the study is required to be conducted again goingback to the concept of containment functions (boundary).
Also as the major requirements to be satisfied for the containment system for the gas phase,the impact of the radioactive materials contained in the gas phase during the fuel debris
retrieval work to the outside is required to be controlled sufficiently, the impact outside must be
limited even assuming the possible accidents, and the integrity of the system is to be secured
under the possible seismic conditions.

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(3)Status of action, evaluation and issue
A. Concept of containment functions
Fukushima Daiichi NPS is currently stable condition by taken post-accident emergency
response and the reasonably feasible measures on premise of the severe environment anddamaged facility by the accident. “Matters to be addressed” of the Nuclear Regulation Authority
specifies a standard (goal) to be complied with at this stage (during normal period) considering
the impact (entire site) caused by the additional exposure at the site boundary. This standards
is based on the dispersion limit from controlled area and radiation control of operating NPS.
Also, applying the standards at the various kinds of accident, the exposure dose at the siteboundary is assumed by at the time of accident.22
A specific limit of the exposure dose at the site boundary and that on workers at Fukushima
Daiichi NPS has not been established from the view point of securing safety and safety
regulation. Therefore, limit for site boundary dose and exposure dose on workers would bestudied based on the concept of containment system and fuel debris retrieval method which
are established under a certain assumption on the state of the regulation.
During fuel debris retrieval work, an increase of FP and alpha nuclides inside the PCV might
be a concern. To that end, the utmost effort needs to be made in order to minimize disperse ofradioactive materials containing alpha nuclides to the environment.
B. Containment system for the liquid phase
1) Current containment system for the liquid phase
Cooling water injected into the RPV flows down to the PCV, then water leaked from the PCV
is accumulated inside the reactor building. The cooling water circulation system collects,
treats and purifies this accumulated water and inject to the RPV as cooling water again. In
this case, the boundaries of the reactor buildings are designed to control the stagnant waterlevel in the reactor building lower that of the groundwater around the building. Accordingly,
groundwater is flowing in (inflow) and water containing radioactive materials are prevented
from being flown outside the building (outflow).
2) Containment functions for the liquid phase during fuel debris retrieval work
Since the liquid phase contains particulate fuel debris with alpha nuclides, containment
functions is to be achieved by the primary containment boundary consisting of the PCV andcells for fuel debris retrieval as well as the secondary containment boundary consisting of
the reactor building and others. Figure 4.5-3 shows example of containment functions forthe liquid phase under consideration. A study is necessary on development for PCV repair(seal of water leakage) technology and the PCV water level control for establishing the
primary containment boundary.
pouring grout for Fuel debris will be retrieved under pouring water or submerged condition,with a controlled PCV water level after repairing of the lower PCV (seal of water leakage byplugging vent pipe or placing concrete inside of S/C), and repair the upper PCV as needed.
The S/C receives the leaked water from the repaired locations and the cooling water
circulation system collects and treats it. Water leakage from the primary containment

22

Reference: "Safety evaluation guide" Partial revision of Former Nuclear Safety Commission ofJapan: Regulatory Guide for Reviewing Safety Assessment of Light Water Nuclear Power Reactor
Facilities, Mar 29, 2001.

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boundary (including a large amount of water leakage due to the accident) flows into the torus
room The secondary containment functions being maintained by collecting and treating theaccumulated water in the torus room by pumping it into the above mentioned circulation
system enables to prevent the contaminated water flowing out to the environment.

Figure 4.5-3: Containment of the liquid phase (example)
3) Importance of the water level control system for the liquid phase (securing water level
controllability)
Regardless of the fuel debris retrieval methods, concentration of radioactive materials in theliquid phase will be raised because fuel debris particles including alpha nuclides are to be
produced. In order to establish the containment functions for the liquid phase in this case,technical development for the PCV repairing (seal of water leakage) has been carried out.
However, the water level control system should be established, assuming that complete sealof water leakage is difficult because it is hard to verify the effectiveness of seal of waterleakage. The water level control system is important for maintaining “water level in the torus
room lower than that of groundwater around the reactor building and the containment system
for the liquid phase is also necessary to keep the condition of inflow of the groundwater andto prevent outflow of the contaminated water even at the accident such as a large amountof leakage.

C. Containment system for the air phase
1) Current containment system for the air phase
Hydrogen explosion has been prevented and the amount of radioactive material dispersion
has been also minimized by the PCV gas management system which has functions ofdeactivating inside of the PCV by the Nitrogen injection; and maintaining the gas phaseinside of the PCV, maintaining inside the PCV under the slightly positive pressure anddischarging gas inside of the PCV after filtering and measuring radioactivity. Consequently,

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the exposure dose at the site boundaries is sufficient low from the results of the evaluation
for the additional dispersion amount. 232) Containment functions for the air phase during fuel debris retrieval
At operating nuclear power stations, static containment functions are generally establishedby the reactor building and the PCV. However, static containment function at Fukushima
Daiichi NPS has been deteriorated by partial damage of the reactor buildings and the PCV.
Therefore dynamic containment functions by the negative pressure control system has been
considered. Figure 4.5-4 shows the example of the containment functions for the air phase
under consideration. It is necessary to proceed with technical development for establishing
the containment boundary by, inflow control, achieved by PCV repair and by the negative
pressure control system to maintain the negative pressure inside of it.
The primary containment function will be established by preventing outflow to theenvironment from the cell and the PCV through maintaining negative pressure inside of themthrough repair of upper PCV as needed as well as collecting retrieving fuel debris into
canisters and installing them into canisterstransport cask are performed within cells (also as
shielding). As the secondary containment functions, there is an idea that collects and treats
radioactive materials leaked from the primary containment to covers and containers installedon the existing reactor building which are maintained in slightly negative pressure.

Figure 4.5-4: Containment of the air phase (example)
3) Importance of the negative pressure control system
During fuel debris retrieval work, alpha nuclides concentration in the air phase is expected
to increase in spite of taking measures to collect radioactive materials containing alphanuclides as much as possible in the peripheral area of fuel debris retrieval. Since the upper
limit of the concentration for alpha is stricter than other radionuclides to prevent internal
exposure especially through breathing, we need more attention need to be paid to23“Evaluation results of additional release rate from the reactor building” by Tokyo Electric Power
Company Holdings, Inc., April 27, 20174-38

radioactive dispersion from the air phase in a form of dust. Consequently, the containment
boundary for the air phase needs to be constructed with careful consideration during fueldebris retrieval.
The negative pressure control system of PCV for ensuring containment in the gas phase isconsidered to be established by using conventional system configurations and devices. The
negative pressure control system has been investigated while studyingconcepts of various
necessary systems related to fuel debris retrieval methods. (Refer to Section 4.5.2.9)
The containment functions at Fukushima Daiichi are summarized in Appendix 4.11
Containment functions, including their history.
(4)Future actions
A. Concept of containment functions
The containment boundary and its concept is to be established in consideration of study resultson feasible containment system of liquid and gas phases during the fuel debris retrieval work
based on current concept for the containment functions.

B. Containment system for the liquid phaseCollecting the particles of fuel debris in the peripheral area when cutting the fuel enables
suppressing debris, and reduction of alpha nuclide diffused in the liquid and radioactive
materials in the contaminated water are to be investigated. The leakage from the PCV is to beprevented by repairing the PCV (seal of water leakage) and water level control. Consequently,
the risk is to be reduced by lowering the amount of radioactive materials that transfer to theaccumulated water in the building.
In the case of a large volume of water leakage, the water level in the torus room may become
higher than that of ground water depending on the water level in the PCV. Therefore, the waterlevel in the PCV and its control methods need to be studied to prevent the situation where thewater level in the torus room becomes higher than that of and ground water.

C. Containment system for the air phase
The negative pressure control system to maintain negative pressure inside the PCV will be
installed to minimize the impact to the outside even if alpha nuclide concentration is increasedin the gas phase during the fuel debris retrieval work. Also the studies need to be conducted
for the containment system in the gas phase including the installation of another negative
pressure control system (secondary boundary) that to control negative pressure inside the R/B.
As common activities to the containment systems for both the liquid and air phases, verificationthat impact on the external environment at the accident condition is small shall be performed
through the impact evaluation.
4.5.2.1.2 Establishing containment by repairing the PCV, etc.
(1)Objectives
Based on the boundaries described in the previous section, the PCV which is the primaryboundary is required to be repaired to the extent possible in order to prevent the impact to the

4-39residents and environment by confining the radionuclides including alpha nuclide caused by thework as well as minimizing and controlling its disperse during the fuel debris retrieval work for along period of time (during the normal operation and accident condition). It is, however, difficult
to achieve a complete seal of water leakage since the cement-based materials will be used for
the PCV repair material which is currently under the development. The establishment of thecontainment system that allows a certain amount of the leakage should be studied as well.
(2)Major requirements
A. Ensure secure long-term water sealing performance for the liquid phase and in-leak control
functions to maintain of seal of water leakage (PCV repair) negative pressure for air phase.
B. Ensure the reliability of seal of water leakage including the monitoring and inspection of the
PCV repair work, detection in case of leakage and establishment of the re-repair method.
C. Establish the water level control system f with function of leaked water collection for preventing
water leakage to outside of the R/B.
(3)Status of action, evaluation and issue
The leakages currently identified are at the sand cushion drain line at the bottom of the PCV,
bellows on the vacuum break line and bellows of the PCV penetrations for MS line and so on.
The possible locations of leakage includes suppression chamber, which is shown in Section 4.3,
as well. Also, since the inspection of the upper part of the PCV has not been carried out due tothe impact of high radiation caused by the accident, the possible locations of damage of each
unit may reach to about 300, which is quite a lot (Refer to Figure 4.3.2-14). The areas whereboundaries are to be established are indicated by the bold red lines in Figure 4.3.2-3 and 4. Since
the leakage may be occurred from the suppression chamber itself or its penetrations of Unit 2,
the possible leak paths are planned to be closed by applying work of seal of water leakage tovent pipes, downcomers, strainers which are located upstream of them. Also the penetrations in
the upper part of the PCV where the leakage are expected are also planned to be sealed waterleakage. It is, however, difficult to repair by welding, since the repair areas are of very high degreeof difficulty in accessing due to a high radiation and many obstacles. Therefore, the grout placingor blasting of sealing material are being investigated as a method for seal of water leakage atarea where leakages are occurred using remote-controlled equipment in order to seal the waterleakage. However, the repair will basically be performed by welding if the areas where welding
is considered applicable is found based on future the on-site inspection.
Figure 4.5-6 shows the status of development of the major technical concept for repairing thePCV, and the current evaluation result and issues related to the technical development aresummarized below.

A. Inflow control to secure function of containment secure and to maintain negative pressure
1) Repair of the lower part of the PCV (below torus room ceiling)
For the submersion method, the repair of lower PCV portion for seal of water leakage thatwithstands against water pressure correspond to submerged method are underdevelopment based on the evaluation on the current containment functions in the liquid
phase. However, that is evaluated of very high degree of difficulty. As well, risk of outflow isalso high because of a large water volume in the PCV. In the case of the partial submersion

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method, the repair work is not evaluated of high degree of difficulties because the waterpressure at repaired areas of the lower PCV is lower than that for the submersion method.
a. Plugging vent pipes
This technology is placing the plug inside of the vent piping which is consist of following
three steps. Firstly, installing a temporary weir of grout filled inflatable bag at the middleof the vent piping; secondary, plugging remaining gaps between the vent piping andtemporary weir by filling supporting materials into them (heavy aggregate, etc.) and finally,
injecting water sealing materials (concrete or rubber materials) at the upstream of the weir.
(Refer to Figure 4.5-6)
The component testing on rubber materials has been performed in addition to cement
materials which were initially examined as a candidate of sealing materials for the ventpipe. Cement material is difficult to achieve complete seal of water leakage but can belimited to a small amount of water leakage against the water pressure correspond to thewater level at the bottom of the torus room ceiling (equivalent to the water level where
outflow from the R/B can be prevented at when all of water outflowed into the torus roomat the accident condition). Moreover, rubber materials as an alternative water sealingmaterial are under investigation. Although it is necessary to continue this development toseal water leakage, it may ensure containment functions by the combination of controlling
proper water level and the water collection systems from the inside of the PCV and S/C
for the partial submersion method.

b. Seal of water leakage by pouring grout for concrete inside of S/CThis is a method for the sealing of water leakage by bury downcomer tips and piping
structures (strainer, quencher, etc.) by pouring grout materials inside the S/C. (Refer toFigure 4.5-6) The seal of water leakage or strainer/quencher, downcomer, and vacuum
break valve will be established by changing the height of filledpouring grout materials
inside the S/C. The mixing ratio of the concrete had been determined by the groutcomposition test. Subsequently performed test for seal of water leakage at thestrainer/quencher by using the concrete has proved the good pouring condition and waterpressure resistance. The seal of water leakage on the downcomers and vacuum breakvalves, is difficult under the water flowing condition correspond to current cooling water
circulation, but that is confirmed to be achievable in no water condition flowing.
Filledpouring grout materials could impact on the seismic performance of the S/C due tothe increased inertia load is under investigation for the various filled groutlevels of pouringlevel inside the S/C and S/C water level above concrete.
Moreover, development of the techniques to place mortar in the torus room (outside S/C)
is in progress to reinforce S/C supports.

2) Repair of the upper part of the PCV (above the 1st floor of the reactor building)
The component test is mainly conducted since inspection has not progressed sufficiently
due to high radiation. Since there are many possible damaged portions at the upper PCV tobe repaired, the radiation exposure dose of site workers is expected to exceed the pastannual total exposure dose significantly, even if the dose rate in the work area could be

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brought down to 3 mSv/h. As dose reduction may remain difficult, remotely operated repair
technology need to be investigated for the upper part as well.
The negative pressure control system (ventilation system with a function to maintainnegative pressure) is required for air phase in the upper part of the PCV. The partial
submersion method requires the negative pressure control system to contain alpha nuclides
inside of the PCV, and its capacity becomes big due to the volume of air phase. However,this system is considered to be feasible.
a. Repair of secure water sealing for D/W penetration and the penetrations to maintainnegative pressure by inflow control
For the development of the repair method (Refer to Figure 4.5-6) bellows at pipe
penetrations and electric penetrations, applicability of improved Polyurethane rubber andadditionally selected Polyurea water sealing materials was examined by using blasting
method.
Polyurethane rubber was confirmed to satisfy water tightness up to 0.45 MPa (correspondto water level up to the Well) as well as to be effective to inflow suppression up to negative
pressure of -1.0 kPa. High-pressure water jet and rust remover by reducing agent are
confirmed to be applicable to remove sludge and rust from portions to be repaired.
Repair of the D/W penetrations is of very high degree of difficulty because repair work is
to be performed under high radiation exposure dose condition and a study is also neededon removing structures interfering with repair of the D/W penetrations and penetration
covers and so on. Therefore, additional study need to be performed on remote-control
devices for the practical application.
b. Seal of water leakage of the equipment hatch
Work of seal of water leakage will be implemented by welding or blasting of sealingmaterial by using a remote-controlled device which is accessible through hole on theshield blocks (Refer to Figure 4.5-6). It was confirmed by the component tests that thisrepair method is effective to flange mating surfaces of the equipment hatch with up to
3mm in gap and 7 mm in level difference. And it is effective to prevent inflow into the PCV
under negative pressure. Blasting or laser beam was confirmed to be effective for
exfoliation of paint and rust as a result of the combining test of the equipment hatch bywelding.
The level of difficulty of this repair work is high, because this repair work need to be
performed by remote-controlled devices at the site and further study on practicalapplication is needed.
3) Seal of water leakage of pipe penetrations through the torus room wall
For the development of the repair method (Refer to Figure 4.5-6) for pipe penetrations
through the torus room wall, applicability of improved Polyurethane rubber and additionallyselected Polyurea water sealing materials was examined by using blasting method.
Polyurethane water sealing materials was confirmed to satisfy water tightness up to 0.15
MPa. High-pressure water jet and rust remover by reducing agent are also confirmed to be
applicable to remove sludge and rust from portion to be repaired.

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However, a study is necessary on access to this wall from T/B and Rw/B, depending on the
progress of reducing dose because access to this wall from the reactor building is currently
impossible due to high dose exposure condition,
Pipe penetrations o through the torus room wall were repaired by filling grout at Units 5 and
6 where the torus room is accessible. However, the inflow of the ground water into the torus
room is just reduced to the half and complete seal of water leakage may be difficult due tounidentified water path. Therefore, seal of water leakage at Units 1-3 is more difficult,because of inaccessibility due to high radiation condition. Other measures may be possible
such as soil improvement around the reactor building or installing land-side impermeable
walls. However, these measures are effective to reduce water inflow but difficult to stop water
inflow completely because inspection of seal of water leakage and leak detection from
improved soil or impermeable walls are difficult.
B. Reliability of seal of water leakage
1) Reliability of seal of water leakage on the S/C, torus room, piping (vent pipe, downcomer,etc.)
Conditions of the inner surface of the location of component to be sealed is yet to be
identified inner surface condition of them needs to be identified and its impact to the seal ofwater leakage is also clarified by test for the on-site application.
2) Long-term performance of seal of water leakage
Technical development of leakage detection and re-repairing (e.g. selection of material for
re-repairing and its injection position) methods needs to be planned.
3) Reliability of work of seal of water leakage
Technical development which enables monitoring of the repair work progress and repair
completion condition is necessary. A study is necessary on repeatability and performance
guarantee method (leak test, inspection, etc.) of work of seal of water leakage, as well asverification of feasibility needs to be verified by test for repetitive works such as plugging on
the vent pipe.
C. Establishment of water level control system for the liquid phase
Section 4.5.2.9 shows the example of the study results on the concept of the cooling water
circulation system with function of water level control in liquid phase under the assumption thatsome amount of water will flow out from the repaired portions. This system has a broad rangeof functions including fuel debris cooling, criticality control, corrosion control of PCV internals,
monitoring/control of the PCV water level. Major function of this system is to maintain waterlevel in the torus room lower than that of underground water and to prevent outflow from thetorus room at the accident by determining the water intake place point from the PCV,
monitoring/controlling of water levels of the PCV, S/C, and torus room.
1) Concept of the water level inside the PCV
The current water level inside the PCV differs among Units 1-3. Its practical application ofwater leakage technology which is under development is also considered to be differed byunit. Therefore, the PCV water level will be determined through comprehensive
considerations on the following requirements.
a. Advantage of cutting/handling fuel debris under water (Prevention of radioactive
materials dispersion including alpha nuclides, into the air)b. Possibility of reversing the level of underground water and in the torus room accident
condition in the event of abnormality

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c. Technical feasibility of work for seal of water leakage sealingd. Technical feasibility of water level control
In any case, a study on technology of seal of water leakage for the lower PCV andtechnologies and procedures to stably control water level at the determined level needs tobe continued.
2) Ensuring water level controllability
Currently, technical development underway for water level control such as changing waterlevel or maintaining water level at a certain level during fuel debris retrieval work. Althoughdevelopment of technology of plugging on the vent pipe is continuing even though it has
technical difficulties in terms of feasibility because it enables to reduce size of primarycontainment boundary by isolating the D/W and S/C.
Appendix 4.12 shows the study of study on water level in the bottom of PCV when retrieving
fuel debris work.
(4)Future actions
Actions specified below will be pursued for developing a method of establishing a containment
boundary by repairing the PCV, etc. and its practical application.
A. In-leak control for secure water sealing performance and maintaining negative pressure
By concentrating on solving aforementioned many issues, in-leak control functions will beestablished to maintain negative pressure of gas phase and to seal water leakage for theliquid phase. For suppressing dispersion of radioactive material during fuel debris retrieval
work, the air-phase control system to maintain negative pressure inside the PCV will beestablished. In order to enhance containment function of the air phase, feasibility of PCV
repair method by welding or blasting of sealing material for controlling in-leak to the PCV will
be investigated. As well, dose reduction by decontamination, removal of radioactive materials,
radiation shielding, etc. and technologies to place repairing device to the target area will be
studied to enable PCV repair work by remote control devices.
B. Reliability of seal of water leakage
Long-term performance and reliability of repair will be ensured by concentrating on thefollowing three items.
1) Verification of performance of seal of water leakage under the conditions simulated theactual unit (rust and sludge on the inner surface of the PCV) and reproducibility of repairing
work.
2) Verification of long-term performance of seal of water leakage including feasibility of rerepair work
3) Verification of reliability of repair work including monitoring technologies (repair work
progress and completed condition)
C. Establishment of water level control system for the liquid phase
As mentioned before, pours a certain amount of water leakage into the torus room needs tobe accepted because it is difficult to complete seal of water leakage by current repairing
method that fills grout materials. Therefore, water level in the torus room is required to be
maintained under that of ground water around R/B during fuel debris retrieval work. In addition,a study is necessary on entire acceptable leaked water volume prediction method based on
estimation of leaked water from repaired portion as well as the liquid phase control system,
including the collecting system of outflowed water. (Refer to Section 4.5.2.9)

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4-45
Figure 4.5-5: Upper PCV penetration (example at Unit 2)(Prepared based on IRID materials)

4-46
Figure 4.5-6 Main PCV repair technologies(Prepared in based on IRID materials)

4.5.2.2 Maintaining cooling functions(1)Objectives
To maintain cooling functions over a long period for stagnant water treatment, repairing PCV andfuel debris retrieval work since fuel debris continuously generates decay heat.
(2)Major requirements
Major requirements are listed follows:

Be sure to assume, control and record the status of cooling fuel debris, and maintain its
temperature less than 100ºC.
To install alternative water injection system and equipment including agile measures forcase that a permanent facility cannot cool fuel debris at the events such as earthquake
and tsunami.
To maintain cooling functions even during repairing PCV.

In addition to above, the following indicates major requirements for the cooling water circulation
system during stagnant water treating period and fuel debris retrieval work.
A. During stagnant water treating period Capable of cooling the reactor vessel, removing of Cs and desalting of the contaminatedwater.
 Capable of treating stagnant water in each building with different floor height and waterlevel. For buildings where treatment has not been completed, capable of maintaining
constantly “stagnant water level inside the building lower than that of the undergroundwater” (monitoring and controlling of each water level).
B. During the fuel debris retrieval work Capable of Circulation of required water flow, collection of excess water and drainage
before the commencement of the PCV repair work.
 Furnish the required functions (e.g. cooling, cleanup, water level control, criticality
prevention) for long-term operation during the fuel debris retrieval works.
 Be studied on the method of treatment of the small pieces of fuel debris flowing into thecirculation loops(3)Status of action, evaluation and issue
Cooling water circulation has been maintained by cooling water the circulation system as shownin Figure 4.5-7, and temperature at RPV bottom and air phase of the PCV at each Unit are being
kept in a stable condition. There is no significant change in operating parameters such aspressure inside the PCV and the amount of radioactive material disperse from the PCV, has been
observed and no sign of disorder of cooling conditions or criticality has been identified. From theabove, it is confirmed that a condition equivalent cold shutdown are generally maintained andthe reactor vessel is in a stable condition.
For improving reliability of cooling function, the work has been carried out to shorten circulationloop, resulting in reduction of the circulating loop from about 3 km to about 0.8 km. (Refer toFigure 4.5-7). In addition, for acceleration the stagnant water treatment in the buildings, thevolume of water injection into the reactor vessel has been reduced to 3.0 m3/h since December

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2016, starting from Unit 1, then Unit 3, and Unit 2. No disorder has been observed in the coolingcondition due to the decrease of water injection24.
(4)Future actions
C. Cooling water circulation system during stagnant water treating period
For further improvement of reliability of cooling function, the proposal for approval of revisionof the “Implementation Plans” was submitted for alteration of some portion of CS water
injection line to polyethylene pipe at Unit 1, and the timing of implementation is now underconsideration. (Applied on March 6, 2017)
D. Cooling water circulation system during the fuel debris retrieval work
The cooling water circulation system shall be designed based on a study results on acceptable
volume of water leakage after PCV repair, volume of water collected from the torus room, thatof circulating for cooling and that to be pumped out based on estimation of water leak location.
Investigation is also required on the location of intake for cooling water, the installation methodof the intake line, and others.
Moreover, it is necessary to establish the cooling water circulation system with the functions toenable retrieving the fuel debris under water or pouring water condition. The cooling water
circulation system needs to be investigated to have the functions such as cooling, criticality
control, removal of radioactive material, suppression of turbidity, water quality control, waterlevel control/monitoring, and interlocks, which are necessary during fuel debris retrieval work,in taking into account engineering work, R&D and preparation for responding to regulations.
(Refer to Section 4.5.2.9).

Figure 4.5-7 Conceptual Diagram of the Present Circulating Loop

24

Tokyo Electric Power Co. “Overview of Countermeasures for Decommissioning & Contaminated
Water Treatment”, The 40th Team Meeting for Countermeasures for Decommissioning andContaminated Water Treatment/Secretariat, March 30, 20174-48

4.5.2.3 Criticality control(1)Purpose
Regarding the fuel debris with no sign of re-criticality under the present condition, the occurrenceof re-criticality shall be prevented, considering possible changes in the condition caused by thework associated with fuel debris retrieval (including preparation work), and measures to prevent
impact on people and the environment shall be taken as the preparedness for re-criticality.
(2)Major requirements
The representative critical conditions under which the substances including nuclear fuel materials
are that the presence of a moderator like water, high concentrations of U-235, Pu-239, etc. (inaddition, a small amount of neutron absorbers), and having a certain size. As for the fuel debris
present at the Fukushima Daiichi NPS, the possibility of re-criticality is extremely low from theengineering viewpoint, because the alternation of molten fuel assemblies are not likely to reachcriticality in terms of the abundance ratio with water, mixture of impurities such as core internals
is expected in the process of core melt, the debris are assumed to scatter in a wide range, insteadof remaining in the core, as a result of the accident evolution. In addition, it is assumed that thefuel debris widely scatters even in the event of re-criticality, etc., therefore, the effect of recriticality is estimated to be small.
Considering a large extent of uncertainty inside the PCV, as the measures against the possibilityof the above-mentioned conditions that may be locally combined because of possible changes
in the shape of fuel debris or in the amount of water, etc., it is required to establish a proper
control method to prevent increasing of exposure risk due to the newly generated radioactive
materials, etc., by understanding the possible conditions leading to re-criticality during fuel debris
retrieval, and combining prevention, detection and termination of criticality.
Based on the above, the requirements to accomplish the purposes are described below: These
requirements are common to individual methods.
A. Establishment of the concept for the criticality control method
1) Understanding of the conditions leading to re-criticality
In order to appropriately control the criticality in the current status where the conditions inside
the PCV is unknown, it is necessary to evaluate the possibility of reaching the re-criticality
under based on the results of the estimation for the amount, location, shape and propertiesof the fuel debris, and identification of the events that may induce re-criticality.
The criticality scenario is to be evaluated based on appropriate conditions assuming
multiple methods.
 To reduce uncertainties in the assessment of possibility of re-criticality, the necessary
information must have been identified, and a plan to obtain the information must havebeen developed and implemented.
2) Evaluation of behavior at criticality

To consider the measures for mitigating impacts in case of re-criticality, it is required toassess the amount of FP generation, exposure doses, etc. with high accuracy, by evaluating
the behaviors during criticality.
 The accuracy of assessment of impacts in case of re-criticality must have been verified.
3) Consideration of the criticality control method
It is necessary to consider the criticality control method based on the result of the abovementioned assessments, and verify the applicability in actual use.
Regarding the system and equipment to be used for criticality control, the applicability

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in actual use must have been confirmed, in line with the review of the retrieval system& equipment and the circulating cooling system.
B. Development of technologies for implementing the criticality control method
1) Criticality prevention Technologies
For the purpose of improving reliability for criticality prevention during the work associatedwith fuel debris retrieval, the technologies have been developed for monitoring of subcriticality and applying of neutron absorbers.
a. Sub-criticality monitoring methods
 The increase in effective multiplication factor of the fuel debris shall be able to be
detected.
b. Application of neutron absorbers
 In the case of applying the soluble neutron absorbers, the system integrity shall bemaintained, considering the corrosion of core internal materials and the effect on the
coolant circulating system.
 To maintain the assumed condition of sub-criticality, the required reactivity shall beidentified and secured.
2) Re-criticality detection and criticality suspension technologies
In case of re-criticality, measures shall be taken to detect the increase of the amount of FP
generation, neutrons, and γ-ray dose to stop re-criticality.
By detecting partial re-criticality of the fuel debris widely distributed, safety shall be
secured in combination with the measures for mitigating impacts (such as criticality
suspension, the containment function, etc.)(3)Action status and evaluations and issues
A. Establishment of the concept for the criticality control method
1) Understanding of the conditions leading to re-criticality


To indicate how high the water level of the PCV can be increased in terms of criticality control,
it has been carried out to assess the concentration of boric acid required for criticality
prevention and to review the possibility of increasing the PCV level by pure water, assuming
the residual fuel at the core (including the partially intact fuel), and the fuel debris at thebottom of the RPV, in the CRD housing, and at the bottom of the PCV.
The concentration of boric acid required for criticality prevention is assessed to be about
6,000 ppm, considering that gadolinium, contained in the fuel assembly as a neutron
absorber, is accompanied. Regarding the assessment of the possibility of a level increaseby pure water, on the other hand, it is indicated that the possibility of re-criticality is low when
considering the core internals and the FP as the composition of fuel debris within a realistic
range, as long as the level increases up to the lower RPV. For the level increase up to thecore, the real composition of the fuel assembly was considered for evaluation, and theconditional range possibly leading to criticality was analyzed, considering the number ofassemblies remaining in the core.
The information required to improve the accuracy of these evaluation results has been
provided for the actions to analyze the internal core condition, and further actions are
required to reflect the information that will be obtained in the future.
Since the work of fuel debris retrieval involves a change in the shape of fuel debris
associated with cutting, etc., the scenario of criticality has been developed for each methodof retrieval, and the criticality control method is considered based on the result.

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Since the information regarding the possibility of re-criticality is essential for making
decisions to determine the method and start the work, a method of comprehensively
estimating the possibility of re-criticality is being reviewed based on limited data, byincorporating statistical handling of the composition of fuel debris, etc., which is consideredas the criticality evaluation condition. In addition, at Unit 1, an action of evaluating the degreeof sub-criticality is being taken through analysis of the radioactivity concentration data for
Kr-87/88 measured by the gas management system. As the evaluation of the degree of subcriticality based on the data in actual use can be valuable basis for decision, reviews should
be continued, taking into account applicability to other units.
2) Evaluation of behavior at criticality
While evaluation of the possibility of re-criticality is now proceeding as described in 1), it isdifficult to completely eliminate a possibility of re-criticality under the current situation ofmany uncertainties present in the information regarding the inside of the PCV; thus,
measures for mitigating effects should be considered as the preparedness for re-criticality.
Consequently, development of a method of evaluating the behaviors during criticality has
been proceeding to assess the response of neutrons and the amount of FP generation after
the occurrence of re-criticality. Until now, construction of models for evaluating the behaviors
during criticality has been carried out, assuming water filling and fuel debris retrieval,considering the each retrieval methods. In addition, aiming at preventing the effects on
people and the environment even in case of criticality, through the evaluation of behaviors
during criticality under representative scenarios, the information required for thedevelopment of measures for effect mitigation and for the fuel debris retrieval system has
been extracted.
To establish a method of evaluating behaviors during criticality, accuracy must be enhanced
by reflecting the review status regarding the information of the inside of the PCV and thefuel debris retrieval system, which will be obtained in the future.
3) Consideration of the criticality control method
Based on the result of evaluation for the possibility of re-criticality and behaviors during
criticality, as mentioned above, review is now being made for the criticality control method
during water filling of the PCV and during fuel debris retrieval (Table 4.5-1).
As the criticality control method during water filling in the PCV, the use of pure water as thecoolant and the use of sodium pentaborate, a soluble neutron absorber, are underconsideration. As the criticality control method during fuel debris retrieval, requirements havebeen extracted such as setting a limit on the one-time amount of retrieval for each method,and suppressing the dispersion of cutting particles, and been proposed for development of
the system and equipment for fuel debris retrieval. In addition, a more reliable method of
criticality prevention is being considered through the technology for monitoring sub-criticality
and the application of insoluble absorber.
Evaluation associated with the feasibility during the PCV level increase and the fuel debris
retrieval is described as follows:a. Criticality control during the PCV level increase Considering mixture of core internals, FP, etc. into the fuel debris, the possibility of

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reaching re-criticality is estimated to be very low in reality, even in case of the level
increase up to the lower RPV. The information required to reduce uncertainties in
evaluation has been provided for the action to analyze the internal core condition, andfurther actions are required to reflect the information that will be obtained in the future.
 Although the possible condition for re-criticality is limited during the level increase up to
the core, in the cases when the presence of residual fuel cannot be negated as in Unit
2 and Unit 3, it is required to ascertain a more feasible method of criticality control,
based on the review made for constant injection of sodium pentaborate as well as theresults of inspecting the inside the RPV. At Unit 1, the amount of residual fuel in the corearea is estimated to be small; thus, the possibility of re-criticality due to the water level
increase is thought to be low.
 If water filling by pure water is adopted, as the measures to be taken at the time when
the work is stopped due to abnormality detected during the work, flexible changes in
the plan should be assumed to continue the work, including the changeover from purewater to sodium pentaborate as the coolant, the change of the assumed PCV level, etc.
b. Criticality control associated with the fuel debris retrieval work
 By assessing the one-time amount of retrieval for fuel debris and the additional reactivity,
the requirements for the device and system for fuel debris retrieval have been extracted,
such as setting a limit on the one-time amount of retrieval, as required. In the future,they should be reflected in the device design.
 To improve reliability of criticality prevention during the work, reviews have been madefor the development of technology for monitoring sub-criticality and the application ofneutron absorbers.
 Flexible reviewing of the control method is required, by deliberately working while
accumulating information relating to fuel debris, based on the information obtained
depending on the progress of the work.
In the future, the key issues to establish a criticality control method include the following:
Regarding the requirements for the system and device for fuel debris retrieval and thecirculating cooling system that were extracted in the process of reviewing the criticality
control method, the result of confirmation of the feasibility of each method shall bereflected in the criticality control method.
 To respond to uncertainties in the information relating to fuel debris, the concept of how
to optimize the criticality control method step by step shall be organized, based on theinformation obtained depending on the progress of the work.
 The progress of technology development, as shown in B. below, shall be reflectedaccordingly.
 Considering the feasibility of other technical requirements relating to fuel debris retrieval
and reduction of exposure suffered by workers during the field working, adequate andrealistic goals of criticality control shall be established through discussions with relevantorganizations.
 When more than one candidate is available as the criticality control method which was
confirmed as feasible, determination shall be made with reference to “Five basic
concepts” described in Chapter 2.
B. Development of technologies for implementing the criticality control method
1) Technology for criticality prevention
To improve reliability for re-criticality prevention during working, the following technology
development is now proceeding.
a. Technology for monitoring sub-criticality
The technology for monitoring sub-criticality is designed to prevent re-criticality throughearly detection of the status of approaching to criticality and transmission of an alarm prior
to reaching criticality, by persuading to stop the work. The proposed specific method of

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operation is shown in Figure 4.5-8. As the result obtained so far, technologies with certain
applicability have been reviewed and a system concept of combining methods, such as
the reactor noise analysis and neutron source multiplication, has been established. In the
future, we must conduct testing to confirm the feasibility and review reflecting of the results
into the procedure of criticality control; it is also required to review the possibility of neutron
detection in the vicinity of fuel debris, together with development of the fuel debris retrieval
device, as described later in 2).
b. Application of neutron absorbers
Constant injection of sodium pentaborate as a soluble neutron absorber has been
considered, and the development of insoluble absorbers is now proceeding.
For the constant injection of sodium pentaborate, the range of concentrations for sodium
pentaborate, which is consistent from both of the viewpoint of corrosion of structural
materials and that of the required concentration for sodium pentaborate to suppress
criticality, has been detected, and reviewing of the co-existence with other systems suchas the circulating cooling system has been started (refer to Section 4.5.2.9). As the futurechallenges for application in actual use, we must continue to review the impacts on waste
and fuel debris storage cans.
For the development of insoluble neutron absorbers, requirement specifications for field
application (such as radiation resistance, absorptivity to fuel debris) have been defined
and the candidate materials have been extracted so far, thereby narrowing down themthrough fundamental properties testing, radiation resistance testing, etc. As a result,durability to the surroundings in the vicinity of fuel debris is estimated to be obtained,
regarding B4C/metal sintered materials and glass materials containing B/Gd, etc. In the
future, testing and evaluation must be conducted to understand the nuclear properties andthe effect on fuel debris storage cans, etc., in order to narrow down to final candidate
materials. In addition, as the method of confirming the reactivity reduction effect throughinput of neutron absorbers, an operation method must be considered, combined with subcriticality measurement using the technology for monitoring sub-criticality
2) Re-criticality detection and criticality suspension technologies
As the criticality detection technology, two methods are under consideration: one is to detect
neutrons, and the other is to measure gamma rays emitted from the short-lived FP gas.
The method of detecting neutrons is of highly fast-response; however, a detector must be
installed in the vicinity of fuel debris due to the short moving distance of neutrons underwater,and review is being made, in line with the development of a retrieval device. In addition, as
the requirement specification for a detector, information of dose rate, etc. at the location of
the detector is required to determine radiation resistance and detection sensitivity.
Consequently, the result of investigation for the inside the PCV, which is supposed to beobtained in the future, must be reflected, while several candidates for a detector must be
considered.
As the method of measuring gamma rays emitted from the short-lived FP gas, it isconsidered to speed up the response to re-criticality detection by measuring Kr-87/88 inaddition to Xe-135, which is currently measured. According to the testing in actual use
conducted at Unit 1, it was confirmed that Kr-87/88 deriving mainly from spontaneous fission
can be measured. Further, the behavior of the FP gas generated in the PCV up to arriving
at the gas management system is analyzed, in order to evaluate the adequacy of the delay

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time in re-criticality detection, which is assumed in the evaluation of behaviors during
criticality, as well as the possibility of further earlier detection.
As for the criticality suspension technology, while the use of the standby liquid control system,
which has already been operated at the site, is kept in mind, reviews are being made for thework of water filling and fuel debris retrieval, including the necessity of optimization. Since
the system capacity and the control value of boron concentration, which should bemaintained during normal operations, may vary depending on the water volume held in the
PCV, it is important to reflect the status of progress made in reviewing the methods,
accordingly.
(4)Future actions
A. Establishment of the concept for the criticality control method
 Information that should be obtained for developing criticality control methods shall be
secured without fail, by determining a specific timing when the information is required,and collaborating with the action of analyzing the internal PCV condition.
 Adequate and realistic goals shall be established for criticality control, and a logic shallbe constructed based on the data showing the adequacy.
 Regarding the requirements for the system and device for fuel debris retrieval and thecirculating cooling system that were extracted in the process of reviewing the criticality
control method, a review shall be made for refinement, based on the fuel debris retrieval
policy which is determined. In addition, the result of confirmation of the feasibility of thesystem and equipment for fuel debris retrieval as well as the circulating cooling systemshall be reflected in the criticality control method during reviewing.
B. Development of technologies for implementing the criticality control method
1) Criticality prevention technologies
a. Sub-criticality monitoring methods
 Although a detector must be installed in the vicinity of the location for processing fueldebris, the achievability of requirements shall be adequately reviewed, due to theuncertain distribution of fuel debris and restriction on the location for installation.
 If achievement cannot be expected, the policy of development shall be reconsidered.
b. Application of neutron absorbers
 Regarding the constant injection of sodium pentaborate, impacts on the environment
due to leakage and effect on storage cans shall be identified, and the method of
operation in actual use shall also be reviewed.
 Review shall be made on the criticality control method to reduce the boron
concentrations as practically as possible, via combination with other technologies, etc.
 To practically apply insoluble neutron absorbers, a method of quantification shall beconsidered for the reactivity reduction effect of neutron absorbers.
 Long-term effect on storage cans, etc. of insoluble neutron absorbers shall be identified.
2) Criticality detection and criticality suspension technologies
 Regarding the re-criticality detection by neutron, requirement specifications shall bedefined with the result of investigations for the internal PCV, which is supposed to beobtained in the future, reflected, and several candidates for a detector shall beconsidered.
 In the case of re-criticality detection through detection of gamma rays, the delay time indetection should be shortened; thus, improvement of accuracy in detection using
krypton gas shall be considered.
 Regarding the boric acid solution injection system used for criticality suspension, its
performance shall be confirmed, taking into account the status of reviewing theconstruction method and criticality control.

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Table 4.5-1 Basic Concept of Criticality Control (draft)

Criticality
Prevention

Concept of generalmanagement method(sample)- Evaluate possibility ofcriticality and apply
neutron absorbers- Limit the reactivity
inserted at one time
while measuring thedegree of sub-criticality
in advance of operation- Suspend the operation
by detecting the stateclose to the criticality
Impacts
Mitigation(Criticality detection)
Identify re-criticality
state by monitoring
neutron and FP gas
behavior(Criticality suspension)
Insert negative
reactivity after
detecting re-criticality

PCV water filling
Pure water
Sodiumpentaborate
Criticality scenario
Evaluate criticality
evaluation
scenario +criticality
prevention bysodiumpentaborate
Limit of water filling
speed Incremental
water-fillingFP gas behavior
monitoring

FP gas behavior
monitoring

Criticality detection
by short-lived FP gas γray

Criticality detection
by short-lived FP
gas γ-ray

End of criticality bysodium pentaborate
or decrease in waterlevel

End of criticality byhigh density
sodiumpentaborate
or decrease inwater level

Fuel debris
retrieval
Evaluate criticality
scenario +criticality
prevention byusing absorber
limitation of
retrieval quantity at
one time
Sub-criticality
monitoring by
Sub-criticality
measurement
systemCriticality detection
by shot-lived FP
gas γ-ray or byneutronEnd of criticality byabsorber*For criticality prevention, any one of the methods mentioned above, or combination of them are under consideration.

Public document: 2017.2 Secretariat of the Team for Countermeasures for Decommissioning andContaminated Water Treatment

Neutron K-effective

Figure 4.5-8: Image of Criticality Prevention by Monitoring for Sub-criticality (Source: IRID)

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4.5.2.4 Ensuring the structural Integrity (aseismicity) of the PCV and reactor building(1)Objectives
The major structures such as reactor building, PCV, RPV of each unit that were damaged dueto the accident shall satisfy the following important safety functions during the fuel debris retrieval
under both normal operating conditions and seismic events.


Reactor buildings shall maintain the supporting functions for the equipment andsystems important for safety, such as the PCVs and RPVs.
Deterioration of the containment function of the PCVs, RPVs and reactor buildings shallbe suppressed, and the mass release of radioactive materials shall be mitigated and/or
prevented.

Figure 4.5-9 illustrates the reactor building, and major components.

Figure 4.5-9 Cross section of R/B and major components(2)Main requirements
A. Evaluation on the aseismic performances of the submersion method and partial submersion
methodThe required functions described in (1) shall be evaluated to be maintained during the fueldebris retrieval work even at large seismic events, applying either the submersion method or
the partial submersion method.
The following conditions need to be taken into consideration in the evaluation of aseismic
performances.

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Consider the damage caused by the accident, degradation of materials due to exposure
to high temperature, corrosion due to seawater injection as well as further degradationand corrosion that may progress to the time of the completion of fuel debris retrieval.
 Take account of the additional weights based on fuel debris retrieval work plan such asfuel debris, water inside of the PCV for cooling and submersion of fuel debris, watersealing material and fuel debris retrieval equipment including, radiation shielding andsupporting structures.
 According to the above, establish appropriate design seismic ground motions andevaluation criteria considering the damaged states of the systems, structures andcomponents induced by the accident.
B. Development of corrosion prevention measures for RPVs/PCVs and related piping systems,
and the verification of its applicability

The prevention measures for corrosion progression in the RPVs/PCVs and related piping
systems over a long period of time during the future fuel debris retrieval work in order tomaintain the current status shall be developed, and their applicability to the actual equipment
shall be verified.
(3)Action status, evaluations and issues
A. Evaluations on seismic resistance of the submersion method and partial submersion method
Following studies have been carried out for the evaluations of aseismic performances of thesubmersion and partial submersion methods so far.
Seismic safety assessment of the reactor buildings taking into account its damage
induced by the hydrogen explosions
Evaluation of the load bearing capacities and stiffness of the reinforced concrete RPV
pedestals which have experienced high temperature exposures during the accident andthen cooling water injection
Evaluation on the aseismic performances of the submersion method and partial
submersion method, considering the above, together with corrosion of the RPV, PCV,etc.

The findings and issues obtained to date are summarized below:
1) Seismic safety assessment of the reactor buildings taking account of its damage
The following studies have been carried out by TEPCO for the seismic safety assessmentof the reactor buildings of the Fukushima Daiichi NPS Units 1-4.
Evaluation of seismic responses of the reactor buildings under the Great East Japan
Earthquake motions (a huge earthquake of the moment magnitude, Mw, of 9.0) 25,26
Seismic safety assessment of the reactor buildings, taking into account the damages
induced by the hydrogen explosions during the accident 27
Investigation on degradation of the reinforced concrete seismic resistant walls of thereactor buildings 28

25

Nuclear and Industrial Safety Agency "Impacts of the Great East Japan Earthquake on the nuclear
power plants (Seismic response analysis results of the buildings, structures, components and thepiping systems) (Fukushima Daiichi and Daini nuclear power plants),” Dec., 2011 (in Japanese)
26
TEPCO “About the Fukushima Daiichi NPS after the Great East Japan Earthquake -- Impacts on
Units 1 - 3 due to the earthquake” July 24, 2012 (in Japanese)
27
TEPCO “ TEPCO “ TEPCO “ TEPCO “Seismic safety evaluation of the main buildings of the unit 1
through 4 of the Fukushima Daiichi nuclear power plants against the design basis seismic groundmotion Ss TEPCO “”, Study group on monitoring and assessment of specified nuclear facilities of NRA(the 4th), Material 5-1, February 21, 2013 (in Japanese)28
For example, TEPCO “Results of Periodic Inspections to Confirm Integrity of Reactor Building of Unit
4 at Fukushima Daiichi NPS (the 9th)” July 31, 2014 (in Japanese)

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The overviews for each item above are described below.
The acceleration time histories of the Great East Japan Earthquake were recorded on thebase mats of the reactor buildings. The analytical responses of the reactor buildings andmajor equipment under the recorded acceleration time histories were evaluated to be well
below the evaluation criteria, although some of the responses of the bearing walls exceeded
the responses slightly under the current design basis seismic ground motion, Ss(600Gal)25,26.
The seismic safety assessment results of the R/Bs taking into account the damage inducedby the hydrogen explosions under the design basis seismic ground motion, Ss(600Gal),
indicates that the responses of the major seismic resistant walls and the SFPs are below
the evaluation criteria with decent seismic margin27.
TEPCO has evaluated the ground motion (900Gal) based on the new regulatory
requirements for light water nuclear power plants issued by NRA also29, and evaluated thatthat the reactor buildings have seismic safety margins against the 900Gal earthquake
ground motion29. However, the R/Bs and PCVs damaged by the accident have great
difficulties for the repair works and reinforcement due to the high radiation dose environment.
Under such conditions, if the design seismic ground motion and evaluation criteria with highdegree of safety margin are used for the design of fuel debris retrieval work, the whole risk
reduction at the site may be delayed due to the extension of the design and construction
period of the structures and facilities, which have to ensure higher aseismic performances.
Therefore, the design seismic ground motions and evaluation criteria shall be selected
appropriately from the perspective of optimum reduction of the entire risk at Fukushima
Daiichi NPS.
The periodic investigation on the degradation of the reinforced concrete seismic resistantwalls28 has been implemented at Unit 4 reactor building, which seems to be most severely
damaged by the hydrogen explosion, since its radiation dose is comparatively low. According
to the investigation results, no harmful cracks that may cause corrosion of the rebar wereobserved on the major seismic walls and SFP walls. In addition, the concrete compressive
strengths of roughly 35N/mm2 or higher are being obtained in every investigation of theconcrete strengths at seismic resistant walls, which are much higher than the design
concrete strength of 22.1N/mm2 and no sign of degradation has been observed. At Units 13, the similar investigations for cracks have also been progressing to some extent. If thesituations are improved due to dose reduction efforts, in-depth deterioration investigations
should be conducted similar to the case of Unit 4. However, the investigation results of Unit
4 and those of Units 1-3 obtained so far indicate that the effects of deterioration seem to be
not significant.

2) Evaluation of load bearing capacity and stiffness of reinforced concrete RPV pedestal
influenced by high temperature exposure
The severe accident analyses have been conducted for Units 1-3 by using analysis codessuch as MAAP and SAMPSON. The analytical results indicate that the molten fuels may be

29

TEPCO “Considerations on protection against the external events for the Fukushima Daiichi nuclear
power plants ”, Study group on monitoring and assessment of specified nuclear facilities of NRA (the27th), October 3, 2014 (in Japanese)

4-58released from the bottom of the RPV to the PCV bottoms inside of the RPV pedestals.30.
Consequently, the load bearing capacities and stiffness of RPV pedestals are considered to
be decreased due to the effects of exposure to the thermal history at high temperature.
The many experiments using various size and types of specimens under several
temperature conditions have been conducted by R&D activities for the evaluation of the
impact of high temperature exposure followed by moistening by cooling water injection. Oneof typical experiments was conducted using 3 sets of 1/6 scale model of reinforced concretepedestals which were submerged in the water after thermal history of ambient temperature,
400 and 800 deg. C, respectively. Cyclic horizontal incremental loadings were applied up to
the failures of the test models under vertical static loads corresponding to the deadweights
of RPV supported by the pedestals31. According to this test results, the maximum resistanceforce (load bearing capacity) of the test model submerged in the water after the thermal
history of 800 deg. C was reduced to approximately 70% of those of the test model ofambient temperature experience32. However, the load bearing capacity of pedestal exposed
to high temperature of 800 deg. C is evaluated to be higher than the load induced by thedesign basis seismic ground motion Ss. (contribution by IRID’s achievements in thedecommissioning and contaminated water management project in FY 2015 and 2016).
In the future, the load bearing capacities of the pedestals should be reevaluated in detail,according to the knowledge acquired by the further investigation of inside the PCVs.

3) Evaluation of aseismic performance of the submersion and partial submersion methods
based on 1) and 2) above
In R&D activities so far, the aseismic performance assessment for the submersion - top
access method (completely submerged up to the PCV top) and the partial submersion - top
access method (current water level), have been conducted for RPVs, PCVs, and peripheral
equipment and systems against the current Ss (600 Gal). The assessment takes into
account the roughly estimated weight of containers and equipment on the operation floor forfuel debris retrieval, cooling water, and materials of water leak blockage31. (The outlines of
evaluation results for aseismicity of RPVs and PCVs are described in Appendix 41-3.)
Based on the outline results of seismic assessments conducted so far, the evaluation resultof the seismic resistance against the basic design earthquake ground motion Ss 600 Gal foreach method is shown below:
a. In the case of Submersion -Top access method
Because of the large weight of the cooling water in the PCV and equipment and systems
on the operation floor, a certain seismic disadvantage is found compared to other methods.
The details are summarized below.
The reactor buildings have relatively large seismic safety margins, despite the damage

30

IRID “FY 2014, Amended National Budget, Subsidized Project of Decommissioning andContaminated Water Management, Completion Report on Enhancement of Estimation of Internal PCVStates through Severe Accident Progression Analysis and Obtained Data”, March 2016 (in Japanese)

31

IRID “FY 2013 Amended National Budget, Subsidized Project of Decommissioning and Contaminated
Water Management, Progress Report on Development of Technologies for Integrity Evaluation ofPressure Vessel and Containment Vessel”, November 2015 (in Japanese)
32 Masaki, et al, “Strength evaluation of reinforced concrete structures subjected to severe
accident”, Atomic Energy Society of Japan, Fall Convention 2016, PP.3E10 - 3E12

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





by the accident.
The evaluations show that the major parts of the PCVs and RPVs have relatively large
seismic safety margins, even considering the thickness reduction in 40 years due tocorrosion.
For the pedestals, the top access submersion method may have a certain disadvantage
since the earthquake seismic response loads onto the pedestal would be increased dueto the additional weight of cooling water in the PCV and the equipment on operating
floor. However, based on the results obtained by the R&D activities so far, it is estimatedthat the pedestals may have seismic safety margins against Ss 600 Gal, even taking
into account the reduction of load bearing capacities due to the high temperature andthe effect of subsequent cooling water injection during the accident.
It is estimated that the seismic safety margins of S/C supports are relatively small.
However, it has been revealed that the seismic safety margins may be more affectedby the repair methods of water leak blockage applied to the lower PCV due to the weightof the leak blockage material and the water inside S/C. Further efforts have been
made using more detailed analysis models. According to the results obtained so far,
it is evaluated that seismic safety margins may be maintained, such as water leak
blockage inside the vent pipes or strainer water leak blockage done by the partial grout
inside the S/C.

b. In the case of Partial submersion -Top access method
For the major parts of the RPV and PCV, it is evaluated that seismic safety margin can be
secured.


For the S/C supports, the strength assessment is now being conducted using detailed
analysis models, as mentioned in a. above. It is evaluated that seismic safety margins
may be maintained, such as water leak blockage inside the vent pipes or strainer waterleak blockage done by the partial grout inside the S/C.
Regarding the pedestal, there is an advantage of decreasing of seismic response load
onto the pedestal, as the weight of cooling water in the PCV is significantly smaller
compared to that of the submersion method; thus, it is evaluated that seismic safetymargin can be secured against Ss 600 Gal..

c. In the case of Partial submersion - Side access method
Since the equipment and devices for the fuel debris retrieval will be mainly installed on
the first floor of the reactor buildings, a certain advantage is expected in terms of seismic
response loads, compared to the top access methods. On the other hand, the impacts on
the openings to be made in the PCV and the reactor buildings have to be considered.
In the R&D activities so far, although the seismic assessment has not been conducted
for the partial submersion - side access method yet, the seismic safety margin for the
method may be larger than the top access methods, if additional openings of PCV are
relatively small or if existing hatches or penetrations without making new openings areused for fuel debris retrieval.
The seismic safety evaluations mentioned above are obtained based on the limited
information up to now on the actual situations inside PCVs and RPVs. In the future, moredetailed seismic safety margin assessments should be conducted, on the basis ofadditional information obtained from the progress of the investigations and design.
Regarding the assessments of impacts on the important functions of the RPV and PCV
shown in (1) against large scale earthquakes, the study of safety scenarios has been
started pursuing the necessary countermeasures for the prevention and the mitigation to

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maintain the important functions, such as in the case of possible failures of the S/C
supports or the RPV pedestal. In 2017, it is planned to come up with the preliminary results
for the countermeasures for these cases.

B. Development of corrosion control measures for RPVs/PCVs and piping systems and theirverification of applicability to the actual units 33
The one of the concerns at the Fukushima Daiichi NPS Units 1-3 is the progress of corrosionof the RPVs/PCVs caused by exposure to the seawater and to the high temperature histories
due to the fallen fuel debris. At present, at the Fukushima Daiichi NPS, the nitrogen injection
into the PCVs has been implemented to prevent hydrogen explosion, and the dissolving
oxygen concentration underwater decreases in the nitrogen atmosphere; thus, it is estimatedthat the progress of corrosion of steel materials is now being suppressed.
Since the PCVs have to be partially opened to atmosphere during the fuel debris retrieval work
period, corrosion may progress due to the inflow of ambient air and increase the dissolved
oxygen in the water. Progress of corrosion may arouse concern that the seismic strength ofequipment and systems, and the containment function of the PCVs including S/Cs may beaffected. Therefore, it is required to confirm the practical applicability of solutions for corrosion
suppression, in order to prevent the progress of corrosion (both thickness reduction due togeneral corrosion and leakage due to local corrosion) on structural materials such as the
RPVs/PCVs and necessary piping during long-term decommissioning, and to keep the current
status.
A promising solution for corrosion suppression is the addition of corrosion inhibitors into thecirculating cooling water. Among representative candidate corrosion inhibitors which are widelyused for water treatment system, the following four kinds of corrosion inhibitors have been
selected through comprehensive judgment on the effect of suppression for general corrosion
and local corrosion, as well as on the impacts when they are used under radiation environment.
Sodium tungstate (oxide film type)Sodium pentaborate (oxide film type)
Mixed phosphate of zinc and sodium carbonate (precipitation film type)
Mixed phosphate of zinc and sodium molybdate (oxide film + precipitation film type)

In addition, the following studies have been done so far to apply the four kinds of corrosion
inhibitors mentioned above in actual use.
Conceptual design for a corrosion suppression system to inject corrosion inhibitors into
the circulating cooling system (a smaller circulation system)
Review of the procedures for the operation and management of corrosion inhibitors(criteria of determining if a corrosion inhibitor must be injected or not; selection of
inhibitors and the combination of them in terms of the surface temperature of fuel debris,dose rate, etc.; target concentration of each corrosion inhibitor, and others)
Establishment of the target value to reduce concentrations of corrosion inhibitors in thewater to be treated, in order to mitigate impacts on the nuclide removal function of theexisting water treatment system

33

IRID “Subsidized Project of Decommissioning and Contaminated Water Management, Final Reporton Development of Technologies to Suppress Corrosion of Pressure Vessel and Containment Vessel”,
March 20174-61

(4)Future course of actions
The future actions to be taken are described below based on the evaluation results obtained by
FY2016 in order to contribute the decision making on the method of the fuel debris retrieval work.
A. Evaluation on aseismic performance of the submersion and partial submersion methods
Through performing of the following studies based on the results of aseismic performance
evaluation by FY2015, the prospects for the seismic performance of the Submersion andPartial submersion method are planned to be obtained.
1) Establishment of safety scenarios against a severe earthquake event
Impact assessments on the important safety functions of the RPVs/PCVs against a severe
earthquake event are to be performed based on the studies up to 2016. If there is any
possibility of failure of the SC supports, for example, its consequent impacts (ripple effects)
need to be evaluated as well as its countermeasures (preventive and mitigation measures)is to be established (establishment of safety scenarios).
2) Development of the assessment methods of the seismic safety of the components andevaluation method of impacts of their failures to establish the safety scenarios
The development of the methods to assess the seismic safety of components and toevaluate the impact of their failures are to be completed for the PCVs/RPVs, S/Cs and RPV
pedestals in order to establish the safety scenarios described in 1).
3) Improvement of safety scenarios
In order to improve the assessment methods of the seismic safety of the components andevaluation methods of impacts of their failures described in 2), efforts are to be also focusedon the verification of the applicability of the developed methods reflecting more realistic
conditions such as the restraint conditions of components and the conditions for seismic
load evaluation. Furthermore, the detailed analyses and experiments are to be conducted
as needed

B. Development of corrosion control measures for RPVs/PCVs and piping systems andverification of their applicability to actual units Development
It is necessary, for the future fuel retrieval work, to verify comprehensively that the corrosion
control system and other required functions of the circulating cooling system are both satisfied.

4.5.2.5 Reducing occupational radiation exposure during on-site works
The environment inside the reactor buildings is extremely high radiation due to the contamination
as a result of the accident. The retrieval of fuel debris and related work are mainly conducted inside
the reactor buildings. Thus, it is true that reducing site workers’ occupational exposure to radiationis crucial for the successful undertaking of the tasks.
Generally speaking, the key policies on the reduction of radiation exposure are “time, distance andshielding.” However this is only applicable to the situation where the radioactive source isunchangeable. In the case of Fukushima Daiichi NPS, the radioactive source itself should be
removed, and the decontamination (in the broad sense of decontamination including the source
removal) is also an important measure. However, due to the high level of radiation inside the reactor
buildings, the decontamination still poses difficulties in its execution. Thus, measures must ensurethat the duration of work by site workers directly on site is reduced, while remote technologies andappropriate shields are deployed.

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By ensuring an appropriate combination of these protective measures, the occupational exposureof site workers must be reduced, and this is the approach what we should aim for. Therefore,
regarding the reduction of radiation exposure of site workers who are engaged inside the reactor
buildings, it is necessary to consider the decontamination, shielding and remote technologies basedon the following points:












Consider first of all the reduction of exposure to radiation by a combination of remotetechnologies and decontamination. Then, plan on-site radiation exposure management
for site workers by the “time, distance and shielding” approach.
In the extremely contaminated areas such as inside the PCV and torus rooms, work
should be pursued by remotely controlled machines, etc. to avoid engaging personnel
inside.
In the areas inside the reactor buildings other than the above, the best combination of
decontamination, shielding, removal of unnecessary objects, remote technologies andreduction of on-site time, with considerations for the balance between the exposure
during the decontamination tasks and the exposure during the PCV repair tasks, tominimize the cumulative dose for overall work.
Where remote technologies are employed, additional work will be required, such as theinstallation of systems, maintenance and technical troubleshooting, which must be
taken into consideration in the above evaluation and planning.
As for the decontamination tasks, the judgment between remote technologies andpersonnel employment must be made based on factors such as the dose rate in thetarget areas, type of contamination, space for work, frequency of use, deployability of
remote technologies and the stages, process and cost of technology development, etc.
A priority must be placed upon areas where work requirements are clearly identified.
Considerations must not be pursued if task requirements are unclear, or in a nonspecific “betterment-oriented” manner such as to aim for an overall reduction ofradiation dose.

Meanwhile, for the reduction of radiation exposure in the long-term decommissioning, it isnecessary to have an overall perspective and comprehensive, shared measures of exposure
reduction. It is important that actual on-site performance and lessons learned from them provide
feedback to the planning of subsequent steps to improve the accuracy of the plans, and to prevent
recurrence of problems such as delays in work procedures. The above approach must be adopted
to develop systems of cross-sectional exposure reduction management and know-how/technology
transfer.
Regarding the reduction of radiation exposure in relation to the fuel debris retrieval, the major
issues are the decontamination inside the reactor buildings for conducting preparatory and otherwork, on the one hand, and the installation of shielding for executing the retrieval, on the other.
4.5.2.5.1 Reduction of radiation exposure inside the reactor buildings(1)Purpose
By decontaminating the work area and access routes (including the shielding and removal ofradiation source), exposure of site workers must be reduced during their work of the internal
investigation and repair of the PCV, preparation for the fuel debris retrieval, etc.
(2)Major requirements
A. Surveillance of contamination

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Considering the needs in the PCV internal investigation and repair work, investigations into
the state of contamination (types of contamination, distribution, and objects requiring
decontamination, etc.) must be conducted where the existing data are not sufficient.
B. Radiation dose reduction plan
To ensure the work environment necessary for the target work areas by applying appropriate
dose reduction technologies (decontamination, removal and shielding), dose reduction plans
must be put in place based on the states of contamination. The target dose rate in the work
areas must be determined with considerations for the work methods, time and required numberof workers, so that it remains below the maximum exposure level by the law (50 mSv/year and
100 mSv/5 years).
C. Radiation dose reduction technologies
The information about radiation dose reduction technologies must be updated regularly.
(3)Action status, evaluations and issues
A. Surveillance of contamination
 Reduction of radiation dose inside the reactor buildings
Regarding the first-floor of the reactor buildings, where access is indispensable in orderto prepare the work areas necessary for fuel debris retrieval, efforts are being made toremove structural rubbles, decontaminate the floors and low/middle sections, and shieldhigh-radiation spots, in order to reduce the dose rate in these work areas. However,based on the results from the dose reduction measures applied in the first-floor areasof each reactor building as shown in Figure 4.5-10, so far the effectiveness of thesemeasures are 1/2 to 1/4. Further contamination from upper sections and narrow gap
areas, as well as the equipment, pipes and ducts that are difficult to remove, contributeto counter-affect the decontamination. Thus, there are many issues to be addressed inorder to pursue further dose reduction measures. Also, most areas above the second
floor and on the basement levels remain untouched. This situation is making it difficult
to carry out sufficient investigation inside the PCV and in preparation for the PCV repair
work.
 Radiation reduction of high dose pipes
Regarding the first floor of the Unit 1 reactor building, the dose is extremely high in theareas near the atmospheric control system pipes (hereinafter referred to as AC pipes)
and the dry-well humidity control system pipes (hereinafter referred to as DHC pipes),making it difficult to perform work in these areas. It is assumed that the AC pipes
became the radiation source because highly contaminated steam was put through thepipes to execute a suppression chamber ventilation. The DHC pipes are assumed to bea source due to the fact that it is connected to the RCW system (reactor component
cooling water system), which itself is a highly radioactive system, which means
contaminated water remains in the pipes.
 Radiation shielding of penetration for internal investigation
Regarding the first floor of the Unit 2 building, high concentration contamination was
identified on the floor near the X-6 penetration (penetration on the PCV for CRD access),considered to be due to the leakage from sealing on a flange as a result of the removalof shielding blocks (June 2015). After removing the melt leak from the flange, the floor
was decontaminated and grinded the surface to reduce the dose, only not to achieve
the target dose level. Therefore, a dose measurement was taken to distinguish theradiation originating from the PCV seeping through the X-6 penetration and that comingfrom the floor contamination. Based on the results, iron palate shielding was installedon the high dose areas of the floor, and the shields around the investigation device were
reinforced, according to the shielding analysis. The work was also made remotelycontrollable (November 2016). As a result, the dose rate during the work was reduced
to approximately 5 mSv/h as opposed to the planned target of 20 mSv/h (February

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

2017). We have learned from these that it is possible to develop the work environment
by applying appropriate shielding to radioactive sources which present difficulties indecontamination and high dose radiation originating inside the PCV.
Surveillance of contamination distribution over operating floors
On the operating floor of the Unit 3, there were areas of high dose over 1000 mSv/hfollowing the accident. When an investigation was conducted to identify the distributionof contamination, not only the environmental dose measurement, but also energy
spectrum was also measured by collimating radiation to ascertain the directions ofsources and identify the nuclide types, as well as to estimate the locations of sources.
These helped to clarify the areas that needed dose reduction (November 2015). Afterthe decontamination, there were locations with more than 100 mSv/h, and the averageradiation level was approximately 40 mSv/h, but the average was brought down toapproximately 2 mSv/h with the installation of shields. These measures enabled toachieve the dose reduction effectively, and work such as putting in the cover for spent
fuel retrieval was made possible (December 2016). Figure 4.5-11 depicts the dose
measurement results before and after the installation of shields.

B. Radiation dose reduction plan Reduction of radiation exposure during PCV repair work
Concerning the radiation exposure during the PCV repair work, the submersion method
requiring water seal to the upper PCV is expected to result in the radiation exposure ofsite workers to become several times higher than the past annual total exposure doseeven if the dose rate in the work area could be brought down to 3 mSv/h, because there
are many possibly-damaged penetrations from the first to fourth floors of the reactor
building. By comparison, the partial submersion method, with water seal at the lower
PCV, would be able to limit the areas in need of repair, allowing for an estimation thatthe workers’ exposure to radiation is less than the past annual total exposure dose. It isnecessary to take further considerations on these situations in deliberating on the dose
reduction and the methods to be employed. It should be noted that, regarding theexposure during the repair work on the PCV in air space, future evaluations arenecessary in response to requirements of the air space containment system.
 Radiation dose reduction plan and processes
In the preparation for the Unit 2 PCV internal investigation, repeated decontamination
was conducted following the high concentration of radioactivity at the X-6 penetrationon the first floor of the reactor building, but its effect remained insufficient. Eventually,
the plan was reviewed and the remotely-controlled investigation device and shielding
were introduced. This resulted in the achievement of the target dose level, however it
took approximately 1.5 years. The areas of high dose are difficult to investigate to
identify the state of contamination, forcing us to consider and develop decontaminationplans based on speculations. If the levels of contamination are found to be unexpectedly
high after the implementation of the plans, this may cause delays in the processes.
Therefore, the dose reduction plan should be reviewed and modified as necessary.
 Limitations concerning decontamination tasks
In the vicinity of the reactor buildings, there is other work underway simultaneously,
such as the preparation of the operating floor and field preparation for the spent fuelretrieval. Also, tasks that use robots or remote devices also require engagement of site
workers directly on the floor for loading/unloading the device. The time for completingthese tasks are limited, and it is difficult to achieve significant improvement on efficiency.
For the dose reduction task, which should be accorded urgency, processes must beorganized with reference to other tasks that interfere one another, to accomplish thetasks efficiently.
 Accumulation of know-how
At present, the dose level on the first floor of the reactor buildings is more influenced by
the ducts and other features in the middle/upper sections and narrow gaps than thefloor. In the Unit 2 building, the duct louvers were removed and the steam
decontamination was performed inside the duct. Decontamination of the south-side

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

HCU (hydraulic control units) by spray and installation of shields also contributed, andthe overall dose has been brought down below 10 mSv/h, while the areas below 5
mSv/h are growing. These showed the importance of decontamination of middle andupper sections. It is important to accumulate and share such know-how gained on site,
to make future dose reduction efforts more efficient.
Manual tasks and remote-control device
There are areas which necessitate manual tasks, while in other areas where remotedevice is introduced, such a device also involves certain manual tasks. Thus, theradiation exposure of site workers during the decontamination tasks is increasing. It isnecessary to improve working methods to reasonably reduce the manual tasks.

C. Radiation dose reduction technologies
 Accessibility of remote-control devices
As part of the research and development efforts, remote-control decontamination
devices (dry ice blast, suction blast and high-pressure water jet) for the first floor ofreactor buildings (for upper section of the first floor and higher floors) were developed,
and the target decontamination performance was verified through a mock-up test.
However, some issues still remain in terms of driving and access to the destination as
the space is limited with the high density of local equipment.
 Development of procedures and feedback
The dose reduction through decontamination and shielding for the operating floor in the
Unit 3 was completed in December 2016. We had learned from previous experiencesthat re-contamination was prevented by gathering and sucking the structural rubbles
from all over the operating floor first, then applying appropriate covers before moving
on to the grinding/chipping task, that foam decontamination was effective on the rusted
surfaces of metal parts, and that the use of anti-dispersion agent was counterproductive
in decontamination. These lessons informed the operating floor decontamination
procedures that were eventually formulated. It is important to provide these experiences,knowledge and procedures as feedback for treating the units in the future.
 Technical database
Information about the technologies for reducing dose has been compiled and published
as a technical catalog34 by an open call. The R&D and on-site task performance that
followed have also provided technical knowledge and issues. It is important toaccumulate such knowledge, and reorganize it, so that it can be leveraged easily.
(4)Future actions
A. Surveillance of contamination
 Reduction of radiation dose inside the reactor buildings
Clarify the environmental conditions for work required to conduct the PCV internal
investigation, PCV repair work, etc., and continue implementing necessary
contamination surveillance. This must be pursued in conjunction with the effort to lower
the radiation exposure of site workers, by employing auto-drive robots for the survey inhigh dose areas.
 Radiation reduction of high dose pipes
When executing a PCV internal investigation in Unit 1 through the X-6 penetration,
further consideration is required for the decision concerning the internal cleaning of theAC pipes and the removal of the contained water in the DHC pipes, as well as the34

The technical catalog was developed through the assistance program of the Agency for Natural
Resources and Energy 2011 (FY2011 3rd amendment of the “Support grant for the research anddevelopment of technologies concerning accidents of power generation reactor, etc.”). This catalog
has compiled from existing technologies gathered within and outside Japan through research and open
calls, aiming to contribute towards the development of equipment for the tasks involved in the
Fukushima Daiichi NPS decommissioning.

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



removal of these sections of equipment.
Radiation shielding of penetration for internal investigation
For the reduction of environmental dose in the work areas inside the reactor buildings
by means, for example, of shielding cells, the shielding must be executed accurately by
the local dose evaluations, as well as the shield analysis based on the results of themeasurement must be conducted.
Surveillance of contamination distribution over operating floors
Regarding the investigation into the contamination distribution, measurement method,which collimates radiation and measure energy spectrum, must be introduced in thereactor buildings to enable the estimation of nuclide types and source locations.

B. Radiation dose reduction plan Reduction of radiation exposure during PCV repair work
Of the methods for fuel debris retrieval, the submersion method is considered to resultin site workers being exposed to radiation to a considerable extent, as it requires the
PCV repair up to the upper water seal. From the viewpoint of reducing the site worker’sradiation exposure, it may be appropriate to also consider the partial submersion
method as the method for fuel debris retrieval in the future because less exposure
associated with repairing PCV is expected in this method, provided that shielding on
cells and other protective measures are executed accurately.
 Radiation dose reduction plan and processes
As for the areas requiring work such as the PCV repair, plans must be developed foreach reactor with considerations for the amount of work involved, timing to start thework, and site worker numbers required (for the inspection of plantmaintenance/management systems). The on-site experience and learned lessons as
feedback should be added to improve the depth and accuracy of the plans.
 Limitations concerning decontamination tasks
In the work of high-priority decontamination and dose reduction, the time slot for
execution must be secured through the rescheduling of overall work procedures.
 Accumulation of know-how
Regarding the removal of equipment, decontamination and shielding, the knowledge
gained through previous on-site work, such as the efficiency of dose reduction, radiation
exposure levels and processes, should be compiled as know-how and leveraged in thedevelopment of dose reduction plans in the future.
 Manual tasks and remote-control device
Taking into account the performance of, and progress in, the dose reduction of the firstfloor in reactor buildings, the effect of decontamination executed in the building is
considered less capable. While further effective decontamination efforts should
continue, in high-dose areas where the radioactivity levels are unlikely to be brought
down to the target level, shielding as well as the optimal combination between manualtasks and remote-control tasks must be considered for implementation to reduce theradiation exposure of site workers. The remote-control devices to be used fordecontamination and dose reduction should be of low-maintenance and high mobility,
thereby reducing the tasks to be executed by site workers, and thus preventing theincrease of exposure dose.
C. Radiation dose reduction technologies
 Accessibility of remote-control devices
In high-dose areas where the levels are unlikely to be brought down to the target level
through decontamination, certain work methods must be considered for introduction,such as a shielding method with enhanced mobility and a remote-control method withhigh accessibility, to minimize the manual tasks for site workers.
 Development of procedures and feedback
The decommissioning is a long-term project. It requires feedback based on the past

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performance, lessons learned and developed know-how. It also needs procedures andmanuals developed to transfer the technologies, which should be revised and updated
regularly.
Technical database
The data on existing decontamination and remote technologies, and the information
concerning dose reduction technologies, must be added and/or updated as appropriate.
Continuing efforts in reducing dose levels and resolving the issues raised, andcontributing to R&D, the technical knowledge gained through the application at
Fukushima Daiichi NPS should be organized into a database so that these may beleveraged in the decommissioning work that takes a long time to complete.

4-68Figure 4.5-10 State of dose reduction on the first floor of each reactor building35 (Source: TEPCO)

35

TEPCO “Report on the duct inside decontamination of the first floor of Unit 2 reactor building, andprogress report on dose reduction on the first floor of Units 1-3 reactor buildings” METI, Secretariat
Meeting of the Team for Countermeasures for Decommissioning and Contaminated Water Treatment
35, October 27, 2016

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Figure 4.5-11 State of dose reduction on the operating floor of Unit 3 reactor building36(Source: TEPCO)
4.5.2.5.2 Reduction of exposure to radiation during fuel debris retrieval(1)Purpose
To reduce the exposure level in the work area and site boundary as low as reasonably achievable
during fuel debris retrieval, and to maintain the radiation-safe environment.
(2)Major requirements
A. Based on the surveillance results on the distribution of fuel debris, FP, radio activated metal,
etc. evaluate the dose rates in the work area and site boundary according to the work
environment at the time of the fuel debris retrieval. Appropriate measures, such as shielding,must be put in place. Consider as one of the radiation sources about the system equipment
containing radioactive materials, such as radioactive materials collection and treatment.
B. Dispersion of radioactive dust that occurs through removal/relocation of objects inside thereactor buildings, or by cutting out the fuel debris, must be evaluated in terms of the dose
exposure attributed to this, measured in the work areas and at the site boundaries.
Appropriate anti-dispersion and containment measures must be implemented accordingly.
(3)Action status, evaluations and issues
A. In terms of the radiation dose released from the activated reactor structures and fuel debris
during fuel debris retrieval, preliminary evaluations are conducted about the effectiveness ofwater shielding on the operating floor, and the necessary thickness of the shields for cells. It
has been predicted that approximately 1 mSv/h on the operating floor may be achieved
36

TEPCO “Installation of fuel removal cover etc. in Unit 3 reactor building at Fukushima Daiichi Nuclear
Power Station,” Nuclear Regulation Authority, 48th The Commission on Supervision and Evaluation of
the Specified Nuclear Facilities, February 20, 20174-70

through shielding based on a premise that the fuel debris is concentrated inside the reactor
core.37 Furthermore, the task of retrieving the fuel debris will be basically remotely controlled,
and it is possible that no differences may result in dose exposure, depending on work methods
or access routes. It should be noted that the installation of shielded cells still needs to beestablished technically. However, from the experience of shielding the X-6 penetration in the
Unit 2 , pursued for the internal investigation of the PCV, it is considered possible to lower the
local dose down to a tolerable level with the shields installed outside the PCV when applying
the partial submersion-side access method.
B. Regarding the assessment of radioactive dust attributed to the cutting of fuel debris, among
various candidate methods for cutting fuel debris, laser-based cutting/chipping method isunder consideration as a demonstrative option among thermal processing methods, which
are relatively likely to generate radioactive dust through execution. Using test samples withsimilar mechanical properties and thermal characteristics to those of the fuel debris, tests
were conducted both with the submersion and partial-submersion methods to measure theproperties and quantity of dust from the cutting/chipping processes. The data were obtained
about particulate diameter distribution and volume in each method. Effective collection of thedust is yet to be established. In order to evaluate the dose level of generated dust, it isnecessary to obtain information not only about the actual fuel debris’ mechanical properties,
but also its chemical composition as a radioactive source. At the present stage, we use astandard condition that is estimated based on the fuel debris properties obtained from thesevere accident progression analysis. In the future, however, it is necessary to improve theaccuracy of information by obtaining new on-site information from fuel debris sampling, etc.
(4)Future actions
A. For considering the radiation source distribution of the fuel debris, FP and radio activatedmetal, etc., it is necessary to improve the accuracy regarding the radioactive sources
information in each reactor, taking into consideration the progress made with the severe
accident progression analysis and PCV internal investigation. At the same time, as opposedto the submersion method, the partial submersion method, which presents more difficulties interms of shielding, must be evaluated for the shielding measures and dose assessment, toidentify the specifications of shielding regarding shielded cells connected to the PCV in eachreactor, depending on the method to be employed. In this way, a reasonable shielding design
must be developed in planning.
B. In order to realize the partial submersion method, there needs to be a management system
to contain radioactive dust, including the alpha emitting radionuclides. It is thus necessary to
determine the feasibility of the system by means of feasibility study based on existing
dispersion data and obtaining additional data on radioactive dust behavior through simulant
sample tests, and by considering effective dust collection/recovery methods. It should be
noted that the results of the evaluation of dust properties and quantity by each cutting method
shall be used to develop the ventilation and air conditioning systems and filter designs for thereactor buildings and PCV negative pressure control.
4.5.2.6 Ensuring occupational safety(1)Objectives
Most of the tasks envisaged before the fuel debris retrieval will take place inside the reactor
buildings. The condition inside the reactor buildings is narrow with insufficient lighting, structuraldebris scattered, while high levels of radioactivity and dust permeate the environment. Also, thetasks envisaged are unprecedented. It is important not to allow industrial accidents to happen
even in pursuing unprecedented tasks under such harsh conditions.
37

Nuclear Damage Compensation and Decommissioning Facilitation Corporation, “Technical Strategic
Plan 2016 for Decommissioning of the Fukushima Daiichi Nuclear Power Station of Tokyo ElectricPower Company” pp. 4-86, July 13, 20164-71

(2)Main requirements
Upon embarking on the unprecedented tasks to be conducted under the extremely harshconditions inside the reactor buildings, conduct mock-up drills, consider countermeasuresagainst unexpected emergencies, and pursue other preparations, planning and training in
advance, even more than the previous approaches to occupational safety.
(3)Status of action, evaluation and issue
In response to the “Guidelines on Occupational Safety and Health Management at the TEPCO
Fukushima Daiichi Nuclear Power Plant (Notification No. 0826-1)” issued by the Ministry of
Health, Labour and Welfare in August 2015, TEPCO and principal employer together reinforce asafety and health management system, with risk assessments, centralized exposure dosemanagement, and effective exposure reduction measures from early stages of workcommissioning, which are proving effective. In this way, they are pursuing further improvementon the safety and health management measures in preparation for the commencement of fueldebris retrieval. 38










The average daily deployment for FY 2016 was approximately 6,000 with very little
fluctuations, while the industrial accidents (the number of the injuries) decreased from
26 cases in FY 2015 to 20 in FY 2016, as a result of the efforts such as strict compliancewith rules at the power station (the 22 rules of safety, TBM-KY education, strict
observation of 5S), the horizontal sharing of accident case studies, the reinforcementof safety management system, etc. While the number of accidents is reduced, the cause
analysis indicates that accidents by similar causes occurred in succession. Therefore,
further efforts to reduce accidents and improve the working environment continue fromthe viewpoint of “awareness, skill-development and management.”
As part of the initiative to prevent cases of heat stroke, in FY 2016, an additional 6
WBGT (parameters that account for the humidity, radiant heat and air temperature, thethree indexes that have a significant impact on human heat exchange) measurement
displays were installed (7 in total), together with the implementation of the reinforcementof heat adaptation measures and health record checks for past history of heat stroke(screening for compromised health). As a result, there were only four cases in FY 2016,
which is a significant decrease from the previous fiscal year where there were 12 cases.
As for the situation regarding exposure management, the average exposure dose
remains below 1 mSv/month from 2014 to 2016, while the monthly average for FY 2016
alone is 0.46 mSv (estimate at the end of February 2017), well below the target
exposure dose of 1.7 mSv/month.
Regarding the dose reduction measures applied on the site, the decontamination andshielding through removal of high-dose debris and surface ground as well as the facing
work, helped to achieve the target dose rate of 5 μSv/h at the end of FY 2015 (except
the precincts of Units 1 to 4 and contaminated waste depository area). The installationof 86 dose rate monitors was complete in January 2016, enabling a real-time monitoring.
Also, the control zone is now equipped with a train of 10 dust monitors to measure theenvironmental radioactive concentration, and thus the non-full face mask area has been
extended up to 90% of the total site since May 2015.
Regarding the evaluation of internal exposure dosage due to the fuel-debris-derived
radiation such as alpha emitting radionuclides (internal exposure evaluation based on
the bioassay analysis), an agreement has been entered into with JNFL to establish a

38

TEPCO, “Improvement on working environment of workers in Fukushima Daiichi Nuclear Power
Station”, Fukushima Prefecture 2017 citizens’ conference on ensuring safety in decommissioning, May
17, 2015

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system of swift response.
As for the fuel stored in the Unit 3 spent fuel pool, further efforts to improve occupational safety
are being made in preparation for the retrieval work scheduled to commence in the middle of FY
2018, including the operational drill using fuel handling system, which will be installed on site, toenhance the safety measures. The occupational safety concerning fuel debris retrieval is closely
related to the retrieval policies, and as the actual tasks are made more specific, further
considerations will be made regarding the points such as the deployment of remote-operation
technology.
For the establishment of technological foundation for remotely operated equipment, a mock-up
test facility was installed at the JAEA Naraha Remote Technology Development Center during
FY 2015, and a 3D VR system was developed, virtually representing the internal space of the
Fukushima Daiichi NPS reactor buildings, aiming to enable effective task planning andpreparatory training (Figure 4.5-12).
(4)Next steps
 Review the tasks of dose reduction inside the reactor buildings and PCV internalsurveillance so far executed, and leverage them in other tasks in terms of their
preparation, planning and training.
 Concerning the plans of tasks under a high-dose environment in which remote devices
cannot be employed and therefore human engagement is inevitable, an evaluation mustbe given with considerations for justification and optimization, and it is important torealize the safest possible working environment taking multi-faceted approaches into
account.
 In particular, it is vital that the 3H work (hajimete: first time; henkou: change; andhisashiburi: work after a long absence) is extensively dealt with in the mock-up training
so that effective procedures and test methods are developed, implemented andevaluated. Therefore, it is considered to be an important point in the occupational
accident prevention that procedures are repeatedly verified using the VR system, andsafe and reliable work plans are developed with clear hold points. For this reason,
improvement of the VR system is also effective by, for example, updating with the latest
method information and site conditions as appropriate.
 The on-site work, including the preparatory tasks for fuel debris retrieval, is expected toentail the decontamination inside the reactor buildings, surveillance of PCV leakage,
repair work at the bottom and top parts of the PCV, development of system equipment,
installation of equipment and devices for fuel debris retrieval, and otherpreparatory/construction work, followed by the fuel debris retrieval andstorage/transportation/safekeeping of the retrieved fuel debris. For this reason, detailedplans must be prepared for each task step, and preventive measures against possible
accidents and trouble must be put in place based on appropriate prior risk assessments.
Also, considerations must be given to unforeseeable eventualities, such as securing aspace for maintenance at all times so that, in an unlikely event of an accident or trouble,
responses are mobilized immediately.
 By the combination of the containment of radioactive materials, decontamination anduse of protective gear, internal exposure to radiation due to the inhalation of radioactive
materials must be prevented. The prevention of contamination on protective clothes andtools is particularly important in the sense that this prevents indirect internal exposure
by inhaling the radioactive materials from the surface of contaminated protective gear,and external exposure as they land on the skin. Regarding the protection using masks
to prevent exposure through inhalation, it must be managed in terms of the physical andchemical properties of target nuclides. It is particularly important that the protection isprovided based on a thorough consideration on the particle diameters most optimal for

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



the filter’s efficiency.
In the event of an inhalation of radioactive substances, the internal exposure dosage
must be measured and evaluated by means of either the systemic measurement using
a whole body counter, or estimated based on the bioassay and the environmental
radioactivity concentration.
At present, mainly Cs-134 and Cs-137 nuclides are under monitoring, but as we draw
closer to the fuel debris retrieval, it is anticipated that alpha/beta radioactivity will
increase in proportion than the current situation based on the fuel debris compositionassessment. 39 Thus, it is necessary to consider the measurement management toaccount for the mixed nature of radioactivity such as the alpha emitting radionuclides,from the hardware and software aspects, and put measures in place before thecommencement of the fuel debris retrieval.

Figure 4.5-12 Flow of work drill using a virtual reality (VR) system (Source: JAEA)

39

Estimation of fuel compositions in Fukushima-Daiichi Nuclear Power Plant, JAEA-Data/Code2012018 (2012）4-74

4.5.2.7 Establishment of access routes to fuel debris(1)Purpose
With respect to the three methods for a major consideration, the access route inside the building
required for fuel debris retrieval work and the route to access the fuel debris inside the PCV fromthe operating floor or the side of the building are to be established.
(2)Major requirements
A. Requirements common to all methods
 Radiation dose at the working areas inside the building is reduced, the obstacles are
removed, and the access route inside the building is established so as to carry-in/out
and install the equipment/devices as well as transporting the retrieval equipment andfuel debris to be used for the fuel debris retrieval.
 Release of radioactive materials from inside the PCV/RPV is prevented on the accessroute.
 Plans must be developed bearing in mind that the fuel debris may be distributed inside
the RPV or at the bottom of the PCV.
B. Requirements for the top access method (submersion-top access method, partial submersiontop access method) Since it is a method to retrieve fuel debris from Operating floor, the existing equipment
located along the access route from Operating floor to the fuel debris must be removed
before establishing the access route from the upper part of the PCV to the fuel debris.
 In order to access the fuel debris situated at the bottom of the PCV, a plan must be
developed bearing in mind that measures may be required to create an opening at thebottom of the RPV.
C. Requirements for the side access method (partial submersion - side access method) Since it is a method to retrieve fuel debris from the first floor of the reactor building, anaccess route to reach the fuel debris must be established through the opening createdon the side of the PCV. Plans must include, as necessary, an installation of an opening
on the sidewall of the reactor building, and an extension of the opening on the side of
the PCV.
(3) Action status and evaluations and issues
A. Common to all methods
Future considerations are necessary to determine how to execute the reduction of dosage
along the access routes and removal of obstacles from the routes. Establishing of accessroutes to the fuel debris inside the PCV, and the development of crucial technologies to containradioactive release from within the PCV and RPV, for each method are developed as shown
below (contribution by IRID’s achievements in the decommissioning and contaminated water
management project in FY 2015 and 2016).

B. Top access method
1) Consideration on access routes
The top-access method entails accessing the fuel debris from the operating floor. In order toestablish the access route, planning is essential in preparing the environment of theoperating floor.
Concerning the establishing of the access route from the operating floor to the fuel debris,
there is an access route from the operating floor to the reactor core structurally secured to4-75

facilitate the fuel replacement work during the periodic inspection. Access is enabled byremoving the reactor well shield plug, PCV top cover, RPV top cover insulator and RPV top
cover. Further removal of internal structures, namely, the RPV steam dryer and gas/water
separator, etc., will enable an access to the top grid directly above the reactor core, thereby
making access to the fuel debris inside the RPV possible. As for the removal of obstructing
internal structures on the access route, there are two possibilities: removing the structurescompletely, or making an opening of a required size in the structures.
Figure 4.5-13 illustrates major structures inside the reactor that may need to be removed
when pursuing the top-access method.

(The structural drawings are provided by IRID)
Figure 4.5-13 Example of the structures inside the reactor that need removing
when applying the top-access method
When executing fuel debris retrieval, the aforementioned access route for regular periodic
inspections will be used in principle. This requires preparations for work under high-dose
conditions (ensuring the containment function and placing remote operation devices, etc.)
to handle the obstacles on the route, such as the reactor well shield plug, internal structuresand other highly contaminated equipment, depending on the circumstances when executing
the tasks. Furthermore, certain pieces of equipment which should be removed are possibly
deformed as they were exposed to high temperature at the time of the accident. No direct

4-76verification has been made as yet, but it is suspected that the deformation, depending on its
extent, may make the pursuit of the work more difficult.
Given the above, the top-access method possibly entails a major-scale task of removing thereactor well shield plug and other obstructing structures inside the reactor, and therefore thismethod is thought to take longer before the access to the fuel debris inside the RPV isachieved than the side-access method, which takes the route to the fuel debris at the bottomof PCV, which will be described later. Also, using the top-access method to reach thepedestal at the bottom of PCV and access the fuel debris necessitates the removal of reactor
core support structures and creating a large opening in the bottom of the RPV in order tosecure an access route, which may pose a highly technical challenge at this point.
Furthermore, the access to the fuel debris outside the pedestal must be made through the
opening in the pedestal, which is considered even more difficult to do.
In R&D activities, the access routes from the top of the operating floor to the fuel debris areunder consideration. Examples of such access routes are shown in Figure 4.5-14(submersion - top access method) and Figure 4.5-15 (partial submersion - top accessmethod).
In Figure 4.5-14, the illustration shows the method in which the openings necessary tosecure access to the fuel debris are made in the internal structures and other obstacles
using an access device from the operating floor (see the following section for the accessdevice). The retrieved fuel debris is placed into a unit container, and the unit containers areplaced into a storage container on the operating floor, before being transferred. In the areafor loading the unit containers into the storage containers, the radiation containment by cells
is in place (refer to the next section for details on the radiation containment by cells).
Figure 4.5-15 depicts a system, in which the internal structures to be removed are
dismantled, and seals to prevent radiation dispersion are activated inside the RPV (refer tothe next section for details on the seal function of the access device inside the RPV). The
unit containers with retrieved fuel debris are loaded into storage containers in the dryer
separator pool (hereinafter referred to as DSP), and transferred out from the operating floor.
By using DSP, contamination of the operating floor is controlled (refer to the next section for
details on the contamination containment technology using DSP).
Note that the methods depicted in Figures 4.5-14 and 4.5-15 are still at a conceptual stage,
and require future engineering and site-applicability evaluation. The concepts are
considered in terms of the submersion - top access method and partial submersion-top
access method separately, but the technologies themselves may be applicable to both. Thus,
we need to consider further when developing the details.

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4-78

Figure 4.5-14 Example of the submersion-top access method (Source: IRID）

4-79

Figure 4.5-15 Example of the partial submersion-top access method (Source: IRID)

2) Development of technology for the establishing of access routes
Concerning the establishing of access routes from the operating floor to the fuel debris, andthe radiation containment on the access routes, in order to verify the feasibility of major steps,
technological development is underway including element testing as part of the developmentof technology for fuel debris retrieval methods. The progress in the technologicaldevelopment for each method is as follows.


Submersion - top access method
Figure 4.5-16 illustrates the possible key items in considering the feasibility of thesubmersion-top access method.

 Feasibility of shielding functions of
the cells to be installed on theoperating floor Feasibility of functions to suppress
release of radioactive materials

 Feasibility of accesstechnology to the RPV
insideFigure 4.5-16 Items for consideration in terms of the feasibility of the access route by thesubmersion-top access method
Regarding the items shown in Figure 4.5-16, feasibility testing of the platform andcells, and element testing of the shielding port to be installed there, are being
performed to verify the feasibility of the cells to be installed on the operating floor interms of their shielding and radiation containment performances. (Refer to Figure 4.522 (4) under Section 4.5.2.8 and Appendix 4.14 (4) for the outline of the element
testing)Element testing is also conducted to verify the feasibility of the devices used to accessthe inside of RPV to retrieve the fuel debris. (See Figure 4.5-22 (3) under Section
4.5.2.8 and Appendix 4.14 (3) for the outline of the element testing)
The status of the element testing is described in Section 4.5.2.8.
Partial submersion - top access method
Figure 4.5-17 illustrates the possible key items in considering the feasibility of the partial
submersion-top access method.

4-80 Feasibility of the operating floorarea sectioning Technical feasibility to prevent
spreading of contamination

 Technical feasibility to containradioactive materials near fueldebrisFigure 4.5-17 Items for consideration in terms of the feasibility of the access route by thepartial submersion - top access method
Regarding the items shown in Figure 4.5-17, element testing is being planned in whicha scale model of approximately 1/4 will be used, to verify the feasibility of the measures
such as operating floor sectioning and the prevention of spread of contamination fortransferring core structures. (Refer to Figure 4.5-22 (7) under Section 4.5.2.8 andAppendix 4.14 (7) for the outline of the element testing) Element testing is also
performed on the shape-following, light-weight shield, which enables easy installation
and removal of the shields placed between the operating floor and PCV. (Refer to Figure
4.5-22 (6) under Section 4.5.2.8 and Appendix 4.14 (6) for the outline of the element
testing)
Similarly, in order to reduce the size of high-dose and highly contaminated areas upon
the retrieval of the fuel debris, radioactivity containment will be installed near the fueldebris. To verify its feasibility, element testing is being performed for the sealing
technology to apply to the devices used to access inside the RPV under development.
(Refer to Figure 4.5-22 (8) under Section 4.5.2.8 and Appendix 4.14 (8) for the outlineof the element testing)
The status of element testing is described in Section 4.5.2.8.
C. Side access methods
1) Consideration on access routes
The side access methods require an access route from the sidewall of reactor buildings to
the side of PCV inside the building, and an access route further down from the side to theinside of PCV to retrieve the fuel debris.
Regarding the access route inside the reactor building, it is necessary to develop a plan toensure efficient execution of the retrieval, including the creation of an opening in the reactor
building wall and an extension of the opening on PCV. Currently, each unit is underconsideration for the layout based on their respective access route, and two scenarios are
postulated: Plan A is to put an opening on the reactor building to secure the working area for
retrieve the fuel debris, and Plan B is to transfer the retrieved fuel debris through the existinglarge hatch to minimize the impact on the building. Figure 4.5-18 illustrates the example
access routes inside the reactor building for each unit according to these plans.

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4-82

Figure 4.5-18 Side access method – Example access routes inside the reactor building for each unit (source: IRID)

Concerning the establishing of access routes inside the reactor building, it is necessary to
pursue dose reduction and obstacle removal in the areas including the side of PCV where
the fuel debris retrieval device will be installed in well-planed way. Also, the progress inthe dose reduction should be reflected in the access route establishing and considerations
on retrieval equipment as necessary.
As for the route inside the PCV, there are several hatches for equipment transfer and for
CRD that provide access into the PCV, mounted on the side of the PCV. These aresomewhat limited in size as an access opening, but some access routes are structurally
secured. In accessing the bottom part of the PCV, moreover, possible obstacles include
the PLR pumps, valves, pipes and supporting materials outside the RPV pedestal, and
CRD replacement rail, operation floor (grating), etc. inside the RPV pedestal. However,access to the fuel debris at the bottom of PCV becomes possible by severing andremoving them.
Figure 4.5-19 depicts the state of obstacles inside the PCV subject to the side accessmethod.

Unit 1 (Source: Hitachi GE Nuclear Energy)
Figure 4.5-19 State of obstacles for the side access method
In order to establishing access routes to the fuel debris at the bottom of the PCV, it isessential to remove the obstacles inside the PCV. However, since most of the obstacles
are pieces of equipment found at the openings and on the access route, it is consideredless problematic than the case of the top access method, which needs handling of thestructures inside the reactor. Nevertheless, internal surveillance and other confirmation of
the internal state are necessary when developing specific plans. With respect to theaccess into the pedestal for the purpose of fuel debris retrieval and waste removal by theside access method, the extension of existing openings on the pedestal, such as the CRD
portal, and/or creating new openings, must be handled with utmost care from the viewpointof the structural strength. Note also that the openings that may be used by the side accessmethod, such as the one on the side of the PCV, are limited in size. Therefore, it isnecessary to consider how to improve work efficiency, including development ofcompatible retrieval device. The PCV water level at Unit 3 is approx. 6 m at present, andlowering it to the appropriate level is required for retrieval. Thus, it needs to be reflectedto the specific plan.

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Meanwhile, accessing the fuel debris inside the reactor using the side access method is
possibly more difficult to execute at this stage than the top access method from theperspective of safety, as certain measures must be put in place, such as prevention of
structural collapses during work.
In terms of the establishing of access routes for the side access method, it is considered
in R&D activities, as in the case of the top access method, according to respective unit’s
layout (Refer to Figure 4.5-18). Figure 4.5-20 shows the example of Plan A.
Figure 4.5-20 illustrates that the fuel debris is retrieved using robotic arms on the rail
installed from the side into the PCV (refer to the next section for the details of the roboticarms). The unit containers with fuel debris inside are retrieved through the side of the PCV,
then loaded into the storage containers by fuel debris retrieval cells. The storage
containers are packed into the casks through the storage container cells and cask cells,
and then transferred outside the reactor buildings. The side of the PCV and fuel debris
retrieval cells are treated with the radioactivity containing welding technology (refer to thenext section for the details of the technology to remotely weld cells and PCV). Note that,in the Plan B as shown in Figure 4.5-18, openings in the reactor buildings and the PCV
are different from those in Plan A, but the tasks involved in the fuel debris retrieval are thesame.
Like in the case of the top access method, Figure 4.5-20 and other cases of side accessmethods are still at the conceptual stage, and future engineering and site-applicability
evaluation are required.

Figure 4.5-20 Example of the side access method (Plan A; source: IRID)
2) Development of technology for the establishing of access routes
Concerning the establishing of access routes from the side of reactor building to the fueldebris, and suppression of radioactive release on the access routes, in order to verify the

4-84feasibility of major steps, technological development is underway including element testingas part of the development of technology for fuel debris retrieval methods.
Figure 4.5-21 illustrates the possible key items in considering the feasibility of the partial
submersion-side access method.

Figure 4.5-21 Items for considering the feasibility of the access route by the side access method
Regarding the items shown in Figure 4.5-21, there is a need to suppress release ofradioactive materials as part of the establishing of access route. Thus, element testing isunderway for remotely welding PCV and cells to be installed. (Refer to Figure 4.5-22 (11)under Section 4.5.2.8 and Appendix 4.14 (11) for the outline of the element testing.)
Similarly, element testing is being pursued to verify the feasibility of robotic arms and rails
as the technology to access the fuel debris inside the pedestal from within the PCV. (Refer
to Figure 4.5-22 (9) under Section 4.5.2.8 and Appendix 4.14 (9) for the outline of theelement testing.)
For the technology to remove obstructing equipment before the execution of the fuel debris
retrieval inside the PCV, element testing is being performed on remotely operated flexible
arms. (Refer to Figure 4.5-22 (10) under Section 4.5.2.8 and Appendix 4.14 (10) for theoutline of the element testing.)
The status of the element testing is described in Section 4.5.2.8.
(4) Future actions
A. Measures common to all methods
As for the measures common to all methods, feasibility of the access routes in the operating
floor and reactor building should be considered based on the levels of difficulties of the tasks
required as preparatory work, namely, the dose reduction and obstacle removal on site. The
locations of access points for the operating floor and reactor buildings planned as accessroutes should also be considered with reference to on-site conditions around the reactor
building and adjusted at the site as necessary.

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The fuel debris is considered to be distributed mainly inside the RPV and at the bottom of thePCV, as explained in Section 4.3, for which it is considered difficult at present to apply single
method to successfully retrieve both parts of the debris. Therefore, it is necessary to consider
combining the top and side access methods. In this case, considering that the execution of
one method will have an impact on the conditions inside the PCV and RPV, all possible
consequences from the combination of methods must be evaluated in advance, andcountermeasures must be prepared as necessary.
Regarding necessary development of technology through the feasibility assessment of theaccess route establishing, as the consideration on each method through engineering advances,verification through further element testing should be fed back to the consideration on theaccess route establishing in a planned manner, if needs arise.

B. Measures in relation to the top access method
It is necessary to propel the planning of access route establishing on the operating floor in the
top access method.
With regard to the establishing of an access route from the operating floor to the fuel debris, itis necessary to carry out engineering and site-applicability evaluation based on the results ofprevious considerations. It is also necessary to establish the policies regarding the removal of
obstacles from the route between the operating floor and fuel debris (reactor well shield plugs,
internal structures, etc.) in order to construct the access route.

C. Measures in relation to the side access method
It is necessary to pursue the dose reduction and obstacle removal on the route in order toconstruct the access route inside the reactor buildings in the side access method.
Similar to the top access method, it is necessary to pursue engineering and site-applicability
evaluation based on the results from previous considerations, for the establishing of accessroute from the side of the PCV to fuel debris. Note that, with the side access method, the routes
are established inside the reactor building, where the radioactivity level is high, the progress
made in the dose reduction and obstacle removal work should be reflected in the considerationon the establishing of access routes if necessary. It is also necessary to establish the policies
regarding the removal of obstacles from the route to the fuel debris inside the PCV in order toconstruct the access route.
4.5.2.8 Development of the fuel debris retrieval equipment and devices(1)Purpose
To develop equipment and devices corresponding to the site conditions to retrieve fuel debris insafe, proven and effective manner.
(2)Major requirements
Fuel debris is envisaged to be inside the RPV and “in and outside” of the pedestal. Following
requirements shall be satisfied on the occasion of developing the functions (e.g. access device,visual and measurement, cutting/dust collection, prevention of radioactive dust scattering,shielding and storage) required for the said three methods.

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

Since fuel debris retrieval work will be performed under high radiation environment,
remote-controlled retrieval equipment shall be adopted. Although retrieval work isbasically being performed by manual but remotely from a safe area, equipment shallhave protection mechanism such as automatic stop in case of abnormal conditions (e.g.
collision avoidance, overload).

Equipment and devices for the fuel debris retrieval shall be designed to have highreliability/redundancy and also functions corresponding to the environment of each unit
other than high radiation field. And, fail safe concept shall be introduced in the design
to the extent possible. Also, existing proven technologies shall be adapted to the extent
possible.



Although inspection/maintenance works for equipment are basically performed byremote control, manual maintenance works after decontamination and shielding shallbe also considered as necessary.

All necessary repair/recovery work shall be performed without any block caused bymalfunction even if equipment/device malfunction happens during the fuel debris
retrieval work.

Specified radiation resistance shall be fulfilled.



Equipment and devices shall be designed to stand against dusty environment.



Waterproof performance corresponding to each PCV water level shall be ensured.



A view under the muddy water (e.g. suspended matter, cut piece,) shall be secured
during the retrieval work.

Considering the safety against earthquakes, installation systems (including platforms,
cells, cranes and other machines) must be evaluated for seismic resistance.

Considering that the fuel debris retrieval takes place over a long period of time,
efficiency in pursuing the debris retrieval must also be taken into consideration.

(3) Action status and evaluations and issues
The following R&D activities are ongoing as part of the technology necessary for the fuel debris
retrieval work, including the element testing to verify the feasibility of establishing access routes(described in 4.5.2.7), (Refer to Appendix 4.14 for details).
A. Common technical development for all methods


Test with a hydraulic manipulator (Figure 4.5-22 (1))
Element tests such as controllability, repeatability, and operating accuracy are being
conducted by using commercially available hydraulic manipulator to obtain basic dataof the manipulator being used for fuel debris retrieval.

Technical developments for cutting/dust collection, visualization and measurement(Figure 4.5-22 (2))
Equipment for cutting / machining the fuel debris, dust collection, visualization andmeasurement being used in high radiation environment are under development.



Development of handling equipment for fuel debris canister (Figure 4.5-22 (5))
A device which places collected fuel debris in a storage canister closes a lid of canisterand transfer it to the outside is under development. This development can be applicableto other methods.

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B. Technical development of Submersion-Top Access Method

Development of the devices to access to the inside of the RPV (Figure 4.5-22 (3))

The upper part hanging device installed on the operating floor and the lower work stage
used in the RPV (being hanged by the said upper hanging device) are underdevelopment.



Development of platform/cell (Figure 4.5-22 (4))
Platform/cell to be installed on the operating floor are under development.

C. Technical development on Partial submersion-Access Method

Development of light-weight and shape-following shielding (Figure 4.5-22 (6))
Shielding being filled with water only when it is used is under development.



Development of utilizing films and sheets for contamination spread prevention method.
(Figure 4.5-22 (7))
Using a 1/4 scale model, films and sheets for partitioning off working areas are underdevelopment to prevent radioactive dust spread.



Development of sealing technology for devices accessing to the inside of the RPV(Figure 4.5-22 (8))
Sealing system of the access device (which moves up and down inside RPV) is underdevelopment to partition RPV into two areas (upper than the device and lower than thedevice).

D. Development of technology for the partial submersion-side access method

Development of device to access inside the pedestal (Figure 4.5-22 (9))
A device to access to the inside of the pedestal though the opening for CRD
replacement machine located in the lower part of the PCV is under development.



Development of flexible structure arm for remote-controlled work (Figure 4.5-22 (10))
A device to disassemble/remove obstacles in PCV is under development.



Development of the PCV remote seal welding equipment for cells (Figure 4.5-22 (11))
A remote sealing device to connect cells and PCV is under development. On theassumption of utilizing the device at X-6 penetration, the welding test in use of thedevice has been carried out in narrow space.

Concerning the above, currently some certain outcomes are gained through element testing(contribution by achievements of IRID, COMEX NUCLEAIRE, TAISEI CORPORATION and
Hamamatsu Photonics K.K. in the decommissioning and contaminated water management
project in FY 2015 and 2016), and it is necessary to proceed with substantiation of thetechnology and applicability evaluation based on engineering needs, and consider practicalapplication through mock-up testing, and the like. Also, the technologies that are developed
according to the need for each method must be considered for compatibility with other methods.
For the future issues to be addressed in the technology development in relation to fuel debris
retrieval, by evaluation the risk levels (refer to Section 4.4) and ascertain the prospects ofhandling the challenges extracted through the considerations on fuel debris retrieval(considerations on the remote technology for the maintenance of fuel debris retrieval devices,4-88etc.) (contribution by IRID’s achievements in the decommissioning and contaminated water
management project in FY 2015 and 2016). Some examples are described below.
Technology for cutting fuel debris and collecting dust
Preparations are necessary to develop technology for cutting the fuel debris in mass tomake it transferable. Also, it is necessary to consider the technology for gathering
radioactive dust deriving from the cutting of the fuel debris and how to handle the alphaemitting radionuclides during the work.

Collecting technology for retrieving fuel debris
It is considered that the fuel debris exists inside the PCV in various forms (block,particulate, powder, etc.) together with other sediments. For this reason, there is apossibility that the technology should be made applicable to the field relatively soon inearly stages of the fuel debris retrieval.



Remote technology for the devices to retrieve fuel debris
The maintenance of fuel debris retrieval devices will be performed remotely in principle,
because the devices will be installed in high-dose areas. Considerations must be madeon the contamination of the devices themselves, and thus maintenance areas will belimited. Furthermore, it is necessary to minimize the waste deriving from the device
maintenance. Therefore, the maintenance technology must be made efficient andversatile to be able to adapt to device conditions.

(4) Future actions
Consistency with overall plan for the fuel debris retrieval methodCoordinating with overall plan, the followings shall be considered. Assuming the
scenario consistent with overall plan that can perform a series of processconsecutively. Picking up the required element technology. In case that there is any
technology which is not addressed currently and/or new issues from the present
development of element technology, the plan shall be revised as necessary. Also, thedevelopment of element technology required to study the feasibility of the scenario
shall be carried out.

In the Fukushima Daiichi NPS, various inspections/analyses are carried out and new
information and knowledge have been obtained. These information and knowledge
shall be fed back to the development of equipment and device.
Applications of existing (proven) technologies

High reliability shall be required for equipment and device used for fuel debris retrieval.
Although development of new technology may be required, existing (proven)technology with high reliability shall be utilized.

Even if some existing technologies lack in performance, it can be compensated suchas by the operation (maintenance). Development of device/equipment shall bedecided by the results of comparison between newly developed technologies andexisting technologies, from the view point of cost and time perspective.
Implementation of mockup testMockup tests shall be required to confirm the on-site applicability of additionallydeveloped equipment and device.



It is also important to develop human resources who operate the developed





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equipment and devices. Since the actual site is under high radiation environment,
education/training shall be conducted by using mockup.
Effective use of Naraha Remote Technology Development Center of JAEA as amockup test site shall be considered from the view point of reducing time and cost.
Future development of remote-controlled equipment

Regarding the radioactivity resistance, some data are being accumulated about theradioactivity resistance of the electronic parts mounted on robots by past PCV internalsurvey etc. They should be developed into databases and put to good use to future
development of equipment.



Based on case study of success and failure on the past inspection works in use ofrobots, besides with careful consideration of actual work performance (e.g.
decontamination work) under high radiation environment, actual on-site environment
and the robotic technologies in the general industry, the equipment development shallbe executed.

The development must focus not only on individual devices, but also on
supporting aspects such as reliable telecommunication, lighting, and otherinfrastructure.

With regard to the technologies developed independently, it is important to create asystem of utilizing them in appropriate manner according to their advantages.

From the perspective of utilizing robotic technology and remote technology, thedevelopments of equipment, functions and elements that can be commonly usedshall be enhanced in future.



As the direction to take in the development, in terms of the surveillance equipment to
be used for a short time, more weight should be placed on the development ofalternative measures for emergency arrangements, as such equipment is deployedin an unpredictable environment. As for the retrieval devices for long-term use,reliability and robustness should be enhanced based on the environmental dataobtained by surveillance equipment. As the site is a high dose area, a mechanism of
rescue from the site must be put in place in case of malfunctions and faults on site.
Furthermore, considerations must be made on the reduction of waste produced
through device maintenance.



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4-91

4.5.2.9 Establishment of system equipment and working areas(1)Purpose
Given the situation that the reactor buildings and equipment such as the PCV are damaged in
the accident, new equipment, devices and tools will be needed to ensure continued, safe retrieval
and storing of the fuel debris, which is distributed in the PCV and RPV. It is thus required todevelop a system that includes systems to ensure the safety features necessary for operating
those devices appropriately. Also, sufficient areas must be secured for their installation.
(2)Main requirements
A. Identify the system, equipment, devices and tools to consist in a system necessary forpursuing the fuel debris retrieval, and clarify the functional specifications required of each ofthese, before developing the design based on them.
B. These system, equipment, devices and tools must be able to be installed on site and operated
appropriately with reference to the following points.










Additional containers and working cells inside the buildings must have the structural
strength required, and the sufficient area for that must be secured.
A sufficient area must be secured to install the system equipment, devices and tools,
with all the required environmental conditions satisfied according to the design
restrictions of them.
A sufficient area must be secured for operating and maintaining the system equipment,
devices and tools, and installing shield, etc. for maintenance management activities andreduction of exposure dose on workers, etc., with all the required environmental
conditions satisfied.
Where existing equipment is used in the fuel debris retrieval, required functions of theequipment must be ensured taking into consideration the impact of the accident andaging.
A storage area must be secured on site for the fuel debris, severely contaminatedstructures to be exported over the course of the fuel debris retrieval work.
Where the spent fuel pool is deployed in the fuel debris retrieval, the pool must becompletely emptied of spent fuel rods, stored objects such as control rods, andobstacles such as fuel racks, etc. (where the top-access method is applied)(3)Status of action, evaluation and issue
As for the safety features required to ensure safety throughout the fuel debris retrieval, as statedin Section 4.2, we are pursuing some considerations presently. For this reason, in theconsiderations of the feasibility of the system equipment and the preparation of areas, temporary
conditions and evaluation indexes are set up based on the past cases observed at nuclear
facilities and items that should be implemented at present-day Fukushima Daiichi NPS, with thecurrent situation taken into consideration. Based on this, the system component equipment,
devices and tools are outlined and the layout of their installations is deliberated (contribution by
IRID’s achievements in the decommissioning and contaminated water management project in
FY 2015 and 2016).
A. System equipment
Safety features that are considered necessary from the viewpoint of safety assurance during
the fuel debris retrieval are temporarily set up. Table 4.5-2 illustrates an example of thecorresponding system equipment and feature requirements as a result of the consideration.
The specifications for the equipment will depend on the fuel debris retrieval methods and theplant conditions of each unit because they will be derived from further detailed considerations
based on the circumstances of the site in relation to the individual technologies that aredescribed in Sections 4.5.2.1 to 4.5.2.8.

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Table 4.5-2 System equipment corresponding to the safety features (examples considered)postulated for fuel debris retrieval
Safety feature
Undernormalconditions
Prevention ofradiation
leakage in the
air phaseIn
accidents

Monitoring
function

Functional requirement
To maintain negative pressure within the
primary containment boundary (PCV)

System equipment
Negativepressurecontrolsystem/working cells
Radioactive
dust
gathering/processing system
To reduce release of radioactive materials
Fuel debris cutting particles local
collection system
To maintain negative pressure within the Negative
pressurecontrolprimary containment boundary (PCV)
system/working cells
To maintain negative pressure within the Ventilation and air conditioning
secondary containment boundary (reactor systembuilding, etc.)
Reactor buildings/containers
Radioactivedust
To reduce release of radioactive materials
gathering/processing system
To monitor primary containment boundary pressure/secondary containment
boundary pressure/released radiation

To maintain PCV water level
Prevention ofradiation
leakage in theliquid phaseUndernormalconditionsIn
accidents

Prevention offire and
explosion

Monitoring
function
Under
normalconditionsIn
accidents
Monitoring
function

Decay heatremovalCriticality control

To control water level differences between
torus room and ground water
To prevent or mitigate leakages from PCV totorus room
To control water level differences between
torus room and ground water
To transfer and store stagnant water in thetorus room

Circulating water cooling system
Torus room water collection system
Torus room water collection system
Torus room water collection
systemTorus room water level/ground water level
To deactivate by nitrogen injection

Nitrogen supply system

To dilute flammable gas by scavenging

Negative pressure control system

To deactivate by nitrogen injection
To dilute flammable gas by scavenging

Nitrogen supply system
Negative pressure control system

To monitor hydrogen/oxygen concentration in D/W gas

Under
normalconditionsIn
accidents
Monitoring
function
Under
normalconditions

To cool by water injection

Circulating water cooling system

To maintain D/W water level

Circulating water cooling system

To cool by water injection
Emergency cooling system
To maintain D/W water level
Circulating water cooling system
To monitor water injection flow/water level, water temperature and air temperaturein the D/W
To maintain the subcritical state through thecontrol of fuel debris shape or prevention of Neutron absorber injection system
criticality by means of neutron absorbers*In
accidents

Criticality suspension due to the loading ofabsorbers

Monitoring
function

To monitor neutron flux, Kr/Xe concentration in the D/W gas, or D/W water level

Emergency neutron
injection systemabsorber*In cases where subcritical credits cannot be obtained due to the shape, etc.
The progress and challenges of major systems from the above system equipment shown in Table
4.5-2 are as follows.

4-931) Negative pressure control system
An example of negative pressure control system considered on a conceptual level is shownin Figure 4.5-23. There are mainly three parts of systems that make up the system as
explained below. Note that the basic configuration of the system would be compatible withthe three methods, but equipment size may be possibly variable depending on theprocessing capacity of each method.
Negative pressure control system: maintains the negative pressure inside the PCV
using ventilators to ensure the containment function.
 Radioactive dust gathering/processing system: removes radioactive cutting dust andparticles generated through the fuel debris retrieval work from exhaust gas and treatsthem.
 Nitrogen supply system: performs dilution by scavenging and supplies nitrogen into the
PCV to deactivate in order to prevent fire and explosions inside the PCV.
The airflow to maintain negative pressure is considered as a margin based on the area sizeof the damaged area estimated from the present plant conditions (with the supply of nitrogengas the pressure inside the PCV is kept positive). Also, the outline of the system isconsidered based on the assumption about the maximum permissible exposure on siteboundary of Fukushima Daiichi NPS during the fuel debris retrieval work and emergency,
and simulation results show that the impact will not be significant. Further considerations in
the following aspects are necessary in the future.
a. The properties and characteristics of alpha emitting radionuclides (substances originatedfrom the fuel debris) contained in the exhaust gas must be identified, as the dust
gathering and processing system may possibly grow in size due to conservativerequirements because of the lack of knowledge about the properties and characteristicsof the substances that originated from the fuel debris.
b. Planning must be pursued with reference to the on-site situations, such as the positionsof the PCV ventilation and equipment/ducts layout plans.
c. To reduce the impact of further environmental contamination, technology for a system togather cutting dust of fuel debris right near the cutting site must be developed.
d. Exposure assessment must be improved based on the outcomes of technology
development.

Figure 4.5-23 Example of negative pressure control system for the primary containment boundary(PCV) as considered (partial submersion - side access method)

4-942) Circulating water cooling system
An example of the circulating water cooling system considered on a conceptual level isshown in Figure 4.5-24. There are mainly three parts of system that make up the system as
explained below. Note that the basic configuration of the system would be compatible withthe three methods, but system size may be variable depending on the processing capacityof each method.
Circulating water cooling system: pumps up and circulates water from the D/W (or S/C)
to RPV for removing the decay heat and controlling the PCV water level.
Circulating water treatment system: removes and purifies cutting dust and eluted
radioactive materials generated through the fuel debris retrieval work from circulatingwater.
Torus room water collection system: collects the circulating water that leaked into thetorus room under normal circumstances or in accidents together with the ground water
to maintain the difference in the water levels.

Further considerations in the following aspects are necessary in the future.
a. The properties and characteristics of alpha emitting radionuclides (substances originatedfrom the fuel debris) leaked into the circulating water must be identified, as the circulatingwater treatment system may possibly grow in size due to conservative requirements
because of the lack of knowledge about the properties and characteristics of thesubstances originated from the fuel debris.
b. Planning must be pursued with reference to the on-site situations, such as the positionsof the water inlet/outlet of D/W and S/C, and equipment/piping layout plans.
c. If the anti-corrosion agent that was considered for the measures to prevent the progressof corrosion of structures inside PCV/RPV is used, as shown in Section 4.5.2.4, its impact
on the circulating water cooling system must be evaluated, and the results be fed backto the system.

Figure 4.5-24 Example of considered circulating water cooling system(Partial submersion - side access method)

4-953) Criticality control system
As described in Section 4.5.2.3, two possibilities are under consideration concurrently
concerning the methods of controlling the criticality during the tasks of fuel debris retrieval(including the preparatory work): to use pure water as coolant, and to mix sodium
pentaborate as neutron absorber into cooling solution. In the case where criticality control ispossible with pure water, the circulating water cooling system may be simplified comparedto the case of using neutron absorbers.
Supposing that sodium pentaborate is used to control the criticality, on the other hand; wecurrently consider how to maintain the boron concentration in the coolant inside the PCV at
the level of approximately 6,000 ppm by combining the circulating water cooling system andneutron absorber injection system, as shown in Figure 4.5-24. Furthermore, as for thecirculating water that leaks into the torus room, considering the possibility that this water isdiluted with the ground water that flows in from outside the reactor building, the concentrationlevel is adjusted by controlling the neutron and collection equipment, and the water isreturned into Liquid Receiver Tank. Note that different methods result in different watercapacities of the PCV, and these differences may impact the volume of neutron absorbers
to be used and/or size of supply equipment.
Also, it is currently under consideration that either a criticality detector monitor is installedinside the PCV and gas sampling equipment is installed on the negative pressure control
system, and if a critical state is detected, sodium pentaborate is injected by the emergency
boric acid solution injection system to suspend the criticality, thereby containing theabnormal situation. Further considerations are necessary in terms of the following aspects.
a. It is necessary to evaluate the impact of using the neutron absorbers on the circulatingwater cooling system, etc., and if an adverse effect is recognized, countermeasures must
be developed.
b. The methods to control and adjust the concentration levels of neutron absorbers inside
PCV must be considered for feasibility based on the future situations on site.
B. Preparation of areas
We postulated rough dimensions of each device within the system and necessary areas forsetting them up based on the system specifications, and we discussed the layout feasibility,with challenges identified with reference to the areas used by the fuel debris retrieval methods
and dosage information. Figure 4.5-25 illustrates an example of layout for adopting the partial
submersion - side access method. The identified challenges are as follows.
1) Feasibility of the layout of equipment inside the reactor buildings
It is necessary to consider the layout in terms of, for example, securing the space byremoving obstacles and reducing dose level, in light of the current situation of each unit.
2) Consideration on the necessity of additional buildings
For the system that cannot be accommodated in the reactor buildings, additional buildings,
etc. must be considered.
3) Consideration on the feasibility regarding the locations of water sources and outlets andpiping routes
The circulating cooling water pipes may contain high-level radioactive substances.
Therefore, measures to shield and/or prevent leakage must be considered at the same
time.

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4) Inter-building connections in the case of distributed layout
It is necessary to consider some issues deriving from the systems that run across several
reactor buildings, such as installation of connections and connection structures.

Figure 4.5-25 Example of the considered necessary space for the layout of each system (partial
submersion - side access method (Unit 2))(4)Future actions
A. About system equipment
As stated before, we identified some issues with each system through consideration. To
address these issues, it is necessary to pursue technology development in light of thesituations on site.

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Also, measurement of each process data, which will be required from the viewpoint of ensuringsafety, is planned through the system consideration process. Monitoring of internal situations
is indispensable upon the execution of fuel debris retrieval, and the measurement systems
thereby required (visualization, pressure, temperature, radiation, criticality [noble gas
concentration, etc.], hydrogen concentration, etc.) are important future challenges, and thesemust be developed in detail.

B. About the preparation of areas
Regarding the preparation of the areas, we calculated the space necessary for installing thesystems. As stated in the issues identified above, it is necessary to consider the layout withreference to the handling of high-dose areas inside the reactor buildings and interference withother tasks, and include the outside of the buildings into consideration. Once the policies for
retrieving the fuel debris are determined, the area layout must be further deliberated for the
installation and operation of component equipment of the systems, for securing the space fortemporary storage and disposal of retrieved equipment, and for the on-site plot plan for thearea to store retrieved fuel debris.

4.5.3

Technical requirements for storage of fuel debris in safe and stable condition

This section provides an overview of the three technical requirements, i.e., collecting, transferringand storing technologies of the fuel debris, safe handling technologies of radioactive waste
generated during fuel debris retrieval work, and the relevant safeguards, those technologies are
essentially required for fuel debris retrieval work, regardless of fuel debris retrieval methods.
4.5.3.1

Fuel debris handling (collecting, transferring and storing)(1)Objectives
To establish an entire system that covers from designing and manufacturing of canisters forcollecting, transferring to storing of the retrieved fuel debris, in order to bring the retrieved fueldebris under safer and more stable storing condition within the site.
(2)Major requirements
To establish the safe and efficient system for collecting, transferring, and storing of the fuel debris
that fulfills the following technical requirements before starting actual fuel debris retrieval work.

To develop easy-to-handling canisters those are suitable for fuel debris retrieval workand storing of them, as well as fulfil general safety requirements related to sub-criticality,
structural integrity and Hydrogen generation.
 To prepare equipment which enable to transfer the fuel debris canisters safely, and toestablish the flow line of them to the storing facility.
 To design the storing facility of the fuel debris canisters within the site in order to keep
the retrieved fuel debris under stable and safe condition.
(3)Action status and evaluations/issues
A. Development of a comprehensive plan and collection of necessary information (R&D)
As part of research and development activities to date, accumulation of information
contributable to development of the system for collecting, transferring, and storing of the fueldebris, and arrangement of input and output from/to other related R&D projects have been

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implemented. The comprehensive R&D plan was also established. As well, investigation andaction to the related technical issues have been performed. In FY 2016, information regarding
sub-criticality evaluation method, drying technology, countermeasures to Hydrogen generationfrom MCCI products and so on, have been collected from overseas. Collected information is
being analyzed and its applicability to management of the retrieved fuel debris is being
examined.
B. Study on basic specifications of the canisters and system/facility for transferring and storingof the canisters (R&D)
 Development of the canisters for retrieved fuel debris and its safety assessment
methods
 The basic specifications such as design conditions, basic functions, outline shape ofthe canisters and so on, has been investigated in line with progress of study on fueldebris retrieval work process. In particular, the inner diameter of the canisters is being
studied from the viewpoint of its maneuverability and efficiency of collecting fueldebris into the canisters. (See to Figure 4.5-26)
 Technical issues related to the canister design were identified and issues that needmore detailed study were selected. (i.e. criticality evaluation, structural integrity
evaluation and countermeasures against generated Hydrogen) Study on appropriate system/facility for collection, transferring and storing of theretrieved fuel debris
 Rough flow line in collecting, transferring and storing system/facility of the retrieved
fuel debris was created in taking various characteristics and condition of the fueldebris based on current understanding on the situation at Fukushima Daiichi NPS
into account. Movement flow of the canisters within the R/B is shown in Figures 4.514, 4.5-15 and 4.5-20 and that in storing facility is given in Figures 4.5-27 and 4.5-28.
The typical storing methods are compared in Figure 4.5-29.
C. The open space within Fukushima Daiichi NPS is mostly occupied by temporary storage
facilities for contaminated water, spent fuel and radioactive waste. Since space forconstruction of new facilities such as collecting and storing facilities for the retrieved fueldebris is limited, consequently, it is necessary to carefully investigate the plot plan for such
facilities as well as counterplans.
(4)Future actions
Along with cooperation with relevant R&D project such as “equipment development for fuel debris
retrieval”, “fuel debris characterization” and others, it is necessary to continue to advance
realization of the system/facility for fuel debris collection, transferring and storing and to adequately
address the identified technical issues as listed below. In addition, the system/facility for fuel debris
collection, transferring and storing need to be optimized as a whole through cooperation with thestudy on transferring path of the canisters and development of installation plan of storing facility
these will be implemented by the operator.







Study on the general safety requirements and basic specifications related to transferringand storing of the fuel debris canisters
Verification of safety feature of the fuel debris canisters
Review of the fuel debris canisters design in accordance with progress of realization ofprocess from retrieval, collection, transferring to storing of the fuel debris
It is necessary to define the handling and segregation standards for the fuel debris andthe fuel debris-attached structures, and to establish the canisters specifications and thefacility design in consistent with these standards
Although wet and dry storing technologies are considered as a candidate for fuel debris
storing method, detailed engineering work of the storing facility for the fuel debris
canisters and others need to be started. It is particularly important to investigate
applicable drying method for the retrieved fuel debris and countermeasures for Oxygen

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








and generated Hydrogen by radiolysis of residual moisture in the fuel debris.
The open space within the site is mostly occupied by temporary storage facilities forspent fuel and radioactive waste or contaminated water. However, it is necessary tosecure necessary area to install facilities for transferring and storing of the retrieved fueldebris through discussion with overlapped work
The safeguards and collection of samples for fuel debris analysis also need to bestudied in the flow line from collection of fuel debris to its storing within on the site.
Safety requirements related to criticality prevention, radiation shielding, heat removal of
the fuel debris and structural integrity of the canisters, etc. are to be established so as
to deal with the regulatory authority.
The design of continuous series of work from grinding and tipping of the fuel debris,collecting into the canisters within the R/B, transferring, to storing them in the storing
facility needs to be consistent with study results of the related projects. Performance offuel debris retrieval work also need to be evaluated.
As a part of R&D mockup tests of fuel debris canisters and canister handling equipment
designed and manufactured for mockup test need to be performed combined with thefuel debris retrieval equipment/devices.

The retrieved fuel debris collected into the canisters is currently considered to be temporarily
stored within the site, because the Mid- and Long-Term Roadmap states that treatment anddisposal methods of the stored fuel debris will be finalized in the 3rd phase.

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Sub-

Safety Requirement

criticality
Heatremoval

Set inner diameter of φ220/400 ㎜ tomaintain sub-criticality and collecting
efficiency
Assume fuel debris temperature to be
300 ℃

Containment

Maintained by transferring container

Radiation

Maintained by transferring container forshieldingweight saving

HydrogenReleased through filter

Fire

Handling in Nitrogen atmosphere or

prevention

under water

FunctionCapable for both wet/dry methods
Install water injection/discharge port and drain tube
Install gripping mechanism for remote handling

Figure 4.5-26 Outline of canister basic specifications (created from information provided by IRID)

Figure 4.5-27

Figure 4.5-28

Example of flow of wet-type storing

Example of flow of dry-type storing

4-101(provided by IRID)(provided by IRID)

4-102
Figure 4.5-29 Example of storage method (provided by IRID)

4.5.3.2

Handling of waste generated during fuel debris retrieval work

(1)Purpose
To ensure safety, classification and storage of the various waste such as structures to bedisassembled / removed from the inside and outside of the reactor and replacement parts on
construction which occur at each stage such as preparation work, removal work, and cleaning offuel debris retrieval work.
(2)Major requirements
To draft a plan for appropriate classification and storage of the several types of wastes generated
in the fuel debris retrieval work while also ensuring safety, as part of the fuel debris retrieval plan
and waste management(3)Action status and evaluations/issues
Currently, the fuel debris retrieval method is being studied for each direction of access, and theestimated waste generated in each access method is as shown in Table 4.5-3.
Table 4. 5-3 Example of waste estimated to be generated in the fuel debris retrieval work
Type

Specific examples

Disassembly/removal

Top accessWell shield plug, PCV top-head, RPV top-head,of structures within

methodsteam dryer etc.

reactor (metal

Side access

Biological shielding units (concrete), CRD housing,structural material,

methodCRD replacement equipment.

and outside theconcrete)
Secondary waste generated during

Filter for ventilation and air conditioning, water filter,tasks

replacement parts for fuel debris retrieval equipment
etc.

The waste generated at each stage of the fuel debris retrieval work described above isconsidered to vary from a high contamination state to a low one. Further, the estimated waste
also includes those that contain fuel debris. It is necessary to keep in mind the fuel debris retrieval
work plan, and transport and storage plan, to carefully plan waste sorting.
(4)Future actions
As a part of the comprehensive waste management study about each type of waste generated
in the fuel debris retrieval work, a summary evaluation of waste amount generation and safety
will be studied and a plan for appropriate classification and storage will be formulated.

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4.5.3.3

Safeguards(1)Purpose
To conduct a technical study in close cooperation among relevant parties so that a transparent
set of safeguards that are well suited to handling of fuel debris will be put in place before start offuel debris removal.
Operators should establish accounting provisions according to law and report about inventory
and inventory fluctuations of nuclear materials to the national government. In the meanwhile,
based on Japan and IAEA Safeguards Agreement40, the government should report to the IAEA
about the actual nuclear material inventory and indicate that the nuclear material will not be takenout undeclared. It is also necessary for operators to accept inspections by the government andthe IAEA and indicate that they are implementing appropriate accounting and management of
nuclear materials.
In Units 1-3 of the Fukushima Daiichi NPS, it is presumed that the fuel assembly inside thereactor has been fused due to accident, turning into a fuel debris state; therefore, the situation
does not favor diversion of nuclear material to nuclear weapons.
Comprehensive consideration of this situation calls for the need to study the realistic safeguards
measures suitable for the fuel debris retrieval method to be established in future.
(2)Major requirements
A. Study of the safeguards policy for fuel debris Study should be conducted to formulate a realistic safeguards policy, which has been
agreed upon with the Government and the IAEA and applicable for all the process
stages from the retrieval and transport to storage of the fuel debris.
B. Management of the safeguards establishment process
 To clarify matters necessary for safeguards related to fuel debris, a plan should beformulated for establishing realistic safeguards measures and managing theimplementation process.
(3)Action status and evaluations/issues
A. Study on the safeguards policy for fuel debris Current status
 The installation of remote monitor cameras and radiation monitors around the R/Bs
in Fukushima Daiichi NPS Units 1-3, and site inspection by the Government and the
IAEA on short –term notice has been accepted.
 Evaluation and issues
 The fuel debris at Fukushima Daiichi NPS Units 1-3 is presumed to consist of aheterogeneous mixture of metal including control rod and concrete structural materialetc. and quantifying the nuclear material in fuel debris with high accuracy is thought
to be difficult.
 Therefore, in order for the government and the IAEA to verify that nuclear material is
not taken out undeclared, it is necessary to keep in mind the fuel debris retrieval
method for the purpose of showing the proposed safeguards that can be agreed with
the Government and the IAEA.
 Keeping in mind the fact that, as mentioned earlier, it is difficult to quantify presence40

The obligations related to establishment and maintenance of items that require safeguards andnuclear material management system were agreed upon.

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of nuclear material in high precision in the fuel debris in Fukushima Daiichi NPS Units
1-3, a method of measurement control that can be agreed with the Government andthe IAEA needs to be examined.
 There will be timely and active cooperation with the Government and the IAEA to deal
with the technological challenges that come to light, such that there are no key issues
left unresolved in work related to fuel debris retrieval.
 At the 9th Fukushima Task Force meeting held at the IAEQ HQ in the spring of 2017,
information on the current direction towards decision on approaches to the retrieval
method was shared. For safeguards applicable to new facilities, there is an idea ofSafeguards by design (incorporate safeguards right from the design stage). As in thecase of decommissioning, in order to build a safeguards method that is feasible,
efficient and effective, it is important that there is close cooperation among relevant
parties right from the study stage of fuel debris retrieval methodology and its collection,transport and storage system, and conduct technical study in this regard.
B. Management of safeguards establishment process It is necessary to study and implement the safeguards taking into account the Mid- andLong-Term Roadmap as well as the actual work progress.
(4)Future actions
To study and implement the safeguards taking into account the Mid-and-Long-Term Roadmap
as well as the actual work progress.

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4.5.4

Evaluation of feasibility of the methods

This section summarizes the evaluation and status of the individual and detailed technicaldevelopment and study for the nine technical requirements for the three methods executed in 4.5.2((1) Submersion-Top access method, (2) Partial Submersion-Top access method and (3) Partial
Submersion-Side access method) i.e. the six technical requirements concerning safety at work andthe three technical requirements concerning method feasibility. Also, a comparative evaluation of
the feasibility of the current method will be done.
At this point, the development of each method is at concept design and essential technology
development stage where the 'safety requirements value' and 'allowable risk level', which form thegrounds for evaluating a method, are not necessarily clear, and therefore the comparativeevaluation has to be based on the method's feasibility and the level of difficulty. In future, it should
be more quantitative as the method and system equipment take shape.
The evaluation results are shown in table 4.5-4 and a summary of each evaluation is given below:
4.5.4.1

Technical requirements for ensuring safety while retrieving fuel debris(1)Perspective on establishment of containment function
For Units 1-3, there is a high possibility that performance is not at the same level as that of thecontainment function required of normal R/B. In Submersion method, water-sealing by PCV
repair is necessary for withstanding water filling in the core level for contaminated water leakage
control. However, in Partial submersion method, PCV repair is for prevention of entry of air
stream to maintain negative pressure, and it is possible that the difficulty level would be reduced.
In both cases, it is necessary to study containment where negative pressure control of gas phasepart is in conjunction with it. Further, PCV water level is to be necessarily controlled and retrievedfor horizontal access. As for the exact containment function, securing a containment function witha primary containment boundary comprising of PCV and retrieval cell and a secondary
containment boundary comprising of the R/B and container etc. are considered in both methods.
The containment function consists of a "liquid phase" for recovering water containing particulate
fuel debris and a "gas phase" for collecting dust and maintaining the inside at negative pressure.
The feasibility of the water level and doing repair works in the liquid phase and gas phase forbuilding a containment boundary is examined below.
A. Liquid phase containment function
High radiation at the site makes access difficult to the upper part of the PCV (above the firstfloor of the R/B) in each unit, and thus surveillance itself has not advanced. Also, there aremany penetrations that need to be repaired, but it is technically difficult to develop a remote
repair technique suitable for each part, with complete implementation of water-sealing andmaintenance over a long period of time.
Regarding of the vent pipe water-sealing technology, the strainer water-sealing technology andthe downcomer water-sealing technology being developed for repairing the lower part of thePCV (underground floor).
Although, it is in the stage of element testing, if the PCV water level is on the lower side of theceiling of the torus room, it was confirmed that the hydraulic pressure applied to the repair
section is low and out leaks from the liquid phase can be prevented/ suppressed. Now for thewater leakage from the water seal, installation of some underwater pumps in the S/C and4-106

collection of it by cooling water injection system are considered. Although it is necessary tocontinue studying the development of water-sealing technology, it will be possible to obtain thecontainment function in the Partial submersion method with PCV repairs and combination ofan appropriate setting of the water level and collecting system.
Reliable construction techniques such as welding of guide pipes on the S/C or welding of cells
to PCV boundary will be developed continuously.
1) Concept of water level in PCV
Actual water level in PCV is different in Units 1-3. Also, at present, development of water
seal technology is underway, but its applicability will vary from unit to unit. Therefore, thewater level in PCV will be determined taking into consideration the following requirements
comprehensively.
a. Advantage of fuel debris handling in water (prevention of diffusion of radioactive
materials, including Alpha emitting radio nuclides, into air)b. Possibility of reverse rotation of underground water level and water level in torus roomat abnormal conditions
c. Technical feasibility of water-sealingd. Technical feasibility of water level control
In either case, to stably control the determined water level in PCV, it is necessary to havewater-sealing technology for PCV lower part and technology to control the water level.
2) Ensuring water level controllability
At present, water level control is being studied to change the water level and maintain aconstant water level during fuel debris retrieval work. Also, although the vent pipe sealing
technology, which is a water-sealing method to isolate D/W and S/C of PCV, is technicallyvery difficult, it has an advantage that the range of primary confinement can be reduced, andtherefore its technical development is being continued.
B. Containment of gas phase
Static containment boundaries such as RPV/ PCV and R/Bs are built in NPS, however, at the
Fukushima Daiichi NPS where this function has been degraded, a combination of the repair of
the static containment, leaking water collection and negative pressure maintenance is being
considered for building the containment function.
Of these, for containment of radioactive materials dominated by the Alpha emitting
radionuclides, based on the current PCV pressure and nitrogen supply quantity, it is thoughtthat the PCV negative pressure maintenance devices (gas circulation system, exhaustequipment with filter etc.) can be implemented within the size range of systems and devices
for general use. Further, in order to enhance sealability, a method for suppressing in-leaks to
PCV (welding repair, sealant application) is also being developed.
A system for collecting leakage from the primary containment boundary by installing a building
cover and container in the R/B as a secondary containment boundary, and controlling it atslight negative pressure is considered.
(2)The perspective of cooling function maintenance
Maintenance of cooling function is required even while retrieving fuel debris as fuel debris releasedecay heat. However, since decay heat reduces over time, the demand for cooling function also
reduces over time. When considered from the perspective of temperature monitoring of the RPV

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lower head, etc. at present it is in a stable state and that securing the cooling capacity required
for retrieving fuel debris is possible. In the future, the cooling water injection system will be
studied and the start of fuel debris retrieval work will be prepared.
(3)The perspective of criticality control
As the level of possibility of re-criticality can change due to rise of PCV water level or shapechange due to fuel debris retrieval work, the possibility of re-criticality is evaluated under thefollowing premise. Also, the management method is examined (occurrence prevention andimpact mitigation) and the feasibility is evaluated.
Since information on fuel debris is limited, in addition to the method to prevent theoccurrence of re-criticality, a method to prevent impact on people and environment by
combining criticality detection and shut-off technology.
 The feasibility of the management method will be enhanced by reflecting on theinformation regarding the internal PCV conditions obtained in future.
A. The rise of PCV water level Considering contamination of the internal structures of the reactor and FP etc. containedin the fuel debris, it is thought that the possibility of re-criticality is low even if the waterlevel rises to the lower part of the RPV. To reduce the uncertainty of evaluation,
information is being provided to carry out the internal RPV condition analysis, and thefuture information also needs to be reflected.
 Conditions that can cause re-criticality due to a rise in the water level to the reactor core
are limited, but if it is not possible to deny the presence of residual fuel like Unit 2 orUnit 3, it is necessary to consider constant injection of sodium pentaborate or identify
criticality control methods with higher feasibility based on the results of RPV internalsurvey. For Unit 1, it is thought that a small amount of fuel remains in the reactor corearea, and the possibility of re-criticality owing to a rise in the water level is considered
to be low.
B. Fuel debris retrieval work The quantity removed at a time while retrieving fuel debris and the additional reactivity
thereby are evaluated and the requirements for the fuel debris retrieval device andsystem, such as setting a limit value for each retrieval quantity as required are examined.
In future, it is necessary to reflect it in the device design.
 In order to increase the reliability of the criticality prevention during tasks, studies arebeing made toward the development of the critical approach detection technology andthe application of the neutron absorber.
 It is necessary to carefully perform tasks while accumulating information on fuel debris
and flexibly revise the management method based on information obtained according
to the progress of the task.
C. Common items for each task
 Simultaneous injection of sodium pentaborate is being considered in the study for thefeasibility of cooling water injection system.
(4)Perspective of seismic performance evaluation of PCV & R/B
For evaluating the seismic performance of the R/B and the PVC at the time of fuel debris retrieval,in addition to considering the effects of damage due to the accident and subsequent corrosion,
it is also necessary to consider the weight of required equipment facility used to retrieve fueldebris, cooling water, water-sealing material used for at the lower part of the PCV etc.
Based on the preliminary earthquake resistance evaluation result based on the conceptual studyof the fuel debris retrieval method through research and development, and the evaluation results

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of TEPCO, the evaluation results of the seismic feasibility of each method for the design basis
seismic ground motion Ss600 Gal of the fuel debris retrieval method are shown below. (Result
known at present time are written, some are still under evaluation)
A. Submersion-Top entry method
Since the equipment facility weight on the operating floor and the cooling water weight in the
PCV is heavy, compared to other methods, this method has some disadvantages for seismic
performance evaluation. The evaluation based on this is given below.
Even if the R/B underwent damage due to accident is considered, it has a comparatively
large seismic margin.
 Main parts such as PCV / RPV, etc. have been able to secure comparably large seismic
margin even after considering thinning for 40 years.
 For R&D, the high-temperature history of reinforced concrete pedestal at the time of
accident and effects of the subsequent cooling water injections are taken into
consideration, and a reduction in proof stress and rigidity is being evaluated throughexperiments and analysis. Even in that case, evaluation that can secure seismic margin
against Ss 600 Gal has been obtained.
 As for the supports of S/C, seismic margin of the structure is relatively small and rather
than the difference in method, it is affected more by PCV lower part repair method.
Keeping this in mind, intensity evaluation by the detailed analysis model is underway.
At present, evaluation results have been obtained showing that seismic tolerance canbe secured in cases where the repair for lower part of the PCV such as vent pipe watersealing or strainers water-sealing that partially grouts in the S/C is carried out.
B. Partial submersion-Top access method
 For the main parts such as R/B, RPV, PCV etc., results that secure seismic margin canbe secured are obtained.
 For pedestal, there are some merits such as the cooling water weight in PCV issignificantly lower than the Submersion method and the seismic force to the pedestal isreduced, and even if considering history of high temperature at the time of the accident
and impact by subsequent cooling water injection, evaluation that can secure seismic
margin against Ss600 Gal has been obtained.
 For the supports of the S/C, since tolerance is relatively small, intensity evaluation by adetailed analysis model is underway. In the case of water-sealing method for vent pipes
and strainers, an evaluation result that can secure seismic performance evaluation of
the supports of the S/C has been obtained.
C. Partial submersion-Side access method Since the fuel debris retrieval device is installed mainly on the first floor, it is considered
to have greater seismic advantage than the Top access method, but on the other hand,
it is necessary to consider the influence of the opening in the PCV and R/B.
 In the R&D done so far, although earthquake resistance of the Partial submersion-side
access method has not been evaluated yet, if the existing hatch or penetrations areused for retrieving fuel debris without making new openings in the PCV and thereinforced concrete bio-shield wall outside the PCV or even if new openings are added,
there is little impact on earthquake resistance if the opening is so small that it does notaffect the proof stress.


The above is the outline of results of the seismic performance evaluation based on the current
fuel debris retrieval method conceptual study, however, there is a need to conduct moredetailed seismic performance evaluation based on the progress of surveillance and design.

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(5)Perspective of reducing occupational radiation exposure
It is essential to reduce occupational radiation exposure for the workers as almost works will
be performing in the areas of high radiation levels. Accordingly, assessment of radiation
exposure due to PCV repair and preliminary work such as decontamination of retrieval work
area and removing of existing facilities, and the spread of contamination/radiation exposuredue to retrieval activities is necessary.
Regarding the occupational radiation exposure during PCV repairing in the case of
upper PCV sealing for submersion method, it is assumed that the occupational
exposure will be several times of the past annual total occupational exposure since
there are so many potentially damaged penetrations in the first to fourth floor of thereactor building even though the dose rate in the work area can be reduced to 3mSv/h.
Meanwhile, it is possible that the occupational exposure during water sealing for lower
PCV repair for partial submersion method will be the same as or smaller than the pastannual total occupational exposure since there are just a few locations to be repaired.
It is required to further discuss the reduction of dose rates and retrieval methods basedon the assessment above.
As for the occupational exposure during fuel debris retrieval operation, there will not bemuch difference depending on the retrieval method and access route as the retrieval
works must be remotely handled anyway. However it is required to establish the secured
technology for installing radiation shielding containment cells.

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Technical requirements for method(1)Building access route for fuel debris
Fuel debris is believed to exist in RPV internal and PCV bottom part, and access routes need to
be built for each of these sites. An evaluation is being carried out for the feasibility of the threemethods considering their work steps respectively. According to these evaluations, with regard
to the top access route, the scale of work involved in opening the upper part of PCV / RPV toreach the fuel debris inside the RPV and removing the structure inside the reactor is large. In
order to access the fuel debris existing at the bottom of the PCV, it is necessary to provide an
cutting at the RPV bottom head from the operation floor to recover the fuel debris which may be
spreading inside and outside the pedestal. For this reason, this is technically more difficult andtime consuming when compared with the lateral access route.
Meanwhile, since the retrieval of the fuel debris from inside the RPV by side access route isconsidered very difficult at this point, it might be necessary to use the top access route. Further,for side access to Unit 3, it is necessary to lower the water level from its current position in the
PCV. From the above, if considering that there is a possibility of fuel debris existing inside the
RPV and at the bottom of the PCV, there is a need to study both /or a combination of side andtop access.
(2)Fuel debris retrieval components/device development
The development of essential technology (cutting device, visual device, remote-controlled seal
welding for cells in Side-access method, etc.) extracted from the study of major devices that will
perform the fuel debris retrieval from RPV internal and PCV lower portion and the operation step
study is in-progress for each method. And development for practical application will also be
continued.
For development, it is considered that it is impossible for a person to work inside the PCV and
RPV, due to high radiation dose, presence of moisture, high dust concentration, etc. Keeping in
mind that remote work by robot etc. is essential, special robustness, maintainability and rescue
function in the event of malfunction is required in the device.
Also, the issues identified while doing method evaluation etc. (treatment of alpha nuclides for
cutting operation, bulk fuel debris cutting, etc.) need to be studied as key issues.
(3)Perspective of system equipment and area
For fuel debris retrieval work, it is necessary to establish an environment for safely performing
work by maintaining negative pressure in the confining section, controlling the water level in PCV
etc. and cooling and purifying circulating water. A conceptual study was conducted for the various
systems necessary for securing the functions such as "gas phase and liquid phase confinement
function", "cooling", "criticality control", etc., which are necessary for ensuring safety and thespace required for installing each of these systems was considered. Regarding the containment
system for gas phase and liquid phase, an evaluation of the impact of the exposure toenvironment at the abnormal conditions and time of fuel debris retrieval has been carried out.
As a result of the study of the containment system, although there remain issues in handlinglarge amounts of water leakage in an abnormal situation in the Submersion method, trials done
as part of an outline evaluation on impact of exposure to the environment when the system is4-111

applied produced no show excessive exposure impact, thus showing that implementation of themethod is feasible.
However, as there is a possibility of internal exposure due to alpha nuclides, it is necessary tofurther study measures for dose reduction through dispersion prevention. Also, while spacerequired for installation of system equipment is being calculated for area construction, in order toexamine the specific installation area, it is necessary to consider handling of high dose area
inside the R/B and interference with other tasks and installation outside the building too. After
determining the fuel debris retrieval policy, applicability based on the actual condition of the siteshould be considered.
As monitoring of the internal condition is necessary for debris retrieval, currently examination of
all measurements is planned in the system study. Specification of measurement systems(visualization, pressure, temperature, radiation, criticality (noble gas concentration and others),
hydrogen concentration, etc.) will be an important issue hereafter.
4.5.4.3

Evaluation summary of method feasibility

From the viewpoint of carrying out safe and reliable retrieval of fuel debris, the following
summarizes the feasibility evaluation of the three methods.
In the realization of the Submersion method, although PCV repair is required to seal thewater, since there are many portions that may be damaged at the upper side of the PCV,
it is difficult to develop remote technology suitable for each portion, and the total
exposure dose for survey and repairs is also very high. Thus, the feasibility of building
a containment function is low at present.
Meanwhile, since it can be expected that containment by Partial submersion method
can be achieved by maintaining negative pressure inside the PCV, it is necessary tocontinue developing technology of negative pressure maintenance function to containalpha nuclide.
In order to complete the removal of fuel debris from the estimated distribution of fueldebris, at this point it is thought that a combination of top access method and side
access method is believed to be necessary. Surveillance and technical developments
are being continued and an optimal method needs to be flexibly explored.

The above is the assessment of feasibility from the technical requirements assumed at present,
but to perform verification to avoid being at high risk at the time of the task, a risk assessment
method is to be developed for fuel debris retrieval task. Evaluation is conducted keeping in mind
an environment where all preparation tasks such as PCV repairs have been completed byexperienced persons who were involved in the Hanford facility's decommissioning to TMI-2 incident,
within the range of information obtained at present. The features of risk obtained from this risk
assessment and their measures are not largely different from the main study, and, the method itself
can be used for future confirmation of the method and the preparation task, so study shall continue.
(Refer Appendix 4.15)

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Table 4.5-4 Evaluation summary of fuel debris retrieval method feasibility (Mar.2017)
1. Submersion-Top Access Method
2. Partial Submersion-Top Access Method

Study Items

3. Partial Submersion-Side Access Method◎Entering fromoperating floor,
cutting fuel debris
underwater andretrieve it fromoperating floor◎Entering fromoperating floor,
cutting fuel debris
underwater or in the
air with waterrunning on it andretrieve it from◎Entering from side of
PCV, cutting fuel debris
underwater or in the air
with water running on it
and retrieve it from sideof PCV*Red:
Potential Fuel Debris*Blue:
Potential Water
Coverage*Yellow:
Vent Pipes Sealing*Red:
Potential Fuel Debris*Blue:
Potential Water
Coverage*Yellow:
Vent Pipes Sealing*Red:
Potential Fuel Debris*Blue:
Potential Water
Coverage*Yellow:
Vent Pipes Sealing【Descriptions of Methods】

Establishing
Containment
1

Capability

Liquid Phase/
PCV Repair
Water Level
Gas Phase/
PCV Repair

2

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3

Retaining Cooling Capability
Criticality Control

Structural Soundness of
PCV & R/B (Seismic Capacity)
4

5

Reduction in

DuringOccupational

Preparation

Exposure

DuringRetrieval

6

7

Establishing
AccessRoutesto
Fuel Debris

In RPV
PCV Bottom

Developing Devices/Equipment
for Fuel Debris Retrieval
Establishing
NeededSystems/
Areas

Negative
Pressure
Controller

Lower PCV Repair: Development of sealing methods that can endure the water pressure in submerged
condition is underway. It is technically difficult.
Upper PCV Repair: Since there are so many potential damaged openings, it is technically difficult toremotely repair them for sealing. In addition, it has a high risk of leak out of PCV due to large amount ofwater contained.
As there are some gas phase parts on the top, air conditioning system with negative pressure controller
is needed. However, the scale of containment facilities for alpha radioactive nuclides will be smaller asalpha radionuclide concentration in the air is relatively low and the range of gas phase is small due tosubmersion activities.
According to current stable state, it is possible to retain cooling capability that has been required so far.
It is necessary to discuss circulation water cooling system.
Criticality prevention, especially during filling RPV with water, is a challenge since the fuel debris whichis exposed to the air will be submerged. For this reason, it is needed to review the study results assuming
constant injection of boric-acid solution/internal investigation and work with limited water filling speed
and strict supervision.
Although the seismic performance is worse than that of partial submersion method due to increased
weight of coolant in PCV and fuel debris retrieval equipment to be installed at upper R/B, seismic margin
assessment has concluded that major components such as R/B and PCV will have large enough seismic
margin. The assessment has concluded that needed seismic performance will be ensured at S/C legsin the case of water sealing with strainer which grouts internal S/C partially.

When it is assumed that the dose rate in the work area can be reduced to 3mSv/h, the occupational
exposure will be still significantly greater than the past total annual dose rates since there are so many
openings that might be damaged and to be worked on for upper PCV, 1st through 4th floors of R/B, watersealing required by submersion methods. Moreover, it is anticipated that occupational exposure during
additional surveys, decontamination/dose rates reduction work in high-dose radiation areas for PCV
repair and building of access route will be increased.
Although water shielding is promising there might be no big difference between exposure doses ofdifferent methods/access routes as fuel debris, basically, will be retrieved remotely. It is required to
implement shielding measures against retrieved fuel debris after installation of shielding cells.
It might take longer to access fuel debris and large scale of work along with opening of PCV/RPV andremoval of inner reactor structures will be required; top access method might be required for fuel debris
located in RPV.
It is needed to remove fuel debris and inner reactor structures in RPV and install an opening at PRV
bottom in order to access fuel debris located at PCV bottom; the scale of work might be greater thanthat of side access methods.
As a part of method-oriented technologies, development of elemental technologies such as for devices
for accessing fuel debris and cells to be installed at operation floor is underway; it is necessary tocontinue the development for practical use.
Installation and management of the equipment is easier than those for partial submersion methods as
the scope of negative pressure control is the cell containers to be installed at upper operating floor.

Lower PCV Repair: This method has less technical difficulty than submersion method due to lower waterpressure.
Upper PCV Repair: The number of sites to be repaired is limited. In accidents, as runoff water will be
stored in torus room, outward leak will be prevented.

Same as Left

Air conditioning system with capability to keep negative pressure in order to contain alpha radionuclides
generated via leakage is needed. Although the large gas phase requires large scale facilities, it isassumed that environmental exposure will not be too high and feasible. Development of technologies
for inward leakage measures such as welding repair and applying of water sealing materials is underway.
Same as Left

Same as Left

Partial submersion method is better than submersion one; it doesn't require to fill reactor core with water.
Although recriticality is unlikely to take place as work proceeds, it is required to work carefully andaccumulate the data since needed information on fuel debris has not been collected yet.

Same as Left

Although the seismic performance is worse than that of partial submersion-side access method due toincreased weight of fuel debris retrieval equipment to be installed at upper R/B, seismic margin
assessment has concluded that major components such as PCV will have large enough seismic margin.
The assessment has concluded that needed seismic performance will be ensured at S/C legs in thecase of water sealing with strainer which grouts internal S/C partially.

Since fuel debris retrieval equipment will be installed mainly on the first floor, side access method hasan advantage of better seismic performance over top access method. Although the evaluation results
regarding partial submersion-side access method have not been obtained, according to the evaluation
results regarding partial submersion-top access method, it is assumed that required seismic capacity
will be ensured; using existing hatches and penetrations with no installation of new openings in PCV
walls/biological shield walls, or newly installing some additional openings, as long as the additional
openings are small enough not to have any impacts on the seismic capacity, that might not significantly
affect seismic capacity.
Same as Left

The occupational exposure may not be greater than the past total annual radiation exposure since thenumber of the sites to be repaired is limited for lower PCV water sealing. It is anticipated thatoccupational exposure during additional surveys and decontamination/dose reduction work in high doseareas for building of access route will be increased. Future assessment is required for occupationalradiation exposure during PCV gas phase portion repair according to requirements of gas phase portion
containment system.
Although water shielding is not promising there might be no big difference between exposure doses ofdifferent methods/access routes as fuel debris, basically, will be retrieved remotely. It is required to
implement shielding measures against retrieved fuel debris after installation of shielding cells.
Same as Left

Same as Left
It seems difficult to access fuel debris located in RPV via side access method so far; it is possiblecombination with top access will be required.

Same as Left

Scale of work might be smaller than that of top access method if adopting side access method foraccessing fuel debris located at PCV bottom.

As a part of method-oriented technologies, development of elemental technologies associated withfeasibility of sealing system for devices (and around the devices) for accessing fuel debris is underway;
it is necessary to continue the development for practical use.
In addition that the scope of negative pressure control at PCV and R/B etc., and large scale of facilities
and management is required, repair might be needed on an as needed basis. It is, however, possible
to ensure containment capability of gas phase using existing technologies. And, it is also feasible fromthe perspective of radiation exposure evaluation; it is assumed that the level of radiation exposure will
not be too high.
The small amount of water in PCV might require smaller scale facilities than those for submersion
methods. And it is possible to ensure containment capability for liquid phase without inversion ofgroundwater level if PCV water level is appropriately set.
The smaller amount of water in PCV requires smaller amount of boric acid and smaller scale facilities
for provision/treatment
Same as Left

As a part of method-oriented technologies, development of remote sealing technologies associated withthe cells to be installed in R/B is underway; it is necessary to continue the development for practical use.

Circulating
Water
CoolerCriticality
Control System
Establishing
Needed Area
Feasibility Judgements

The large amount of water in PCV requires large scale facilities. Highly reliable repair considering
long-term soundness for upper PCV through-holes is necessary for ensuring containment capability for
liquid phase as inversion of the groundwater level is possible in an accident.
The large amount of water in PCV requires a lot of boric acid and a large scale facilities forprovision/treatment of it.
The area required for system installation has been discussed. Applicability of the area required for waste
storage facilities is to be tested according to the site condition, also.
It is difficult to develop remote leak sealing technologies, and total occupational exposure is too high.

Remarks

1）The study results are based on the information on site condition and development of technologies as of March 2017, and will be revised as further research and development proceed.
2）Subsequent methods might become technically easier as they will be based on the findings through further researches into site condition.

8

Same as Left

Same as Left

Same as Left
Same as Left
Same as Left

Development of negative pressure control technologies for containment of alpha radionuclides is to be continued. Both of top and side access methods might be required.

4.6 Comprehensive evaluation based on Five Guiding Principles
For fuel debris retrieval, technical feasibility for each technical requirement was evaluated in
Section 4.5. In this section, based on the Five Guiding Principles for advancing successive risk
reduction in the decommissioning of Fukushima Daiichi NPS, including the evaluation of thetechnical feasibility in the preceding paragraph, multiple methods are comprehensively evaluated.
Based on the results produced, recommendations for deciding fuel debris retrieval policy for each
unit are compiled in section 4.7.
Fuel debris retrieval policy is the basic concept to start full-scale access to fuel debris in each unit
keeping in view entire process from the start to finish, which will be the fundamental concept for
the fuel debris retrieval method that can be supposed at present. This fundamental concept
includes factors for reducing the risks in the whole process of fuel debris retrieval as well as thelocation from where fuel debris is to be removed, order and combination of access, water level of
PCV at fuel debris retrieval, etc.
4.6.1

Evaluation Method(1)Five Guiding Principles
The following Five Guiding Principles are indicated in section 2.2 as the basic ideas for risk
reduction task in decommissioning. In works which include many uncertainties as in thedecommissioning of the reactor of Fukushima Daiichi NPS, these five points are in trade-off
relationship. Advantages and disadvantages of each of these points must be understood and acomprehensive evaluation must be conducted.






Safe- Reduction of risks posed by radioactive materials and secure work safety
Proven- Highly reliable and flexible technologies
Efficient- Effective utilization of resources (e.g. human, physical, financial and space)
Timely- Awareness of time axis
Field-oriented- Thorough application of Three Actuals (the actual place, the actual parts
and the actual situation)

Among these, "Safety" and "Proven" are examined in the technical feasibility evaluation in Section
4.541. As for 'Efficient', since it is difficult to quantitatively evaluate the scale of the method at present,
judgment will be made on the extent of exposure to workers and the risk of rework due toconstruction.
Required is a process which explores the maximum rationality by performing the engineering work
based on the fundamental concept recommended as the retrieval method.
For 'Timely', evaluation is made by considering the length of time required for constructing theaccess route and whether it is possible to start retrieving fuel debris at a relatively early stage.
Regarding 'field-oriented', although constant evaluation is being conducted based on the range
known at present based on the circumference space, the situation of the contamination inside theR/B, etc., further investigate is required in preliminary engineering as described later.

41

For "Safety", a decision was taken to evaluate how to secure of occupational safety at the stage when
the construction steps will become more specific.

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4.6.2

Evaluation result(1)Evaluation related to water level during fuel debris retrieval
The Submersion method that fills PCV with water up to the height where fuel debris exists has,
not to mention the example of TMI-2, the advantages from the viewpoint of radioactive dust
dispersion prevention and radiation shielding effect 42 , and if possible, the same method is
expected to be adopted.
However, each of the three damaged reactors has many penetration holes with different
accessibilities and structures in the upper part of the PCV. R&D is going on to repair themremotely to avoid liquid phase leakage through these penetration holes. However, in order toapply it practically on-site, it is necessary to solve the problem in terms of both amount of workand performance guarantee. In particular, based on the current condition of decontamination
progress in the R/B, worker exposure during repair is expected to become enormous. From theviewpoint of "Safety" that should reduce worker exposure, application of Submersion method
based on the current technology is considered quite difficult at this point.
Meanwhile, in the Partial submersion method which is a combination of works underwater andin the air with pouring water by controlling water level, the same advantage as Submersion
method is expected in underwater work. In the work in the air with pouring water, as describedin paragraph 4.5.2.9, the containment of the radioactive materials in gas phase by constructing
a negative pressure control system (a system maintaining negative pressure in PCV andcontaining gas for purification) is important.
Thus, assuming the current condition of R&D and that inside and outside of the R/B, it isnecessary to further accelerate R&D towards fuel debris retrieval by Partial submersion method
for all the Units 1-3 and examine its on-site applicability.
Incidentally, it is necessary to continue efforts to lower the dose in building, while keeping in mind
the feasibility of the Submersion method to be discussed again in future. In order to prepare forsuch a case, the knowledge obtained in the previous R&D and the remaining subjects should beappropriately stored as a database for possible future use.
(2)Access route
As shown in paragraph 4.5.2.7, since it is difficult to access and retrieve the fuel debris in the
RPV from the side at this point, access from the operating floor (from above) is necessary.
Also, for the fuel debris at the bottom of the PCV, since retrieval by accessing from the operating
floor (from above) takes time to reach the fuel debris and the distance of the remote operation islong, a high degree of technical difficulty is anticipated. In addition, pedestal is a physical barrier
to retrieve fuel debris outside the pedestal from the operating floor (from above).
Meanwhile, on the first floor of the R/B, all the surrounding seismic walls are sound, and theconcept of shielding cell is being studied to connect the cells to the bio-shield wall and PCV.
Also, from the experience of tasks carried out in the surveillance of the bottom of the PCV, it canbe considered that in case of side access, it can be started with relatively small devices andinstallations.

42

The shielding effect of radiation is an advantage which cannot be ignored from the viewpoint of theradiation resistance of equipment/devices.

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Therefore, fuel debris at the bottom of the PCV can be more realistically accessed and retrieved
from the PCV side (R/B 1st floor). From the viewpoint of reduction of worker exposure andmaintenance, side access can be thought of as rational, and it is necessary to have a moredetailed basic design and confirm its feasibility.
(3)Locations in which to start fuel debris retrieval first
In Units 1-3, though there is a difference in level, it is thought that fuel debris exists both insideof the RPV and at the PCV bottom, and both fuel debris should be retrieved and the risk must be
reduced.
Even if starting with either fuel debris, if amount of fuel debris finally removed is similar, there isno great difference in risk reduction effect.
However, a difference will emerge from the viewpoint of quickly reducing the risk of fuel debris
while minimizing the risk increase associated with retrieval depending on the order of retrieval
from the inside of the RPV or at the bottom of the PCV, and so it is necessary to set the order ofthe retrieval location taking the following points into consideration.
First, the feasibility of realistic engineering is required to be considered. A certain level ofsurveillance is being carried out inside the PCV, but the survey research technology for the insideof the RPV is currently under development, and it will take time to understand the actual situationof the RPV.
Through surveillance at the bottom of the PCV, certain level of knowledge about the side accessroutes to the bottom of the PCV in Units 1-3 was obtained, but the feasibility of the Top accessroute to the inside of the RPV is going to be based on the progress of surveillance to be carriedout hereafter.
Secondly, it is necessary to consider the time period for reaching the fuel debris after starting the
preparation work. In fuel debris retrieval over the long term, taking into consideration therepetition of retrieval work and technical study while advancing several types of surveillance, fueldebris should be collected at as early a stage as possible, analyzed for specific properties,
chemical characteristics, etc. as it is desirable to eliminate uncertainty to the extent possible infuture technical studies. It is expected that the results of the analysis and the examination will
reduce several types of project risks, such as making the fuel debris retrieval work practical on
the whole and decreasing the rework.
Within the RPV, a relatively high radiation dose of Co - 60, Cs - 134, Cs - 137, etc. is assumed
compared to the PCV bottom. Many large-scale structures that become physical obstacles suchas well shield plug, PCV head, RPV head, reactor internals etc. are recognized even before
reaching the fuel debris. That is, when accessing inside of the RPV from above, it is expectedthat it will take longer to reach fuel debris after the starting the task than when side access ismade to the bottom of the PCV.
Thirdly, the rationality of the decommissioning process is required to be considered as a whole.
Although it is considered that taking out fuel from the pool takes priority over taking out fuel debris
from each unit, preparations for side access towards the bottom of the PCV can be done inparallel with the removal of fuel from inside the pool, while ensuring safety in both upper andlower work simultaneously done.

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On the other hand, preparation of top access for RPV internal is expected to interfere with thetask of fuel removal from inside the pool. Also, when accessing the bottom of the PCV, theconcerned condition of pedestal from the viewpoint of structural integrity and the damage
condition of the bottom which would be the next target could be investigated. Furthermore, by
obtaining a certain level of time margin up to the fuel debris retrieval from the inside of the RPV,
it is possible to expect a reduction in dose caused by Co - 60 of reactor internals.
As described in Section 4.4, the risk level of fuel debris is greatly impacted by the existence ratioof forms with high mobility and easiness to be taken not human body, such as fine particles,
sludge, aerosols, etc.
At the bottom of PCV, there is concern about the presence of MCCI products with high uncertainty
in its property, and the possibility of existence of fuel debris affecting PCV boundary and pedestal,etc.
The risk level of fuel debris fallen into PCV can be said to be high due to these factors of
uncertainty. At present, not sufficient information has been obtained, and therefore, it would bemeaningful to understand the level of risk by investigating the condition and properties of the fueldebris at the bottom of PCV in each unit.
Regarding the amount of fuel debris to be retrieved, it is assumed that, in Units 1 and 3, PCV
bottom contains a higher amount of fuel debris as compared to inside RPV. In Unit 2, on the other
hand, it is assumed that the quantity of fuel debris inside RPV is higher, and the one is low in
PCV bottom. Results of Unit 2 PCV internal survey conducted recently suggested that a certain
amount of fuel debris was at the bottom of the pedestal.
From the above discussion on the location of first approach of fuel debris retrieval, considering
several types of factors comprehensively, it is judged that the fuel debris retrieval should bestarted towards the bottom of the PCV first.

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4.7 Proposal for deciding fuel debris retrieval policies and efforts after decision(strategic proposal)
4.7.1

Proposal for deciding fuel debris retrieval policies

For "Determination of fuel debris retrieval policies for each unit", there are many elements ofuncertainty regarding the situation within the R/B; however, the fuel debris situation in each unit
has been gradually clarified and results were obtained to some extent in the R&D for fuel debris
retrieval.
In considering the current fuel debris retrieval policy, it is necessary to have the following points in
mind.



It is necessary to reduce the potential risk of fuel debris as soon as possible.
Information pertaining to the fuel debris of Fukushima Daiichi NPS, development oftechnology for retrieval etc. are still limited and there is a huge uncertainty in studying
the retrieval now.

A comprehensive plan which is totally optimized from the preparation work to retrieval work,transport/treatment/storing and cleaning up is required to study the fuel debris retrieval that aims
at early risk reduction. To steadily advance in such a study, it is important to work on the premiseof high uncertainty and develop a precise short-term plan. At the same time, it is also important todevelop a long-term plan that shall proceed in a framework, which can respond flexibly touncertainties from a panoramic view.
For this reason, study of fuel debris retrieval involves determining the safest and most probabledirection (basic concept) at that point of time. Based on this, the engineering task for preliminary
determination of the feasibility of construction (hereinafter referred to as "preliminary engineering")
is required to be carried out; this shall be carried out prior to the basic design, which is normallydone when starting construction. After that, the internal survey results, which are obtained in parallel
while progressing with the access and retrieval of fuel debris, should be reflected occasionally inthe subsequent access and retrieval. This helps in increasing the accuracy of fuel debris retrieval.
In other words, the actual fuel debris retrieval work shall be in tandem with the surrounding
conditions and surveillance of the next target; the amount of necessary information should be
increased gradually by a step by step approach in which the scale is expanded gradually while
analyzing successful cases whatsoever small, and proceed flexibly according to the situation.
It is necessary to have a mutual understanding of this idea when progressing with the study of fueldebris retrieval at Fukushima Daiichi NPS.
Based on the above, the recommendations for deciding upon the fuel debris retrieval policy are
given below.
(1)Develop a comprehensive fuel debris retrieval plan aimed to optimize the entire retrieval process,from preparation work and transfer from the site to treatment, storage and cleanup, including
coordination with other works in the field.
(2)Move forward in a flexible manner according to the information gained little by little via a step bystep approach after deciding the retrieval method to be focused on.
 The results of PCV surveillance and RPV surveillance to be implemented in the future
shall be reflected in the study of the fuel debris retrieval method in a timely andappropriate manner; the retrieval work itself shall be in tandem with the surrounding
conditions and the fuel debris surveillance, which shall be the subsequent target.
 In the fuel debris retrieval work, small-scale retrieval should be started, and verification
and review of the specific method of retrieval should be conducted based on acquired
fuel debris properties and work experience and newly obtained internal situation, etc.

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The scale should then be expanded gradually and moved to large-scale retrieval.
(3)Assume that combination of a variety of methods will be required to complete the fuel debris
retrieval.
 Fuel debris is assumed to be found in the bottom of the PCV and inside RPV
respectively in each unit. However, it is at present difficult to retrieve both efficiently witha unified method. It is necessary to combine the side access method that is effective in
retrieving fuel debris from the bottom of PCV and top access method that is
advantageous for the debris inside RPV.
 To give shape to this method, it is important to progress with R&D for the future, flexibly
capture various ideas in Japan and overseas and search for optimum methods.
(4)Promote preliminary engineering and R&D focusing on the partial submersion methods.
 Based on the studies undertaken so far, the Submersion method poses tough technicalchallenges at present, from the viewpoint of reducing the exposure at the time of thetask and the technical difficulty of water-sealing for the upper part of the PCV. However,keeping in mind the merits, it is important to consider a discussion in anticipation of theadvancement of R&D in the future.
(5)Firstly, focus on retrieving the fuel debris located at the bottom of the PCV and keep reviewing
the methods based on the newly gained expertise/experience through the retrieval.
 The fuel debris distribution for each plant is estimated by comprehensively analyzing
and evaluating the results of on-site inspection such as muon measurements andcontainment vessel surveys, analysis results of plant data, results of accident
progression analysis, etc.
Unit 1: It is thought that fuel debris exists at the bottom of the PCV.
Unit 2: It is thought that some fuel debris exists at the bottom of the PCV too.
Unit 3: It is thought that fuel debris exists at the bottom of the PCV.
 From the fuel debris surveillance experience and technical development situation, it isreasonable to first consider the approach for dealing with fuel debris in the PCV as
primary from a safety viewpoint.
 The information and knowledge obtained during the primary work of surveillance andthe fuel debris retrieval work at the bottom of the PCV helps in understanding thesituation in the RPV and contributes effectively to the comprehensive study of fueldebris retrieval.
(6)At first, focus on the route from the side of the PCV (the side-access method) for the first accessto the fuel debris located at the bottom of the PCV. The following are the points to be stressed
about the method.
 The side access method has the best accessibility to the bottom of the PCV and could
enable moving to fuel debris retrieval even earlier.
 However, it is essential to be aware that the following issues have to be cleared to use
the side access method.
A. Reduction of radiation on work area
The environmental radiation dose of first floor in the R/B is particularly high in Units 1 and 3
when compared to Unit 2. Therefore, it is necessary to adjust access route, reduce the dose,and develop and apply remote technology to avoid excessive exposure to the workers.
B. Establishment of water level control technology
Even during the fuel debris retrieval work, water level control is important as water supply
continues for cool-down and showering onto fuel debris. In particular, the water level at Unit 3
is higher compared to Units 1 and 2 and it is necessary to reduce its water level first. It is also
important to set the water level applicable and suitable to Partial submersion method according
to the situation of each unit.

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C. Establishment of cell connection technology and securing of area
It is necessary to connect the hermetic cell to the PCV side and take all possible measures forthe containment of radioactive materials represented by alpha particle.
As mentioned above, the issues for each unit are to be identified in the early stage through thepreliminary engineering etc. and be solved through R&D activities. At the same time, its
feasibility is required to be identified to contribute to the final determination of the fuel debris
retrieval method.

4.7.2

Efforts after deciding fuel debris retrieval policies

Initiatives should be taken with focus on the following items after deciding upon the fuel debris
retrieval policy, for "Determination of fuel debris retrieval methods for the first implementing unit"
following the fuel debris retrieval policy decision and for acceleration of the design of the actual
construction plan.
(1)Preliminary engineering
Based on the fuel debris retrieval policy, in preliminary engineering, on-site applicability of theresults of R&D and system concept analysis will be studied, and the fuel debris retrieval process
will be defined.
In the on-site applicability study, in addition to considering infrastructure development,
maintenance, etc. it is necessary to properly grasp and study the site condition, understand theplacement, flow lines, etc. inside and outside of the building and predict the feasibility of themethod so that there would be no rework in the basic design stage. Furthermore, to implement
the side access method, environmental improvement of the first floor of the R/B including dose
reduction will be an important theme of preliminary engineering.
The fuel debris retrieval method will be revised as required based on the results of preliminary
engineering.
(2)Acceleration of the technical development and its practical application through selection andrefinement of R&D
Advancing R&D is necessary to determine fuel debris retrieval method and implement the actual
work that follows. Below is of special importance.


Additional implementation of PCV internal survey
Constructing an access route that is larger than the one currently used in
PCVinternal survey helps in relaxing the dimensional constraints of the introduced
measuring instruments and facilitates an even more detailed internal survey.
Implementation of RPV internal survey
In addition to the development of a method to survey the inside of the RPV by making
a hole from the operating floor, the development of the survey method by drilling a holefrom the PCV side should be carried out. This will help in obtaining information more
quickly.
Identifying the feasibility of the alpha nuclides control system required for the Partial
submersion method.
The control system that contains the alpha nuclides (negative pressure control systemand cooling water injection system) is necessary to achieve the Partial submersion
method. The study should be started based on existing airborne dispersion rate data
and underwater transfer rate data. On the other hand, it is necessary to acquire detailed
data on dust behavior and confirm feasibility of the system to optimize and advance thesystem.

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







Promotion of R&D necessary for the implementation of the side access method
As part of the R&D required for the implementation of the side access method, PCV
water level control technology, and PCV and shielding cell connection technology forsecuring containment function etc. are to be continued.
It is also necessary to conduct R&D etc. regarding collection of fuel debris, preparationof system for transport and storage, preparation of storage facility, R&D for waste
generated from retrieval work, all of which would regulate fuel debris retrieval work
efficiency.
Foundation and basic research required for understanding the property and behavior offuel debris should be promoted intensively after setting the time axis useful for the actual
fuel debris retrieval work.
Importance of mockup
Mockup test is mandatory before deploying the developed equipment and components
on site. The facility and installations for the mockup test are envisaged to be large-scale
and the preparation should be started at an early stage.

(3)Path to commencement of fuel debris retrieval
To start the fuel debris retrieval process, in addition to preparing the fuel debris retrieval
equipment, activities such as on-site verification through internal survey, on-site environment
improvement, preparation of storage facility after retrieval, etc. and related permissions andimplementation are also required. It is also necessary that a draft plan based on practicalsafeguard measures is established in view of the actual circumstances at the Fukushima
Daiichi NPS.
For this reason, in the interest of finalizing a fuel debris retrieval policy, communication
between domestic regulatory authorities and international organizations such as the IAEA is
important.
Although there are various considerations as mentioned above before starting fuel debris
retrieval, the following points should be kept in mind when proceeding with the fuel debris
retrieval project.
 Consideration regarding sustainability of the project
The fuel debris retrieval work is a long-term project that spans from task preparation to
retrieval work as well as collection, transport and storage. Therefore, it is necessary to
pay attention to the continuity of project itself. For this reason, it is necessary to set up
an environment where the tasks (preparation of installations with high maintainability)
and management (resource management, human resource development, technicaltransfer, etc.) can be continued in a stable manner. This will help in achieving the targetof a more stable and controlled state through fuel debris retrieval as early as practicallypossible.
 Total optimization
For optimal implementation of fuel debris retrieval from the viewpoints of "Safety","Proven", "Efficient", "Timely", "Field-oriented", it is not just enough to give consideration
to unit-wise optimization (sharing of technology and equipment facilities between fueldebris retrieval at PCV bottom and fuel debris retrieval within RPV), but it is also
necessary to take into consideration the integration of installations between the units
and the optimization of task for decommissioning across the Fukushima Daiichi NPS.
In addition, following the initial unit retrieval, it is necessary to aim for optimization of the
project as a whole from viewpoints such as which part of fuel debris of which unit would
be reasonable to be subsequently targeted, rationality of the task, utilization ofexperience, etc.
 Close communication with the local people and the society
When formulating and implementing the operation plan for fuel debris retrieval, it isnecessary to place emphasis on ensuring safety while also taking into full account theunderstanding of the local people and the society from the security perspective.

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Fig 4.7-1 Path to commencement of fuel debris retrieval

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5. Strategic Plan for waste management
5.1 Study policy on the Strategic Plan for waste management
Concerning the area of waste management, the Mid- and Long-Term Roadmap specifies that thebasic concept of solid waste management should be compiled in FY 2017 and that the prospectsof a processing/disposal method and technology related to its safety should be made clear byaround FY 2021. The waste processing and disposal is a long-term project, characterized by thefact that the prospects for implementing the final disposal needs to be identified while a risk
reduction is achieved in each stage of the commitment.
Solid waste43 from the Fukushima Daiichi NPS accident is deemed different from waste generated
by normal operation of nuclear power stations. The characterization of solid waste are being
continued, future plan for processing and disposal of it is under study, and solid waste storage
measures are being performed, in accordance with the principle of risk reduction.
Under such circumstances, it is important for us to accumulate information on the characteristics
of solid waste and to summarize the principles of securing safety in radioactive waste management
to show the policy for dealing with matters that may affect future solid waste management.
From this point of view, we will examine the Strategic Plan for waste management in the following
procedures:(1)The principles of securing safety against radioactive waste management compiled
internationally are summarized to clarify matters to be considered when managing solid waste
resulting from the accident at Fukushima Daiichi NPS.
(2)Current situation of efforts on solid waste management based on the Mid- and Long-Term
Roadmap is shown. Taking this into account, issues that may affect future efforts are extracted.
(3)Based on the internationally shared principles stated in (1) above and the issues extracted in (2)above, recommendations that contribute to compilation of basic concepts of processing anddisposal of solid waste are made.
In accordance with this procedure, this chapter is composed of the following sections.
Section 5.2 discusses how the safety of radioactive waste is ensured in other countries, which will
be helpful to examine how to manage solid waste.
Section 5.3 summarizes the current status of solid waste management measures exercised basedon the Mid- and Long-Term Roadmap.
Section 5.4 discusses the characteristics of solid waste that are presumed from the current statusof the management efforts discussed in Section 5.3. Based on those characteristics, and based on
the safety principles summarized in Section 5.2, solid waste management policies and specificmeasures to be taken according to the policies will be proposed.
The Strategic Plan may be reviewed and more specified according to the progress in the approachof waste management.

43

The term solid waste in the Mid- and Long-Term Roadmap refers to the waste consist of “rubble, etc.”
generated after the accident (although some of them may be reused on the site and may not be
considered as waste or radioactive waste), and solid radioactive waste that have been stored in the
Fukushima Daiichi NPS site before the accident. ”This Strategic Plan uses the term “solid waste” withincluding “secondary waste from water treatment” which is to be managed based on the same policies.
Fuel debris, however, is not included.

5-1

5.2 International safety principles on radioactive waste management
An international convention 44 intended to achieve and maintain high level safety in the
management of radioactive waste45 etc. requires that proper measures be taken in every stage ofradioactive-waste management to ensure that individuals, society, and environment are properly
protected from hazards of radioactive materials and non-radioactive substances. The basic
principle of ensuring safety on radioactive waste management has been built internationally
including the experience on countermeasure of managing radioactive waste resulting from
accidents occurred before the case of Fukushima Daiichi NPS. The radioactive waste specified tobe controlled include waste resulting from accidents.46
Therefore, for Japan that has never experienced major nuclear accidents like Fukushima Daiichi
NPS accident, it is helpful to examine solid waste management based on the international principles
of radioactive waste management and radioactive protection.
The international principles on ensuring safety of radioactive waste management compiled
internationally by the IAEA and ICRP are introduced below.
5.2.1 Basic principles for ensuring the safety of radioactive waste
Unless people and radioactive materials are close, people will not be exposed to radiation.
Accordingly, the IAEA Safety Requirements47 SSR-5 “Disposal of Radioactive Waste”48 specifies
the following basic principle of ensuring safety concerning radioactive waste:

The preferred strategy for the management of all radioactive waste is to contain it andto isolate it from the accessible biosphere. This strategy does not preclude thedischarge of effluents, arising from waste management activities, that contain residual
amounts of radionuclides, or the clearance49 of the materials that meet the relevant
criteria.
Radioactive waste may arise initially in various forms. In waste management activities,
the waste is generally processed to produce stable and solid forms; and reduced involume and immobilized, as far as practicable to facilitate their storage, transport anddisposal.

44

Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste
Management (1997). (Japan joined this convention in Nov. 2003.)
45
IAEA defines radioactive waste management as all activities, administrative and operational, that are
involved in the collection, segregation, pretreatment, treatment, conditioning, transport, storage anddisposal of radioactive waste.
Cf. International Atomic Energy Agency, Radioactive Waste Management Glossary, IAEA, Vienna（2003 Edition）p. 50.
46International Atomic Energy Agency, Predisposal Management of Radioactive Waste, IAEA Safety
Standards, No. GSR-Part 5, IAEA, Vienna（2011）47
The IAEA's safety standards are categorized into three levels: Safety Fundamentals, Safety
Requirements and Safety Guides. Safety Fundamentals present basic safety objectives and ten
principles on protection and safety to form the basis of safety requirements. Safety Requirements
define current and future requirements that must be satisfied to ensure public and environmental
protection. Safety Guides provide recommendations and guidance on how to comply with safety
requirements and present an international consensus that it is necessary to take recommended
measures.
48
International Atomic Energy Agency, Disposal of Radioactive Waste, IAEA Safety Standards, No.
SSR-5, IAEA, Vienna（2011）49
The term clearance is used in Japan to mean that, if the dose due to the radioactivity of a small
amount of radioactive substances included in a given material is sufficiently small compared with theradiation level in the natural world, and if its risk to human health is deemed negligible, the material is
excluded from the scope of regulation framework because it is not necessary to treat it as a radioactive
material. Nuclear Safety Commission, “About the clearance level in major reactor facilities”, March
1999.

5-2

5.2.2 Principles for ensuring the safety concerning predisposal management of radioactive
waste
In IAEA Safety Standards GSR-Part 5, “Predisposal Management of Radioactive Waste”46,
predisposal management of radioactive waste is specified to include all the radioactive waste
management stages performed before disposal, including processing, storage and transport. The
terminology related to radioactive waste management released by the IAEA is shown in Figure 5.21. In the predisposal management stage, radioactive waste processing is divided into pretreatment,
treatment, and conditioning.

Figure 5.2-1 Terms Regarding Radioactive Waste Management (IAEA)505152

Concerning specific measures to meet GSR-Part 5 Safety Requirements, IAEA Safety Guide SSG40 “Predisposal Management of Radioactive Waste from Nuclear Power Plants and Research
Reactors”53 provides measures for predisposal management of radioactive waste generated from
nuclear power plants and so on. Such international safety principles in radioactive waste
predisposal management can be summarized as follows:

Predisposal management of radioactive waste covers all the steps in the managementof radioactive waste from its generation up to disposal, including processing(pretreatment, treatment and conditioning), storage, and transport, and the radioactive

50“IAEA Safety Glossary, Terminology Used in Nuclear Safety and Radiation Protection,” 2007 Edition,
p. 216.

51
52( missing number )( missing number )53

International Atomic Energy Agency, Predisposal Management of Radioactive Waste from Nuclear
Power Plants and Research Reactors, Specific Safety Guide, SSG-40, IAEA, Vienna (2016).

5-3











waste shall be characterized and classified.
The main purpose of processing radioactive waste is to enhance safety by producing awaste form, that fulfills the acceptance criteria for safe processing, transport, storage
and disposal of the waste management, and to ensure safety of radioactive waste
disposal.
The processing of radioactive waste shall be based on appropriate consideration of thecharacteristics of the waste and of the demands imposed by the different stages in its
management (pretreatment, treatment, conditioning, transport, storage, and disposal).
The anticipated needs for any future steps in radioactive waste management have tobe taken into account as far as possible in making decisions on the processing of thewaste. Various factors are to be considered not only that of human health and safety onradiation exposure, but also environmental effects, and social and economic factors on
non-radioactive contained materials.
Decisions have to be taken within the overall approach to the predisposal managementof radioactive waste on the extent to which the waste has to be processed, with accounttaken to the quantities, activity and physical and/or chemical nature of, the radioactive
waste to be treated, the technologies available, the storage capacity and the availabilityof a disposal facility.
Storage is an option that should be considered in the waste management strategy.
Proper storage should be provided at all stages in waste processing, to ensure isolationand for environmental protection. Storage has to take place between and within thebasic steps in the predisposal management of radioactive waste. Storage is used tofacilitate the subsequent step in radioactive waste management, to act as a buffer
between and within waste management step, to allow time for the decay ofradionuclides prior to clearance etc., or to hold waste pending decisions on its future
management.
Waste shall be stored in such a manner that it can be inspected, monitored, retrieved
and preserved in a condition suitable for its subsequent management. Due accountshall be taken of the expected period of storage, and to the extent possible, passive
safety features shall be applied ensured by passive methods where possible. If it isstored for a long period of time, measures should be taken to prevent the degradationof waste containment performance.

Consequently, it is necessary to characterize and segregate radioactive waste at all the stages in
predisposal management of radioactive waste. In particular, storage is important as a means togive flexibility to management methods, which should be done in a way that can be inspected andmonitored. In addition, processing should be decided taking the amount and characteristics oftargeted radioactive waste, as well as the disposal requirements, into consideration. When making
decisions on processing, every requirement to be anticipated in each management stage in the
future should be considered as much as possible.

5-4

5.2.3 Principles for ensuring the safety of radioactive waste disposal
The ICRP describes the concept of radiological protection in relation to radioactive waste disposal
in Publ. 46 (1986) 54 , Publ. 77 (1998) 55 and Publ. 81 (1998) 56 in a systematic manner. The
introduction of ICRP Publ. 81 states that radioactive waste disposal strategies can be divided into
two conceptual approaches: “dilute and disperse”57 or “concentrate and retain” The Publication
states that the two approaches are not exclusive to each other and that either of which is more
appropriate than the other. The body of Publ. 81 deals with the radiological protection of the public,
after the disposal of long-lived solid radioactive waste using the “concentrate and retain” strategy.
On the other hand, the IAEA discusses Specific Safety Requirements No. SSR-548, the aims ofdisposal of solid radioactive waste can be summarized as follows:






To contain the waste;
To isolate the waste from the accessible biosphere and to reduce substantially the
likelihood of, and all possible consequences of, inadvertent human intrusion into thewaste;
To inhibit, reduce and delay the migration of radionuclides at any time from the waste
to the accessible biosphere; andTo ensure that the amounts of radionuclides reaching the accessible biosphere due toany migration from the disposal facility are such that possible radiologicalconsequences are acceptably low at all times.

It is possible to implement measures not to give possible radiological consequences, based on one
or more of these principles. Specific Safety Requirements SSR-5 includes more concrete
explanation regarding its requirements:

Waste shall be contained and isolated from the accessible biosphere to the extent that
this is necessary. The biosphere is that part of the environment that is normally inhabitedby living organisms, while the accessible biosphere is taken generally to include thoseelements of environment, including groundwater, surface water and marine resources,that are used by people or accessible to people.
The designs of disposal facilities for radioactive waste may differ widely, depending on
the types of waste to be disposed of and the host geological formation and/or surface
environment.

54
ICRP, 1985. “Principles for the Disposal of Solid Radioactive Waste.” ICRP Publication 46, Japan
Radioisotope Association.
55
ICRP, 1997. “Radiological Protection Policy for the Disposal of Radioactive Waste.” ICRP Publication
77. Ann. ICRP 27 (S).
56
ICRP, 1998. “Radiation Protection Recommendations as Applied to the Disposal of Long-lived Solid
Radioactive Waste.” ICRP Publication 81. Ann. ICRP 28 (4). (Publ. 81 was published to supplement andrevise the recommendations provided in Publ. 46 in light of international progress in radioactive waste
disposal, and to communicate the concept more clearly. Although the ICRP released Publ. 103 (The2007 Recommendations, International Commission on Radiological Protection) and then Publ. 122
Radiological protection in geological disposal of long-lived solid radioactive waste (2013) in which Publ.
103 is applied for geological disposal, these publications state that Publ. 81 is still valid.)
57
Concerning actions that correspond to ICRP’s “dilute and disperse” actions where radioactive
gaseous and liquid materials are involved, IAEA defines that "dilution and disperse" means releasing
effluent into the environment with environmental conditions and processes that will definitely reduce theconcentration of radionuclide to such a level that the radiation effect of the released material becomes
acceptable in the environment. (IAEA Safety Requirement GSR - Part 5 (4.1))

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In Specific Safety Guide SSG-14, Geological Disposal Facilities for Radioactive Waste 58 andSpecific Safety Guide SSG-29, Near Surface Disposal Facilities for Radioactive Waste,59 include
the following guidance to meet requirements on Specific Safety Requirements SSR-5:

In developing the designing of the disposal facility and the safety case,60 the operator
has to take account of the characteristics and quantities of the radioactive waste to bedisposed of, the prevailing geological environment, the engineering and mining
techniques available, and the national legal infrastructure and regulatory requirements.
In developing the design of a safe, near surface disposal facility, the operator should
establish a safety strategy that will clearly set out how the facility is to comply with all
the safety requirements. The strategy should indicate how the safety principles will beapplied and should take into consideration the characteristics and quantities of theradioactive waste to be disposed of, the characteristics of the available site or sites, theavailable engineering techniques, and the national legal infrastructure and regulatory
requirements.

Based on these principles, containment and isolation should be the basic principle for disposal, and
it is important to examine disposal concepts based on the characteristics of radioactive waste. Notethat disposal facilities are needed to be designed in consideration of radioactive waste
characteristics and volumes.

5.3 Current Status of Solid Waste Management
The current status of actions concerning storage, waste characterization and processing/disposal
based on the Mid- and Long-Term Roadmap is shown below.
5.3.1 Storage(1)Current status on storage of solid waste
Solid radioactive waste that has been generated up to now consist of “rubble, etc.” and secondary
waste from water treatment. “Rubble, etc.” attributed to scattering and diffusion of nuclides which
originated from core fuel. Secondary waste results from the absorption treatment of contaminatedwater in which the nuclides, mainly originated from fuel debris, are dissolved (Figure 5.3-1).
“Rubble, etc.” are segregated into rubble, felled tree, used protective clothing, etc., and are stored
separately. The rubble in the volume of 350,000 m3 is stored in temporary storage areas,according to its surface dose rate. On the other hand, among the secondary waste generated
from water treatment, 3,600 adsorption columns are temporarily stored in the used adsorption
column storage facilities, provided with additional shielding and water tightening, depending on
the column types. Table 5.3-1 shows the data of solid waste storage conditions.
Of “rubble, etc.”, used protective clothing, etc. has been incinerated since March 2016 at a newly
built miscellaneous solid waste incinerator to reduce volume. The incinerator ash after volume
reduction is packed in drums, and stored in the solid waste storage building.

58
International Atomic Energy Agency, “Geological Disposal Facilities for Radioactive Waste,” Specific
Safety Guide, SSG-14, IAEA, Vienna (2011).
59
International Atomic Energy Agency, “Near Surface Disposal Facilities for Radioactive Waste,”Specific Safety Guide, SSG-29, IAEA, Vienna (2014).
60
The term safety case refers to a collection of theoretical grounds and evidentiary bases in science,technology, management, etc. A safety case is built along with the development of facilities, and used
as a ground for making decisions related to development, operation, and shutdown.

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Of secondary waste from water treatment, used vessels from the cesium adsorption apparatus
are stored in a box culvert, used vessels from the second cesium adsorption apparatus and thehigh-performance multiple radionuclide removal system are stored in a rack (steel mount) fixed
to the deck slab. The slurry from the pretreatment system of the multiple radionuclide removal
system is stored in high-integrity containers (HICs), and the box culverts in which the HICs will
be placed are built to be watertight. The volume of slurry stored in each HIC is limited, to preventfrom leaking supernatant water due to slurry expansion. Regarding waste sludge, the volume ofabout 600 m3 is temporarily stored in waste sludge storage facilities (in the basement of mainprocessing building; as of the end of March 2017). The sludge storage facilities are provided withagitation function and heat-removing function to prevent the decay heat concentration, and withventilation equipment to prevent hydrogen stagnation. Protection measures against tsunami arealso considered.
These temporary storage areas are currently dotted widely across the Fukushima Daiichi NPS
site. (Figure 5.3-2(a))

Figure 5.3-1 Overview of Solid Waste Contamination Sources and Nuclide Migration Pathways61

61

IRID Supplementary budget for 2014, “Management Waste Contaminated and Decommissioning ofProject (R&D for processing and disposal of solid waste), Final Report, August, 2017.

5-7

Table 5.3-1 Solid Waste Storage Status62(a)Management status of rubble, felled tree, used protective clothings, etc. (as of Apr. 30, 2017)

RubbleSurface dose rate
（mSv/h）

Storage method≦0.1

Outdoor storage

≦1

1～30

＞30

Storage volume(m3)/ Storage capacity(m3)(Percentage)
147,900 / 214,300（69%）
30,900 / 71,000（44%）

Outdoor sheet covered
storage
Soil covered temporarystorage facility, Temporarystorage tent, Outdoor
container storage
Container(in Solid waste
storage building)Total

20,800 / 27,700（75%）
8,300 / 12,000（69%）
207,900 / 325,000
(64%)

----

Felled tree
Category

Storage method

Root

Outdoor storage

Branch/leaves

Temporary storage pool

Total

----

Storage volume(m3)/ Storage capacity(m3)(Percentage)
79,500 / 144,500(55%)
19,600 / 24,900(79%)
99,100 / 169,400(59%)

Used protective clothing, etc.
Storage volume(m3)/ Storage capacity(m3)Storage method(Percentage)
67,500 / 71,200(95%)

Container(b) Management status of secondary waste generated from the water treatment (as of May 18, 2017))

Used Vessels
Storage
Place

Outdoortemporarystoragearea ofu s e dvessels

Type of Used Vessels
Cesium Absorption apparatus
2nd Cesium absorption apparatus
HICs from multiple radio-nuclides removal
systemHICs from improved multiple radio-nuclides
removal system
Used vessels from high-performance
multiple radio-nuclides removal system
Used column from multiple radio-nuclides
removal system
Used vessels and filters from mobile-type
strontium system

Storage
NumberStorage
Number/Capacity(Percentage)

758
188
1-365
1,044

3,628 / 6,239
(58%)

73
9
191

SludgeSludge storage facility (Indoor)

597

m3

597 / 700
(85%)

9,379

m3

9,379 / 10,700(88%)

Concentrated waste liquid
Concentrated waste liquid storage tanks (Outdoor)

62

Decommissioning and contaminated water management team meeting (42th), Attachment 3-4,
Radioactive waste processing and disposal, May 25, 2017.

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(2)Solid Waste Storage Management Plan and its revision
In March 2016, TEPCO released the Solid Waste Storage Management Plan (hereinafter the“Storage Management Plan”),63 which forecasts the amount of solid waste that will be generated
in the next ten years or so, and indicated the policy for the construction of facilities to store thatwaste. The Storage Management Plan was revised 64 in June 2017, reflecting the reviewed
forecast of waste generation, based on the latest storage records of “rubble, etc.”, and the latest
construction plan.
In the Storage Management Plan, the amount of solid waste generated in the next ten years isessential, and a plan is shown to reduce volume of waste stored outdoor as much as possible
and transfer to indoor storage in order to achieve further reduction of the risk on storing solid
waste.
For “rubble, etc.”, TEPCO plans to reduce as much volume as possible and transit to the statusthat it is stored in solid waste storage facilities in FY 2028, excluding contaminated soil andrecyclable waste (those with a surface dose rate of less than 0.005 mSv/h). TEPCO also plans
to construct large waste storage facilities that is possible to store large, heavy secondary wastefrom water treatment, such as absorption columns to realize indoor storage.
When these plans are realized, many temporary storage areas which are dotted will bedisappeared, and the waste storage areas will be integrated as shown in Figure 5.3-2(b).
However, the waste whose future generation are not yet estimated and are not be included in
the Storage Management Plan will be reflected to the forecast when the plan become concrete
and action such as removal of the waste may be done within ten years after the Storage
Management Plan is implemented.
The waste generation forecasts may be fluctuated with the progress of decommissioning workin the future, and as TEPCO indicates, the forecasts need to be reviewed at least once a year
and the Storage Management Plan needs to be updated as appropriate .

63Tokyo Electric Power Company, “The Solid Waste Storage Management Plan for TEPCO Fukushima
Daiichi Nuclear Power Station,” March 31, 2016.
64
Tokyo Electric Power Company Holdings Inc., “The Solid Waste Storage Management Plan for
TEPCO Fukushima Daiichi Nuclear Power Station,” June 2017 Version, June 29, 2017.

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Figure 5.3-2 Current and Future (in FY 2028) Status of “Rubble, etc.” and Secondary Waste of
Water Treatment in the Fukushima Daiichi NPS Site64
5.3.2 Waste characterization
The solid waste attributable to the Fukushima Daiichi NPS accident are deemed to differ incharacteristics from the waste generated by normal operation of nuclear power stations, inconsideration of the contamination sources and the migration pathways of nuclides shown in Figure5.3-1. As discussed in Subsection 5.2.2, in every control stage of the predisposal management ofsolid waste, it is necessary to understand characteristics of solid waste such as nuclide composition
and radioactivity concentration.

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For this purpose, samplings were carried out mainly on rubble generated after the accident andsecondary waste from water treatment, and samples are analyzed at existing analysis facilities.
Analysis results of about 300 samples have been accumulated over the past 6 years (Table 5.3-2).
Figure 5.3-3 shows examples of the sampling location.
Based on the characteristics of the analysis data obtained so far, the characteristics of part ofcontamination is becoming presumable.65 Because the number of the samples that can be used
for estimation is still limited, further sampling and analysis are required to improve the accuracy of
the assumption. Simplified analyses for measuring environmental samples and management ofregular radiation work has been conducted in Fukushima Daiichi NPS, in addition to analyses
aimed at characterization of the solid waste mentioned above.
Analysis data on samples of rubble in the reactor buildings
 Fission products such as H-3, Sr-90, I-129 and Cs-137, alpha nuclides such as Pu238, and activation products such as C-14 and Co-60 were detected for samples fromthe first and fifth floors of Units 1 and 2 and on the first floor of Unit 3,.
 Analysis data on samples obtained from PCVs and spent-fuel pools The radioactivity concentration of Co-60 and alpha nuclides to that of Cs1-37 tends
to be high in ratio for upstream stagnant water (in the PCVs of Units 2 and 3) is,
compared with downstream stagnant water (in the main processing building and hightemperature incinerator building of the centralized waste processing facilities).
 In the Unit 4 fuel pool, the ratio of Co-60, some activation products, to Cs-137, isgreater in radioactivity concentration ratio, compared with other samples in thereactor buildings.
 Analysis data on samples of turbine building basement sludge
 The radioactivity concentration of Sr-90 to Cs-137 tends to be higher for sludge in thebasement of the turbine building, compared with the rubble in reactor buildings.
 The sludge sediment from the underground stagnant water of the turbine building
tends to include Cs-137, Sr-90 and alpha nuclides. On uranium U, the influence of
natural uranium is high based on the ratio of the nuclides.
 Analysis data of samples of rubble near reactor buildings
 The radioactivity concentration of H-3, C-14, Co-60 and Sr-90 in the rubble near Units
1 and 3 tends to be proportional to the radioactivity concentration of Cs-137.
At present, a method of understanding the characteristics of solid waste through complementary
use of data evaluated based on nuclide migration model66 and solid waste analysis data, is also
being developed in order to promote characterization.
The migration model–based evaluation data initially had many uncertainties because it used
documented values estimated under conditions that may not come from the same contamination
65

IRID/JAEA, Fukushima Daiichi Nuclear Power Station Solid Waste Sample Analysis (Achievement toDate), May 25, 2017, p. 4–5.
A model simulated the contamination processes inside reactor building and stagnant water, causedby nuclides released from damaged fuel after the accident, in order to calculate radioactivity of eachnuclide in solid waste. This model is built considering the burnup calculation for estimating radioactivityof each nuclide in a reactor, the ratio of nuclides released into a reactor as airborne, the ratio of nuclides
migrated to stagnant water, and the ratio of nuclides released to outside of reactor building. Calibration
using accumulated analytical data reduces the uncertainties of nuclide migration parameters such as
release rates, to improve accuracy of estimation on nuclide composition and radioactivity in waste
continuously. The evaluation based on the nuclide migration model means that evaluation done byestimating nuclide composition and radioactivity of waste calculated using nuclide migration model.
66

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migration process as that of Fukushima Daiichi NPS accident. The accuracy has been improving
by conducting calibration using analysis data.

Figure 5.3-3 Example of Sampling Locations65

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Table 5.3-2 Samples on Analysis for Characterization of Solid Waste67 (as of March 31, 2017)
Category

Rubble, etc.

Contaminated water

Samples

In reactor building

Numberofsamples

Unit 1, floors 1 and 5

Rubble, etc.

29

Unit 2, floors 1 and 5

Rubble, etc.

7

Unit 3, floor 1

Rubble, etc.

11

Unit 4, spent fuel pool

Rubble2

In turbine building

Unit 1

Sludge and sand7

Around reactor
building

Around Units 1-3, 4

RubbleSoil covered
temporary storage
facility

Pools 1 and 2

RubbleIn reactor building

Units 2 and 3, in PCV

Stagnant water4

In turbine building

Unit 1

Sludge andstagnant water6

In centralized
waste processing
facilities

Basement

Stagnant water

Treatment system

Cesium Absorption apparatus(incl. the second)

Treated water

Decontamination system

Treated water

3

Desalination system (RO)Treated water2

Evaporative concentration
systemTreated water(Improved) multiple radionuclides removal system

Treated water

15
10

Slurry

12
27

3
18

Secondary wastefrom water
treatment(Improved) multiple radio-nuclides removal system

Combustibles

Incinerator ash of used protective clothing

5

Soil

Soil

6

PlantsFelled tree

Branch/leaves

Tree

Branch/leaves, fallen leaves, surface soil

6

5
123
Total

301

5.3.3 Processing and disposal
Of the solid waste, some of secondary waste from water treatment has high fluidity, includes
materials and components that have not experienced processing or disposal in Japan, and isrelatively in high dose rate. Therefore, the technologies for stabilization and immobilization havebeen developed on a priority basis.
Technologies used for processing radioactive waste on an actual scale have been selected toperform basic cold test of immobilization using simulated waste. The basic tests conducted on theuniaxial compressive strength and waste content (ratio in weight of waste to solidified body) wereused as the criterion for determining whether immobilization is possible or not. The nuclide-leachingrate and the amount of hydrogen gas attributable to radiation decomposition were used as criteria
67

IRID/JAEA, “Fukushima Daiichi Nuclear Power Station Solid Waste Sample Analysis (Report of theCurrent Status),” May 25, 2017, p. 3.

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to evaluate the excellence of the technologies, thus examining the applicability of individual
immobilization process technologies61.
To judge whether these technologies are applicable, data samples under the environmental
conditions of actual waste, such as dose effect assessment and heat generation effect evaluation
are not sufficient. Therefore, more tests and evaluations are required to obtain them in the future.
For HIC slurry, stabilization treatment (drying and filtration), extraction, transfer, and HIC cleaning
were tested, among stabilization processes61. It is necessary to study technical feasibility as astabilization plant and methods of storage of slurry dehydrate after stabilization treatment.
For waste sludge, domestic and international technologies seemed to be applicable have been
examined, and candidate technologies are being screened.68
To study how to dispose of solid waste, it is necessary to set the nuclide composition andradioactivity concentration of solid waste based on the data obtained through waste
characterization. To date, we have been trying to examine classification and evaluation of solid
waste from the viewpoint of disposal, based on provisional nuclide composition and radioactivity
concentration61.
In the above study, we tried to check whether it is possible to classify solid waste from the viewpointof ensuring safety after the control period, referring to IAEA's Safety Guide GSG-1 "Classificationof Radioactive Waste"69. This has been applied to the target disposal concepts, considering theextent of classification and the uncertainty of nuclide composition and radioactivity concentrationof solid waste. It was shown that there is a possibility of obtaining effective information for efficientcharacterization, such as solid waste of which data should be acquired on a priority basis as wellas important nuclides. In the present situation, disposal concept based on the solid waste
characteristics is not ready for examination, and the above test was conducted concerning theexisting disposal concept.

68

Study for Processing Secondary Waste of Water Treatment, 5th Radioactive Waste Regulation
Review Meeting of Specific Nuclear Facilities, Document 2, TEPCO Holdings Co., Ltd., February 10,
2017.
69International Atomic Energy Agency, Classification of Radioactive Waste, General Safety Guide,
No.GSG-1, IAEA, Vienna. 20095-14

5.4 Proposal for compilation of the basic concept of solid waste management(strategic proposal)
Subsection 5.4.1 discusses the characteristics of solid waste presumed based on the accident
conditions at Fukushima Daiichi NPS, actions taken for decommissioning after the accident, andthe results of waste characterization to date. Subsection 5.4.2 proposes solid waste management
policies based on the presumed characteristics. The purpose of solid waste management is toprocess and dispose of it in safe manner, in addition to reduce the volume. It should be focused on
predisposal management such as characterization, storage and preceding processing until
establishing concrete prospect for disposal. Subsection 5.4.3 discusses specific measures in linewith this policy.

5.4.1 Characteristics of solid waste
The characteristics of solid waste presumed based on accident conditions, actions taken fordecommissioning after the accident, and the results of characterization so far, are discussed below.










The amount of solid waste is greater than that of solid radioactive waste generated bynormal operation of nuclear power stations, and a lot of solid waste may have relatively
high dose rate. As an example, when fuel debris retrieval starts, the materials andequipment removed from around fuel debris will appear as high-dose solid waste.
As the major source of contamination is fuel debris, radioactive concentration of alphanuclides and beta/gamma nuclides in solid waste does not exceed that of spent fuel.
Therefore, it is possible to proceed with study, and research and development of solid
waste processing and disposal, using the experience and knowledge concerning thoseof radioactive waste that have accumulated in Japan and other countries.
The results of waste characterization obtained so far suggest that solid waste nuclide
composition and radioactive nuclide concentration of solid waste have more varieties,
compared with those of solid radioactive waste generated by normal operation of
nuclear power stations. In order to verify this, further solid waste characterization should
be conducted.
Some of secondary waste from water treatment has high fluidity and contains
radioactive materials with high dose rate that may cause generation of hydrogen.
Furthermore, most of them have never been dealt with in Japan.
The waste generated immediately after the accident are presumed to include theingredients of tsunami seawater that entered the reactor buildings, boron included in
the boric-acid water poured to prevent re-criticality, and the anti-scattering agent usedduring recovery work after the accident.
It is estimated that some of the materials might contain substances that possibly impact
on the safety; e.g., substances that may lower vessel containment performance during
the predisposal management period, or disposal facility containment performance
during the control period of disposal. It is also estimated that some materials may impact
on the environment by its chemical hazard.
The overall profile of the solid waste that is essential for the study of disposal (generated
volumes and characteristics) will be revealed sequentially with the progress of work,
such as fuel debris retrieval, contaminated water management and otherdecommissioning-related work as well as the establishment of the plan. Additional work
will be planned in the future, through which new characteristics will be gradually
identified.

5.4.2 Solid waste management policies
Considering the international safety principles for radioactive waste management and radiologicalprotection, as well as the characteristics of solid waste which have been presumed based on the

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results of the activities for decommissioning and of the waste characterization after the accident,
the following solid waste management policies are proposed:(1)Thorough containment and isolation
Based on the safety ensuring principles specified in Subsection 5.2.1, solid waste management
should ensure their thorough containment of radioactive materials to prevent their dispersion(leaking) and their complete isolation to prevent human access, thus preventing humans fromhaving significant radiation exposure from solid waste.
(2)Solid waste volume reduction
Reducing the amount of solid waste generated by decommissioning as much as possible in orderto ease the burden of solid waste management.
(3)Promotion of waste characterization
To proceed with study on processing and disposal methods, such as clarifying the nuclide
composition, radioactive concentration of solid waste is essential through waste characterization.
In addition to the fact that solid waste is large in volume, they have varied nuclide composition
and radioactive concentration, because of the influence of the accident. With the progress of
decommissioning process, the number of samples to be analyzed will increase. Therefore, only
accumulation of data with methods such as sampling, radiochemical analysis and measurement
will require a length of time for solid waste characterization.
In order to deal with future increase in analysis samples, it should be performed from a long-term
perspective to prepare analysis facilities and equipment and to ensure and train sufficient
analysis personnel for improved analysis capacity.
Furthermore, it is beneficial to build a method for understanding the characteristics of solid waste
by complimentarily combining analysis data and evaluation data based on nuclide migration
model. In addition, more efficient waste characterization should be performed, by research anddevelopment on the optimization of the numbers of analyzed samples and the simplification andspeed-up of analysis methods.
(4)Thorough storage
Storage Management Plan should be implemented steadily in order to improve the safety of solid
waste storage.
To dispose of solid waste, it is essential to understand the overall profile (volumes andcharacteristics) of the solid waste to be disposed of, and to establish specifications of disposal
facilities and technical requirements for waste packages. However, the overall profile of solid
waste will become clear step by step, with the future progress and plan established on fuel debris
retrieval work, contaminated water management and other decommissioning-related work.
Therefore, safer and more reasonable storage should be performed for solid waste generated in
the future, taking the characteristics of solid waste into account based on the principle inSubsection 5.4.2(1) stated above. The storage capacity should be ensured to make sure of thestorage of solid waste within the site of the Fukushima Daiichi NPS.

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Preparing for the full-scale retrieval of fuel debris in the future, we will also examine how to store
and manage the solid waste generated during fuel debris retrieval, together with how to retrievefuel debris.
(5)Establishment of a selection system of preceding processing methods in consideration of
disposal
It is desirable to perform waste processing based on the technical requirements for disposal after
they have been established. However, there will be the cases that processing for stabilizationand immobilization (preceding processing) may be required before the technical requirements ofdisposal are established, in order to store and manage solid waste in safer and more reasonable
manner.
In such a case, a possibility of rework should be taken into account, considering that the
specifications of solid waste on which the preceding processing conducted may not fit, whentechnical requirements of disposal are established.
To minimize the possibility that the processing may not meet the technical requirements ofdisposal, for each specification of solid waste on which preceding processing was applied, thesafety against some disposal methods is evaluated, and based on the results, the system forselecting the processing methods should be established.
Through such a procedure, preceding processing concerned with the stabilization andimmobilized of solid waste in consideration of future disposal may be enabled during the storage
period while the possibility of failing to meet technical requirements for disposal is minimized.
The United Kingdom and France have already introduced similar approaches, which can be used
as references.
Among the solid waste to be evaluated by the system, development of necessary technology
should be promoted for the solid waste that highly requires stabilization and immobilization but
for which even candidate methods are not yet identified, in parallel of implementing the system.

5.4.3 Present efforts and R&D based on solid waste management policies
Specific measures to be taken in line with the above-mentioned policies of solid waste managementin Subsection 5.4.2 are shown below.
(1)Thorough containment and isolation
Based on the policy described in Subsection 5.4.2(1), safety should be ensured by the following
methods in the solid waste management:
Solid waste is contained by measures such as storing in containers or immobilization
as needed for prevention of flying apart or leakage.
 Solid waste is isolated by storing it in the storage place appropriately, and suitable
management such as monitoring is performed.
(2)Reduction in the volume of solid waste (reduction in amounts and volume reduction processing)At the Fukushima Daiichi NPS, to reduce the amounts of solid waste based on the policy
described in Subsection 5.4.2(2), it is necessary to continue such effort as carry-in control,
reuse/recycling and volume reduction.
The reduction of packaging materials and other materials to be carried into the site has beenachieving a certain level of results and should be continued. In addition, in future

5-17

decommissioning work, dismantling and decontamination methods that will lower amounts ofsolid waste will be used as far as reasonable, to reduce the amounts of waste to be generated.
Solid waste generated should be segregated to reduce handling burdens in the subsequent
management stages. For the waste whose contamination level is very low, they should be reused
and recycled, to reduce the amounts of solid waste to be managed. For the solid waste thatcannot be reused or recycled, volume reduction processing should be performed as necessary
for those whose volumes can be reduced to ensure the capacity of storage.
(3)Promotion of waste characterization
Based on the policy specified in Subsection 5.4.2(3), analytical capability and efficiency of waste
characterization should be improved. To improve analytical capability, development of new
facilities/equipment and utilization of existing ones should be promoted systematically from a
mid-to-long term perspective. At present, the number of analysis personnel with necessary
abilities is not sufficient. Since it takes time to secure and foster new staff, it is necessary toestablish a system for human resource development (HRD) and technology transfer at an earlystage.
Regarding R&D for improving characterization from the perspective of efficiency, develop acharacterization method with complimentarily combining analytical data and evaluation data
based on nuclide migration models, and facilitate R&D for optimizing the number of analyticalsamples and for simplifying analytical methods to speed up analysis.
As for the development of characterization method with complimentarily combining analytical
data and evaluation data based on nuclide migration models, analytical data should beaccumulated and managed based on the analysis plan, and incorporated the variation ofcontamination estimated by analysis data and contamination mechanism into the nuclide
migration model to improve the accuracy of the evaluation data. By supplementing with theevaluation data, it is expected to reduce the number of necessary analysis data, compared withcharacterization based only on analysis data.
Regarding the optimization of the number of analytical samples, the evaluation methods from theviewpoint of the representativeness of analysis data should be developed. In addition, byproceeding with the evaluation of analysis data related to waste characterization and with theunderstanding of contamination mechanism, the nuclides to be analyzed will be reviewed. R&D
should be done to develop such as more simplified and speedier analysis methods based on therevised set of nuclides.
In the near future, after reviewing the contents of waste characterization (to grasp contamination
distribution, nuclide composition, etc.) based on the progress of solid waste management,
analyzed nuclide and measurement precision etc. will be reviewed as appropriate, according to
the contents. The results should be reflected in the analysis plan.
Concerning the solid waste that will be generated with the near-future decommissioning work,
records including the place of generation and position data will be collected and managed, toenable utilization in solid waste characterization, storage, management, stabilization andimmobilization.
Waste characterization involves sampling high dose level samples as well as handling of highdose level solid waste. To reduce personnel’s radiation exposure, remote-control should be
introduced into sampling.

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The waste characterization is mainly focused on radioactive substances. In the future, study
should also conducted on non-radioactive substances that may affect such as the safety ofprocessing and disposal.
(4)Thorough storage
Based on the Storage Management Plan, further safety improvement concerning solid waste
storage should be realized, by reducing the volume of solid waste temporarily stored outside
solid waste storage facilities as far as possible, and by transferring them in solid waste storage
facilities. The annual review of the Storage Management Plan should incorporate changes in theconditions, such as the progress of decommissioning activities.
Based on the previously mentioned policy in Subsection 5.4.2(4), it will be required to discuss
the methods of evaluating the amount of hydrogen gas generated from secondary waste fromwater treatment during the period of storage, to estimate timing of implementing additional safetymeasures and to consider what kind of measures will be required, considering the characteristics
of solid waste to be generated sequentially in the future. For containments in storage, further
safety improvement is examined and evaluated concerning storage, such as for anticorrosion.
When fuel debris retrieval starts, solid waste, including waste and equipment removed from near
fuel debris, is expected to be generated. As Table 4.5-4 suggests, this solid waste will include
high dose and high radioactivity waste.
Study storage methods for solid waste to be generated by fuel debris retrieval parallel to thestudy on fuel debris retrieval/storage methods.
(5)Establishment of a selection system of preceding processing methods in consideration of
disposal
Based on the previously mentioned policy described in Subsection 5.4.2(5), the system forselecting the processing method should be established, in view of disposal.
Firstly, some disposal methods that are reasonable, feasible, and suitable to the characteristics
of solid waste should be established, based on the evaluation results of solid waste
characterization, with referring to the disposal study on examples in other countries, and not
restricted in specific location or scale of facilities. In the next, safety evaluation methods(scenarios, models, data and other factors related to safety evaluation) suitable for theestablished disposal methods should be developed. In addition, study should be conducted to
examine the sufficiency of the data required for the evaluation related to preceding processing
and to evaluate substances that may influence safety such as safety in disposal, and required
information should be obtained, as needed.
At the same time, the solid waste processing methods evaluated by this procedure should bescreened, and taking into account uncertainty on the evaluation results of waste characterization,
specifications of solid waste on which preceding processing has applied should be set. The solid
waste to be evaluated by this procedure should be that whose characteristics are known to someextent, such as secondary waste from water treatment with high fluidity (e.g., sludge generated
from such apparatus as the Advanced Liquid Processing System and waste sludge from the
Simplified Active Water Retrieve and Recovery System).
Safety evaluation for each set of solid waste specifications should be conducted after preceding
processing method is applied based on the processing method screened for the establisheddisposal methods. Based on the results of this evaluation, an appropriate processing methodshould be selected. Depending on the safety evaluation results, established disposal methods

5-19and processing methods should be reviewed where necessary. Based on this selection results,
preceding processing for stabilization and immobilization should be promoted.
Concerning the solid waste whose candidate stabilization and immobilization processing
methods are not established among the solid waste to be evaluated by this procedure, technicaldevelopment for stabilization and immobilization should be promoted. The technical developmentof secondary waste from water treatment should be on a priority basis. In technical development,
we should extensively study domestic and overseas technologies, and evaluate to what extent it
is efficient as a real scale processing facility, how realizable it is in the situation of Fukushima
Daiichi NPS, whether it is applicable to disposal, and how much it has been demonstrated to befeasible to date.
(6)Efficient implementation of R&D projects from the perspective of overall solid waste
management
To efficiently proceed with R&D concerning the solid waste management for safe disposal, close
cooperation should be realized between research and development fields such as waste
characterization, processing and disposal. Share the issues and discussions progress among
the all players and proceed with R&D, reviewing required R&D themes with a bird’s-eye-view ofoverall solid waste management.
(7)Development of a system for continuous operations
Develop a continuous operational system including development of adequate facilities andhuman resources, which are concerned with solid waste management, in order to continue safe
and steady solid waste management.
(8)Measures for reducing exposure of workers to radiation
To proceed with solid waste management steadily, it is important to ensure the safety and healthof workers engaged in the work. Therefore, radiation exposure control, safety management andhealthcare program should be implemented thoroughly based on the relevant laws/regulations.
Figure 5.4-1 shows the near-future measures as well as research and development plan basedon this Section.
The above measures and R&D plans should be closely associated with the decommissioning
process including fuel retrieval.

5-20Compilation of basic
concept ofprocessing/disposal

Phases/FY
2015

2016

2017

Clarification of prospects onprocessing/disposal method
and safety technology

2018

2019

2020

2021

2022 and later

On-site work
Technical studies for on-site construction
Research and development

1. Reduction of solid waste

Carry-in control, reuse, recycling and volume reduction

2. Waste characterization
Sampling(1) Analysis data acquisition,
management, etc.

Analysis and measurement for verification of waste for disposal
Development/establishment of methods on verification of waste for disposal
Planning/updating analysis plan, conducting analysis, characterization, accumulation/evaluation/management analysis data

Development of sampling technique etc. (Method of sampling high-dose samples)

Establishment of Database system

Development of radioactive material analysis and research facilities(2) Improvement of analysis

Training for analysis personnel

capacity
Development of characterization method with complimentarily use of data evaluated from nuclide migration modeland solid waste analytical data

(3) Improvement in efficiency ofwaste characterization

Optimization of the number of analytical samples, review of nuclides to be analyzed, simplified and speedieranalysis methods, etc.

3. Storage(1) Storage Management Plan
(2) Study and evaluation ofstorage methods(3) Study and evaluation ofstorage methods of solid
waste generated during
fuel debris retrieval
4 Development of selection
system of preceding
processing

Storage
Development and review of the Storage Management Plan

Study on storage of secondary waste from water treatment

Study on safety
improvement
measures

Study on storage method, storage containers, and the way of putting waste into containers

Study of processing/disposal

Establishment of a selection system of preceding processing
Survey and setting of disposal method based on characteristics of solid waste
Survey and development of safety evaluation method for each disposal method
Survey on conditioning
Technical development of treatment methods related to stabilization and immobilization
technology

5. Efficient Promotion ofresearch and development
overlooking the entire solid
waste management
6. Development of continual
operational system
7. Measures for reducing
radiation of personnel’
exposure

Status of study and challenges are shared to confirm the necessary R&D tasks through close collaboration among each fields

Continuous operational system regarding solid waste management including development of related facilities and human resources

Control and management of exposure, safety, health checkup, etc., based on the related laws and regulations

Figure 5.4-1 Near-future Measures and R&D Plan for Waste Management

5-21

Use ofselectionsystem6. R&D initiatives
6.1 Basic policy for R&D
6.1.1 Basic policy
The decommissioning of the Fukushima Daiichi NPS involves many highly technical challenges. It
is necessary to develop new technologies to deal with those challenges and highly reliable
technologies for practical use in order to implement various sound measures according to the Midand-Long-Term Roadmap. Therefore, through the governmental subsidies for research projects
and facilities, multiple R&D projects aimed at practical application of technologies, research
centers/facilities construction by JAEA, and basic/generic researches and advanced researches by
research institutes including universities, etc. are under way. TEPCO has been working by itself as
well.
NDF has determined the R&D duties implementation policy based on the NDF Act. According tothis policy, NDF gathered national/international expertise and managed a variety of R&D projects,
such as clarifying the inner reactor conditions, feasibility study of retrieval methods and so on, that
support steady implementation of measures based on the Mid-and-Long-Term Roadmap.
As fuel debris retrieval policies for each unit will be fixed, such R&D methodology shall enter a newphase. And as detailed processes towards the decommissioning will be clarified, the roles of eachR&D player should be more clarified. In this case, it is necessary to arrange the division of roles
and responsibilities between the government and the operator in an appropriate manner in orderto steadily implement R&D results on the decommissioning site. And it is considered to be further
expected that the government and the relevant research institutes should establish a center ofbasic research/research infrastructure based on a mid-to long-term perspective. Research
institutes are expected to enhance the technologies required for decommissioning throughconsiderations on the status of the project and the fundamental R&D activities according toscientific and technological issues (needs) regarding the decommissioning.
It is required to mobilize all available resources/expertise by further encouraging a variety of
initiatives/efforts from different angles such as effective R&D activities, strengthening cooperation
with the relevant institutes including foreign ones, promoting use of research facilities and HRD inorder to deal with this unexplored challenge.

6.1.2 Entire perspective of R&D
There are variety of institutions engaged with the Fukushima Daiichi NPS decommissioning R&D
projects through the areas of basic/generic research, applied research and development
research/utilization. (See Figure 6-1 and 6-2). NDF has been undertaking optimization of R&D
activities carried out by each institution for a better effectiveness and efficiency as a whole.

6-1

Figure 6.1 A conceptual picture of the division of roles among main R&D institutions for 1F
decommissioning
Regarding gathering national and even international expertise, METI’s Project of Decommissioning
and Contaminated Water Management includes foreign companies as its players,andinternational joint-research projects in the Nuclear Energy Agency of the Organization for Economic
Co-operation and Development (hereinafter referred to as “OECD/NEA”) such as BSAF, a research
project of severe accident progression analysis, are ongoing. Moreover, joint-research projects withoverseas institutes are underway as a part of MEXT’s World Intelligence Project for Nuclear S&T
and Human Resource Development.

6-2

6-3

Figure 6.2 Whole Picture of R&D structure of 1F Decommissioning (as of FY2017)

6.2 Promotion of R&D aimed at utilization for the decommissioning process
There are two types of R&D activities towards practical use for successfully implementing thedecommissioning of Fukushima Daiichi NPS: the engineering activities including R&D conductedby TEPCO via subcontracting, etc. and the Project of Decommissioning and Contaminated Water
Management carried out by selected subsidiary companies.
6.2.1 R&D activities performed by TEPCO
It is required for TEPCO, as a responsive operator of the Fukushima Daiichi NPS decommissioning,
to implement R&D activities, and also to promote introduction and demonstration of necessary
excellent technologies with economic feasibility, with consideration for the site conditions.
Some of the R&D activities TEPCO has conducted up to now include: investigation on equipmentsin the reactor buildings using robots developed on the basis of smartphones, and R&D contributing
to the operation and maintenance of contaminated water purification equipment, etc. TEPCO is
expected to continue actively implementing highly feasible R&D projects necessary fordecommissioning, and to promote engineering projects for applying component technologies
developed in the Project of Decommissioning and Contaminated Water Management as well.
Especially, TEPCO is expected to develop practical equipment as of the progress of the engineering
necessary for fuel debris retrieval.
6.2.2 Project of decommissioning and contaminated water management
METI’s Project of Decommissioning and Contaminated Water Management has achieved certain
results. It is important to flexibly reorganize this kind of R&D projects, depending on the accurate
grasp of the “needs” in the decommissioning operations, latest conditions in the reactor, thedecommissioning process or the actual progress status of each R&D project, to make the entire
project realistic and effective. Therefore, NDF has been watching the process of the Project ofDecommissioning and Contaminated Water Management as needed and supporting it technicallyat every occasion such as interim/final reporting session of each project and performance test at
developing site. NDF has been developing annual R&D plans that are mindful of the timeline of thedecommissioning activities with METI for every fiscal year and reporting it to the Secretariat of theTeam for Countermeasures for Decommissioning and Contaminated Water Treatment as well(seeTable 6-2, Figure 6-4 through 6-17).
Once the fuel debris retrieval policies for each unit are determined, engineering activities, such as
application designing for actual units, shall commence. As these processes progress, it is needed
to continuously revise R&D projects in a manner of proper division of the roles between thegovernment and the operator in order to make it work more closely with engineering processes.
More specifically, it is required to focus on important R&Ds towards practical use at actual sites inan accelerated manner based on the fuel debris retrieval policies to be decided for each unit, andto promote transfer of technologies that were completed through the Project of Decommissioning
and Contaminated Water Management into the engineering process smoothly. It is also necessary
to focus on and conduct thoroughly the R&D of technologies that can be a critical path of the futuredecommissioning activities as mentioned in Section 4.7 including the one rose up throughpreliminary engineering.

6-4

6.3 Reinforcing cooperation in R&D
Decommissioning of Fukushima Daiichi NPS holds a variety of issues accompanied with R&D
challenges. Figure 6-2 shows the agencies and institutions of industry/government/academia thathave been taking part in the R&D initiatives. A sound understanding of the R&D activities andsharing information about them among the relevant agencies/institutions plus close collaboration
between decommissioning site and R&D site are critical for linking these activities of both ends
organically and solving these issues raised at decommissioning site effectively via R&D efforts.
Therefore, NDF regularly holds the Decommissioning R&D Partnership Council that consists of
NDF, JAEA, TEPCO, IRID, plant manufacturers, relevant experts and agencies according to thedecision of the Team for Countermeasures for Decommissioning and Contaminated Water
Treatment. The council deals with the following issues:






Sharing information on R&D needs among relevant agencies/institutions
Sharing information on prospective R&D seeds
Arrangement of R&D initiatives based on decommissioning needs
Facilitating R&D collaboration among the relevant agencies/institutions
Facilitating HRD collaboration among the relevant agencies/institutions

It deals with these issues by collecting national/international expertise and addresses these issues
comprehensively and systematically in close collaboration among the relevant agencies/institutions.
In addition, NDF has launched a website in August 2016, English version in May 2017, to provide
a unified collaborative platform for sharing information on R&D needs/seeds and current activities
based on the discussions in the Decommissioning R&D Partnership Council in order to promote
two-way collaboration and active participation of national/international researchers. The website
provides information about issues/challenges the decommissioning site currently faces, R&D
initiatives and achievements, relevant R&D efforts and relevant data on the site including
environmental condition in order to improve convenience and access to credible information for the
researchers.

Japanese site

English site

Figure 6-3 Information Portal for the Research and Development for the Fukushima Daiichi
Decommissioning (http://www.drd-portal.jp/)

6-5

Furthermore, cross-institutional collaboration in academia has been deepened; the “Platform of
Basic Research for Decommissioning” consisting of JAEA/CLADS and research institutes such asuniversities/colleges was founded in February 2016 as a consultative body for R&D promotion. And
in May 2016, “Academic Network Contributing to Fukushima Reconstruction and Reactor
Decommissioning” 70 composed of 32 academies/associations was established with the aim offacilitating activities for Fukushima recovering/decommissioning in a more effective and efficientmanner via information sharing and collaboration.
These R&D organizations have been participating in the R&D initiatives regarding Fukushima
Daiichi NPS decommissioning voluntarily sustained by a crisis awareness. To add attractive
incentives71 in order to make them keep motivated and participate in R&D continuously is one of
the next issues as well.
6.3.1 R&D priority agenda based on needs and its strategic promotion
In the meetings of Decommissioning R&D Partnership Council, it was pointed out that it is important
to make the needs clear and to facilitate prospective seeds to develop until applied/practical levels.
The needs-oriented basic research for achieving specific goals including clarifying phenomenons
is essential, particularly. To facilitate the Fukushima Daiichi NSP decommissioning project that may
continue for over three to four decades in a safely steady and effective manner, it is essential todevelop mid-and-long-term R&D strategies including scientific and technological investigation
based on understandings of the principles and the theories. For this purpose, NDF has built a task
force on research collaboration and specified the Essential R&D Themes that should be
preferentially and strategically targeted. Now, sectional meetings have been established in the
Platform of Basic Research for Decommissioning and discussions to compile R&D strategies forthe themes are underway. (See Table 6-1 and Appendix 6.1.) The task force is still working on theselection of further top essential R&D themes.
6 sub-sessions were established in the Platform of Basic Research for Decommissioning in
February 2017, following the selection of the 6 essential R&D issues. And discussions to develop
an R&D strategy that stipulates the approach to the selected essential R&D themes are underway.
These Essential R&D Themes are fundamental R&D issues to be dealt with based on scientific andtechnological investigation on the issues (needs) regarding the decommissioning project and thestatus of the project. To implement the themes, it should be required to establish a sustainable
framework of COE with outstanding staff or core researches. It is expected the government will
start to conduct or sponsor the R&D activities associated with the themes and consider developinga scheme to have those R&D projects run more effectively in accordance with the status of the
progress of the R&D strategy of the Essential R&D Themes.

70

http://www.anfurd.jp/
As a part the effort, NDF has been holding the International Forum on the Decommissioning of the
Fukushima Daiichi NPS since 2016 (see Section 7). And JAEA/CLADS launched FRC, Fukushima
Research Conferences, aiming at holding the highest status conferences for each area concerning basic
research for decommissioning through whole the year, from 2016. (The conferences were held 4 times
in 2016.)71

6-6

Table 6-1 6 Essential R&D Themes that should be worked on for the clarification of principles in a
strategic and preferential manner
Essential R&D Themes

Descriptions / Background issues

To identify process ofcharacteristic changes in fueldebris over time

The fuel debris retrieval is scheduled for 2021 onward, 10 years after the fueldebris production. And since it is anticipated that the retrieval will require a long
period of time, the fuel debris will remain inside the reactors over 10 years. We
also need to remember that the retrieved debris must be stored safely. Choosing
the best possible methods of retrieve/transmission/storage of fuel debris requires
predictions of characteristic changes of fuel debris over time.

To elucidate corrosion
mechanisms underunusual/extreme
circumstances

It is required to collect data on corrosion under a variety of circumstances withconsideration of the circumstances specific to 1F decommissioning such as highradiation levels and unsteady routes of cooling water in order to prepare forpotential corrosion during decommissioning.

Radiation measurement
technologies adopting
innovative approaches

The radiation levels are still extremely high inside the 1F reactors/buildings due tothe accident and the existing measurement devices do not meet thecapability/functional requirements to provide accurate figures. It is vital to develop
an innovational device adopting brand-new ideas/principles based on 1F needs.

To clarify behavior ofradioactive particulates
generated duringdecommissioning (incl. alphadust treatment)

As thermal cutting of the fuel debris via machine or Laser may produce a large
amount of alpha dust, it requires safety measures and dust containment solutions.
It is necessary to understand physical/chemical properties of alpha dust, to predict
the amount of dust to be produced for each method, and to consider how to seal
the dust according to the results in order to make sure the retrieval will be
conducted in a safe and effective manner.

To understand fundamentallymechanisms of radioactive
contamination

To figure out the mechanism of radioactive contamination towards effective
decontamination; it is critical to implement effective approaches ofdecontamination based on the mechanism of the contamination to radiation
sources, and to decrease the volume of radioactive wastes as well.

Environmental fate studies ofradioactive materials
generated duringdecommissioning

It is essential to clarify the behavior of radioactive materials such as absorption,
dispersion, moving along with groundwater flow in shallow underground in order
to conduct environmental fate studies to ensure they will not affect theenvironment.

6.3.2 Building R&D Infrastructure Based on Mid-and-Long–Term vision
It is essential to work on developing R&D infrastructure and accumulating technological knowledge,
such as previously mentioned Essential R&D Themes, developing generic technologies andcollecting basic data, building up research centers, facilities and equipment, and human resource
development in order to thoroughly implement the long-term decommissioning of Fukushima
Daiichi NPS from technical aspects.
The building for international research collaboration of JAEA/CLADS has opened in April 2017 in
Tomioka-machi, Fukushima prefecture as a place where domestic and overseas universities,
research institutions and industry form a network to promote R&D activities and human resource
development in an integrated manner. The government and the organizations involved should
implement measures to further utilize JAEA/CLADS.
It is also important to develop research infrastructure of hardware. JAEA’s Naraha Remote
Technology Development Center put in service in Naraha-machi, Fukushima Prefecture in April
2016 is equipped with a variety of facilities for developing and demonstrating remote control
devices/equipment. In particular, full-scale mockup test prior to using the devices under extreme
circumstances that human being cannot access is necessary, not only for performance verification
but for training and establishing operating procedure also. Operators are expected to facilitate
active use of the facility.
Fukushima Prefectural Centre for Environmental Creation where Fukushima Prefecture, JAEA andNational Institute for Environmental Studies have their offices opened in Miharu-machi, Fukushima

6-7

Prefecture in July 2016 as a comprehensive center where three different agencies/institutions
collaborate. And in Okuma-machi, Fukushima Prefecture, construction of JAEA Okuma Analysis
and Research Center (radioactive material analysis and research facility) is well underway.
As mentioned above, Research facilities for the decommissioning, contaminated water
management and environmental decontamination measures have been set up mainly inFukushima Prefecture, and those facilities are forming global centers for decommissioning
research and development on the mid-and-long term vision of R&D.
6.3.3 Developing and securing human resource
It is important for long-term sustainable R&D activities to facilitate programs of human resource,
such as for developing and providing human resources of the researchers and engineers.
Our duties toward students are to show them many paths to successful career in the field of
Fukushima Daiichi NPS decommissioning: including continued collaborated efforts between
industry and academia to have them well understand the nuclear industry and its attraction, topresent the “attraction” of Fukushima Daiichi NPS decommissioning as an unprecedented andextraordinarily high level challenge that requires extremely high level of technologies, and to build/clearly offer many different career paths to become established researchers/engineers. To this end,
MEXT’s the Center of World Intelligence Project for Nuclear S&T and Human Resource
Development has been proactively facilitating HRD via research activities among
universities/colleges. (see the column: Conference for R&D Initiative on Nuclear Decommissioning
Technology by the Next Generation (NDEC), and the column: Creative Robot Contest for
Decommissioning.) Additionally, for human resources retention and expansion through the entire
industry, various initiatives such as “Nuclear Dojo” providing nuclear education programs incollaboration among 16 Japanese universities and “Nuclear Facilities Tour for College/High SchoolStudents towards Brighter Future” have been carried out.
In long-term and large-scale projects, such as the decommissioning project at the Fukushima
Daiichi NPS, it is important to develop core personnel for research and development who can
perform scientific and engineering investigation from the academic perspective and personnel witha panoramic perspective (system integrators) who can integrate individual technology seeds into asystem with practical functionality. This activity is being implemented by working on the previously
mentioned Essential R&D Themes. Meanwhile, not only students/researchers, it is also important
to train and develop the engineers of private companies who are working in job sites. Consideringthat the decommissioning of Fukushima Daiichi NPS is such a large-scale complexed project
including many kinds of issues, special engineers who can manage their projects not only within
their specialty but also from a comprehensive viewpoint with the relation between other projects
considering whole scene of the decommissioning are expected.
A subgroup of Professional Engineer, in the Council for Science and Technology of MEXT proposed
directions for reform of the qualifying exam of Professional Engineers (PE) considering
competences required to PE and ability to deal with complicated engineering in their report “The
Future System of Professional Engineer.” (December 22, 2016) In the report, they proposed that
an elective subject “Nuclear Reactor Systems and Facilities” in the second stage of the qualifying
exam for the fields of nuclear power/radiation should include “Reactor Decommissioning including
Handling after Severe Accident,” and another elective subject “Nuclear Fuel Cycle and Radioactive
Waste Management and Disposal” should include “Decommissioning and Nuclear
Fuel/Radioactive Waste Management and Disposal after Severe Accident.” (see Appendix
6.2)Following this, qualifications of the major positions such as PEs are asked in the public offerings
for the Project of Decommissioning and Contaminated Water Management beginning on March2017. It is expected that companies will continue to make effort to expand the employees’

6-8capabilities by encouraging them to acquire related qualifications such as PE, Chief Engineer ofReactors, or Radiation Protection Supervisor, etc.

Column: Conference for R&D Initiative on Nuclear Decommissioning Technology by the Next Generation (NDEC)
Through the MEXT’s HRD program, 7 different institutions, e.g. Tokyo Institute of Technology, The University of Tokyo, Tohoku University, Universityof Fukui, National Institute of Technology Fukushima College, Fukushima University and Japanese Geotechnical Society, are currently engaging indecommissioning R&D initiatives and educating/training young researchers/engineers who will take roles in the future decommissioning. The NDEC,
Conference for R&D Initiative on Nuclear Decommissioning Technology by the Next Generation for students, initiated by the 7 institutions has been
held since FY2015. The NDEC contributes the young students’ career path developing; they are motivated by improving their capabilities via
presentation of their achievements and exchanging their opinion with the decommissioning staff who are actually working and facing serious challenges
at the site.
The 1st conference was held at Aobayama Campus of Tohoku University on March 16, 2015. The 2nd conference was held at Ookayama Campus ofTokyo Institute of Technology on March 7, 2016. The awards were given for excellent oral presentation and poster session at the conferences.

Oral Presentation

Poster Session

Award Ceremony

Column: Creative Robot Contest for Decommissioning
Decommissioning Human Resource Development Consortium, a council of technical colleges to promote research and education related todecommissioning founded by National Institute of Technology Fukushima College, a participant of the MEXT’s HRD Program, called for technicalcolleges and held the Creative Robot Contest for Decommissioning for students at JAEA Naraha Remote Technology Development Centeron December 3, 2016. It was aimed not only at attracting the students’ interest in decommissioning throughout robot building, but developing theircreativity, problem detection/solving skills, as well.
Three months prior to the contest, the participants, 15 teams of students from 13 different technical colleges, took a field trip to Fukushima Daiichi
NPS and JAEA Naraha Remote Technology Development Center (Naraha Summer School) to better understand the current state/technical issues ofthe decommissioning and increased their motivation for the contest.
Assuming very severe condition with extremely high radiation levels that human being cannot access and limitation of the duration of the operationof semiconductor devices, the contest applied some particular rules such as “The robot must be controlled via the footage of the camera installed on
itself, with limited field of vision.” and “the time limit will be extended for devices with radiation shield.” The team of Osaka Prefecture University College
of Technology received an award from the Minister of Education, Culture, Sports, Science and Technology, the highest award in the contest.
A special program featuring the contest was broadcasted on NHK television network. The Consortium, plans to hold the 2nd contest with the same
conditions in FY 2017.

Naraha Summer School

ParticipantsStandardized Testing Field(Photos/Images Provided Courtesy of National Institute of Technology, Fukushima College, Japan Broadcasting Corporation(NHK), Japan Atomic Energy Agency)

6-9

Table 6-2: METI’s Project of Decommissioning and Contaminated Water Management
Project
Purposes and Overviews
R&D Concerning Internal Investigations
Upgrading level of grasping
We perform comprehensive analysis and assessment of fuel debris within reactors and fission
state inside reactor (Figure 6-4) product (FP) to attain a firmer understanding of them, in order to contribute to policy anddecision on fuel debris retrieval, and to confirm the stable condition of the plant.
Development of technologies
We perform analysis and assessment of fuel debris properties, in order to contribute tofor grasping and analyzing
comprehensive analysis and assessment of state inside reactor, to policy and decision on fuelproperties of fuel debris (Figure debris retrieval, and to development of technologies for containing/transportation/storage of
6-5)
fuel debris. Also we perform tests using effective mock debris and develop the necessary
technology to analyze and measure fuel debris that is actually retrieved from reactors in the
future.
Development of investigation
We develop the equipment to inspect and verify the condition inside of the pedestal inside the
PCV and perform plant verification, contribute to policy on fuel debris retrieval. Also we maketechnology of inside of PCV(Figure 6-6)
the development plan in order to carry out the more in-depth actual unit inspection for fueldebris retrieval, and perform the elemental test.
Development of technologies
We develop the equipment and more advanced inspection technology in order to get moreprobable and more extensive understanding of the distribution of fuel debris inside PCV andfor in-depth investigation of
PCV inside (Figure 6-7)
the condition inside and outside of the pedestal and perform the validation in the actual plant,
to contribute to decision on the method for fuel debris retrieval.
Development of investigation
We develop inspection technology to understand the condition of fuel debris, etc. inside theRPV, to contribute to decision on the method for fuel debris retrieval.
technology of inside of RPV(Figure 6-8)
R&D Concerning Fuel Debris Retrieval (Developing Retrieval Methods)
Advancement of retrieval
In technologies for advancement of retrieval method and system of fuel debris and internal
method and system of fuelstructures, we develop technologies to solve the problem extracted as result of concept study,
debris and internal structuresand carry out elemental test as needed.
(Figure 6-9)
Advancement of fundamental
We carry out necessary elemental technology development and test based on past study
technologies for retrieval of fuel results about method and equipment, etc. for retrieval of fuel debris and internal structures.
debris and internal structures(Figure 6-10)
R&D Concerning Fuel Debris Retrieval (Improvement of Work Environment)
Development of sampling
We make the scenario of sampling of actual fuel debris, and study on/develop samplingtechnologies for retrieving fuelequipment to contribute to criticality control, equipment design and rationalization ofdebris and internal structures
construction procedure for the construction of fuel debris retrieval.
(Figure 6-11)
Development and management We establish safety scenario about the important equipment inside RPV and PCV in the caseof evaluation method of seismic of large-scale earthquake and develop evaluation method for measures to prevent and reduce
performance/impact of RPV
the influence, in order to contribute to policy and decision on fuel debris retrieval.
and PCV (Figure 6-12)
Development of criticality
We establish criticality evaluation techniques and develop criticality control technologies such as
control technologies of fuelclose monitoring technique, detection technologies, and prevention technologies using neutron
debris (Figure 6-13)
absorbents, in order to contribute to decision on the method for fuel debris retrieval.
We develop the repair technology for leakage sections to construct confinement function inside
Development of repair
technology for leakage sections PCV and to maintain stable condition in terms of preventing spread of radioactive materials,
radiation shielding and keeping cool, and judge the feasibility to actual plant, in order toin PCV (Figure 6-14)
contribute to policy and decision on fuel debris retrieval.
We perform full-scale test to confirm technique for repairing bottom part of PCV developed by
Full-scale test of repair
technology for leakage sections the project of “Development of repair technology for leakage sections in PCV” in order tocontribute to policy and decision on fuel debris retrieval.
in PCV (Figure 6-15)
Development of technologies
We develop the systems to contain, transport and store retrieved fuel debris safely and surely
for containing, transportationin order to contribute to discussions on scenarios/options regarding fuel debris retrieval
and storage of fuel debris
including policy and decision on fuel debris retrieval.
(Figure 6-16)
R&D Concerning Waste Management
Research and development of
We characterize the solid waste efficiently considering properties of accident-oriented waste,
processing and disposal of
conduct investigation/study towards providing processing technologies, disposal concepts andsolid waste (Figure 6-17)
safety assessment technique and develop the technologies needed for risk reduction regarding
solid waste storage management in order to gain technical views regarding processing/disposal
strategies and the safety by around 2021
※ Created on the basis of Document 4-2 “Progress Status and Direction of Next Plans of R&D Projects” (39th Meeting/Secretariat of theTeam for Countermeasures for Decommissioning and Contaminated Water Treatment, February 23, 2017)6-10

2nd Phase (up to the commencement of fuel debris retrieval)

Topics / Fiscal year
2014 2015
Major events on the current
Roadmap

2016

2017

2018

After 2019

Maintenance/ management of stable plant condition
Preparation of fuel debris retrieval

Decision on the method for fuel debris
Decision on the policy for fuel debris▽ retrieval from the initial Unit
▽ retrieval from each Unit

Understanding and analysis of amount, locations, and properties of fuel debris and FPs<Upgrading Level of Grasping
State inside Reactor>
1.Comprehensive
analysis/evaluation of theconditions inside the reactors

Improvement and advancementof severe accident analysis codes
Comprehensive analysis/evaluation based on actual units data and achievements of other projects
Fuel debris and FPs distribution
Establishment of database for comprehensive analysis/evaluation

2.Estimation/evaluation of fueldebris behavior and fission
products behavior/property

Reduction of uncertainties using analysis method ▽Improvement of reliability of reactor event estimation
Evaluation of chemical properties of FPs ▽Evaluation of chemical properties of FPs
Utilization of domestic/foreign knowledge
OECD/NEA/BSAF‐1 through the international joint project ▽Estimation of fuel debris behavior
OECD/NEA/BSAF‐2‐Measured data of actual units‐Achievements of other projects
e.g. Development of Investigation
Technology of Inside of PCV,
Grasping properties of Fuel Debris: Field work(including engineering): R&D

Figure 6-4 Process chart for “Upgrading level of grasping state inside reactor”

Figure 6-5 Process chart for “Development of technologies for grasping and analyzing propertiesof fuel debris”

6-11

2nd Phase (up to the commencement of fuel debris retrieval)

Topics / Fiscal year
2014 2015
Major events on the current
Roadmap

2016

2017

2018

After 2019

Maintenance/ management of stable plant condition
Preparation of fuel debris retrieval

Decision on the method for fuel debris
Decision on the policy for fuel debris▽ retrieval from the initial Unit
▽ retrieval from each Unit

A2 inspection (on‐site validation)▽

On‐site inspection
A2’ inspection (on‐site validation)▽▽(on‐site validation for in‐depth inspection)<Development of Investigation
B2 inspection▽
Access route search forTechnology of Inside of PCV>
(on‐site validation)
in‐depth inspection ▽inspection of inside of Unit 3▽
1.Making and update of theinspection/development plan
・Evaluation/rearranging ofDevelopment of Technologies for In‐depth Investigation of PCV Insideinspection result of inside
Evaluation/rearranging of inspection result・Continuous update of(Next phase inspection plan shifted to the project of “Development
Making and update of theinspection planof Technologies for In‐depth Investigation of PCV Inside”
inspection plan
・ Development plan for the
Making of plan (including(Next phase development plan shifted to the project of “Development
next phase inspection ofelemental test plan)of Technologies for In‐depth Investigation of PCV Inside”
inside・Preparation of access to PCVConcept design, trial manufacturing (The development continues in the project of “Development offor next phase inspection ofof the device and elemental test
Technologies for In‐depth Investigation of PCV Inside”
inside2.Development ofaccess/inspection device andsystem to specific areas
Design/manufacturing/test(1)Inspection of platformbottom of Unit 2 (A2’)(2)Inspection in the pedestal of
Design/manufacturing/test
Unit 3・ Inspection in the pedestal of
Unit 2(A2)・ Inspection outside thepedestal of Unit 1(B2)Concept study and elemental test
(The development continues in the project of “Development of・Concept study and elementalof access device and inspection
Technologies for In‐depth Investigation of PCV Inside”
test for underground floor
technology
inspection: Field work(including engineering): R&D

Figure 6-6 Process chart for “Development of investigation technology of inside of PCV”
2nd Phase (up to the commencement of fuel debris retrieval)

Topics / Fiscal year
2014 2015
Major events on the current
Roadmap

2016

2017

2018

After 2019

Maintenance/ management of stable plant condition
Preparation of fuel debris retrieval

Decision on the method for fuel debris
Decision on the policy for fuel debris▽ retrieval from the initial Unit
▽ retrieval from each Unit

A2 inspection (on‐site validation)▽
A2’ inspection (on‐site validation)▽
B2 inspection▽<Development of Technologies
for In‐depth Investigation of
PCV Inside>(on‐site validation)

inspection of inside of Unit 3▽▽On‐site inspection (on‐site validation)
Access route search forin‐depth inspection ▽Preparation of Preparation of
on‐site search on‐site inspection

Inspection result of the project “Developmentof Investigation Technology of Inside of PCV”
1.Making of theinspection/development Inspection/development plan
Update of plan
Continuous review of planof the project “Development
plan▽of Investigation TechnologyofInsideof
PCV”
・Making of the plan based(Basic design) Update of plan/design/manufacturing/teston the policy on fuel debris
retrieval, etc.
・Preparation of the inspection
for access to PCV
2.Development of device andtechnology
・Development ofaccess/inspection device
・Applicable inspection ofelemental technology(Basic concept design) Update of plan/design/manufacturing/test

Review of plan based on
inspection result, as needed

The development continues
according to development plan

(Plan/elemental test) Update of plan/elemental test/design/manufacturing: Field work(including engineering): R&D

Figure 6-7 Process chart for “Development of technologies for in-depth investigation of PCV
inside”

6-12

2nd Phase (up to the commencement of fuel debris retrieval)

Topics / Fiscal year
2014 2015
Major events on the current
Roadmap

2016

2017

2018

After 2019

Maintenance/ management of stable plant condition
Preparation of fuel debris retrieval

Decision on the method for fuel debris
Decision on the policy for fuel debris▽ retrieval from the initial Unit
▽ retrieval from each Unit
On‐site inspection▽
Evaluation of on‐site feasibility/on‐site study/preparation of on‐site inspection<Development of Investigation
Technology of Inside of RPV>
1.Making of theinspection/development plan

Equipment manufacturing for actual reactor (including preparation, etc.)/mock‐up test
Confirmation offeasibility of bothinspection methods On‐site applicable plan

Confirmation of development
plan of inspection method bylateral perforating

2.Development of equipment
Concept design ofto access the reactor coreinspection method by
Basic design
upper part perforating
Detailed design2‐1) Development ofElemental test/confirmation of partial mock‐up
equipment to access fromupper part
Study on applicability
Basic design/of inspection method
detailed design
by lateral perforating Concept design2‐2) Development ofequipment to access from
Elemental testside
detailed design

Basic design

3.Development/selection of
inspection method up to coreregionConcept design

4.Overall system design forinspection equipment andmethod planning

Overall system design/method planning/study of safety ensuring

Elemental test

: Field work(including engineering): R&D

Figure 6-8 Process chart for “Development of investigation technology of inside of RPV”
2nd Phase (up to the commencement of fuel debris retrieval)

Topics / Fiscal year
2014 2015
Major events on the current
Roadmap

2016

2017

2018

After 2019

Maintenance/ management of stable plant condition
Preparation of fuel debris retrieval

Decision on the method for fuel debris
Decision on the policy for fuel debris▽ retrieval from the initial Unit
▽ retrieval from each Unit

<Advancement of Retrieval
Study result until last fiscal year
Method and System of Fuel
Debris and Internal Structures>

1. Technology development
about confinement function
2. Technology development
about collection/removal ofdust derived from fuel debris
3. Study on monitoring system
of α nuclide by fuel debris
retrieval
4. Optimization study on safety
ensuring of method andsystemTechnology development about building of confinement function

Technology development for reduction/removal of dust (including rearranging of data requirement)

Concept study on the system/making of development plan

Study on applicability ofexisting technology

Optimization study on safety ensuringof method and system<Advancement of Fundamental
Technologies for Retrieval of
Fuel Debris and Internal
Structures>
engineering

Basic design ofthe system: Field work(including engineering): R&D

Figure 6-9 Process chart for “Advancement of retrieval method and system of fuel debris andinternal structures”

6-13

2nd Phase (up to the commencement of fuel debris retrieval)

Topics / Fiscal year
2014 2015
Major events on the current
Roadmap

2016

2017

2018

After 2019

Maintenance/ management of stable plant condition
Preparation of fuel debris retrieval

Decision on the method for fuel debris
Decision on the policy for fuel debris▽ retrieval from the initial Unit
▽ retrieval from each Unit

<Advancement of Fundamental
Technologies for Retrieval of
Fuel Debris and Internal
Study on access method/elemental technologies development/study on safety system
Structures>
1. Development of common
fundamental technologies
concerning retrieval of fueldebris and internal structures

Development of collection system of fuel debris
Development of cutting/dust collection system of fuel debris

1.1 Technology development
for prevention of dust
diffusionDevelopment of prevention method of dust diffusion
Development of setting method of work cell

1.2 Elemental technology
development for setting of
retrieval equipment

Development of obstacle removal technology at the time of fuel debris retrieval
Development of monitoring technology at the time of fuel debris retrieval

1.3 Development ofmonitoring technology at thetime of fuel debris retrieval

Development of remote maintenance technology for fuel debris retrieval
equipment

2. Development of remote
maintenance technology forfuel debris retrieval
equipment<Advancement of Retrieval
Method and System of Fuel
Debris and Internal
Structures>

Manufacturing/setting of fuel debris retrieval equipment
Mock‐up: Field work(including engineering): R&D

Figure 6-10 Process chart for “Advancement of fundamental technologies for retrieval of fueldebris and internal structures”
2nd Phase (up to the commencement of fuel debris retrieval)

Topics / Fiscal year
2014 2015
Major events on the current
Roadmap<Development of Sampling
Technologies for Retrieving
Fuel Debris and Internal
Structures>

2016

2017

2018

After 2019

Maintenance/ management of stable plant condition
Preparation of fuel debris retrieval

Decision on the method for fuel debris
Decision on the policy for fuel debris▽ retrieval from the initial Unit
▽ retrieval from each Unit

Remark)*1 If the technology developed in the project of “Development of
Investigation Technology of Inside of PCV” is available, that technology is used
Formulation of scenario

1.Study and formulation ofsampling scenario of fueldebris2.Design/trial manufacturing ofsampling system andequipment of fuel debris in
PCV
2.1 Basic design of samplingsystem of fuel debris
2.2 Design/trial manufacturingof access equipment
2.3 Design/trial manufacturingof collection equipment offuel debris sample

System design
Design of access equipment*1
Design of collection equipment for
sample such as pebble/sandTrial manufacturing ofaccess equipment*1
Trial manufacturing of equipment
collection equipment for sample
such as powder/column

Formulation of concept
3. Concept design of samplingsystem of fuel debris in RPV

Elemental test

On‐site inspection(on‐site validation): Field work(including engineering): R&D

Figure 6-11 Process chart for “Development of sampling technologies for retrieving fuel debris
and internal structures”

6-14

2nd Phase (up to the commencement of fuel debris retrieval)

Topics / Fiscal year
2014 2015
Major events on the current
Roadmap

2016

2017

2018

After 2019

Maintenance/ management of stable plant condition
Preparation of fuel debris retrieval

Decision on the method for fuel debris
Decision on the policy for fuel debris▽ retrieval from the initial Unit
▽ retrieval from each Unit

Development of RPV/PCV seismic resistance and impact assessment method (engineering)
Actual unit condition evaluation<Development andManagement of Evaluation
Method of Seismic
Performance/Impact of RPV
and PCV>
1.Establishment of safety
scenario in the case of large
scale earthquake
2.Development of seismic
resistance/impact
assessment method toestablish safety scenario
3.Enhancement of safety
scenario

Study onsafety scenario▽(Interim evaluation)

Safety scenario▽

Study ofcountermeasure▽(Interim evaluation)

Measures forsafety scenario▽

Analysis ofpedestal▽(Interim evaluation)▽Seismic resistance and impact assessment method▽(Interim evaluation)▽Seismic resistance and impact assessment method

Analysis of
S/C supports

Verification
conformation ▽(Interim evaluation)measures for
upgrading▽Advancement of safety scenario<Development of Repair
Technology for Leakage
Sections in PCV><Advancement of Retrieval
Method and System of Fuel
Debris and Internal
Structures>: Field work(including engineering): R&D

Figure 6-12 Process chart for “Development and management of evaluation method of seismic
performance/impact of RPV and PCV”
2nd Phase (up to the commencement of fuel debris retrieval)

Topics / Fiscal year
2014 2015
Major events on the current
Roadmap<Development of Criticality
Control Technologies of Fuel
Debris>1.Establishment of criticality
evaluation technique
2.Development of criticality
control technology(1)Sub‐criticality monitoring
methods(2)Re‐criticality detection
technology(3)Criticality prevention
technology

2016

2017

2018

After 2019

Maintenance/ management of stable plant condition
Preparation of fuel debris retrieval

Decision on the method for fuel debris
Decision on the policy for fuel debris▽ retrieval from the initial Unit
▽ retrieval from each Unit

Detailed evaluation of criticality control
method of the initial Unit

Feasibility evaluation on criticality control
method for multiple methods

Elaboration based on the latest findings▽

Study on criticality control method
Study on criticality evaluation/criticality scenario(reflected the latest findings)▽Indicating data to assess applicability to the actual unit
▽Indicating application method to the actual unit
▽Indicating data to assess applicability to the actual unit
▽Indicating application method to the actual unit
▽Insoluble neutron absorber
Selection of applicable material tothe actual unit

▽Indicating of feasibility validation data

<Upgrading Level of Grasping
State inside Reactor><Advancement of Retrieval
Method and System of Fuel
Debris and Internal
Structures>: Field work(including engineering): R&D

Figure 61-3 Process chart for “Development of criticality control technologies of fuel debris”

6-15

2nd Phase (up to the commencement of fuel debris retrieval)

Topics / Fiscal year
2014 2015
Major events on the current
Roadmap

2016

2017

2018

After 2019

Maintenance/ management of stable plant condition
Preparation of fuel debris retrieval

Decision on the method for fuel debris
Decision on the policy for fuel debris▽ retrieval from the initial Unit
▽ retrieval from each Unit

<Development of Repair
Assessment of applicability
Technology for Leakage
to the actual unit▽
Sections in PCV>
1.Process study/planning of(Complete concept study) →Carry out in accordance with the project of “Upgrading
filling water up to PCV
2.Development of technique for Approach and System for Retrieval of Fuel Debris and Internal Structures”
repairing bottom part of PC(Interim evaluation Assessment of Applicability to the(1)Test for long‐term water▽ (I.E.))▽ actual unit (A.A.) Study of repair method
sealing, etc.
▽ A.A.
a)Reinforcement of S/C supports
Improvement of equipment/combination test
▽I.E.
b)Filling vent pipe
▽ A.A.
▽I.E.
Improvement of equipment/combination testc)Placing S/C injection material Feasibility check ▽I.E.
(F.C.)▽▽A.A.
(2)Filling vacuum break line
▽I.E.
F.C.▽
▽A.A.
(3)Construction of boundary forconnection piping
▽I.E.
F.C.▽
▽A.A.
(4)Wall surface of torus room
piping opening
3.Development of technique forrepairing upper part of PCV
F.C.▽
▽I.E.
▽A.A.
(1)Sealed part
▽F.C. of air seal
F.C. of water seal▽▽A.A.
(2)Piping bellows▽ ▽F.C.
▽A.A.
(3)D/W shell repair
4.Cocept study of environment Study of repair method
improvement for application
to the actual unit of repair
method・Method to improve
▽A.A.
▽F.C.
environment, etc. according
to the application method foreach repair areas<Full‐scale test of Repair
Technology for Leakage
Sections in PCV><Advancement of Retrieval
Method and System of Fuel
Debris and Internal
Structures>: Field work(including engineering): R&D

Figure 6-14 process chart for “Development of repair technology for leakage sections in PCV”

2nd Phase (up to the commencement of fuel debris retrieval)

Topics / Fiscal year
2014 2015
Major events on the current
Roadmap<Full‐scale test of Repair
Technology for Leakage
Sections in PCV>
1.Full‐scale test of repair andwater sealinga)Reinforcement of S/Csupports

2.Integrity check forreinforcing materials/water
seal materials after testing
3.Maintenance of VR data forpreliminary simulation test

2017

2018

After 2019

Maintenance/ management of stable plant condition
Preparation of fuel debris retrieval

Decision on the method for fuel debris
Decision on the policy for fuel debris▽ retrieval from the initial Unit
▽ retrieval from each Unit

Development of water seal technology
for application to the actual unit

Full‐scale test of S/C support reinforcement
Manufacturing
Test implementationof test unitb)Water sealing of vent pipesc)Water sealing of S/C injection
material(Water sealing of downcomer)

2016

Technology
development
Technology
development

Full‐scale test of water sealing of vent pipes
Test implementation
Full‐scale test of water sealing of S/C injection material
Test implementation

Collection of data/validation
Complete making
basic data ▽▽ to the actual unit
Improve basic data accuracy<Development of Repair
Technology for Leakage
Sections in PCV><Advancement of Retrieval
Method and System of Fuel
Debris and Internal
Structures>▽Plan on completing development: Field work(including engineering): R&D

Figure 6-15 Process chart for “Full-scale test of repair technology for leakage sections in PCV”

6-16

2nd Phase (up to the commencement of fuel debris retrieval)

Topics / Fiscal year
2014 2015
Major events on the current
Roadmap

2016

2017

2018

After 2019

Maintenance/ management of stable plant condition
Preparation of fuel debris retrieval

Preparation for storage, transportation and repository<Development of Technologies
for Containing/Transportation/Storage of Fuel Debris>
1.Survey and making ofdevelopment plan abouttransportation and storage
2.Study on fuel debris storing
system
3.Development of safety
evaluation method
4.Development of storage
technology for fuel debris
5.Development of transfer/storage technology forstorage canister<Development of Technologies for
Grasping and Analyzing
Properties of Fuel Debris><Advancement of Retrieval
Method and System of Fuel
Debris and Internal Structures>

Decision on the method for fuel debris
Decision on the policy for fuel debris▽ retrieval from the initial Unit
▽ retrieval from each Unit

Manufacturingof actual unit
Mock‐up test

Planning▽

Planning and update based on related projects, etc.

Study of system optimization based on study result on related projects, etc.
Determination of safety
evaluation method▽

Structure verification test

Implementation of safety
evaluation/additional test

Determination of basic
specification of storage canister▽
Optimization of specification of storage canister based onrelated projects, etc.※

Determination of
specification of storage
canister/handlingequipment for mock‐up▽ Trialmanufacturing

Optimization of specification of handling based on related
projects, etc.※
Determination of safety requirements abouttransportation/repository facilities※To be appropriate design based on
information from related projects andadjustment between projects

Trialmanufacturing: Field work(including engineering): R&D

Figure 6-16 Process chart for “Development of technologies for containing, transportation andstorage of fuel debris”

6-17

2nd Phase (up to the commencement of fuel debris retrieval)

Topics / Fiscal year
2014
Major events on the current
Roadmap<Research and Development ofProcessing and Disposal of Solid
Waste>
Ⅰ. Characterization

2015

2016

2017

2018

Establishment of basic concept ofprocessing/disposal for solid radioactive wastes ▽After 2019Technical prospect for processing/disposal measures and the safety ▽Preparation of samplingof rubble, ALPS, soil,
incineration ash andhigh dose sample, andexhibition of data

Efficiency of sampling/analysis method of rubble, ALPS, soil,
incineration ash, sample in R/B and high dose sample, andestablishment of database

Response toprogress ofsampling/analysis

2.Accuracy improvement ofanalytic evaluation method

Development of evaluation
method for secondarywater treatment waste,
rubble, trees and soil

Accuracy improvement of analytic inventory evaluationreflecting the variation of analysis result

Upgrading ofevaluation method

3.Comprehensive report ofinventory evaluation

Making/update ofanalysis plan

1.Collection/management, etc. ofanalysis data

4.Response to disposal influencematerial, etc.

Comprehensive evaluation of the estimate of analysis data and radioactivity
inventory/estimation of inventory/confirmation of update flow
Rearranging of the concept about temporary acceptance density
concerning management before the disposal and disposal facilities

Ⅱ. Management before thedisposal

Storage

1.Applicability evaluation ofTechnology survey/test/making
solidification technology suitable of catalogue /presentation ofcandidate technology
for character of solid waste
2.Study/evaluation on storage/management method suitablefor character of solid waste(1)Study on storage measures ofhigh dose waste(2)Stabilizing treatment
technology of secondary water
treatment waste

Soundness evaluation aboutstorage of Cs adsorption
tower, etc./study andpresentation of measures
Study and selection ofstabilization technology of
ALPS pretreatment slurry

Survey and evaluation of applicability/performance test/basic test about influence of radioactivity and heat
Study and presentation about reduction measures of hydrogen
outbreak and requirements of vent, etc. at the time of hydrogen
outbreak/study on storage method of waste such as rubble atthe time of fuel debris retrieval
Applicability evaluation of stabilization technology of waste sludge,
etc. and performance test

Study of disposal concept
Ⅲ. Study on disposal concept and and safety evaluation
Survey about disposal method at homemethod in the country
safety evaluation method
and abroad
suitable for solid waste character existing

Study on waste stream

performance test
and evaluation

Study on theevaluation/study
on measures
based on on‐site
condition

Survey/study on pollution measurement
and evaluation method for reduction of Study of theamount of waste
evaluation

3.Study on reduction technologyof amount of waste

Ⅳ. Integrating of R&D results

Preparation ofanalysis evaluationof influence

Making of draft plan
/reflection of result/review

Study on disposal
concept andStudy on disposal concept and safety evaluation
safety evaluation scenario
model, etc.

Comprehensive evaluation of progress/adjustment/problem of R&D

Evaluation basedon progress of
R&D: Field work(including engineering): R&D

Figure 6-17 Process chart for “Research and development of processing and disposal of solid
waste”

6-18

Figure 6-18 Structure of MEXT’s ‘The Center of World Intelligence Project for Nuclear S&T and
Human Resource Development’
Table 6-3: Adopted projects of the Acceleration Program for Reactor Decommissioning R&D (Domestic research)Core InstituteName ofRepresentative

Project name

FY2016【Theme 1】Research on Fuel Debris Retrieval
Shizuoka University

MinoruWatanabeDevelopment of a radiation-hardened embedded system used for robots
decommissioning nuclear reactors【Theme 2】Research on environment countermeasures including those against waste
Hokkaido University

Tamotsu
Kozaki

Nagaoka University ofTechnology

KazuyukiTakaseJapan Atomic Energy
AgencyTakahito
Osawa

Study for Dismantling of Radioactive Concrete and Reasonable Disposal of
Resulting Waste
R&D on Technology for Reducing Concentration of Flammable Gases Generated
in Long-Term Waste Storage Containers
Development of Automatic Preprocessing System for Radionuclide Analysis in
Wastes by Robot Control System

FY2015【Theme 1】Research on Fuel Debris Retrieval
Hokkaido University

TadashiNarabayashi

Tokyo Institute ofTechnology

Yoshinao
Kobayashi

Japan Atomic Energy
AgencyIkuo Wakaida

Development of High Efficiency Multi-Nuclide Aero Filter for Reducing
Occupational Radiation Exposure
Investigation on the Fracture Behavior of Core Structure Materials by Molten
Corium in Boiling Water Reactor Plants during Severe Accidents
Advanced study on remote and in‐situ elemental analysis of molten fuel debris indamaged core by innovative optical spectroscopy【Theme 2】Research on environment countermeasures including those against waste
Tohoku University

Naoki Asao

Decontamination of radiation-tainted water with newly-designed metal oxide
nanomaterials

Japan Atomic Energy
Institute

Kazuki Iijima

Estimation of on-site radionuclides inventories of Fukushima Daiichi NPS based on
their off-site distribution

6-19

Table 6-4: Adopted projects of the Acceleration Program for Reactor Decommissioning R&D (Joint
research between Japan and the UK)
Core InstituteName ofRepresentative

UK Side Core
Institute

Project name

FY2016【Theme 1】Research on Fuel Debris Retrieval
University of Tokyo

Akira
Yamaguchi

Imperial College
LondonInvestigation of the safe removal of fuel debris: risk
assessment and multi-physics simulation【Theme 2】Research on environment countermeasures including those against waste
Hokkaido University

Tsutomu Sato

Long-term performance of cement disposal systems forsynthetic zeolites and titanates arising from reprocessing
University of Sheffieldof contaminated water

FY2015【Theme 1】Research on Fuel Debris Retrieval
Tokyo Institute ofTechnology

Hiroshige
KikuraUniversity of Bristol

Nagaoka University ofTechnology

Jun Katakura

Lancaster University

An ultrasonic measurement system and its robotic
deployment into vessels for the combined assessment ofdebris condition and water leakage
Technology development to evaluate dose rate distributionin PCV and to search for fuel debris submerged in water【Theme 2】Research on environment countermeasures including those against waste
Kyushu University

Advanced Waste Management Strategies for High Dose
Yaohiro Inagaki University of Sheffield Spent Adsorbents

Japan Atomic Energy
AgencyYoshihiro
MeguroDevelopment of solidification technique with minimisedwater content for safe storage and disposal of secondary
University of Sheffieldradioactive aqueous wastes in Fukushima

Table 6-5: Adopted projects of the Acceleration Program for Reactor Decommissioning R&D (Joint
research between Japan and the US)
Core InstituteName ofRepresentative

US Side Core
Institute

Project name

FY2016【Theme 1】Research on Fuel Debris Retrieval
Japan Atomic Energy
AgencyToshihiko
Onuki

Texas A&M University

6-20

Using Radioiodine Speciation to Address Waste Stream
Stabilization Problems at the Fukushima Daiichi Nuclear
Power Plant and a DOE Site

Table 6-6: Adopted projects of the HRD and Research Program for Decommissioning of the Fukushima Daiichi NPSCore InstituteName ofRepresentative

Project name

FY2015
University of Fukui

Yoshinari
Anoda

Fukushima University

Yoshitaka
Takagai

National Institute ofTechnology, Fukushima
College

Katsuhiro
AoyagiJapanese GeotechnicalSocietyIkuo Tohata

Research and human resource development for analysis of fuel debris anddecommissioning technology of Fukushima Daiichi nuclear power plants

Development of analytical specialist by multi-phases educational system and therapid measurement system for hard-to-measure nuclides in the decommissioning
support techniques
NIT Fukushima College R&D and Education Program for the Decommissioning ofFukushima Daiichi NPP under the Collaboration of Nationwide NIT colleges,
Universities, National Institutes and the Local Private Companies

Geotechnical perspectives towards the solution for Fukushima No. 1 Nuclear Power
Plant

FY2014

Fundamental Research and Human Resources Development Program for Nuclear
Decommissioning Related to Integrity Management of Facilities including PrimaryContainment Vessel and Buildings, and Fuel Debris Processing and Radioactive
Waste Disposal

Tohoku University

Nobuyoshi
Hara

University of Tokyo

Koji Okamoto

HRD for The Fukushima Daiichi Decommissioning based on Robotics and Nuclide
Analysis

Tokyo Institute ofTechnology

Toru Obara

Advanced Research and Education Program for Nuclear Decommissioning(ARED)

6-21

7. Enhancement of international cooperation
The Fukushima Daiichi NPS, which experienced a core meltdown accident and hydrogen
explosions, was brought into a cold shutdown state after taking emergency measures. However,there exists a great amount of radioactive materials even today. Its decommissioning is anextremely complicated and difficult task that we have never experienced in Japan. On the other
hand, there are a lot of experiences and expertise in the world on decommissioning and cleanup
of nuclear facilities and sites that suffered accidents or contamination. To actively learn and makeuse of these leanings will help to facilitate safer and faster decommissioning of Fukushima Daiichi
NPS, and it is important for relevant organizations in Japan to proactively enhance international
cooperation.

7.1 Integrating and utilizing wisdom(1) Purpose
For the Fukushima Daiichi NPS decommissioning, in order to make good use of the knowledge
and experiences of other countries, it is important to facilitate information exchange, technology
introduction, R&D cooperation, and evaluation and advice from international experts.

(2) Current efforts
As a strategic formulation organization, NDF has implemented various activities in collaboration
with relevant organizations in Japan, to make full use of knowledge and experience of othercountries in the fields such as fuel debris retrieval, waste management andimplementation/management of decommissioning. Specifically, NDF has been engaged in theinternational collaboration activities, including participation in the activities of the internationalorganizations such as OECD/NEA and IAEA, signing the co-operation agreement with NDA, UK
and memorandum of understanding with CEA, France for information exchange, and also
participation in the discussions under the governmental framework between UK, France, US,
Russia and Japan, as shown in Table 7-1. NDF also invites distinguished experts from UK, US,
France and Spain as International Special Advisers to receive advice from them. TEPCO and IRID
also conduct similar international collaboration activities, as an approach to get advice fromoverseas experts according to their respective roles (see the column: International Special
Advisors/International Expert Group/International Advisors).
NDF also holds the Decommissioning R&D Partnership Council, in order to promote effectively andefficiently the research and development projects of individual organizations to realize overall
optimization. Table 7-2 shows the examples of the international studies and knowledge sharing
activities conducted in the past by the domestic relevant organizations. Furthermore, in October
2016, NDF joined the TMI-2 workshop held at the Idaho National Laboratory in cooperation withthe domestic relevant organizations. NDF interacted with attendees from Department of Energy(DOE), Nuclear Regulatory Commission (NRC), and General Public Utilities Nuclear (GPUN) who
actually took part in the fuel debris retrieval at TMI-2, and learned the importance of flexibility inimplementation of the project from their experience of TMI-2 fuel debris retrieval.

7-1

(3) Efforts in the near future
As technically challenging works progress and appropriate project management is required toensure the decommissioning for a mid- and long term, the necessity to enhance international
cooperation and learn from overseas knowledge and experiences is increasing.
In the UK, NDA was established in 2005 as a non-departmental public organization responsible fordecommissioning of nuclear facilities and ensuring the management of radioactive wastes. The
projects such as cleanup of the First Generation Magnox Storage Pond in Sellafield72 are currently
under way. In the US, the retrieval of fuel debris from a damaged reactor at TMI-2 which
experienced a core meltdown accident in 1979 (refer to Appendix 7) was conducted. Furthermore,
the legacy clean-up activities in more than one hundred nuclear weapons production facilities havebeen conducted for over 60 years. In France, long-term decommissioning activities such as CEA’s
project at the Marcoule nuclear site, including the decommissioning of reprocessing plant UP-173
are undergoing. Russia also has experiences in taking responses to accidents in nuclear reactors
and reprocessing plants.
In the future, when understanding of the PCV internal is difficult, it is important to work on thesevarious challenges: fuel debris retrieval, radioactive waste management, effective and flexible
approaches to decommissioning with uncertainties, risk-informed prioritization of decommissioning
tasks, optimizing the decommissioning program, and building accurate understanding among the
local residents on decommissioning. From the viewpoint of promoting these efforts smoothly, it isnecessary to learn knowledge and experience, both successes and failures, from the abovementioned countries and relevant organizations that have experienced decommissioning. In
addition, in order to progress each effort, it is important to share the latest knowledge to contributeto R&D and to seek for applicable technologies.
In collaboration with overseas organizations, the information provided by the UK includes the G6
framework consisting of relevant organizations (Sellafield Ltd., NDA, ONR, etc.) to share solutions
to achieve common targets and thus realize the early reduction of risks, and the effectiveness ofthe concept of enabling regulation74. The US provided the information on the effectiveness to work
as one team known as GEND (GPU, EPRI, NRC, and DOE), to take required measures after TMI2 accident. The proper relationship with domestic relevant organizations is important, and it isnecessary to learn the efforts in other countries. In addition, at the workshop on stakeholder
involvement in the decision-making in the nuclear energy sector held by OECD/NEA, theimportance of dialogue with local residents was reconfirmed. Based on these overseas efforts, it isnecessary to strengthen the efforts through the International Forum on the Decommissioning of the
Fukushima Daiichi Nuclear Power Station, etc.
As mentioned above, the domestic relevant organizations, such as the government, NDF, IRID,
JAEA, and TEPCO, have been building cooperative relationships with overseas organizations, and72

At the First Generation Magnox Storage Pond at the Sellafield site, Magnox fuels were erodedfor decades, and about 1,500 m3 of sludge accumulated. In 2015, the first radioactive sludge wasremoved.
73 Reprocessing Facilities UP-1 were shut down in 1998. It is planned to disassemble major plantcomponents by around 2035 and to complete the retrieval of remaining wastes until around 2040.
74 The term “enabling regulation” is used by the Office for Nuclear Regulation (ONR) to express
its regulatory approach. ONR has open and constructive dialogues with dutyholders and relevant
stakeholders, aiming to effectively realize clear and prioritized safety targets. This approach
reduces delay in ensuring safety and enables best use of resources, by executing the fit-forpurpose solutions and by exercising proportionate regulations.

7-2

establishing relationship with them and obtaining advice from international experts according totheir individual roles specified in Chapter 1. Continuously, it is necessary to gather and utilize
knowledge and experiences on decommissioning activities of other countries and to be evaluatedby the international standards, and thus to move forward properly and steadily with thedecommissioning at the Fukushima Daiichi NPS by integrating and utilizing wisdom from Japanand overseas.

Table 7-1
Framework

Intergovernmental Frameworks between Japan and Other Countries
Description
Japan-side
Counterpart
organization
organization
Annual Japan-UK ・This dialogue is held based on the
MOFA
FCO
METI
BEIS
Nuclear Dialogue
appendix to the joint statement of
MEXT
EPSRC
the Japan-UK top level meeting in
MOE
DEFRA
April 2012, “Japan-UK Framework
NRA
ONR
on Civil Nuclear Energy
NDA
Cooperation” (1st meeting: February NDF
TEPCO
NNL
2012).
Sellafield Ltd
Japan-France
Nuclear Energy
Committee

・It was established under the jointstatement of Japan–France toplevel meeting in October 2012 (1stmeeting: February 2012).

MOFA
METI
MEXT
MOE
NRA
NDF

DGEC
CEA
ASN
IRSN
ANDRA

Japan-US
Decommissioning
andEnvironmental
Management
Working Group・After the Fukushima Daiichi NPS
accident in March 2011, theestablishment of the US-Japan
Bilateral Commission on Civil
Nuclear Cooperation (the BilateralCommission) was announced in
April 2012 based on the close
relationship between Japan and the
US to further reinforce bilateral
cooperation. Under thiscommission, the Decommissioning
and Environmental Management
Working Group (DEMWG) wasestablished (1st meeting:
December 2012).

METI
MEXT
MOE
NDF
JAEA
TEPCO

DOE
DOC
EPA
SRNL

Japan-Russia
Nuclear Working
Group・The Nuclear Working Group wasestablished after confirming that
Energy is one of the eight areas ofcooperation plan approved at theJapan-Russia top-level meeting in
September 2016, (1st meeting:
September 2016).

MOFA
METI
MEXT
NDF
JAEA
TEPCO

ROSATOMTENEXFCNRSPDC UGRRosRAO
MES
Kurchatov
Institute
Physics andEnergetics
Institute

7-3

Table 7-2 Examples of International Research and Knowledge Sharing Activities
Activity
Description
Relatedorganizations
・Research organizations and governmental organizations InstituteofOECD/NEA
from eleven countries joined to conduct benchmark Applied
BSAF
study using severe accident analysis codes developed Energy (IAE)by these organizations to find out how the accident in NRA
the Fukushima Daiichi NPS progressed and how the fuel JAEA
debris and FPs spread inside the reactors. Knowledge
andfindings
relatedtothemodelingof CRIEPI
phenomenological issues obtained by member NDF
TEPCO
countries’ organizations are being utilized.
・ Data measured during the accident and information
database regarding the post-accident radiation levels
are shared.
・
Regulations and standards in various countries and NRA
OECD/NEA
operators’
efforts are shared and summarized on how to NDF
WGAMA-LTMNPP
ensure safe and stable status in the nuclear power TEPCO
plants in which fuel remains after severe accidents.
・Knowledge available to understand the characteristics of JAEA
OECD/NEA
fuel debris is shared and the methods for safety NDF
PreADESevaluation on fuel debris sampling and retrieval are TEPCO
organized.
・It is planned to analyze information obtained from the JAEA
OECD/NEA
inside of reactor building/PCV and information about NRA
ARC-F
water sample and to share the useful knowledge to NDF
understand the status of progress of the severe accident TEPCO
and the conditions of the reactor buildings and PCVs in
depth. (In progress, details to be determined)
・ From the viewpoint of proceeding with basic, JAEA
OECD/NEA
fundamental research for the Fukushima Daiichi NPS, a
TCOFF
thermodynamic database, which is suited for thematerial science analysis to understand the migration
behavior of molten fuel and FPs and the characteristics
of fuel debris, is improved and enlarged.
OECD/NEA
・Expansion of knowledge for waste management and
NRA
EGFWMD
decommissioning at the Fukushima Daiichi NPS
JAEA・ Advice to Japan’s R&Ds regarding waste in the TEPCO
Fukushima Daiichi NPS
METI
NDF
IRID
・Knowledge and experience obtained from the efforts on NRA
IAEA DaRoD
challenges of decommissioning and recovery of NDF
damaged nuclear power facilities (regulations,
technologies, systems, and strategies) are shared
among the relevant countries.
・Japan-UK joint nuclear research adopted in FY 2015 and MEXT
Bilateral
FY 2016
government-based
Theme:
programs- Joint research on fuel debris retrieval- Joint research on environmental measures including
measures for the management of radioactive waste
・Japan-US joint nuclear research adopted in FY 2016
Theme:- Joint research on environmental measures including
measures for the management of radioactive waste
・Improvement of the accident analysis codes (US)Cooperation withIRID
・Development of technologies for prevention of leakage
overseas(France, US)organizations at・Developmentof technologies for remotethe stage of IRID’s
decontamination(UK, US)practical

7-4

application
research

Technical
development withoverseasorganizations by
TEPCO

・Development of internal investigation technologies (UK,
Canada, US, France, Germany, Russia)
・Internal fuel debris detection and characterization (US,France, Kazakhstan)・Fuel debris, structural component retrieval technology
development (US, UK, France, Germany)
・ Criticality control technology development (Hungary,
Russia, France)
・ Fuel debris containment, transportation, storage
technology development (France, US)
・ Solid waste processing and disposal technology
development (US)・Geographic Information System: GIS (UK)・Fuel handling facilities for removal from the fuel pool in
Unit 3 (US)TEPCO

Column: International Special Advisors/International Expert Group/International Advisors
In order to smoothly and promptly proceed with the very complex and difficult project ofdecommissioning multiple reactors that caused severe accidents, it is necessary to integrate andutilize a wide range of wisdom. To do so, NDF invites the experts in strategy development, R&D,
program/project management and safety regulation as the International Special Advisors (Mike
Weightman, former ONR Director; Paul Dickman, Senior Policy Fellow at Argonne National Laboratory,
US; François Gauchè, Director of Nuclear Power Development, CEA, France; and Juan José Zaballa,
Director of ENRESA, Spain).
TEPCO also invites the specialists who have expertise and experience as the International Expert
Group (Sam Armijo, former Chairman of the Advisory Committee on Reactor Safeguards, the US
Nuclear Regulatory Commission, US; Rosa Yang, Fellow, Electric Power Research Institute (EPRI); Dr.
Adrian Simper, NDA Director, UK; Joel Pijselman, President, Enrichment Technology Company (ETC),France; and Nicolai Steinberg, independent consultant in Ukraine) in order to obtain advice fromthem to support safe and effective decommissioning and R&D of the Fukushima Daiichi NPS.
IRID also invites distinguished experts as the International Advisors (Luis Echávarri, former Secretary
General of OECD/NEA; Melanie Brownridge, Head of Research and Development, NDA; and Lake H.
Barrett, former TMI Local Representative Director, NRC) for evaluation on the status of design review
of ongoing R&D and the enhancement of information dissemination and communication, as well as
to obtain knowledge including past failures.

International Special Advisors

International Expert Group

7-5

7.2 Active information dissemination to international communities(1) Purpose
In order to carry out the responsibility to international community and proceed withdecommissioning by integrating and utilizing wisdom from all over the world, domestic relevantorganizations should actively disseminate information to international community.

(2) Current efforts
In order to introduce our efforts on decommissioning of the Fukushima Daiichi NPS, NDF has been
participating in a variety of events in close collaboration with the government, holding side eventsat the IAEA General Conference, participating in WM Symposium, etc. Additionally, information on
the decommissioning of the Fukushima Daiichi NPS is presented in the OECD/NEA activities as
described in Table 7-2 above. NDF has also decided to organize the International Forum on theDecommissioning of the Fukushima Daiichi Nuclear Power Station as a way to build trusting
relationship with the local communities and as a means of disseminating the latest information on
decommissioning to the world. The 1st forum was held in April 2016 and the 2nd in July 2017 withthe cooperation of the government and other relevant organizations (www.ndf-forum.com) (column:
the International Forum on the Decommissioning of the Fukushima Daiichi Nuclear Power Station).
In addition, the Technical Strategic Plan and the decommissioning R&D information portal site(http://www.drd-portal.jp/) are created and released in English and Japanese.

(3) Efforts in the near future
It is important to carry out decommissioning in a manner open to the international community,
informing the data obtained through decommissioning and countermeasure to manage
contaminated water, as well as the site situation, quickly and accurately to be understood easily, inorder to fulfill our responsibility as the country that had the accident. Furthermore, it is also importantto actively disseminate information on the progress and accomplishment of decommissioning
through the side event at IAEA General Conference, OECD/NEA programs, and international
symposia, etc. to get advice and evaluation. At the same time, since it is necessary to attract theattention from companies and education/research institutes that will be able to leaddecommissioning, it is also important to enhance the structure to integrate and utilize the wisdom
such as TEPCO CUUSOO (https://tepco.cuusoo.com) which disseminates the issues in theprocess of decommissioning to the people inside and outside of Japan.
When disseminating information, it is necessary to make effort to enable international community
to understand accurately and prevent reputational damage. Based on the viewpoint of stakeholder
involvement described in Section 7.1, in order to inform the complicated situation of Fukushima to
local residents in an easy-to-understand manner, NDF has been making efforts such as holding
the International Forum on the Decommissioning of the Fukushima Daiichi Nuclear Power Station
with the support of the communication specialists. Domestic relevant organizations should also
promote dissemination of information that is easy for people to understand.

7.3 Close cooperation between relevant organizations
The domestic relevant organizations such as the government, NDF, IRID, JAEA and TEPCO havebeen engaging in each of their international collaboration activities according to their roles as stated

7-6

in Chapter 1. However, these organizations should further enhance close cooperation for theeffective acquisition and use of information. In this respect, it is important to make efforts to present
and share information required by each organization.

Column: The International Forum on the Decommissioning of the Fukushima Daiichi Nuclear
Power Station
In April 2016, as a function of sharing technical knowledge on decommissioning and as a place ofdeveloping common understanding with domestic and international relevant people, NDF organized
the 1st International Forum on the Decommissioning of the Fukushima Daiichi Nuclear Power
Station in Iwaki city, Fukushima prefecture, jointly with the METI Agency for Natural Resources andEnergy. It was attended by 641 people from 15 countries, and decommissioning under the themesof “local communication” and “top-level technical research for decommissioning” was discussed.
It has become clear that the efforts such as fuel debris retrieval and waste management related todecommissioning of the Fukushima Daiichi Nuclear Power Station are important subjects, not onlyin the technological aspects but also in the social aspects such as communication with the local
community, which Japan has to overcome by integrating wisdom from all over the world. A
decommissioning activity is an intergenerational project. In order to promote this, it is not sufficient
to merely disseminate the information on the decommissioning unilaterally. It is also important tocreate opportunities for local residents to find such information in the community and to transmit
information to them in an easy-to-understand manner. BY doing so, understanding will be sharedwidely through interactive communication.
Additionally, the 2nd International Forum on the Decommissioning of the Fukushima Daiichi
Nuclear Power Station was held by NDF. The session intended mainly for the local community washeld in Hirono-town and the dialogue sincerely to face their questions and anxiety was conducted.
This is the first international decommissioning forum held in a city that suffered the nuclear power
accident.

Representatives of local community

Representatives of experts

7-7

8. Future actions
The Fukushima Daiichi NPS decommissioning program is to enter the next phase based on thetwo strategic proposals provided in the Strategic Plan 2017. The significance of the project
management capabilities for an appropriate and effective decommissioning has been understoodand recognized and TEPCO has been trying to further improve the situation. It is getting more andmore important as we enter a more technically difficult phase including fuel debris retrieval.
As discussed in the Chapter 1 above, NDF will be further involved in the project management,
based on the revised NDF Act, as an administrator/supervisor of the decommissioning and takemore roles/responsibilities such as (1) the appropriate management of funds for decommissioning,(2) the supervision of an appropriate system for executing the decommissioning, and (3) steadywork management based on the decommissioning reserve fund system.
Earning trust from the local communities is a key factor for the steady progress of the project, too.
The government, NDF, TEPCO and other stakeholders need to maintain sincere and detailed
communication, sharing their roles appropriately, to eliminate fear and worry of the local residents
and build positive relationships with them again.

8.1 Enhancing project management capabilities
For Fukushima Daiichi NPS decommissioning, a variety of projects such as contaminated water
management and fuel debris retrieval from spent fuel pool are running in parallel and they areaffecting each other. We are now to focus on more technically difficult projects including fuel debris
retrieval that require us to carefully consider the relations and consecutiveness with the preceding
projects. It is essential to manage this large scale, multi layered and very complicated project
consisting of many smaller projects as one integrated project based on the correlation among thosesmaller ones.
Furthermore, it is also important to manage R&Ds, HRD, supply chain, regulations, and relationshipwith local communities that are all indispensable parts of the project in an integrated manner.
Since the decommissioning project is an unprecedented challenge in the whole world and anecessary condition for the local recovery, it has great significance. Managing the project in an
effective and integrated manner holds the key to the success as the project is very complicatedand difficult to carry out.
To promote the decommissioning project steadily in the decommissioning of Fukushima Daiichi
NPS with a high-level of uncertainty, it is especially critical to identify the risks affecting progress ofthe decommissioning project, analyze and prioritize them, and take required actions to
avoid/reduce them as a part of risk management procedure.
For Fukushima Daiichi NPS decommissioning, a variety of smaller projects are running in parallel
and they are affecting each other. Moreover, each smaller project is composed of a variety ofactivities. It is very important to always consider progress and current situation of every single
activity in addition to the original planning and assumptions when executing the projects with great
uncertainty. However, if there is a huge gap between the information/assumptions collected uponplanning and the information gained through the actual activities, the activities may not proceed as
expected. It is vital to consider project risks of such kind all the time. The delays caused by theappearance of risks may lead to more delays in the following activities and then in the entire project.
Furthermore, for Fukushima Daiichi NPS decommissioning in which a variety of smaller projects
are running in parallel, balancing between those activities is a serious challenge. It is essential to8-1

better deal with such project risks as affect stability and continuity of the entire decommissioning
project.
In the US/UK decommissioning cases, various measures against those risks have been discussed
and adopted. In addition, varieties of efforts have been carried out to assess the impact on theschedule of the entire project and the costs with potential appearance of different risks in mind. For
Fukushima Daiichi NPS decommissioning project that includes great uncertainty, it is absolutely
meaningful to make efforts to deal with project risks while studying the past cases from theviewpoints of appropriate progress/funds management.
The project should rebuild public trust, forming the basis for the successful project. NDF andTEPCO, as the administrator/supervisor and the implementer of the decommissioning project
respectively, need to clarify the roles/responsibilities and accountability to improve the governanceof the entire project in order to maintain transparency in the activities and fund management andachieve stable progress through the project; we both need to remember that it is absolutely
imperative to win uninterrupted public trust and that we have mandate to use the reserve funds
appropriately.

8.2 Engagement with stakeholders
The domestic/overseas professionals and organizations that have experienced decommissioning
insist that good communication with local communities is indispensable to steady and successful
decommissioning. The government, NDF, TEPCO and all other parties should fully understandeach other’s position and consider their overseas experiences to appropriately share theroles/responsibilities. Even a minor safety trouble may increase the fear/concern of the local
residents, lead to decreased motivation for restoring/returning to home, increase reputational risks
and affect the local economy. To avoid these, it is absolutely critical to invest a great deal of effort
to avoid any kind of trouble with extreme caution, firstly. And if any troubles occur, it is vital toprovide sincere and detailed explanations. Providing accurate and easy-to-understand informationis crucial to achieve this. It is important to keep providing information proactively about
implementation of countermeasures and improvement in safety management not only right after
the trouble happens but also onwards. It is also essential to put more effort in improving quality of
information delivery factoring in the past case in which information provision was not performed ina fully appropriate/careful manner when the January and February 2017 investigation into the insideof the Unit 2 discovered high dose radiation.
Furthermore, it is required to share the concepts of the risk reduction policies with the local
residents in each and every phase of decommissioning. It is also important to have common
understanding among the relevant parties that the risks must be prioritized and classified as “risks
to be removed as soon as practicable” and “risks to be dealt with in a careful manner” based on
the risk reduction strategy.
Just sharing information between the providers and receivers is not enough to establish goodcommunication. It is necessary for both sides to make good conversations and mutual effort to
reduce the gap between them repeatedly. The government holds the “Fukushima Advisory Board
on Decommissioning and Contaminated Water Management” as a forum for providing information
to local stakeholders, strengthening communication, and discussing how the public relations
activities should be. In addition to this effort, the government creates PR videos and brochures"Important Stories on Decommissioning" to convey accurate information to be understood easily.
At the present stage, NDF has undertaken activities and programs 1) to listen to the locals, provide

8-2them with easy-to-understand explanation of Fukushima Daiichi decommissioning, and hold “TheInternational Forum on the Decommissioning of the Fukushima Daiichi NPS” constantly in order toshare the current status and technical achievements of the decommissioning widely among
national/international professionals, 2) to directly explain to/make conversation with the locals, and
3) to successfully continue offering explanation to local government officials who have frequentcontact with the locals. In addition, we have been discussing specific ways to make proactive
communication with various stakeholders. TEPCO’s management and risk communicators haveconstantly been explaining/talking to local residents at “Prefectural Safety Assurance Conference”
organized by Fukushima prefecture. TEPCO also has been providing explanation for the progressof the Mid-and-Long-Term Roadmap and making conversation with neighborhood associations’leaders. Furthermore, TEPCO has been preparing to accept the investigators/visitors including the
locals who have been evacuating into Fukushima Daiichi NPS.
Just anxiety about potential risks may cause serious harm to the reputation. Even worse, it is often
pointed out that only the significance of the image just after the accident will be highlighted even if
around six years have already passed and it remains to be regarded as risk with the frequency
ignored, typically.
The delay in taking appropriate actions and increase of the occupational exposure of the workers
and cost through the approaches to existing reputational damages and reducing the risks posedby radioactive materials may harm public image/reputation of the decommissioning. And theconsequences may cause further delay in implementing measures and trigger a downward spiral.
To avoid further reputational damages, it is more significant, than anything else, to manage
radioactive materials in an appropriate manner in order not to let them leak and reduce the existingrisks promptly. It is also crucial to provide accurate and precise information not only to the local
residents, the press, market players and distributors but to domestic/overseas consumers as well.

8.3 Consideration for project sustainability
It is the lifeline for Fukushima Daiichi decommissioning to ensure the project sustainability for longterm period, so it is important to succeed the technical knowledge/information and maintain thehuman capability/motivation. Therefore, it is necessary to ensure the mechanism that enables
continuous project management, R&D, engineering, and site management, etc. and to secure alarge variety of human resources taking the roles in these works. In particular, it is desirable tomanage the knowledge and experiences, establish the database, archive the data, and build thesystem to utilize and succeed them, as well as to build an environment which enables people
involved in the project to develop their careers with pride, motivation, and the sense of reassurance.
We hope that added value will be created by implementing the decommissioning project based on
the mechanisms and the improved environment, and then industry, academia, local communities
and global society will have beneficial spin-off effects. We also hope that great talents with strong
leadership for contributions to society, problem solving and innovation, not only in the area of
decommissioning but in different fields also, will be developed based on the experiences gained
through the project and the pride. It is desirable that the spin-off effects will bring broader expertise
into the project, attract workers/experts in other areas and keep the project sustainable with a wider
range of human resources. Furthermore, it is expected that the decommissioning project will be
implemented in a stable manner without interruption whilst creating added value.

8-3

＜Appendix＞
Appendix 1: Japanese Government's Organizational Framework for Decommissioning the
Fukushima Daiichi NPSAppendix 2: Information on the Strategic Plan
Appendix 3: Overview of SED indicator
Appendix 4.1: Periodic measurement of plant data
Appendix 4.2: Results of PCV internal survey
Appendix 4.3: Survey results of S/C and torus room
Appendix 4.4: Results of muon detection
Appendix 4.5: BSAF project outline and results so far
Appendix 4.6: Characteristics of MAAP code and SAMPSON code and analysis results
Appendix 4.7: Summary of heat balance method and estimation result
Appendix 4.8: Location of fuel debris estimated from plant parameter trend
Appendix 4.9: List of properties of fuel debris
Appendix 4.10: Status of the comprehensive analysis and evaluation of the conditions inside
the reactor in each unit
Appendix 4.11: Studies on containment function
Appendix 4.12: Study of water level at the bottom of PCV when retrieving fuel debris
Appendix 4.13: Preliminary Seismic Margin Evaluation of RPV and PCV.
Appendix 4.14: Development of fuel debris retrieval equipment and devices
Appendix 4.15: Risks accompanied by fuel debris retrieval tasks
Appendix 5: Status of Solid Wastes Management, and Their Storage Management Plan
Appendix 6.1: Interim Report by Research Partnership Task ForceAppendix 6.2: Changes in elective subjects of secondary examination for professional engineers(category of atomic energy and radiation)
Appendix 7: Conditions at TMI-2 and Chernobyl Unit 4

Appendix 1: Japanese Government's Organizational Framework for Decommissioning
the Fukushima Daiichi NPS
In response to the accident at the TEPCO’s Fukushima Daiichi NPS on March 11, 2011, theJapanese government established a Nuclear Emergency Response Headquarters according to the
Act on Special Measures concerning Nuclear Emergency Preparedness (Act #156 of 1999). The
headquarters was aimed to press ahead with emergency measures to address the nuclear
emergency caused by the accident.
The government established a Council for the Decommissioning of Fukushima Daiichi NPS so that
under the leadership of the Nuclear Emergency Response Headquarters, it would focus all its
energies on fundamentally solving the issues of contaminated water and the decommissioning of
the Fukushima Daiichi NPS, without leaving these issues to TEPCO. The council is deliberating
and deciding on important matters associated with the Mid- and Long-Term Roadmap to thedecommissioning of the power station.
The framework of the Japanese government to decommission the Fukushima Daiichi NPS is shown
below.

Fig. A1-1: Organizational Framework of the Japanese Government Regarding theDecommissioning of the Fukushima Daiichi NPS

Fig. A1-2: Regulatory Framework Relating to the Fukushima Daiichi NPS

A-1

Appendix 2: Information on the Strategic Plan
(1) History of the Mid- to Long-Term Roadmap
The mid- to long-term plan originates with the "Results of Reviews Regarding Mid- to Long-Term
Actions for the Tokyo Electric Power Co., Inc. Fukushima Daiichi NPS" report (dated Dec. 7, 2011)
prepared by the Advisory Committee for Formulating Mid- and Long-term Strategies to Clean up
the Fukushima Daiichi NPP of TEPCO, which was installed at the Atomic Energy Commission ofJapan; the first such report to be presented to the government, TEPCO, and other organizations.
Later, on Dec. 21, 2011, the Government and TEPCO’s Mid-to-Long Term Countermeasure
Meeting formed under the Nuclear Emergency Response Headquarters finalized the first edition of
the Mid- to Long-Term Roadmap, which has undergone three revisions up to the current time (Table
A2-1).
The Mid- to Long-Term Roadmap is being revised on an ongoing basis to reflect field conditions,R&D outcomes, and other factors.
Table A2-1: History of the Mid- to Long-Term Roadmap
Atomic Energy
Commission
This was a report issued by the Advisory Committee for Formulating
Advisory Committee
Mid- and Long-term Strategies to Clean up the Fukushima Daiichi NPP
for Formulating Midand Long-termof TEPCO formed under the Atomic Energy Commission. This was theStrategies to Clean up first report to review and report on the mid- to long-term plan for thethe Fukushima Daiichi Fukushima Daiichi NPS.
NPP of TEPCO report
dated Dec. 7, 2011
The Government and TEPCO’s Mid-to-Long Term Countermeasure
Meeting concluded the Mid- to Long-Term Roadmap (first edition) on
Dec. 21, 2011 based on the fulfillment of the Step 2 target (See note) inthe "Accident Response Roadmap for the Tokyo Electric Power Co., Inc.
Mid- to Long-Term
Fukushima Daiichi NPS -- A Roadmap for Approaches at this Time." This
Roadmap dated Dec.
roadmap was compiled by TEPCO, the Agency for Natural Resources
21, 2011
and Energy, and the then Nuclear and Industrial Safety Agency, based(First edition)on the Atomic Energy Commission, Advisory Committee for Formulating
Mid- and Long-term Strategies to Clean up the Fukushima Daiichi NPPof TEPCO report.
(Note) The Step 2 target reads "The release of radioactive materials isunder control, and radiation dosages are being limited significantly."
After the completion of Step 2, the first revision of the Mid- to Long-Term
Mid- to Long-Term
Roadmap was created on Jul. 30, 2012 reflecting the specific plan
Roadmap dated Jul.
prepared by TEPCO regarding priority matters that must be addressed30, 2012
to improve mid- to long-term reliability and the progress of efforts made(First revision)
up to that point.
The Council for the Decommissioning of TEPCO’s Fukushima Daiichi
was installed at the Nuclear Emergency Response Headquarters. It wasMid- to Long-Termdecided that field operations and R&D progress management would be
Roadmap dated Jun.
implemented in a unified manner with the involvement of other relevant
27, 2013
agencies in addition to the government and TEPCO. The Mid- to Long(Second revision)
Term Roadmap dated Jun. 27, 2013 (second edition) was created by thisCouncil.
This revision defined that ongoing risk reduction actions will be taken toMid- to Long-Termprotect persons and the environment from the risks that arise from
Roadmap dated Jun.
radioactive materials. It was also decided that optimum
12, 2015
countermeasures would be implemented based on the categorization(Third revision)and prioritization of risks. The Mid- to Long-Term Roadmap dated Jun.
12, 2015 (third edition) was created by this Council.

A-2

(2) What is the Strategic Plan?
As described in Chapter 2, the Strategic Plan developed by the NDF focuses on considerations
from the technical viewpoint according to the role of NDF, namely technical support. It is a total planthat covers not only on-site operations but also necessary research and development activities andtechnical considerations associated with site work.
The plan provides not only a “strategy” but also concrete policies and requirements necessary to
carry forward the strategy, review on available resources, indicating the plans on the on-site work
and R&D activities required.

Fig. A2-1: "Strategy," "Policies," "Plans" and the Strategic Plan
(3) Reviews of the Strategic Plan to date
The Strategic Plan 2015, which was announced in April 2015, declared the "Ongoing and swift
reduction of risks resulting from radioactive material at the Fukushima Daiichi NPS" as its basic
policy. The five guiding principles that would be key to achieving this policy (safe, proven, efficient,
timely, and field-oriented) were also presented.
In order to reduce risks in a steady manner, a broad range of radioactive materials (sources of risk)
were identified, and then grouped into three categories and prioritized by referencing the SED scoredeveloped by the British Nuclear Decommissioning Authority (NDA). Of these, countermeasures
were already underway for sources of risks which needed to be addressed as expeditiously as
possible, such as contaminated water, and therefore, it was decided that the Strategic Plan would
review the retrieval of fuel debris, a risk source that would require thorough preparations and posenumerous challenges, as well as countermeasures for waste materials which are a risk source thatwould require long-term action.
Since reviews on fuel debris retrieval would extend to a diverse range of areas including analyzing
conditions in the PCV, ensuring safety in the fuel debris retrieval process, and what methods to use
for fuel debris retrieval, the structure of all necessary requirements were compiled into a logic tree
to provide an overall view of the situation.
In the area of fuel debris retrieval, methods of fuel debris retrieval to be given priority review wereselected from among available methods, and then reviews were conducted for the status of actions
currently underway with regard to the technological requirements of various submersion and partial
submersion methods that were outlined from the logic tree, as well as reviews for avenues of action
moving forward. Internal PCV condition analyses were then carried out based on results frominvestigations of the PCV interior, analyses results, and parameters of the actual unit. Additionally,
reviews were carried out on the feasibility of the technological requirements for fuel debris retrieval,
as well as policies for retrieval.

A-3

In the area of waste management, given that it is critical to review the safe and stable storage andmanagement of solid wastes that were generated in the accident, as well as processing
methodologies and disposal concepts based on a mid- to long-term vision, steps were taken tooutline the basic, internationally outlined ideas that underly the ensuring of safety in radioactive
waste disposal in general, as well as the modalities of processing that must be kept in mind.
Additionally, the current status of actions being taken based on the current Mid- a Long-Term
Roadmap were outlined, and reviews were carried out on the policies of actions and future actions
based on a mid- to long-term perspective.
In order to move forward with the R&D required in the two areas stated above, an overall plan for
R&D that would need to be pursued was presented, and steps were taken to strengthen
management operations for improving R&D efforts and the effectiveness of these efforts.
Additionally, steps were taken through the Decommissioning R&D Partnership Council in the areaof training human resources, as well as strengthening collaboration with universities and researchorganizations who engage in basic and generic research, IRID which conducts practicaldevelopment, and TEPCO who will be conducting the decommissioning.

A-4

Appendix 3: Overview of SED indicator
The SED indicator referred in the risk analysis of the strategic plan is explained hereinafter. NDA
uses this indicator as one of the indicators for risk reduction prioritization among the diverse
facilities at 17 sites owned by NDA75.
The SED indicator is expressed by the following formula. The first term of the formula is referred to“Hazard Potential” of the risk sources and the second term is referred to “Safety Management”.
Each factor of the formula is explained in the following. CHP in the formula below is the hazard
potential of chemical substance, that is not applied to the risk assessment in the strategic plan.

SED  RHP  CHP   FD  WUD 

4Radiological Hazard Potential (RHP) is an indicator expressing the potential impact of radioactive
material and expressed by the following formula. RHP is defined as the potential impact of radiation
exposure to the public in case of total release of the radioactive materials contained in the risk
source.

RHP Inventory  Form Factor
Control Factor

Inventory is defined by the product of radioactivity of the risk source and the Specific Toxic Potential(STP)76 as shown in the following formula.

 Inventory m 3  Radioactivity TBq  Specific Toxic Potential m 3 TBq

Specific Toxic Potential (STP) is, defined as the volume of dilution water of 1TBq of radioactive
materials, that will result in an exposure dose of 1mSv by ingesting a fixed amount of this diluted
water in a period of one year, as shown in the formula below.
⁄

⁄

⁄

⁄
0.001
The Inventory is therefore defined as a basic volume representing the extent of impact ofradioactive material to the public. The SED indicator conservatively uses the larger dose coefficientof inhalation or ingestion.
FF: Form Factor, as shown in Table A3-1, is a factor that describes the rate of radioactive materials
is actually released depending on the difference in property, i.e. gaseous, liquid, or solid state etc.
Based on the experimental measurement data, the factor is set that, in case of total loss ofcontainment function, 100% of gaseous and liquid materials would be released, while 10% of
powder materials was supposed to be released. Solid materials had no clear basis but was
proposed as a sufficiently small value to show the unlikelihood of being released.
CF: Control Factor, as shown in Table A3-2, is a factor describing time allowance available, until
recovery of safety functions ensuring the current stable state of the risk sources, or until any risk is
actualized when safety functions required ensuring the current stable state are lost. It considers

75

NDA Prioritization – Calculation of Safety and Environmental Detriment score, EPGR02 Rev.6, April
2011.
76
Instruction for the calculation of the Radiological Hazard Potential, EGPR02-WI01 Rev.3, March 2010.

A-5

heat, corrosive, flammable, hydrogen production etc., reaction with water or air, and criticality etc.
as reactivity specific to a risk source.
FD: Facility Descriptor is a factor that describes sufficiency of containment function of a facility. It
categorizes risk sources in terms of the combination of elements such as integrity of facility,
multiplicity of containment function, and safety response state etc. of each facility. As shown in
Table A3-3, the state of each facility is classified into 10 categories, and a Score is given to eachcategory.
WD: Waste Uncertainty Descriptor is a factor that shows whether or not any impact is generated
when there is a delay in retrieving the risk source. Risk sources are categorized in terms of the
combination of degradation, reactivity, package state and monitoring state etc. of the risk source.
As shown in Table A3-4, the state of each risk source is classified into 10 categories, and a Scoreis given to each category.
Table A3-1

FF (Form Factor)

Property

Table A3-2
Score

Class

CF (Control Factor)
Time scale

Score

Gas, liquid

1

Hour

1 hour1

Sludge, powder

0.1

Day

24 hours

10

0.00001

Week

168 hours

100

0.000001Month

730 hours

1,000

Discrete solids
Largemonoliqueandactivatedcomponents

Table A3-3
Class
1
2
3
4
5
6
7
8
9
10

Year

8,760 hours

10,000Decade87,600 hours

100,000

FD (Facility Descriptor)

Definition
Building past its original design life, single containment, significant defect, limited contingency provisions,
Building is not qualified to withstand modern design standards.
Same as Class 1. However, with no significant defect.
Same as Class 2. However, with sufficient contingency provisions.
Same as Class 3. However, with double containment.
Same as Class 4. However, while it hasn’t exceeded the design life, it will exceed the design life at the timeof retrieval of the risk source.
Same as Class 5. However, it won’t exceed the design life at the time of retrieval of the risk source.
Same as Class 6. However, while it meets the modern design standards, execution of the safety case islimited in scope.
Same as Class 7. However, while the safety case is thoroughly executed, it is influenced by the adjacentbuilding.
Same as Class 8. However, it influences the adjacent building.
Same as Class 9. However, it is not influenced by the adjacent building, nor is the adjacent building
influenced by it.

Table A3-4

Score
100
91
74
52
29
15
8
5
3
2

WUD (Waste Uncertainty Descriptor)

Class
1
2

Definition
Score
Degrading* unpacked waste. It is not monitored or controlled.
100
Same as Class 1. However, it is packed.
90
Reactive** unpacked waste. Its existence, quantity and location is unknown, and it can’t practically be
374
confirmed.
4Same as Class 3. However, it can be confirmed through sampling etc.
50
5Same as Class 1. However, it is monitored or controlled.
30
6Same as Class 2. However, it is monitored or controlled.
17
7
It is not active, but deteriorating unpacked waste. It is not monitored or controlled.
98Same as Class 7. However, it is packed.
5
9
It is neither active nor deteriorating unpacked waste. It is monitored or controlled.
3
10
Same as Class 9. However, it is packed.
2*Degrading: A property that is likely to cause possibility of change in removal method, increase in exposure dose, or
occurrence of criticality etc. through dissociation or dispersion in the future.
**Reactivity: A property causing an abrupt change such as generation of heat and explosion etc.

A-6

Table A3-5 shows the overview of the containment function, safety equipment, andcontrol/monitoring state etc. of each risk source that were used when evaluating Safety
Management. On the basis of these, risk sources are categorized into 10 categories using relative
comparison, and set the Score for modified FD and modified WUD.
The uncertainty is defined to have a range corresponding to the difference in the Scores of adjacentcategory. The score shows a linear relation on a logarithmic scale, then the uncertainty isapproximately constant without regard to the category. As regards rubble etc. (placed outdoors),
the uncertainty is defined depending on consideration of various storage condition.
Table A3-5 Characteristics of each risk source concerning Safety Management.
Risk source
Fuel debris

Spent fuelContaminatedwater

Secondary wastefrom water
treatment
systems

Solid waste

Characteristics
No significant damage has been observed in PCV. Redundancy of criticality
control, cooling, hydrogen explosion prevention systems is secured. In
addition, important parameters such as Xe concentration, temperature, andhydrogen concentration etc. are monitored.
The spent fuel pool of each Unit is designed to maintain subcriticality, andredundancy of the cooling system is secured. Falling of rubble and heavy
objects defect of ceiling of the building, and seawater injection etc. have been
experienced in some Units
Common pool and dry casks as well as their buildings are not affected by theearthquake and the tsunami.
Containment function is maintained by management of water level differenceof the contaminated water in the building and the ground water.
Concentrated liquid waste is liquid waste generated by concentrating
concentrated saltwater through the evaporation-enrichment system, and has
high concentration of radioactive material and salinity. It is stored in weldedtype tanks and the tank is located in the weir.
The waste adsorption column is a shielded carbon-steel vessel storing Csadsorbent zeolite, which is stored again in a shielded vessel, and placed inbox culvert or on the frame. It doesn’t need any control such as decay heatremoval etc.
Waste sludge is stored in the storage of pit structure tank for granulated solid
waste located in the main process building, and leak monitoring, decay heatremoval and hydrogen exhaust operations are carried out.
HIC slurry is stored in polyethylene containers, which are further stored in
SUS reinforced vessels and placed in box culverts. As HIC slurry contains
water, hydrogen is released to the atmosphere through the vent. Decay heatremoval is not needed, but monitoring is continued because an event
occurred where generated hydrogen caused water to drip.
Rubble (in the storage facility) is high dose rubble stored in containers andare placed in the storage facility, requiring no special control.
Rubble etc. (placed outdoors) includes wastes of various dose level which
are stored outdoors in various forms, requiring monitoring.

A-7

Appendix 4.1: Periodic measurement of plant data
From plant data on the dose rate temperature, hydrogen concentration, PCV pressure, radioactive
materials concentration, and other items inside the PCVs that have been continuously obtainedsince the accident occurrence, it is estimated that Units 1-3 in the plant are kept in a stable cold
shutdown condition.

(1) Dose rate
Figure A4.1-1 shows the variations in the dose rate summarized based on information publishedby the TEPCO. After the accident, the radiation dose rate has gradually decreased over time due
to decay of radioactive materials or leaching to contaminated water, but it is still over a few
hundred mSv/h in the D/W, which is still high to carry out work. Although the rate in the torus
room is lower than that in the D/W at around several tens mSv/h, the condition is not suitable fordoing long tasks. It can be assumed that the radiation dose rate will further decrease with thepassage of time.
The Containment Atmospheric Monitoring System (hereafter referred to as "CAMS") (A System)in the S/C of Unit 3 shows a change in trend since October 2015 towards a decrease in theradiation dose rate. This is believed to be because the number of water feed pumps forcontaminated water were increased in each unit during this period. The CAMS in the torus room
is exposed to radiation from the radioactive materials attached to the wall surface/ components
and radiation from the stagnant water in the S/C and the torus room. Effect of the stagnant waterin the torus room is considered to have decreased since a decline in the dose rate has causedby change in water feeding. In Unit 2, on the other hand, no change was seen in the CAMS valueeven though the water feed pumps for contaminated water were similarly increased during thesame period. Therefore, it appears that the effect from the radioactive materials attached to thewall surface/ components and the contaminated water in the S/C is larger than that of thecontaminated water in the torus room.

(2) Temperature
Figure A4. 1-2 shows the variations in the reactor's ambient temperature. Temperature in the
PCV reduced from what it was immediately after the accident and reached below 100°C in six
months after the accident. Since then, it has exhibited a gradual declining trend each year
following the seasonal fluctuations in air temperature and water temperature. Figure A4. 1-3
shows the heat (decay heat) from the elements that constituted the fuel assembly loaded in theaccident. Immediately after reactor shutdown, decay of nuclides with short half-life is high, but
because they change over to stable nuclides through the process of repeated decay, when 5
years have passed since the accident, the amount of heat generated drops to 1/1000th of theamount at the time of the accident. Nuclides with long half-life remain inside the PCV, and since

A-8their decay progresses slowly, the decay heat is also expected to decrease gradually. The
temperature can be expected to decrease further over time.

(3) Hydrogen concentration and PCV pressure
Irradiating water with gamma rays causes hydrogen to be generated by radiolysis. The PCVs are
injected with water to cool the fuel debris inside them. In addition, the doses inside them are high.
With these factors, there is a fear that hydrogen may be generated inside them. Based on thefact that hydrogen has a lower explosive limit of 4%, hydrogen explosion is prevented byintroducing nitrogen to dilute the hydrogen in PCV after the accident. Figure A4. 1-4 and figure
A4. 1-5 show the changes in hydrogen concentration in PCV and changes in PCV pressure
respectively. The hydrogen concentrations are low enough, indicating that hydrogen has beeneffectively diluted by nitrogen injection. From the viewpoint of confining FPs, it may be effective
to discharge hydrogen out of the PCVs. However, doing so may cause the pressure inside the
PCVs to be lower (negative pressure) than the atmospheric pressure. This may allow air that
includes oxygen to enter the PCVs through, for example, sealed sections of them, resulting in
mixture of hydrogen and oxygen. For this reason, the pressure inside the PCVs is kept slightly
higher (slightly positive pressure) than the atmospheric pressure.

(4) Concentration of radioactive materials
Figure A4. 1-6 and figure A4. 1-7 show the variations of Xe1-35 concentration and Cs-137 andCs-134 concentration in the PCV, respectively. In each unit, Xe-135 has been detected in the
PCV Gas Management System, but its presence was sporadic and in trace amounts in eithercase and it is considered to have been generated mainly by spontaneous fission of Cm-242, Cm244 etc. Concentrations of Cs-137 and Cs-134 have been declining with time. The half-life of Cs137 is approx.. 30 years and that of Cs1-34 and approx.. 2 years. Over time, Cs-134, which has
a shorter half-life, is getting detected less frequently than Cs-137. It can be assumed that theconcentration and detection frequency will further reduce in the future.

(5) Reduction in the amount of water injection and temperature of each part
From December 2016 to March 2017, an operation was carried out to reduce the amount of water
injections from 4.5m3/h to 3.0m3/h. Temperatures of the main parts of each unit at that time andthe amount of water injection are shown in figure A4. 1-8. The amount of water injection was
reduced three times in each unit, but no noticeable rise in the temperature was observed. Further,
it is seen that a general change in the temperature of each part tends to depend on thetemperature of the water injection.

With these facts, it is estimated that Units 1-3 are kept in a stable cold shutdown condition.

A-9

Figure A4. 1-1

History of radiation dose rate around the reactor in the Fukushima Daiichi NPS(Prepared based on data published by TEPCO)

A-10

Figure A4. 1-2

Changes in the ambient temperature of the reactor in the Fukushima Daiichi
NPS(Prepared based on data published by TEPCO)

A-11

Figure A4.1-3

Heat from the fuel, FPs, and irradiated materials inside the Reactors(Prepared based on data published by JAEA-Data/ Code 2012-018)

Figure A4. 1-4

Changes in the hydrogen concentration inside the PCV(Prepared based on data published by TEPCO)

A-12

Figure A4.1-5

Changes in the pressure inside the PCVs

(Prepared based on data published by the Meteorological Agency and TEPCO)

Figure A4.1-6

Changes in Xe-135 concentration in the PCVs

(Prepared based on data published by TEPCO)

A-13

Figure A4. 1-7

Changes in concentrations of Cs-137 and Cs-134 in the PCVs
(Prepared based on data published by TEPCO)

A-14

Figure A4. 1- 8

Changes in the temperature during decline of amount of water injection(Prepared based on data published by TEPCO)

A-15

Appendix 4.2: Results of PCV internal survey
Status of the PCV internal survey so far is summarized below.

(1) PCV internal survey of Unit 1
A. Conducted in October 2012 and April 2015
1) Purpose: Obtaining information on "steel grating of 1st floor in PCV".
2) Method: The survey device was inserted through PCV penetration (X-100B penetration).
PCV’s internal survey by a CCD camera and stagnant water sampling were performed in
October 2012, whereas pedestal outer survey (B1 survey) was done in April 2015 using ashape-changing robot.
3) Information obtained:a. No major damage to the existing facilities (PLR pump, wall surface in PVC, HVH etc.)
was observed.
b. Radiation dose rate on the steel grating was approx. 5 - 10 Sv/h.
c. It was observed that PLR piping shielding units (lead wool mat) have fallen. It can thus
be estimated that the temperature on the steel grating of 1st floor might have exceeded
328°C, which is the melting point of lead.
d. Access route to the bottom of D/W was investigated, but sediments are widely distributedat the bottom of D/W. Fuel debris was not observed in this survey.

Figure A4. 2-1

Results of PCV internal survey (B1 inspection) (Unit 1)[Reference: TEPCO "Development of primary containment vessel internal survey technology”
pedestal survey on 1st floor steel grating (B1 survey) on-site demonstration testing results"]

A-16

B. Conducted in March 20171) Purpose: Observation of the spread of fuel debris from the bottom of the PCV to outsidepedestal as indicated in the analysis and check whether fuel debris has reached PCV shell.
2) Method: In March 2017, a remote survey device was inserted through PCV penetration (X100B Penetration) and a CCD camera and a dosimeter were suspended from the steel
grating of 1st floor outside pedestal to carry out survey (B2 survey) and collect images anddose data.
3) Information obtained:
a. Sediments were observed at the bottom of PCV and in the piping etc.
b. Since the sediments did not stir up when images were taken closely the sediments, it
was assumed that they have a certain amount of weight.
c. The dose decreased when the dosimeter entered the water but increased when it
approached the bottom of PCV.
d. The height from the bottom of PCV at which the dose starts increasing varies dependingon the point where it is measured.
e. Radiation dose rate on the steel grating was approx. 4-12Sv/h. It has not changed muchsince the last inspection (April 2015) and no significant damage to the existing structure
was observed.

Figure A4. 2-2

Results of PCV internal survey (B2 survey) (Unit 1)[Reference: TEPCO "Unit 1 PCV internal survey"]

A-17

(2) PCV internal survey of Unit 2
A. Conducted in March 2012 and August 2013
1) Purpose: Observation of any fallen objects on the platform or any damage and the conditionof the access route near the bottom of the PCV.
2) Method: In March 2012 and August 2013, radiation dose rate measurement and PCVinternal survey with a CCD camera, and stagnant water sampling through PCV penetration(X-53 penetration) were carried out.
3) Information obtained: Dose rate varied from place to place. 31-73 Sv/h was measured in March 2012 and 2436 Sv/h was measured in August 2013.
 A survey inside the pedestal (A2 survey) in the PCV was planned, but since eluted
material was found near CRD hatch (X-6 penetration) and the peripheral dose rate wassignificantly higher than expected values. Radiation dose reduction measures were
undertaken and the survey was postponed to 2016.
B. Conducted in January and February 20171) Purpose: The survey device was inserted through PCV penetration (X-6 penetration) andthe condition of the platform in the pedestal (observation for deformations), condition offuel debris fallen on platform and CRD housing and condition of the structure in thepedestal were observed.
2) Method: From January to February 2017, a survey was conducted by inserting a guide pipewith a CCD camera from X-6 penetration and a survey (A2 survey) was conducted byinserting a remotely operated survey device to acquire images, temperatures and dose
data through the CCD cameras, thermometer and dosimeters attached to the survey
device.
3) Information obtained:
The remotely operated survey vehicle could not reach the platform because of the sedimentson the CRD replacement rails but data such as images, radiation dose and temperature on
the CRD replacement rails could be obtained. Also, images of the inside of the pedestal
were obtained in the preliminary survey with a guide pipe that was conducted before
inserting the remotely operated survey device. The main information obtained by this survey
is given below.
a. In the steel gratings inside pedestal, it was observed that had fallen out or some that had
deformed so much that even the square appeared to be deformed. A lot of sediments
were also seen.
b. No major damage was observed to the CRD housing support near the entrance ofpedestal.
c. Objects were observed that seem to have stuck to the CRD replacement machine andTraversing In-core Probe (TIP) guide pipe support near it.
d. Abnormalities such as cracks were not observed on the wall surface in the pedestal of
the pedestal platform.
e. It was observed that steam was rising from the lower part of the steel gratings.
f. The maximum radiation dose rate of the CRD replacement rails was approx. 70 Sv/h.

A-18

Figure A4. 2-3

Results of PCV internal survey (A2 inspection) (Unit 2)[Reference: TEPCO "Unit 2 primary containment vessel internal survey - Additional inspection by
image analysis"](3) PCV internal survey of Unit 3
A. Conducted in October 2015
1) Purpose: A survey was conducted mainly to observe the condition of cool-down in PVC andalso gather information that would contribute to the future study for the investigation
methods.
2) Method: In October 2015, a camera, a thermometer and a dosimeter were inserted by PCV
penetration (X-53 penetration) to measure the temperature and radiation dose rate. The
survey for the PCV internal structures and the bottom of PCV, and sample stagnant water
were conducted.
3) Information obtained:a. No damage was observed on the structure/ wall surface in PCV within the range of theobservation.
b. No damage was observed in CRD rail and X-6 penetration within the range of theobservation.
c. Sediments were observed on the CRD rail and the steel grating of 1st floor.
(Transparency under water in PCV was good.)
d. Water level in the PCV was OP: approx. 11800 mm, which was almost consistent withthe estimated value.
e. Temperate inside PCV was approx. 26-27°C in the gas phase and approx. 33-35°C in
the water.

A-19

f. Dose rate of the PCV gas phase was approx. 0.8-1 Sv/h. Radiation dose in the PCV wasat the lowest among Units 1-3 which seemed to be an effect of shielding caused by ahigh level of stagnant water.
g. According to the water quality results of the stagnant water in the PCV, PCV does nothave a severe corrosive environment, but rather has low corrosive environment.
Currently an internal survey by inserting an ROV from X-53 penetration is underconsideration to survey the inside of the pedestal in the PCV.

Figure A4.2-4

Results of PCV internal survey (Unit 3)[Reference: TEPCO - Fukushima Daiichi NPS, Unit 3 Primary Containment Vessel (PCV) internalsurvey results]

B. Conducted in July 20171) Purpose: Survey device was inserted through PCV penetration (X-53 penetration) toobserve the status in the basement floor of pedestal.
2) Method: An underwater ROV was inserted through X-53 penetration in July 2017 to carryout survey by taking pictures.
3) Information obtained:
The underwater ROV reached inside the pedestal, and it was possible to gather images of
the condition inside the pedestal for the first time. The main information obtained from thissurvey is as given below:
a. Presence of molten material that appears to have solidified was observed inside thepedestal.
b. It is possible to access the inside of the pedestal through the pedestal opening.

A-20

Figure A4.2-5 Results of PCV internal survey (Unit 3)［Reference: TEPCO, IRID Unit 3 Primary Containment Vessel (PCV) internal survey results(summary of rapid report)］(4) Stagnant water sampling results
Figure A4.2-6 shows the analysis results of concentration of Cs-137 in the stagnant water
samples collected during the main surveys inside S/C and torus room and the PCV internalsurvey. Since the sampling time and the units where they were collected are different, it is difficult
to predict the trend of a specific unit. However, since water treated with cesium and strontium
adsorption equipment has been injected into the PCV as cooling water, concentration of Cs1-37
is overall declining. Figure A.4.2-7 shows the analysis results of concentration of Cs-137 in thestagnant water samples from the T/B, process building and high temperature incinerator building.
Although there are effects of the penetrations in the T/B and the connection of drainage from theneighboring facilities, the concentration is gradually decreasing in the stagnant water in the T/Bof Unit 1, Unit 2 and Unit 4. And the concentration of Cs1-37 in the stagnant water in Units 2 and3 is showing a common tendency wherein the concentration inside of the PCV < that inside of
the Torus room.

A-21

Figure A.2-6 Concentration of Cs-137 in the sampled water(Prepared based on data published by TEPCO)

Figure A4.2-7

Concentration of Cs-137 in water sampled in the T/Bs and water injection
system diagram(Prepared based on data published by TEPCO)

A-22

Appendix 4.3: Survey results of S/C and torus room
Amongst the surveys conducted so far in S/C and torus room, the main status is summarized below.

(1) Survey around the lower part of vent pipe in Unit 1
A. Period: November 2013
B. Purpose: Observation whether there is a water flow from the edge of vent pipe sleeve andsee the condition (external observation) of sand cushion drain pipe etc. In addition, measure
the radiation dose in lower part of vent pipe.
C. Method: Survey with a water boat equipped with a camera and dosimeter.
D. Information obtained: The following information was obtained.
1) It was observed that water was flowing down the S/C surface from the top of S/C aroundvent pipe X-5E.
2) It was observed that the sand cushion drain pipe around vent pipe X-5B has come off andwater is flowing out.
3) Radiation dose rate was approx. 0.9 - 2 Sv/h.

(2) Survey of the top of S/C and wall surface of torus room in Unit 1
A. Period: May - June 2014
B. Purpose: Observation for water leakage from the top of S/C near vent pipe X-5E, where awater flow was observed in the survey of November 2013 and observation for water leakage
from penetration of the torus room's east wall surface. The radiation dose measurement in
the torus room.
C. Method: Survey using an inspection device equipped with a crawler having a camera, sonar(ultrasonic sensor) and dosimeter
D. Information obtained: The following information was obtained.
1) A leakage was observed from the protective cover of the expansion joint of vacuum breakline which connects to vent pipe X-5E at the top of S/C.
2) No leakage was observed in the penetrations of the torus room’s east wall surface. Sonar
inspection was reconsidered since the place to be surveyed was narrow.
3) Maximum radiation dose rate was approx. 2.4 Sv/h.

(3) Survey in torus room in Unit 2
A. Period: April 2012
B. Purpose: Visual inspection of torus room (obtaining images), dose measurement andcollecting acoustics from torus room.
C. Method: Survey using a remote control robot equipped with a camera, dosimeter and acoustic
device.
D. Information obtained: The following information was obtained. The acoustic device could not
be retrieved because communication with the remote control robot got cut off during survey.
1) Images were obtained on steel grating.
2) Maximum radiation dose was approx. 118 mSv/h.

A-23

(4) Survey in torus room in Unit 2
A. Period: April 2013
B. Purpose: Measurement of the radiation dose and temperature in torus room and obtaining
images.
C. Method: Survey by inserting a thermometer, dosimeter and camera through a hole opened in
the floor of the RHR heat exchanger (B) room on the south side of the reactor building's 1st
floor
D. Information obtained: The following information was obtained.
1) Dose rate was measured at intervals of approx. 1m and the maximum dose was approx.
134 mSv/h near water surface.
2) Water level was around OP. 3260 (depth approx. 5.3 m).
3) Temperature was measured at the height intervals of approx. 1m. Air temperature wasapprox. 20°C and underwater temperature approx. 25°C.
4) Stagnant water was transparency more than approx. 100cm.
5) Although rust was found on the structure in torus room as far as it could be observed, nomajor damage was found. Image of the bottom part could not be obtained because it
interfered with the torus room stairs at approx. 1.5 m underwater.

(5) Survey in torus room in Unit 3
A. Period: July 2012
B. Purpose: Visual inspection of torus room (obtaining images), radiation dose measurement
and collecting acoustics from torus room.
C. Method: Survey using a remote control robot equipped with a camera, dosimeter and acoustic
deviceD. Information obtained: The following information was obtained. The acoustic device could not
be retrieved because communication with the remote control robot got cut off during
inspection.
1) Images were obtained on steel grating.
2) The maximum radiation dose rate was approx. 360 mSv/h.

A-24

Appendix 4.4: Results of muon detection
Fuel debris distribution was measured using muon detection technology through the transmission
method for Unit 1 in 2015 and for Unit 2 in 2016. Measurement results are summarized below.
Measurement for Unit 3 has been started in May 2017.

(1) Unit 1
For Unit 1, fuel debris distribution was measured twice from February to May and from May to
September 2015 by the transmission method of muon measurement.
With the measurement results, it is reasonable to assume that no fuel fragment larger than 1 m,
which is maximum muon detection capacity by transmission method or no water exist at theoriginal reactor core

Figure A4.4-1

region.

Results of fuel debris measurement by transmission method of muon detection

for Unit 1 [Source: Quick Report on the Measurement Results for Unit 1 from Development ofTechnology for Detecting the Fuel Debris Location inside Nuclear Reactors from TEPCO](2) Unit 2
For Unit 2 fuel debris distribution was measured by the transmission method of muon
measurement from March to August 2016. The following was observed from these results.


It was observed that high density material believed to be fuel debris is present at thebottom of the RPV.
Based on the evaluation conducted by comparing with the simulations, it was presumed
that most of the fuel debris is at the bottom of the RPV. It also indicated that some high-A-25

density material believed to be fuel debris is present in the lower part and outer
periphery of the core. However, some uncertainty in the test remains because of theeffects of the reactor building's structure.

Figure A4.4-2

Unit 2 Results of fuel debris measurement by transmission method of muon

detection [Reference: TEPCO - Fukushima Daiichi NPS Unit 2, Locating fuel debris location bymuon measurement]

A-26

In addition, muon radiography using nuclear emulsion counter was conducted by joint researchof Toshiba corporation and Nagoya University in 2015, and by comparing the analysis value ofmuon count and the measured value, the residual ratio of the core part was obtained as (9 to 36)± 51%.

(3) Unit 3
Muon detection has being carried out for several months since May 2017. The current results ofthe evaluation being carried out indicate the possibility of presence of some fuel debris in both
the reactor core region and the bottom part inside the RPV, but the presence of any high-density
substance has not been observed so far.

Fig A4.4-3 Results of fuel debris measurement by transmission method of muon detection for Unit
3 ［Reference: TEPCO - Fukushima Daiichi NPS Unit 3, Measurement status regarding locating
fuel debris location by muon measurement (interim report)］

A-27

Appendix 4.5: BSAF project outline and results so far(1) Outline of BSAF project
The BSAF project of OECD/NEA aims to improve the severe accident progression analysis code
and analyze the accident progression in Unit 1 to Unit 3 of the Fukushima Daiichi NPS and theircurrent condition to provide useful information for decommissioning of the reactors. Phase 1 wascarried out with 8 participating countries (France, Germany, Korea, Russia, Spain, Switzerland,
U.S.A and Japan) for 3 years from April 2012.
After phase 1, phase 2 has being carried out since April 2015 with 11 participating countries(Canada, China, Finland, France, Germany, Korea, Russia, Spain, Switzerland, U.S.A andJapan). The final report will be summarized in March 2018.

(2) Result of phase 1
In phase 1 of the BSAF project of OECD/NEA, 13 institutions in Japan and abroad conducted
severe accident progression analysis of Unit 1 to Unit 3 for 6 days after the earthquake. Resultsof this analysis are shown in Table A4. 5-1.

Table A 4.5-1
Area

Result of fuel debris distribution evaluation in BSAF phase 1 [unit: ton]
Unit 1

Unit 2

Unit 39 institutions

6 institutions

3 institutions

4 institutions

5 institutions

Reactor core

0-3

0-14

0-32

0-21

0-36

RPV lower head

0-8

0-91

0

8-81

0

PCV

105-1640147-2400140-268(Provided by IAE)

The analysis results of Units 2 and 3 were categorized into two cases, one where the fuel debris
remains in the RPV, and one where it falls down to the PCV. The result of Unit 2 depends on themodel of the fuel debris relocation from the reactor core to the lower plenum and the assumed
amount of the water injection by the fire engine which is highly uncertain. With regard to Unit 3,
the results were affected by the difference in the assumption of HPCI water injection behavior(amount of water injected in reducing the RPV pressure), that is, the maximum quantity of thesteam flow rate that drives HPCI and its cycle were significantly different among the institutions
that performed the analysis.

(3) Status of phase 2
In phase 2, the latest knowledge is being incorporated to carry out severe accident progression
analysis for three weeks after the earthquake, with an aim to make advancements in analysis.

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Also, knowledge about severe accident research analysis is being shared through workshops
regarding MCCI or the state of Appendix on the FP reactor internals. Progress meetings (PRG)
and workshops are held twice a year and the final report will be compiled in March 2018.

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Appendix 4.6: Characteristics of MAAP code and SAMPSON code and analysis
results
Characteristics of the SAMPSON code and MAAP code, the severe accident progression analysis
codes, are shown in Table A 4.6-1 and the analysis results by both codes as well as noteworthy
comments about the analysis results are given in Table 4.6-2.
Table A 4.6-1
Characteristics of MAAP code and SAMPSON code
Code
MAAP code
SAMPSON code
Developed by:
U.S. EPRI
Japan NUPEC (Institute of Applied
Energy is continuing thedevelopment at present)
Analysis targets
In+Ex Vessel
In+Ex Vessel
User adjustment
None (with dependency on mesh
Multiplefactordivision)
Calculation time
Short
20 to 30 times the real time
Characteristics
 Users can obtain the expected Analysis results do not depend
analysis results by combining theon the user since it is built with aadjustment factors.
theoretical and mechanistic
 Analysis results often differ fromphysical model.
user to user.
Verifying the User groups will continue to Participated in OECD/NEA's
individual model,
improve the code.
international benchmark problem
code etc.
(ISP: International Standard
Problem) and was highly
acclaimed.
 It has also been verified byseveral other experimental
analyses.
[Reference: Koji Okamoto, Atomic Energy Society of Japan,
SA evaluation research expert committee's documents]

Table A.4.6-2

Analysis results using accident progression analysis code
(Implemented in FY2015)
Unit 1

Location

Unit 2

Unit 3[unit: ton]

MAAP

SAMPSON

MAAP

SAMPSON

MAAP

SAMPSON

Reactor core

0

0

0

13

0

29

RPV lower

15
10
25
58
25
79
head
Inside109(78)79(130)92(37)
76(14)
103(51)53(20)
pedestal
Outside33(52)52(0)102(4)
5(0)
96(6)
0(0)
pedestal
Total value(including
287
271
260
166
281
181
concrete)
Note) The weight shown for inside and outside pedestal is the fuel/ structure material weight(without concrete weight) and concrete weight is shown inside (

A-30

).








In Unit 1, RPV damage began before the start of water injection. Both the MAAP code andthe SAMPSON code show a trend of most of the debris falling into the RPV pedestal.
In Units 2 and 3, analysis using the SAMPSON code indicated that the granular particles of fueldebris were cooled by water injection and accumulated inside the RPV. At the core region andbottom of the RPV, amounts of debris estimated by the SAMPSON code are larger than that
estimated by the MAAP code.
In the SAMPSON code model, granular particles debris are modeled as a mass which cools
down when water is present. It may be possible that the cooling water gets distributed unevenly
and granular fuel debris which does not come in contact with cooling water melts again anddrops on the lower plenum.
Results by the SAMPSON code shows that some fuel in the outer periphery of the core hasremained in stub-shape, but that is because there is water outside the reactor core shroud whichis contributing towards the cooling. The fuel will melt and drop if there is no water.

[Reference: IRID Completion report for "Improvement of recognition regarding the internal PCV
condition using severe accident progression analysis and actual plant data"]

A-31

Appendix 4.7: Summary of heat balance method and estimation result
The heat generation is still continued due to decay heat from the fuel debris and water injection
into the RPV to cool it down. As temperature of water injection goes up to the RPV temperatureand from RPV temperature to PCV stagnant water temperature, cooling has been continued withincrease in temperature.
Heat balance method is a method that estimates the fuel debris proportion inside the RPV and PCV
assuming it maintains (balance) the state that enables decay heat (heat generation) = Sum of
temperature rise of cooling water (heat release), That is, the heat balance method means toestimate the proportion between the fuel debris in the RPV and PCV based on the assumption thatthe temperature of the cooling water injected into the RPV is raised to the temperature of theaccumulated water by the heat source (fuel debris) inside the RPV and PCV, namely theassumption that the heat input (heat of the injected cooling water and decay heat) comes intobalance with heat radiation (heat radiated out to the building or into the air through the PCV walls
and cooling-water temperature rise caused by fuel debris). Figure A4.7-1 is a conceptual diagramof evaluation based on the heat balance method.
Shown below is the fuel debris distribution estimated by the heat balance method for each unit.

Figure.A4.7-1

Conceptual diagram of evaluation by the heat balance method[Reference: IRID Completion report for "Sophistication of the internal investigation of the PCVs
using accident progression analysis and actual plant data"](1) Evaluation results for Unit 1
For Unit 1, an evaluation based on the heat balance method was performed on the assumptionthat no heat source exists inside the RPV (i.e., the RPV decay heat in the equation above = 0)
according to the analysis results based on the MAAP code. The evaluation used the quantity of

A-32

the heat of the injected and leaked water calculated from actually measured temperatures of bothwater, with the degree of contribution of decay heat to the rise of the accumulated watertemperature used as an evaluation parameter.
The evaluation results indicate that if it is assumed as a heat source equivalent to 45% of decay
heat, the actually measured changes in stagnant water temperature are almost reproducible as
Figure A4.12-2 shows; thus it is determined that a significant heat source exists inside the PCV.
It is, however, deemed that the fuel debris distribution is significantly affected by the uncertainty
about the decay heat of the fuel debris that fell at the bottom of the PCV (according to evaluationby the JAEA, the decay heat decreases approximately 60% if all of the highly-volatile nuclides
are released), possibility of heat release from fuel debris to the floor concrete, and uncertainty
about the evaluation of the thermal conductivity of the heat radiation from the PCV into theoutside air.

Figure A4.7-2

Example of an evaluation based on the Heat Balance Method for Unit 1(Provided by IRID)(2) Evaluation results for Unit 2
The evaluation was made based on the quantity of the heat of the injected and leaked water
calculated from actually measured temperatures of both water, using as evaluation parameters
the degree of contribution of decay heat to the rise of the accumulated water temperature andthe ratio between the amounts of the heat sources (fuel debris) inside the RPV and PCV.

A-33

Figure A4.7-3 shows an evaluation result example. With the ratio between the quantity of theheat sources (fuel debris) inside the RPV and PCV used as a parameter, the evaluation results
indicate that if it is assumed that 30 to 60% heat source remains inside the RPV, the changes instagnant water temperature inside the RPV and PCV are almost reproducible.

Figure A4.7-3

Example of an evaluation based on the Heat Balance Method for Unit 2(Percentage of Heat Source inside the RPV: 60%) (Provided by IRID)(3) Evaluation results for Unit 3
As with Unit 2, the evaluation was performed based on the quantity of the heat calculated fromthe actually measured water temperatures.
Figure A4.7-4 shows an example of evaluation results. The evaluation results indicate that if it isassumed that 20 to 70% heat source remains inside the RPV, the changes in stagnant watertemperature inside the RPV and PCV are almost reproducible. They also indicate, however, thatsince the changes in the temperature of the stagnant water in the RPV do not agree with that of
the injected water, the quantity of the fuel debris existing inside the RPV as a heat source is alower than estimated.

A-34

Figure A4.7-4Example of an evaluation based on the Heat Balance Method for Unit 3(Percentage of Heat Source inside the RPV: 40%)
(Provided by IRID)(4)Summary of evaluation results from Units 1-3
Table A4.7-1
Unit
Unit 1:Results of estimation of fuel debris distribution by the heat balance method
Estimation results

Heat source equivalent to 45% of decay heat may be present in the PCV.
(Based on analysis result by MAAP code, it is evaluated that there is no heat sourcein the RPV (decay heat in the RPV is zero))

Unit 2:30-60% heat source (fuel debris) may be present in the RPV.

Unit 320-70% heat source (fuel debris) may be present in the RPV. However, since thetemperature of the RPV stagnant water does not follow the water injection
temperature, the amount of debris present in the RPV as a heat source may be evensmaller.

Note:

It is deemed that the fuel debris distribution is significantly affected by the uncertainty

about the decay heat of fuel debris that fell at the bottom of the PCV (according to evaluation by
the JAEA, the decay heat decreases approximately 60% if all of the highly-volatile nuclides are
released), the possibility of heat radiation from fuel debris to the floor concrete, and uncertainty
about the evaluation of the thermal conductivity of the heat radiation from the PCV into the outside
air.
(Provided by IRID)

A-35

Appendix 4.8: Location of fuel debris estimated from plant parameter trend
The heat source in RPV (fuel debris) was estimated based on the trend of the temperature aroundthe RPV in the post-accident condition, water temperature of S/C and amount of water injected in
FDW system and CS system. Figure A4.8-1 shows the difference between flow paths of the FDW
system and CS system. FDW system is a system that introduces the cooling water during thenormal operation of BWR into the RPV. After flowing into the RPV, cooling water is accumulated in
the space between reactor core shroud and the RPV (annulus), it will be flowing into the jet pumpat the time when the water level reaches to the upper part of the jet pump mixer. If the integrity ofthe bottom of the RPV is maintained, the cooling water flowed in will accumulated inside the RPV
and the water level the accumulated water will be raised. However, since increase in the water level
was not observed, the bottom of the RPV has been damaged and the cooling water is estimatedto be flown down from the damaged portion into the inside of the pedestal. That is, FDW systemwater injection cannot cools down the BWR reactor core portion but the bottom part of the RPV,
even after the accident. While on the other hand, CS system is a core spray system in case of theloss of coolant accidents and is installed along the inner walls of the core shroud immediately above
the core. In the CS system water injection, the cooling water flowing down the space from thereactor core to the bottom of the RPV and the space can be cooled down. Based on the above, thefuel debris locations were estimated for each Unit.

Figure A4.8-1

Flow path of FDW and CS systems［Reference: Public data released by TEPCO］

Shown below is the fuel debris distribution for each unit estimated from the trend of the plantparameters.
(1) Evaluation results for Unit 1

A-36

Figure.A4.8-2 shows temperatures at several positions inside the PCV of Unit 1 along withchanges in the quantity of injected water and measurement locations. In response to thechanges in the quantity of injected water, the following characteristic changes in temperature
were observed.
1) Compared to Units 2 and 3, the ambient temperature of the RPV decreased at a faster rate,
which decreased below 100°C in five months after the accident.
2) The ambient temperature of the RPV did not rise that corresponds to the decrease in theamount of water injected by the FDW system.
3) With increases in the quantity of water injected for the FDW system, the ambient
temperature of the RPV dropped to below 50°C and the S/C water temperature rose.
4) With decreases in the amount of water injected for the FDW system, the ambient
temperature of the RPV rose.
With the characteristics shown in 1), 2), and 3) above, it is estimated that the heat source inside
the RPV is probably small. From 3) and 4), it is assumed that a heat source may exist in thewater injection channel of the FDW system and the heat removed in response to water injection
has transferred to the S/C.

Figure A4.8-2

Changes in plant parameter and measurement locations for Unit 1(2) Evaluation results for Unit 2
Figure 4.8-3 shows the temperatures at several positions inside the PCV of Unit 2 along withchanges in the amount of injected water and measurement locations. In response to the changes
in the amount of injected water, the following characteristic temperature changes were observed.
1) Compared with Unit 1, the ambient temperature of the RPV is high, which was higher than
100°C even six months after the accident.
2) The temperature of the lower RPV head sensitively responded to the decreases in theamount of water injected for the FDW.
3) With the start of water injection for the CS system, the ambient temperature of the RPV
decreased and the S/C water temperature rose.

A-37

4) With decreases in the amount of water injected by the CS system, the ambient temperatureof the RPV rose. The biggest rise in temperature at this time was seen in the lower part of
the RPV lower head.
5) With increase in the amount of water injected by the CS system, the ambient temperatureof the RPV dropped.
With the characteristics shown in 1), 3), 4) and 5), it is estimated that a certain quantity of heatsource may exist inside the RPV; from (2), it is assumed that the lower RPV head is closer tothe heat source than the water supply nozzle (N4B). From (3), it is deemed that the removed
heat had transferred to the S/C.

Figure A4.8-3

Changes in plant parameter and measurement locations for Unit 2(3) Evaluation results for Unit 3
Compared with Unit 1, the ambient temperature of the PRV is high, which was higher than 100°C
even six months after the accident; as with Unit 2, it is assumed that a certain percentage of fueldebris exists in both of the RPV and PCV. The procedure for the above estimation is shown
below:
1) Compared with Unit 1, the ambient temperature of the PRV is high, which stayed at a level
higher than 100°C even six months after the accident.
2) Although the amount of water injected by the FDW system is the highest in all units, theambient temperature of the RPV decreased at a low rate.
3) With the start of water injection by the CS system, the ambient temperature of the RPV
dropped rapidly.
4) With a decrease in the amount of water injected by the CS system, the temperatures of thewater feed nozzle (N4B) and the RPV lower head rose.
With the characteristics shown in 1), 3), and 4) above, it is estimated that a heat source may
exist inside the RPV.

A-38

Figure A4.8-4

Changes in plant parameter and measurement locations in Unit 3.

(4) Summary of evaluation results from Units1-3

Table A.8-1 Results of estimation of fuel debris locations based on the trend of plant parameters
Unit
Estimation results
Unit 1: It is likely that the heat source in the RPV is small.
It is possible that heat sources are present in the flow path of water injected by the
FDW system, and the amount of heat removed by water injections is transferring to
S/C.
Unit 2:A certain quantity of heat source may be present in the RPV.
The distance to the heat source is the closer from the RPV lower head than the
feedwater nozzle N4B.
The removed heat is transferring to S/C.
Unit 3:A certain quantity of heat source may be present in RPV.
Since there is a large amount of stagnant water in the PCV there is not much change
in the parameter because of which it is difficult to estimate.

A-39

Appendix 4.9: List of properties of fuel debris
The estimation result of shape of fuel debris is shown below as an example of a list of propertiesof fuel debris.

A-40

A-41

References (Estimation of Fuel debris properties)
Eidam: G. R. Eidam, “Core Damage” Chapter 5 of “The Three Mile Island Accident,” 1986 American Chemical Society, Volume 293.
EPRI: W. C. Holton, C. A. Negin and S. L. Owrutsky “The cleanup of Three Mile Island Unit 2 A Technical History: 1979 to 1990”,
EPRI NP-6931 (1990)
H26:
IRID et al. “Operating grants for Government-led R&D Program on Decommissioning and Contaminated Water Management(Research Report of Fuel Debris Characterization and Development of Fuel Debris Retrieval Processing Technology)” (2015)
H27:
IRID et al. “Operating grants for Government-led R&D Program on Decommissioning and Contaminated Water Management(Research Report of Fuel Debris Characterization) (Interim report)” (2016)(Provided by IRID)

A-42

Appendix 4.10: Status of the comprehensive analysis and evaluation of theconditions inside the reactor in each unit

A-43

[Reference: TEPCO “Estimation of current status inside RPV and PCV at Fukushima daiichi NPS”
(The 2nd International Forum on the Decommissioning of the Fukushima Daiichi NPS)]

A-44

Appendix 4.11: Studies on containment function
The accident at the Fukushima Daiichi NPS resulted in a long-term loss of power that deprived thepower station of its cooling capability, resulting in a core meltdown. This was followed by the lossof the RPV and PCV containment boundaries as well as the loss of containment function of the
R/B, resulting in severe environmental contamination. Under a situation where the confinement
capability (boundaries) is incomplete, our critical challenge is to reduce the radiation dose and its
impact, i.e., the risk level, through emergency measures, stabilization of the facilities, and cleaning
after the accident.
This section gives more information about the containment function such as the concept on which
the power station was originally designed, what measures were taken at the time of accident, what
the present situation is and what challenges are involved now that we are entering into a new phaseof ‘fuel debris removal’, which is summarized below.

(1)Containment function of a NPS under a normal condition
In order to examine containment, its targets must be first clarified. A typical power station contains
mainly two types of substances: fission products (FP) and corrosion products (CP).
FPs are produced inside fuel pellets by nuclear fission. Some of it leaves the fuel pellets and iscontained inside the fuel cladding. If this tube is damaged, FPs leak out into to the water, thecoolant, in the nuclear reactor. Amongst the FPs, while special attention is paid to noble gas andiodine as they are likely to leak out in a gaseous state, if a leak is detected, the reactor is stopped
to prevent the cooling water from being contaminated with FP, and measures are taken to replacedamaged fuel assemblies.
On the other hand, CPs exist in the coolant as products that are resulted from corroded structurematerials, adhere to the fuel at the core, and then become liberated. The CPs that requireparticular attention are Co-60 and Mn-54. CPs are securely controlled with a treatment system
for liquid waste to prevent them from having an impact on the outside of the power station. In
addition, since they greatly expose workers to radiation, measures for reducing radiation
exposure are being carried out such as use of low-cobalt materials as structural materials,
improvement of reactor water chemistry, and removal of corrosion products.
During normal operation, the FPs and CPs mentioned above, including small amount of leaks,
are kept within the radiation controlled area and treated with a waste treating system to ensurethat they are below the reference value before they are released.
During scheduled outage, RPV and PCV are opened and the reactor well is filled up with waterto replace the fuel and inspect the inside of the reactor. In this case also FPs and CPs are kept
within the radiation controlled area and treated with a waste treating system to ensure that they
are below the reference value before they are released. Radioactive dust may be a problem

A-45during construction, and therefore its release from the radiation controlled area is restricted by
setting up barriers or separate air-conditioning etc.

(2)Containment function at the time of accident
Originally, the containment function is expected to prevent FPs from being released when an
accident occurs. As a containment function, FPs are contained by 5 layers of walls: (1) fuel pellets,(2) fuel cladding, (3) reactor coolant-pressure containment boundary (RPV containment
boundary), (4) primary containment vessel (PCV containment boundary), and (5) R/B (secondary
containment). Out of these containment functions, the RPV boundary is for containing liquidphase and the PCV boundary and R/B are for containing gas-phase. In order to prevent hydrogen
explosion in BWR plant in the event of an accident, the PCV is filled with nitrogen and inactivatedduring operation. According to the scale of the accident, the release is prevented by the fuelcladding tube, RPV boundary, or PCV boundary. Various accidents at BWR plants so far have
not resulted in significant core damage caused by core re-submersion.
This even applies to an accident that undergoes a guillotine break of the RPV boundary piping
resulting in a loss of the coolant. If an accident occurs, for the liquid phase, a circulation loop is
established as a boundary that allows cooling water that has flown out from the ruptured part tomove to the S/C and then be poured into the reactor by the ECCS system. Regarding the PCV
gas phase part, while containment is maintained, radioactive substances leaking at slightly below
the allowable value are retained inside the R/B and released while being processed in an
emergency gas treatment system. Regarding this, the dose standard for accidents at the siteboundary is determined as 5 mSv / accident.

The accident at the Fukushima Daiichi NPS, however, resulted in a long-term loss of power,
leading to conditions more severe than those assumed in the design phase.
The loss of the cooling capability of the reactors and the damage to the cores resulted in lossesof two walls: (1) fuel pellets and (2) fuel cladding tube. Furthermore, the molten core damaged
the RPV, resulting in a loss of the (3) RPV containment boundary. Subsequently, the temperatureand pressure inside the reactor containment vessels became high, which damaged the (4) PCV
containment boundary, causing leakage of steam containing radioactive substances andhydrogen, produced by a water-zirconium reaction, into the R/B. Since the standby gas treatment
system did not work either due to the loss of power, hydrogen explosions and other factors
damaged the (5) R/B.
However, although damaged, the PCV still existed which suppressed the release of gases otherthan highly volatile gases such as the noble gas and iodine. Although the reactor core melting
could not be avoided, reactor core cooling was possible. This development was never seen in
Chernobyl.

A-46

As a result, the release of cesium to the air outside the PCVs was controlled under 2%. For theother nuclides, the release amounts were even lower.
On the other hand, it is necessary to keep pouring in water to cool the fuel debris but that waterleaks out from the breakages in the PCVs. This causes radioactive substances contained inleaked water to continuously flow out and mix with the stagnant water in the buildings, which
presents a problem of contaminated water caused by degradation of the containment function
for the liquid phase.
Since the R/B’s water seal function has also degraded, the ground water keeps flowing in andincreases the contaminated water. This has further complicated the problem. To prevent
contaminated water from increasing and radioactive substances from flowing out to the outsideof the system, a cooling water injection system is used that purifies contaminated water and thenreuses it to cool reactor cores with the building used as a boundary. The level of the accumulated
water inside the building is kept lower than that of ground water so that ground water will flow in(in leak) so as to prevent radioactive substances from flowing out (out leak) to the outside of thebuilding. This continuous water injection successfully cooled fuel debris and sufficiently
decreased the temperature inside the containment vessels to below 100°C. This remarkably
suppressed the release of radioactive substances, achieving a so-called stable cold shutdown
state.

(3)Containment functions at the Fukushima Daiichi NPS at present
The present containment functions at the Fukushima Daiichi NPS is as follows.
The containment function for the liquid phase uses a cooling water injection system that purifies
the water accumulated inside the building using a water treatment system and then reuses the
purified water to cool the reactor core with the PCV used as the primary boundary to retain
cooling water poured into the reactor and the building used as the secondary boundary to retain
water leaked from the PCV. The level of the accumulated water inside the building is kept lower
than that of ground water so that ground water will flow in (in leak) so as to prevent radioactive
substances from flowing out (out leak) to the outside of the building. This causes the groundwater that has flowed into the reactor to be redundant water, which must be stored as water
treated by the water treatment system. This requires a storage tank to be kept inside the premises,
which presents a problem with contaminated water. To counter this problem, subdrainage
equipment is installed near the building to pump up ground water and release it into the harborthrough a purification system, thus lowering the level of ground water to reduce the amount thatflows into the building. Multiple layers have also been set up by putting up land-side impermeable
walls to reduce the groundwater near the R/B. Transfer system for stagnant water inside thebuilding is also installed so that the level of the accumulated water inside the building can becontrolled. To reduce the amount of ground water that comes in and prevent contaminated water

A-47

from flowing out, it is required to be able to securely control the level of water inside and outside
the buildings under any condition.
As for boundary for the gas phase, a nitrogen filler is used to inactivate the inside of the PCV anda PCV gas monitoring system is installed to extract gases inside the PCVs, filter them, andmeasure radiation from them before releasing them. This arrangement keeps the pressure in thegas phase slightly positive to prevent a hydrogen explosion and minimizes the release ofradioactive substances. With these measures, the estimated radiation dose at the boundary of
the premises due to the radiation released from Units1-4 is sufficiently low--approx.0.00024
mSv/year.

(4)Issues concerning the containment function during the fuel debris retrieval work
Removal of fuel debris involves cutting the debris, which probably causes α fine particulates tomove to the liquid or gas phase inside the PCVs. In addition, changing the level of water inside
the PCV is being considered, meaning that the requirements for the containment function may
change.
For alpha particulates, the upper limit of concentrations associated with internal exposure
through inhalation, in particular, is stricter than that for the other nuclides. For this reason,
attention must be paid to particulates that may be released as dust from the gas phase and thegas-phase containment boundary must be arranged based on careful consideration. Hence, it iseasier to control the alpha particles if they are moved to liquid phase by doing the task of cutting
fuel debris under water or with pouring water as far as possible. In this case, raising the level ofwater inside the PCVs must be considered. This is one of the major aims for which intensive
R&D for PCV repairs (water sealing) is in-progress at present.
Regardless of the fuel debris retrieval method, alpha particles are basically being converted to
the liquid phase wherever possible, which inevitably raises the concentration of radioactive
substances in the liquid phase. This raises an issue of how to build a liquid phase containment
function while considering the current control for the difference in the water level between theinside and outside the R/B.
As for the containment function for the gas phase, although an action is taken to convert alphaparticles to the liquid phase as much as possible while removing fuel debris removal, theconcentration of alpha nuclides in the gas phase is expected to increase. Now the challenge isto develop a system wherein a cell is installed in the upper part the PCV to keep the insidepressure negative (primary containment boundary) while a container is also installed in the R/B
to set up an air conditioning system that controls the inside pressure to keep it negative(secondary containment boundary).
While developing these containment functions, not only normal operations but measures in theevent of an emergency must also be considered.

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Appendix 4.12: Study of water level at the bottom of PCV when retrieving fueldebrisAssuming that it is possible to control water level by installing of underwater pumps, water level at
the time of fuel debris retrieval at the bottom of PCV is illustrated and explained by dividing into twocases, that is, achieving and not achieving vent pipe water seal*1 (Fig A4.12- 1).
The upper part of the diagram shows the current water level, the middle part is the conceptual
diagram when water sealing in the lower PCV has been achieved by vent pipe repairing, and thelower part is the conceptual diagram when water sealing of vent pipe could not be achieved.
It is necessary to setup and control the water level depending on the success or failure of therepairs to the vent pipe.
*1: Inability to perform water sealing for vent pipe also includes the cases in which water sealtechnology cannot be applied at on-site as construction is not possible due to high dose of radiation.

[Water level at the bottom of PCV in Unit 1]
When water sealing of vent pipe successful - The expected water level when water sealing for ventpipe is done is shown in the middle part of the diagram. The water level in the PCV can be
maintained to the present level (about 2 m from the bottom of PCV), and fuel debris retrieval canbe carried out underwater. In order to improve the seismic resistance margin of S/C while retrieving
fuel debris, it is preferable to dry up the inside of the S/C (if it is dried up in the beginning, that state
can be maintained). If the water level in the PCV lowers, it can also be carried out in a state ofpartial submersion where water is poured in.
In case water sealing for vent pipe could not be done, water sealing of strainers are carried out
since downcomer water seal has a poor seismic margin. From the viewpoint of seismic resistanceof S/C, it is desirable to continuously dry up the inside of S/C. As shown in the lower part of the
diagram, water level of bottom part of PCV cannot be maintained. For example, it will be low as inUnit 2 (About 0.3 m from the bottom part of PCV). Therefore, the fuel debris retrieval is carried out
in partial submersion state with water being poured in.

[Water level at the bottom of PCV in Unit 2]
If vent pipe water sealing is done, it is possible to increase the water level at the bottom of PCV.
For example, if the level is set at current level of Unit 1 (approx. 2m from the bottom of PCV), it will
be possible to carry out fuel debris retrieval under water. It will also be possible to carry out work
under partial submersion, with water being poured in, by maintaining the water level inside PCV in
the current state.
If vent pipe water sealing cannot be done, one method to stop water could be to apply a downcomer
water seal in which high-flowing concrete is poured into S/C to bury the tip of the down comer andA-49

separate the D/W and S/C. The performance of downcomer water seal has been confirmed by
testing, but its seismic resistance, although promising, needs to be confirmed in future detailed
testing. Also, though leakage from S/C is expected in Unit 2, prevention through water sealing
method will be planned. Fuel debris retrieval work is done underwater or in a state of partial
submersion where water is poured in. The lower part of the diagram shows retrieval tasks being
carried out under water.

[Water level at the bottom of PCV bottom in Unit 3]
If vent pipe water sealing is done - If side access method is to be adopted, the current water level(approx. 6m from the bottom of PCV) needs to be reduced to the current level of Unit 1 (approx.
2m from the bottom of PCV) (middle stage in the diagram).
In this case, it will be possible to carry out fuel debris retrieval work under water.
However, it is necessary to obtain technical support for keeping this water level stable during theretrieval work. It will also be possible to lower the water level further and carry out the task inpartially submersed state with water being poured in.
If vent pipe water sealing is cannot be done - Like Unit 2, a water sealing method that can beconsidered is to apply downcomer water seal for separate D/W and S/C. Fuel debris retrieval work
is done underwater or in a state of partial submersion where water is poured in. The lower part ofthe diagram shows retrieval tasks being carried out under water.

A-50

Figure A4.12-1

Water level at the bottom of PCV and water sealing for vent pipes(Conceptual diagram)The current water levels inside PCV are: Unit 1: water depth of about 2m; Unit 2: water
depth of about 0.3m; Unit 3: water depth of about 6m.
If water sealing can be achieved in vent pipes, the task of retrieving fuel debris can be
performed under water in each unit. The water level inside PCV is controlled by operating
the circulation system from D/W for raising the water level of Unit 2 or lowering the waterlevel of Unit 3. If the water level has been lowered, the task can be carried out in partiallysubmersed state with water being poured in.
If vent pipe water sealing is not done, strainer water seal is applied in Unit 1, and water

A-51

Appendix 4.13: Preliminary Seismic Margin Evaluation of RPV and PCV.

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A-59

Appendix 4.14: Development of fuel debris retrieval equipment and devices
Objective(1) Test with a hydraulic manipulator

Test details

Checking the basic feasibility of the control characteristics helpful in designing amanipulator for removing fuel debris, and identifying their challenging items

Present situation• Evaluating the effect of the hose length (10 m or 100 m) based on the assumption thata long distance is required between the fluid-pressure controller and access unit when
the latter must be remotely controlled• Evaluating the effect of the load (15 kg) to be mounted at the end of the manipulator• Setting the moving speed of the laser end as the operating speed (2 mm/second or so)

Evaluation andTests up to FY2016 complete. It was possible to confirm the basic features of fuel debris

challenges

retrieval manipulator.

Objective

For practical application, it is necessary to verify its controllability, considering its use incombination with tip tools.

Overview

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Designation(2) Technical developments for cutting/dust collection, visualization and measurement

Objective• For easy removal, fuel debris must be processed into easy-to-handle sizes. Identifying
the characteristics of cutting technologies that seem to be effective for fuel debris
processing and developing technologies for collecting dust and aerosols produced
during processing• Developing a radiation-proof camera to be used in the environment of fuel debris
removalTest details• Improvement bit: Checking the stability in early stages of cutting and the amount ofchips to be produced, and verifying the processibility of the mock fuel debris• Non-core bit: Checking the processibility of mock fuel debris• Laser: Measuring the processing efficiency and the weight and the particle size
distribution of secondary products (fumes) moving into water and air
Analytic evaluation to select a method for preventing fumes that move into the air fromspreading
Checking the effect of removing fume from gases• Camera: Irradiation test on the pickup camera tube and radiation-resistant camera

Present situation

Tests up to FY2016 complete. With respect to cutting technology, it was possible to verify
cutting performance using laser and boring. With respect to visual component, irradiation
test of imaging tube and radiation-proof camera was conducted and its resistanceagainst radiation was verified.

Evaluation andSpecific equipment needs to be designed (including combinations with robotic arm etc.)

challenges

for application at on-site. It is necessary to continue studies on an efficient dust collection
technology.

Overview

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Designation(3) Development of the devices to access to the inside of the RPV

Objective

Developing a device that brings the fuel debris handler handing device nearby fueldebris and supports the reaction force during fuel debris removal (processing andcollection)

Test details• Prototyping a 1/1 scale element of the lower table, which is a common platform forwork inside the RPV• Checking the basic action of the support mechanism as well as the performance ofremove at a (single) malfunction• Identifying the most appropriate supporting method (contact or pressing support)• Prototyping an upper table and conducting a test on it in or after fiscal 2017 reflecting
the design conditions, such as the interfaces cell

Present situation

Tests up to FY2016 complete. Verified the basic feasibility of the concept by producing
a test specimen of the device to RPV.

Evaluation and

It is necessary to verify workability (mockup test etc.) assuming specific tasks.

challenges
Overview

A-62

Designation(4) Development of platform/cellObjective

Developing a system that contains the radioactive substances generated from the PCV
and shielding from radiation by using a floor door installed between the cell installed on
the refueling floor and the PCV

Test details• Prototyping a 1/1 scale element of the shielding port
• Checking the action of the opening/closing door
• Checking the airtightness of the seal

Present situation

Tests up to FY2016 complete. By creating prototype of the actual scale of theplatform/cell, it was possible to verify the operation of closing door and shielding door
inside the cell and the airtightness.

Evaluation andInstallation feasibility and remote maintenance performance etc. needs to be studied forchallenges

practical application.

Overview

A-63

Designation(5) Development of handling equipment for fuel debris canister

Objective

Developing equipment that can be remotely controlled to transfer container cans to
the RPV, close the lids, and clean the surfaces of the cans

Testing details• Prototyping a 1/1-scale element of a device for closing the lids of container cans
• Checking the basic action of the bolt tightening mechanism• Checking the appropriate tightening method and procedure• Checking the easiness of removing, disassembling and carrying out for
maintenance of the remote device

Present situation

Tests up to FY2016 complete. It was possible to verify sealing and remote operation
performance with the test specimens for canisters and sealed containers’ closing
mechanism and closing jig.

Evaluation and

Design conditions have been determined for the equipment for handling canisters.

challenges

But, it must be designed by maintaining its consistency with the entire removal
system.

Overview

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Designation(6) Development of light-weight and shape-following shielding

Objective

Developing equipment for facilitating installation and removal of a shield to beinstalled between the refueling floor and PCV in the upper-access method

Test details• Reviewing the required strength through stress analysis based on distortion
simulation• Reviewing the removing and recovering method a installed water-filled shield• Reviewing the shape of the drainage nozzle and drainage pressure conditions

Present situation

Tests up to FY2016 complete. It was possible to verify installation and removal byremote operation and verify the basic feasibility with 1/4 scale model.

Evaluation and

It is necessary to establish an exact method for verification of tasks etc. for on-sitechallenges

application.

Overview

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Designation(7) Development of utilizing films and sheets for contamination spread prevention
methodObjective

Developing equipment for partitioning work areas and preventing radioactive dust
from spreading removal of large structures

Test details• Checking the mechanisms of equipment for preventing contamination fromspreading, large opening/closing equipment, and remote equipment using modelsof an approximately 1/4 scale model of the actual equipment and checking theoperating procedure for the equipment• Checking the airtightness of film and sheet to be used as partitions between areas• Checking the weldability and airtightness of contaminated equipment during cure

Present situation

Tests up to FY2016 complete. Verified the performance by testing characteristic offilms and sheets and verified the work procedure 1/4 scale model.

Evaluation and

It is necessary to establish an installation method that is suitable for on-sitechallenges

application.

Overview

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Designation(8) Development of sealing technology for devices accessing to the inside of the RPV

Objective

Developing equipment for preventing radioactive dust spreading at access unit andalso at the interface between the access unit and RVP, in the in-air upper accessmethodTest details• Checking the sealing performance by partial simulation tests (including tests on
slewing and opening and closing of the port) using a full-scale model• Checking the sealing mechanism inside the RPV• Checking the sealing mechanism at the bottom of the equipment

Present situation

Tests up to FY2016 complete. It was possible to verify the feasibility of basic
prevention of radioactive dust dispersion and certain sealing performance.

Evaluation and

It is necessary to check the impact on performance when combining remote-control

challenges

operation procedures and maintenance method and the in-furnace work equipment
for onsite application.

Schematic
Overview drawing

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Designation(9) Development of device to access inside the pedestal

Objective

Checking the basic feasibility of a series of processes from installation of anaccess rail inside the pedestal to cutting of fuel debris using a robot arm thehorizontal access method

Test details• Checking the positioning accuracy of the robot arm (in with the conditions ofend load of 2 t and an arm length of 6.5 m)• Checking whether a rail can be installed remotely• Checking the accessibility to the inside of the pedestal• Checking the cutting action inside the pedestal

Present situation

Tests up to FY2016 complete It was possible to verify the basic performance ofaccess equipment within pedestal contributing to the design of practicalapplication prototype machine.

Evaluation and challenges

It is necessary to verify the controllability in combination with advanced tool,
proceeding with the verification of installation method of robot arm at on-site andmaintenance method.

Overview

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Designation(10) Development of flexible structure arm for remote-controlled work

Objective

Developing the arm and its ancillary devices for removing obstacles that block fueldebris removal work inside the PCVs for the horizontal access method

Test details• Checking the accessibility, remote operability, and handleability by using mockup
facility simulating the condition of the pedestal inside• Conducting element tests with the in-air horizontal removal method in mind to checkthe applicability of the method under review for removing fuel debris and devices

Present situation

Tests up to FY2016 complete. It was possible to verify the operability of the remote
operation using prototype and handling etc.

Evaluation and

It is necessary to study work methods and maintenance techniques to match thechallenges

condition of the site and confirm the plan for removal of interfering objects within PCV.

Overview

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Designation(11) Development of the PCV remote seal welding equipment for cells

Objective

Developing welding equipment for building boundaries between the cell being installed
at a side of the PCV and the PCV connection in the horizontal access method

Test details• Checking the feasibility of welding for the narrow section simulating the actual working
conditions• Checking the quality of cross sections of welds• Checking the tensile strength of weld samples• Weld performance of 30-m hydraulic head pressure (0.3 MPa)

Present situation

Tests up to FY2016 complete. It was possible to verify the basic design feasibility
necessary for PCV and cell boundary construction.

Evaluation and

In order to apply this on-site, it is necessary to study regarding specific work

challenges

procedures and how to ensure welding quality etc.

Overview

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Appendix 4.15: Risks accompanied by fuel debris retrieval tasks
During fuel debris retrieval, there are risks such as abnormal radioactive material leakage
accompanied by the tasks, hence in order to carry out the task safely, it is important to conduct risk
analysis and have clarity on improvement points in advance, then take them into account ofreviewing methods in future.
Further, the risk here is defined as the influence and likelihood of released radioactive material asa result of events that can be assumed during debris retrieval.

(2)Risk assessment method
Since installations and procedures for fuel debris retrieval are not decided, experts with experiencein TMI-2 accident and decommissioning of Hanford facility evaluated the risk based on theirengineering judgment with the following procedures.
A. Assume an overview of installations and procedures.
B. Determine the types of events that can be expected (refer to (2)), and use the knowledge ofengineering to define the likelihood of their occurrence in 5 stages (historically occurrence israre - not expected to occur even without prevention measures taken - occurs once if
preventive measures are not taken.)
C. Considering the distribution and properties of fuel debris, environmental conditions (on thesurface of the water, submersion) where fuel debris is placed, the release route assumed in
the event, etc. evaluate the dose that the public receives from inhalation and seafood intake
as well as the dose that that the workers receive from inhalation and direct ray respectively
based on released amount and judge in five stages (engineering judgment by experts whenquantification is difficult)((Can be ignored – Cannot be ignored but less than normal dose standard – At normal dosestandard level.)
D. Evaluate risks in five stages based on their likelihood of occurrence and impact.
(2)Events which can be expected (refer Table A4.15-1)
Firstly one of the possible events in which radioactive materials could be released while retrieving
fuel debris are assumed, is a case in which the enclosed contaminated gas and liquid may bereleased . Also, like the event mentioned above, leakage of radioactive dust caused by falling ofheavy objects can be considered.
Besides these, release by criticality or hydrogen explosion can be considered.

(3)Evaluation results and risk reduction.
The result of applying method (1) to the event (2) is shown in Table A4.15-2.
Since the effect of dust dispersion prevention achieved by flooding by partial submersion method
is small, there is an increase in radioactive material concentration in the air which increases therisk as the in impact at the time of contaminated gas outflow increases.

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In submersion method, the PCV water level is increased as part of preparation, fuel debris isflooded and there is a possibility of becoming critical. Also, lot of fuel debris comes in contact
with water which generates hydrogen due to radiolysis and the hydrogen concentration increases
since the gas phase section is small. This leads to an increased risk as it increases the possibilityof a hydrogen explosion.
There are two ways to reduce risks, i.e. reduce the impact and possibility of occurrence. In order
to effectively reduce the risk, it is better to reduce high impact or possibility of occurrence which
is a major cause of the risks. As an example of mitigation measures in the event of outflow ofcontaminated gas, installation of a cutting particles collection system for reducing concentrationin the air and an installation of a filter to a negative pressure control system are considered. As
an example of measures to prevent criticality, installation of sub-critical state monitoring andneutron absorber injections are considered. And as for the measures for preventing hydrogen
explosion, scavenging with inert gas and monitoring the hydrogen concentration are considered.
Further, considering risk in terms of environmental pollution rather than exposure, there is higher
risk in the event of large amount of radioactive materials emission, especially events with large
release of liquid that are more difficult to diffuse than gas. Considering cases in which large
amounts of liquid radioactive material is released into the environment, one possible scenario is
where there is an outflow due to some catastrophic failure such as PCV damage. Especially, incase of submersions method, the risk is higher because the amount of water held in PCV is morethan the partial submersion method.
Characteristics of these risks and countermeasures are generally consistent with Section 4.5"Considering feasibility of fuel debris retrieval methods". Since it is believed that this method canbe used in future to realize methods or verify the safety of preparatory tasks, the study on thiswill be continued.

Table A4.15-1
Event type
Outflow ofcontaminatedgas
Falling ofheavy
objects
Outflow ofcontaminatedliquidExample of specific events

Example of specific events
Incorrect connection of exhaust installations, loss of confinement function due
to erroneous operation, outflow due to exhaust installation failure, outflow due
to disturbance in exhaust balance, outflow caused by installing an opening in
PCV (only for side removal), HEPA filter damage.
PCV shield plug falling, PCV head falling, RPV head falling, dryer or separator
falling, RPV internal structure falling on operating floor, fuel debris falling in
CRD housing due to cutting of the fuel debris in RPVRPV, fuel debris falling on
pedestal due to cutting of the fuel debris in RPV on CRD housing, RPV lower
head falling due to cutting of it to remove pedestal's fuel debris (top accessmethod).)Outflow due to small piping damage, outflow due to large piping damage,outflow due to catastrophic failure such as PCV damage, outflow due totsunami (only for side access method)A-72

Event type
Criticality

Hydrogen
explosion

Example of specific events
Criticality due to cutting fuel debris re-assembly in critical shape inside water,
criticality due to re-assembly of cutting particles after movement, criticality due
to rising water level (only in case of submersion method), criticality caused byan increase in the water level by mistake (only in case of partial submersion
method), criticality caused by gathered fuel pieces
Explosion by hydrogen accumulation due to ventilation airflow blockage,
explosion due to hydrogen release from fuel debris, explosion due tohydrogen accumulation in piping.

Table A4.15-2
Event type
Outflow ofcontaminatedgas (including
dust due tofalling of heavy
weight object)
Outflow ofcontaminatedliquidCriticality

Hydrogen
explosion

Main causes of risk and examples of risk reduction measures

Main causes of risks and examples of risk reduction measures
Since the effect of dust dispersion prevention achieved by flooding issmall with partial submersion method, concentrations in the air increase
and the impact during outflow increases. As a measure to mitigate theimpact for reducing airborne concentration, setting installing cut particles
collection system and setting up filters in negative pressure control
systems are considered.
Even if contaminated liquid in the building or piping containing theradioactive material flows out in the site, measures such as theinstallation of seaside impermeable wall are taken and the possibility ofoutflow to ocean is low. Even if it is assumed that direct ocean outflow is
made, the effect on the public is extremely small due to the diffusion
effect in the ocean.
When PCV water level is raised as preparation for submersions method,
fuel debris is flooded and there is possibility of becoming critical. As apreventive measure, subcritical state monitoring and neutron absorbing
injection are considered.
In submersion method, lot of fuel debris comes in contact with water,which generates hydrogen due to radiolysis and the hydrogenconcentration also increases since the gas phase section is small.
Compared to partial submersion method, there is increased possibility ofa hydrogen explosion. As prevention measures, scavenging with inert gas
and monitoring hydrogen concentration are considered.

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Fig. A5-1 Management status of rubble, etc. and secondary waste from water treatment
Source: Tokyo Electric Power Co., Inc., Document 3-4, "Management Status of Rubble and Felled Trees (as of April 30, 2017)" from the42nd meeting of the Reactor Decommissioning and Contaminated Water Countermeasures Team of May 25, 2017.

Appendix 5: Status of Solid Wastes Management, and Their Storage Management Plan

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Fig. A5-2 Overall schematic of the storage management plan for solid waste from the Fukushima Daiichi nuclear power station.
Source: Tokyo Electric Power Co., Inc., Document 3-4, p. 22, "Tokyo Electric Power Co., Inc. Solid Waste Storage Management Plan from the
Fukushima Daiichi Nuclear Power Station (Revised in June 2017)" (body text) from the 43rd meeting of the Reactor Decommissioning andContaminated Water Countermeasures Team/bureau conference of June 29, 2017.

Appendix 6.1: Interim Report by Research Partnership Task Force(Tentative translation)

November 28, 2016
Research Partnership Task Force

1. Background
The “Technical Strategic Plan for Decommissioning of the Fukushima Daiichi Nuclear Power
Station of Tokyo Electric Power Company 2016” (“Technical Strategic Plan”) points out that it isimportant to create research plans that target whole of the phases from basic/generic research tothe applied research and the practical development by broadening our horizons to include thetechnical development issues from the strategical viewpoint and by widely collecting those basic
and generic technologies that are thought to be necessary in the future.
At present, the main budgetary measures by the Government related to the research anddevelopment to foster the decommissioning of the Fukushima Daiichi Nuclear Power Station (1F)
are the “Project of Decommissioning and Contaminated Water Management” (the “National
Project”) by the Ministry of Economy, Trade and Industry (METI) and the “Center of WorldIntelligence Project for Nuclear S&T and Human Resource Development” (“World Intelligence
Project”) by the Ministry of Education, Culture, Sports, Science and Technology (MEXT).

TheInternational Research Institute for Nuclear Decommissioning (IRID), the Japan Atomic Energy
Agency (JAEA), research institutes such as universities, etc. are carrying out research anddevelopment projects that are thought necessary while Tokyo Electric Power Company (TEPCO)
Holdings is also carrying out their own required research and development.

On the other hand,the 1F decommissioning project, which may continue for 30 to 40 years long, currently has many
uncertain and unclear issues and thus the whole image of the needs for research and development
considering the time base is not clarified yet. Thus, since systematic search for seeds is not easy,
it seems that there are many potential research seeds that are still not investigated.
Therefore, NDF launched a “Research Partnership Task Force” consisting of a small number ofexperts in order to identify, evaluate and prioritize the further research and development issues andneeds on which we should work in a strategic and preferential manner toward decommissioning of
1F, and the task force had discussions.

2. Basic Ideas
2-1. Necessity of understanding of principles and investigations based on theories
The technically difficult issues include those that can be cleared by application of the existing
technologies only if time and cost are spent and those that can be cleared by understanding of theuniversal principles lying in the backgrounds of the phenomena and by investigation based on thetheories.

In the future device development, etc., we should stand on these scientific investigations

for achievement of the spec targets, etc. in order to firmly drive the development without having

A-76critical problems and large delays. In other words, in order to foster the 1F decommissioning
project that will last for 30-40 years long, it is required to plan a mid-and-long-term research anddevelopment strategy including understanding of the principles and scientific investigations of thetheories as above mentioned.
For the basic researches related to the 1F decommissioning project, on the other hand, the project
configuration is based on consciousness of the needs, such as participation of the NDF in the
proposal selection and the interim evaluation.

To prevent the selected tasks from configuring only

by following the conventional research themes on the side of the seeds (adopted organizations), it
is expected to produce achievements that are more conscious of the applicability to the 1F
decommissioning site.

2-2. Clarification of needs and search for seeds
The 1F decommissioning project takes 1F only as its target to which the technologies are appliedbut on the other hand, it is a comprehensive engineering derived from a variety of scientific andengineering technologies and knowledge including nuclear physics, chemistry, machinery,architecture, physical properties and measurement. The potentiality of the seeds is unlimitedly
open. The task to find the prospective seeds from them requires wide knowledge, critically highjudgment, and a great deal of labor.
Therefore, it is the most certain and efficient way that some experts who are familiar with both thesite condition of the 1F decommissioning project and the responses to the reactors damaged by
the accident firstly extract various scientific and engineering issues (hidden needs) that may
become critical in the future in the 1F decommissioning project from the medium-to-long term
viewpoint, and then proceed to search for the seeds based on those issues.

3. Methods
The Platform of Basic Research for Decommissioning (secretariat: Collaborative Laboratories for
Advanced Decommissioning Science (CLADS), JAEA) has been breaking down and systematicallyorganizing the needs while creating an R&D map to clarify the relationship between the needs andthe seeds.

These activities correspond to the above-mentioned idea and therefore, it seems

appropriate to continue our actions by using the Platform of Basic Research for Decommissioning
as the basis.
The Platform of Basic Research for Decommissioning should found a task subcommittee(provisional name) consisting of experts from both the needs side including TEPCO and the seeds
side including universities, etc. for Essential R&D Themes extracted as an issue that could becritical in the future in the progress of the decommissioning project. It should be investigated mainly
by comprehensive system integrators or candidates who can be a communication hub between the

A-77both needs and seeds sides with a panoramic perspective and integrate the technical seeds andbring them to the practical phases.
This R&D strategy should suppose to found a system that forms a COE centered with corepersonnel or core studies and it would continue to engage with the project for a long time in orderto support the long-term 1F decommissioning.

As contents of the strategy, we expect acomposition of concrete core research themes (research scopes) and approaching methods
thereof for application to the 1F decommissioning project, i.e., a comprehensive R&D strategy thatconsiders a research conducting system centered on a COE, research facilities and equipments,
development of young human resources through appointing them as a core personnel andpromoting research, implementing timeline and achievement target that are calculated byconsidering the Mid- to Long-Term Roadmap, the technical strategic plan, and the needs, a desired
way of research evaluation, and a required budget, with consideration on the international situationof the R&D.

In this case, since these are R&D planning that are generally made by researchers,

it is desired that the task subcommittees be operated by R&D institutions, etc. that have special
knowledge of each task field.

In addition, these institutions should collaborate with the NDF and

the Research Partnership Task Force to go on while fully considering the needs.
What is expected to TEPCO are to send persons as the above mentioned experts on the needs
side who are familiar with the 1F decommissioning site and can sufficiently communicate with theexperts on the seeds side from each fields, and to take consideration that these persons will beable to play their roles for a long time as core personnel on the needs side.

4. Essential R&D Themes
Based on the above mentioned points, while considering the logic tree in the technical strategy
plan, etc., the Research Partnership Task Force discussed the essential R&D Themes on which
we should work strategically and preferentially for clarification of principles. As a result, the task
force extracted the 6 themes for now as shown on the table and identified the related technicalseeds.

Note that some other essential issues remain including the remote controlling techniques

and the storage and disposal of fuel debris. Therefore, we will keep extracting the essential R&D
themes.
These Essential R&D Themes have different levels of specification of research scope, different
emergency levels, different timelines of application, and different research periods.

Therefore,

considering these points, the above-mentioned task committees should review the R&D strategy
in a flexible and agile manner.
year (FY2016).

If necessary, NDF should start initial R&D activities within this fiscal

In addition, based on the intent of this interim report, the Governmental andrelated institutions are expected to start or support the R&D actions and to consider a system thatis necessary for carrying out this R&D strategy in a more effective manner.

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6 Essential R&D Themes that should be worked on for the clarification of principles in a strategic
and preferential manner
Essential R&D Themes

Descriptions / Background issues

To identify process ofcharacteristic changes in fueldebris over time

The fuel debris retrieval is scheduled for 2021 onward, 10 years after the fueldebris production. And since it is anticipated that the retrieval will require a long
period of time, the fuel debris will remain inside the reactors over 10 years. We
also need to remember that the retrieved debris must be stored safely. Choosing
the best possible methods of retrieve/transmission/storage of fuel debris requires
predictions of characteristic changes of fuel debris over time.

To elucidate corrosion
mechanisms underunusual/extreme
circumstances

It is required to collect data on corrosion under a variety of circumstances withconsideration of the circumstances specific to 1F decommissioning such as highradiation levels and unsteady routes of cooling water in order to prepare forpotential corrosion during decommissioning.

Radiation measurement
technologies adopting
innovative approaches

The radiation levels are still extremely high inside the 1F reactors/buildings due tothe accident and the existing measurement devices do not meet thecapability/functional requirements to provide accurate figures. It is vital to develop
an innovational device adopting brand-new ideas/principles based on 1F needs.

To clarify behavior ofradioactive particulates
generated duringdecommissioning (incl. alphadust treatment)

As thermal cutting of the fuel debris via machine or Laser may produce a large
amount of alpha dust, it requires safety measures and dust containment solutions.
It is necessary to understand physical/chemical properties of alpha dust, to predict
the amount of dust to be produced for each method, and to consider how to seal
the dust according to the results in order to make sure the retrieval will be
conducted in a safe and effective manner.

To understand fundamentallymechanisms of radioactive
contamination

To figure out the mechanism of radioactive contamination towards effective
decontamination; it is critical to implement effective approaches ofdecontamination based on the mechanism of the contamination to radiation
sources, and to decrease the volume of radioactive wastes as well.

Environmental fate studies ofradioactive materials
generated duringdecommissioning

It is essential to clarify the behavior of radioactive materials such as absorption,
dispersion, moving along with groundwater flow in shallow underground in order
to conduct environmental fate studies to ensure they will not affect theenvironment.

(Note) The “Descriptions/Background issues” describes the contents of the issue that the Research Partnership
Task Force suppose at present. Compositions of research theme (scopes of study) and the methods of approachshould be established in a more concrete manner in the R&D strategy by the task subcommittees.

(Reference 1) Members of Research Partnership Task ForceKoji Okamoto
Professor, Graduate School of Engineering, The University of Tokyo
Toru Ogawa
Director, Collaborative Laboratories for Advanced Decommissioning Science(CLADS), Japan Atomic Energy Agency (JAEA)
Jun Matsumoto
Corporate Officer, Executive Vice President, Chief Technology Officer,
Fukushima Daiichi Decontamination and Decommissioning Engineering
Company, Tokyo Electric Power Company
Hajimu Yamana President, Nuclear Damage Compensation and Decommissioning Facilitation
Corporation (NDF)(In the order of the Japanese syllabary)(Reference 2) Held meetings of Research Partnership Task Force First Meeting: Monday, September 26, 2016
Agenda・ Free discussion about matters of concern with regard to partnership in
research and development for decommissioning
・ Others
 Second Meeting: Monday, October 31, 2016

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Agenda・・・・Report of hearing survey by secretariat
Hearing from experts with regard to important issues
Free discussion
Others Third Meeting: Monday, November 28, 2016
Agenda・ Interim report draft
・ Candidates of Essential R&D Themes to be added in the future
・ Others

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(Source) “The Future System of Professional Engineer” by the Professional Engineer Subcommittee, the Council for Scienceand Technology, dated December 22, 2016
(Tentative translation by NDF)

Appendix 6.2: Changes in elective subjects of secondary examination for
professional engineers (category of atomic energy and radiation)

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Appendix 7: Conditions at TMI-2 and Chernobyl Unit 4
In addition to the Fukushima Daiichi nuclear power station, severe accidents happened incommercial reactors in the world were TMI-2 and the Chernobyl nuclear power station. Differences
between these power stations and the Fukushima Daiichi nuclear power station are summarizedin Table A7-1.
Both power stations differs from the Fukushima Daiichi nuclear power station in points of reactor
type, power output, UO2 amount and progress in the severe accident. However, since fuel debris
retrieval in TMI-2 has already been completed, learning from TMI-2 is a lot for consideration ofprocess, system and safety measures during fuel debris retrieval in the Fukushima Daiichi nuclear
power station. On the other hand, in the Chernobyl nuclear power station, the first molten coreconcrete interaction (MCCI) was happened in world history. It is reported that spontaneous failureof fuel debris has progressed with time and dust has been generating 77 , 78 . Such ageing
phenomenon of fuel debris is also helpful for studying characteristics of fuel debris, safetymeasures. The outline of both power stations up to the present that more than 30 years has passed
from the accident is summarized as following.
TMI, located on a sandbank in the Susquehanna river in Pennsylvania, US, experienced an
accident on its unit 2 on March 28, 197979,80. Due to equipment failure and human error, the primarycoolant in the RPV was released as steam, exposing approximately two-thirds of its core 81 .
Approximately 45% of the core underwent a meltdown, which formed a melt pool at the center,producing fuel debris. Based on a strategy for retrieving the fuel debris, treatment of thecontaminated water that was produced as well as internal PCV condition analyses were carried out
between 1979 and 1984, and a decision was made to employ the method of flooding the RPV withwater and erecting a platform above it to retrieve the fuel debris. Fuel debris retrieval was carriedout from 1985 to 1990. The retrieved fuel debris was shipped to the Idaho National Laboratory (INL),
low level radioactive waste was transported to a commercial treatment facility in Barnwell, andsecondary waste produced during the treatment of contaminated water was transported to Hanford.
Approximately 9,000 tons of tritium water that remained after the elimination of contamination was
treated by evaporation due to a lawsuit prohibiting the release of this water into the river filed by
the city of Lancaster that was located in downstream from the site on the Susquehanna river. The
fuel debris shipped to the INL was wet-stored, after which it was high-temperature vacuum dried,
and is currently being dry-stored on INL premises. TMI unit 2 is currently under post defuelingmonitored storage (PDMS), with dismantling preparations to begin roughly in 2040 in conjunction
with the decommissioning of TMI-1 and reactor decommissioning scheduled to be completed by
205382. In May 2017, however, it is published that decommissioning of TMI Unit 1 will be accelated.
Decommissioning plan of TMI Unit 2 is paid attention.

77“Twenty-five Years after Chornobyl Accident: Safety for the Future” National Report of Ukraine,
Ministry of Ukraine of Emergencies, p. 222 (2011)78
Boris Burakov, “Study of Chernobyl “lava”, corium and hot particles: experience of V. G. Khlopin
Radium Institute(KRI) ”, Handout of the 2nd International Forum on the Decommissioning of the
Fukushima Daiichi Nuclear Power Station, (2017)
79“The Cleanup of Three Mile Island Unit 2 A Technical History: 1979 to 1990”, EPRI NP-6931, (1990)
80“Three Mile Island Accident of 1979 Knowledge Management Digest Recovery and Cleanup”,
U.S.NRC, NUREG/KM-0001, Supplement 1, (2016)
81
Seiji Mizukoshi, and Takeshi Yamanaka, “Study on Decommissioning of Reactor Facilities Damaged
by Accidents”, JNES-RE-2013-2015, Japan Nuclear Energy Safety Organization, (2013)82“Three Mile Island Nuclear Power Station, Unit 2 Post-Shutdown Decommissioning Activities Report
Revision 2”, GPU Nuclear, Docket No.50-320, TMI-15-093, p.12 (2015)

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The Chernobyl nuclear power station is located in the northern part of the Ukrainian Republic (thena part of the Soviet Union), and experienced an accident on April 26, 198683. The reactor's output
increased rapidly, which resulted in a steam explosion, hydrogen explosion and fire. Most of thecore melted, and flowed down to the lower levels of the reactor building. MCCI was caused andfuel debris was produced. Since containment vessel to contain this reactor was not built81, large
amounts of radioactive materials were released into the environment. In order to prevent therelease of radioactive material into the environment, a shelter was built as a containment structure.
However, the shelter eventually got holes due to corrosion and other forms of degradation, and rainwater entered into the shelter through these holes. The site received support from the European
Bank for Reconstruction and Development after the Ukrainian Republic gained independence in1991. Later, as part of the many measures to counter the instability of the shelter, construction of a
new safe confinement84,85 (NSC) with arch-shaped began in 2007. The NSC with reportedly 100
years of durability77,85 was slid to above unit 4 in November 2016, and operations are scheduled to
begin in November 2017. Once operations commence, stabilization measures consisting of theremoval of portions with collapsing risk using the remote controlled crane inside the NSC will be
undertaken through 202386, and this is planned to be followed by fuel debris retrieval77,86, the detailsof which have yet to be decided.
Table A7-1: Comparison of the Fukushima Daiichi NPPs, TMI and Chernobyl NPPs Accidents

Unit 1

Unit 2

Unit 3

Unit 2Chernobyl nuclear
power plant
Unit 4

Reactor type

BWR

BWR

BWR

PWR

RBMK-1000

Power output

460MW

784MW

784MW

959MW

1 000MWOperation dateMar. 26,
1971

Mar. 27,
1976

Dec. 30,
1978

Mar.
1984

Years of operationto accident

40 years

Jul. 18,1974
36 years,
8 months

35 years

3 months

2 years

Average burnup

26GWd/t

23GWd/t

22GWd/t

3-3.7GWd/t

10.3GWd/t

Spent fuel392 rods

615 rods

566 rods

0 rodsHydrogen
explosion

Large scale

None

Large scale

Small scale

Large scale

Fire

None

None

None

None

Yes

RPV / PCV
damage

Yes

Yes

Yes

None

MCCI

Yes

Yes

Yes

None

Yes

Seawater injection

Yes

Yes

Yes

None

None

UO2 amount

77t

107t

107t

94t

190t

Fukushima Daiichi NPS

83

TMI

20 tons(in a separate building)Damage to RPV.
No PCV installed“Environmental Consequences of the Chernobyl Accident and their Remediation: Twenty Years of Experience”,
Report of the Chernobyl Forum Expert Group ‘Environment’, Radiological Assessment Reports Series, IAEA (2006),
ISBN 92-0-114705-8
84 “TRANSFORMING CHERNOBYL”, European Bank for Reconstruction and Development, (2015)
85 “Chernobyl New Safe Confinement: a one-of-a-kind project”(Nov. 29, 2016 press material), EBRD, VINCI,NOVARKA, BOUYGUES and SSEChNPP (2016)
86 “NSC: DEVELOPMENT, STRUCTURAL AND TECHNOLOGICAL FEATURES”, International research andpractical seminar “FROM DESTROYED ChNPP UNIT 4 TO THE NEW SAFE CONFINEMENT”, (Dec 16-18, 2017,
handouts)

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