Technical Strategic Plan 2021 for Decommissioning of
the Fukushima Daiichi Nuclear Power Station of Tokyo
Electric Power Company Holdings, Inc.

October 29, 2021

Nuclear Damage Compensation andDecommissioning Facilitation Corporation

All rights reserved except for reference data. No part of this document may be reproduced,edited, adapted, transmitted, sold, republished, digitalized or used in any other way that violates
the Copyright Act for any purpose whatsoever without the prior written permission of Nuclear
Damage Compensation and Decommissioning Facilitation Corporation.

2

Looking back on NDF’s efforts to the decommissioning of Fukushima
Daiichi Nuclear Power Station
Hajimu Yamana
President
Nuclear Damage Compensation andDecommissioning Facilitation Corporation

March 2021 marked the 10th anniversary of the Fukushima Daiichi Nuclear Power Station
accident. Among the ongoing reconstruction and decontamination efforts led by those in theaffected areas and other stakeholders over the past ten years, action to bring the power station
accident to an end and to decommission the reactors has been continuous. As a result, thesemeasures have been successful at lowering radiation risk, and have allowed work intended todecommission the reactors over the mid- to long-term to proceed steadily.
Established in 2011, the Nuclear Damage Compensation Facilitation Corporation has
conducted research and development, as well as provided advice, guidance, andrecommendations, required for reactor decommissioning following legal reforms executed in2014, and has served as the Nuclear Damage Compensation and Decommissioning Facilitation
Corporation (NDF).
Nevertheless, we must take seriously the fact that many of those from the evacuation orderarea are still living in refuge. Under the fundamental principle of “Balancing between
Reconstruction and Reactor Decommissioning,” NDF believes that in order to proceed intandem with these two efforts and to reassure the people in the areas affected by the disaster,
we must work tirelessly while maintaining a constant awareness of the expectations andconcerns of local communities and society.
NDF has published the Technical Strategic Plan annually since 2015 with the intent ofbroadly sharing a technical understanding of the complex reactor decommissioning processwith stakeholders and the public. The Plan has also served to clarify the technical grounds forassisting the government in formulating its reactor decommissioning policies and the Tokyo
Electric Power Company Holdings, Inc. (TEPCO) in making more solid engineering judgments.
In addition to sharing the latest technical information, the Plan aims to help realize a safe,
reliable reactor decommissioning process in line with the government's Mid-and-Long-term
Roadmap by systematically compiling mid- to long-term issues and strategies. To this end, wedrew up the Plan through exchanges of views among stakeholders, including opinions from theDecommissioning Strategy Committee and the expert committees of this Corporation.

3

We believe that the Technical Strategic Plan has thus far been successful in helping toprogress reactor decommissioning and to realize future plans. Specifically, the Plan has
achieved these by making recommendations for determining fuel debris retrieval policies andfinalizing fuel debris retrieval methods for the first implementing unit, as well as for compilingthe basic concepts concerning solid waste processing/disposal, on the basis that risks caused byradioactive materials are to be successfully reduced in the short-term, as well as over the midto long-term.
In addition to these technical efforts, we have provided advice on overall project
management, thereby enabling TEPCO to properly and reliably proceed with decommissioning
projects through the management of the Reserve Fund for Decommissioning established in2017. Although these efforts have allowed TEPCO to begin enhancing its systems for managing
and implementing projects to achieve its targets, we believe TEPCO must take further steps toenhance its overall capabilities going forward.
At the same time, in order to steadily promote the decommissioning process, which involves
highly challenging issues, over the mid- to long-term, we are striving to promote anunderstanding among both the international and local communities. These efforts include
holding the International Forum on the Decommissioning of the Fukushima Daiichi NPS todisseminate information to Japan and the world and to gather wisdom from around the globe,
as well as holding dialog with local communities.
In the future, we will begin tackling new challenges such as making the shift to full-scale
retrieval of fuel debris, an unprecedented undertaking anywhere in the world. NDF will provide
technical assistance for the trial retrieval of fuel debris and to further expand the scale ofretrieval. In addition, we are committed to contributing to the next Mid-and-Long-term
Roadmap by advancing a review of solutions for issues in anticipation of what lies beyond thestart of fuel debris retrieval. This review will be based on the prospects of processing/disposal
method and technology related to its safety.
Finally, we would like to ask for your ongoing support as all stakeholders continue to work
under the all-Japan framework to make steady progress in their efforts regarding
decommissioning.

4

Table of Contents
1. Introduction ................................................................................................................................... 9
1.1 Structures and systems toward the decommissioning of the Fukushima Daiichi Nuclear
Power Station....................................................................................................................... 10
1.2 The Technical Strategic Plan................................................................................................ 12
1.2.1 Positioning of the Technical Strategic Plan ................................................................... 12
1.2.2 Overall structure of the Technical Strategic Plan 2021 ................................................. 12
2. Concept on risk reduction and safety assurance for decommissioning of the Fukushima Daiichi
NPS............................................................................................................................................ 15
2.1 Basic policy for the decommissioning of the Fukushima Daiichi NPS ................................. 15
2.2 Concept of reducing risks caused by radioactive materials ................................................. 15
2.2.1 Quantitative identification of risks .................................................................................. 15
2.2.2 Risk reduction strategy .................................................................................................. 19
2.2.2.1 Interim targets of the risk reduction strategy ........................................................... 19
2.2.2.2 Progress of risk reduction ....................................................................................... 20
2.2.2.3 Basic approach to risk reduction ............................................................................. 21
2.3 Approach to ensuring safety during decommissioning ........................................................ 23
2.3.1 Basic policy for ensuring safety based on the characteristics of Fukushima Daiichi NPS....................................................................................................................................... 23
2.3.1.1 Promulgating the “Safety First” principle that safety perspective comes first......... 24
2.3.1.2 Optimization of judgement with a safety assessment as its basis and ensuring
timeliness in decommissioning ............................................................................... 25
2.3.1.3 Ensuring safety by incorporating “the operator’s perspective” ............................... 26
2.3.1.4 ALARP judgment based on safety .......................................................................... 26
2.3.2 Preliminary implementation and utilization of the obtained information in the latterstages ............................................................................................................................ 27
2.3.3 Approach to address a temporary increase in risk level associated withdecommissioning operations ........................................................................................ 29
3. Technological strategies toward decommissioning of the Fukushima Daiichi NPS .................. 30
3.1 Fuel debris retrieval .............................................................................................................. 30
3.1.1 Targets and progress ..................................................................................................... 30
3.1.2 Key issues and technical strategies to realize them...................................................... 34
3.1.2.1 Trial retrieval and PCV internal investigation, and gradual expansion of fuel debris
retrieval ................................................................................................................... 35
3.1.2.2 Continuation of accident analysis activities (clarification of events that occurred at
the accident and the process of accident progression) ......................................... 43
3.1.2.3 Technical issues for technical requirements and future plans ................................ 44
3.1.2.4 Summary of key technical issues ............................................................................ 68
3.2 Waste management.............................................................................................................. 70
3.2.1 Targets and progress ..................................................................................................... 70
3.2.1.1 Current status of Storage in Fukushima Daiichi NPS ............................................. 71
3.2.1.2 Prospects of processing/disposal method and technology related to its safety ..... 73
3.2.2 Key issues and technical strategies to realize them...................................................... 78
3.2.2.1 Issues based on the Technical Prospects ............................................................... 78
3.2.2.2 Technical strategy by sectors .................................................................................. 79
3.2.2.3 Summary of key technical issues ............................................................................ 81
3.3 Contaminated and treated water management .................................................................... 83
3.3.1 Targets and progress ..................................................................................................... 83
3.3.2 Key issues and technical strategies to realize them...................................................... 86
3.3.2.1 Issues in the future treatment of stagnant water in buildings ................................. 86
3.3.2.2 Issues of contaminated water management considering the decommissioning
process such as fuel debris retrieval ...................................................................... 88
3.3.2.3 Issues for discharging ALPS-treated water into the ocean ..................................... 89
3.3.2.4 Summary of key technical issues ............................................................................ 94
3.4 Fuel removal from spent fuel pools ...................................................................................... 96

5

3.4.1 Targets and progress ..................................................................................................... 96
3.4.2 Key themes and technical strategies to realize them .................................................... 99
3.4.2.1 Fuel removal from SFPs ......................................................................................... 99
3.4.2.2 Decision of future processing and storage methods ............................................ 101
3.4.2.3 Summary of key technical issues .......................................................................... 102
4. Analysis strategy for promoting decommissioning................................................................... 103
4.1 Uncertainty of fuel debris, etc. and importance of analytical results .................................. 103
4.2 Three elements of analysis strategy................................................................................... 104
4.3 Current status of establishing an analysis system and strategy ........................................ 105
4.4 Improvement of the quality of analysis results, diversification and expansion of analysis
methods ............................................................................................................................. 107
4.4.1 Improvement of the quality of analysis results............................................................. 107
4.4.2 Diversification and expansion of analysis methods in light of sample size and quantity
constraint ..................................................................................................................... 108
5. Efforts to facilitate research and development ......................................................................... 111
5.1 Significance and the current status of research and development .................................... 111
5.2 Key issues and strategies................................................................................................... 113
5.2.1 Updating the R&D medium-and-long term plan........................................................... 113
5.2.2 Initiatives for the Project of Decommissioning and Contaminated Water Management..................................................................................................................................... 113
5.2.2.1 The Project of Decommissioning and Contaminated Water Management ............113
5.2.2.2 Next-term R&D Plan ...............................................................................................114
5.2.2.3 Research and development implementation system for the Project ofDecommissioning and Contaminated Water Management...................................114
5.2.3 Promotion of cooperation between decommissioning sites and universities/ researching
institutions ................................................................................................................... 115
5.2.3.1 The Nuclear Energy Science & Technology and Human Resource Development
Project....................................................................................................................115
5.2.3.2 Collaboration between the Project of Decommissioning and Contaminated Water
Management and the World Intelligence Project, and initiatives for businessacademia collaboration by TEPCO .......................................................................116
5.2.3.3 Establishment of the centers of basic research/research infrastructure................117
6. Activities to support our technical strategy............................................................................... 119
6.1 Further strengthening of project management and improvement of capability required as adecommissioning executor ................................................................................................ 119
6.1.1 Significance and current status of project management ............................................. 119
6.1.2 Key issues and strategies to be strengthened in the future ........................................ 121
6.1.2.1 Safety and Operator’s Perspectives and promulgating the "Safety First ............. 121
6.1.2.2 Owner's engineering capability ............................................................................. 122
6.1.2.3 Securing and developing human resources ......................................................... 127
6.2 Strengthening international cooperation ............................................................................ 131
6.2.1 Significance and the current status of international cooperation................................. 131
6.2.2 Key issues and strategies ............................................................................................ 133
6.2.2.1 Integrating and utilizing wisdom and knowledge from around the world .............. 133
6.2.2.2 Maintaining and developing the international community's understanding of andinterest in decommissioning and cooperative relationships................................. 134
6.3 Local community engagement............................................................................................ 136
6.3.1 Significance and the current status of local community engagement ......................... 136
6.3.1.1 Basic concept ........................................................................................................ 136
6.3.1.2 Specific measures under the current situation...................................................... 136
6.3.2 Key issues and strategies ............................................................................................ 138
List of Acronyms/Glossaries ......................................................................................................... 141
Attachment……………………………………………………………………………...………………..146

6

Table of figures
Fig.1 Division of roles of related organizations responsible for decommissioning of the Fukushima
Daiichi NPS............................................................................................................................ 11
Fig. 2 Positioning of the Technical Strategic Plan based on the Reserve Fund ............................ 12
Fig. 3 Reduction of risks present in the Fukushima Daiichi NPS .................................................. 18
Fig. 4 Example of risk levels assigned to the major risk sources at the Fukushima Daiichi NPS . 19
Fig. 5 Risk reduction process for major risk sources and its progress (example as of March 2021)............................................................................................................................................. 21
Fig. 6 ALARP centered on safety (conceptual diagram) ................................................................ 27
Fig. 7 Estimated fuel debris distribution, access route and surrounding structures of Units 1 to 333
Fig. 8 Image of fuel debris retrieval system ................................................................................... 34
Fig. 9 Image from retrieval to temporary storage of fuel debris ..................................................... 34
Fig. 10 Schematic drawing of isolation chamber to be installed at penetration X-6 ...................... 36
Fig. 11 Schematic drawing of enclosure to penetration X-6 .......................................................... 36
Fig. 12 Enclosure and dual arm manipulator ................................................................................. 39
Fig. 13 Image diagram of the process for narrowing down retrieval methods............................... 42
Fig. 14 Flow of retrieval method study (conceptual diagram) ........................................................ 42
Fig. 15 Example of containment function (gas-phase) by negative pressure control.................... 47
Fig. 16 Build example of containment function (liquid phase) ....................................................... 51
Fig. 17 Example of examining the liquid-phase particle (collection/removal) in the Project ofDecommissioning and Contaminated Water Management ............................................... 64
Fig. 18 Technical issues and future plans on fuel debris retrieval (progress schedule) ................ 69
Fig.19 Summary of waste hierarchy at the NDA, UK, and countermeasures at the Fukushima
Daiichi NPS ......................................................................................................................... 75
Fig. 20 Development of methods to reasonably select safe processing/disposal methods of solid
waste .................................................................................................................................. 78
Fig. 21 Main technical issues and future plans on waste management (progress schedule) ....... 82
Fig. 22 Outline of contaminated water management ..................................................................... 84
Fig. 23 Results of contaminated water generation by factor and main measures to reduce theamount ............................................................................................................................... 87
Fig. 24 Water treatment systems for stagnant water in buildings and measurement results of total
α-nuclides ........................................................................................................................... 88
Fig. 25 Outline of the underwater recovery method for zeolite sandbags ..................................... 91
Fig. 26 Annual discharge of tritium in nuclear facilities in Japan and abroad ................................ 91
Fig. 27 Treatment flow up to discharging waste liquid containing tritium in nuclear facilities........ 92
Fig. 28 Conceptual diagram of the discharge system planned by TEPCO ................................... 95

7

Fig. 29 Key technical issues and future plans on contaminated/treated water management(progress schedule) ............................................................................................................. 97
Fig. 30 State of the collapsed rubble on the Unit 1 operating floor................................................ 97
Fig. 31 Fuel removal method from Unit 1 SFP............................................................................... 98
Fig. 32 Fuel removal method from Unit 2 SFP............................................................................... 99
Fig. 33 Storage status of spent fuel (As of March 2021) ............................................................. 102
Fig. 34 Key technical issues and future plans for fuel removal from SFPs (Progress schedule) 104
Fig. 35 Incorporation of analyses and investigation results, and their relationships ................... 104
Fig. 36 Three elements of the fuel debris analysis strategy ........................................................ 108
Fig. 37 Conceptual diagram from fuel debris analysis to evaluation/countermeasures .............. 109
Fig. 38 Estimation results of the cumulative mass of fuel debris to be retrieved from Unit 3 andsamples to be analyzed ....................................................................................................110
Fig. 39 Example timing of in-situ analysis and non-destructive measurement in the process fromfuel debris retrieval to storage stage.................................................................................112
Fig. 40 Scope of R&D for decommissioning and implementation entities ....................................112
Fig. 41 Overview of the R&D structure of the decommissioning of Fukushima Daiichi NPS Design
flow of fuel debris retrieval (conceptual) .......................................................................... 124
Fig. 42 Design flow of fuel debris retrieval (conceptual)…………………………………………….125
Fig. 43 Annual meeting of NDF with foreign organizations concerned (held online in April 2021)………………………………………………………………………………………………….134
Fig. 44 Structural drawing inside Reactor building…………………………………………………..145
Fig. 45 Structural drawing inside Reactor Pressure Vessel (RPV)………………………………...1458

1. Introduction
The overall approach to the decommissioning of the Fukushima Daiichi Nuclear Power Stationof Tokyo Electric Power Company Holdings, Inc. (hereinafter referred to as the “Fukushima Daiichi
NPS”) started under the Mid-and-Long-term Roadmap1 towards the Decommissioning of TEPCO’s
Fukushima Daiichi NPS Units 1 to 4, released by the Japanese Government in December 2011,
and it has been 10 years in March 2021.
The most urgent issues, such as treating contaminated water and removing fuel from the spent
fuel pools (hereinafter referred to as “fuel removal from SFP”), have been given top priority in thiseffort. However, to complete the decommissioning, long-term measures are required including fueldebris retrieval work, and so it is essential to prepare a mid- and long-term strategy. On August 18,
2014, the former Nuclear Damage Compensation Facilitation Corporation was reorganized into the
Nuclear Damage Compensation and Decommissioning Facilitation Corporation (hereinafterreferred to as “NDF”), a new organization responsible for technical studies needed to proceed withthe decommissioning properly and steadily from the mid- to long-term perspective. NDF’s duties
include, in addition to those assigned to its forerunner, conducting R&D of decommissioning
technologies, and providing advice, guidance and recommendations for ensuring the appropriate
and steady implementation of the decommissioning.
At present, contaminated water management, which required urgent measures immediately after
the accident, have been stabilized, including the reduction of the amount of contaminated water
generated and the completion of the treatment of stagnant water in the buildings (other than thereactor buildings of Units 1-3, the main process building, and the high-temperature incinerator
building). Fuel removal from SFPs has been completed in Units 3 and 4, the decommissioning work
is steadily in progress. In addition, countermeasures against disasters such as earthquakes andtsunamis are progressing. During the earthquake that struck off the coast of Fukushima Prefecture
on February 13, 2021, the water level in the primary containment vessels of Units 1 and 3 dropped,
and the medium- and low-concentration tanks and storage tanks for stagnant water in Units 5 and6 slid (shifted), but As for the ALPS-treated water, the government has announced a policy ofdischarging it into the ocean, on the premise that safety will be ensured and that measures against
reputational damages will be thoroughly taken.
On the other hand, due to the impact of the new coronavirus infection, the fuel debris retrieval
from Unit 2, which was scheduled to start within 2021, is expected to be delayed by about one year,
and efforts are currently being made to minimize this delay. The Tokyo Electric Power Company,
Incorporated (hereinafter referred to as “TEPCO”) has updated its Mid-and-Long-term

1 The first edition was formulated in December 2011 as a way for the government to define the basic concept andmain target processes for promoting the decommissioning and the contaminated water management ofFukushima Daiichi NPS. It has been revised five times based on the progress of the decommissioning and thecontaminated water management, and the latest 5th edition (revised on December 27, 2019) can be found at thefollowing link.
https://www.kantei.go.jp/jp/singi/hairo_osensui/dai4/siryou2.pdf

9Decommissioning Action Plan, which was announced in 2020, in March 2021 to specify the outlook
for decommissioning work based on the above.
The "Technical Strategic Plan 2021 for Decommissioning of TEPCO's Fukushima Daiichi Nuclear
Power Station" (hereinafter referred to as the "Technical Strategic Plan 2021") describes the issues
to be addressed for the trial retrieval of fuel debris to minimize the impact of the new coronavirus
infection, the issues to be discussed for the selection of a method for further expansion of retrieval
scale, and the efforts for the ALPS-treated water, while offering the prospects ofprocessing/disposal method and technology related to its safety (hereinafter referred to as
the ”Technical Prospects”), that was to be presented around FY2021 in the " Mid-and-Long-term
Roadmap for Decommissioning of TEPCO's Fukushima Daiichi Nuclear Power Station" (hereinafterreferred to as the "Mid-and-Long-term Roadmap").
With regard to the impact of the new coronavirus infection, which has been ongoing since
FY2020, TEPCO has established a task force to deal with the new coronavirus infection, and is
implementing measures based on the business continuity plan and measures to prevent infection
and spread2. Although there have been cases of infected employees and workers engaged in thework at the Fukushima Daiichi NPS, the safety of decommissioning work has not been affected.
There have been delays in the start of the trial retrieval of fuel debris and the lack of directcommunication regarding international cooperation and regional symbiosis, but efforts are being
made to minimize these impacts.

1.1 Structures and systems toward the decommissioning of the Fukushima
Daiichi Nuclear Power Station
In order to make systematic and steady progress in addressing issues of decommissioning workover the mid-to-long-term, TEPCO has been working to build and strengthen the project
management structure. In April 2020, though its organization was reorganized, and the
management structure and scheme were established, after this, it is important to enhance andupgrade the management methods and make them effective and rooted in the field operations.
From the financial perspective, the Nuclear Damage Compensation and Decommissioning
Facilitation Corporation has been carrying out the management of a reserve fund fordecommissioning since October 2017 to ensure immediate decommissioning work. The
management task aims are, in every fiscal year, (1) TEPCO will deposit the amount at NDF that is
specified by NDF to implement decommissioning appropriately and steadily, as well as thatapproved by the Minister of Economy, Trade and Industry, and (2) based on the “Withdrawal Plan
for Reserve Fund for Decommissioning” (hereinafter referred to as “Withdrawal Plan”), that was
jointly prepared by NDF and TEPCO and approved by the Minister of METI, TEPCO will withdraw
the reserve fund and implement decommissioning. At present, three years have passed since the

2 TEPCO, “Status of the outbreak of New Coronavirus infection in our group”

https://www.tepco.co.jp/press/news/2021/1612325_8971.html

10decommissioning reserve management was introduced, and it is contributing to the proper andsteady implementation of decommissioning. (Fig.1）
Under this management task, NDF will (1) appropriately manage the fund for decommissioning,(2) manage the implementation structure for proper decommissioning and (3) manage thedecommissioning work based on the Reserve Fund for Decommissioning appropriately, and NDF
assumes responsibility as an organization to manage and oversee TEPCO’s decommissioning
activities. NDF prepared “the Policy for Preparation of Withdrawal Plan for Reserve Fund for
Decommissioning" (hereinafter referred to as “The Policy for Preparation of Withdrawal Plan”),
which was drawn up based on the “Technical Strategic Plan”, and presented to TEPCO the workgoals and main activities to be incorporated in the Withdrawal Plan, and evaluated theappropriateness of TEPCO's efforts in the process of jointly preparing the Withdrawal Plan fromthe perspective of symbiosis and communication with the community, etc. (Fig. 2).
In addition to the operation of these systems, the division of roles among the organizations
directly involved in the decommissioning of the Fukushima Daiichi NPS, including the Japanese
government, NDF and TEPCO, as well as organizations specializing in R&D, such as theInternational Research Institute for Nuclear Decommissioning [IRID] and the Japan Atomic Energy
Agency [JAEA], is shown in Fig.1, which also indicates how the abovementioned systems are
implemented. Among these roles, R&D are discussed in Chapter 5, and dialogue with local
residents and communities is described in Chapter 6.
Nuclear Emergency Response Headquarters (Director: Prime Minister)Council for the Decommissioning of TEPCO's Fukushima Daiichi Nuclear Power Station*1(Chairperson: Chief Cabinet Secretary)

Mid-and-Long-Term Roadmap

The Team for Countermeasures for Decommissioning
and Contaminated Water Treatment(Head of Team: Minister of Economy, Trade and Industry) Progress management of each countermeasure according toMid-and-Long-Term Roadmap

Progressmanagement

Granting of
project budget

Apply forapproval

Approve

Nuclear Damage Compensation and Decommissioning Facilitation Corporation

 Development of a mid-to-long-term strategy

Withdrawal plan

Technical  Tech. support for progress management of important
issues (in collaboration)
Strategic  R&D planning and progress management

plan

International Research Institute for
Nuclear Decommissioning (IRID)*2
Share progress
Japan Atomic Energy Agency (JAEA), and challengesetc.
 R&D implementation
Report Strengthen international partnershipReserve Fund for Management of Reserve Fund for
Decommissioning
Decommissioning
Advice, guidance, recommendations,supervision and support
Reserverelated to project management, etc.

Disburse-ment

Apply for report

R&D organization

Mid-to-Long-Term risk map

Minister of Economy, Trade and Industry
Presentation ofimportant issues

ReportNuclear Regulation Authority
 Approval of implementation
plan, pre-service inspection,
welding inspection, etc.
 Providing goals fordecommissioning

Monitoring, examination

Report on theoutcomeReportBidirectional dialogue

Local residents and communities

Organizations related to decommissioning Decide the direction of decommissioning and contaminated water
treatment

Tokyo Electric Power Company Holdings, Inc. (Fukushima Daiichi Decontamination and Decommissioning Engineering Company)
 Contaminated water and ALPS-treated water management, fuel removal from SFP, fuel debris retrieval, waste management, entire sitemanagement, etc.
Mid-and-Long-term
 Through TEPCO's management reform, the funds required for decommissioning should be reserved
Decommissioning Action Plan
responsibly (Implementation of the 4th special business plan) Detailed measures for facility safety and protection of nuclear fuel material (Development of implementation plan for specified nuclear facility*1 In response to the ALPS-treated water disposal policy decided on April 13, 2021, “Council for the Decommissioning of TTEPCO’s Fukushima Daiichi NPS toward steady implementation of basic policy on ALPS-treatedwater disposal” was founded.

*2 TEPCO, a decommissioning project operator, participates as a member of IRID and shares the needs, challenges, and results of research and development.

Fig.1 Division of roles of related organizations responsible for decommissioning of the Fukushima
Daiichi NPS

11Government

Mid-and-Long-Term Roadmap*
Proposals for Revise* Continually revised based on the progress of programs,
in the past revised once a couple of years
Mid-and Long-term R&D Plan, Next R&D Plan
(The Government-led R&D program on
Decommissioning and Contaminated Water Management)

NDF

Technical Strategic Plan (annual)
The Policy for Preparation ofWithdrawal Plan for Reserve Fund for
Decommissioning (annual)

Feedback

TEPCO

Withdrawal Plan for Reserve Fundfor Decommissioning (annual)

Planning from
Mid-and-long-term aspect

Jointly created with NDF and approved by METI
• Contaminated water management PG• Spent Fuel removal PG• Fuel debris retrieval PG• Waste management PG• General management of the site PG• Facility maintenance/Operational Management

Mid-and-Long-term
Decommissioning Action Plan(annual)On-site work

Engineering

Feedback

R&D

Promotion of Decommissioning project

Fig. 2 Positioning of the Technical Strategic Plan based on the Reserve Fund

1.2 The Technical Strategic Plan
1.2.1 Positioning of the Technical Strategic Plan
NDF has been compiling the Technical Strategic Plan every year since 2015 for the purpose ofproviding a solid technical basis for the Mid-and-Long-term Roadmap, contributing to its smoothand steady implementation, consideration of revisions, achievement of the goals of the Nuclear
Regulation Authority (NRA)'s "Target Map for Reducing Medium-Term Risk at TEPCO's Fukushima
Daiichi NPS," and providing a basis for the Policy for Preparation of Withdrawal Plan (Attachment
1). The Technical Strategic Plan makes suggestions that contribute to the annual revision of the
Mid-and-Long-term Decommissioning Action Plan from a technical perspective.
The Technical Strategic Plan 2021 presents a technical strategy from a medium-to-long-term
perspective that overlooks the overall efforts at the Fukushima Daiichi NPS in order for project
operators to steadily implement decommissioning work toward achieving the goals of the Mid-andLong-term Roadmap that was revised in 2019. In particular, since fuel debris retrieval, which is ahighly difficult task, is approaching, and the roles of the government, NDF, TEPCO, research
institutes, etc. are becoming more significant to realize this task. The Technical Strategic Plan
designed with these perspectives in mind as well.
1.2.2 Overall structure of the Technical Strategic Plan 2021
The Technical Strategic Plan 2021 consists of six chapters.

12

Chapter 1 (Introduction) describes the current situation 10 years after the accident and theprogress of the efforts to realize the milestones of the Mid-and-Long-term Roadmap, as well as themain points of the Technical Strategic Plan 2021.
Chapter 2 (Concept on risk reduction and safety assurance for decommissioning of the
Fukushima Daiichi NPS) presents the basic policy as the approach to risk reduction and ensuringsafety. In implementing risk reduction strategies, it also describes the immediate targets, basic
approach to risk reduction, and the approach on the order of priority, as well as the approach toensuring safety, including the basic policy of ensuring safety based on the characteristics of the
Fukushima Daiichi NPS, incorporating the safety perspective and the operator’s perspective, andthe preliminary implementation and utilization of the information obtained in the later stages.
In Chapter 3 (Technological strategy toward decommissioning of the Fukushima Daiichi NPS),
sector-specific goals are set for each of the four areas of fuel debris retrieval, waste management,
contaminated and treated water management, and fuel removal from SFP.
Section 3.1 of Chapter 3 (Fuel debris retrieval) describes the issues to be addressed for the trial
retrieval to minimize the effects of the new coronavirus infection and the gradual expansion of theretrieval scale, as well as the efforts to study the retrieval method for further expansion of theretrieval scale.
In Section 3.2 of Chapter 3 (Waste management), in accordance with the basic concept of solid
waste processing/disposal, in addition to the Technical Prospects presented in the TechnicalStrategic Plan 2021, issues based on the achieved status and matters to be implemented thereafter
are described.
Section 3.3 of Chapter 3 (Contaminated water and treated-water management) describes theinitiatives and direction of countermeasures for the newly detected relatively high total alpha nuclide,
reinforcement and renewal of the entire water treatment system from the perspective of service life
and aging, and issues for discharge of ALPS-treated water into the ocean.
Section 3.4 of Chapter 3 (Fuel removal from spent fuel pools) describes the new goals for fuelremoval from SFPs and the appropriate and specific work plan depending on the situation of each
unit. This section also describes the direction of efforts to determine the future processing andstorage methods, such as securing the necessary capacity to properly store the removed fuel onsite and evaluating the long-term integrity of the fuel in SFP.
In Chapter 4 (Analysis strategy for promoting decommissioning) shows, as the significance andsystem of analysis, the importance of building and maintaining analytical facilities and functions
required for handling waste and fuel debris, and establishing a system including human resource
development, as well as the characteristics of fuel debris, the policy for reducing uncertainty andthe issues of sample analysis and the policy for handling them.
Chapter 5 (Efforts to facilitate research and development) describes the individual research anddevelopment described in Chapters 3 and 4 with a view to the medium and long term as a whole,and summarizes the approaches expected of the government, project operators, and related
research organizations. This chapter also describes the importance of ensuring the continuity of13research and development and accessibility to past results after the termination of IRID, and theefforts to strengthen the system for these purposes.
Chapter 6 (Activities to support our technology strategy) describes the significance, current status,and major issues and strategies in the following areas: further enhancement of project management,
improvement of capabilities as a decommissioning operator, enhancement of international
collaboration, and regional symbiosis.
Section 6.1 of Chapter 6 (Further strengthening of project management and improvement ofcapability required as a decommissioning executor) describes the need for constant awareness tocomply with corporate ethics in response to the problem of non-conformity with the nuclear
materials physical protection program of at the Kashiwazaki-Kariwa Nuclear Power Station and theimportance of insourcing of technology through decommissioning work.
Section 6.2 of Chapter 6 (Strengthening of international cooperation) describes the necessity ofstrengthening international collaboration, such as strengthening partnerships with relevantorganizations in each country that are engaged in decommissioning in overseas legacy sites, inorder to concentrate wisdom in Japan and abroad, and the current status of efforts affected by the
new coronavirus infection.
Section 6.3 of Chapter 6 (Local community engagement) describes the approach to be taken
when related organizations, mainly TEPCO, work in cooperation with each other based on theprinciple of "Balancing between reconstruction and decommissioning" in order to continue
decommissioning the Fukushima Daiichi NPS for a long time. This report describes the approach
to be taken by TEPCO and other related organizations when they work together.

142. Concept on risk reduction and safety assurance fordecommissioning of the Fukushima Daiichi NPS
2.1 Basic policy for the decommissioning of the Fukushima Daiichi NPS<Basic policy for the decommissioning of the Fukushima Daiichi NPS>
Continuously and quickly reduce the risks arising from the radioactive materials caused by theaccident that do not exist in normal nuclear power plants
The Fukushima Daiichi NPS has the necessary safety measures required by the NRA in place
for the matters for which measures should be taken and it is being maintained in a state with acertain level of stability.
However, there are still enormous risks at the Fukushima Daiichi NPS because fuel debris andspent fuel still remain in the reactor buildings damaged by the accident, part of the status of the
NPS has not yet been sufficiently ascertained, and the site has radioactive contaminated water andenormous amounts of extraordinary radioactive wastes. If left unaddressed, these risks may
increase due to aging degradation of the facilities and other factors. Quickly and swiftly reducing
these risks is an urgent matter for the NPS.
Accordingly, the basic policy for the decommissioning of the Fukushima Daiichi NPS is “tocontinuously and quickly reduce the risks arising from the radioactive materials caused by theaccident that do not exist in normal nuclear power plants” by taking measures specifically designed
to reduce risks. Generally, the following measures are effective for reducing risks at facilities wherean accident has occurred; (1) Improving the containment functions of the damaged facilities; (2)Changing the properties and form of the contained radioactive materials to be more stable; and (3)Strengthening monitoring and control over the equipment to better prevent or mitigate theoccurrence or propagation of abnormalities. In order to achieve these measures in an integrated
way, in addition, (4) Collecting radioactive materials from the damaged facilities or insufficient
containment conditions and placing them in more robust storage is effective.
Since the accident, these diverse measures for risk reduction have been taken with careful
preparations aimed at preventing accidents and radioactive exposure of workers (Attachment 2).

2.2 Concept of reducing risks caused by radioactive materials
2.2.1 Quantitative identification of risks
The term “risk” has various meanings depending on the field or situation in which it is used. In
general, in the context of appropriate risk management, “risk” can be understood as an expectation
value of the negative impact of an event. In other words, the magnitude of a risk (risk level) posedby a subject (risk source) can be expressed as the product of the level of impact and the likelihoodof occurrence of an event.
The Technical Strategic Plan uses a method based on the Safety and Environmental Detrimentscore (hereinafter referred to as “SED”) developed by the Nuclear Decommissioning Authority15(hereinafter referred to as “NDA”) to express the magnitude of risk (risk level) for radioactive
materials. The risk level expressed by SED is given by the calculation formula below.
Risk Level expressed by SED = “Hazard Potential” × “Safety Management”“Hazard Potential” here, is an index of the impact of the event, namely, the impact of internal
exposure in the event of human intake of radioactive material contained in the risk source. It canbe expressed as the product of Inventory, which is the amount of radioactive material contained in
the risk source (taking account of toxicity of the radioactive material), and factors that depend on
the form of the risk source and the time allowable until the manifestation of the risk. “Safety
Management” is an index of the likelihood that an event will occur. It is determined by factors that
depend on the integrity and other aspects of the facility and on the packaging/monitoring status of
the risk source (Attachment 3).
The major risk sources of the Fukushima Daiichi NPS are summarized in Table 1, and Fig. 3
shows the risks of the Fukushima Daiichi NPS as the sum of these risk sources. The current risk
levels assigned to the respective risk sources are expressed in Fig. 4 with “Hazard Potential” and“Safety Management” as the axes.
In the Mid-and-Long-term Roadmap, management of these risk sources is broadly classified
into three major categories: (1) Relatively high risks given high priority (stagnant water in buildings
and fuel in SFPs), (2) Immediate risk unlikely, but risk may grow when handling with haste (fueldebris), and (3) Increased risk unlikely in the future, but appropriate decommissioning efforts are
required (solid waste such as sludge generated by the decontamination device). Their priorities areset, and appropriate measures are being taken. In Fig.4, above (1) is represented in pink, (2) inyellow, and (3) in green, with the risk sources in the “sufficiently stable management” region (inpale blue area) are shown in light blue.
Major risk sources identified at the Fukushima Daiichi NPS are shown in Table 1. In addition,
the overall decommissioning work over the long term includes waste that existed before theaccident and the risk sources that have low hazard potential but are not adequately controlled in astable manner. Since the Technical Strategic Plan 2019, these issues have also been presented.
In particular, regarding the facilities containing risk sources that were not expressly considered
before, investigations and examinations are being conducted in consideration of external events
such as earthquakes, tsunamis, and rainwater (Attachment 4).
It is also important to identify risks that have not been anticipated before. Although it is not easy
to identify such risks, when an unexpected event occurs, analyzing the event to clarify causes that
had not been anticipated before provides a clue for risk identification.
At the event of total-β contamination leakage in the rubble temporary storage area reported on
March 25, 20213, leakage of radioactive materials occurred from a container whose contents were3

Tokyo Electric Power Company Holdings Inc., “Report on the accident event of the drainage at the shallow draft
wharf and storage of rubble”, Study group on monitoring and assessment of specified nuclear facilities meeting(90th), Material 4, April 19, 202116

not identified. Before this event, it was assumed that solid content such as rubble would not
immediately transfer radioactive materials to the environment due to container damage. In light ofthis event, however, it is important for risk identification to understand physicochemical state andits changes over time, in addition to the location of the risk sources and radioactivity. At the time ofthe earthquake on February 13, 2021, with its epicenter off the coast of Fukushima Prefecture4,
lowering of the PCV water levels at Units 1 and 3, and sliding of tanks on site exceeding the sliding
amount evaluated at the time of tank installation were observed. For the PCVs for which the current
state is not well-understood, understanding the damage condition by internal investigation andassessment of the situation at the accident, and estimation of aging by monitoring/evaluation areuseful for risk identification. Regarding external events such as natural disasters, it is necessary tothoroughly evaluate in advance the consequences of and the necessity of countermeasuresagainst beyond-design-basis events in existing/new systems.
Although none of the above events resulted in significant consequences, it is important tocarefully analyze the events using methods such as root cause analyses, and to identify risks that
had not been anticipated in order to help prevent the occurrence of significant consequences. For
this purpose, TEPCO needs to make efforts to learn from the unexpected events as describedabove.
Table 1 Major risk sources at the Fukushima Daiichi NPS
Fuel debris
Spent fuel

Fuel in SFPs

Contaminatedwater, etc.

Fuel in the Common
Spent Fuel Storage Pool
Fuel in dry casks
Stagnant water inbuildings

Zeolite sandbags

Secondarywaste
generated bywater
treatment

Stored water in weldedtanks
Residual water inflanged tanks
Waste sorption vessels,etc.

ALPS slurry

Sludge generated bydecontamination device
Concentrated waste
liquid, etc.

4

Fuel debris in RPVs/PCVs in Units 1 to 3
Fuel assemblies stored in the spent fuel pools (SFPs) in Units 1
and 2
Fuel assemblies stored in the Common Spent Fuel Storage Pool
Fuel assemblies stored in dry casks
Contaminated water accumulated in the reactor buildings of Units
1 to 3, process main building and high-temperature incinerator
building, and sludge containing α-nuclides at the bottom ofbuildings of Units 1 to 3
Sandbags containing zeolite placed on the basement floors of theprocess main building and high-temperature incinerator building
Strontium-treated water and ALPS-treated water, etc. (ALPStreated water and water under treatment) stored in welded tanks
Concentrated saltwater and sludge containing α-nuclides left at
the bottom of flanged tanks
Spent sorbents used in a cesium sorption apparatus, a second
cesium sorption apparatus, a third cesium sorption apparatus,
advanced multi-nuclide removal equipment, mobile-type strontium
removal equipment, a second mobile-type strontium removal
equipment and mobile-type treatment equipment, etc.
Slurry produced during treatment by the multi-nuclide removal
equipment and added multi-nuclide removal equipment, andstored in high integrity containers (HIC)
Flocculated sludge generated during the operation of thedecontamination system
Concentrated waste liquid generated by evaporative concentrationof concentrated salt water with further volume reduction by
concentration, and carbonate slurry collected from theconcentrated waste liquidTokyo Electric Power Company Holdings Inc., “Additional system inspection and seismic evaluation in responseto the earthquake on February 13 at the Fukushima Daiichi Nuclear Power Station”, Study group on monitoring
and assessment of specified nuclear facilities meeting (90th), Material 5-1-3, April 19, 2021

17

Rubble, etc.

Solid waste storage
facility
Soil-covered temporarystorage, etc.
Outdoor storage, etc.

Contaminated structures, etc., in thebuildings*1*2

Rubble (30 mSv/h and above) stored in the solid waste storage
facility
Rubble stored in the soil-covered temporary storage facility andcontainers (1-30 mSv/h), felled trees stored in the temporarystorage pool
Rubble stored in outdoor sheet-covered storage (0.1-1 mSv/h),
rubble stored in outdoor storage (below 0.1 mSv/h), felled trees
stored in outdoor storage
Structures, pipes, components, and other items (shield plugs,
standby gas treatment system pipes, etc.) inside the reactor
buildings and PCVs/RPVs that are contaminated with radioactive
materials dispersed due to the accident; and activated materials
generated from operation before the accident

The risk level was high due to fuel debris right after the accident, however, it became significantly lower
because the hazard potential was decreased a lot by attenuation of the radioactive materials inside thefuel debris during the one year after the accident.
In the evaluation eight years after the accident, as a result of incorporating the insight that the rise in thewater temperature after cooling shutdown was slower than expected, the risk associated with fuel in
SFPs is lower than previously estimated, because the time margin before the risk becomes apparent isincreased.

Fig. 3 Reduction of risks present in the Fukushima Daiichi NPS

18

Color-code for therisk sources basedon the Mid-andLong-term Roadmap** Risk sources that are “relatively high risks given high priority” are shown in pink, those that are “immediate risk unlikely, butrisk may grow when handling with haste” are shown in yellow, those that are “increased risk unlikely in the future, but appropriate
decommissioning efforts are required” are shown in green, and those that are in the “sufficiently stable management” region are
shown in light blue.
The red letters present risk sources that have changed significantly since last year, and the dotted lines or the starting points of
the arrows indicate the locations in the previous year.

Fig. 4 Example of risk levels assigned to the major risk sources at the Fukushima Daiichi NPS
2.2.2 Risk reduction strategy
2.2.2.1 Interim targets of the risk reduction strategy
Measures for risk reduction include the reduction of the “Hazard Potential” and the reduction of
the “Safety Management” level. Examples of reduction of the “Hazard Potential” include thedecrease in inventory and decay heat associated with radioactive decay, and changing the form ofliquid and gas into a less moveable form. Processing contaminated water to change it into
secondary waste is an example of form change.
Examples of reduction of the “Safety Management” level include transferring fuel in SFPs to the
Common Spent Fuel Storage Pool, and placing rubble stored outdoors into storage. Of the various
risk reduction measures, reduction of the “Safety Management” level is generally considered to beeasily realized. Consequently, the decommissioning of the Fukushima Daiichi NPS, which is
implemented under the basic policy of “to continuously and quickly reduce the risks arising fromthe radioactive materials caused by the accident and that do not exist in normal nuclear power
plants” (refer to Section 2.1), should first focus on steadily managing risk sources by keeping themin higher-integrity facilities to lower their Safety Management levels. The interim target of the risk
reduction strategy is to bring the risk levels into the “Sufficiently stable management” region (thepale blue area) as shown in Fig. 4.

19

Regarding the progress from the Technical Strategic Plan 2020 towards this goal, Fig. 4 shows
the completion of transfer of the fuel from Unit 3 SFP to the Common Spent Fuel Storage Pool in
February 20215; the completion of treatment of the remaining water (ALPS-treated water, etc.) at
the bottom of the flanged tank in July 2020 (transfer to welded tanks)6; and the decrease of thestagnant water in buildings (transfer to sorption vessels, etc.). In addition, Table 1 specifies that theshield plugs and the piping of the standby gas treatment system, whose state have been verified
through field investigations by the NRA7, are included in the contaminated structures, etc., in thebuildings.
In considering the station-wide risk reduction strategy for the Fukushima Daiichi NPS, the abovementioned SED is a quantitative indicator of risks attributable to radioactive materials at a certain
time, and is an effective method for prioritizing risk sources for risk reduction.
In response to the “Direction of risk consideration for external events” described in the TechnicalStrategic Plan 2020, TEPCO is proceeding with studies related to natural disasters, including
measures against tsunami and large-scale rainfall, and building integrity evaluations, as stated in
the Mid-and-Long-term Decommissioning Action Plan 20218. TEPCO should continue verificationon the system and facility integrity against external events including natural disasters, and develop
their actions commensurate with the degree of risk.
2.2.2.2 Progress of risk reduction
Fig. 5 shows the process to bring major risk sources into the “Sufficiently stable management”
region as the interim goal, and the decommissioning work progress along this process. Fig. 5(a)shows the outline flow of the decommissioning work to date and the future plans to represent theoverall decommissioning process in a comprehensive way. Using the coloring in Fig. 4 to indicate
the risk level of each risk source, Fig. 5(a) also shows the flow of risk reduction. Based on this flow,
it is possible to visualize how the risk sources have changed compared with the time of the accident
by applying it to fuel debris, spent fuel, and Cs-137 released during the accident. The number ofspent fuel assemblies as an indicator to make the work progress easier to see in Fig. 5(b), and for
Cs-137, the estimated radioactivity (Bq) common to various risk sources as an indicator in Fig. 5(c),both indicate the progress of the decommissioning work by representing the status of transition to
the “Sufficiently stable management” region in a pie chart format.

5Tokyo Electric Power Company Holdings Inc., “Completion of fuel removal at Unit 3”, at the 88th Meeting of theSecretariat of the Team for Countermeasures for Decommissioning and Contaminated Water Treatment, March
25, 2021
6 Tokyo Electric Power Company Holdings Inc., “Status of storage/treatment of the accumulated water containing
high-concentration radioactive materials at the Fukushima Daiichi Nuclear Power Station (459th report), July 13,
2020
7 The 19th Study Committee on Accident Analysis of the Fukushima Daiichi Nuclear Power Station, Material 4,“Interim report on investigation/analysis of the accident at the TEPCO Fukushima Daiichi NPS (draft)”, March 5,
2021
8 Tokyo Electric Power Company Holdings Inc., “Mid-and-Long-term Decommissioning Action Plan 2021”, March
25, 2021

20

(a) Risk reduction process(b) Number of fuel assemblies of spent fuel(Units 1 to 4)(c) Radioactivity of Cs-137 released at the accident (Units
1 to 3)Fig. 5 Risk reduction process for major risk sources and its progress (example as of March 2021)(Attenuation in Fig. 5(c) takes into account radioactive decay of Cs-137 from the time of theaccident to the end of March 2021)

2.2.2.3 Basic approach to risk reduction
The decommissioning of the Fukushima Daiichi NPS is a project that involves considerable
uncertainties. To date, the internal status of the Primary Containment Vessels (hereinafter referred
to as “PCVs”) of Units 1 to 3 has been estimated to some extent through simulation of the accident
development process, estimation of the places with fuel debris by muon-based fuel debris detection
technology, placement of investigation equipment into the PCVs, radiation dose measurement andvideo photographing in the buildings, and other means. There are however still significant
uncertainties. Eliminating these uncertainties requires many resources and, in particular, a21

considerable amount of time. In order to realize prompt reduction of risk, it is necessary to promote
the decommissioning work through a flexible and prompt approach, based on the directions
determined with previously obtained experience and knowledge, and with experiment and analysisbased simulation, making safety the top priority, even though uncertainties remain to a certain
extent.
Regarding the perspective from which these comprehensive decisions will be made, NDF
summarizes the following five guiding principles:(Five guiding principles) SafeReduce the risks posed by radioactive materials and ensure work safety(Issues such as containment of radioactive materials [environmental impact],
exposure of workers to radiation, assessment of the effect of risk reduction)
 Proven

Highly reliable and flexible technologies(Issues such as conformity to requirements, effectiveness and flexibility
against uncertainty)

 Efficient

Use resources effectively (e.g., people, things, money and space)(Issues such as reduction of waste generation, cost, efficiency, securing
necessary work area and site) Timely

Be conscious of time
(Issues such as the early start of fuel debris retrieval and estimation of timerequired for fuel debris retrieval) Field-oriented Comprehensive three-reality policy by checking actual site, actual things,
and actual situation(Issues such as workability including environment-friendliness, accessibility,
and operability, and maintainability including ease of maintenance andtroubleshooting)
In applying the five guiding principles to the actual site, it is important to proceed with thedecommissioning operation after greatly emphasizing safety assurance for the purpose ofprotecting human beings and the environment from the radioactive materials associated with theoperations, thoroughly conducting radiological impact evaluations, and taking appropriate
radioprotective measures. (“Safe” in the five guiding principles)
In the decommissioning of the Fukushima Daiichi NPS, because the public risk level is risingwith time as the degradation of facilities damaged by the accident progresses, controlling this risk
to be as low as reasonably achievable (“Efficient”) as promptly as possible (“Timely”) in light of thesituation at the site, and proceeding with the decommissioning in a reliable manner (“Proven”) by
feasible ways in the harshest on-site state (“Field-oriented”) will lead to ensuring safety in themedium-to-long-term.
As for the result judged based on these guiding principles, it is also important to work todisseminate information carefully so that the results will be widely accepted by society.

222.3 Approach to ensuring safety during decommissioning
2.3.1 Basic policy for ensuring safety based on the characteristics of Fukushima Daiichi
NPS
Decommissioning of the Fukushima Daiichi NPS containing the reactors involved in the accident
is an unprecedented activity that takes place in a peculiar environment different from that of anormal reactor, and therefore, to ensure safety, the following characteristics (peculiarities)
regarding safety should be fully recognized:•

A large amount of radioactive material (including α-nuclides that have a significant impact
in internal exposure) is in an unsealed state, as well as in unusual (atypical) and various
forms

•

Barriers for containing radioactive materials, such as reactor buildings and PCVs, areincomplete•

Significant uncertainties exist regarding the state of these radioactive materials andcontainment barriers, etc.

•

Difficulty in accessing the site and installing instrumentation devices to obtain on-site
information due to constraints such as high radiation levels on site•

Since the current level of radiation is high and further degradation of containment barriers
is a concern, it is necessary to take measures in consideration of the time axis without
prolonging the decommissioning activities

Consequently, TEPCO, as the operator of the decommissioning project, needs to pay special
attention to the following points in proceeding with the decommissioning based on the five guiding
principles.
Firstly, with regard to “safety”: There is great uncertainty about the state of radioactive materials
and containment barriers, and on-site access and installation of instrumentation devices to reduce
the uncertainty are also restricted. Under these circumstances, a large amount of atypical andunsealed radioactive material will be handled in an incomplete state of containment. Therefore, thestarting point for all reviews should be confirmation of the feasibility of ensuring safety with a wide
range of possibilities (cases) assumed. At the same time, with regard to “safety”, it is important not
to prolong the work period in light of risk reduction over the entire work period. Therefore, it isnecessary to avoid excessive safety measures and to take optimum safety measures (ALARP9).
Such perspective on “safety” (the safety perspective) should be reflected in the decommissioning
work review.
Secondly, with regard to “field-oriented”:•

The on-site environment is in a peculiar state that includes a high level of radiation, andtherefore attention should be paid to the feasibility of construction/implementation of safetymeasures on site.

•

An approach through design alone has limitations due to significant uncertainties.

9Abbreviation of As Low As Reasonably Practicable. This is the principle that the radiological impact must be as

low as reasonably achievable.

23

From the above-mentioned reasons, it is essential to accurately apply the information gained onsite into engineering. In order to ensure the implementation of unprecedented engineering such asfuel debris retrieval, the views and feelings of the individuals and organizations (operators) that are
responsible for the on-site work (including operation, maintenance, radiation control,
instrumentation, analysis, etc.) and very familiar with actual site should be highly respected.
Moreover, it is important to respect their perspectives and judgements directly based on the site(the operator’s perspective). In promoting the prolonged decommissioning work, it is necessary tomaintain and strengthen the operator's perspectives/feelings, and TEPCO themselves should
inherit their perspectives. Therefore, TEPCO needs to take action that always accounts for theworksite in the overall decommissioning work process, such as by inviting outside experts andtechnicians with operator’s perspectives for coaching/educational training, including experienced
workers in difficult operations and those who experienced on-site operations.
In the actual study of the decommissioning work, TEPCO, as the decommissioning project
executor, should clearly define the “requirements” for the work in advance, and should consider
specific safety measures to achieve them. In doing so, it is essential to apply the safety perspective
and the operator’s perspective to handling the characteristics (peculiarities) of decommissioning
the Fukushima Daiichi NPS. Specifically, requirements should be established in consideration of“the safety perspective” and “the operator’s perspective”, and specific safety measures selected
for the work that satisfies the requirements, considering the two perspectives. In each stage ofreviewing the decommissioning work, in this manner, sufficient attention should be paid to “thesafety perspective” and “the operator’s perspective”.
In this decommissioning work with significant uncertainties, it is frequently difficult to clearly
define requirements in advance. Even in such cases, the decommissioning work should be carriedout flexibly and promptly with by verifying and improving the selected, specifical safety measures
as the “preliminary implementation and utilization of the obtained information in the latter stages”
and “iteration-based10 engineering” as described later.
This section describes first the importance of the safety assurance measures in terms of thecharacteristics of Fukushima Daiichi NPS based on safety assessment which includes theoperator’s perspective. Then, it describes the operator’s perspective-specific importance thatshould be incorporated at multiple levels in the safety assurance process while requesting spreadof “Safety first” performed by project operator. Lastly, the section refers to the importance of ALARP
judgment.
2.3.1.1 Promulgating the “Safety First” principle that safety perspective comes first
The use of any method or device is basically unacceptable unless the safety perspective issufficiently reflected in them. Therefore, it is important that all who work in the processes (projects)
leading up to the use of methods and devices on the site, keep the safety perspective first in mind
as they engage in their work (safety first). The specific application of the general “safety first”
10Method that gradually increases the level of completion of engineering by obtaining the next result from one

result and repeating this process.

24

principle in the projects means, “Conducting extensive assessments on safety matters associatedwith methods and devices when reviewing any project and, upon verifying that necessary andsufficient levels of safety have been ensured, taking into account factors such as technical reliability,
reasonableness, speed, actual site applicability and project risks in a comprehensive manner todecide which methods or devices to use, and which safety measures to apply consequently”.
Since the accident at Fukushima Daiichi NPS, leaders of nuclear operations at TEPCO have
stepped up to the plate and continue to work hard to raise awareness on the issue of nuclear safety,
such as through dialogue amongst themselves, as well as through messages that they
communicate to other TEPCO employees. In order to thoroughly disseminate the “safety first”
principle to all persons involved in the projects including on-site workers, the attitude of top
management (the approach to reiterating the special nature of nuclear safety and that special
attention is needed accordingly) is important.
2.3.1.2 Optimization of judgement with a safety assessment as its basis and ensuring
timeliness in decommissioning
With an aim of reducing risk through decommissioning, it is most important to take appropriate
measures and ensure the safety of work in which a large amount of radioactive material is handledthat is technically difficult and has significant uncertainties, such as fuel debris retrieval. Thus,
decommissioning work should be carried out with such “safety perspective”.
Specifically, when designing safety measures for each decommissioning activity, it is essential
to make decisions based on the five guiding principles after conducting a thorough safety evaluationand confirming that the required safety is ensured. As mentioned above, the decommissioning workof the Fukushima Daiichi NPS is unprecedented and has significant uncertainties. Using
deliberated safety evaluation as the basis for making decisions regarding safety measures, thedecisions will not be significantly unstable (that is, without devoting too little or excessive resources),and thus necessary, sufficient, and reasonably feasible safety measures can be realized(optimization of judgment based on safety assessment). In regard to reasonably feasible safetymeasures, it is particularly important in the safety assessment of the Fukushima Daiichi NPS to
conduct safety assessment with incorporating operator’s perspective stated in the section 2.3.1.3.
In addition, the importance of making progress in the decommissioning work without delay (theimportance of time-axis-conscious action) can be mentioned as “the safety perspective” unique tothe decommissioning of Fukushima Daiichi NPS. Considering the high radiological impact that has
already materialized, and the possibility of further degradation of containment barriers, etc., making
progress in the decommissioning work without delay will have great significance for ensuring thesafety of the entire decommissioning process from a medium-and-long term perspective. Therefore,
it should be noted, for ensuring safety, that different perspectives from normal reactors are required,
which have a certain margin in terms of human, physical, and financial resources and have low
radiological impacts and high stability. On the condition that safety is secured, rational judgement
should be made on resource allocation and the time-axis-conscious progress in the25decommissioning work without delay based on the relationship with the overall balance (ensuring
timeliness in decommissioning activities).
2.3.1.3 Ensuring safety by incorporating “the operator’s perspective”
To ensure that safety measures are truly effective, it is necessary to satisfy the needs from thestandpoint of those who actually perform the operations and tasks on site, “the operator’s
perspective” (perspectives and judgements from the standpoint of those who are familiar with thesite and perform operations and tasks on site) is important. In addition to this standpoint, theFukushima Daiichi NPS is a facility that has suffered from an accident, and unlike a normal reactor
its decommissioning requires an unprecedented approach that can be carried out in a peculiar
environment, such as one with high radiation levels. Therefore, when determining the feasibility ofsafety measures on site, the peculiarities of the on-site conditions, such as high radiation levels
and the environment, shall be considered.
“The operator’s perspective” is also important for ensuring safety from the following perspectives,
which are different from those of normal reactors.
• Complementation of design by operations, including operating controls:
Due to significant uncertainties, there is a limit to addressing all situations by design alone.
Therefore, it is effective to supplement the design with operator response and on-site operation,and improve safety collectively with operation.
• Utilization of information in design obtained through monitoring, analysis, etc.:
To cope with significant uncertainties, it is important to utilize information obtained through onsite operations such as monitoring and analysis, etc., in designing safety measures. Whenutilizing such information, it should be linked with calculational evaluation, etc., to “make acomprehensive use of it”.
• Handling an abnormality:
Although it is essential to take all possible measures to prevent progress of an abnormality, onsite response as a precaution to prevent the occurrence of an abnormality is effective
considering the characteristics that the progress of abnormalities is moderate and there issufficient time to respond11.
2.3.1.4 ALARP judgment based on safety
For safety, there is a minimum level of safety standards that must be met before the relevant
retrieval method or equipment can be used. At levels above the level that meets this minimum level,
there is a range of choices and, within this range, retrieval methods and equipment to be adopted
will be determined based on trade-off between the safety level to be achieved and project cost andduration (note that retrieval method and equipment involving long-term large safety measures are

11Since a long time has passed since the accident at the Fukushima Daiichi NPS, the intrinsic energy (decayheat) that drives dispersion of radioactive materials is small. Therefore, in general, abnormalities have thecharacteristic of progressing slowly so there is a large time margin to respond.

26not necessarily beneficial to safety, especially at Fukushima Daiichi NPS), a kind of ALARP. Thereis also an issue as to whether such retrieval methods and equipment are feasible in the field.
Based on the above, in the process of determination, it is important to decide retrieval methods
and equipment to be employed eventually through the cycle of “defining the safety standards(safety perspective)”, “indicating the feasibility on-site (operators’ perspective)”, and “examining
and discussing at projects (project management)” as shown in Fig.6. As shown in this figure, thesafety perspective and the operator’s perspective are not independent from each other. The ALARP
judgement made by the project based on the safety perspective will be linked to the decision of theretrieval method and equipment after going through the feasibility check based on the operator’s
perspective. The operator's perspective is essential to actually incorporate the safety perspective
into the site, and the judgment based on the safety perspective is needed to utilize the operator's
perspective.

① Define the safety standards
to be achieved(Legal/regulatory requirements)

Set management
standards from theoperator’s perspective,
with a margin to safety
standards.
(Operator’s perspective)

Inform retrieval methods andequipment proposals② Examine and discuss
methods and equipment
to be adopted in the
project (ALARP)③ Examine feasibility onsite (by experts).
(Operator’s perspective)

Provide retrieval method andequipment options to meetproject operator’s targets.

④ Determine retrieval methods
and equipment at the project.

Inform availability(if possible)(Safety perspective)

Inform availability(if not possible)
Review management targets in some cases.

Fig.6 ALARP centered on safety (conceptual diagram)
2.3.2

Preliminary implementation and utilization of the obtained information in the latterstagesThe on-site conditions at the Fukushima Daiichi NPS containing the reactors involved in theaccident includes considerable uncertainties. If the whole operation of a large-scale project suchas fuel debris retrieval is to be designed only with existing knowledge, assumptions of an extremely
large safety margin and wide range of technical options will be needed. Thus, extension of the work
period or the risk of rework will be unavoidable. As a result, the feasibility or predictability of theentire project may be reduced, leading to a delay in the entire decommissioning, a rise in thedecommissioning cost, or increased radiation exposure of workers.
However, considering the current environment with an already high radiation level, further
deterioration of containment barriers, and the possibility of future major natural events (such as
earthquakes or tsunamis), it is necessary to immediately improve the state of such risks and reduce
uncertainties. Therefore, a “sequential type approach” is important where the whole operation isdivided into several stages, “operation at first stage” is implemented for which practical safety canbe ensured, and then the information obtained there is utilized in the next stage. With this27

approach12, operation proceeds with safety ensured through monitoring the condition inside reactor,
restricting operational actions and flexible on-site responses13 at each stage of the process. The
information obtained at each stage of operation is utilized in the design of subsequent stages. Thisapproach reduces uncertainties in the operations in subsequent stage as well as improve thereliability of safety assurance and rationalize design.
TEPCO should actively introduce an approach like this into actual engineering and project
management14.
In a case that uses a similar approach, TEPCO has been conducting reactor water injection
shutdown tests since FY 2019. One of the purposes of these tests is to contribute to determining
whether or not to terminate water injection in the future, which also takes into account maintaining
flexibility in selecting fuel debris retrieval methods. Knowledge on whether or not to terminate water
injection has been accumulated by identifying different risks (rising temperature of fuel debris andthe RPV bottom, increased dust scattering outside the PCV, and re-criticality when resuming water
injection), and by gradually extending the testing time while taking certain risks.
These reactor water injection shutdown tests have resulted in the clarification of the relationshipbetween the lowering of the PCV water level and pressure in Unit 1. According to the informationon the damaged piping that was obtained from the on-site investigation conducted before the tests,
it was assumed that the PCV pressure might decrease when the water level reached the pertinent
damaged area. In fact, the occurrence of the event, which developed as predicted, further
enhanced the degree of confidence of the presumed cause that “the PCV pressure dropped as aresult of exposing the damage area due to the lowering of the PCV water level”.
This example of the reactor water injection shutdown test yielded results that led to reducing theuncertainty through examination in combination with the information gained from on-site
investigations, although this information is not directly related to the purpose of the test.
Hereafter, it is recommended to make it clear as a policy that the information to be gained
through on-site operation should be fully incorporated and accumulated as knowledge inconsecutive activities for ensuring safety. For example, the same applies to risk identification
associated with hydrogen at the time of fuel debris retrieval. Testing to reduce nitrogen supply foran experimental purpose may help identify hydrogen risk, and determine requirements on thenecessary amount of nitrogen supply and reliability of the exhaust systems to ensure safety. It is
important to accumulate successful/unsuccessful experience gained in the process of thesesequential approach as a track record, allowing gradual reduction in major uncertainties in theoverall decommissioning work in the future. This will lead to steady progress in decommissioning
and contribute to ensuring safety in decommissioning the Fukushima Daiichi NPS from theperspective of risk reduction in the medium-and-long term.
12This is also used in UK, for example, for the decommissioned facilities in Sellafield, and is called Lead & Learn.
13 Some example measures include installing nuclear instrumentation to the extent feasible; limiting the amount ofdebris retrieving; and setting the value for managing radioactive dust concentration and regulating operations.
14 This is stated in the Decommissioning Implementation Plan (March 17, 2021, Tokyo Electric Power Company

Holdings Inc.,), which summarizes the policies on implementing decommissioning at the Fukushima Daiichi
Nuclear Power Station. https://www.tepco.co.jp/press/release/2021/1585525_8711.html

28

2.3.3 Approach to address a temporary increase in risk level associated withdecommissioning operations
While the decommissioning work is striving for prompt risk reduction from a medium-and-longterm perspective, careful deliberation of the possibility that the performance of decommissioning
work may temporarily change the risk levels and may increase the radiation exposure of workers
is required. Executing the decommissioning work involves taking some action on the current
situation of the NPS, which is maintained in a state with a certain level of stability despite somerisks. Such risks may materialize, depending on the way action is taken. For example, accessing
the inside of the reactor to retrieve fuel debris will affect the current containment status, and thespecial operations and maintenance performed in the retrieval work will increase the exposure ofworkers involved in these activities.
The possibility of a temporary increase in the risk level and a rise in workers’ exposure arisingfrom such decommissioning work must be addressed by taking measures to prevent and restrict
them. In particular, as for the radiation safety of workers, it is imperative to limit the increase in therisk level during decommissioning as much as practicably possible by thorough preparations as
achieved through application of the concept of ALARA (to suppress radiation exposure to As Low
As Reasonably Achievable).
Note that the basic stance for promptly implementing the decommissioning must stand firm
because if the decommissioning work is delayed excessively, it means that existing major risks willremain over the long term and their risk levels may gradually rise as the buildings and facilities
deteriorate over time. Therefore, with regard to the selection of work methods, the design andmanufacture of equipment and safety systems, and the development of work plans for thedecommissioning work, cautious and comprehensive decision making is required for early
implementation of decommissioning in consideration of many constraints such as time, cost, andworker’s exposure needed for relevant preparations and work, while giving priority to limiting therisks involved in the decommissioning work (Attachment 5). The approach to risk reduction andensuring safety in the decommissioning of the Fukushima Daiichi NPS, as described in this chapter,
RA), NDF, TEPCO, and others to cooperate with each other to reduce risks based on the approachto ensuring safety based on their respective positions. In doing so, it is important to establish asystem for on-going risk monitoring which enables a wide range of people to easily understand how
the overall risks at the site have been continuously reduced through the decommissioning work,and to communicate such progress to the public. In addition to sharing the status of risks throughthe Technical Strategic Plan on a constant basis, NDF is considering providing the status of risk
reduction along with the progress of the decommissioning work described in 2.2.2.2. TEPCO also
needs to develop a mechanism to identify risks for the entire site and become aware of the needto take action to communicate the status of risk reduction to society in a proactive manner.

29

3. Technological strategies toward decommissioning of the
Fukushima Daiichi NPS
3.1 Fuel debris retrieval
3.1.1 Targets and progress(Targets)(1) Retrieve fuel debris safely after thorough and careful preparations, and bring it to a state ofstable storage that is fully managed.
(2) Trial retrieval in Unit 2 was scheduled to begin within 2021, but the process has been delayed
due to the COVID-19 infection. In order to limit the delay to about one year, preparations will
be made for starting retrieval. Continue a series of work including the gradual expansion offuel debris retrieval to acquire knowledge and experience necessary for the further expansionof fuel debris retrieval (For the target of fuel debris retrieval, see Attachment 6).
(3) With regard to the further expansion of fuel debris retrieval, consideration will be given to themethods including those for containing, transferring, and storing of fuel debris, by assessing
fuel debris retrieval at the first implementing unit, internal investigations, research anddevelopment, and the on-site environmental improvement, etc.
(Progress)
Fig. 5 shows the estimated fuel debris distribution, access route and surrounding structures ofUnits 1 to 3. The progress in each unit is also shown below.
(1) Unit 1
Within FY 2021, further detailed information inside PCV is planned to be gained by inserting aboat-type access investigation device with diving capabilities (hereinafter referred to as“underwater ROV”) into PCV and investigating the internal status such as distribution of deposits
widely scattered at the bottom of outside the pedestal, presence or absence of fuel debris involved
in deposits, and structures inside the pedestal. In preparation for the start of this investigations,removal of obstacles within the PCV is being promoted while taking measures to control dust
dispersion and monitoring dust concentration by considering the change in dust concentration
when opening the inner door at the penetration X-2. Also, at the beginning of 2021, during the
preparation of the obstacle investigation in PCV, for the pressure decreasing that were expected tohave occurred due to the addition of external force to the outer door of the penetration X-2, removalof obstacles within the PCV is being promoted while performing measures to suppress theoccurrence. As the immediate response to the PCV water level drop caused by the earthquake off
the coast of Fukushima Prefecture on February 13, 2021, a study is underway to implement
continuous water level monitoring. Specifically, when conducting PCV internal investigation using
underwater ROV, consideration is being given to increase the amount of injection water once to thelevel before the earthquake because the risks of interference with deposits increases when thewater level is low, and to return to the current water level after PCV internal investigation.
30

As forward-looking measures, lowering of the water level is planned to improve the seismic
resistance of the suppression chamber (hereinafter referred to as “S/C”), and preparations are
currently being made to lower the water level, such as by water intake using existing piping andwater quality surveying of the S/C water.
(2) Unit 2
The 2019 Med-and-Long-term Roadmap specified Unit 2 as the first implementing unit for fueldebris retrieval, and its trial retrieval was supposed to launch within 2021. However, the process
has been delayed due to the new coronavirus infection. In order to limit the delay to about one year,
preparations are being made for starting fuel debris retrieval. As for dust dispersion prevention, in
consideration of the changes in dust concentration while drilling the inner door of the penetration
X-2 for Unit 1 PCV internal investigation (June 2019), preparations are underway for low-pressure
water cleaning of deposits, spray curtains on the penetration outlet and countermeasures to reducepressure in the PCV. The arm-type access equipment (robot arm) has arrived in Japan and started
testing. In addition, in order to allow the arm-type access equipment to be used to enter from thepenetration X-6, removal of the deposits in the penetration X-6 is planned. In 2020, deposit contact
investigation and 3D scanning investigation in the penetration X-6 showed that the deposits deform
due to contact, the deposits are accumulated on the slope from the building side toward thepedestal, and the remaining cable is not stuck and can be lifted. Based on these results, removalof the deposits in the penetration X-6 is under consideration.
A plan for the gradual expansion of fuel debris retrieval is also underway, and the retrieval device
will be improved by increasing the weight capacity and enhancing accessibility while complying
with specifications for the devices for trial retrieval and PCV internal investigations. In this plan, therequirements related to the arm performance of the arm-type access equipment and the interface
between the equipment and enclosures have been clarified and examined. The retrieved fuel debris
will be stored in unit cans in the enclosure, and then transferred to the receiving/delivery cells onsite and stored in temporary storage cells. In addition, some of the fuel debris will be collected in
the receiving/delivery cells for analysis and transported to the facility for analysis. Designing of theretrieval device, receiving/delivery cells, and temporary storage cells is in progress (Fig. 8, Fig. 9).
In response to the unprecedented approach to retrieving fuel debris from the first implementing
unit, NDF continues working while verifying the actual site applicability of the device and the resultsof the review on modifications to the safety system from the perspectives of safety, reliability,
reasonability, timeliness, and a field-oriented stance, in accordance with the progress ofengineering at TEPCO.
(3) Unit 3
As for Unit 3, due to high water levels in the PCV, it is planned to gradually lower the PCV waterlevels, taking into account the improved seismic resistance of the S/C and conducting PCV internal
investigation. The result of sampling the S/C water (conducted in 2020) using the piping connectedto the S/C shows that the concentration of radioactive materials such as Cs-137 is higher than that31

of the stagnant water in buildings. Therefore, its impact on the treatment of contaminated water isbeing considered, and the analysis results are being incorporated into the PCV water intake system.
In addition, the conceptual study on further expansion of fuel debris retrieval is in progress.
(4) Impact of and response to the earthquake on February 13, 2021, with its epicenter off thecoast of Fukushima Prefecture
Due to the earthquake that occurred off the coast of Fukushima Prefecture on February 13,
2021, lowering of the PCV water level was observed in Units 1 and 3, however, water injection into
the reactor is continuing. As there are no significant changes in plant parameters, the coolingcondition of fuel debris is considered to be intact. It is assumed that this drop in water level may
have been caused by an increase in leakage from inside of the PCV due to changes in the conditionof the damaged areas previously identified in the PCV and its newly damaged areas. Going forward,
the change in parameters such as water level will be checked by water injection shutdown testing,and expansion of knowledge be considered.
In light of the recent earthquake, it is necessary to enhance monitoring systems to observechanges in plant conditions, perform impact assessment to maintain/manage systems andbuildings over the med-and-long-term, and promote technical development required to understandthe situation. (Refer to 3.1.2.4.1.5: Issues in the structural integrity of PCVs and buildings)

32

(Prepared in reference to “Material 4-1: Progress of treatment of stagnant water in buildings”, the 81st meeting of the Study group on
monitoring and assessment of specified nuclear facilities)

Fig. 7 Estimated fuel debris distribution, access route and surrounding structures of Units 1 to 3

33

(TEPCO material edited by NDF)

Fig. 8 Image of fuel debris retrieval system(Trial retrieval and subsequent gradual expansion of fuel debris retrieval)(TEPCO material edited by NDF)

Fig. 9 Image from retrieval to temporary storage of fuel debris(Gradual expansion of retrieval scale)
3.1.2 Key issues and technical strategies to realize them
Since the understanding the situation inside the PCVs is still limited, the current design and theplan for on-site operations related to fuel debris retrieval should be continuously reviewed basedon knowledge that will be obtained in the future, and it is also important to accurately incorporate
the results of studies, research and development toward fuel debris retrieval.
In Unit 2, trial retrieval and PCV internal investigation will be performed, and implement thegradual expansion of fuel debris retrieval based on the findings. In addition, a conceptual study on
the further expansion of fuel debris retrieval for Unit 3 is planned.

34

3.1.2.1 Trial retrieval and PCV internal investigation, and gradual expansion of fuel debris
retrieval
3.1.2.1.1 Development status and prospects of trial retrieval/internal investigationequipment
For the trial retrieval and PCV internal investigation in Unit 2, the operation will be performed by
opening the flange of the penetration X-6 to make a larger opening than before, through which thearm-type access investigation equipment is moved in/out to retrieve fuel debris inside the PCV. In
this operation, an expansion will be made to provide an isolation chamber (composed of a robot
carrying-in room, etc.) to be built during opening the penetration X-6 (Fig. 10), and an enclosure to
be newly provided (which will enclose an arm-type access investigation equipment, etc.) (Fig. 11),
since the conventional containment barrier was located in the blank flange part of the penetration
X-6. Although small in scale, this is a fundamental form of site construction for future retrieval work,in which an opening will be newly provided in the PCV to extend the containment barrier outside
the PCV. This presents an approach that enters a new stage.
Thus, as the opening of penetration X-6 is a task related to maintaining to containment barrier
function, careful consideration should be given to safety in particular. From the preparatory workstage, detail of the work should be thoroughly checked, potential risks should be identified, andcountermeasures should be considered. For its implementation, it is important to conduct more
careful and thorough preparation, verification and training.
The fuel debris retrieval through the trial retrieval and PCV internal investigation was scheduled
to begin within 2021, but the process has been delayed under the impact of the COVID-19 infection,
considerations are being made to limit the delay to about one year. The performance confirmation
test of the arm-type access investigation equipment in the UK has been delayed. As there was a
possibility of further delay, reconsideration is given to cancel the performance confirmation test andmockup test simulating the on-site environment, transport the equipment to Japan, and conduct
the performance confirmation test, mock-up test and training domestically. No changes due to thisreconsideration have been made to the initial plan for testing and training menus.
Although it is important to make efforts to minimize delays caused by the COVID-19 infection,
mockup testing that takes full account of uncertainties on site is also important in terms of actualsite applicability and ensuring safety. In addition to simulating the severe environment on site, thismockup testing should clarify the parts that cannot be simulated and make enough preparations
for practical application. NDF will also confirm the test plan and whether measures are sufficient.

35

(TEPCO material edited by NDF)

Fig. 10 Schematic drawing of isolation chamber to be installed at penetration X-6(during opening the Penetration X-6)

(TEPCO material edited by NDF)

Fig. 11 Schematic drawing of enclosure to penetration X-6(during trial retrieval and PCV internal investigation)
The key technical issues, countermeasures, and points to consider are described below.
・ Dust dispersion prevention associated with removal of deposits in the penetration X-6.

36

In order to prevent dust dispersion due to removal of deposits inside the penetration X-6,
measures such as low-pressure water cleaning of deposits, spray curtains to the penetration outlet
and lowering pressure in the PCV are in preparation, in consideration of the changes in dust
concentration during drilling the inner door of the penetration X-2 for Unit 1 PCV internal
investigation (June, 2019). During AWJ cutting of obstacles from the penetration X-6 to the pedestal
opening (ground floor), it is also planned to perform the work after lowering PCV inner pressure.
For the removal of deposits in the penetration X-6 and AWJ cutting of obstacles, it is essential
to develop a detailed work plan in accordance with the preliminary implementation and use it in thelater stages as mentioned in 2.3.2 and ensure safety accordingly (for example, break down work
steps; proceed to the next step after confirming there is no problem with the dust dispersion
monitoring results in each step; and take countermeasures and proceed to the next step if any signof abnormality is observed).
NDF will confirm whether a sufficient level of safety is secured for performing work, including
whether the work plan is well-developed by TEPCO, the plan is fully implemented without fail, andwork is stopped when necessary.
・ Considerations for the risk of spreading impact of the COVID-19 infection
The performance confirmation test and mockup test simulating the on-site environment in the
UK have been canceled, and tastings by bringing the equipment into Japan are being promoted.
For the performance confirmation test in Japan, it is essential to secure the UK engineers, and it isnecessary to maintain the backup system on the UK side in the event of a defect, while sharing
information and communicating smoothly with the UK engineers.
It is also important to make all possible preparations for the risk of the COVID-19 pandemic
expanding in Japan. NDF will confirm these responses.
・ Considerations in project management
It is important to proceed with the project while paying attention to the process progress
management of the contractors including overseas enterprises and subcontractors. As part of theirproject management activities, TEPCO needs to make further efforts to perform prior-evaluation ofrisk of delays, and develop alternative plans and measures to prevent the occurrence of risks. NDF
also participates in meetings with contractors and their subcontractors to closely check the status
and support risk assessment.
・ Limitations in the scope of trial retrieval and internal investigation, and incorporation into
gradual expansion of the retrieval scale
In the PCV internal investigation using a robot arm, it is planned to ascertain the state of existing
structures, and the distribution of deposits inside the pedestal (3D data), the distribution of gamma
rays and neutron counts at the bottom and on the platform, in as wide a range as possible. However,since more structures and platforms in the pedestal remained than the initial design plan, the range
in which the arm can access the bottom of the pedestal is limited. Thus, the possible range ofneutron measurement and trial retrieval from the bottom of the pedestal is limited. Assuming thatdebris at the bottom of the pedestal cannot be retrieved, it is also planned to retrieve the deposits
37

on the platform which are highly likely to be fuel debris, as same as those at the bottom of thepedestal. Given the limited scope of investigation and trial retrieval, greater consideration isrequired in advance to determine what information is needed to gradually expand the retrieval scale
as a next step for promoting the retrieval work in a reliable manner.
・ Human resource development and technology transfer for the next step (gradual expansion of
the retrieval scale)With regard to the trial retrieval, there are uncertainties and difficulties in the development of therobot arm and the removal of deposits and obstacles due to a limited understanding of theconditions inside the PCV. Therefore, when performing such work, it is necessary for TEPCO andrelated institutions to utilize human resources with a wealth of field experience, including thoseinvited from outside as needed, to develop human resources to foster field-oriented
perspectives/feelings, and to transfer techniques cultivated through these activities.
3.1.2.1.2 Development status of gradual expansion of retrieval scale and prospects
The retrieval equipment to be used for gradual expansion of the retrieval scale will be improvedby increasing the weight capacity and enhancing accessibility while complying with specificationsof the devices for trial retrieval and PCV internal investigation.
It is planned to expand the range of retrieval step by step while making achievements, startingwith retrieval of fuel debris that can be gripped and sucked, and expanding it to fuel debris retrieval
with cutting. Consideration will also be given to the possibility of cutting platform beams and therange of cutting. The enclosure containing the arm-type access equipment, etc. is connected to thePCV via penetration X-6 connection structures to secure containment functions. In order to bringfuel debris into the enclosure, it is necessary to consider shielding, measures against hydrogenand prevention of the spread of contamination, methods for transferring fuel debris from theenclosure, and methods for confirming the maintenance of boundary and dynamic equipment
functions and for remote maintenance.
From the perspectives of research/development and engineering by TEPCO, and in terms of
ensuring actual site applicability and safety, NDF continues to observe and check the status oftechnology development and preparations for application to the site in a timely manner.
The key technical issues and countermeasures are described below.
・ Ensuring containment performance of enclosures for fuel debris
In the retrieval operation, the process is repeated from carrying debris retrieved from the PCV
into the enclosure, storing in unit cans and to carrying out to the outside of the enclosure for onsite transportation. As a result, the enclosure gradually becomes contaminated, and it is importantto secure the containment performance of the enclosure.
This work is performed by controlling the pressure in the enclosure as the arm-type accessequipment is moved in/out. Therefore, in order to confirm airtightness performance and operation
reliability, through the duration, it is important to perform prior mockup testing, post-installation
testing of the equipment, and subsequent abnormality monitoring.

38

・ Ensuring reliability of a dual arm manipulator (Fig. 12)
The dual arm manipulator to be installed in the enclosure plays an important role in performing
various operations and maintenance in the enclosure (cleaning and decontamination of the arm,
replacement and maintenance of cutting/recovery device, storage of fuel debris in containers, carryout of storage containers, decontamination in enclosures, replacement of parts, etc.), and thus it isimportant to ensure its reliability. Therefore, it is necessary to improve the reproducibility of workthrough a wide range of operation/maintenance training in advance, and to train operators.

(IRID material edited by NDF)

Fig. 12 Enclosure and dual arm manipulator
・ Ensuring maintenance of devices and countermeasures during the in-service period
To expand the retrieval scale in a gradual manner, in addition to periodic maintenance, repair or
replacement is required in case of failure. Since the radiation dose in the Unit 2 reactor building,where the enclosure will be installed, is high and it is difficult to perform maintenance in that place.
Therefore, it is planned to construct a maintenance building outside the building, transfer theequipment or enclosure itself, and decontaminate, dismantle, repair or replace it inside themaintenance building.
In addition, since a dual arm manipulator that performs various operations may be repaired or
replaced during the in-service period, the device to carry out the manipulator to the maintenancebuilding is under development.
Since it is extremely important to ensure the maintenance of equipment/devices and theirmeasures, including repairs, NDF will check the examination and preparation status for them in
TEPCO. It is also important to leverage the experiences gained through the in-service maintenanceof equipment/devices for further expansion of the retrieval scale. Therefore, a system that canreliably preserve maintenance records, including failure histories and their measures, should beestablished.
39

3.1.2.1.3 Further expansion of fuel debris retrieval
Toward further expansion of the retrieval scale, methods should be selected based on the
viewpoint that “fuel debris retrieval is an important process in decommissioning, and its retrieval ina reliable manner affects the success/failure of the decommissioning project”, and from acomprehensive standpoint (in anticipation of technical feasibility as well as business continuity). In
addition, TEPCO should take responsibility for selecting the method. Therefore, this section
describes in detail how to select methods.
At the Fukushima Daiichi NPS, where uncertainty still exists, the uncertainty of the condition
inside the PCV hinders examination, which forces preconditions to be set to perform examination.
In judging technical feasibility in the future, it is important to clarify and examine the requirements(boundary conditions) and constraints (site use area, existing system interface, etc.) for methods
and systems, including criticality control, dust containment, shielding, and heat removal.
TEPCO has been engaged in the conceptual study for further expansion of the retrieval scale.
As part of this conceptual study, TEPCO will consider scenarios for fuel debris retrieval and, at theend of FY 2021, identify promising methods (top/sub candidates). Subsequently, based on theinformation obtained from the survey results, TEPCO will finally narrow down the possible methods(top/sub candidates), and decide the method with its design in mind.
The following are the points to be considered for examining retrieval scenarios and methods.
・ How to select retrieval methods
In selecting the method, based on the five principles mentioned in Section 2.2.2.3 (Safe, Proven,
Efficient, Timely and Field-oriented), a determination should be made not only to satisfy the targetof safety level (radiation exposure dose of the general public and workers, robustness against
natural events such as earthquakes and tsunamis) but also to use the attributes (evaluation items)
such as cost and schedule as indexes for selection. It is necessary to quantify these evaluation
items as much as possible by using a multi-attribute utility analysis method15, etc. It is believed thatthe most important factors in the process of selecting methods are the evaluation items used asindexes for determination and how to weight these indexes. In setting these indexes, a decision
should be made not only based on the approach of TEPCO as an operator, but also from acomprehensive viewpoint developed through discussions with experts. Normally, the selection of
methods should be based on the results of internal investigation. However, in a situation with many
uncertainties like at the Fukushima Daiichi NPS, it is considered necessary to proceed withexamination based on the currently available information and then to feed back the results gained
from the investigation. As for the results of method selection, it is also important to make efforts todisseminate information in a careful manner so that the evaluation results will be widely acceptedby society.
15Method for determining the relative merits and demerits for decision-making based not only on one attribute(evaluation item) but also multiple attributes (evaluation items). This methodology is applied to the process of
selecting methods, and those with a high score calculated from “Σ (evaluation of each attribute (evaluationitem)) x (weight of each attribute (evaluation item) = importance)” will remain as promising methods.

40

・ Development of retrieval scenarios
Given the limited understanding of the situation in the PCV, it is important to examine several
scenarios of fuel debris retrieval by each unit and to clarify several paths from start to completion.
In considering fuel debris retrieval scenarios, different cases are assumed, where fuel debris will
be retrieved by the side-access or top-access methods, or combining both. Then, examinations
including internal investigation required for each case will be made to consider several paths. This
study intends to assume in advance different results obtained from PCV/RPV internal investigations
or technical studies in the future, and then conduct examination based on the preconditions of using
such results.
After reviewing these numerous paths, it is important to narrow down promising candidates for
retrieval methods at a certain point of the path (e.g., narrow down promising candidates for theside-access method), and then further narrow down the path to take according to the information
obtained afterward (e.g., perform internal investigation, and narrow down promising candidates forthe top-access method).
It is also important to formulate a specific schedule with considering these fuel debris retrieval
scenarios.
・ Clarification of requirements
With regard to further expansion of the retrieval scale, the methods will be considered, including
those for containment, transfer, and storage of fuel debris, based on the findings from fuel debris
retrieval in Unit 2 (trial retrieval, gradual expansion of the retrieval scale), PCV/RPV internal
investigation, research and development and the on-site environment improvement, etc. In doingso, operations, devices and equipment, and facilities will be larger than, and the scope of
construction will be wider than in the case of retrieving fuel debris from Unit 2. Therefore, much
more attention should be paid in overviewing the entire Fukushima Daiichi NPS, including otherwork. In addition, because of the high radiation dose on site and the limited understanding of thesituation inside the PCVs, the scope of work may also be extensive. Therefore, it is important tospecify the requirements (containment, criticality, operability, maintainability, throughput 16 , etc.)
required for operations and devices more clearly and proceed with the work. Attention should also
be paid to the interaction between the requirements.
・ Process for narrowing down promising retrieval methods
Based on the development results of the Project of Decommissioning and Contaminated Water
Management, the latest findings at home and abroad are taken into consideration to derive ideas.
Then, primary/secondary screening will be performed to gradually narrow down the methods. The
primary screening verifies compliance with requirements and constraints, while the secondary
screening quantifies and weighs each evaluation item. A multi-attribute utility analysis method,etc., is utilized for such screening process (Fig. 13). Diverse ideas are expected to be derived, but

16 It represents the ability to retrieve fuel debris and indicates the processing time and work efficiency of theretrieval operation.

41

it is important to conduct objective evaluation based on this process and to narrow down themethods.
End of FY2021
Ideas

Narrowing down
promising methods(First and second candidates)

Candidates

Method A
Method B

Method A

Method C

Method C

Method D
Method E
Method F

Primary screening
Meet therequirements or not

Method D
Method E Secondary screening

Method A

Method G

Method C

Method G

Method H

Method H

Method I

Method I

Method K

Quantification andweighting of eachindicator

Method E+G

Method J
Method K
Fig. 13 Image diagram of the process for narrowing down retrieval methods
Based on the above considerations, NDF will evaluate the validity of the results of TEPCO'sconceptual studies (Fig. 14).
On-site work such as retrieval offuel debris from the first
implementing unit and internal
investigations
 New insight obtained from fueldebris retrieval from the first
implementing unit and internal
investigations

Research and development
 Progress made in research anddevelopment (by theGovernment-led R&D program on
Decommissioning andContaminated Water
Management) and technologicaldevelopment (TEPCO ‘svoluntary project)

Status in the improvement of
workplace environment

Status of site-wide planning Contaminated water Status in the on-site environment
management(radiation dose, water level in PCV)  Plan for solid waste Various new installations
 Interference with other works, etc.
(including buildings)
Prospect of site securing(planning for removal of obstacles
including rubble), etc.

Evaluation on the results of the retrieval conceptual study (including the overall conceptual study)●Target:• Requirements• The methods of retrieval, accessing, anddebris recovery• Safety system• Facilities for collection, transfer and storage, etc.

●Viewpoints in evaluating the actual site applicability
and feasibility:• Safety• Reliable• Reasonable• Prompt• Field-oriented

Comprehensive evaluation●Comprehensive evaluation from the perspectives of site-wide optimization

Fig. 14 Flow of retrieval method study (conceptual diagram)
Issues in each technical field for further expansion of the retrieval scale are described in
Section 3.1.2.4.

42

3.1.2.2 Continuation of accident analysis activities (clarification of events that occurred at
the accident and the process of accident progression)
Analysis of deposit samples collected by the previous internal investigation for fuel debris
retrieval is in progress17. The information obtained by such investigation and analysis is directly
incorporated into fuel debris retrieval methods and storage management. Moreover, examinationand study in light of the accident history will promote understanding of phenomena, contribute todetermining the cause of the accident and decommissioning, and indirectly to the enhancement of
nuclear safety.
TEPCO and JAEA are cooperating in implementing activities for estimating and verifying
individual events that occurred at the accident, including overheating, melting, chemical reactions
and hydrogen explosions, the process of their progression over time, and the operation status of
emergency cooling and depressurization equipment, by comparing the results of sample analyses
with mock-up testing on accident progression and past scientific knowledge18. Moreover, TEPCO
is independently investigating the operating floor 19 , 20 , 21 and standby gas treatment system 22(hereinafter referred to as “SGTS”) and so forth in each unit.
The NRA responsible for continuing accident investigations/analyses23 established “The Study
Committee on Accident Analysis of the Fukushima Daiichi Nuclear Power Station” and investigatedthe operating floor in Unit 2, the inside of the reactor building in Unit 3, and SGTS filter lines, etc.,with cooperation from TEPCO, and compiled an interim report24. Based on these investigations,
the NRA has assessed that a large amount of Cs exists between the first and second layers of theshield plug installed on the operating floor in Units 2 and 3. Since the information related to such
accident analysis may be affected by the dismantling of the facility due to the progress of
decommissioning work, “Liaison Council on the accident investigation of Fukushima Daiichi
Nuclear Power Station” 25 , 26 has been established, and communication and coordination for
accident analysis and decommissioning has been conducted among NRA, Agency for Natural17

The 84th Secretariat Meeting of the Team for Countermeasures for Decommissioning and Contaminated Water
Treatment, “Material 3-3: Analysis results of samples related to the internal investigation of the Unit 1 to 3 PCVs”
18 PowerPoint report on the development of analysis and estimation technologies for fuel debris characterizationin the national project (To be updated as the material will be posted online around September 2021)
19The 69th Secretariat Meeting of the Team for Countermeasures for Decommissioning and Contaminated Water
Treatment, “Material 3-2: Investigation on obstacles and well plugs in the SFP of the Unit 1 reactor building”
20The 88th Secretariat Meeting of the Team for Countermeasures for Decommissioning and Contaminated Water
Treatment, “Material 3-3: The 24th meeting of the Team for Countermeasures for Decommissioning andContaminated Water Treatment/Secretariat, “Material 3-2: Preliminary report on investigations of the operating
floor of the Unit 2 reactor building”
21 The 24th Secretariat Meeting of the Team for Countermeasures for Decommissioning and Contaminated Water
Treatment, “Material 3-2: γ-ray spectrum measurement results of the operating floor of the Unit 3 reactor
building”
22 The 84th meeting of the Study group on monitoring and assessment of specified nuclear facilities, “Material 4-3:
Surveillance results for removal of SGTS piping at Units 1 and 2”
23Article 4, Paragraph 1, Item 11 of the Act for Establishment of the Nuclear Regulation Authority stipulates thescope of authority as “Affairs concerning investigations of causes of accidents that have resulted from theoperation, etc. of reactors and causes of damage that has arisen from nuclear accidents”.
24The 19th Study Committee on Accident Analysis of the Fukushima Daiichi Nuclear Power Station, “Material 4:
Interim report on investigation/analysis of the accident at the TEPCO Fukushima Daiichi NPS”
25 NRA, " Material 3, Continuous survey and analysis on accident of Fukushima Daiichi Nuclear Power Station”(27th meeting)
26 NRA, The 7th Committee on the accident analysis of the Fukushima Daiichi Nuclear Power Station “Material 2,
Continuous survey and analysis on accident of Fukushima Daiichi Nuclear Power Station”」

43Resources and Energy, TEPCO and NDF. As an example of approaches to balance
decommissioning work and accident investigation, in the partial removal of SGTS piping for Units
1 and 227, TEPCO is planning to take measurements of the radiation dose rates and with a gamma
camera28, and to collect samples of the piping for analysis after cutting. As a countermeasureagainst radioactive dust emission during cutting, injection of urethane form and local exhaust
ventilation are being considered to ensure that the decommissioning and accident investigation will
not affect the surrounding residents and the environment.
TEPCO is also planning to systematically conduct on-site investigation while estimating the stateof the reactor core and PCV, and unsolved issues29. After a lapse of 10 years since the accident,
the radiation dose has decreased due to the decay of fission products (hereinafter referred to as"FP”) and environmental improvements on site, and accessibility to the reactor buildings has
improved. However, there are still many places with high radiation doses. In order to locate FP
released at the accident with a low radiation exposure dose, it is important to collaborate with eachorganization, and continue these activities to a reasonable extent to clarify the events that occurredat the accident, the process of their progression and the equipment operation status. When newfacts about the accident are revealed through further investigations and other activities, it is also
important to deepen and incorporate knowledge by performing severe accident progression
analysis evaluations, etc.
Since the shield plugs are installed on the operating floor, , PCV internal investigation andgradual expansion of the retrieval scale from the first floor of the reactor building are not directly
affected by the trial retrieval. In order to further expand the retrieval scale, however, it is importantto fully recognize high radiation doses in the shield plugs, taking into account the possibility thataccess from the operating floor (top-access) is required, and examination of retrieval methods withdecontamination, shielding, and containment in mind.
3.1.2.3 Technical issues for technical requirements and future plans
3.1.2.3.1 Technical issues for ensuring safety of fuel debris retrieval work
In general, when considering ensuring safety at a nuclear facility, a series of evaluations are
conducted by assuming accident scenarios in which the potential hazards of the facility becomematerialized, evaluating that these scenarios fall within the safety standards, and confirming thatthe safety measures are appropriate. In an ordinary nuclear power plant, such a series of safetyassessment procedures are established and standardized by national regulations and guides.
However, since there are no established and standardized regulations and guides for thedecommissioning work at the Fukushima Daiichi NPS, it is necessary to organize an approach to

27 NRA, The 21st Committee on the accident analysis of the Fukushima Daiichi Nuclear Power Station “Material5-

1, Partial removal of SGTS piping of Units 1 and 2 of Fukushima Daiichi Nuclear Power Station

28 A camera that visualizes gamma rays by superimposing the gamma-ray detection results from the radiation

detector and the images from the camera

29 Tokyo Electric Power Company Holdings Inc., “Estimation of the state of the reactor core and PCV in Units 1 to

3 of the Fukushima Daiichi NPS, and discussion on outstanding issues, the 5th progress report”

44

ensuring safety based on the safety-related features of the Fukushima Daiichi NPS and to share itwith the parties concerned.
Decommissioning work of the Fukushima Daiichi NPS containing the reactors involved in theaccident is an unprecedented activity that takes place in a peculiar environment different from thatof a normal reactor, and therefore, to ensure safety, the following characteristics (peculiarities)
regarding safety should be fully recognized:
・ A large amount of radioactive material (including α-nuclides that have a significant impact
in internal exposure) is in an unsealed state, as well as in unusual (atypical) and various
forms・ Barriers for containing radioactive materials, such as reactor buildings and PCVs, areincomplete
・ Significant uncertainties exist regarding the state of these radioactive materials andcontainment barriers, etc.
・ Difficulty in accessing the site and installing instrumentation devices to obtain on-site
information due to constraints such as high radiation levels on site・ Since the current level of radiation is high and further degradation of containment barriers
is a concern, it is necessary to take measures in consideration of the time axis without
prolonging the decommissioning activities
Based on these characteristics, NDF is organizing an approach to ensuring safety with thefollowing as the basis:• Optimization of judgement with safety assessment as its basis:
When making decisions by comprehensively considering technical reliability, reasonableness,
promptness, etc., safety assessments should be fully used to avoid a great variance in decisions
for safety measures (to avoid too excessive or too insufficient resource input).
• Ensuring timeliness in decommissioning activities:
While paying attention to the prevention of accidents and the mitigation of their impacts,
measures should be taken with concentration on the time axis so as not to prolong thedecommissioning period, in consideration of high radiological impacts that have already become
apparent, as well as further deterioration of containment barriers, etc.
• Complementing design by operating controls, monitoring, analysis, and on-site operation in theevent of an abnormality:
Due to significant uncertainties, there is a limit to addressing all situations by design alone. For
this reason, the information obtained at the operation stage, including that obtained throughmonitoring and analysis, should be utilized in design, and design should be complemented byoperators’ efforts and on-site operation to enhance safety in total with operations. In preparations
for abnormalities, consideration should be given to on-site response considering thecharacteristics that the progress of abnormalities is moderate and there is sufficient time torespond.
45

In addition, along with organizing the concept for ensuring safety, technical requirements havebeen established for ensuring safety of fuel debris retrieval and intensive studies are being
conducted as shown in the following Sub-sections 3.1.2.4.2 to 3.1.2.4.7.
3.1.2.3.1.1 Establishing the containment functions (gas-phase)
Dispersion of radioactive material in ordinary operating nuclear power plant is prevented bykeeping interior of a reactor building under negative pressure against the ambient air (active
containment function by negative pressure control) and pressure in a PCV is maintained to be equal
to that inside of a reactor building (passive containment function). However, the reactor buildings,
PCVs, etc. of the Fukushima Daiichi NPS were partially damaged by the hydrogen explosion andtheir containment function is deteriorated. Due to this, establishment of an active containment
function by negative pressure control is being considered during fuel debris retrieval work.
Moreover, from the perspective of prevention of hydrogen explosions due to steadily generatedhydrogen by the process of radiolysis of water and of corrosion (inactivation) of structural materials
due to the presence of oxygen, nitrogen is injected into the PCV to maintain it in a nitrogen
atmosphere. As for the exhaust from inside the reactor buildings, the release of radioactive
materials has been prevented by the PCV gas control system, which is furnished with filters toremove radioactive materials and measure radioactivity30.
We expect that existing safety systems will be able to cope with the retrieval of fuel debris, suchas gripping and sucking, in the case of a trial retrieval or gradual expansion of fuel debris retrieval.
In the subsequent work such as fuel debris cutting, it is necessary to construct the containment
function of the gas phase system in consideration of re-scattering of Cs, etc., that adhere to theequipment and structures in the PCV, aerosolization of water containing radioactive materials, andgeneration of short-lived iodine and noble gases if criticality should occur.
In addition to re-scattering of Cs, etc., the fact that dispersed fine particles (α-dust) containing
α-nuclides may be generated and the radioactivity concentration in the PCV gas-phase may
increase is a concern. Therefore, dispersion of α-dust from inside the PCV must be suppressed asmuch as possible, and a function for containing the gas-phases should be provided to make theradiation dose impact on workers and the public fall within the allowable value.
Accordingly, it is reasonable to expand the retrieval scale while understanding the tendency of
α-dust dispersion at each stage of expanding the fuel debris retrieval scale, and verifying theappropriateness of the containment function built in the subsequent stage. In TEPCO’s engineering
work, the improvement of dust monitoring installations inside the reactor buildings and the study oflowering pressure or reducing pressure in the PCVs using existing equipment are in progress basedon the outcome of the Project of Decommissioning and Contaminated Water Management. In the
future, the effect on the surroundings will be assessed based on the monitoring results of the

30 TEPCO, evaluation results of the additional release amount from the Units 1 - 4 reactor buildings (June, 2020),

the Team for Countermeasures for Decommissioning and Contaminated Water Treatment/Secretariat (80th)material 3-6, May 28, 2020
https://www.tepco.co.jp/decommission/information/committee/roadmap_progress/pdf/2020/d200528_11-j.pdf

46changes in condition such as the dispersion of α-dust associated with the work, and the retrieval
scale of fuel debris will be gradually expanded.
In the process, TEPCO is considering establishing a secondary containment function andstudying its necessity through their engineering work, while assuming the possibility of anincreasing impact on the surroundings.

Fig. 15 Example of containment function (gas-phase) by negative pressure control
In establishing the containment function (gas-phases), technical issues to be addressed
immediately for further expansion of fuel debris retrieval are as follows:(1) Understanding of the dispersion rate of α-dust
As described above, for the purpose of fuel debris retrieval work, it is necessary to collect datasuch as α-dust dispersion rate and to establish measures to suppress the transition of α-dust to thegas phase as much as possible based on the collected data.
To collect data such as the dispersion rate of α-dust, it is necessary to plan demonstration andconfirmation of the dispersion rate measurement during a trial retrieval and the gradual expansionof fuel debris retrieval. Moreover, in order to proceed with the review of technologies and R&D
activities for the fuel debris retrieval method/system under conditions where such demonstration
data have not been obtained, the general behavior related to α-dust dispersion should be roughly
understood. For this purpose, verifications are currently underway by using the mock-up debris fordeveloping analysis and estimation techniques to characterize the fuel debris.
In order to suppress the transition of α-dust to the gas phase, it is desirable to submerge thefuel debris and to fabricate it underwater as much as possible. However, the water level in the PCV
is to be adjusted with other technical requirements such as the building of the containment function47

of the liquid phase described in the next section. Therefore, not all fuel debris can be fabricatedunderwater, and transition of α-dust to the gas phase is considered to be mitigated by splashingwater on the fuel debris that is not submerged.
(2) Ascertaining the feasibility of negative pressure control in the PCVA.

Technical feasibility of negative pressure control based on the site conditions
In order to maintain negative pressure in the PCV, sufficient gas exhaust system capacity thatconsiders the condition of PCV damage is required. Although damaged parts have not been
identified yet, the exhaust capacity is currently established based on the relationship between
actual nitrogen supply volume and actual PCV internal pressure. At this time, it is necessary tomaintain sufficient differential pressure to respond to the internal pressure rise of the PCV due toabnormal events such as an internal temperature rise or shutdown of the gas exhaust system. In
order to achieve this, repair of the damaged parts of the PCVs will be considered as necessary,
but some difficulties such as remote work or exposure of workers are assumed due to work underhigh radiation dose conditions.
In this way, it is necessary to ascertain the technical feasibility of maintaining the negative
pressure in the PCV based on the site conditions and the information obtained during trial retrievalor the gradual expansion of fuel debris retrieval.
B.

Effect of air flow into the PCV during negative pressure control
Since air flows into the PCV in case of negative pressure control, measures of maintaining

inactivated condition in the PCV by increasing nitrogen gas supply into the PCV will be examined
as necessary based on evaluation on occurrence of accidents such as fire and hydrogen
explosion using accumulated information regarding volume of hydrogen generated by radiolysis
of water in the PCV and inflow volume of air (oxygen) into the PCV.
C. Study on the necessity of a secondary containment function
As illustrated in Fig. 15, for fuel debris retrieval, it is assumed that a working cell is newly
installed as connected to the PCV under negative pressure control, and the work from retrieving
the fuel debris to collecting the container for fuel debris retrieval into a transfer cask is performedin this cell. The PCV and this working cell have a primary containment function to prevent α-dust
to the exterior (out-leakage).
In addition to this, in order to respond to an event in which radioactive materials are dispersed
from the containment boundary caused by loss of primary containment function through negative
pressure control, the necessity of the secondary containment function has been investigated byinstalling building covers and containers in the existing reactor buildings and maintaining slightly
negative pressure inside of the reactor buildings to recover and treat radioactive materials.
However, a large capacity gas exhaust system is considered to be required to maintain negative
pressure in the secondary containment boundary since the reactor building has a large volume
and its leak tightness may have deteriorated due to the accident. Therefore, based on theaccumulated results of the tendency of dust dispersion obtained hereafter, it will be necessary to48

ascertain the required functions to establish a secondary containment function and to proceedwith research and development accordingly.
D. Deterioration control of the containment function of the PCV
In order to maintain negative pressure in the PCV during the fuel debris retrieval work, it isnecessary to manage risks of earthquakes and aging after considering how to handle
deterioration of the primary containment function consisting of the PCV and the attached cell. Thisis outlined in Section 3.1.2.3.1.5.
(3) Study on exhaust gas management
In control of exhaust gas associated with negative pressure control, it should be confirmed thatradioactive materials in gases that may contain nuclear fuel materials derived from fuel debris aremaintained below the radiation dose standard for the public in the vicinity of the facility bymeasuring and controlling the release concentration and the release amount. In addition, αnuclides derived from fuel debris should be included in the assessment and constantly
monitored/measured during fuel debris handling so as to evaluate their normal fluctuation range inadvance. By using such data, a system for early detection of abnormal events such as leakage andimplementing appropriate impact mitigation measures should be established for preventing any
impact on environment and workers.
The reliability and accuracy of the mechanical properties and chemical composition of the fueldebris needs to be improved because these are essential information for designing thedecontamination equipment for efficient collection of radioactive dust.
As described above, regarding establishing a containment function (gas-phase), TEPCO has
been ascertaining the feasibility of negative pressure control in the PCV through their engineering
work, and in parallel, has started to examine the necessity of a secondary containment function,and since 2020 consideration is being given to establishing a secondary containment function. In
the future, TEPCO will incorporate the results gained from the Project of Decommissioning andContaminated Water Management, trial retrieval and gradual expansion of the retrieval scale, andgive shape to the system specification necessary for establishing a containment function throughtheir engineering work. (See Chapter 3.1.2.3.2.3 for the progress status of installing covers on theside and top of the Unit 3 reactor building to improve its containment performance.)
3.1.2.3.1.2 Establishing the containment functions (liquid-phase)
To mitigate the dispersion rate of generated α-dust and to minimize the transition to the gas
phase, fuel debris cutting, etc., would be performed by pouring water over the fuel debris for fueldebris retrieval. Existing safety systems are expected to be capable of fuel debris retrieval bygripping and suction. For the subsequent work such as fuel debris fabrication and removal of
obstacles, a large amount of α-particles will flow into cooling water (liquid phase). To prevent thecooling water containing α-particles from leaking to the environment, it may be of great importanceto establish a cooling water circulation/purification system, and a liquid phase containment function
in consideration of prevention of spread of contamination (Fig. 16).
49

For this reason, it is necessary to examine technologies for removing soluble nuclides that may
be leached from fuel debris to the circulating cooling water as well as treatment technologies forsolid matter trapped by the filter equipped in the circulating cooling water system. Accordingly, thec Decommissioning and Contaminated Water Management31 has been promoting research anddevelopment. In parallel with this, the establishment of a PCV circulating cooling system that takes
water from the PCV and injects it into the reactor for cooling, which is beneficial in terms of
preventing the spread of cooling water containing α particles, was considered in research anddevelopment32 by the Project.
To establish a reasonable containment function of liquid-phase in each stage of the scale
expansion of fuel debris retrieval, it is rational to monitor the radioactive concentration of coolingwater by stage and verify the validity of the containment function to be built in the subsequent stage
based on the results (information on debris properties, etc.) obtained from research anddevelopment by the Project of Decommissioning and Contaminated Water Management. As withthe containment function (gas phase), from the viewpoint of verifying/investigating the impact of theretrieval work on the liquid phase, TEPCO, through engineering work, has been discussing system
addition/installation, etc., for the purpose of monitoring the circulating water system according to
the results of the Project33. With regard to the effects on the liquid phase during fuel debris retrieval
operations, the scale of fuel debris retrieval will be expanded gradually, based on the results ofmonitoring changes in the state of waste liquid containing α-nuclides. The water level in the reactor
building is required to be maintained lower than the groundwater level to prevent the outflow ofcooling water to groundwater and to appropriately control the water level in the PCV. Safetysystems are to be established taking this into consideration.
In establishing the containment function (liquid-phases), the technical issues to be addressed
immediately for further expansion of fuel debris retrieval are as follows:

31 IRID, additional subsidies in FY 2018 for the Project of Decommissioning and Contaminated Water

Management, “Technology development for further expansion of technologies to retrieve fuel debris/in-core
structures”, actual results in FY 2019, August 2020.
https://irid.or.jp/wp-content/uploads/2020/09/2019008kibonosaranarukakudai.pdf
32 IRID, additional subsidies in FY 2017 for the Project of Decommissioning and Contaminated Water
Management, “Development of technology to establish primary containment vessel water circulation system(full-scale test), final report for FY 2019, August 2020.
https://irid.or.jp/wp-content/uploads/2020/09/2019006mizujyunkan.pdf
IRID, additional subsidies in FY 2017 for the Project of Decommissioning and Contaminated Water
Management, “Development of technology to establish primary containment vessel water circulation system(full-scale test), final report for FY 2019, August 2020.
https://irid.or.jp/wp-content/uploads/2020/09/2019007mizujyunkanjitukibo.pdf
33 IRID, supplementary budget in FY 2016, “subsidies for the Project of Decommissioning and Contaminated
Water Management”, sophistication of retrieval method and system of fuel debris and internal structures, final
report in FY 2018, July 2019.
http://irid.or.jp/_pdf/20180000_13.pdf

50

Fig. 16 Build example of containment function (liquid phase)(1) Suppression of radioactive concentration increase in cooling water due to fuel debris retrieval
work
As an approach to assess the radioactive concentration in the cooling water, it is necessary toplan to investigate the impact on waste liquid when gradually expanding the scale of retrieval.
To suppress an increase of radioactive concentration in cooling water in the PCV, cutting
particles generated by fabrications such as cutting is planned to be collected through the PCV
circulating cooling system to mitigate dispersion of dust. It is recommended that the results ofmonitoring by the water circulation system are taken into account and incorporate them into
modifications of the PCV circulating cooling system as necessary in the stage of gradually
expanding the scale of retrieval.
(2) Setting of water level in the PCV
Due to a low seismic margin of S/C support columns, it is recommended that the water level in
the S/C is lowered, and thus, considerations are being given to lower the water level. In this instance,
it is necessary to appropriately set and manage the water level in the PCV with consideration forthe damaged condition of the PCV in each unit and prevention of the outflow of cooling water togroundwater (maintaining the water level in the reactor building below the groundwater level), etc.,
and to confirm that safety is ensured in terms of fuel debris cooling and control of dust dispersion.
As described above, regarding establishing a containment function (liquid-phase), based on the
results of the Project of Decommissioning and Contaminated Water Management (technology forremoving soluble nuclides that are considered to have leached from fuel debris into circulating
cooling water, technology for treating solids collected by filters of the circulating cooling water
system, etc.), since 2020 construction of a PCV circulating cooling system has been underinvestigation through TEPCO’s engineering. Going forward, based on the monitoring results
51

obtained from trial retrieval and gradual expansion of the retrieval scale, consideration will be given
to modifications required for the PCV circulating system while verifying the validity of thecontainment function to be established in the next phase.
3.1.2.3.1.3 Issues in cooling functions
Fuel debris generates heat due to the decay of radioactive materials. In Unit 2 after a lapse of
10 years since the accident, for example, the amount of heat generation has reduced to 1/1000 of
the level at the accident. However, the maximum heat generation is still estimated to be 69 kW34.
Therefore, unless cooling is continued, the surrounding materials gradually absorb the generated
heat and there is concern that the following events may occur:• Oxidation caused by increasing the temperature of the uranium oxide in fuel debris (increased
O/U ratio) proceeds resulting in volume expansion, crack generation, and thereby progressionof pulverization.
• Moisture in concrete structures is also dissipated and dried by heat, and then cracks occur todecrease the concrete strength.
• The inside of the PCV becomes dry, causing radioactive dust to be easily scattered and sent
airborne.
• When water is injected after the PCV becomes dry, the water that contacts fuel debris turns
into steam and raises the PCV internal pressure, which leaks from damaged areas withradioactive dust.
In order to prevent the above four events and excessive temperature rise of the fuel debris, thetemperature is being maintained below 100°C (cold shutdown state) currently by conducting
circulating cooling with water injection into the reactor. In FY 2019, water injection into the reactor
was temporarily suspended with the aim of optimizing operation/maintenance management ofcooling systems and emergency response procedures, etc. From FY 2020 onward, based on the
results of the water injection shutdown test conducted in FY 2019, testing will be planned andperformed depending on the purpose for the condition of each unit. Based on this policy, water
injection shutdown tests are being performed for each unit (Unit 1: 5 days from November toDecember 2020; Unit 2: 3 days in August 2020; Unit 3: 7 days in April 2021), and based on thecondition of lowering PCV water level during termination of water injection, it is planned to study
water injection methods for the future such as further reduction of the amount of water injection.
In addition, during the fuel debris retrieval work, it is necessary to keep the temperature below
the level at which the fuel debris retrieval device can continue to work without any problems for along period of time.
However, it should be considered that the injection of cooling water may become redundant in
the future due to further decrease in the amount of decay heat along with reduction in the remaining
amount of fuel debris, as the fuel debris retrieval progresses.
34 Kenji Nishihara et al., "Evaluation of fuel properties at the Fukushima Daiichi Nuclear Power Station”, Japan

Atomic Energy Agency (JAEA), JAEA-DATA/Code 2012-018(2012).

52

In maintaining this cooling function, the technical issues to be addressed for the time being
include setting of the target temperature inside the PCV to make each task feasible, as well as thecountermeasures to be taken under the assumption of cooling function abnormalities when eachtask is performed. While essential countermeasures would be to continue cooling by early recoveryof the cooling water circulation system or by mobile equipment, etc., it is necessary to evaluatechanges in the PCV internal condition based on the time margin in an emergency and to consider
emergency response measures and procedures, etc., including collection of devices.
As temperature monitoring in the PCV is also evaluated by the temperature of devices and wateraround the fuel debris, the fuel debris temperature is not directly measured. In preparation for
reducing the amount of cooling water, a temperature measurement of fuel debris must be made or,if direct measurement is difficult, examining techniques to estimate it from the temperature of thedevices and water used for the current assessment. Therefore, examinations are being madethrough research and development by the Project of Decommissioning and Contaminated Water
Management.
In addition, during fuel debris retrieval operation, the processing of cutting fuel debris while
spraying water is conceivable from the perspective of dust dispersion control, and attention should
also be paid to water level control inside the PCV, as well as controlling of the contaminated water
generated.
From the above, monitoring parameters and their criterion need to be studied and preparedthrough TEPCO’s engineering work in order to carefully promote fuel debris retrieval work whileobserving how this work will affect the existing circulating water cooling and purification system, as
well as its cooling function.
3.1.2.3.1.4 Issues in criticality control
At present, monitoring of the concentration of Xe-135, which are short-half-life fission products,
has shown no sign of criticality as the concentration remains lower than the criticality criterion of 1
Bq/cm3. In addition, the possibility of re-criticality of the fuel debris at the Fukushima Daiichi NPS
is presumed to be low based on the expected condition of the existing fuel debris in engineering,
because the alternation of molten fuel assemblies is not likely to reach criticality due to theabundance ratio with water, and the mixture of impurities, such as internal structures, can be
expected in the course of core meltdown. Furthermore, the fuel debris is presumed to be scatteredin a wide area beyond the core as a result of the accident progression. Even assuming thepossibility that control rods might have melted down before the fuel elements in the course of core
meltdown and that the optimum mixing of incidentally crushed fuel debris with water occurs, thepossibility of criticality is considered to be small.
Though the possibility of criticality is low, fuel debris retrieval alters the shape, etc., of fuel debris.
Therefore, it is essential to ensure reliable criticality management during retrieval without fail by
investigating what conditions would lead to criticality if shapes, etc., of fuel debris change, and toestablish an appropriate control method for ensuring prompt detection and shutdown in case of an
unexpected criticality.
53

In the initial stage of fuel debris retrieval work, fuel debris should be retrieved by limiting theprocessing amount based on methods that will not significantly change the fuel debris shape, suchas by gripping and sucking, as well as the estimated fluctuation of reactivity. Also, in the process of
expanding the retrieval scale and cutting, the retrieved volume of fuel debris will likely be increased,
while taking measures such as measurement of pre-work subcriticality and preparation for insertionof neutron absorbers. In addition, in the retrieval operations overall, unless criticality is unlikely tooccur in consideration of the retrieval conditions, it is necessary to ensure the safety of fuel debris
retrieval while evaluating for criticality by checking the amount of neutron flux fluctuation in thevicinity of the fuel debris, as well as reliable criticality prevention (criticality monitoring) thatcombines measures in design with operator monitoring and judgment to stop/resume work.
To maintain the subcritical condition more reliably for fuel debris storage after retrieval, it isimportant to store it stably while controlling the shape and size, for instance, by storing the debris
in containers. In TEPCO’s engineering work, the concept design is in progress for circulating water
system configurations and system specifications for further expansion of fuel debris retrieval.
Criticality assessment of equipment needs to be performed early in order to avoid rework at a laterstage, and criticality prevention measures that greatly affect the equipment specifications havebeen extracted in advance.
For this criticality control, technical issues to be addressed for the time being are as follows.
(1) Establishment of criticality evaluation methods
An evaluation method has been developed to obtain the information on the conditions forreaching re-criticality of fuel debris based on the information to be obtained from each stage of fueldebris retrieval, including internal investigations, and to estimate the conditions for subcriticality
and the degree of influence of criticality if it should happen. In conducting these evaluations, a planshould be made so that information on the critical parameters with high impact on the criticality
evaluation can be obtained in the course of internal investigation and retrieval work, and be revised
appropriately through information updates as needed for incorporating on-site information during
preliminary work into safety assessment in later stages.
(2) Local neutron measurement around retrieval point
There are various kinds of existing neutron detectors according to application (fission chamber,
B-10 proportional counter tube, semiconductor detector, etc.). Taking advantage of each feature,
selection of neutron detectors for each stage has been considered. The required specifications forthe neutron detectors for criticality monitoring are; (1) ability to survive the accumulated radiationdose (Gy) for the operation period and (2) installability of the assumed equipment (size/weight,
cable diameter) or installability at a work site (size/weight, cable routing); and (3) the guaranteed
detection efficiency at the required level (time, accuracy). Accordingly, optimal detector selection
will be made according to the information on PCV radiation dose rate obtained by internal
investigation and the progress of the equipment development of each unit.

54

In addition to subcriticality measurement, a small detector is highly likely to be used as astandalone detector for constant monitoring of local neutron measurement. Toward practicalapplication of local and constant monitoring, studies are underway in the Project ofDecommissioning and Contaminated Water Management, including specifications of neutron
detectors and their actual location and quantity, and approach-to-criticality monitoring technologies
in combination with evaluations based on the obtained data. In FY 2020, confirmation testing plan
was formulated to verify the applicability of neutron detectors to the on-site environment. As specificoperation

methods,it

isnecessary

to

set

up

determination

criteria

onoperationsuspension/resuming if fluctuation of the neutron flux is detected, and on boron injection as aneutron absorber.
The possibility of criticality needs to be examined in places other than retrieval locations. For
example, criticality may occur in places where fuel debris cutting particles accumulate and cannot
be collected in the circulating water cooling system (i.e., outside the PCV bottom pedestal, in piping,water filters, waste water receiving tanks, etc.). Although criticality can be detected by the PCV gas
management system, countermeasures will be considered in accordance with the feasibility of
approach-to-criticality monitoring, criticality risk scenarios and evaluation results.
(3) Feasibility study of measuring the degree of subcriticality
When measuring the degree of subcriticality, in addition to the required specification of (2), it isnecessary to select a highly sensitive detector in order to measure a weak neutron signal, capturing
a neutron fluctuation in a very short time under a gamma ray environment. From examination
performed to date, considering the equipment mountability (size/weight/electromagnetic noise) andoperation methods (measurement time and duration, installation location, etc.) by sensitivity are
key issues due to the necessity of lead shielding in a high ɤ-ray environment (assuming 1000 Gy/h).
In the future, it will be necessary to consider neutron detector selection and optimization, applying
the constraints (weight, size, cable handling, interference with arms, balance between
measurement and fabrication time, etc.) introduced by the fuel debris retrieval methods andsystems. Meanwhile, improving the actual site applicability of the detectors is under review in lightof an approach to assess the ɤ-ray radiation dose rate and neutron count rate in the vicinity of thefuel debris as well as constant monitoring, and by downsizing with the combined use of several
detectors, etc.
In addition, in order to verify the applicability to fuel debris, where various mixes of compositions
and properties of the fuel debris are expected, it is necessary to assess the technical feasibility,
including applicability, by planning and demonstration.
(4) Feasibility study of neutron absorber
Based on the information obtained in each stage of scale expansion, in preparation for cases
where the level of fuel debris criticality is high or the applicability of measuring the degree ofsubcriticality has limitations, assessments on the required boron concentration and feasibility
engineering studies on installations are in progress by TEPCO for cases of filling with sodium
55

pentaborate during normal fuel debris retrieval. As a result, environmental impact in the event ofleakage and compatibility with concrete as structural materials has been evaluated35.
While examining operation details to maintain the boron concentration in consideration of the
impact on the PCV circulation cooling system as well as the impact on systems and waste during
segregation, collection, re-use and processing of boric acid, it is necessary to verify the applicability
on site for the case of sodium pentaborate injection, together with the approach-to-criticality
monitoring technologies described in (2) Local neutron measurement around retrieval point.
In addition, in the case that criticality should occur and emergency sodium pentaborate injection
would be used to achieve a subcritical state, the methods for maintaining subcriticality after
transition (lowering the water level, maintaining the boric acid concentration, etc.) as well as
recovery methods need to be examined.
Moreover, to ensure maintenance of subcriticality when the margin for criticality of fuel debris issmall, development is also underway for a non-soluble neutron absorber that can locally limit the
impact on the PCV circulating cooling system. Until now, fundamental property testing and radiation
resistance testing have been conducted to list B4C metal sintered materials, glass containing B/Gd,
Gd2O3 particles, and sodium silicate/Gd2O3 granulated powder as candidates for non-soluble
neutron absorbers. For these candidates, the impact of long-term irradiation on the integrity ofcontainers during storage of fuel debris, the method of spraying on fuel debris for debris crushing,
and the effect after spraying have been verified. In FY 2020, the feasibility of a boric acid control
system for a circulating water system was examined. In the future, technical development for onsite operation in combination with neutron detectors will be promoted.
To introduce non-soluble neutron absorbers, it is necessary to study the impact on PCV
corrosion and environmental impact in the event of an environmental release.
(5) Detection of criticality by PCV gas management installations
Immediate criticality monitoring and sophistication of detectors in the PCV gas monitor isrequired to detect when approaching criticality and criticality in the vicinity of retrieval, and to detect
criticality due to the fall of fuel debris and/or accumulation of powder debris in locations other thanfuel debris retrieval. By measuring Kr-87/88 with high trackability for reactivity changes, in addition
to Xe-135 that has already been measured, criticality detection can be accelerated and the level ofsubcriticality of the entire PCV can be presumed. Therefore, a method of practical application onsite in combination with detection by neutron detector needs to be examined in the future36.
3.1.2.3.1.5 Issues in the structural integrity of PCVs and buildings
As for the main equipment in the PCV/RPV pedestal, etc., and reactor buildings, their structural
integrity has been evaluated in post-accident studies by TEPCO and the Project of
35IRID, supplementary budget in FY 2017, subsidies for the Project of Decommissioning and Contaminated Water

Management, sophistication of retrieval method and system of fuel debris and internal structures (Technology
development for establishing criticality control methods), final report, July 2019.
http://irid.or.jp/_pdf/20180000_04.pdf
36IRID, supplementary budget in FY 2017, subsidies for the Project of Decommissioning and Contaminated Water
Management, “Sophistication of retrieval method and system of fuel debris and internal structures (Technology
development for establishing criticality control methods), final report, July 2019.
https://irid.or.jp/_pdf/20180000_04.pdf

56

Decommissioning and Contaminated Water Management. As a result, it has been confirmed thatthe main equipment and reactor buildings have a certain level of seismic margin.
Hereafter, the existing main equipment and reactor buildings, as well as equipment/systems andbuildings (including modified areas of the existing equipment/systems and buildings) to be newly
installed for fuel debris retrieval over a relatively long period, should satisfy the functional
requirements and (1) be capable of performing operations safely and (2) ensure the required levelof safety against external events such as earthquakes and tsunamis. Assuming (3) long-term
maintenance management, in addition, it is important to (4) feedback new knowledge to be gained
from planned PCV internal investigations and debris analysis results, etc., to the design of fueldebris retrieval systems and the study of retrieval methods. The following shows the key functional
requirements as examples.
 Existing equipment and buildings (including modified areas. The impact of aging is also
considered as necessary)• Control deterioration of containment functions of PCV, RPV and reactor buildings, etc., andcontrol/prevent large release of radioactive materials (maintaining containment functions).
• Reactor buildings, etc., safely support equipment/systems to be newly installed in the reactor
buildings for fuel debris retrieval in addition to the existing main equipment (maintaining
support functions).
 Equipment/systems and buildings to be newly installed for fuel debris retrieval (including
connections to the existing equipment/systems)• Have functions according to design requirements and control/prevent large release ofradioactive materials (ensuring containment functions).
• Safely support equipment/systems to be installed for fuel debris retrieval (ensuring support
functions).
• New buildings, etc., provide a safe work environment as required (ensuring shielding
performance, etc.).
In FY 2020, TEPCO has formulated a long-term maintenance management plan for existing onsite systems/equipment and buildings in consideration of the progress of time-related deterioration,and started implementation of the plan. When new facts about the accident are revealed

through further investigations and other activities, it is also necessary to clarify the impact of
the accident, especially damage, by performing severe accident progression analysis
evaluations, etc. and to secure functions throughout the decommissioning period in
consideration of the progress of time-related deterioration. Moreover, regarding the existing
equipment/systems and buildings, the cooling function was maintained during the earthquakewith its epicenter off the coast of Fukushima Prefecture that occurred on February 13, 2021,37
37 Earthquake with its epicenter off the coast of Fukushima Prefecture. The maximum intensity of upper 6 wasobserved in Miyagi and Fukushima Prefectures. At the Fukushima Daiichi NPS, a seismometer installed on the57although lowering of the water level in the PCVs of Units 1 and 3 was confirmed. In light of thisearthquake, in order to maintain and manage the equipment/systems and buildings with the above
functions over the medium-and-long term, it is necessary to conduct impact assessments on theaccident impact, aging and external events (earthquakes and tsunamis, etc.) anticipated during
the decommissioning period. In view of the fact that the past assessment of these effects waslimited, it is necessary to make maximum use of existing techniques and evaluation results forplanning and implementing an investigation plan, in which remote control under a high radiation
environment is a challenging issue, and to develop underlying technologies to understand thesituation. In so doing, while giving priority to safety, it is useful to actively introduce the latest
knowledge and achievements not only in the nuclear field but also in other fields.
As for existing and new equipment/systems and buildings, the loading conditions (layout, size,
weight of the new equipment/systems, new openings on PCV/biological shielding walls, etc.) during
fuel debris retrieval will be specified with further progress in designing. In order to ensure thestructural integrity of equipment/systems and buildings, while considering the state of the site,examination will be promoted steadily based on the latest design information.
In the specific designing of new equipment/systems and buildings, it is important to define
seismic classes and perform seismic evaluation accordingly. However, it is still challenging to repair
and reinforce buildings and main equipment damaged by the accident in a high radiation dose
environment. Because of that, earthquake ground motion and design criteria used in the design will
be appropriately defined, taking into account the perspective of risk assessments.
3.1.2.3.1.6 Issues in reduction of radiation exposure during work
In accordance with the Mid-and-Long-term Roadmap and TEPCO’s Mid-and-Long-term
Decommissioning Action Plan, removal of obstacles and radiation dose reduction in the reactor
buildings are in progress as improvement of the work environment in work areas/access routes. In
the future, as work related to fuel debris retrieval, reduction of exposure during work such as
removal of high-radiation dose equipment, etc. is an issue, and research and development has
been promoted by the Project of Decommissioning and Contaminated Water Management tosupport TEPCO’s engineering.
The main work areas related to fuel debris retrieval are high radiation dose areas such as inside
the reactor buildings. Also, there comes the need to handle nuclear fuel materials containing αnuclides from fuel debris with a large dose contribution in the case of internal exposure. Accordingly,
enhanced control of not only for external exposure but also for internal exposure is essential for
reduction of exposure.
Specifically, it is important to prevent excessive exposure to workers through appropriate
radiation protection schemes depending on the working environment and working style. Regarding
protection from external radiation exposure, the radiation exposure dose is evaluated considering
the radiation sources and the radiation dose rate in the work area. Then, based on the three
2nd basement floor (on the foundation plate) of the Unit 6 reactor building recorded the quake with its maximum
acceleration of 235 gal. This is equivalent to the response level of about half of the seismic response analysis
results of the buildings against the design basis earthquake ground motion (Ss).

58principles, namely “time, distance, and shielding”, taking measures to reduce radiation exposure to
as low as reasonably achievable will be needed.
Therefore,anappropriate combinationofexposure

reductionmeasures such as

decontamination, shielding, remote technology etc. is to be selected, with the following ideas inmind.
・ Priority should be given to reducing radiation exposure through a combination of remotetechnologies and decontamination. Then, plan on-site radiation exposure management forsite workers by the “time, distance and shielding” approach.
・ In the extremely high-radiation dose areas such as inside the PCV and torus rooms, work
should be pursued by remotely controlled machines, etc. to avoid engaging personnel
inside.
・ With regard to the inside of the reactor buildings, except for the areas mentioned above,
consideration should be given to the optimal combination of decontamination, shielding,removal of unnecessary objects, remote technology, and reduction of working time in orderto keep the accumulative radiation dose for the entire project at a low level.
・ 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 radiation dose rate in
the target areas, type of contamination, space for work, frequency of use, applicability anddevelopment situations of remote technologies, schedule and cost, etc.
・ 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 non-specific“betterment-oriented” manner such as to aim for an overall reduction of radiation dose.
Regarding protection from internal exposure, measures such as suppressing dispersion ofradioactive dust and prevention of the spread of contamination are being taken and appropriate
protective measures are to be selected depending on the target nuclides, airborne concentrationand surface contamination density in the work area, to prevent inhalation ingestion and bodycontamination. In the event of intake, the effective radiation dose should be assessed using
external counting (lung monitor) and bioassays. For this reason, it is important to select α-nuclides
that are important for exposure assessment in advance, and to incorporate them into the control ofairborne concentrations, standards for wearing protective equipment, and equipment calibration
management. Controlling the surface contamination density in the work environment and the bodyof workers entering/leaving contaminated areas is also important to early detect the spread ofcontamination beyond the area division and to prevent an intake of re-suspended dust from loose
contamination.
With the objective of dose reduction in long-term decommissioning, it is important to accumulateknowledge such as on-site operation experiences and lessons learned and to hand down knowhow.
59For further expanding fuel debris retrieval, it is necessary to develop a database that enables
sharing of information and prompt feedback for the next work plan.
In particular, in fuel debris retrieval operation, access to the PCV should be made from thepenetration X-6, etc., after the work environment in the reactor building is sufficiently secured. To
reduce the radiation exposure of workers in the reactor building, it is important to conduct sufficientinvestigations on the radiation dose distribution and state of contamination, including thecontribution from the surroundings of the subject areas, to identify the source locations and intensity
as much as possible and to build the radiation dose reduction plan. Upon adequate verification on
the operation feasibility, the target dose rate in the work areas and access routes shall be set inconsideration of the margin for the radiation exposure dose limit (50 mSv/year and 100 mSv/5
years) for workers specified by laws and regulations. In the radiation dose reduction plan for highradiation dose areas, it is important to take management measures to reduce the total radiation
exposure dose to as low as reasonably achievable and accomplish operations with respect to work
hours in accordance with dose limits and required work hours to accomplish operations. Based onthese, and as R&D tasks by the Project of Decommissioning and Contaminated Water
Management, the development of technologies to identify radiation sources using environmental
survey data and to digitize the environment and radiation source distribution visualized by digitaltechnology, for the formulation of safe and efficient work plan has been in progress since FY 2021.
Moreover, in developing remote technologies for environmental improvement and removal of
obstacles under high radiation dose, the obstacles to be removed have been selected and theelemental technologies have been extracted in accordance with the required functions, andtechnical investigation is underway since FY 2020.
In addition, adequate radiation exposure control should be conducted after formulated a longterm work plan that includes not to concentrate workers’ radiation exposure on individual workers
and to help reduce whole workers’ radiation exposure. Support should be provided to develop a
database that can improve work plan efficiency and radiation exposure control, and to establish asystem that manages and operates various information on the entire Fukushima Daiichi NPS in an
integrated, step-by-step manner.
3.1.2.3.2 Technical issues related to fuel debris retrieval methods
3.1.2.3.2.1 Issues in securing access routes
For carrying in, installing, and carrying out devices and equipment used for fuel debris retrieval
work, and transporting fuel debris and waste, access routes should be established by removing
obstacles on the access routes and reducing the radiation dose in the R/B to the level at whichsuch tasks can be performed. When establishing new openings in the PCV or the like to constructaccess routes to fuel debris, suppression of the release of radioactive materials from the PCV and
RPV and maintaining the integrity of existing structures should be kept in mind taking into accountof the gas phase containment function described in Sub-section 3.1.2.3.1.1.

60

The Mid-and-Long-term Roadmap indicates that the first implementing unit would be Unit 2 andtrial retrieval begins toward gradual expansion of fuel debris retrieval. Accordingly, TEPCO iscurrently proceeding with specific engineering studies to conduct an access route from penetration
X-6 in Unit 2.
On the other hand, toward a further expansion of fuel debris retrieval, studies are underway on
the construction of access routes from the side opening of the PCV to fuel debris (the side accessmethod), based on the results of research and development conducted to date by the Project ofDecommissioning and Contaminated Water Management. In the side-access method, the issue isto address containment, shielding for connecting structures between newly installed heavy
structures and the side-opening of the PCV, and seismic displacement, and technical developmentof the method is underway using lightweight cells and fixed rails as well as access tunnel systems.
As for the construction of access routes including top access (top-access method), in addition
to side access, the technology for removing obstacles and transportation methods that can shorten
the preparation processes for retrieval are under study in order to enhance throughput. Since FY
2020, the Project of Decommissioning and Contaminated Water Management has been examining
the feasibility of a method to cut, retrieve and transport interfering structures as a single or large
unit while ensuring containment and shielding. The NRA pointed out the possibility that a large
amount of Cs might exist on the undersurface of the shield plugs of Units 2 and 3 from theirevaluation of TEPCO's radiation dose measurements on the operating floor. Therefore, the topaccess method requires consideration of constructing access routes with this in mind.
In the future, based on the above-mentioned issues, it is necessary to clearly define the accessroute to be built at the next stage from the data obtained at each phase of scale expansion. In
particular, when cutting the inner door of the penetration X-2 in unit 1, the dust concentration in the
PCV increased more than expected before the start of the work, and the pressure in the PCV
dropped while installing a camera chamber for investigating obstacles. Therefore, not only
countermeasures to prevent dust dispersion, but also the extra time required for responding towhen faced with such an unexpected situation need to be considered and planned.
Since the fuel debris retrieval policy stipulates that the optimum combination of retrieval methods
should be selected depending on the location where fuel debris exists for each reactor Unit, it isimportant to proceed with research and development toward a planned scale expansion.
3.1.2.3.2.2 Issues in development of devices and equipment
In each phase of trial retrieval, gradual and further expansion of fuel debris retrieval, devices
and equipment for fuel debris retrieval need to be developed with emphasis on safety, reliability,and efficiency. To flexibly respond to the situation inside the RPV and the PCV bottom where thefuel debris is predominantly present, the specifications of devices/equipment to be developed inthese phases should be established in consideration of radiation resistance, dust resistance,waterproofness, range of temperature, remote inspection/maintainability, remote operability,
securing visual field, seismic resistance, protection mechanism for collision avoidance or automatic

61

shutdown in case of abnormality, high reliability, appropriate redundancy, a rescue mechanism that
does not disturb the subsequent work when a problem occurs, and efficiency of fuel debris retrieval.
Equipment development for trial retrieval and gradual expansion of fuel debris retrieval has
progressed as part of research and development of the Project of Decommissioning andContaminated Water Management. After the gradual expansion of fuel debris retrieval, TEPCO
needs to take over and substantiate the development results. TEPCO is proceeding with theengineering of the robot arms, etc., to be applied to Unit 2, while preparing education/training
scheduled to be started from 2021 for the fuel debris retrieval operations using these remotedevices. Prior to installing remote equipment such as robot arms on site, adequate performanceverification and operation training are essential by using mockups simulating the expected PCVinternal environment. For this purpose, development of mockup systems is required.
As for devices and equipment for further expansion of fuel debris retrieval, development isunderway for retrieval methods for improving efficiency, retrieval/handling systems according todiverse conditions of fuel debris, and dust collection systems for dust generated during fuel debris
fabrication. Specific examples of ongoing development activities include: research anddevelopment for removal of large-scale structures by the top access and removal of obstacles by
top and side access methods; development of fuel debris cutting/fabrication systems (mechanical,
thermal) and dust collection/dispersion control systems; development of retrieval/collection
systems according to diverse conditions of fuel debris (fragments, sludge, fine powders, etc.);
remote operation support for robot arms, etc.; on-site transportation of unit cans; removal of soluble
nuclides in circulating cooling water; processing of deposits collected from the PCV; and predictionof dust behavior in the PCV. Furthermore, techniques for installing the devices and equipment used
for fuel debris retrieval are required. Furthermore, techniques for installing the devices andequipment to be used for fuel debris retrieval are required. On the assumption of remote operation,
research and development has been in progress since FY 2020 for installing work cells to establish
radiation shielding and gas phase containment functions and for connection methods with theexisting structures. Since the stage of further expansion of the retrieval scale is prolonged, and inorder to ensure safe, efficient and continuous retrieval work, it is also important to develop
integrated support technologies for decommissioning of the Fukushima Daiichi NPS, such as
development of rational maintenance technologies for a wide variety of remote devices andequipment, and system development to continuously monitor environmental changes in the PCV
along with work progress. These development activities have been promoted since FY 2021.
TEPCO is currently conducting a concept study for selecting a method for further expanding thescale of retrieval, and plans to narrow down the number of methods for the subsequent design
study by the end of FY 2021. Further development of devices and equipment should be plannedand promoted in light of the selected method.
As for how to proceed with development, it is necessary to flexibly promote operations in thesubsequent phase based on the information gradually obtained from preceding investigations andretrieval work, and to continue development for emerging important issues. The developed devices
62and equipment need to be combined as a system and undergo a series of mockup tests todemonstrate that they can realize their performance safely and reliably at the actual site. These
mockup tests need to be implemented in a facility simulating the on-site environment in order toverify the applicability of remote equipment and operability/maintainability of the entire remotesystem under severe environmental conditions containing significant uncertainties. Therefore, in
cooperation with the organizations concerned, NDF and TEPCO have been engaged in examining
how to proceed with the remote mockup test plan, the test plan review, the scope of the mockup
facility to be maintained, the time required, and operation management, etc. From 2021, it isexpected that TEPCO will take the initiative in promoting examination and materialization.
3.1.2.3.2.3 Issues in system installations and working areas
Assuming that safety functions are ensured, and considering avoidance of excessive system
specifications, it is necessary to examine the establishment of system installations, etc., takenecessary measures such as system additions based on the results of such examination, and thento operate them properly. In carrying out examination, sufficient areas should be secured to satisfy
the required environmental conditions while considering installing shields for reducing radiation
exposure for workers in addition to system installation, operation/maintenance management.
The system installations include a negative pressure control system required for establishingcontainment functions (gas phase), a circulating water cooling and purification system required formaintaining the containment functions (liquid phase) and cooling functions, and a criticality control
system required for controlling criticality. Moreover, realization of measurement systems (forpressure, temperature, water level, radiation, etc.) to monitor the PCV internal state is a significant
issue, which is essential for fuel debris retrieval. In order to build safety systems incorporating theabove, TEPCO, through its engineering work, has tentatively set assumptions (system design
conditions) based on research and development by the Project of Decommissioning andContaminated Water Management, and has been examining system design and layout.
Fig. 14 shows an example case of research and development for the liquid-phase system by
the Project38. In this project, the system configuration and the outline of its specifications for particle
collection/removal are under consideration. In engineering work by TEPCO, a basic plan for system
configuration has been developed in reference to the results of the Project. A case study is also
conducted on the feasibility of the system layout (installing devices in the reactor building or
installing only some devices in the reactor building if there are restrictions on arrangement) under
the reactor building environment (ease of securing equipment installation space, radiation dose in
the subject area, etc.). Based on the result of this examination, further investigation will be given
on feasibility as a safety system. As described above, in establishing safety systems, steady efforts

38 IRID, supplementary budget in FY 2016, subsidies for the Project of Decommissioning and Contaminated Water

Management, sophistication of retrieval method and system of fuel debris and internal structures, final report,
July 2019.http://irid.or.jp/_pdf/20180000_13.pdf

63are required to improve the quality of design by repeatedly adjusting system layout design, andreconsidering device specifications as the situation demands.
It is important to ascertain the feasibility of systems designed to secure the safety functions
necessary for proper operation of equipment and devices for safe retrieval and storage of fueldebris, and to steadily proceed with studies on the fuel debris retrieval method through a reliableapproach.

(Source : IRID)
Fig. 17 Example of examining the liquid-phase particle (collection/removal) in the Project ofDecommissioning and Contaminated Water Management
To further expand the retrieval scale, the working area required for installing fuel debris retrieval
equipment/related devices and system installations is now being investigated in TEPCO’s
engineering work. In considering the handling of high radiation dose areas in the reactor buildings
and interference with other tasks, installing systems outside of the existing buildings is also being
considered.
In order to improve the containment performance of the Unit 3 reactor building, constructing
covers on the side and upper part of the building is under investigation. In this process,consideration is given to the surrounding environment in addition to securing the space necessary
for cover construction. As the radiation dose around the reactor building is currently high,construction of the cover by remote operation is under discussion. Verification that simulates theactual condition is required to achieve accurate construction of the cover. However, it is expectedthat there will be a limit to improvement of airtightness. Therefore, in parallel with the investigationof the structure and workability, the environment around the reactor building should be improved64

from a comprehensive viewpoint so that workers can be engaged in key operations and airtightness
can be enhanced.
3.1.2.3.3 Technical issues related to safe and stable storage of fuel debris
3.1.2.3.3.1 Issues in handling fuel debris (containing, transferring, and storing)
Before initiating fuel debris retrieval work, a comprehensive system should be established thatconsists of a series of steps from containing and transferring to storing of retrieved fuel debris
furnished with safety functions such as maintaining subcriticality, containment functions,countermeasures against hydrogen generation, and cooling. Accordingly, examination of thefollowing is being progressed until the end of FY 2020:39
・ Development of basic specifications for the container, such as overall length in considerationof handling, internal diameter, quality of materials and lid structure in light of work efficiency
and maintaining subcriticality, etc., and demonstration of the structural integrity of thecontainer by testing.
・ Examination of a practical and rational prediction method of hydrogen generation from fueldebris stored in containers; determination of a vent mechanism for hydrogen gas release on
the container lid by using the said prediction method; and establishment of safe transferring
conditions with consideration for accumulation of hydrogen gas in transferring casks.
・ Development of efficient drying technology applicable to fuel debris in unit cans, andconsideration on a drying system using this technology.
Moreover, in reference to the results of these studies, TEPCO continues their activities tomaterialize systems and devices/installations used for the process from containing to storing fueldebris, which are required for gradually expanding the scale of retrieval, in coordination with otherassociated projects. In addition, specific transferring routes, storing technologies/forms and its
locations in light of the usage plan for the entire site are taking shape.
Since information and knowledge on the fuel debris properties is limited, the systems anddevices/installations will be designed by conservatively assuming the fuel debris properties.
Therefore, for rationalization it is important to incorporate to the extent possible the information anddata on fuel debris properties (amount of hydrogen generated, data on fuel debris properties), as
well as knowledge and experience on handling fuel debris during the operations from receiving thefuel debris by containers for on-site transferring to temporary storage that can becollected/accumulated during trial retrieval and gradually expanding the scale of fuel debris retrieval
work into the design of systems and facilities for containing, transferring, and storing fuel debris atfurther expansion of fuel debris retrieval.
Unit cans and containers in which fuel debris is stored should be handled and maneuvered bya remote device in a safe and reliable manner. Therefore, the assumed mockup operations need
39IRID, supplementary budget in FY 2018, “Subsidies for the Project of Decommissioning and Contaminated

Water Management (Development of Technology for Collection, Transfer and Storage of Fuel Debris)”, FY 2020
final report, June 2021
https://irid.or.jp/wp-content/uploads/2021/06/2020008syuunouisouhokanFIX_20210615r2.pdf

65

to be executed (e.g., handling of unit cans and containers as well as fuel debris sampling foranalysis by using actual or similar equipment/device including a remote device) in the initial stageof detailed design. Moreover, in terms of preventing design change/modification, the approachthrough the mockup operations, which determines the specifications/sizes of these remote devices
and the devices/equipment required for containing, transferring, and storing fuel debris, their layout,
and the flow of fuel debris handling, is considered to be useful. In developing the specificinstallations and systems for handling and storing retrieved fuel debris, it is also necessary to giveconsideration to the installations with respect to applying safeguards requirements.
The Mid-and-Long-term Roadmap stipulates that the processing/disposal method of theretrieved fuel debris shall be investigated and fixed during the third phase after starting the fueldebris retrieval work.
3.1.2.3.3.2 Issues in sorting out fuel debris and radioactive waste during fuel debris
retrieval
In the fuel debris retrieval work, obstacles and structures to which molten fuel are partiallyadhered will also be retrieved from the PCV in addition to fuel debris in which molten core fuel are
mixed with metals and solidified, and compounds (MCCI product) produced by mixing molten corefuel with concrete at the PCV bottom. Of these, if substances on which a small amount of molten
fuel is adhered are all deemed as fuel debris, the amount would be enormous. It may become an
obstructive factor in advancing decommissioning because scales of facilities and sites for fueldebris storage become larger. This requires the development of sorting techniques for fuel debris
and radioactive waste, i.e., sorting scenarios (sorting is performed in which process from retrieval
to storage), sorting criteria and necessary measurement technique/devices.
It is determined that those requiring special attention and facilities/systems for handling andstoring to maintain subcriticality can be considered as fuel debris. For that purpose, it isrecommended to aim for sorting out fuel debris based on the measurement results of the amountand concentration of nuclear fuel materials. The following studies were conducted in response to
this40.
・ Consideration of which steps in the operation processes, from retrieval to storage, sorting isfeasible to separate retrieved materials from the PCV (fuel debris, structures, etc.) into fueldebris and radioactive waste (consideration of sorting scenarios)
・ Investigation of techniques/equipment that may be capable of measuring the content of
nuclear fuel materials contained in the materials retrieved from the PCV (investigation of
possible measurement techniques)
Based on these studies, it is currently considered as incredibly difficult challenge to measure or
estimate the amount and concentration of nuclear materials in the retrieved materials from the PCV,
which required innovative technology development.
40IRID, supplementary budget for FY 2018, subsidies for the Project of Decommissioning and Contaminated

Water Management, “Technology development for further expanding the retrieval scale of fuel debris/in-core
structures”, actual results in FY 2019, August 2020.To be revised once the 2021 report is released.
https://irid.or.jp/wp-content/uploads/2020/09/2019008kibonosaranarukakudai.pdf

66

In developing measurement techniques and devices, it is important to understand themeasurement errors first. Major factors affecting measurement errors include the fuel debris
properties and the condition and location of nuclear fuel material in fuel debris storage condition inunit cans and containers, and there are many factors other than errors depending on themeasurement device itself. However, the extent of such impact is highly uncertain at present due
to a lack of knowledge on the fuel debris properties and the storage condition of fuel debris.
Therefore, in parallel with the development of measurement technology and actual measurement
device by iterating actual measurement using mockup fuel debris, etc., it is considered beneficial
for R&D in terms of cost and time saving to identify factors influencing measurement errors and its
strength accumulated by a number of numerical experiments by computer and incorporate findingssuch as the extent of influence into the development of actual measurement technologies/devices.
Thus, the impact of fuel debris with different properties and storage conditions on the measurement
errors as well as modifications/improvements (e.g., specifications of shielding materials and theirinstallation location) of measurement techniques/devices to reduce measurement errors can also
be examined through numerical experiments by computer. Such research and development
activities have just started since FY 2020 in the Project of Decommissioning and Contaminated
Water Management, and this R&D is greatly desired to be accelerated in the future.
Knowledge on the properties and actual storage conditions of fuel debris is anticipated to beaccumulated through trial retrieval and gradually expanded-scale of fuel debris retrieval. By
comparing the results of numerical experiments using the improved analytical model and expanded
analysis condition based on these accumulated knowledge with the results of actual measurements
using mock fuel debris manufactured based on the knowledge on the properties of actual fuel debris,
it is possible to improve the accuracy of measurement technologies/devices and to further
accelerate its development.
It is desirable that development of measurement techniques/devices required for sorting fueldebris and radioactive waste is to be continued in this manner. It is also important to continue
activities to enhance the effectiveness and practical applicability of sorting methods (sorting criteria,
scenarios, measuring techniques/devices) by leveraging knowledge and information on the fueldebris properties obtained by planned PCV internal investigation and analysis results of fuel debris
samples taken during trial retrieval, and the gradual expansion of fuel debris retrieval and so on.
3.1.2.3.3.3 Issues in examining safeguards strategies
At the Fukushima Daiichi NPS, the safeguards implemented before the accident are notapplicable due to the damage on fuel assemblies and destruction of reactor cores and reactor
buildings. The appropriate safeguards have been applied along with the progress of thedecommissioning process through cooperation and information sharing between Japan
Safeguards Office (“JSGO”) of the NRA, International Atomic Energy Agency (“IAEA”) and TEPCO.
As a result, the IAEA and JSGO confirmed that there had been no diversion of nuclear materials
and undeclared nuclear materials or nuclear activities (For the concept of the safeguards, refer toAttachment ).
67

New material accountancy and safeguards are expected to be applied to the retrieved fuel debris
depending on its properties and state. As it is unprecedented, TEPCO may face technical issues
in examining and applying them to the site.
In response to that, NDF will conduct wide-ranging surveys on existing technologies related tomaterial accountancy and safeguards for all nuclear materials handled at the Fukushima Daiichi
NPS to prepare in case that TEPCO needs technical assistance. NDF will also check the progressof the project from an engineering perspective to confirm that the application of safeguards tosystems has not affected the decommissioning process.
3.1.2.4 Summary of key technical issues
The main technical issues and plans described in this section are summarized as shown in Fig.
18.

68

Short term (approx. 3 years)
24
Fiscal Year

2021

2022

Gradual expansionof fuel debris retrieval(Unit 2)25

Mid-to-Long-term (FY2024 to End of FY2032)
26
27
28
29
30
31
2032(year)
2026▽Start of the fuel debris retrieval in theinitial unit (by the end of 2021)* Due to the impact of coronavirus infection,
the start will be delayed in about 1 year.

RM milestones

Trial retrieval(Unit 2)2023

Improve environment
inside the building: Work pe riod
: The period during whichcha nge is expected

Technical studies on indivi dual items re lating to on-site construction

Trial retrieval andinternal investigations

Manufacture and install
the retrieval device

On-s ite
ope ra tion

R&D

: To be conducted a s required

Technical issues in the Government-led R&D program on
Decommissioning and Contaminated Wa ter Management

Characterize the fueldebris
Improve environment
inside the building

Gradual expansion of fuel debris
retrieval

Fuel debris retrieval facility/safety system/fuel debris
temporary storage facility/maintenance facility

Design and production

Installation

Characterize the fuel debris

Review of the program on Decommissioning
廃炉・汚染水対策事業、東電自主事業
and Contaminated Water Management andにおける検討TEPCO's voluntary project--PCV内部詳細調査技術の開発
Development of technology for detailed
investigation inside PCV--燃料デブリの段階的に規模を拡大した
Development of technology for thegradualexpansion of fuel debris retrieval
取り出し技術の開発

Technical study by TEPCO- Technological issues for gradual expansion of fuel debris retrieval(1) Preparing retrieval system (improving durability of equipment, transfer test by transport truck)(2) Manufacturing equipment for gradual expansion of fuel debris retrieval(3) Providing safety system (Gaseous system, water treatment system, criticarity management system)(4) Providing temporary storage facility of fuel debris
Improvement of the environment inside and outside the Unit 1 building(Inside of the building: Dose reduction/removal of obstacles;
Outside of the building: Removal of exhaust towers of Units 1 and 2/removal of transformers, etc.)

Further expansionof fuel debris retrieval(Units 1/3)

Improvement of the environment inside and outside the Unit 3 building(Inside the building: PCV water lowering/dose reduction;
Outside the building: removal of the exhaust towers of Units 3/4, removal of transformers, etc.)

Fuel debris retrieval facility/safety system/fuel debris storage facility/maintenance facility/training facility and others
Review on the concept,

Actual site applicability verification
development (Remote installation,
dust diffusion control, etc.)

DesignProduction, installation, and retrieval

Review of the program on Decommissioning

廃炉・汚染水対策事業、東電自主事業
and Contaminated Water Management andにおける検討TEPCO's voluntary project- -原子炉建屋内・外の環境改善のためのDevelopment of technologies forenvironmental improvement of inside and
技術開発
outsidereactor building- -１号機PCV内部詳細調査技術の開発
Development of technology for detailed
investigation inside Unit 1 PCV- Development of investigation technology-原子炉圧力容器内部調査技術の開発
inside RPV- -燃料デブリ性状把握のための分析技術
Development of analytical and estimation
techniques for characterization of fuelの開発
debris- -燃料デブリ・炉内構造物の取り出し規模
Development of technologies for retrieving
fuel debris and internal structures towardの更なる拡大に向けた技術の開発
further expansion of fuel debris retrieval- -遠隔装置保守技術の開発
Development of technologies formaintaining remote equipment- Development of technologies for contai ning,-燃料デブリ収納・移送・保管技術の開発
transferring and storing fuel debris- Development of support technology for-１Ｆ廃止措置統合管理のための支援技
integrated management of 1F術の開発
decommissioning

Technical study by TEPCOTechnical issues toward further expansion of fuel debris retrieval(1) Development of fuel debris retrieval method (Side access/top access/of common)
Development of access facility installation, dismantlement technology, technology development for realization of large piece
removal cocept, crashing and containing technology(2) Development of safety system (Liquid system/ gas system and critiality control system)
Strengthening function to remove radioactive material, development of boric acid management technology such as boric acidrecovery, evaluation of dust behavior and recovery(3) Strengthening and sofistication of containment technology
Development of containment technology for radioactive material by passive control(4) Development of safety assessment technology
Data aquisition of debris dispersion rate and analytical method for evaluating radiation exposure* It is assumed that Unit 3 will be examined in advance and expanded to Unit 1.

Fig. 18 Technical issues and future plans on fuel debris retrieval (progress schedule)693.2 Waste management
3.2.1 Targets and progress(Targets)(1) The Solid Waste Management Plan (hereinafter referred to as the “Storage Management
Plan”) is appropriately developed, revised and implemented including waste prevention,
volume reduction and monitoring, with updating the estimated amount of solid waste to begenerated in the next ten years periodically.
(2) Countermeasures integrated from characterization to processing/disposal of solid waste arestudied from the expert point of view, and the prospects of processing/disposal method andtechnology related to its safety should be made clear by around FY 2021.

<Key points of "Basic Policies on Solid Waste”>(1) Thorough containment and isolation
Thoroughly containment and isolation of radioactive materials to prevent human access tothem, in order not to cause harmful radiation exposure.
(2) Reduction of solid waste volume
To reduce the amount of solid waste generated by decommissioning as much as possible.
(3) Promotion of characterization
Proper characterization addressing an increase in the number of analysis samples to proceedwith studies on processing/disposal method of solid waste.
(4) Thorough storage
Generated solid waste should be stored safely and reasonably according to its characteristics.
Storage capacity should be secured to ensure that the waste can be stored within the site of
the Fukushima Daiichi NPS.
(5) Establishment of selection system of preceding processing methods in consideration of
disposal
To establish selecting methods of processing for stabilization and immobilization (preceding
processing) and then select preceding processing methods before technical requirements ofdisposal are established.
(6) Promotion of effective R&D with an overview of overall solid waste management
To confirm required R&D tasks after cooperating with each R&D field in characterization andprocessing/disposal and overviewing the overall management of solid waste.
(7) Development of continuous operational framework
To establish the continuous operational framework including development of relevant facilities
and human resources in order to continue safe and steady solid waste management.
(8) Measures to reduce radiation exposure of workers
Thorough implementation of radiation exposure control, health and safety management basedon the relevant laws/regulations.

(Progress)
Waste management is a long-term effort that needs to attain the prospect of implementing final
disposal, while reducing risks in every stage from generation, storage, processing to disposal.
Since a large amount of solid waste with various characteristics is generated in association withdecommissioning of the Fukushima Daiichi NPS, the efforts are being made based on the Basic

70

Policies on Solid Waste summarized in the Mid-and-Long-term Roadmap. TEPCO is required toensure safe and reasonable storage of the solid waste generated. Led by NDF, the organizations
concerned are promoting efforts based on each role to advance technical examination of integrated
measures from characterization to processing/disposal of solid waste. The development forestablishing a flexible and reasonable waste stream (the flow of the integrated measures fromcharacterization to processing/disposal) is underway in addition to improvements incharacterization analysis abilities. The goal is to provide the Technical Prospects by around FY2021.
3.2.1.1 Current status of Storage in Fukushima Daiichi NPS
Table 2 shows the current state of solid waste storage. To store the solid waste properly, TEPCO
releases its Storage Management Plan, and estimates the volume of solid waste that will be
generated in the next ten years, and shows their policy such as on installing waste management
facilities to be required based on the volume.
According to this Plan, temporary outdoor storage of the solid waste will be eliminatedcompletely by FY 2028, except for secondary waste generated by water treatment, metal andconcrete rubble, and scrapped flanged tanks with extremely low surface radiation dose rates forreuse/recycling. Facilities needed to achieve this goal are under development (Attachment 10).
Among the targets of reuse/recycling, concrete rubble is crushed and recycled as roadbed
material after confirming that the surface dose rate is equivalent to the background radiation dose.
In addition, such as by melting is under consideration as a decontamination method for recycling
metal.
Secondary waste generated by water treatment is planned to be transferred to store in a building,and a large waste storage building is being constructed as a storage facility for sorption vessels.
Moreover, the slurry generated at ALPS (hereinafter referred to as “ALPS slurry”) generated by themulti-nuclide removal equipment, etc., and the waste sludge generated at the water purification
system (hereinafter referred to as “waste sludge”) sludge have high water content and mobility.
For safer storage, the slurry will undergo stabilization (dehydration) treatment (Scheduled start ofoperation in FY 2022), while the sludge will be transferred from the current storage area, anunderground storage tank in a building, to higher ground (to be completed in FY 2023).
Such solid waste will continue to be generated with some exceptions, and additional solid waste
will be generated from fuel debris retrieval.
Table 2 Status of solid waste storage(a) Storage of rubble, felled trees, used protective clothing, etc. (as of June 30, 2021)
Storage method
Outdoor storage (surface radiation dose rate ≤ 0.1
mSv/h)
Outdoor sheet covered storage (surface radiation dose
rate 0.1 - 1 mSv/h)Soil-covered temporary storage facilities, outdoor
container storage (surface radiation dose rate 1 - 30
mSv/h)

71

Stored volume (m3) / Storage
capacity (m3)(Percentage)
226,400 / 270,200（84%）
40,900 / 71,000（58%）
17,900 / 24,600（73%）

25,600/ 39,600（65%）

Containers* (in solid waste storage building)
Total
----

310,700 / 405,400(77%)

Felled trees
Classification

Storage method

Trunks, roots,
branches, leaves

Outdoor storage

Branches, leaves

Temporary storage pool

Total

----

Stored volume (m3) / Storage
capacity (m3)(Percentage)
99,500 / 134,000(74%)
37,300 / 41,600(90%)
136,800 / 175,600(78%)

Used protective clothing
Stored volume (m3) / Storage
Storage method
capacity (m3)(Percentage)
33,700 / 68,300
Container(49%)* Including secondary waste generated by water treatment (e.g. small filter)
Note that the storage volume is rounded to the nearest 100m3, so the total and the breakdown may not be
consistent.

(b) Management status of secondary waste generated by water treatment (As of July 1, 2021)
Sorption vessels, etc.

Outdoor temporary storage area ofused sorption vessels

Storage place

Storage volume

Cesium sorption apparatus

779

2nd Cesium sorption apparatus

244

3rd Cesium sorption apparatus

9
Existing

1,923

Expansion

1,888

High
performance83

Number ofvessels17

Number ofColumns

215

215

HICs from multi-nuclide
removal system
Used vessels from highperformance multi-nuclide
removal system

Number ofvessels and filters
Number ofvessels and filters
Number ofvessels and filters
Number ofcontainers
Number ofcontainers

Used columns from multiExisting
nuclide removal system
Used vessels and filters from mobile typestrontium system

Storedvolume/capacity(Percentage)

5,158 / 6,372(81%)

Waste sludge
Stored volume (m3) / storage
capacity (m3)(Percentage)454 / 700
(65%)

Storage place
Sludge storage facility (indoor)

Concentrated waste liquid
Storage method
Concentrated waste liquid storage tanks (outdoor)

72

Stored volume (m3) / storage
capacity (m3)(Percentage)
9,380 / 10,300(91%)

3.2.1.2 Prospects of processing/disposal method and technology related to its safety
The Mid-and-Long-term Roadmap states that Technical Prospects will be provided in theTechnical Strategic Plan by around FY 2021. Specifically, it will “present measures toward reducing
the volume of solid waste”, “develop analytical and evaluation methods for efficientcharacterization”, and “develop methods to reasonably select safe processing and disposal
methods at the time when the necessary information such as solid wastes’ properties are proven”.
On “present measures toward reducing the volume of solid waste”, the Basic Policy for Solid
Waste specifies that TEPCO should reduce the burden of overall waste management (for the entire
process, from generation, storage, and processing to disposal) to the extent possible. The
Technical Strategic Plan 2020 takes into consideration the possibility of further efforts based on
precedents in other countries.
“Develop analytical and evaluation methods for efficient characterization” and “develop methods
to reasonably select safe processing and disposal methods at the time when the necessary
information such as solid wastes’ properties are proven” mean to develop and establish themethods necessary for disposing of materials that become waste even after undergoing volume
reduction. NDF has organized and examined the following specific goals in the Technical Strategic
Plan 2018. Of the wide variety of types of solid waste, the secondary waste generated by water
treatment, which is highly mobile and for which there is no precedent for processing or disposingof it in Japan, has been selected as the main object for examination.
(1) Establish a safe and reasonable disposal concept based on characteristics and volume of thesolid waste generated in the Fukushima Daiichi NPS with its applicable processing technology,
and develop safety assessment methods that apply the features of the disposal concept, withconsidering examples of foreign countries.
(2) Clarify radiological analysis and evaluation methods for characterization.
(3) Clarify processing technology which practical application could be expected for stabilizationand immobilization, considering disposal for important waste streams such as secondary
waste generated by water treatment.
(4) Establish methods of reasonably selecting processing technology to stabilize and immobilize
waste based on the above methodology before the technical requirements for disposal are
determined (i.e., preceding processing).
(5) Have the prospect of setting processing/disposal methods for solid waste for which theprocessing technology considering disposal is not clarified, using a series of methods to be
developed by around FY 2021.
(6) Clarify issues and measures concerning storage of solid waste until it is conditioned
The following shows the Technical Prospect based on the results of these studies.
3.2.1.2.1 Approach for volume reduction
If a large volume of solid waste exists, not only does segregation and analysis take time, but the
number of storage containers and the scale of storage facilities increase, and so does the load of73waste management. Therefore, it is extremely important to reduce the physical volume of solid
waste as much as possible.
At the Fukushima Daiichi NPS, it is important to instill initiatives on volume reduction, aimed atdecreasing the burden of the overall solid waste management, into the overall decommissioning
activities by referring to the examples of overseas countries that have implemented the wastehierarchy concept (Attachment 11).
Specifically, the priorities for measures to be taken as waste management are (1) prevention ofwaste generation, (2) minimization of waste volume, (3) reuse, (4) recycling, and (5) disposal. In
waste management, it is important to prioritize (1) as much as possible, and consider (5) disposal
as the last option (Fig.19) for volume reduction of waste to be stored, processed, and disposed of.
In terms of reducing waste generation, it is important to consider in the design and construction
plan to reduce the volume of materials to be used. It is also important not to bring in substancesthat affect processing/disposal as much as possible. To minimize physical volume, considerationon

preventing

contamination

throughstrictsegregation,

maintenance/managementofmanufactures, extension of product life, and waste volume reduction is important. Reuse should
be promoted after contamination checks, decontamination, repair and parts replacement, and it isuseful to consider ease of reuse from the design stage. Considering alternative uses is also
beneficial. In recycling, it is important to consider the contamination condition of contaminatedvaluable sources, separate and process recyclable materials, and use them as new materials andproducts.
As shown in Fig.19, TEPCO has also been implementing initiatives corresponding to thisconcept. As new measures to be implemented as described in 3.2.1.1, among rubble accumulated
outdoors (surface radiation dose rate ≦ 0.1 mSv/h) (Table 2), reuse/recycling of metals, concrete,and scrapped flanged tanks with extremely low surface radiation dose rates are underconsideration. As part of this activity, decontamination methods for recycling metals are being
examined. In promoting safe and reasonable waste management, it is important to examine further
possibilities based on the characteristics of solid waste at the Fukushima Daiichi NPS in reference
to precedent cases overseas.

74

Summary of the Waste Hierarchy

Preferred
Approach

Examples of countermeasuresat the Fukushima Daiichi NPS

Waste Prevention

Installation of a vehicle service station,
prevention of carrying in packing material,and common use of construction equipment

Waste Minimisation

Classification, segregation, etc., of waste

Re-use of Materials

Examination on re-use of metals, concrete,and dismantled pieces of flange tanks with avery low surface dose

Recycling

Application of volume reduction equipment(an incineration system, a crusher, and a
metal cutting machine)

Disposal

Source: Strategy Effective from April 2011 (print friendly version), arranged by NDF

Fig.19 Summary of waste hierarchy at the NDA, UK, and countermeasures at the Fukushima
Daiichi NPS
3.2.1.2.2 Development of analytical/evaluation methods for implementing efficientcharacterization
Since the solid waste at the Fukushima Daiichi NPS is characterized by diverse nuclide
compositions and activity concentrations, and a large physical volume, efficient characterization isrequired. With the aim of developing the analytical/evaluation methods necessary for this purpose,
the Project of Decommissioning and Contaminated Water Management and other initiatives havedeveloped analytical methods for simplified and speed-up data acquisition. Furthermore, as amethod to perform characterization with small analysis data, the R&D program has established amethod for quantifying uncertainties in evaluation values (to identify variable distribution and thewidth) using statistical method with a method for efficiently identifying the inventory in combination
with analytical data and contamination mechanism (Attachment 11).
As an analytical method for simplified and speed-up data acquisition, automation of
pretreatment for analytical samples and a method using the triple quadrupole inductively coupledplasma mass spectrometry (ICP-MS/MS) (simplified method compared to the conventional
radioactivity measuring method) have been developed. These results will be incorporated into the
Building #1 of the Radioactive Materials Analysis and Research Facility under construction.
As characterization using statistical methods, in addition to using analytical data, statisticalmethods have been applied to the method for efficient characterization by combining with transition
models 41 , resulting in establishing the method to quantify uncertainties in evaluated values. A
method that combines the Data Quality Objectives (hereinafter referred to as “DQO”) process42
41 The process of contamination in reactor buildings and stagnant water by nuclides released from damaged fuelafter the accident is modeled to calculate the radioactivity of each nuclide in solid waste.

42 DQO process: A method developed by the US Environmental Protection Agency (US EPA) to plan sampling ofanalysis for decision-making. A sampling plan will be developed in seven steps. Planning and analysis arerepeated as necessary to achieve the objectives. This process can be applied to various problems that can besolved by analysis.

75

with statistical methods has also been examined/tested to develop a medium-to-long-term
analytical plan, and its effectiveness has been confirmed. In addition, a database called FRAnDLi(Fukushima Daiichi Radwaste Analytical Data Library) has been created, which contains
information related to analytical data (sample information (type, sampling location, date, etc.) andanalytical values of activity concentration, etc.), allowing data accumulation on a constant basis.
As described above, analytical/evaluation methods have been developed for efficientcharacterization and will be applied to characterization of solid waste.
3.2.1.2.3 Establishment of methods to reasonably select processing/disposal methods
In selecting processing/disposal methods in a reasonable manner, based on the waste
properties, an appropriate combination of processing (waste form) and disposal (disposal facility)
methods should be clarified so that the risk of buried solid waste to the public and environment canbe maintained sufficiently low in the future.
In the case of solid waste from a normal reactor, its properties can be estimated to some extent
by the previous findings (data) or analytical methods. Accordingly, the appropriate combination ofprocessing (waste form) and disposal (disposal facility) methods can sufficiently reduce the risk to
avoid a significant impact on the public and surrounding environment.
Even in the case of solid waste from the Fukushima Daiichi NPS, molten nuclear fuel is a major
source of contamination, and the radioactivity concentration does not exceed that of spent fuel.
Therefore, the risk can be sufficiently reduced by understanding the overall picture of the target
solid waste (properties such as nuclide composition, activity concentration by waste, waste volume),
and selecting a proper combination of processing (waste form) and disposal (disposal facility)
methods, while utilizing the experience and knowledge on radioactive waste processing/disposal
accumulated in domestic and overseas.
However, the overall picture of solid waste to be disposed of, including that which will be
generated, will gradually become clear as the progress and plans for fuel debris retrieval,contaminated water management, and other decommissioning work are clarified. Therefore, it isnecessary to repeatedly examine processing/disposal methods and safety assessments, startingfrom the waste for which properties have been clarified; to give consideration to making
processing/disposal methods more appropriate; and to accumulate knowledge to consider safe andreasonable processing/disposal methods for diverse solid waste collectively. Aiming for safer andnot extremely conservative storage of waste with high mobility such as slurry waste, processing(preceding processing) for stabilization/immobilization may be required before determining thedisposal method (disposal facility). Reprocessing would be necessary if the specifications of thewaste form even after preceding processing did not conform to those required by the disposal
method (disposal facility) to be determined. Therefore, in order to minimize such possibility, aselection method for the preceding processing method with disposal in mind is needed.

76

As mentioned below, study on an appropriate combination of processing and disposal methods,
or preceding processing methods, is considered for the waste for which properties have been
identified to some extent.


Establish several feasible disposal methods suitable for waste characteristics (without
specifying the feature of facilities such as their locations and sizes)



In parallel, establish several processing methods suitable for waste characteristics to beconsidered, and set the specifications of waste form after applying each processing method.



Evaluate the safety of several selected disposal methods based on the specifications ofwaste form after processing to verify whether risk to the public and environment can be
sufficiently low, and to consider more effective processing/disposal methods based on theevaluation results.

The above examination steps are repeated to narrow down disposal methods and specifications
for waste form after processing. Clarifying the overall characteristics of solid waste concurrently
with characterization helps identifying an appropriate combination of processing/disposal. When
preceding processing becomes necessary, candidate processing methods will be selected inconsideration of the status of examination and open issues at that point.
It is also important to consider the period during which pre-disposal management is to be
implemented, taking into account the risk reduction during that period, and examining necessary
and feasible technologies. Since storage is important to provide flexibility to respond to the progressof processing/disposal, and to reduce radiation exposure of workers due to the decay of theradioactive materials, it is important to consider storage strategies as part of this examination
process.
A series of these studies is represented as a flowchart shown in Fig.20. Technical knowledge
and evaluation methods necessary for these studies (establishment of processing technology andwaste form appropriate for waste, safe, reasonable and feasible disposal methods, disposal safetyassessment) have been developed through research/development (verification of applicability ofprocessing methods using engineering-scale test equipment conducted mainly on the secondary
waste generated by water treatment, establishment of reasonable and feasible disposal methods
based on the waste properties and applicable processing technologies, and development of safetyassessment methods) by the Project of Decommissioning and Contaminated Water Management,
and has been established as a series of methods to select processing/disposal methods in a safeand reasonable manner (Attachment 11).

77

Fig.20 Development of methods to reasonably select safe processing/disposal methods of solid
waste
3.2.2 Key issues and technical strategies to realize themAfter providing for future issues based on the prospects of a processing/disposal method andtechnology related to its safety, the technical strategies by category to realize them are show below.
3.2.2.1 Issues based on the Technical Prospects
As volume reduction is extremely important for the safe and reasonable management of solid
waste according to the progress of decommissioning work in the future, the measures in progressshould be continued steadily. Since solid waste continues to be generated, it is important tocontinuously examine further possibilities by referring to advanced cases of overseas’ for morevolume reduction. It is recommended that volume reduction is realized in consideration of the
expected outcome and feasibility.
For the development of analytical/evaluation methods for efficient characterization, it isnecessary to improve evaluation methods and continuously incorporate them into solid waste
management, including processing/disposal, while accumulating analytical data using efficient
analytical methods established through achievements in research/development. In this case,efforts should be made for low-activity waste such as rubble, as well as high-activity waste such as
secondary waste generated by water treatment and waste generated from fuel debris retrieval,according to the characteristics of each type of waste.
To establish methods to select safe processing/disposal methods in a reasonable manner, themethods shown in Fig.20 developed in 3.2.1.2.3 should be used to proceed with the examination
toward the determination of waste form specifications and manufacturing methods for Phase 3, as
specified in the Mid-and-Long-term Roadmap. Specifically, through these methods, the trial
78examples of optimization/rationalization of processing/disposal methods will be accumulated bywaste stream according to the progress of characterization and with the assumption of ensuringsafety to widely acquire findings on optimization by waste stream. Moreover, consideration will begiven to specify strategies for optimization/rationalization of the overall picture covering the entire
waste stream, allowing clarification of approaches toward such purposes. In doing so, it is important
to flexibly examine appropriate measures in consideration of actual use and economic feasibility
by reflecting the latest findings and applying the concept of Best Available Techniques43. As theexamination progresses, and the processing/disposal methods for the overall picture of waste are
finalized, it will be important to share the examination process for optimization, such as by sharing
the awareness of problems with local communities and society.
3.2.2.2 Technical strategy by sectors
3.2.2.2.1 Characterization
For low-activity waste such as rubble, it is not so challenging to perform the analysis work itself.
However, it takes an immense amount of time to measure entire quantity because of the enormousvolume of waste, and therefore, volume reduction and a corresponding efficient analysis strategy
are needed. For that purpose, it is important to take an approach that efficiently ensures therequired accuracy. In order to achieve this, efficient analyses should be promoted by making themsimplified/speed-up, and inventory evaluation methods that combine the DQO process withstatistical methods be established.
For high activity waste, sampling and analysis themselves are difficult, and the amount ofanalysis data to be obtained is limited. Thus, inventory assessment based on the transition model
becomes more important. It is necessary to obtain actual sample data such as by the ongoing
efforts for sampling from Cs sorption vessels and its analysis, which are currently in progress. The
application of inventory evaluation methods, which combine the DQO process with statisticalmethods, and the priority of data to be collected should also be considered to enhance the accuracyof the transition model.
Following the phase of analyzing samples that are easy to collect, characterization is now in thephase of collecting/analyzing samples that are important for waste management. Going forward, it
is important to develop a medium-to-long-term analysis strategy that defines the solid waste to beanalyzed, its priority, and quantitative targets for analysis, etc., and to proceed withanalysis/evaluation accordingly. It is useful to accumulate trial results and verify their validity inorder to establish a flow from the development of a medium-to-long-term analysis plan using
43 The term is defined as follows, as the concepts proposed in environmental protection and internal discussions.

・For the purpose of preventing environmental pollution, it should be defined as “a process, facility or method ofstate-of-the -art operation”, “with the principle of limiting emissions, discharge and disposal”, “taking into
account technical and economic feasibility”.
・It can change in accordance with technological, economic and social factors, as well as with changes inscientific knowledge and understandings.
(Daisuke Sugimoto, Hiroshi Hasegawa, “The concept of ‘Best Available Techniques (BAT)’ for disposing
radioactive waste – Review of overseas cases and suggestions for Japan-“, L06001 Report by Central
Research Institute of Electric Power Industry (November 2006))

79statistical methods; analysis and data acquisition; the incorporation of the acquired data into
examination of processing/disposal methods and evaluation of the outcome; to the development of
the next medium-to-long-term analysis plan based on the evaluation results.
As for facilities for analysis, in addition to the existing facilities in the JAEA’s Ibaraki area, etc., itis planned to establish the Radioactive Material Analysis and Research Facilities underconstruction, as well as facilities for analysis by TEPCO, allowing characterization of a variety ofsolid waste in parallel. Since the target nuclides, analysis items, accuracy, and the number ofsamples for analysis depend on the target solid waste, a structure should be established based on
the appropriate division of roles and according to the characteristics of facilities.
3.2.2.2.2 Storage
For storage of all waste, it is important to reconsider measurement items and timing, etc., interms of diverse information for characterization, while acquiring necessary information throughmonitoring and surveillance of the storage status commensurate with the risks involved.
With regard to high-activity waste, such as waste generated from fuel debris retrieval, the issues
and countermeasures assuming the further expansion of the fuel debris retrieval scale have beenclarified according to the results of research/development as of FY 2021. Going forward, reviews
should be performed along with the examination of the fuel debris retrieval methods. Measures
should be taken to ensure storage of solid waste that is expected to be generated during fuel debris
retrieval (trial retrieval, gradual expansion of the retrieval scale) before full-scale retrieval.
The site also has solid waste stored before the accident, and a large volume of dismantled waste
is expected to be generated after the completion of fuel debris retrieval. Only increasing storage
capacity for solid waste will eventually reach the limit, so efforts should be made to reduce thevolume of solid waste to be generated as much as possible.
In considering further possibilities of volume reduction, with an aim to reuse/recycle metals withextremely low surface radiation dose, chemical decontamination (decontamination by phosphoricacid), physical (mechanical) decontamination (steel blasting), and decontamination by melting(decontamination by melting slag), are under consideration as metal decontamination methods forrecycling.
As metal recycling with decontamination by melting slag has already been used in many
Western countries, it is considered a promising candidate technology. Thus, it is important to focus
on the areas where the conditions are different between Western countries and the Fukushima
Daiichi NPS (target nuclides, etc.), and to evaluate the applicability of the method.
3.2.2.2.3 Processing/disposal
The objective is to establish safe and reasonable processing/disposal methods for all solid
waste in which diverse waste streams exist, and to widely obtain knowledge for optimizing eachindividual stream. Therefore, it is necessary to continue development/research of processing anddisposal technologies required for the series of studies as shown in Fig.20.

80

Regarding the processing technology, outstanding issues in low and high-temperatureprocessing technology, for which research/development is promoted, should be addressed. Waste
streams, for which the application of low and high-temperature processing technologies has notbeen investigated, will be evaluated as necessary, and performance such as leachability forsolidified substances to be produced will be evaluated. As for low-temperature processing
technology, consideration is given to transformation of solidified substances as well as inspection
methods to verify the possibility of solidification. In the case of high-temperature processing
technology, the feasibility of the whole processing system, including supply and exhaust systems,
is an issue in addition to the solidification process, and therefore it is necessary to carry out
examination in a timely manner according to the start time of processing. Furthermore, in order toexpand technological options, it is important to examine the possibility of low-temperaturesolidification after interim treatment such as steam reforming.
Regarding disposal technologies, in order to establish reliable safety assessment techniques,
important issues specific to solid waste at the Fukushima Daiichi NPS will be explored and identified
based on the understanding of the sensitivity structure of parameters to radiation dose and thelong-term transition behavior of disposal facilities. Then, priorities will be examined andincorporated into research plans. Development and improvement of the proposed disposal options
consisting of proposed disposal concepts and waste to be disposed will also be promoted inreference to practices at home and abroad. Furthermore, while improving its reliability, after
expanding the target of waste streams, on which trial assessments will be performed by applying
safety assessment technology, group of disposal options will be examined with a bird-eye-view ofall solid waste at the Fukushima Daiichi NPS. Then, contributions will be made to considering
appropriate measures for overall waste management, in coordination with areas other thandisposal such as presenting targets for waste form performance and the accuracy required forcharacterization.
3.2.2.3 Summary of key technical issues
The main technical issues and plans in the future described in this section are summarized asshown in Fig. 21.
The Mid-and-Long-term Roadmap stated that the properties of solid waste will be analyzed andthe specifications of waste form and their production methods will be determined in the Phase 3,.
Therefore, in Phase 3-1, as a systematic approach toward this goal, studies will be conducted topresent appropriate measures for overall management of solid waste. Specifically, the first step isto create processing/disposal options of solid waste by examining unimplemented issues related
to processing technology, interim treatment, and disposal options. Then, the options will becompared and evaluated using the property data that are becoming clear, and examinations will be
conducted to identify waste stream that are suitable for characteristics of solid waste.

81

Fiscal year

2021

2022

2023

2026

2031 (Year)

On-site work
Technical studies for on-site construction

Related Milestones

Research and development

Processing/disposal methods
and technical prospects
regarding their safety

Outdoor storage will be eliminated completely

1. Waste characterization(1) Acquisition/managementof analysis data

Sampling
Development of sampling technology forhigh activity samples
Analyzing

Development/establishment of waste confirmation method
Analyses and measurements
for waste confirmation(2) Enhancement of efficiency
on waste characterization

Investigating the applicability of statistical inventory estimation methods and improving their accuracy
Study on analysis plan (Developing mid-to-long-term/yearly analysis plan, study on nuclides to be analyzed,
lower limit of quantification and number of samples to be collected, etc.)(3) Improvement of analysis
capacity

Development of Radioactive Material
Analysis and Research Facilities
Human resources development for analysis

Storage

2.Storage(1) Storage Management Plan

(2) Study and evaluation ofstorage methods(3) Reduction in volume

Development and review of the Storage Management Plan

Review according to fuel debris retrieval (Evaluation of type, volume, etc.)

Reduction of carry-in, re-use, recycling, volume reduction treatment, etc.

Development of technology for volume
reduction/recycling

3. Processing/disposal(1) Processing technology

Development of processing technology, Evaluation of long-term stability of solidification

Development of interim treatment technology(2) Disposal technology

Creating disposal concept suitable for solid waste

Improving reliability of safety assessment technology(3) Study on waste stream

Identifying waste stream
suitable for solid waste
Building a methodlogy for safetyassurance

Study on appropriate measures for entire waste management

Fig. 21 Main technical issues and future plans on waste management (progress schedule)823.3 Contaminated and treated water management
3.3.1 Targets and progress(Targets)(1) Under the three principles concerning the contaminated water issues (“Removing”
contamination sources, “Redirecting” fresh water from contamination sources, and “Retaining”
contaminated water from leakage), to reduce the stagnant water in the reactor buildings in FY
2022 to FY 2024 to about the half of the amount of the end of 2020 while continuing theoperation of the constructed water-level management system and controlling the generation
amount of the contaminated water to 100 m3/day or less in 2025. Moreover, to ensure stable
implementation of contaminated water management, measures against large-scale natural
disaster risks, such as tsunamis and storm rainfall, will be implemented in a planned manner.
(2) To arrange the relationship with a decommissioning process including full-scale fuel debris
retrieval beginning in the near future, and to promote examination of the measures of thecontaminated water management for medium-and-long term prospects.
(3) ALPS-treated water currently being stored in tanks will be handled in accordance with thegovernment’s basic policy announced in April 2021.
(Progress)
Fig. 22 shows the outline of the contaminated water management. Stagnant water in buildings,
that is, the contaminated water with a mixture of cooling water contacted with the fuel debris andgroundwater/rainwater flowed into the buildings is liquid containing a considerable amount of the
dissolved radioactive materials (inventory) from the perspective of measures to reduce the risk fromradioactive materials. Therefore, its hazard potential is high and so is the Safety Management level,
as the storage condition of such stagnant water deviates from what is originally intended (refer to
Section 2.2). Of this stagnant water in buildings, excluding the reactor buildings of Units 1 to 3,
where circulating water injection is ongoing, the process main building and high-temperatureincinerator building storing contaminated water temporarily for purification treatment, the treatmentof stagnant water in buildings was completed in 2020, and the inventory was significantly reduced.
However, the hazard potential is still high.
Currently, the following three measures are being implemented as contaminated water
management:83

(Source : TEPCO)
Fig. 22 Outline of contaminated water management44(1) Efforts to promote contaminated water management in accordance with the three principles(“Removing” contaminant sources, “Redirecting” fresh water from containment sources, and“Retaining” contaminated water from leakage”)
The groundwater level in the vicinity of the reactor buildings was stably controlled at low levels
through multilayered contaminated water management such as land-side impermeable walls andsub-drains. The increase in the amount of contaminated water generated during rainfall also tended
to be controlled by repair of damaged roofs and facings on site. As a result, the amount ofcontaminated water generated decreased from approx. 490 m3/day (FY 2015) before the measures
were taken to approx. 140 m3/day (2020). In order to reduce the amount of contaminated water to
100m3/day or less, roof repair and expansion of facing range are being addressed while adjusting
interference with other decommissioning work. 45 shows the results of the amount of contaminatedwater generated by factor, and the countermeasures to achieve the target by factor. Along with thecompletion of treating the remaining water, including ALPS-treated water, at the bottom of theflanged tanks in July 2020, measures and observation for risk reduction are being implemented
through monitoring of groundwater and the harbor.

44 Secretariat Meeting of the Team for Countermeasures for Decommissioning and Contaminated Water/Treated

Water Treatment (92nd), Material 2, “Outline of decommissioning, contaminated water and treated water”, July29, 2021
45Committee on Countermeasures for Contaminated Water Treatment (23rd meeting), Material 4, “Issues andcountermeasures concerning contaminated water management at the Fukushima Daiichi NPS”, June 25, 2021

84

Factors in the contaminated water generation(Category)2020 [FY2020*1]
Contaminated water
generation①

Inflow into the buildings(Rainwater/groundwater inflow)

Approx. 100m /day[about 90m3/day]- Repairing wrecked parts of the rooftop- Facing around the buildings- Lowering water level in subdrain

Approx. 50m /day②

Amount trandsferred from the El.
2.5m ground to the buildings

Approx. 10m3/day
3[about 10m /day]- Decrease in operational water level
Reducing the amount of water transferred to theturbine buildings by securing a margin for waterlevel rise during heavy rain

Approx. 10m3/day

Chemical injection volume during③

2

purification with ALPS *3

3

Approx. 10m3/day- Reliable maintenance[less than about 10m3/day ]3

Approx. 10m /day④-1

Amount of transfer generated byto
decommissioning work

Approx. 10m3/day[about 10m3/day]- Scheduled treatment of stagnant water in trenches
and so forth④-2

Amount of generation transferred(Drainage at the shallow draft wharf)

Approx. 10m3/day
7[about 30m /day]- Reliable operation and management of facilities- Scheduled treatment of stagnant water

Amount of contaminated water generated

Reference

Rainfall(mm)Targets for FY2025
3(m /day)3

Measures to achieve the amount of 100m /dayApprox. 140m3/day
3[about 140m /day]3

Approx. 10m /dayLess than 100m3/day

1,339m (3.7mm/day)[1,329mm(3.7mm/day)]Average rainfall (3.9mm/day)*1 Data obtained through March 31, 2021
*2 Chemicals being injected into preprocessing facility for multi-nuclide removal system(TEPCO material editted by NDF)

*3 Invludes sprinkling of water on the operating floor, inflow into the buiding outer the frozen soil and transfer of stagnant water in trench

Fig. 23 Results of contaminated water generation by factor and main measures to reduce theamount45(2) Efforts to complete stagnant water treatment
In 2020, the treatment of stagnant water in buildings, excluding the reactor buildings of Units 1
to 3, the process building and high-temperature incinerator building, was completed.
It is planned to lower the water level in the reactor building while continuously lowering the subdrain water level in order to reduce the amount of stagnant water in the reactor building by half. In
association with this, the importance of issues in handling sludge containing α-nuclides (α-sludge)
is increasing. As the particle size distribution and chemical composition of the α sludge at thebottom of the reactor building have been analyzed, it is expected that most of the sludge can beremoved by a filter with an appropriate pore diameter. In order to complete the treatment of stagnant
water in the process building and the high-temperature incinerator building, moreover, methods forradiation dose rate surveillance or recovery are under consideration for high-dose zeolite sandbags
located on the lowest floor. In the case of a building where the treatment of stagnant water has
been completed and the floor surface has been exposed, a method for recovering sludge locatedon the floor is being studied.
(3) Efforts for stable operation of contaminated water management
The measures against tsunami have been implemented, including installation of tsunami tidewalls in the Japan Trench; closure of building openings; reinforcement of land-side impermeable
walls (suppression of brine46 leakage by electrifying closing valves of the brine supply piping); andrelocation of water collection systems such as sub-drains from the revetment side to higher ground.
As a countermeasure for heavy rain, reinforcement of discharge functions of the existing drainage,etc. is underway.

46 Cooling liquid that circulates through underground frozen ducts for ground freezing85

(4) Treated water management
For handling water treated with the multi-nuclide removal equipment, so-called ALPS-treatedwater 47 , discussions by experts, including technical aspects and the social impact, such as
reputational damage, were held at a national subcommittee, and a report was published in February
2020. In April 2021, based on opinions from a wide range of people through subsequent opinion
exchanges with local governments and agriculture, forestry, and fisheries industries, thegovernment announced the basic policy of discharging the ALPS-treated water into the ocean48 on
the premise that safety will be ensured and that measures against reputational damage will bethoroughly taken, while TEPCO published “Actions by TEPCO in response to the Government’s
basic policy on disposal of the treated water by multi-nuclide removal equipment”49. Currently, theequipment needed and actions against reputation are under consideration, TEPCO reported thestatus of these examinations and efforts in August 202150, 51.
3.3.2 Key issues and technical strategies to realize them
3.3.2.1 Issues in the future treatment of stagnant water in buildings
The following three points are the key issues for the future treatment of stagnant water inbuildings.
(1) Prevention of spreading α-nuclides
At the bottom of the torus room of the reactor building, stagnant water containing α sludge andionized α-nuclides exists, and relatively high concentration of α-nuclides has been detected (Fig.
24). These α-nuclides has significantly high effective dose factors when ingested by inhalation, and
it is necessary to suppress this spread to a limited extent. At present, the properties of α-sludge
and ionized α-nuclides are being analyzed, and removal methods for preventing the spread of αnuclides are being studied. However, the chemical form of α-nuclides may change depending onwater quality and coexisting substances. In order to ensure removal of α-nuclides, it is necessary
to collect samples from as many places as possible and to understand the variation in theirproperties. Since α-sludge contains a high concentration of Cs -137, in addition, it is important toconsider measures to reduce radiation exposure dose of workers, maintainability, and secondary
waste.

47Water that is treated with multi-nuclide removal equipment until the level of radioactive materials other than tritium

fall below the regulatory standard on safety, without fail.

48 Inter-Ministerial Council for Contaminated Water, Treated Water and Decommissioning Issues (5th meeting),

Material 1, “Basic policy for disposing of treated water by multi-nuclide removal equipment at the TEPCO
Fukushima Daiichi Nuclear Power Station (draft)”, April 13, 2021
49TEPCO, Attachment 1, “Actions by TEPCO in response to the Government’s basic policy on disposal of the treatedwater by multi-nuclide removal equipment”, press release, April 16, 2021
50 TEPCO, Status of examining dealing with water treated with multi-nuclide removal equipment at the Fukushima
Daiichi Nuclear Power Station, press release, Report by the subcommittee dealing with water treated with multinuclide removal equipment, press release, August 25, 2021
51 TEPCO, Attachment 3, Handling of compensation in the event reputational damage caused by discharging
treated water by multi-nuclide removal equipment, press release, August 25, 2021

86

Fig. 24 Water treatment systems for stagnant water in buildings and measurement results of total
α-nuclides52(2) Treatment of high-concentration stagnant water
In treating stagnant water in buildings, water at the bottom of the buildings has not been pumped.
Therefore, not only α-nuclides but also highly concentrated stagnant water containing radioactive
materials and salt, which is close to the state immediately after the accident, exists at the bottomof the torus rooms of the reactor buildings. The S/C also contains contaminated water with higher
concentration of radioactive materials and salt than usual stagnant water.
It is necessary to dispose of contaminated water containing such high concentration ofradioactive materials and salt as soon as possible in terms of risk reduction. In order to suppress
the fluctuation range of the contaminated water concentration for the stable operation of the water
treatment system, however, contaminated water of high concentration and that of low concentration
are mixed for treatment. For this reason, in order to accelerate the treatment of highly contaminatedwater, it is necessary to proceed steadily with careful planning that takes into account the waterbalance with low-concentration contaminated water.
(3) Recovery of zeolite sandbags
As for the zeolite sandbags in a high radiation dose state53 placed on the lowest floor of theprocess building and the high-temperature incinerator building, we are considering a method of
underwater recovery that directly transfers them to the first floor by remote heavy machine or ROV,
dehydrates them and places them in containers, prior to stagnant water treatment (Fig. 15). In thismethod, securing of visibility in turbid water and improvement of recovery rate of zeolite are issues,and prior examination using mockups is important. An impact assessment must also be performed
for the case where part of the zeolite could not be recovered and to consider alternative measures
for radiation dose reduction. The basement floors of these buildings have functions for equalizing
the concentration of stagnant water, settle fine particles and temporarily store the water during
52 The 89th meeting of the study group on monitoring and assessment of specified nuclear facilities, Material 4-3,

Progress of treatment of stagnant water in the buildings, March 22, 2021

53 Total weight of approx. 20 t. The maximum radiation dose rate on the sandbag surface is approx. 3,000 mSv/hin the process building and approx. 4,400 mSv/h in the high-temperature incinerator building.

87

heavy rain. Stagnant water will be transferred to the temporary storage tanks to proceed withstagnant water treatment. It is also necessary to consider the functions and operation method of
temporary stagnant water storage tanks upon estimation of changes in water quality and volume.

Fig. 25 Outline of the underwater recovery method for zeolite sandbags54
3.3.2.2 Issues of contaminated water management considering the decommissioning
process such as fuel debris retrieval
The following three points are the key issues of contaminated water management considering
the decommissioning process such as fuel debris retrieval.
(1) Handling of high-concentration α-nuclides
The possibility that fine particles containing high-concentration α-nuclides are produced andmixed into the water treatment system in association with cutting, fabrication, etc., of fuel debris
cannot be denied. Therefore, it is necessary to take measures, such as strengthening themonitoring of water treatment systems, installing collection systems for fine particles containing αnuclides, and monitoring for criticality. Further, in handling such α-nuclides, the chemical
characteristics of actinide elements constituting the α-nuclides needs to be sufficiently understood.
Actinide elements change into various chemical forms depending on the water quality (pH,dissolved oxygen concentration, coexisting materials, etc.), and the dissolution rate tends toincrease significantly especially in the oxidizing environment. Therefore, it is important tounderstand and control the water quality. When examining the method of removing α-nuclides,
moreover, more effective designing becomes possible by sampling even a small amount of actual
liquid at an early stage and conducting performance verification using samples. It is necessary toevaluate the safety of recovered high-concentration α-nuclides from the perspective of criticality,
heat generation, radiation dose, hydrogen generation, and waste management.
(2) System configuration for sustainable stable operation
The water treatment system at the time of fuel debris retrieval needs to secure reactor coolingwater and purify a large amount of radioactive materials containing α-nuclides. Therefore, as a
54 The 87th meeting of the study group on monitoring and assessment of specified nuclear facilities, “Material 3- 3,

Status of study on progress for zeolite sandbags, January 25, 2021”88whole system, it is recommended that the cooling water circulation system has a simple system
configuration that focuses on reliability. For the purification system which has a complex
configuration, it is recommended to have a system configuration that takes and purifies a portionof contaminated cooling water, enabling both purification and supply of reactor cooling water. As
for the purification system equipment, it is necessary to consider the evaluation of the volume ofsecondary waste generation and maintainability under a high radiation dose, as well as theevaluation of purification performance, and to examine purification methods in the medium-andlong term. In particular, system planning for the entire purification system equipment should bepromoted with the view to improving and updating existing systems, based on the operation
experience of existing systems, changes in the water quality environment up to the present, andpredictions of changes.
(3) Medium-and-long term measures for contaminated water management systems
It is necessary to ensure that periodical inspection and updating of equipment, including landside impermeable walls and sub-drain systems, is implemented in order to maintain theeffectiveness of contaminated water management over the medium-to-long term. For this purpose,
it is important to anticipate various risks, such as deterioration of system functions caused by aging,metallic fatigue due to traffic loads, damage of piping caused by natural disasters; toprocure/arrange backup and alternative items for the enhanced structure for monitoring and earlyrecovery, and for stable operation; and to promote maintenance/management and system updates
in a planned manner.
While the current contaminated water management is shifting to a certain stable state, in
addition, it takes a long time to complete fuel debris retrieval. Along with the selection of methods
for further expanding the scale of debris retrieval currently in progress, it is important to see a
medium-and-long term, overlook the current contaminated water management anew, and examine
the principles of more stable contaminated water management and more appropriate
maintenance/management.
3.3.2.3 Issues for discharging ALPS-treated water into the ocean
As described in 3.3.1 (4), Treated Water Management, in April 2021 the government announced
to discharge ALPS-treated water into the ocean48 on the premise that safety will be ensured andthat measures against reputational damage will be thoroughly taken, and TEPCO is making
preparations for such implementation49. For the peaceful uses of nuclear energy, the discharge of
waste liquid into the ocean with a sufficiently small radiological impact on the human populationand the natural environment is in accordance with the internationally recognized concept of“Dilution and Dispersion”55, which is a method widely adopted in Japan and abroad56.

55 ICRP, 1998. Radiation protection recommendations as applied to the disposal of long-lived solid radioactive

waste. ICRP Publication 81. Ann. ICRP 28 (4).

56 Ministerial Ordinance for Commercial Nuclear Power Reactors concerning the Installation, Operation, etc.,

Article 90, Ordinance of the Ministry of International Trade and Industry 77 of 197889

At the Fukushima Daiichi NPS, efforts are being made to reduce risks for fuel debris retrieval inenvironments where the radiation level is still high. Under these circumstances, it is extremely
difficult to reduce the amount of contaminated water generation to zero. The contaminated water
generated is purified by multi-nuclide removal equipment, etc., and stored in tanks located onhigher ground as ALPS-treated water containing large-volume tritium (ALPS-treated water/water
under treatment). As most of the contaminated water is stored in welded tanks, the possibility ofleakage is very low, but the risk is not zero. There is also concern that the tanks occupy a large
portion of the site, making it impossible to secure enough space for stable, temporary storage ofspent fuel as well as safe fuel debris retrieval.
On the premise that safety will be ensured and that measures against reputational damage willbe thoroughly taken, the announced government policy on the discharge of ALPS-treated water
into the ocean aligns with the international concept mentioned above, and is an important decisionin terms of ensuring the sustainability of decommissioning work. In particular, discharging ALPStreated water into the ocean enables reduction of risks associated with tank storage and allocationof limited resources to other high-risk operations, contributing to the steady progress of
decommissioning work.
On the other hand, it is also a fact that there have been concerns about reputational damage dueto the discharge of ALPS-treated water into the ocean. Therefore, efforts should be continued todeepen understanding to eliminate such concerns. The reliability of TEPCO has declined due toinappropriate incidents in terms of physical protection of nuclear materials at the KashiwazakiKariwa Nuclear Power Station and insufficient provision of information during earthquakes at the
Fukushima Daiichi Nuclear Power Station. TEPCO needs to take this reality seriously and respond
more carefully than before. One of the reasons for this incident might be insufficient communication
with local communities and commercial distributors. Therefore, greater transparency is required,for example, by repeatedly providing explanations in an easy-to-understand and careful manner,
mainly by TEPCO, in order to increase understanding of (1) an operation plan for offshore
discharge; (2) the effects of tritium contained in the water to be discharged to the ocean on thehuman body; and (3) the method for verifying the operation status as the basics for implementing
safe discharge, and by verifying these through reliable third parties such as IAEA in cooperation
with organizations concerned, and by delivering accurate information.
(1) Operation plans for offshore discharge (for details, see the Attachment 12)
Fig. 26 shows the annual discharge of tritium in nuclear facilities in Japan and abroad57. These
nuclear facilities discharge tritium by means that can minimize the radiological impact on thesurrounding environment as much as possible. Fig. 27 shows the treatment flow up to the dischargeof waste liquid containing tritium in nuclear power plants in Japan. Waste liquid generated in a
power plant is [purified] to reduce the concentration of radioactive materials in the waste water as
much as possible by filtration, evaporation, adsorption by the ion-exchange resin method, etc., and

57 Report by the subcommittee on handling of ALPS-treated water, February 10, 202090

decay of radioactivity with time. The purified liquid is collected in a sample tank where theconcentration of the radioactive material is [analyzed and verified] by monitoring and sampling.
Subsequently, waste liquid containing tritium is mixed with seawater used for cooling and thendischarged to the environment. As such, [purification] → [analysis/verification] → [discharge] is thebasic process, and the discharge is in compliance with the regulatory standards on theconcentration of radioactive materials. In nuclear power plants in Japan, radiation monitors are
installed to measure γ-rays before radioactive materials are discharged into the environment tomonitor concentration.

Fig. 26 Annual discharge of tritium in nuclear facilities in Japan and abroad58［ Purification ］［Analysis
Verification］［Release］

Environmental
ReleaseWaste liquidstorage tanks

Purification
devices

Sample tanks

Seawaterfor cooling

Source : Prepared by NDF with reference to materials on domestic nuclear power plants

Fig. 27 Treatment flow up to discharging waste liquid containing tritium in nuclear facilities
58The 1st Expert Meeting for marine monitoring on ALPS-treated water, Reference 2, June 18, 2021

91

The conceptual diagram of the discharge system planned by TEPCO is shown in Fig.28. In thisdischarge facility, the concentration of radioactive materials other than tritium in the ALPS-treatedwater is confirmed to be below the regulatory standard value for the discharge to the environment
in the measurement and verification facilities. Then, the discharged water is mixed and diluted withseawater so that the tritium concentration is less than 1,500 Bq/L, which is 1/40 of the regulatory
standard value (60,000 Bq/L).
TEPCO will evaluate the tritium concentration after the dilution from the tritium concentration
obtained by the analysis at the measurement and verification facilities and the flow rate ratio of the
ALPS-treated water and seawater. In addition, in order to confirm that the tritium concentration in
the ALPS-treated water after seawater dilution is below 1,500 Bq/L, the tritium concentration will
be confirmed by sampling every day during the discharge and will be announced immediately. (Fordetails, refer to Section 3.4(2) c Dilution and Discharge in Attachment 12). In addition, monitoringin other areas of the ocean will be enhanced by adding measurement points and increasing thefrequency of measurement50.

(TEPCO material edited by NDF)
Fig. 23 Conceptual diagram of the discharge system planned by TEPCO59
A set of discharge systems covers from [purification], [analysis/verification], [dilution], [discharge]
to [continuous monitoring in the sea area]. In addition, an emergency shut-off valve is planned to
be installed in the process between [analysis/verification] and [dilution] so that the discharge canbe suspended in the event of system failure or detection of any abnormal values during marine

59TEPCO.: Attachment 2 [Summary version] Actions by TEPCO in response to the Government’s basic policy fordisposing of treated water by multi-nuclide removal equipment”, April 16, 2021

92

monitoring. The annual discharge of tritium is planned to be within 22 trillion Bq, a target value ofcontrolling discharge set by the Fukushima Daiichi NPS before the accident.
This discharge system planned by TEPCO is based on the past results in Japan and overseas,and safe offshore discharge will be possible by ensuring that these results are reflected in thesystem, developing operational procedures and manuals for equipment operation and analysis,thoroughly educating, training, and observing operators, and strictly adhering to the implementation
plan. By strictly following the implementation plan, safe offshore discharge will be possible.
(2) The effects of tritium contained in the water to be discharged to the ocean on the human body
Tritium is a radioisotope of hydrogen, and is contained in any water, including water vapor in theatmosphere, rainwater, seawater, tap water, and even in the human body. In addition, its properties
are similar to those of ordinary water molecules. As its characteristic, therefore, it is difficult toseparate tritium from water. The impact of tritium on the human body due to the disposal of ALPStreated water has also been evaluated using the UNSCEAR60 method, and it has been confirmedthat the impact is extremely small compared with that of natural radiation (2.1 mSv/yr)57. It is alsoplanned to conduct breeding tests of marine organisms in the ALPS-treated water diluted withseawater, with the aim of fostering understanding of the discharge of ALPS-treated water into theocean and reducing the impact of rumors. It is also planned to conduct webcasting of the breeding
situation50. TEPCO will review the annual discharge of tritium according to the progress of
decommissioning, etc.
(3) The method for verifying the operation status
In the operation phase, it is necessary to confirm that the planned facilities are installed andoperated reliably, that the analysis is conducted reliably in accordance with the establishedprocedures, etc., and to publish the results. Moreover, it is necessary to increase the transparencyof the implementation status of the plan by involving third parties other than TEPCO in
analysis/verification before the discharge, and marine monitoring after the discharge. For thirdparty involvement, it is important to develop a quality assurance system because the results mayvary depending on the institution and facility for analysis due to the fluctuation of the lower detection
limit caused by the detection accuracy of measuring instruments or analysis techniques. For
example, it is effective to analyze common samples in an integrated way before the discharge to
see the fluctuation range of the result, and verify and disclose the quality of each analysis institution
beforehand. It is also necessary to consider a structure for verifying the operation status of thedeveloped system.
(4) Further initiatives
TEPCO's planned discharging system, if operated reliably in accordance with the implementation
plan approved in the review by the NRA, will have no adverse effects on humans and the
60UNSCEAR (United Nations Scientific Committee on the Effects of Atomic Radiation): The Committee reports to

the U.N. General Assembly on the effects of radiation among the human population, and on the naturalenvironment from a scientific and neutral standpoint. The 27 member countries include Japan, the UK, the US,
France, Russia, Germany, China, and South Korea.

93

environment, including other radionuclides, and therefore it is an important issue to operate thesystem "reliably" “as planned”. Going forward, TEPCO will need to proceed with the following
preparations. During actual operation, it is necessary to ensure the implementation of the plan(system, operation, information distribution, etc.) established in the preparation stage, to perform
check and review, and to review and expand, the plan as needed, as well as to ensure its
transparency.
 In the operational phase, develop a series of operation plans including system operation,
analysis of ALPS-treated water, flow control of the treated/diluted water, marine monitoring,
maintenance, and troubleshooting, and then develop a system plan that minimizes risks
and eliminates social concerns
 Perform a radiological impact assessment on the human population and the naturalenvironment, and disclose evaluation results based on the specific discharge plan Verify safety by experts from the International Atomic Energy Agency (IAEA) and otheragencies Develop a plan to strengthen marine monitoring, and perform marine monitoring before
the discharge Education and training on system operation and analysis, etc., for parties concerned
including contractors (TEPCO)
 Development of strategies to provide accurate and understandable information
domestically and internationally without causing anxiety from a social perspective, andtimely dissemination of the status of preparations
 Ensuring implementation of measures against reputational damage as set forth in the
Government’s basic policy announced in April 2021
NDF will provide technical and professional support for TEPCO in designing the discharge
system and considering discharging methods, while promoting distribution of accurate informationand increasing understanding through various opportunities in Japan and abroad in line withchanges in the interests of those who will receive the information. We will also make sure thatTEPCO implements measures to minimize reputational damage, and that TEPCO takes
appropriate and sufficient actions with compensation in the event of reputational damage.
3.3.2.4 Summary of key technical issues
The main technical issues and plans described in this section are summarized in Fig. , and Fig.
18 shows the future plans in regard to water treatment system for fuel debris retrieval.

94

Fiscal year
Relatedmilestones
Contaminatedwater generation
amount

2021

2022

2023

2026

To control the contaminated water generation amount to approximately 150 m3/day (in 2020)
To complete the processing of stagnant water in the building (in 2020)

2032(year)

To control the contaminated water generation amount to 100
m 3/day or less(in 2025)

To reduce stagnant water in the reactor building to approximately half that of the end of 2020 (FY 2022 to FY 2024)

Maintenance management operation of a groundwater bypass, a sub-drain, and a land-side impermeable wall
Site pavement inside the land-side impermeable wall (mountain-side)

Site pavement inside the land-side impermeable wall (sea-side)

Repair broken parts of roofs (incl. the installation of a large cover on Unit 1 R/B)

Stagnant water in buildings

Confirming stagnant
water characteristics

Units 1 to 3
R/B

Water level lowering for a 50% reduction

Implementing required measures in accordance with the stage of fuel debris retrieval

Simple measuresagainst α-nuclides

Designα-nuclides removal equipment
Manufacture and installation

Operation▽ Accomplished in December, 2020

Units 1 to 4
T/B, etc.
Processmainbuilding
Hightemperatureincinerator
building

Measuresagainst stagnant
water

Water level lowering for
exposing the floor surface

Examination of methods for collecting sludge on the floor

DesignConceptual
designManufacturing and installatin of equipment for collecting sludge on the floor

Alternative tank
Manufacture and installation
Measures for zeolite sandbags
Design＜Legend＞:The period of on-site work

Manufacture
and installation:Technical study on on-site work

Collection
Water level lowering for
exposing the floor surface:The period in which change is
expected

Removing stagnant water on premises
Dismantling underground water storage tankConceptual study

Design/removal

Fig. 29 Key technical issues and future plans on contaminated/treated water management(progress schedule)

953.4 Fuel removal from spent fuel pools
3.4.1 Targets and progress(Targets)(1) While the return of residents and reconstruction in the surrounding area is gradually advanced,
to carry out a risk assessment and ensure safety including preventing dispersion of radioactive
materials and to start removal of fuel in spent fuel pools (SFPs) in FY 2027 to FY 2028 for Unit
1 and FY 2024 to FY 2026 for Unit 2. To complete fuel removal from SFP in FY 2020 for Unit 3.
(Completed in February 2021)(2) The fuel in Units 1 to 4 that were affected by the accident are removed from the SFPs andtransferred to the Common Spent Fuel Storage Pool, etc., where they are appropriately stored
so that they are in a stable management state. In order to secure the Common Spent Fuel
Storage Pool capacity, the fuel stored there is transferred to and stored in Dry Cask Temporary
Custody Facility.
(3) To perform the evaluation of long-term integrity and the examination for processing for theremoved fuel and to decide the future processing and storage method.
(Progress)
TEPCO is working on the work plan indicated in the Mid-and-Long-term Roadmap and the Midand-Long-term Decommissioning Action Plan.
In Unit 1, due to the hydrogen explosion, roof stabs, building materials, such as steel frames,
which constituted the upper part of the building, an overhead crane, etc., have collapsed as rubble
on the operating floor as shown in Fig. . While the residents were returning, from the perspectiveof further reduction of radioactive dust dispersion risk, the whole operating floor was covered witha large cover for fuel removal in Unit 1 SFP. The removal method was changed to one in which
rubble removal and fuel removal from SFP are carried out inside the cover. Fig. 24 shows aconceptual drawing of this method. Currently, removal of the building cover (remaining parts) thatinterferes with installation of the large covers has been completed, and installation of the large
covers is in progress within the area where the cleanup around the reactor building has been
completed. In addition, in order not to affect fuel in SFP, measures to prevent and mitigate thedropping of rubble, such as installation of supports for overhead cranes and fuel handling machines,
and curing of SFP, were completed in November 2020.

96

(TEPCO material edited by NDF)

Fig. 30 State of the collapsed rubble on the Unit 1 operating floor(TEPCO material edited by NDF)

Fig. 24 Fuel removal method from Unit 1 SFPFor Unit 2, a new method in which the upper part of the operating floor will not be dismantled
and with access from the south side of the reactor building was adopted from the perspective offurther reduction of radioactive dust dispersion risk similar to Unit 1. Fig. shows a conceptual
drawing of this method. At present, yard cleanup and soil improvement work are underway as
preparatory works for installing the fuel removal platform. In addition, based on the measurement
results of the air dose rate and surface contamination of the operating floor conducted in Marchand April 2021, decontamination of the operating floor is ongoing in order to further reduce radiation
dose.

97

(TEPCO material edited by NDF)

Fig. 32 Fuel removal method from Unit 2 SFP
Removal of all fuel from Unit 3 was completed in February 2021. During that time, various
problems occurred in the fuel handling equipment and operation, such as a problem in the
procurement stage including crane voltage setting error; failure of mast rotation during fuel removal;
and irregular winding of mast wire ropes. It is important to roll out the knowledge and lessons
learned from these experiences concerning procurement and remote control to planning andimplementation of future decommissioning work, including fuel removal from Units 1 and 2 and fueldebris retrieval from Units 1 to 3. Therefore, in addition to keeping records of all operations, it isimportant to organize the results, including efforts for troubleshooting and recurrence prevention.
In particular, measures for strengthening quality control, improving device development, plans andresults of preliminary tests, and consideration of contingency plans were compiled as a document
for a reference. Furthermore, as part of its approaches to strengthen quality control, TEPCO hasrevised “The Basic Design Management Manual”, which is common to all Fukushima Daiichi NPS
to improve its design and procurement process.
For Units 5 and 6, fuel will be appropriately stored in the SFPs of the units for the time being.
Then, they will be removed in a range so as not to affect the work in Units 1 and 2.
Securing the available capacity of the Common Spent Fuel Storage Pool and transfer of some
fuel in the Common Spent Fuel Storage Pool to Dry Cask Temporary Custody Facility are required
to remove all the fuel in SFPs, including Units 5 and 6, and store them in the Common Spent Fuel
Storage Pool. For this purpose, TEPCO is working on expanding storage capacity of Dry Cask
Temporary Custody Facility and off-site transportation of new fuel. The available storage capacity
in the Common Spent Fuel Storage Pool and Dry Cask Temporary Custody Facility is shown in
Fig. .
Such efforts will be made to complete fuel removal in all Units in 2031.
As described above, fuel removal from Unit 3 was completed but other high-radiation dose
equipment such as control rods, channel boxes and filters are also stored in SFP. Although coolingis not necessary for them, shielding is required, and there is a risk that the source in the pool will
be exposed if the pool water leaks. Therefore, in terms of risk reduction, removal of such high98

radiation dose equipment is needed following the fuel in SFP. In this case, it is efficient to utilize thedevice used for fuel removal and rubble removal. Therefore, as soon as preparations for removal
are completed, including securing storage facility, removal work should be implemented
immediately. Thereafter, pool water can be excluded from management by draining the pool. Prior
to draining the pool, however, the radiation dose and dust dispersion from the pool after drainage
should be evaluated to confirm safety. This leads to smooth fuel debris retrieval in the later stages
because of the increased flexibility of use of the operating floor, etc.
As with Unit 3, for Units 1, 2 and 4, high radiation dose equipment is stored in each SFP. Giving
priority to fuel removal from SFP (removal is completed in Unit 4), fuel removal and drainage of thepool should be performed in a planned manner. In designing fuel handling facility, it is necessary toconsider not only the fuel in SFP but also the plan and process for removing the high-radiation dose
equipment.

Number of spent fuel, etc.

10000

8000

Number of available capacities as of
now in entire common spent fuelstorage pool and Dry Cask Temporary
Custody Facility ：1995

For fuel removal from Units 1, 2, 5 and 6(4433), efforts are underway to expand
storage capacity for Dry Cask Temporary
Custody Facility and to export new fueloff-premises (360 planned).

Spare: 63

6000
In all Units 1, 2, 5 and 6: 4433

4000

Stored6671
Unit 6 :1884

Spare
1932

2000
Unit 2 : 615

0Stored2033

Unit 5 :542

Unit 1 : 392

Units 1/2 SFP

Units 5/6 SFP

Common Spent Fuel Storage
Pool

Dry Cask Temporary
Custody Facility

Fig. 33 Storage status of spent fuel (As of March 2021)
3.4.2 Key themes and technical strategies to realize them
3.4.2.1 Fuel removal from SFPs
For Units 1 and 2, it is necessary to advance the work steadily to realize the determined new
removal method.
In promoting the project, it is important to make assessment of safety in association with workand confirming that necessary and sufficient safety is ensured. Moreover, it is essential tocomprehensively consider technical reliability, rationality, promptness in the work schedule, actualsite applicability and project risk, etc.
(1) Unit 1
For Unit 1, the installation of the large cover, and the removal of leftover objects such as rubble
on the operating floor will be promoted. Although overhead crane support is installed on the upper99

part of the operating floor for fall prevention, it is still in an unstable state. Therefore, removing theoverhead crane in a safe and reliable way is one of the main issues to prevent it from collapsing
onto the fuel handling machine and falling into SFP. Therefore, in ongoing examination of how to
remove the overhead crane, it is necessary to perform all safety assessments as an assumption,and it is important to carry out a comprehensive examination based on the perspectives of (i)formulating specific work procedures and work plans enabling identification of risk items, (ii) therisk scenario assumed from (i) and the measures, (iii) identification of points to consider such as
exposure of workers, from an operator's perspective, and (iv) rationality and impact on other work.61
Regarding the contamination state of the well-plugs of Unit 1 to 3, the Study Committee on
Accident Analysis of the Fukushima Daiichi NPS pointed out that the well- plugs have important
implications in safety and decommissioning work due to the high level of their contamination62.
Although the well-plugs of Unit 1 has been evaluated by the above mentioned Study Committee tobe about two orders of magnitude less contaminated than several tens of PBq of Units 2 and 3,
those in Unit 1 become deformed and unstable due to the impact of the explosion at the accident.
For this reason, TEPCO is studying the impact of falling well-plugs on the PCV in the event of anearthquake. It is necessary to make a comprehensive decision on how to handle these well-plugs,
based on the study results, and by taking into consideration the impact on the removal of fuel from
SFP and fuel debris retrieval in the later stage, and by performing thorough safety assessments.
While applying overseas findings, a detailed handling plan for 67 fuel assemblies with damaged
cladding tubes, which have been stored in Unit 1 SFP since before the accident, is underdevelopment toward the completion of fuel removal in 2031. In particular, efforts should be madeto ensure verification of the post-accident condition, examination/development of handling methods,
and risk study associated with handling.
(2) Unit 2
In Unit 2, fuel in SFP will be removed from the opening on the south side of the operating floorusing a fuel handling machine composed of a boom-type crane-system, which has not yet beenused for nuclear facilities in Japan. Since it is a new system, it is important to do the following: (i)to set up a design schedule with appropriate margins, (ii) to perform mockup tests fully simulating
on-site situations and operation methods and ensure feedback on the results to design andproduction, and (iii) to be sufficiently familiar with the operation and functionality of systems
beforehand in preparation for removal by remote operation63.
As mentioned above, fuel removal from SFP basically assumes unmanned operation by remotecontrol, however, manned operation is also assumed for system installation or system
61NDF, "Evaluation on the selection of fuel removal methods (plan) at the Fukushima Daiichi Nuclear Power

Station Unit 1”, Secretariat Meeting of the Team for Countermeasures for Decommissioning and Contaminated
Water/Treated Water (73rd), Material 3-2, December 19, 2019.
62NRA, "Draft revision of the interim report (draft) based on the results of public comments”, The Study Committeeon Accident Analysis of the Fukushima Daiichi Nuclear Power Station (19th meeting), Material 3 (pages 81 to83), March 5, 2021
63 NDF, "Evaluation on the selection of fuel removal methods (plan) at the Fukushima Daiichi Nuclear Power
Station Unit 2”, Secretariat Meeting of the Team for Countermeasures for Decommissioning and Contaminated
Water/Treated Water (71st), Material 3-2, October 31, 2019100

troubleshooting. In addition, it is recommended that the radiation dose on the operating floor is
reduced as much as possible, taking into account of high contamination of well-plugs indicated by
the above-mentioned the Study Committee on Accident Analysis of Fukushima Daiichi Nuclear
Power Station.62 In the air dose rate measurement and surface contamination measurement of theoperating floor in March and April 2021, it was confirmed that the radiation dose rate was reduced
by about 20% compared with the result in FY 2018 due to the results of previous cleanup work. It
is important to incorporate these survey results in decontamination and shielding installation
methods for further dose reduction of the operating floor.
3.4.2.2 Decision of future processing and storage methods
The future processing and storage methods for the fuel in SFPs need to be decided after
considering the impact of seawater and rubble exerted during the accident. The impact of seawaterand rubble has been evaluated for the fuel removed from Unit 4, and it is expected that the impact
is small. However, based on the situation of the fuel to be removed, it is necessary to advance theevaluation of long-term integrity and the examination for processing and to decide the future
processing and storage methods.
It is planned to transfer fuel in SFPs of all Units to the Common Spent Fuel Storage Pool by
2031. Considering tsunami risk, however, it is recommended that fuel in the Common Spent Fuel
Storage Pool be transferred to higher ground. Thus, TEPCO is considering fuel storage on higher
ground. It is necessary to prepare this storage facility from the perspective of safety and operator.
In the case of storage on higher ground, it is recommended that dry storage is used, in which
natural convection (ventilation) of the air is used for cooling, and system and maintenance/
management can be simplified. However, consideration should be given in reference to overseas
findings, including handling of fuel with damaged cladding tubes.

101

3.4.2.3 Summary of key technical issues
The main technical issues and plans described in this section are summarized as shown in Fig..
Fiscal year

2020

Related milestones

Unit 1

2021

2022

2023

2026

Complete installation of Unit 1 large cover(around FY 2023)▽

Removal ofbuilding cover(remaining parts)

Start Unit 2 fuel retmoval(FY 2024 to FY 2026)

2032(Year)

To complete fuel removal(in 2031)

Start Unit 1 fuel removal(FY 2027 to FY 2028)

Installation of large cover
Removal of rubble, the overhead crane, etc.
/well-plug disposal /decontamination and shielding of an
operating floor/installation of fuel handling machine, etc.

Fuel removal
Removing high-dose equipment/ Draining the pool

<Legend>

Unit 2

Yard cleanup/
Soil improvement, etc.

: Fuel retmoval work
: The period in whichchanges are expected: Works other than fuelretmoval: The period in whichchanges are expected

Decontamination and shieldingof operating floor
Installation of a working platform on
the south side and an opening/fuelhandling machine installation, etc.

Fuel removal
Removing high-dose equipment/ Draining the pool
Unit 3 fuel removal

Units 3 and 4
Removing high-dose
equipment in Unit 3/ Draining the pool

Units 5 and 6

Removing high-dose
equipment in Unit 4/ Draining the pool

Suspended while
work is performed at
Unit 2

Unit 6 fuelremovalSuspended while
work is performed
at Units 1 and 2

Unit 5 fuel removal

Dry cask manufacturing

Common Spent Fuel
Storage Pool
Cask

Transfer from the Common Spent Fuel Storage Pool to the
Dry Cask at the Temporary Cask Custody Area
Expanding of storage capacity in Dry
Cask Temporary Custody Facility

Examination, design and installation of dry storage facility (for Common Spent Fuel Storage Pool)
Carrying out fuel from Common Spent Fuel
Storage Pool/Dry storage on higher groundFig.34 Key technical issues and future plans for fuel removal from SFPs (Progress schedule)102

4. Analysis strategy for promoting decommissioning
4.1 Uncertainty of fuel debris, etc. and importance of analytical results
The accident at the Fukushima Daiichi NPS was the first core meltdown accident at a BWR in
the world, and there are no records of temperature and other plant parameters due to loss of powerat the accident. In addition, many uncertainties remain regarding the state inside the reactors, thestate of the fuel debris, and FP release paths, etc., due to the unclear operational status of thesafety equipment and the injection of seawater to bring the accident under control. Since there is
uncertainty in the formation process of fuel debris and it is not under human control, fuel debris isconsidered to have heterogeneity in various physical properties such as chemical composition,
microstructure and density.
If the uranium content is unknown and 97 - 98% of the uranium content of the fuel assembly
before the accident should be set as the value for criticality control or off-site transportation, andsafety evaluations/measures should be implemented accordingly. According to the calculation
results of the severe accident (hereinafter referred to as "SA") codes and the videos and photos of
PCV internal investigations, it is clear that the uranium content decreases by melting and mixing
with the surrounding structural materials. However, since there is no value to be used for theevaluation, an excessive margin is included in the safety measures. If the range of such uncertainty
can be reduced, there is no need to include excessive safety margins in safety assessments andsafety measures, and thus, the promptness and rationality of decommissioning can be improved.
In addition to conventional sample analyses, studies on reducing uncertainty of fuel debris
properties by other measurement methods have already started by the Project of Decommissioning
and Contaminated Water Management.
The analysis results of solid waste are important basic information for the study on
processing/disposal methods for various kinds of waste generated by the accident. The analysis
results of fuel debris are applied in a number of areas, including retrieval methods, storage
management, necessity of processing, investigation to determine the cause of the accident, andimprovement of nuclear safety. As shown in Fig. , their relationship changes with the progress madein decommissioning of the Fukushima Daiichi NPS. It is important to correctly recognize that theanalytical results are one of the important criteria for decisions for reducing the range of uncertainty
in the above examination for facilitating decommissioning. TEPCO, incorporating analysis results,
should take the lead in establishing and developing analysis systems, facilities, and functions thatcan efficiently collect and evaluate analysis results.

103

Fig. 35 Incorporation of analyses and investigation results, and their relationships

4.2 Three elements of analysis strategy
To safely and steadily proceed with decommissioning of the Fukushima Daiichi NPS, it isnecessary for TEPCO to establish and develop facilities for analysis and the functions required forhandling of solid waste or fuel debris. In addition, it is important to build a system that effectively
utilizes analyzed results for each decommissioning operation.
In order to obtain good analysis results, as shown in Fig. , it is effective to properly maintain (i)the methods and systems for analysis, (ii) the quality of the analysis results, and (iii) the size andquantity of sample.
・Establishing facilities, systems and coordination
・Assigning roles according to permission and facility
・Selecting the right analysis method for each purpose, etc.

Methods andsystems for analysis

・Increase the amount ofcollectable samples
・Examine the minimum
amount for analysis
・Consider number ofrepetitions, etc.

Size andquantity ofsampleGoodanalysis
results

Quality(accuracy)of analysis
results

・Reduction of insoluble
residues
・Countermeasures forinhibitory nuclides
・Reducing uncertainty, etc.

Fig. 36 Three elements of the fuel debris analysis strategy

104

4.3 Current status of establishing an analysis system and strategy
TEPCO currently utilizes three facilities, which are the laboratory for Units 5 and 6, the chemical
analysis building, and the environmental management building (for pretreatment only), andperforms the analyses required for facility administration and progressing with decommissioning.
They are also planning to establish a facility for the analysis necessary for smooth performance ofroutine analyses related to processing/disposal of solid waste and fuel debris retrieval in the future.
Moreover, as fuel debris retrieval proceeds, the risk of α-nuclides intake is assumed to gradually
increase, and establishment of a bioassay function that contributes to assessment of internal
exposure is planned.
Some deposit samples and solid waste samples are currently transported to JAEA’s facility for
analysis in Ibaraki prefecture, private facilities for analysis (Nippon Nuclear Fuel Development Co.,
Ltd. (hereinafter referred to as “NFD”), and Nuclear Development Co., Ltd. (hereinafter referred toas “NDC”), and analyzed there. It takes a lot of time and is costly to transport samples from the
Fukushima Daiichi NPS to these facilities for analysis, i.e., off-site transportation using public roads.
As an essential facility for decommissioning of the Fukushima Daiichi NPS, the JAEA is
proceeding with the construction of Radioactive Material Analysis and Research Facilities (facility
management building, building #1, building #2) adjacent to the Fukushima Daiichi NPS64 under the
supplementary budget of the Government (FY2012)65. At commencing operations in buildings #1
and #2, they will be designated as controlled areas of the Fukushima Daiichi NPS, which has theadvantage that off-site transportation is not required. Leveraging this, it is effective to promptly
identify basic physical properties such as nuclide inventory, chemical composition, chemical form,
shape and density, and incorporate them into safety assessment and work procedures 66 .The
purpose of building #1 is to analyze solid waste, and building #2 is to analyze fuel debris. The
facility management building began its operation from 2018, building #1 commenced
comprehensive functional tests from February 2021, and building #2 is undergoing safety review.
However, the operation start of building #1 scheduled for June 2021 has been delayed due to a
malfunction of the air supply/exhaust system. 67 Therefore, it is necessary to examine further
utilization of facilities for analysis in the Ibaraki area and to organize division of roles according to
the characteristics of the facilities for analysis in and near the Fukushima Daiichi NPS and those in
the Ibaraki area. However, since all the facilities for analysis in the Ibaraki area have been in
operation for more than 30 years, considerations on measures are required for aging facilities thatwill be used continuously.

64 The 52nd Secretariat Meeting of the Team for Countermeasures for Decommissioning and Contaminated Water

Treatment, “Material 3-4: Opening of the Okuma Analysis and Research Center facility management building”
65The 24th meeting of the study group on monitoring and assessment of specified nuclear facilities, “Material 3-1:

Development of R&D hub facilities for decommissioning”

66The 68th meeting of the study group on monitoring and assessment of specified nuclear facilities, “Material 1-4:

Okuma Analysis and Research Center”
67The 87th Secretariat Meeting of the Team for Countermeasures for Decommissioning and Contaminated Water

Treatment, “Material 4: Shortage of the air volume of the air intake/exhaust system in the building #1 of theRadioactive Material Analysis and Research Facility, and reconsideration of the time of operation start”

105Not only the facilities for analysis in the Ibaraki area, but also the Radioactive Material Analysis
and Research Facilities to be operated in the area adjacent to the Fukushima Daiichi NPS are shortof the human resources required for stable facility operation, and the securing and maintaining ofanalytical engineers needs to be considered. In this respect, it is necessary to consider in advancethe qualities expected of analytical engineers in various analytical works, and to develop a plan sothat the required roles are appropriately achieved.
In a normal nuclear power plant, fuel is sealed in fuel cladding, and before the accident, theunsealed α-nuclides were not directly handled in the Fukushima Daiichi NPS. The fuel debris
generated by the accident contains unsealed fuel, FP and activation products, and there is a riskof internal/external exposure or spread of contamination in performing analyses. Therefore, in
addition to knowledge on general chemical analysis and electronic equipment analysis, a wide
range of knowledge is required, including radiation protection, laws and regulations related tonuclear power and radiation, chemical reactivity of fuel and materials, physical, chemical andbiological properties of radioisotopes, properties of various types of radiation and measurement
methods. Although such knowledge is acquired over time, TEPCO must develop human resources
for fields where there is little experience in as short a time as possible. Until now, JAEA and Japan
Nuclear Fuel Limited have accumulated sufficient knowledge and experience on the handling of αnuclides and fuel analysis techniques. Therefore, with the cooperation of JAEA and Japan Nuclear
Fuel Limited, it is necessary for TEPCO to promptly work on developing analytical technicians.
Personnel exchanges have already started between TEPCO and JAEA for the purpose ofdeveloping human resources engaged in handling and analysis of α-nuclides. In FY 2020, two
TEPCO employees were transferred to JAEA, and eleven JAEA employees were transferred to
TEPCO, and such transfers will continue.
However, since there are few talented personnel (analysis evaluators) who can design theanalytical range and items in anticipation of how to use the analysis results in advance, it is also
important to make efforts in increasing such personnel. TEPCO, which is responsible fordecommissioning the Fukushima Daiichi NPS and who needs to evaluate the analysis results andincorporate them into each process, should take the lead in increasing the number of analyticalevaluators.
Blind tests68 by JAEA, NFD and NDC have been performed on simulated samples produced by
Tohoku University in order to improve not only analytical accuracy but also a wide range ofanalytical techniques. 69 As a measure for increasing the number of analytical evaluators, it iseffective to expand the scope of knowledge by performing this blind test and the comprehensive
evaluation on fuel debris properties described in 4.4. Through these nationwide initiatives, it isexpected that cooperation among organizations will be promoted for analysis, effective use of68

A blind test is a test conducted to acquire an objective evaluation of a product without notifying the subject of the
product information.
69 IRID (2021) Supplementary budget in FY 2020, “Subsidies for the Project of Decommissioning and Contaminated
Water Management”, (Development of technologies for improving analytical accuracy and estimation of thermal
behavior of fuel debris), final report in FY 2020, August 2021

106

facilities, development of analytical human resources and technical improvement, and expansionof knowledge.

4.4 Improvement of the quality of analysis results, diversification andexpansion of analysis methods
4.4.1 Improvement of the quality of analysis results
Fuel debris contains difficult-to-measure nuclides, interfering elements, insoluble materials, etc.,
and it is considered difficult to conduct complete composition analysis. It is also an important
perspective to question the analytical result of the samples in consideration of the impact of theerror factor. Monitoring data, sampling analyses, PCV internal/on-site investigation, analyses using
SA codes, and past knowledge and experimental results have been accumulated. As part ofverification of sample analysis results, deriving consistent property evaluations in reference toanalysis, surveillance and test results leads to improving reliability of analysis results, and thereby
quality of the analysis results.
Fuel debris caused by the accident is a mixture of fuel and materials in the core. It is important
to comprehensively review/evaluate at what stage of the accident progression the substance wasproduced, what elements were mainly contained, and what properties they have. This also enables
provision of feedback on instructions for necessary sampling to cascade analysis results, insteadof unnecessarily increasing the number of samples to be analyzed. As shown in the conceptual
diagram shown in Fig. , it is expected that a flow will be established from [(1) Fuel debris sampling],[(2) Sample analysis], [(3) Evaluation of the characteristics] to [(4) Safety Assessment] and thecycle from [(1) Fuel debris sampling], [(2) Sample analysis], [(3) Evaluation of the characteristics]
to [(6) Instructions for the next sampling]. TEPCO and JAEA are already cooperating in
implementing forensic activities that estimate accident behavior and causes by comparing theresults of sample analyses with mock-up testing on progression of meltdown and past scientificknowledge, and it is recommended to further expand these activities70.

70

IRID (2021), Supplementary budget in FY 2018, “Subsidies for the Project of Decommissioning andContaminated Water Management”, (Development of analytical and estimation techniques for characterizationof fuel debris), Results for FY2020 implementation107

(1) Fuel debris sampling(6) Instructions forthe next sampling(2) Sample analysis(analysis results)

Monitoring data
Analysis and investigation(3) Evaluation of the characteristics
based on analysis results

SA code calculation
Past knowledge and experiments(7) Request of analysis
result for evaluation

Monitoring data

(4) Safety assessment on retrieval, storageand management of fuel debris

SA code calculation
Past knowledge and experiments(5) Safety measures for retrieval, storageand management of fuel debris

Fig. 37 Conceptual diagram from fuel debris analysis to evaluation/countermeasures
4.4.2 Diversification and expansion of analysis methods in light of sample size andquantity constraint
The current sample analysis is mainly performed using an electronic microscope after
transporting smear samples71 to a facility for analysis in Ibaraki area72. Since the density, hardness,and other items cannot be measured for micro or small samples, it is necessary to increase thesize and quantity of the samples in accordance with the progress of the fuel debris retrieval process.
It is difficult to analyze a sufficient number of samples because the analysis is performed by using
a manipulator in each process in a hot cell73, and by about 0.5 to 1 samples per month in one facility.
Fig. shows the estimation results of the accumulated mass of fuel debris in Unit 3 and samples tobe analyzed. In the figure on the left, the blue area indicating the cumulative mass of the sample is
not visible, but in the figure on the right with the vertical axis enlarged, the blue area becomesvisible. The fuel debris in Unit 3 is estimated to be approx. 364 tons74, while the total amount thatcan be analyzed in the facility for analysis is estimated to be approx. 45 kg/reactor. This means that
only about 0.01% of the fuel debris to be retrieved is analyzed, and there is a large gap between
the amount to be retrieved/stored and the amount to be analyzed. As one of the criticality measures

71 Smearing is a method of examining surface contamination by measuring the amount of free radioactive

materials by wiping a surface where contamination is detected with filter paper, etc. This is called the smear
method.
72 The 84th meeting of the Team for Countermeasures for Decommissioning and Contaminated Water
Treatment/Secretariat, “Material 3-3: Analysis results of samples related to the internal investigation of the Unit
1 - 3 PCVs”
73 A hot cell is a compartment surrounded by shielding materials such as concrete and iron to analyze and test
highly radioactive materials. In order to prevent leakage, the inside of the compartment is maintained below
atmospheric pressure, and radioactive materials are handled by a manipulator and tongs.
74 Supplementary budget in FY 2014, the subsidies for the Project of Decommissioning and Contaminated Water
Management (Sophistication of internal PCV state analysis by severe accident progression analysis and actual
equipment data, etc.), completion report (March 2016), Collaborative Innovation Partnership, International
Research Institute for Nuclear Decommissioning (IRID), The Institute of Applied Energy (IAE)108

in the hot cell, the upper limit is set for the amount of fuel handled and stored, and this calculation
also assumes to set the upper limit.
*The right figure shows an enlarged

Approx.0.01%
of fuel debris

Fig. 38 Estimation results of the cumulative mass of fuel debris to be retrieved from Unit 3 andsamples to be analyzed
Since fuel debris has heterogeneity, the analytical values vary depending on the sampled parts,and the situation is such that a sufficient amount of fuel debris cannot be analyzed, resulting in awide range of uncertainty in evaluation. Applying this situation to Fig. 36, it is effective to diversify
and expand the analysis methods and to perform comprehensive evaluation in order to expand thearea of good analytical results, where three circles overlap, regardless of the restrictions on the
improvement of analysis quality and sample quantity. For example, in addition to performing
unsealed sample analysis in a facility for analysis, it is recommended that in-situ analysis 75(simplified analysis) and non-destructive measurement are performed to increase the amount of
information on the sample, including coordinating information during collection, and to consolidate
the results to perform evaluations for controlling the range of uncertainty within a certain level.
However, there are no practical examples of in-situ analysis or non-destructive measurement offuel debris, and remote operation is required because of the high radiation dose of the subject.
Therefore, it is necessary to promote research and development focusing on specific items suchas uranium content.
Fig. illustrates an example timing of in-situ analysis and non-destructive measurement in theprocess from fuel debris retrieval to storage. In-situ analysis should focus on specific items at or in
the vicinity of the retrieval worksite before transporting samples to the facility for analysis. For
example, the content of uranium entering and fusing into concrete or steel materials becomes zeroat the end of the retrieval work. Samples must be transported to a facility for analysis to confirm bycompositional analysis that fuel debris retrieval at the worksite has been completed. This is the final
stage of fuel debris retrieval and, by that time, the properties of fuel debris could be identified
through sample analyses. Therefore, rather than investing a lot of time and resources intransportation, it is effective for the whole process to perform in-situ analysis at many parts and
75 An analysis performed by bringing analytical equipment to the location where the sample can be obtained, as

opposed to the analysis that is performed by taking the sample for analysis back to the location where the analysis
equipment is located.

109

quickly verify the existence of uranium. This is important in terms of the compatibility between
shortening of transportation process to the facility for analysis, cutting process of structures, andincrease in analytical information.
In the case of non-destructive measurement, moreover, it is appropriate to provide nondestructive measurement after retrieving fuel debris, adding neutron absorber materials, and drying
it, but before transition from the fuel debris retrieval stage to the storage stage. If the timing of nondestructive measurement is before sealing the storage container, fuel debris and solid waste canbe sorted out. If it is after the storage container is sealed, pre-operation inspection of the long-term
storage phase can be conducted leveraging the condition with zero risk of the spread of α-nuclides
contamination. Since the significance, purpose, and difficulty of nondestructive measurement offuel debris differ before and after sealing them in the storage container, thorough examination of
the applicability to fuel debris containing neutron absorbers as well as the feasibility of combining
various nondestructive measurement methods is required.

PCV internal investigation* Record of coordinate information and visual inspection
Sample analysis

Trial retrieval• Suitable items for analysis
and measurement

Contribution to reduction in
number of cutting and relieve
future load, not mandatory

Removal of obstacles
Database of fueldebris characteristics

In-situ analysis

U≦ND
U>ND

Solid waste

Gradual expansion of
Fuel debris
fuel debris retrieval
燃料デブリ* Record of coordinate

Sample analysis* Increase in items

Dryinginformation and visual inspection

Further expansionof retrieval scale

Sample analysis* Decrease in items

U>ND

Important for storage andmanagement of fuel debris

Neutron absorber

Non-destructive
measurement

Categorization* Content rate and distribution of U

Storage and management* Record of coordinate
information and visual inspection

In-situ analysis

U≦ND
End of fuel debris retrieval

Contribution to reduction in numberof transportations, not mandatory

Storage Phase
Fuel debris retrieval phase
Phase
Phase 保管・管理の
燃料デブリ取り出しの*※固体廃棄物については、別途、固体廃棄物をSolid waste will be analyzed and measured
separately
for storage and management.
保管・管理するための分析・測定が行われる。

Fig. 39 Example timing of in-situ analysis and non-destructive measurement in theprocess from fuel debris retrieval to storage stage1105. Efforts to facilitate research and development
5.1 Significance and the current status of research and development
There are many difficult technical issues requiring research and development to promote thedecommissioning of the Fukushima Daiichi NPS from the perspectives of safe, proven, efficient,
timely, and field-oriented. At present, when trial retrieval of fuel debris is imminent, it is necessary
to accelerate research and development in consideration of the practical application for a gradualexpansion of retrieval scale and further expansion of fuel debris retrieval.
In order to solve these technical issues, basic/fundamental research and application research by
universities in and outside of Japan and researching institutions such as JAEA, practical application
research and field demonstrations by IRID, manufacturers including overseas enterprises and
TEPCO, etc., are being performed by various industrial-academic-governmental institutions (Fig.
40).
The Government is supporting highly-difficult ones among application research, practicalapplication research and field demonstrations by “The Project of Decommissioning andContaminated Water Management”, and ones regarding basic/fundamental research by “The
Nuclear Energy Science & Technology and Human Resource Development Project” (hereinafterreferred to as "World Intelligence Project").
TEPCO is engaged in technical development that directly leads to practical application. NDF
considers R&D medium-and-long term plans, next-term R&D plans, and promotes these plans andsupports the World Intelligence Project. With institutes involved as its members, moreover, NDF
has established the Decommissioning R&D Partnership Council, which considers information
sharing on needs and seeds for R&D, adjustment of R&D based on the needs of decommissioning
work, and issues on promoting cooperation in R&D and human resource development. Moreover,coordination between the Project of Decommissioning and Contaminated Water Management andthe World Intelligence Project has been promoted through the Decommissioning R&D PartnershipCouncil. Based on the discussions at this Council, NDF set up “a task force on researchcollaboration”76, and identified six Essential R&D Themes mainly in the area of basic/fundamental
research to be addressed with priority, in light of investigational issues and problem awareness on
the side of the needs (Attachment 13). The R&D implementation system is shown in Fig. 41. Also,in promoting research and development, it is important to make use of the Naraha Remote
Technology Development Center, the Okuma Analysis and Research Center, and Collaborative
Laboratories for Advanced Decommissioning Science of JAEA (hereinafter referred to as“JAEA/CLADS”) and to arrange the decommissioning R&D base, including international
perspectives.

76It was established in NDF in 2016 and held 6 times in total. It consists of NDF, intellectuals fromuniversities/researching institutions, and TEPCO. Six Essential R&D Themes were identified, which should
strategically and preferentially ascertain the principle in promoting decommissioning at the Fukushima Daiichi
NPS, and the interim report on the task force on research collaboration was developed (November 30, 2016).

111

It is important for the promotion of effective/efficient R&D to create a database of knowledge on
R&D results, etc., and thus it is necessary for decommissioning operators and R&D institutions toaddress it. It is also important for these institutions to collaborate and cooperate to compile andarchive information/research results obtained during the decommissioning work of the accident
reactors, and to make them available to those who will be engaged in the associated research anddevelopment. NDF is currently examining the method.

Fig. 40 Scope of R&D for decommissioning and implementation entities

Fig. 41 Overview of the R&D structure of the decommissioning of Fukushima Daiichi NPS

112

5.2 Key issues and strategies
5.2.1 Updating the R&D medium-and-long term plan
In October 2020, NDF and TEPCO developed the R&D medium-and-long-term plan overlooking
the overall decommissioning research and development for about ten years in order toappropriately manage the extraction and execution of the required research and development forfuel debris retrieval and waste management after the Government’s Medium-and-Long-term
Roadmap indicated the course of action for about ten years and the promotion policy of researchand development to support it, and based on the Medium-and-Long-term Decommissioning Action
Plan 2020 of TEPCO. Attachment 14 shows the R&D medium-and-long-term plan updated basedon the Medium-and-Long-term Decommissioning Action Plan 2021 revised in March 2021,
progress in R&D and Prospects of processing/disposal methods and technology for waste related
to its safety.
In reviewing the R&D medium-and-long-term plan, TEPCO, the Agency for Natural Resources
and Energy, and NDF have held R&D planning meetings, and shared the plan with the Ministry of
Education, Culture, Sports, Science and Technology (MEXT) to discuss the required results, timing,
implementation structure, etc.
The plan will be updated and expanded continuously based on the information made clear by
the progress of PCV internal investigation and fuel debris analyses.
5.2.2 Initiatives for the Project of Decommissioning and Contaminated Water Management
5.2.2.1 The Project of Decommissioning and Contaminated Water Management
The Ministry of Economy, Trade and Industry is providing support through the Project ofDecommissioning and Contaminated Water Management for research and development to solve
technical issues that are difficult to solve in decommissioning. As shown in Fig. 40, the Project
selects project operators, such as IRID, and other companies conducting R&D, and provide
necessary assistance through corporations with funds and the secretariat. Through the Project todate, significant results, such as the progress in understanding the internal state by PCV
investigations, have been obtained for promoting decommissioning work. Attachment 15 outlines
the implementation of the Project up to now.
In the Project of Decommissioning and Contaminated Water Management, since last fiscal year,
NDF has been participating in the secretariat to strengthen the functions of project planning andprogress management, while TEPCO jointly applies for issuance in cooperation with researchleadership and provides project management to incorporate requirements in terms of actual site
applicability.
As a result of strengthening such functional enhancement, project proposals have been made,which incorporate TEPCO’s on-site needs and actual site applicability in more detail, and the
projects have been executed so that the results of such proposals can be effectively applied to theengineering work by TEPCO. As the secretariat for the Project of Decommissioning and113

Contaminated Water Management, NDF will provide progress management, including verification
as to whether the expected results are achieved in time of need, to meet the intended project
objectives.
While actively participating in this system, it is important for TEPCO to increase their focus on
their own R&D activities, and strengthen their structure. In August 2021, TEPCO has enhanced thesystem to promote the examination and implementation of technology development in the future.
5.2.2.2 Next-term R&D Plan
In order to support the Project of Decommissioning and Contaminated Water Management, andbased on the R&D medium-and-long-term plan, every fiscal year NDF discusses the R&D to be
carried out in the next two years with parties concerned at the R&D planning meeting, and thendevelops the Next-term R&D Plan. At that time, the plan is first deliberated on by the Fuel Debris
Retrieval Expert Committee and Waste Management Expert Committee, and then by theDecommissioning Strategy Committee. After this, it is summarized as an NDF proposal. This plan
was reported by the Ministry of Economy, Trade and Industry (METI) to the Secretariat Meeting of
the Team for Countermeasures for Decommissioning and Contaminated Water/Treated Water
Treatment, and The Project of Decommissioning and Contaminated Water Management has been
implemented accordingly.
In considering the next-term R&D plan, it is necessary to evaluate the R&D results to date,moreover, to identify issues that should improve its achievement and issues to be newly addressed,
as well as to identify emerging issues and organize technical issues with a view of the R&D
medium-and-long-term plan. When identifying the issues, it is also important to identify themexhaustively, to confirm whether each issue is in line with the needs of TEPCO, as the entity
responsible for decommissioning, and to aim for R&D results to be utilized for TEPCO’s engineering.
Based on the next-term R&D plan, there are ongoing projects for gradual expansion of retrieval
scale, further expansion of fuel debris retrieval, and waste management. In connection with the
projects for “development of maintenance technology for remote operation devices” and“development of assistive technologies for integration management of decommissioning of 1F(development of continuous monitoring system in PCV)” started this fiscal year, it is particularly
important to determine the direction for development, and therefore, opinions from the Fuel Debris
Retrieval Expert Committee have been solicited to promote development.
5.2.2.3 Research and development implementation system for the Project ofDecommissioning and Contaminated Water Management
The IRID has played a major role in research and development for decommissioning for about
ten years since the time that the post-accident situation inside the reactors was unknown. In
particular, the IRID has established a good track record in internal PCV condition analysis throughits internal investigations, and in developing fuel debris retrieval equipment and storage containers.
Meanwhile, as the engineering work by TEPCO progresses, the situation in the reactors and theneeds are gradually becoming clear. In addition, development is currently being promoted with the114

engineering work by TEPCO, which is a shift from joint activities through the Collaborative
Innovation Partnership. In light of this changing situation, consideration is being given to a structure
for after summer 2023, which is its deadline set by IRID.
In the Project of Decommissioning and Contaminated Water Management, it is important toestablish a proper R&D implementation structure. The continuity of R&D activities that have been
conducted mainly by the IRID should be ensured, and researchers/developers should cooperate
more closely with TEPCO.
The results of the IRID through the Project of Decommissioning and Contaminated Water
Management have been achieved with the national budget, and decommissioning of the
Fukushima Daiichi NPS is a national/social issue. Therefore, it is important to establish an easy toaccess system that enables organizations involved in the research and development fordecommissioning to make effective use of the R&D results.
5.2.3 Promotion of cooperation between decommissioning sites and universities/
researching institutions
5.2.3.1 The Nuclear Energy Science & Technology and Human Resource Development
Project
Universities/researching institutions tasked with basic/fundamental research are expected tomaintain and develop human resources, knowledge and infrastructure to make a quick responsewhen technical issues requiring scientific knowledge occur. It is important that universities/
researching institutions share awareness of issues in the field of decommissioning. In order tofacilitate the long-term decommissioning project of the Fukushima Daiichi NPS, it is important to
conduct scientific and technological investigations based on understanding of the principles andtheories from the medium-and-long term perspectives.
In this background, MEXT has been promoting fundamental/basic research and humanresource development activities, which contribute to solving issues such as the decommissioningof the Fukushima Daiichi NPS, by bringing together domestic and overseas intelligence fromuniversities/researching institutions as the World Intelligence Project since FY 2015, crossing
barriers of the nuclear field, and through close coordination and alignment including international
joint research. From new subjects adopted in 2018, the leadership was transferred from the MEXT
to JAEA/CLADS to strengthen cooperation between JAEA/CLADS and universities, and establish
a system to implement medium-and-long term R&D and human resource development, contributing
to decommissioning more stably and continuously.
The six Essential R&D Themes mentioned above have been discussed and further deepened
through the Platform of Basic Research for Decommissioning 77 , a council to promote basic/
fundamental research, and utilized in the public invitation to the World Intelligence Project. In
soliciting applications for the current World Intelligence Project, JAEA/CLADS is using the “overall
77 Council for promotion of basic/fundamental research through joint management of JAEA/CLADS and theMinistry of Education, Culture, Sports, Science and Technology's program for strengthening research andhuman resource development in the World Intelligence Project. Refer to:https://clads.jaea.go.jp/jp/about/platform.html

115map of the basic/fundamental research”, which provides an overview of the entire decommissioning
process from contaminated water management to waste processing/disposal, including these six
Essential R&D Themes, and identifies the R&D needs and seeds required. Furthermore, as
described in 5.2.1, MEXT has participated in R&D planning meetings to share information on theimportant basic/fundamental R&D, and used such information for project planning in the Project ofDecommissioning and Contaminated Water Management.
5.2.3.2 Collaboration between the Project of Decommissioning and Contaminated Water
Management and the World Intelligence Project, and initiatives for businessacademia collaboration by TEPCO
Some basic/fundamental research contributing to problem-solving in on-site decommissioning
have recently obtained outstanding research results mainly in the World Intelligence Project. It is
an important issue to directly apply the results in on-site of decommissioning. In order to achievethis, it is essential to match the needs of the decommissioning site with the seeds of universities
and researching institutions, build bridges for excellent results, while making use of the overall mapof the basic/fundamental research. Based on such awareness, JAEA/CLADS has been tacklingassessment of actual site applicability for its adopted subjects in the World Intelligence Project todate, and on-site application as the result of its assessment in cooperation with TEPCO. Moreover,the “steering committee78” that presents the basic policy for the operation of the World Intelligence
Project has been established since July 2020 by the Ministry of Economy, Trade and Industry incharge of the Project of Decommissioning and Contaminated Water Management and domestic
decommissioning related makers. In FY 2020, TEPCO also joined as Program Officer (PO) that is
responsible for research management as well as the screening member for selection in the public
invitation process, leading to strengthening and accelerating the efforts to apply the perspective of
the needs of decommissioning in the Project.
At the 9th Decommissioning R&D Partnership Council held in February 2021, it was proposedthat the MEXT and METI work together to resolve decommissioning issues to further deepen
needs-driven R&D. In response to this, it is important to further strengthen ties by sharing
information on R&D planning meetings and the results of both projects, and to deepen alignment
between the World Intelligence Project and the Project of Decommissioning and Contaminated
Water Management. In this manner, the Decommissioning R&D Partnership Council is required tocontinuously serve as a general coordinator, such as by promoting effective alignment between the
R&D seeds and the needs of the decommissioning work as the initial task of the Council. For thispurpose, it should be noted that TEPCO’s decommissioning needs are becoming clearer along withthe progress in decommissioning compared to the time when the Council was established in 2015.
TEPCO is required to make more effective use of this Council according to the progress of
decommissioning, for example, by communicating the needs of decommissioning work moreproactively at Council meetings and strengthening interactive information sharing with
universities/researching institutions, manufacturers, and academic societies.
78 Consisting of program directors, intellectuals from universities/researching institutions, NDF, TEPCO, andmanufacturers, and established in the Ministry of Education, Culture, Sports, Science and Technology.

116

Since FY 2019, TEPCO has also started joint research with universities based on the results of
the World Intelligence Project to discover the technological seeds owned by universities fordecommissioning 79 . The Government, JAEA/CLADS, NDF, TEPCO, and other organizations
involved need to further strengthen their cooperation for better matching needs with seeds andserve as a bridge to share outcomes.
5.2.3.3 Establishment of the centers of basic research/research infrastructure
In order to make the long-term decommissioning of the Fukushima Daiichi NPS proceed steadierin technical aspects, it is essential to work on developing R&D infrastructure and accumulatetechnological knowledge, develop generic technologies and collect basic data, including theessential R&D themes, building up research centers, facilities and equipment, and human resource
development. Decommissioning of the Fukushima Daiichi NPS is an opportunity for trialing stateof-art science and technology and accumulation of such activities is expected to become a sourceof innovation.
In the building for International Research Collaboration of JAEA/CLADS (Tomioka-machi,
Fukushima Prefecture), universities, researching institutions, industries, etc., inside and outside
Japan form a network and promote research and development and human resource development
in an integrated manner. It is expected that a network will be formed for the exchange of diverse
human resources from Japan and abroad, such as universities, researching institutions, andindustries, and that JAEA/CLADS will become a central organization serving as a hub for suchactivities. For this reason, the World Intelligence Project has been implemented as a JAEA/CLADS
project since FY 2018, transferred from the Ministry of Education, Culture, Sports, Science andTechnology. The “Program for Decommissioning Research by Human Resource Development” wasnewly established in the World Intelligence Project in FY 201980. It created a hub of research andhuman resource development (Collaboration laboratory) in both educational researching
institutions and JAEA/CLADS, and has started R&D and a human resource development projectthat connects these organizations by cross-appoint system 81 . It contributes to functional
enhancement of JAEA/CLADS bases.
It is also important to build research and development infrastructures as hardware. The Naraha
Remote Technology Development Center of JAEA, which began full-scale operation in Naraha
Town, Fukushima Prefecture, in April 2016, is a facility with various systems such as mock-up
facilities for the development and demonstration of remote-control devices and equipment. In
particular, it is essential to conduct full-scale mock-up tests, prior to the introduction of equipment
into a severe environment that cannot be accessed by humans, not only for performanceverification but also for training and establishment of operating procedures, etc. Therefore, active

79Tohoku University, Fukushima University, the University of Tokyo, and Tokyo Institute of Technology (Joined in

FY 2020 except the University of Tokyo)

80 For JAEA, the bases in Tokai and Oarai can also be used in addition to the one located in Fukushima

Prefecture.

81 A system that enables universities, public researching institutions, and private companies to enter into

employment contracts with each other and operate under the responsibility of each institution.

117

use of these tests by operators is recommended. Hereafter, a mock-up test of the arm-type accessequipment for trial retrieval of fuel debris is scheduled. Use of this facility will become even more
important for facilitating decommissioning at the Fukushima Daiichi NPS.
Furthermore, in Okuma Town, Fukushima Prefecture, construction of the JAEA Okuma Analysis
and Research Center (Radioactive Material Analysis and Research Facility) is ongoing. So far,starting with the opening of the facility management building, the construction of the building #1,
which mainly handles low-radiation dose samples such as rubble, is well underway. The detailed
design of the building #2, which handles high-radiation dose samples such as fuel debris, is inprogress.
In this way, research facilities related to decommissioning projects are located in Fukushima
Prefecture, where a global center for R&D for decommissioning has been established, and R&D
infrastructures for medium-and-long-term prospects have been built.

118

6. Activities to support our technical strategy
6.1 Further strengthening of project management and improvement ofcapability required as a decommissioning executor
6.1.1 Significance and current status of project management
In order to smoothly proceed with the entire decommissioning project while coordinating andharmonizing, it is necessary to establish a management system in which the organizations involvedin the project work together to achieve the goals and enhance their overall capabilities.
The individual work in each work area of a decommissioning project generally proceeds throughthe following processes: research and development, conceptual design, basic design, detailed
design, manufacturing, on-site installation, inspection, and operation. In addition, the NRA willconduct reviews and inspections, as necessary. In order to carry out such a series of processes
without omission or delay, it is effective to set up the major workflow defined in the long-term plan
as individual projects, which are management units of an appropriate scale. It is then important to
optimize the interrelationships and chronological relationships among the projects, and to proceedwith overall consistency under a sophisticated project management system so that the risks
inherent in the projects can be appropriately managed. From this perspective, TEPCO has beenworking to build and strengthen its project management system, and in April 2020, then it was
reorganized, and the general framework of the management system and the structure wasestablished. In the future, it is important to enhance and upgrade the management methods androoted in the on-site operations as an effective system. The following are examples of major
approaches until FY2020.
① Strengthening the authority of project managers through reorganization, and improving safety
and quality levels
In April 2020, the Fukushima Daiichi Decontamination and Decommissioning Engineering
Company was reorganized to establish a program82/project structure and the Decommissioning
Safety & Quality Office was set up directly under the Chief Decommissioning Officer (CDO). As aresult, the authority of project managers has been strengthened by assigning full-time project
managers and granting budget implementation authority. In terms of operation, efforts were madeto improve the operational propelling force of the decommissioning project by establishing a system
in which management and other related parties share information on the progress, issues and risksof each program and project every month. Moreover, the Decommissioning Safety and Quality
Office was set up to ensure the safety of decommissioning work and to maintain and improve thelevel of work quality in the face of the extremely uncertain and technically challenging tasks suchas fuel debris retrieval.
However, there are many issues for project-based operations to reach a mature level. For
example, workforce and work content should be properly controlled so as not to ruin motivation of

82 Program is the upper level of project, and it is a project in which multiple projects are organically combined torealize the overall mission.

119

members due to fluctuations in workload between projects. Operations in the decommissioning
safety/quality assurance office, a wide range of activities are needed, including establishing asystem involved in engineering work by each lead group and developing members specializing insafety.
② Preparation of a plan focusing on long term perspective (Mid-to-Long-term Decommissioning
Action Plan)
Since the accident at the Fukushima Daiichi NPS, TEPCO has conducted decommissioning
projects in accordance with the requirements based on the Act on Special Measures Concerning
Nuclear Emergency Preparedness and the Nuclear Reactor Regulation Law83, and the target
process of the Mid-and-Long-term Roadmap determined in the Inter-Ministerial Council for
Contaminated Water, Decommissioning and Treated Water issues. In March 2020, looking at thetarget process, TEPCO made the decommissioning project transparent to local communities andsociety and clarified its proactive attitude toward decommissioning while clarifying the complex
andlong-term

work

outlookbypreparingandpublishingtheMid-and-Long-Term

Decommissioning Action Plan 2020. In March 2021, TEPCO published the Mid-to-Long-Term
Decommissioning Action Plan 2021, which was updated based on the progress of the work.
③ Formulation of R&D medium-to-long-term plan
The decommissioning of the Fukushima Daiichi NPS has entered a "strategic stage of foresight"
and is entering a "phase of systematically tackling unexplored areas" such as the retrieval of fueldebris whose location and properties are unknown. For this reason, collaboration with R&D is
becoming increasingly important, and TEPCO is preparing the R&D medium- to long-term planjointly with NDF starting in FY2020. Going forward, TEPCO will need to strengthen its R&D
planning system to promote the examination of technological development issues andimplementation plans, in order to further strengthen its R&D planning and examination functions,etc.
④ Enhanced budget planning
Each fiscal year, budget plans for program/project work, non-program work (maintenance andutility facilities), and operating costs necessary for decommissioning are formulated, and thesteady implementation of work necessary for decommissioning and the improvement of budget
accuracy have been achieved through the appropriation of subjects based on thedecommissioning mid- to long-term action plan and efforts to finalize designs at an early stage.
On the other hand, there are some discrepancies between the budget and the actual results dueto the gap between the plan and the actual results in the budget year and the weakness of thedesign forecast, etc. Therefore, it is necessary to continue the efforts to improve the budget
accuracy.

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

120

⑤ Addressing issues across projects
TEPCO constantly identifies and shares risks and issues, and each project deals with them, but
for issues that do not fit into individual programs/projects, a cross-divisional system is establishedto deal with them. For example, in the decommissioning of the Fukushima Daiichi NPS, theprevention of the spread of alpha-nuclides to the environment and the prevention of internal
exposure of workers have been identified as future issues, and a cross-program/project working
group has been established to deal with these issues.
6.1.2 Key issues and strategies to be strengthened in the future
As shown in 6.1.1, various measures are being implemented to ensure the safe and reliable
implementation of the decommissioning project, and it is necessary to continuously upgrade andimprove the operation of each of these measures. The following is the important items to bestrengthened in the future.
6.1.2.1 Safety and Operator’s Perspectives and promulgating the "Safety FirstIn April 2021, TEPCO received an order from the NRA to prohibit the transfer of specified nuclear
fuel materials due to non-conformities in the protection of nuclear materials at the KashiwazakiKariwa Nuclear Power Station (hereinafter referred to as “Kashiwazaki-Kariwa”). As a result,
TEPCO's nuclear regulatory inspection response category became Category 4 ("the objectives of
the activities in each monitoring area are satisfied, but there is a prolonged or significant
deterioration in the safety activities conducted by the operator"), and many inspections will now be
conducted under the supervision of the NRA.
The operation of Fukushima Daiichi NPS differs from that of the Kashiwazaki-Kariwa in terms of
the form of business, and as for the management system, the full-scale project management has
been implemented in the Fukushima Daiichi, which is also different compared to KashiwazakiKariwa. In February 2021, on the other hand, a non-conformance event occurred, where
information on the failure of the seismometer installed on a trial basis in the Unit 3 reactor buildingof the Fukushima Daiichi NPS was not shared within the organization, and it was not fixed/restored
for a long period of time. Although there were no issues with work safety or exposure, etc., a seriesof events are receiving severe external criticism. At the Fukushima Daiichi NPS, the management
themselves are engaged in conversation with all site personnel to understand the organizational
issues behind the events. In order to establish safety as an organizational culture, not only a slogan,
it is not enough to just ask employees to prepare themselves, but each and every employee needs
educational materials and opportunities to learn about safety in a systematic way.
NDF has just started to prepare a safety evaluation guide for decommissioning of the Fukushima
Daiichi NPS (a booklet summarizing the peculiarities of the decommissioning of the Fukushima
Daiichi NPS, the concept of ensuring the safety of the decommissioning of the Fukushima Daiichi
NPS, safety evaluation methods, decision criteria, examples, etc.) with reference to the safety
evaluation guide of the U.S. DOE. NDF will consider how to roll out the guide to project activities

121

once it is developed, in coordination with the Office of Decommissioning Safety and Quality, and
the Project Management Office.
Ultimately, it is the people and organizations (operators) who handle all aspects of the"Fukushima Daiichi NPS site" (operation, maintenance, radiation control, instrumentation, analysis,etc.) that will realize safety. At Fukushima Daiichi NPS, the effectiveness and sufficiency of radiation
exposure reduction and personal safety measures are confirmed at "ALARA Meeting84" and "Safety
pre-evaluation Meeting 85 ". However, it is important that operators (including not only TEPCO
employees but also employees of subcontractors) who are familiar with the work site participate inthese meetings, and that the work process is designed so that safety can be comprehensively
checked from the "on-site perspective" (experience) of the various operators. In other words, it isimportant to establish the business process as a mechanism, such as incorporating into the gate
process the business process of "systematizing and manualizing experience and related
information from the field about procedures, rules, and precautions to ensure safety at work, andhaving the personnel and operators in charge of managing and operating the field work familiarize
themselves with them. It is important to establish such a system.
6.1.2.2 Owner's engineering capability
In the case of large scale projects, where the technology is mature and the performancerequirements are clear, such as the construction of nuclear power plants, waterfall-type 86
engineering is generally used. In this type of waterfall engineering, it is effective to clearlycommunicate the requirements of the client (TEPCO) to the potential contractors (manufacturers,
general contractors, etc.) in the supply chain as far as possible upstream of the project to proceedwith the project smoothly.
However, since the fuel debris retrieval is an inexperienced approach, the target setting andrequired specifications from TEPCO, the executor of the decommissioning project, are not always
clear at the time of engineering, and the degree of performance requirement setting, physicalfeasibility of the method and equipment, and performance assurance must be trial and error.
Therefore, in addition to “the sequential approach” that addresses “work in the initial stage” andthen rolls out the information gained to the next stage, the operation executor’s performancerequirements, the establishment of the supply chain functions, and its engineering process should
be iteration approach-based, to some extent87. In this regard, it would be effective to proceed by

84 The purpose of this meeting is to reduce the radiation exposure of workers and to discuss the effectiveness andsufficiency of engineering measures (physical measures) such as removal of radiation sources and installationof shielding.
85 The purpose of this meeting is to prevent the occupational accidents and to deliberate on the effectiveness andsufficiency of personal safety measures using risk management methods (risk identification, risk analysis andrisk assessment).
86 It is a method of increasing the accuracy of the design in stages, from conceptual study to basic design andthen to detailed design. Since the process proceeds in a single direction from upstream to downstream, it has
the advantage of easy control of budget, process and resources (personnel), but it also has the disadvantage of
not being able to go back once the next phase is underway. It is called the waterfall type, because it resembles
a waterfall where water falls from the top to the bottom.
87 A method of gradually increasing the percentage of completion of engineering by finding the next result basedon a certain result and repeating this cycle.

122referring to case examples in other industries that are also implementing the same iteration-type
engineering. For iteration-type engineering, the contract between a project executor and a supply
chain is not conventional88. Therefore, TEPCO, as a project executor, is strongly required to “makea judgment on engineering and is responsible for the results.” To do so, in addition to project
management capability, TEPCO needs to improve the capabilities that the project executor should
have, to optimize the entire supply chain, specifically the engineering ability that TEPCO, the project
executor, proactively performs as the owner89 (owner's engineering capability) such as the ability
to make engineering judgments, the ability to evaluate business risks, and the ability to specify
order specifications. For example, in the evaluation of fuel debris retrieval method selection, it isnecessary to evaluate from the viewpoints of quality (fuel debris retrieval status, safety, etc.), project(cost, time and other visions), and technical feasibility. Alternatively, a project that would previously
have been ordered by a single company can be divided among multiple companies, and TEPCO
can oversee and manage the entire construction. It is also important for the enhancement ofowner's engineering abilities to accumulate and feedback various field experiences, such as bythese activities and the promotion of insourcing to be described later.
Fuel debris retrieval is not a job like the design and construction of a nuclear power plant, where
the finished product is delivered with a performance guarantee. Therefore, unless TEPCO, theexecutor of the project, bears the technical and business risks at the end of the project, the cost
would be astronomically inflated. The fact that the project executor bears the technical risk alsomeans that the project executor itself must have the ability to assess the reliability of the functionalsettings and engineering design, which requires more technical skills. Here, the most important
point is to incorporate “safety and operator's perspectives” as upstream as possible in engineering.
(Fig.42)88 In the conventional construction of nuclear power stations, a supply chain has delivered completed products to

a project executor after guaranteeing the performance (Full turnkey contract).
89 An owner here has three positions of a party responsible for disaster occurrence, a specified nuclear facility

licensee, and a facility owner. TEPCO is executing the decommissioning project from these three positions. (Aproject executor of decommissioning)123Setting boundary
condition from "Safety
perspective and operators'◎ Operator Requirement Specifications〇 Performance requirements(Functional requirements if unable toindicate here)a) Safery : Criticality, exposure ・・・b) Operability : ・・・c) Easy of maintenance : ・・d) ・・・〇 Verifying perfomancea) Safery : Criticality, exposure ・・・b) Operability : ・・・c) Easy of maintenance : ・・d) ・・・〇 Quality assurancea) QA planb) Sub-vendor list (below secondary
contractorsc) Procurement control for subvendor (specially foroverseas vendors)

Elaborated boundary condition based on"Safety perspective and Operators'

Input：Design process of actual equipment：R&D Process

Conceptual review
Confirmation on how the "Safety

Output 1 perspective and operators'
perspectives“ are being

Actual machine

R&D◎ Conceptual review drawingsand docments• Performance requirements
confirmation document• Safety assessment document(Criticality, exposure, dust
dispersion, heat removal ...）

◎Specifying development
specifications• Identify technical issues andtechnical trend(Technical matuarity)①〇〇 Developed②△△ In the practival application
stage
・・・
• Method developed②・・TEPCO's voluntary PJ by 〇〇③・・Government-led PJ by 〇〇 ・Specification developed
Budget, schedule…

Route 1

• Conceptual
review

Route 2・Performancerequirements
revised・On-site

Output 2

No

All the performancerequirements
satisfied?
Yes

Elaborated boundary

Conceptual
review
redoneIncorporating R&D

Next step (basic design, etc.)
OutputRoute 3• The status of satisfaction can be left
as pending and verification on whether
they satisfy persormancerequirements may be performed at a

◎ Basic design drawing• System design
specifications• Machine design
specifications• Shielding calculation
sheet• P&ID•…

Confirmation of how the"Safety perspective andoperators'
perspectives" are being

Fig.42 Design flow of fuel debris retrieval (conceptual)
6.1.2.2.1 Ability to assess and manage process risks
It is fundamental to project management to identify and respond to project risks such as cost,
process, and safety, which have a significant impact on project execution, and to complete theproject within a certain margin of error by making full use of forecasting skills and know-how, while
giving top priority to safety and taking into account the constraints and uncertainties surrounding
the project. It is necessary to complete the project within a certain margin of error in terms of period
and cost. In particular, project risk management for work with high technical difficulty anduncertainty (e.g., fuel debris retrieval) will become even more important in the future.
As a precondition for this, it is essential for TEPCO employees to improve their skills. It is also
necessary to eliminate elementary errors such as process delays due to interference from multiple
construction work caused by a lack of consideration in the construction planning stage, or failure
to assess the progress of processes simply due to insufficient on-site checking.
In the system aspect, TEPCO has commenced operation of the process management system,
which enables assessment of the impact on important milestones in an objective way when any
delay in a project process becomes apparent. TEPCO has also implemented a process in which
each project evaluates the impact of changes, and an appropriate approver authorizes the changes
if any revision in milestones is required. However, these initiatives are still in the early stage. It isnecessary to further improve the ability to assess and manage process risks, and the operation of
the process management system will be further developed to enable detailed input processes,
linking of processes between projects and programs, critical path analysis, risk-based process124planning, and EVM90 (Earned Value Management). It is important to further develop the operationof the system to enable critical path analysis, risk-based process planning, and EVM (Earned ValueManagement). In addition, for systematic risk reduction, it is necessary to increase the sensitivityof each individual to identify risks, to understand the daily changes at the site, to understand thesocial situation, and to continue to improve technical capabilities. Given the large uncertainties
inherent in the projects, including rubble removal on Unit 1operating floor, zeolite sandbags
recovery, fuel debris retrieval, etc., are upcoming, the project management should be further
strengthened in order to be able to grasp the status of delays and assess the impact moreaccurately.
6.1.2.2.2 Acquisition management capability
The decommissioning of the Fukushima Daiichi NPS must be carried out reliably over a long
period of time under changing site conditions, and it will be difficult to cope with conventional
contracts, especially for high-risk project work such as fuel debris retrieval. Therefore, it isnecessary to prepare a contract method based on a new concept, in which both the contractor andthe recipient cooperate, share the contractual risks, and aim for the agreed-upon goals. In terms ofprocurement, instead of one-way Buying from an ordering party to an order-receiving party, bothparties should bear in mind the concept of Acquisition of the final result by Making, withconsideration of all steps from development, manufacturing, to even operation/maintenance (Table
3).
To deal with such a making type project, it is necessary to improve the owner's engineering
capability, such as the ability to concretize specifications, as well as to become familiar withacquisition management, which focuses on "acquisition.
Based on the shared awareness that conventional Buying-oriented project management alone
is not enough to properly control projects with large uncertainties such as fuel debris retrieval,
TEPCO and NDF are actively learning the method with the cooperation of external experts startingfrom the last fiscal year, and trying to proceed with a step-by-step approach the acquisition
management adopted by the U.S. federal government as a benchmark to further develop current
project.
Table 3 Difference between “Making” and “Buying”

What to call theorder receiving
parties and theirroles

MakingBuyingAcquisition of the outcome of the project（Acquisition）

Purchase products (things) that meet the
specifications
Vender, who supplies equipment that
conforms to specifications

Contractor, a partner who is responsible
for obtaining the outcome of the project90 It is a tool for analyzing the cost status of a project. It analyzes the current cost status and predicts how muchthe budget will be when the project is finally completed.

125

How to decide
who will receive
the order
Contract method

Select based on proposal content andfeasibility

Select by priceContracts in line with risk allocation
Data-based cost estimation (analogy:
analogy, integration, parametric :sensitivity, analysis, etc.)“Do the Right Things”* When doing things, always think about
what the right purpose, the right goal,
and the right means.

Fixed price contract
Quotation/price list, etc.

“Do Things Right”* Do things right, by the rules, by theprocedures.

6.1.2.2.3 Promotion of insourcing
In order to meet the many challenges of decommissioning, it is essential to strengthen
engineering capabilities, and TEPCO is promoting insourcing as a means to achieve this.
“Insourcing” means to develop an ability that enables TEPCO to implement planning, design,
maintenance, and operation on its own. It aims at reducing unreasonableness and waste, further
deepening the level of productivity improvement, and improving the operation quality of TEPCO
employees including the quality of design and procurement. The improvement of these qualities
will ultimately contribute to the improvement of safety. While insourcing is effective for a wide range
of issues, it is advisable to proceed with an awareness of the issues that are expected in the future
to gain more benefits. Examples of future issues are listed below.
・Improvement of operation and facility quality through internal integration of operation andmaintenance know-how
In terms of operation and maintenance of facilities, it is expected that the number of operation
tasks will increase due to the installation of new facilities and that maintenance costs will increase
due to the aging and deterioration of facilities. It is necessary to devise more "Smart Maintenance"
methods to keep maintenance costs low while not degrading reliability. For example, in the caseof a new system to be used for a long period of time, sensors (vibration, heat, sound, etc.) for
online monitoring of the operating status can be mounted in advance, a large amount ofaccumulated maintenance data be processed by computer to be used for prediction maintenance,and maintenance data obtained from existing systems be utilized. In this way, systematization andsophistication of maintenance knowledge management facilitate accumulation of maintenanceknowhow internally.
Furthermore, for long-term operations such as debris retrieval, where highly radioactive
materials (and nuclear fuel materials) are handled, TEPCO employees, who are the license
holders, should perform the operations themselves, and the operator should ensure radiation
safety and strive to internally accumulate operational know-how.

126

・Enhanced capability to make an engineering judgment
For a task like fuel debris retrieval, which is inherently subject to great uncertainty and for which
TEPCO has no previous experience, TEPCO needs to make engineering judgments on its own,
and to do so, it must have a firm grasp of the details of design, construction, maintenance, andoperation. Therefore, to take design as an example, TEPCO should improve its competence byworking on system design and determining the required specifications of equipment from thecurrent stage.

6.1.2.3 Securing and developing human resources
6.1.2.3.1 Securing and developing human resources for smooth implementation of
decommissioning projects(1) Securing and developing human resources based on the medium-and-long-term humanresource development plan
The development of human resources is essential as a basis for the smooth implementation of
long-term decommissioning projects. For this purpose, project management and engineering
capabilities should be enhanced. In addition, it is important to assume occupational categories, thenumber of engineers and the time required in the future (design, operation, maintenance, chemical
analysis, safety assessment, radiation control, etc.) in light of the Medium-and-Long-term
Decommissioning Action Plan, to summarize them as the medium-to-long term human resources
development plan, and to promote human resources development and securing of staff
systematically.
As an initiative to enhance project management capabilities, TEPCO has given responsibility andauthority to project managers so that they can devote themselves to project management. In
addition to basic education for project management, TEPCO is expanding the efforts such as to
newly introduce training program for systematic learning in accordance with international standards.
As an initiative to enhance engineering capabilities, while collaborating with other electric powercompanies, as well as manufacturers, general contractors and engineering companies, TEPCO
has been making use of the knowledge of external experts including overseas experts, andaccumulating and transferring their technical capabilities and know-how including engineering
capabilities. Based on the medium-and long-term decommissioning action plan, measures arebeing considered to develop human resources and to secure personnel, assuming the number ofengineers and the time to be required in the future.
(2) Secure and develop human resources with a "safety and operator’s perspectives
It is necessary to secure and train human resources with "safety and operator’s perspectives"
systematically and promptly.
TEPCO recognized early on that it is important to secure and train human resources who arewell versed in the field and have acquired sufficient knowledge of the field to have an "operator’s
perspective. The “core technologies” required for the decommissioning of the Fukushima Daiichi
NPS have been established, and the “Decommissioning Core Technology Course” has been held127

to foster them. Since 2015, the company has been implementing initiatives to improve on-site
capabilities with in-house veteran instructors. In addition to the above general familiarity with the
Fukushima Daiichi NPS decommissioning site and acquisition of on-site knowledge, in case where
special skills specific to the facility or work to be handled are required, it is necessary to secure andtrain personnel with such special skills individually. For example, in the case of remote handling of
unsealed radioactive materials such as fuel debris retrieval, which has never been experienced in
the operation of power plants, it is necessary to consider securing and training personnel with suchspecial skills individually through operation at mock-up facilities in Japan and overseas, and actual
work in removal of radioactive materials on a gradually expanding scale. It is necessary to consider
how to secure and train them individually. With regard to human resource development for fueldebris retrieval operations, TEPCO has started efforts to learn about remote operation systems andtraining methods overseas (RACE, UK). It is also essential to secure and develop human resources
capable of planning the necessary analyses according to analysis needs and also equipped withanalytical techniques (including handling of samples) for analysis work associated with fuel debris
retrieval. Given the prolonged decommissioning work, it is recommended to utilize local humanresources. Assuming an analysis system by function of the facility for analysis, TEPCO is planningto secure techniques by accepting outside engineers, transfer and send their personnel to outside
analysis institutions, and systematically allocate chemical and material careers and new employees,
and is starting some of these initiatives.
It is the personnel (operator) or organization engaged in the entire site of the Fukushima Daiichi
NPS (operation, maintenance, radiation control, instrumentation, analysis, etc.) that ultimately
achieves the perspective of both “operator” and “safety” on site. For this reason, it is desirable that
personnel with an "operator’s perspective" simultaneously acquire a "safety perspective" such as
criticality management and confinement of alpha-nuclides. To this end, there is an urgent need todevelop human resources who have the following qualities and who can play a central role inexpanding the "safety perspective".
① To know the basic concept to achieve reactor safety.
② To know the general methods to achieve the basic concept③ To know how the basic concept and general methods of safety are applied and realized in
actual design.
④ To have a field-oriented viewpoint of safety assurance.
⑤ To have basic technical knowledge about handling of unsealed radioactive materials(including alpha nuclides), criticality, and measurement (detection) methods. As evidence ofsuch knowledge, the person should have a qualification such as a chief nuclear fuelhandling officer (or a qualification established internally as equivalent).
⑥ The person should have knowledge of international precedents.
⑦ The person should have knowledge and experience in negotiating with the NRA.

128

6.1.2.3.2 Fostering the next generation who will be responsible for the futuredecommissioning of Fukushima Daiichi NPS
In order to continue decommissioning of the Fukushima Daiichi NPS for a long period of time
and to continue R&D activities necessary for that, it is important to secure solid technicalcapabilities on a constant basis. For that purpose, it is essential to train and secure future
researchers and engineers, and ensure technical transfer.
Primarily, the assumption is that TEPCO will establish the decommissioning of the Fukushima
Daiichi NPS as a planned and sustainable project, under which activities of human resource
development and acquisition, including their employees and contractors, will be promoted
independently and continuously. Human resources involved in decommissioning of the Fukushima
Daiichi NPS are required not only to specialize in nuclear energy but also to possess science andtechnical knowledge in fields other than nuclear energy. TEPCO is required to secure and develop
human resources with such diverse technical backgrounds in the course of decommissioning.
For this reason, in addition to the importance of strongly encouraging human resource
development and recruitment directly within TEPCO and contractors, it is most important thatexcellent human resources who have graduated from universities, graduate schools, technicalcolleges, high schools, etc., and have specialized in science and technology are continuously
sourced to these companies. In order to achieve this in a stable manner, it is necessary for higher
and secondary educational institutions to create an opportunity for learning and acquiring
peripheral knowledge in addition to professional knowledge, and to maintain associated systems
and structures so that they can function as a whole, including teachers.
With this basic understanding, it is important for industrial-academic-governmental institutions,as a whole, to steadily promote efforts according to each level of higher and secondary education.
For students in higher education such as universities, graduate schools, and technical colleges,
it is important for the industry and higher education institutions to cooperate and continuously
implement activities to promote understanding of the nuclear industry. In recent years, due to thereorganization of faculties and departments related to nuclear energy, it has been pointed out thatthe human resource development function in higher education institutions in the nuclear field is
weakening. Therefore, attention must be paid to maintaining and enhancing the human resource
development function in the nuclear field nationwide. In particular, it is necessary to communicate
the idea that the decommissioning of the Fukushima Daiichi NPS is an extremely advanced
technical challenge, which is unprecedented in the world, and to establish various career paths foryoung researchers and engineers to thrive in the nuclear industry including decommissioning, andto specifically demonstrate and make them feel a career view. For the development of nextgeneration human resources, it is fundamentally important that young researchers/engineers areconstantly produced from these higher education institutions.
In particular, the World Intelligence Project by MEXT and JAEA/CLADS has introduced a system
in which students and young researchers are made to be aware of decommissioning as animportant research area, and is engaged in decommissioning research. From the perspective ofhuman resource development, support has been provided for young researchers and teachers in129

preparing and implementing lecture curriculums related to decommissioning. After more than five
years since the launch of the World Intelligence Project, it has produced great results in terms ofboth research and human resource development. At “the Conference for R&D Initiative on Nuclear
Decommissioning Technology by the Next Generation (NDEC)”, a conference for students topresent their research findings as part of the World Intelligence Project, and the Creative Robot
Contest for Decommissioning for technical college students, students present their research results,
exchange views with researchers and engineers involved in the decommissioning of the Fukushima
Daiichi NPS, and are given awards for excellent performance on a constant basis. As for an
initiative for young researchers in the World Intelligence Project, in addition to the existing program
represented by young researchers under the age of 39, a new program was launched in FY 2021
that requires young researchers to assume commensurate research responsibilities within R&D
projects as a requirement for application. In order to encourage autonomous research activities, an
initiative has been taken since April 2021 where young researchers engaged in the WorldIntelligence Project are allowed to be involved in independent research activities up to 20% of theefforts devoted to the project with the consent of the research representatives (relaxation of theobligation of full-time commitment).
It can be said that the mechanism of and implementation by the World Intelligence Project haveproduced some results for researchers and students in higher education institutions. Hereafter, it
is important to implement this project so that the perspectives of decommissioning sites in TEPCO
and those of the activities in higher education institutions can be more aligned.
For junior and high school students in the stage of secondary education, it is important tointroduce appealing points of engaging in the nuclear energy field including decommissioning, andto make efforts to attract their technical interests with a focus on decommissioning, as well as toincrease their understanding of decommissioning and reconstruction of the Fukushima Daiichi NPS,
and in a broad sense, of the career path in science and technology fields. The secondary education
stage is a stage in which, as a preparatory stage for participating in and contributing to society,students can develop their individuality and competence while being influenced by researchers,
engineers, and science teachers who are active in society, and can make independent choices anddecide their career paths. From this perspective, NDF holds the “International Mentoring Workshop
Joshikai in Fukushima” for middle and high school female students mainly in Fukushima Prefecture
in cooperation with the Organization for Economic Cooperation and Development and the Nuclear
Energy Agency (hereinafter referred to as “OECD/NEA”) as an effort to increase interest in scienceand engineering, especially in decommissioning, etc., through exchanges with female researchersand engineers in order to enhance understanding among women and help increase their motivation
to participate in development studies. In 2020, the "International Mentoring Workshop Virtual
Joshikai in Fukushima 2020" was held using the online system, even in the midst of the global
outbreak of the new coronavirus infection, and will continue to be held using the online system in
2021. In addition, “student sessions” are held for high school students, etc., to give thought to thereconstruction of Fukushima along with the International Forum on the Decommissioning of the130

Fukushima Daiichi NPS (hereinafter referred to as “International Forum”). Through these efforts,high school students, etc., are given an opportunity to think about activities to achieve both
decommissioning of the Fukushima Daiichi NPS and reconstruction, enabling them to increase
their awareness that decommissioning is an important issue in reconstruction of local communities,
and foster interest in and willingness to contribute to decommissioning and reconstruction efforts.
Such activities have achieved some positive results. These sessions and forums will continue tobe held in 2021 by utilizing an online system.
For the development of such technical personnel associated with decommissioning of the
Fukushima Daiichi NPS, it is also necessary to expand into wider fields including fundamental
research and related research. In the course of raising the level of the entire fundamental
technological base in Japan, it is expected that initiatives to deal with the nuclear legacy andnuclear safety will take deeper root.
Institutions concerned are continuously required to promote and strengthen their efforts to secureand develop human resources for the next generation according to their respective roles and levels.
6.1.2.3.3 Dissemination of basic knowledge and promoting the people’s understanding fordecommissioning and radiation safety involved in decommissioning
It is important for many citizens and local residents to acquire basic knowledge of the accident
and decommissioning, disaster response, radiation safety, food safety, etc., related to the
Fukushima Daiichi NPS from the perspective of future resilience of the whole country. This isbecause it will serve as basis for discussions on decommissioning, and related radiation safety,
etc., based on accurate information and for promoting public understanding. In addition, although
it is not directly aimed at fostering human resources who will play a leading role in the nuclear fieldin the next-generation, it is also an aspect of indirectly broadening the range of human resources
who are interested in not only nuclear energy but also science in general. Particularly in the fieldof nuclear energy, it is necessary to learn about the relationships in local communities and society
through various opportunities according to the developmental stage of children, as well as toacquire knowledge/experience on nuclear energy and decommissioning. In doing so, since it isimportant that children take an interest through the knowledge and experiences of adults aroundthem, such as teachers and parents. Therefore, it is important to further spread knowledge on
nuclear energy and decommissioning based on scientific evidence, which to a wide range of people
including those involved in primary education institutions.

6.2 Strengthening international cooperation
6.2.1 Significance and the current status of international cooperation
In recent years, nuclear reactors and nuclear fuel cycle-related facilities built at the dawn of theuse of nuclear energy have reached the end of their operational life, and decommissioning of thesefacilities is in full swing in many countries. Among the reactors that have experienced severe
accidents are the Windscale Pile-1 reactor in the UK, the Three Mile Island Unit 2 reactor (TMI-2)
in the US, and the Chernobyl Unit 4 reactor (ChNPP-4) in Ukraine. These facilities have been131

undergoing stabilization work and safety measures for many years. In addition, there are large
uncertainties in the management of a wide variety of radioactive materials at legacy sites overseas,
and decommissioning and environmental remediation efforts are expected to take a long time. Eachcountry continues to face challenges such as "unknown unknowns," long-term project management,
and securing large amounts of funding.
In order to steadily proceed with the decommissioning of the Fukushima Daiichi NPS, which
deals with difficult engineering issues, it is important to learn lessons from precedentdecommissioning activities, etc. and apply them to the decommissioning of the plant as a risk
reduction strategy, and to utilize the world's highest level of technology and human resources, i.e.,
to gather and utilize the wisdom of the world.
In addition, the decommissioning of the Fukushima Daiichi NPS is a process to solve unexplored
engineering problems by combining knowledge from various fields, not limited to the nuclear field,
and it can be interpreted that the decommissioning of the Fukushima Daiichi NPS can be a powerfulplace to create innovation. The concentration of diverse knowledge and experience from aroundthe world in Fukushima is, in the first instance, an important effort to steadily advance thedecommissioning of the Fukushima Daiichi NPS itself. This is also an important initiative from theperspective of building a symbiotic relationship with the local community, which is essential for the
long-term progress of decommissioning.
To bring together the wisdom of the world, it is important to maintain and develop the international
community's continuous understanding, interest, and cooperation in decommissioning. Therefore,
it is important to gain the confidence of the international community by disseminating accurate
information on the progress of decommissioning, etc., and to promote decommissioning in amutually beneficial manner that is open to the international community by actively and strategically
returning to the international community the knowledge, etc., gained through the accident at the
Fukushima Daiichi NPS and decommissioning.
Specifically, it is important to promote bilateral cooperation in line with the circumstances of eachcountry and to utilize the framework of multilateral cooperation through the IAEA and OECD/NEA.
These international organizations have an important role in establishing international standards fordecommissioning. It is important to participate in the formulation of international standards basedon Japan's experience in decommissioning to promote the decommissioning of the Fukushima
Daiichi NPS in an internationally open manner, and it is also expected to fulfill part of Japan's
responsibility to the international community by sharing Japan's experience with other countries.
From this perspective, Japan has been holding an annual dialogue and establishing a conference
body to share information with other countries as an intergovernmental framework. NDF has beenworking on disseminating information on decommissioning through participation in side events of
the IAEA General Conference and speaking at major international conferences such as the
OECD/NEA Steering Committee. By securing the confidence of the international community andpromoting mutually beneficial decommissioning, we are trying to maintain and develop the132international community's continuous understanding and interest as well as cooperative
relationships. (Attachment 16)
In addition, although the global outbreak of new coronavirus infections is a major obstacle to theabove-mentioned international cooperation, many meetings and events such as the IAEA General
Conference and its side events, OECD/NEA, etc. are held online, and Japan has participated byutilizing the online system. NDF has also been actively utilizing the online system, etc., to gather
the world's wisdom, maintain and enhance the international community's continuous understanding
and interest, and maintain and develop cooperative relationships with international society.
Specifically, opportunities for continuous exchange of information have been secured, for example,
by conducting online exchanges of opinions and meetings with overseas experts invited to NDF
every year, and by holding online annual meetings with decommissioning-related organizations inother countries. For example, in a short period of two weeks from May 17 to May 28, 2021, bilateralmeetings between several domestic organizations concerned and the OECD/NEA were realized.
In the future, it is important to further expand the opportunities for communication with othercountries by taking advantage of the experience gained so far. Fig.43）

Fig.43 Annual meeting of NDF with foreign organizations concerned (held online in April 2021)
6.2.2 Key issues and strategies
6.2.2.1 Integrating and utilizing wisdom and knowledge from around the world
The decommissioning of Fukushima Daiichi NPS is expected to take a long time, and thedecommissioning of legacy sites has many points of reference in terms of technology and operation
as a leading model. In each country that has a legacy site, public organizations related todecommissioning are playing a central role in promoting decommissioning to cope with issues suchas the necessity of new technologies and expertise that differ from the operation and maintenanceof nuclear reactors and nuclear fuel cycle facilities. Therefore, it is necessary for Japan to conclude
cooperative agreements and arrangements among domestic and overseas organizations under theintergovernmental framework, and to continue to implement regular information exchange. NDF
will continue to maintain and develop long-term partnerships with public decommissioning
organizations that play a central role in each country, such as NDA in the UK, CEA in France, and
DOE in the US.
133With regard to the decommissioning of the Fukushima Daiichi NPS, Japan needs to learn fromthe wisdom of the world not only in terms of technology but also in terms of operations, including
systems and policies, strategy formulation, project planning and management, safety assurance,and regional communication. To this end, as we move forward with decommissioning, Japan hasreceived support from the international community and has received various kinds of assistancefrom foreign governmental organizations and experts, through the dissemination of information on
issues related to decommissioning to the international community and participation in international
joint activities such as the DAROD project by the IAEA and the joint project by the OECD/NEA.
More than ten years have passed since the accident, and it is necessary to continue the mutually
beneficial relationship while also working to return the know-how and results accumulated so far to
the internal community.
The difficulty in traveling to and from other countries due to the pandemic outbreak of newcoronavirus infections has become an obstacle to the continuation of such mutually beneficial
relationships. It is important to secure opportunities for communication and work to maintain anddevelop relationships by utilizing online systems and other means so that the unprecedented
situation, not limited to the new coronavirus infection, does not dilute relationships with relevantorganizations, experts, and international organizations in other countries.
Through a series of the International Forums, NDF has collected the wisdom including lessons
learned and technologies derived from decommissioning around the world. It is important tocontinue to make the International Forum an effective opportunity to gather wisdom from all over
the world by using the online system, even in the situation where direct participation from othercountries is difficult.
The decommissioning of nuclear power plants, both in Japan and abroad, is being carried out
under contracts between many companies and decommissioning executors, and the global market
for decommissioning is expanding greatly. As the engineering of Fukushima Daiichi NPS is in fullswing, it is important to grasp the latest status of excellent technologies and human resources in
the world and to utilize them effectively. In this context, TEPCO has been actively engaged intechnological exchanges with overseas private companies. It is necessary for TEPCO to continue
to keep abreast of the latest information from around the world, including the situation in the private
sector, and to engage in continuous communication with these private companies, sharing
information on the progress of decommissioning work and forming an environment in which thenecessary technologies can be accessed when needed.
6.2.2.2 Maintaining and developing the international community's understanding of andinterest in decommissioning and cooperative relationships
In order to mobilize the wisdom of the world for the decommissioning of the Fukushima Daiichi
NPS, it is important to maintain and develop the understanding, interest, and cooperative
relationship of the international community. To this end, it is important for the government and otherdomestic organizations to disseminate accurate information on the decommissioning of the
Fukushima Daiichi NPS, considering the fact that more than 10 years have passed since the134accident and the interests of the recipients of the information have changed since the time of theaccident. Specifically, for experts, it is necessary to provide information not only on the progress of
decommissioning but also on the lessons learned from the decommissioning work. In addition,
consideration should be given to providing easy-to-understand information not only for experts but
also for non-experts, and to adding appropriate devices that consider the level of understanding of
the recipients regarding the background of the accident. In addition, it is also important to participate
in international joint activities and disseminate information to return the knowledge obtained in thecourse of decommissioning to the international community. In participating in international joint
activities, it is necessary to work on the premise that decommissioning, which is Japan’s top priority,
will be steady implemented, and the interests of the international community can also be secured.
From the aspect of returning the results, it is important to maintain the level of interest while
responding to the changes in the international community, such as the growing interest in not only
the accident and decommissioning itself but also the application to other issues.
On the premise of ensuring safety and taking thorough measures against reputational damages,some countries, in fact, issued comments expressing concern about the impact on the environmentand questioning the transparency in disposal in response to the Japanese government's
announcement of a policy to discharge ALPS-treated water into the ocean. On the other hand,foreign governments, related organizations, and international organizations that understand thesituation of Japan's decommissioning have issued comments supporting Japan's decision. In
addition, the IAEA announced that it would actively cooperate with the discharge of ALPS-treatedwater into the ocean from a third-party standpoint by dispatching a review mission and supporting
environmental monitoring. These actions are pushing to gain international understanding ofoffshore discharge, and the importance of ensuring transparency and building cooperative
relationships through accurate information dissemination was reaffirmed.
It is important to continue to obtain the understanding and cooperation of the international
community for the steady implementation of decommissioning, including the future discharge of
ALPS-treated water into the ocean, and it is necessary for the government and other relevant
domestic organizations to disseminate accurate and easy-to-understand information atinternational conferences, international joint activities, and on websites. It is also important for NDF,
through various opportunities, to disseminate information that is accurate meets the changing
interests of the recipients, and to return to the international community the knowledge obtained
through the decommissioning process.

1356.3 Local community engagement
6.3.1 Significance and the current status of local community engagement
6.3.1.1 Basic concept
The fundamental principle for the decommissioning of the Fukushima Daiichi NPS is "Balancing
between reconstruction and decommissioning”. In the areas where the evacuation order has been
lifted, progress toward reconstruction is gradually being made, not only by the return of residents
and the resumption of business activities, but also by the promotion of migration and settlementfrom outside the area and new investment. While giving top priority to further reducing risks to thesurrounding environment and ensuring safety. It is necessary to strengthen communication andpromote coexistence with local communities to gain the trust of the community. Decommissioning
should not be allowed to have a negative impact on the reconstruction process due to anxiety anddistrust of decommissioning, in other words, decommissioning should never be a hindrance toreconstruction efforts.
Therefore, it is important to deepen the understanding of local residents and reassure them about
the decommissioning through interactive communication: not one-way dissemination of information,
but sincere listening to the concerns and questions of local residents and promptly providing themwith accurate information in an easy-to-understand manner to eliminate them. In addition, toaccomplish the decommissioning over a very long period of time, the continuous cooperation ofcompanies, especially local companies, is essential. At the same time, the participation of local
companies in the decommissioning project is an important pillar of TEPCO's contribution to thereconstruction of Fukushima, as it will not only revitalize decommissioning-related industries in theregion and create employment and technology, but also lead to the spread of the results to otherregions and industries. In light of this, TEPCO will contribute to job creation, human resource
development, and the creation of industrial and economic infrastructure in the region throughdecommissioning, while collaborating with efforts to realize the "Fukushima Innovation Coast
Framework Promotion Organization", which sets the accumulation of decommissioning-relatedindustries in the Hamadori region as a priority field, and aims to achieve "Balancing between
reconstruction and decommissioning”.
6.3.1.2 Specific measures under the current situation(1) Communication initiatives
The government has been exchanging opinions with local related organizations at the"Fukushima Advisory Board on Decommissioning, Contaminated Water and Treated Water" andother meetings held by the government, disseminating information on the current status ofdecommissioning through videos, websites, brochures, etc., and holding briefings and roundtablediscussions for local residents and related local governments.
NDF is holding the International Forums for the purpose of frank exchange of opinions on
decommissioning with participants including local communities and organizations concerned, andfor sharing of the latest knowledge, technical achievements, and issues on decommissioning with136

experts in Japan and overseas (the Forum was postponed in FY 2020 due to the COVID-19
infection). They also hold briefings on the progress of decommissioning at meetings hosted by thenational and local governments. In order to promote the exchange of opinions at the International
Forums, a “hearing activity” is held every year to hold a conversation with local communities
including high school and technical college students before the International Forum is held. Then,
their real voice is collected, summarized and edited as a booklet, and distributed as the “Voicefrom Fukushima” at the International Forums.
TEPCO has been making efforts to provide explanations and dialogue to regional
representatives at conferences hosted by the government and Fukushima Prefecture, as well as
to hold regular press conferences and lectures for the media, and to disseminate information
through its website and brochures. In addition, the company is actively accepting visitors to the
Fukushima Daiichi NPS and holding roundtable discussions, as it is very effective in forming acommon understanding to have people see the current status of decommissioning as it is and to
exchange frank opinions (number of visitors: 18,238 in FY2019, 4,322 in FY2020). In addition, the"TEPCO Decommissioning Archive Center" established in Tomioka Town as a place where people
can learn about the process of the nuclear power plant accident and the progress of
decommissioning has about 70,000 visitors as of the end of April 2021, and since last fiscal year,
the center has been collaborating with the "The Great East Japan Earthquake and Nuclear
Disaster Memorial Museum" opened by Fukushima Prefecture in Futaba Town.
In addition, a virtual tour of the decommissioning site of the Fukushima Daiichi NPS has been
available on TEPCO's website since 2018. In the current situation where direct observation islimited due to the new coronavirus infection, such simulated experience programs are more usefulthan ever.
(2) Approach to create regional industrial and economic infrastructure through decommissioning
Based on their "commitment to the people of Fukushima to achieve both reconstruction anddecommissioning” established in March end, 2020 (hereinafter referred to as “Commitment”),
TEPCO has summarized their efforts for the accumulation of decommissioning work into thefollowing 3 categories: (1) Increased participation of local enterprises, (2) Support for local
enterprises to step up and (3) Creation of new local industries, and has started to implement themin a phased manner. With regard to (1) and (2), in cooperation with the Fukushima Innovation
Coast Framework Promotion Organization and the Fukushima Soso Reconstruction Promotion
Organization, TEPCO has set up and are operating a joint consultation service to support
matching between local companies interested in participating in decommissioning projects andprime contractors who are considering placing orders with local companies. In addition, it is also
conducting a survey of the needs of both prime contractors and local companies regarding humanresource development, and has started joint research with several universities. Moreover, thecontents of the "Medium- to-Long-Term Procurement Outlook in the Decommissioning" preparedin September 2020 are being updated as necessary to reflect the progress of decommissioning
work, and briefing sessions are being held not only for prime contractors, but for local governments,137

commercial and industrial organizations, and local contractors, paying close attention to the
spread of the new coronavirus infection.
In addition, with regard to (3), in order to build an integrated decommissioning project
implementation system locally, from "development and design" to "manufacturing," "operation,""storage," and "recycling," TEPCO is planning to establish and operate several new facilities in
the 2020s, so that technologies and products of relatively high difficulty and importance, whichhave been ordered outside Fukushima Prefecture, including overseas, can be completed in the
Hama-dori region (announced on May 27, 2021). In particular, with regard to local manufacturing,
TEPCO has set-up a joint venture with partner companies that have a proven track record in thefield of high-function products that had to be manufactured outside of the prefecture, the plan is
to set up a manufacturing base in the Hama-dori region with the aim of creating local employment,
placing orders with local companies, and promoting collaboration.
At the same time, in order to steadily promote initiatives for local community engagement,
TEPCO has reformed their organization as needed. The local partnership promotion group wasestablished in the Fukushima Daiichi Decontamination & Decommissioning Engineering Company
in April 2020, TEPCO set up the specialized department working to engage with local community
at the Fukushima Daiichi NPS in October 2020, and the Hama-dori decommissioning industry
project office that directly reports to the president was also established. Based on the division ofroles, they are engaged in internal/external coordination, field response to local communities andconsideration of medium-and-long-term direction.
6.3.2 Key issues and strategies(1) Communication issues and strategies
Misunderstandings, concerns, and rumors caused by the inappropriate dissemination of
information on the decommissioning of the nuclear power plant will lead to a loss of reputation andtrust in the decommissioning of the nuclear power plant not only in the local community but also insociety as a whole, which will not only delay the decommissioning of the nuclear power plant but
also hinder the reconstruction of Fukushima. For this reason, TEPCO needs to continue to takevarious measures to promptly communicate the current status of decommissioning in an easy-tounderstand manner. In this regard, while the impact of the new coronavirus infection are expected
to continue for the foreseeable future, TEPCO will make active use of tools such as virtual tour
programs and online conference systems, and it is also important to strengthen communicationthat is possible even in non-face-to-face and non-contact situations, such as by further enhancing
photo and video content.
In addition, the government, NDF, and TEPCO must work to build trust with local communities
by providing information more carefully under appropriate coordination. Therefore, capturing
opportunities to hold round-table talks and join local meetings/events, it is necessary to have directinteraction with local communities. Efforts should also be made for two-way communication by
conversation, including listening to their concerns and questions carefully through events such as138

International Forums, and to deliver accurate information in an easy-to-understand and carefulmanner.
In particular, the disposal policy of the ALPS-treated water has been the subject of anxiety andconcern not only from locally but also domestically and internationally, and local governments andrelated organizations are strongly urged to provide accurate information and take all possible
measures to prevent rumors. In light of these circumstances, TEPCO must do its utmost to
suppress rumors by steadily implementing the measures outlined in the "TEPCO's response
based on the government's basic policy on the disposal of treated-water by multi-nuclear removal
facilities" (released on April 16, 2021), and by making additions and revisions as necessary.
(2) Issues and strategies related to the creation of regional industrial and economic infrastructure
through decommissioning
As mentioned in 6.3.1.2(2), TEPCO is making various efforts to realize the "Commitment," butthese efforts will not produce visible results immediately and will require a certain period of time.
In addition, for “(3) Creation of new local industries” which will be promoted in the future, it is arelatively large-scale investment and is expected to have a great economic effect on the Hamadoriarea. However, as advanced techniques are required to produce high-performance products, theissue is how to eliminate the technical gap between experienced prime contractors and local
communities and connect this to promoting active participation in local companies. Therefore, forthe time being, it is important to continue and strengthen the current activities in a credible manner,
including "(1) Increased participation of local enterprises" and "(2) Support for local enterprises tostep up". It is also important to carefully explain to local governments, commercial and industrialorganizations, and other organizations concerned the location and scale of new decommissioningrelated facilities, the schedule from construction to operation, and the status of considering
engagement with local communities in terms of employment, cooperation and order placement,
and to proceed with the activities while gaining understanding and cooperation.
Also, with the understanding of prime contractors, it is necessary to consider specific methodsof ordering and contracting that will make it easier for local companies to receive orders, and to
implement these methods on a trial basis. As a result of interviews conducted with local companies
last year, it became clear that local companies do not necessarily want to be the main contractor,
but tend to want to enter the market as a subcontractor to gain technology and experience. Afterproperly understanding the intentions and needs of these local companies, a scheme can beestablished to benefit both parties by not only approaching local companies, but also encouraging
existing prime contractors to place orders with local companies, including technical guidance. This
will contribute to the promotion of orders from local companies by adopting methods that are
beneficial to both parties. At the same time, with regard to human resource development, theFukushima Decommissioning Engineer Training Center of the Fukushima Nuclear Energy
Suppliers Council, which was established in 2018 and has been providing education of radiation
protection and special education on specific matters such as low-voltage electricity handling,should be used to provide the training. In parallel, specific studies and preparations for training139

specifically for local companies should be accelerated. It is important to steadily promote thesevarious efforts while responding to changes in the situation as appropriate, and to build afoundation for local industry and economy through the decommissioning project and to develop
local companies and human resources.
In addition to research and development related to decommissioning, as companies fromoutside the region move into the region and provide technical guidance to local companies, thenumber of engineers and researchers visiting and staying in the region is expected to increase.
Therefore, it is necessary to establish the necessary environment and support system so that suchexternal personnel can integrate into the local community and play an active role as a member ofit. In particular, it is necessary to take into consideration a wide range of functions such as daily
life and education so that not only single people but also families can live together with peace ofmind.
In this regard, in order not only to promote the return of residents to their homes, but toaccelerate the reconstruction of the evacuated areas by encouraging migration and settlement in
the wider area, Fukushima Prefecture has established the “Fukushima 12 Municipalities Migration
Support Center”, and assists migration and settlement of people mainly from outside theprefecture to the 12 municipalities, disseminating information to people nationwide who areinterested in migration and providing various types of support for those who wish to move to the
12 municipalities. It is important to consider the possibility of collaboration and cooperation withthese local initiatives.
To steadily promote these efforts for coexistence with local community, it is essential tostrengthen the organizational structure within TEPCO and to have close cooperation between
each department. As mentioned in Section 6.3.1.2(2), TEPCO has been reorganizing itself to set
up specialized departments for regional symbiosis, and efforts to promote local industries throughdecommissioning are gradually moving forward, and gaining a certain level of recognition fromthe local community. It is important to keep this trend going steadily, while further strengthening
the internal structure as necessary.
In addition, it is necessary to further strengthen cooperation and collaboration with local
governments, including Fukushima Prefecture, and local related organizations, including the
Fukushima Innovation Coast Framework Promotion Organization and the Fukushima Soso
Recovery Promotion Organization, which are operating a joint consultation service and co-hosting
matching business meetings. NDF will provide appropriate support to TEPCO's efforts for regional
symbiosis, and will strive to strengthen cooperation and collaboration with local governments andrelated organizations.

140

List of Acronyms/Glossaries
Acronym

Official NameALARPAs Low As Reasonably Practicable：Risk should be reduced as far
as reasonably practicable including risk/benefit criteria or cost while
taking feasibility of risk reduction measures into account.

ALARA

As Low As Reasonably Achievable：The principle of radiologicalprotection in which it advocates that all radiation exposure must be
maintained as low as reasonably achievable in consideration of socialand economic factors.

D/W

Dry Well

DOE

United States Department of Energy

FP

Fission Products

IAEA

International Atomic Energy Agency

ICRP

International Commission on Radiological Protection

IRID

International Research Institute for Nuclear Decommissioning

JAEA

Japan Atomic Energy Agency

JAEA/CLADS

JAEA Collaborative Laboratories for Advanced Decommissioning
Science

NDA

Nuclear Decommissioning Authority

NDC

Nuclear Development Corporation

NDF

Nuclear Damage Compensation and Decommissioning Facilitation
Corporation

NFD

Nippon Nuclear Fuel Development Co., Ltd

OECD/NEAOECD Nuclear Energy Agency

PCV

Primary Containment Vessel

RPV

Reactor Pressure VesselSA

Severe Accident

S/C

Suppression Chamber

SED

Safety and Environmental Detriment

SGTS

Standby Gas Treatment System

TMI-2

Three Mile Island Nuclear Power Plant Unit 2

Penetration X-2

Penetration X-2 of PCV

Penetration X-6

Penetration X-6 of PCVCenter of the WorldIntelligence project

The project that promotes nuclear science and technology andhuman resource development gathering wisdom and knowledge

Operating Floor

Operating Floor of the buildings

Commitment

The commitment to the people of Fukushima for achieving bothreconstruction and decommissioning

Kashiwazaki-Kariwa

Kashiwazaki-Kariwa Nuclear Power Station

141

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

Prospects of processing/disposal method and technology related toits safety

International Forum

International Forum on the Decommissioning of the Fukushima
Daiichi Nuclear Power Station

Submersible ROV

A remotely operated submersible survey vehicle (Remotely Operated
Vehicle)Mid-and-Long-term
Roadmap

Government-developed “Mid-and-long-term Roadmap” toward thedecommissioning of TEPCO’s Fukushima Daiichi Nuclear Power
Station Units 1 to 4

TEPCO

Tokyo Electric Company Holdings, Inc.

Withdrawal Plan

Withdrawal plan for reserve fund

The Policy of
Preparation ofWithdrawal Plan

The Policy of preparation of withdrawal plan for reserve fund fordecommissioning

Fukushima Daiichi NPS

Fukushima Daiichi Nuclear Power Station of Tokyo Electric Company
Holdings, Inc.

Fuel removal from SFP

Fuel removal from spent fuel pool

Glossary

Description

Inventory

Amount of radioactive material contained in the risk source (radioactivity,
concentration of radioactive material, or toxicity possessed by theradioactive material)

Well plug(Shield plug)

A top cover to screen upper part of Primary Containment Vessel made ofconcrete (It is the floor face of the top floor of reactor building inoperation)

Engineering

Design and other work to apply technical elements to the site

Cask

Special container used for transporting and storing spent fuelSubdrain

Wells near the building

Sludge generated at
decontamination
device (waste
sludge)Sludge containing high level of radioactive material generated at thedecontamination device (AREVA), which was operated for contaminatedwater treatment from June to September 2011

Spray curtain

Watering to contain dust and allow it to settle

SludgeMuddy substance, dirty mud

SlurryA mix of dirty mud and mineral, etc. in water

ZeoliteSorbent used to recover radioactive materials such as cesium

Torus room

A room that houses a large donut-shaped suppression chamber that
holds water for emergency core cooling system.

Fuel debris

Nuclear fuel material molten and mixed with a part of structure inside
reactor and re-solidified due to loss of reactor coolant accident condition

Bioassay

A method for evaluating the types and amounts of radionuclides ingested
into body by analyzing samples from the human body, such as
excrement142

Glossary

Description

Facing (paving)

Covering the ground surface in the power station with asphalt, etc.

Platform

Footing for work installed under RPV inside pedestal

Flanged tankBolted assembly tanks

Pedestal

A cylindrical basement that supports a body of reactor

Manipulator

Robot arm to support fuel debris retrieval

Mock-up

A model which is designed and created as close to real thing to possible

Boric acid, sodium
chloride

Soluble neutron absorber (boric-acid solution)143

Well shieldplug

Operating floor
Spent
Fuel
Pool
(SFP)

Component
Pit

Primary
Containment
Vessel (PCV)

4th floor

ReactorPressure
Vessel (RPV)

3rd floor

2nd floor

Pedestal
Vent Tube

1st floor

1st basement Fl.

Suppression
Chamber (S/C)

Torus room

（Courtesy of IRID）

Fig. 44 Structural drawing inside Reactor building
Top cover of RPV
Steam dryer
RPV body
Steam separator
Water-feed sparger
Core spray piping
Upper grid plate
Fuel assemblyCore shroud

Core support plate
Fuel assemblyControl rod

Neutron flux measuring guideJet pump

Differential pressure detector/
Piping of Boric acid solution
injection system

Control Rod Drive Mechanism

Pedestal for RPV（Courtesy of IRID）

Fig. 45 Structural drawing inside Reactor Pressure Vessel (RPV)144

List of Attachment
Contents
Attachment 1 Revision of the Mid-and-Long-term Roadmap and the earlier published TechnicalStrategic Plan ......................................................................................................... 146
Attachment 2 Major risk reduction measures performed to date and future course of action..... 148
Attachment 3 Overview of SED indicator ..................................................................................... 154
Attachment 4 Risk sources that are not explicitly addressed in the major risk sources .............. 160
Attachment 5 Change in risk over time ........................................................................................ 162
Attachment 6 Coverage of fuel debris retrieval ............................................................................ 163
Attachment 7 Concept of Safeguards .......................................................................................... 165
Attachment 8 Terms related to radioactive waste management .................................................. 167
Attachment 9 Disposal of radioactive waste,, ............................................................................... 168
Attachment 10 Overall image of Storage Management Plan for the Fukushima Daiichi NPS .... 171
Attachment 11 Prospects of processing/disposal method and technology related to its safety .. 173
Attachment 12 Efforts toward discharging ALPS-treated water into the ocean .......................... 241
Attachment 13 Basic direction of 6 Essential R&D Themes ........................................................ 276
Attachment 14 R&D medium-to-long-term plan ........................................................................... 282
Attachment 15 Current progress of the Project of Decommissioning and Contaminated Water
Management ........................................................................................................ 283
Attachment 16 Major activities related to enhancing international collaboration......................... 287

145

Attachment1 Revision of the Mid-and-Long-term Roadmap and the earlier published TechnicalStrategic Plan
[1st Edition of the Mid-and-Long-term Roadmap (December 21, 2011)]
 In response to completion of Step 2 described in “the Roadmap towards Restoration from the Accident at
the Fukushima Daiichi NPS” compiled by the government and Tokyo Electric Power Company (TEPCO)
after the accident, the necessary measures to be progressed over the mid-and-long-term, including efforts
to maintain securely stable conditions, fuel removal from spent fuel pools (SFPs), fuel debris retrieval, etc.
were compiled by three parties of TEPCO, Agency for Natural Resources and Energy, and Nuclear andIndustrial Safety Agency and conclude at The Government and TEPCO’s Mid-to-Long-Term
Countermeasure Meeting.

 Basic principles towards implementation of mid-to-long efforts were proposed and targets with timeschedules were established by dividing the period up to completion of decommissioning into three parts;
the period up to spent fuel removal start (1st period), the period up to fuel debris retrieval start fromcompletion of the 1st period (2nd period) and the period up to completion of decommissioning fromcompletion of the 2nd period (3rd period).
[Mid-and-Long-term Roadmap Revised 1st Edition (July 30, 2012)]
 “Specific plan on the matters to be addressed with priority to enhance mid-and-long-term reliability”
developed by TEPCO after completion of Step 2 was reflected and revised targets based on the state ofwork progress were clearly defined.
[Mid-and-Long-term Roadmap Revised 2nd Edition (June 27, 2013)]
 Revised schedule was studied (multiple plans were proposed) based on the situation of each Unit
concerning fuel removal from SFP and fuel debris retrieval, and R&D Plan was reviewed based on theabove.
[Technical Strategic Plan 2015 (April 30, 2015)]
 The 1st edition of the Technical Strategic Plan was published to provide a verified technological basis to theMid-and-Long-term Roadmap from the viewpoint of proper and steady implementation of decommissioningoftheFukushima

DaiichiNuclearPower

Station.

(NDF was inaugurated on August 18, 2014 in response to reorganization of existing Nuclear Damage
Compensation Facilitation Corporation)
 Decommissioning of the Fukushima Daiichi Nuclear Power Station was regarded as “Continuous risk
reduction activities to protect human beings and environment from risks caused by radioactive materials
generated by the severe accident”, and Five Guiding Principles (Safe, Reliable, Efficient, Prompt, Fieldoriented) for risk reduction were proposed.
 Concerning the field of fuel debris retrieval, feasible scenarios were studied by regarding the following
methods as the ones to be studied selectively; the submersion-top entry method, the partial submersion-top
entry method, and the partial submersion-side entry method.

 Concerning the field of waste management, policies for storage, control, etc. were studied from a mid-andlong-term viewpoint based on the basic concept for ensure safety during disposal or for a proper
processing method.
[Mid-and-Long-term Roadmap Revised 3rd Edition (June 12, 2015)]
 While much importance was placed on risk reduction, priority-setting for actions was performed so that risks
could definitely be reduced in the long term.

 Targets for several years from now were concretely established including policy decision on fuel debris
retrieval (two years later from now was targeted), volume reduction of radioactive materials contained in
the stagnant water in the buildings by half (FY2018), etc.
[Technical Strategic Plan 2016 (July 13, 2016)]
 In response to the progress state of decommissioning after publication of the Technical Strategic Plan
2015, concrete concepts and methods were developed based on the concept and direction of the efforts

146

of the Technical Strategic Plan 2015 to achieve the target schedule specified in “Policy decision on fueldebris retrieval for each unit” which is expected to be completed by about summer 2017 defined in the
Mid-and-Long-term Roadmap, “Compiling of the basic concept concerning processing/disposal ofradioactive waste” which is expected to be complete in FY2017, etc.
[Technical Strategic Plan 2017 (August 31, 2017)]
 Feasibility study was conducted on the three priority methods for fuel debris retrieval. Recommendations fordetermining fuel debris retrieval policy were made and efforts after policy decision including preliminary
engineering were recommended as strategic recommendations.

 Recommendations were made for compiling the basic concept concerning solid waste
processing/disposal.
[Mid-and-Long-term Roadmap Revised 4th Edition (September 26, 2017)]
 Policy on fuel debris retrieval and immediate efforts were decided based on NDF technicalrecommendations.
 Basic concepts concerning solid waste processing/disposal were compiled.

 Individual work was defined based on the viewpoint of “Optimization of total decommissioning work”.
[Technical Strategic Plan 2018 (October 2, 2018)]
 The Plan added contaminated water management and fuel removal from SFP, and presented the direction
from mid-to-long-term perspective that overlooks entire efforts of decommissioning of Fukushima Daiichi
NPS.
[Technical Strategic Plan 2019 (September 9, 2019)]
 The plan presented the strategic recommendation for determining fuel debris retrieval methods for the first
implementing unit as well as the direction from mid-to long-term perspective that overlooks entire efforts ofdecommissioning of Fukushima Daiichi NPS including waste management, etc.
[Mid-and-Long-term Roadmap Revised 5th Edition (December 27, 2019)]
 The first implementing unit and the method of fuel debris retrieval were determined.
 The methods of fuel removal from SFP in Units 1 and 2 were changed.

 TEPCO maintains the current target to suppress the amount of contaminated water generation to about
150m3/day within 2020, in addition, set the new target to less than 100m3/day within 2025.
[Technical Strategic Plan 2020 (October 6, 2020)]
 The plan characteristically included providing of the Mid-and-Long-Term Decommissioning Action Plan,
identifying of requirements for the study of fuel debris retrieval methods toward further expansion of thescale, clarifying of the concept for ensuring safety in decommissioning operations, and strengthening ofmanagement system in response to the growing importance of R&D.

147

Attachment2 Major risk reduction measures performed to date and future course of action
Change in the risk level over time assessed and expressed by SED for the entire Fukushima
Daiichi NPS is shown in Fig. A 2-1. The vertical axis in the top graph in the figure shows the risk
level in common logarithmic scale and the horizontal axis shows number of years after the accident.
Although the risk level at the time of zero year after the accident was at high level caused by thefuel in SFP which lost its cooling function and the molten nuclear fuel, over the time of 0.5 years
after the accident the risk level has been reduced with a significant decrease in both Hazard
Potential and Safety Management, because of implementation of safety measures including
cooling function restoration of SFPs, cooling of fuel debris with water injection by core spray system,
nitrogen injection, etc. (in 2011) as well as the contribution of inventory and decay heat decrease
due to decay of radioactive materials.
The risk level in 0.5 to 2.5 years after the accident is shown in the enlarged graph (the verticalaxis is in linear scale) with the breakdown of major risk source (fuel debris, fuel in SFP andcontaminated water, and the others) at the bottom left in the figure and the similar graph since 3
years after the accident is given in the bottom right with the risk level multiplied by 10. These graphsdemonstrate that a continuous risk reduction has been achieved.

Evaluation of fuel in SFP 8 years after the accident occurred reflects the results of water temperaturerise in the testing on SFP cooling shutdown. (For detail, see Fig. 3 in Chapter 2 of main part.)

Fig. A2- 1 Reduction of risks contained in the Fukushima Daiichi NPS

148

Change in the risk level with further breakdown of major risk sources over time since 0.5 years
after the accident is shown in Evaluation of fuel in SFP 8 years after the accident occurred reflects theresults of water temperature rise in the testing on SFP cooling shutdown. (For detail, see Fig. 3 in
Chapter 2 of main part.)

Fig. A2- 12. With a logarithmic scale, risk sources can be indicated that are too small to be
displayed in the linear scale of Evaluation of fuel in SFP 8 years after the accident occurred reflects
the results of water temperature rise in the testing on SFP cooling shutdown. (For detail, see Fig. 3 in
Chapter 2 of main part.)

Fig. A2- 1. Fuel in the Common Spent Fuel Storage Pool and the Dry Cask Temporary Custody
Facility are not shown which stay in the region of sufficiently stable management. The “stagnant
water in buildings + zeolite sandbags” shown in Fig. A2- 2 was assessed based on the informationon the stagnant water in buildings for the period of 0-8 years after the accident. However, since 9
years after the accident, the condition of zeolite-containing sandbags placed in the basement of
the process main building and the high-temperature incinerator building has become clear, and thisinformation was incorporated into the assessment.
Among the major risk sources, fuel debris, fuel in SFPs, and the stagnant water in buildings andzeolite-containing sandbags have relatively high risk levels. Although, in recent years, the treatmentof the stagnant water in buildings has progressed and the risk level of the “stagnant water inbuildings + zeolite sandbags” has been on a declining trend, attention should be paid to zeolite
sandbags laid with a high dose because they may hinder future decommissioning work. In addition,
as for the water stored in tanks (flanged tank and welded tank), the risk of leakage andconsequently the risk level has been greatly reduced as a whole because the treatment of the water
stored in the flanged tanks, which have a higher risk of leakage than welded tanks, has proceeded
well.

Risk level (Logarithmic scale)

Site total

0

Fuel in SFPs
Fuel debris
Stagnant water in buildings + Zeolite
sandbags
Contaminated water in trenches
Stored water in tanks
Secondary waste generated bywater treatment
Contaminated structures, etc. inbuildings

2

4

6

8

10

Year after accident

Fig. A2- 2 Change in the risk level for each major risk source(1) Fuel in SFPs
From one year after the accident, rubble was removed and a cover for fuel removal was installed149

at Unit 4 in preparation for fuel removal, thereby enhancing the functions of reducing the risk of fueldamage by rubble in SFP and controlling the dispersion of damaged fuel. Further, 2.5 years after
the accident, fuel removal was started and the fuel was transferred into the Common Spent Fuel
Storage Pool with low Safety Management, and the risk level was lowered (completed in 2014)91.
Although the effect of risk level reduction was observed due to the decrease in Safety
Management through the diffusion control function of the building cover at Unit 1 (installed in 2011),
this effect has been currently lost because the building cover was removed (in 2015) in preparation
for removal of fuel in SFP92. In order to prevent dust scattering during rubble removal, a large cover
will be installed by FY2023, and fuel removal from SFP is planned to start in FY2027 to FY2028.93.
For Unit 2, a gantry for fuel removal will be installed on the south side of the reactor building,and the removal of the fuel in SFP is scheduled to start in FY2024 to FY202693.
In Unit 3, a cover for fuel removal was installed in 2018 after rubble removal was performed in
preparation for fuel removal from SFP, then fuel removal from SFP was started from April 2019.
After that, transfer to the Common Spent Fuel Storage Pool was completed in February 202194.
In case cooling fuel in SFPs is stopped, the pool water temperature may rise and the pool waterlevel may lower due to decay heat. In and after the 8th year after the accident, as a result ofincorporating the observation that the rise in water temperature after cooling shutdown of SFPs
was slower than expected, the risk level of fuel in SFPs is lower than previously estimated, because
the time margin before the risk of water level lowering becomes apparent increases.
(2) Fuel debris
Although fuel debris was at a high risk level just after the accident because it was at molten
state, and in addition, radioactive materials were released, the risk level was reduced, not only bydecay of the radioactive materials, but also by reduction of Hazard Potential and Safety
Management because of restoration and strengthening of cooling function.
As described in (1), the diffusion control function of the building cover of Unit 1 reduced the risk
associated with the dispersion of fuel debris, and lowered the risk level due to the decrease in
Safety Management; however, this effect is currently lost.
(3) Stagnant water in buildings + Zeolite sandbags
Although stagnant water in buildings is generated by cooling of fuel debris and immersion ofgroundwater into the buildings, etc., the risk level has been lowered due to the start of operation ofcesium sorption apparatus (KURION) and Second cesium sorption apparatus (SARRY), the effectof subdrains and land-side impermeable walls, water drainage in condensers, and the start of the91 Decommissioning project, Status of the decommissioning work, Fuel removal work of Unit 4, (Website), Tokyo

Electric Power Company Holdings, Inc.

92 The 57th Study Group on Monitoring and Assessment of Specified Nuclear Facilities, Reference 7 “State ofprogress of Unit 1 of Fukushima Daiichi NPS and rubble removal on the north side of the operating floor”, Tokyo
Electric Power Company Holdings, Inc.
93 Mid-and-Long-term Decommissioning Action Plan (March 25, 2021), Tokyo Electric Power Company Holdings,
Inc.
94 Decommissioning project, Status of the decommissioning work, Fuel removal from spent fuel pool in Unit 3,(Website), Tokyo Electric Power Company Holdings, Inc.

150

operation of Third cesium sorption apparatus (SARRY-II). This stagnant water treatment in thebuildings so far significantly contributes to risk level reduction of the total site following contribution
by fuel removal in SFP.

151

(4) Contaminated water in trenches
Although the contaminated water of high concentration has been stagnated in the seawater pipe
trenches in Units 2 to 4 since immediately after the accident, the trenches were blocked and thetreatment of the stagnant water has been completed (in 2015)95. With regard to the seawater pipe
trench of Unit 1, the concentration of which is lower than that of Units 2 to 4, purification of thestagnant water is under consideration 96。
(5) Stored water in tanks
There are several types of stored water in the tank with different radioactive material
concentrations depending on the stage of purification treatment. First of all, the strontium treatedwater generated from the purification process of the water in the buildings by KURION, SARRY
and SARRY Ⅱ is stored as welded tank water. After that, the risk level is further reduced by multiradionuclide removal equipment (ALPS), etc., and the water is stored in welded tanks as ALPStreated water, etc. (ALPS-treated water and water under treatment). For the concentrated waste
liquid generated from the evaporation-enrichment system, which operated only for a short period
immediately after the accident, the precipitated slurry with a high concentration of radioactive
materials (concentrated waste liquid slurry) was separated, and the remaining liquid (concentrated
waste liquid) is transferred to welded tanks, thereby reducing the leakage risk and lowering the risk
level.
The treatment of the concentrated salt water generated from the treatment with KURION before
ALPS came into operation was completed in 2015 through the operation of ALPS and the advanced
multi-nuclide removal equipment (Advanced ALPS)97。
Risk level of these stored water in the tanks are also lowered by raising and duplexing the weir(for the existing tanks completed in 2014), transferring from flanged tanks to welded tanks, andtreating the Sr-treated water remaining at the bottom of the flanged tanks (in 2019), and treating
ALPS-treated water (in 2020). The remaining concentrated salt water at the bottom of flanged tanks
is being treated for dismantling the tanks.
(6) Secondary waste generated by water treatment
Many radioactive materials have moved from contaminated water to secondary waste throughwater treatment. What has been generated includes the sludge from decontamination device, thewaste sorption vessels by operation of KURION and SARRY (in 2011) and by the SARRY-II (in2019), ALPS slurry by operation of ALPS (in 2013), the waste sorption vessels by the advanced
ALPS (in 2014), waste sorption vessels by the mobile-type treatment system that treated seawater
95 Decommissioning project, Status of the decommissioning work, Removal of contaminated water in seawater pipe

trenches, (Website), Tokyo Electric Power Company Holdings, Inc.
91st Study Group on Monitoring and Assessment of Specified Nuclear Facilities “Reference 1 ”Work
Schedule for Instruction Items to be Considered Based on the Measures for Mid-and-long-term Risk Reduction
at TEPCO’s Fukushima Daiichi NPS (Risk Map), (March 2021 Edition)”, Tokyo Electric Power Company Holdings,
Inc.
97 Decommissioning project, Status of the decommissioning work, Purification of contaminated water, (Website),Tokyo Electric Power Company Holdings, Inc.
96 The152

pipe trenches, etc. Although the sludge from decontamination device greatly contributes to the risk
level, sludge is not newly generated at present, and thus, the risk level of the total secondary waste
generated by water treatment is not on an increasing trend. As a tsunami countermeasure, thedecontamination sludge stored in the main process building (T.P. 8.5m) will be extracted (planned
for FY2023), placed in a storage container, and transferred to the elevated area (T.P. 33.5m) 98.
Although the concentrated waste liquid slurry separated from the concentrated waste liquid wasstored in horizontal welded tanks without the weir and placed on the ground without the base, its
risk level has been lowered due to the approach to safety taken by installing the reinforced-concretebase and the weir.
(7) Contaminated structures, etc., in the buildings
There is no significant change at the present moment in the risk level of contaminated structures,etc. in the buildings comprised of structures, piping, components, etc. (shield plug, piping of
emergency gas processing system and the like) in the reactor buildings, PCVs or RPVs that arecontaminated by dispersed radioactive materials caused by the accident.
(8) Rubble, etc.
Rubbles, etc. as solid waste are stored under a variety of conditions such as in solid waste
storage, in temporary waste storage and by outdoor accumulation. Each has different Safety
Management, and the rubbles stored in outdoor sheet covered storage and outdoor accumulation
are of the highest risk level. In the past, the facilities with better management condition have beenenhanced by soil covered temporary storage facilities (in 2012), felled tree temporary storage pool(in 2013), expansion of solid waste storage facilities (in 2018), etc. In addition, the rubble from
temporary storage facilities was transferred to the better-controlled solid waste storage facility (in2020). Furthermore, outdoor temporary storage is planned to be discontinued by the end of FY
2028 by increasing incinerators, volume reduction installations and solid waste storages, etc., in
accordance with the Solid Waste Storage Management Plan99.

98 Safety Monitoring Council for the decommissioning of Fukushima Daiichi NPS (March 8, 2021), Reference 5-1
99 The Solid Waste Storage Management Plan at the Fukushima Daiichi NPS (July 2021 Edition), Tokyo ElectricPower Company Holdings, Inc.

153

Attachment3 Overview of SED indicator

Risk analysis targeting various risk sources, which have diverse characteristics and exist all overthe site, was conducted in reference to the SED indicator100 developed by the NDA. The SED
indicator is an important factor to decide priority to implement risk reduction measures. It waspartially modified (refer to the following pages) so that unique characteristics of the Fukushima
Daiichi NPS could be easily reflected when it was applied to the Fukushima Daiichi NPS. Overview
of the SED indicator and the modified part to be applied to the Fukushima Daiichi NPS aredescribed below.

The SED indicator is expressed by the following formula. The first formula is the one widely used
for waste assessment and the second is for contaminated soil assessment. In each formula, thefirst term is referred as to “Hazard Potential” and the second as “Safety Management” of risk
sources.
𝑆𝐸𝐷 = (𝑅𝐻𝑃 + 𝐶𝐻𝑃) × (𝐹𝐷 × 𝑊𝑈𝐷)or𝑆𝐸𝐷 = (𝑅𝐻𝑃 + 𝐶𝐻𝑃) × (𝑆𝑆𝑅 × 𝐵𝐸𝑅 × 𝐶𝑈)
Each indicator is explained below. Although CHP stands for “Hazard Potential” of the chemical
substance, details are not given here as it is not used in this section.

(1) Hazard Potential
Radiological Hazard Potential (RHP) is an indicator representing the potential impact ofradioactive materials and represents the impact to the public by the following formula when the
total amount of radioactive materials is released.

𝑅𝐻𝑃 = 𝐼𝑛𝑣𝑒𝑛𝑡𝑜𝑟𝑦 ×

𝐹𝑜𝑟𝑚 𝐹𝑎𝑐𝑡𝑜𝑟
𝐶𝑜𝑛𝑡𝑟𝑜𝑙 𝐹𝑎𝑐𝑡𝑜𝑟

Inventory is defined as shown below by Radioactivity of risk sources and the Specific Toxic
Potential (STP) and corresponds to the effective radiation dose101 . The STP is defined as thevolume of water required to dilute 1TBq of radioactive materials and corresponds to the radiationdose coefficient. Ingestion of a certain amount of such diluted water throughout the year will resultin a radiation exposure dose of 1mSv. The SED indicator conservatively uses the larger radiationdose coefficient between ingestion and inhalation.
𝐼𝑛𝑣𝑒𝑛𝑡𝑜𝑟𝑦(𝑚 ) = 𝑅𝑎𝑑𝑖𝑜𝑎𝑐𝑡𝑖𝑣𝑖𝑡𝑦(𝑇𝐵𝑞) × 𝑆𝑇𝑃(𝑚 ⁄𝑇𝐵𝑞)

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

154Form Factor (FF), as shown in Table A3-1, is an indicator representing how much radioactive
material is actually released depending on material form, such as gas, liquid, solid, etc. The
indicator is set assuming that 100% of radioactive material is released in the case of gas and liquid
when containment function is totally lost and that 10% of radioactive material is released in thecase of powder based on the measurement data. Because of no clear basis, the indicator in caseof solid is set to a sufficiently small value assuming that the solid materials are less easily released.
In Table A3-1, several expected forms, especially for fuel debris, are added to the definition usedby the NDA. The scores for the form of No.4 and No.5 are newly established.
Control Factor (CF), as shown in Table A3-2, is an indicator representing time allowance available
before restoration when safety functions maintaining current stable state are lost. CF is taking into
account exothermicity, corrosivity, flammability, hydrogen generation, reactivity with air or water,
criticality, etc. which are typical characteristics of risk sources. CF is the same as the one definedby the NDA.
(2) Safety Management – FD and WUD
Facility Descriptor (FD) is an indicator representing whether containment function of the facility
is sufficient or not. Risk sources are ranked by score based on a combination of the factors including
integrity of the facility, redundancy of containment function, safety measure condition, etc.
Waste Uncertainty Descriptor (WUD) is an indicator representing whether any impact isgenerated or not when the risk source removal is delayed. Risk sources are ranked by score basedon a combination of the factors including degradation or activity of the risk source, packaging state,monitoring condition, etc.
As these indicators are difficult to be applied to the Fukushima Daiichi NPS if they are used as
defined by the NDA, they are re-defined as shown in Table A3-3 and Table A3-4 respectively.
(3) Safety Management - SSR, BER and CU
The definition of SSR, BER and CU used for Safety Management assessment for contaminatedsoil is the same as the one defined by the NDA and each score is shown in Table A3-5.
Speed to Significant Risk (SSR) is an indicator concerning the time until the public is affected
through such as distance to the site boundary, groundwater flow conditions, etc. and to assess
urgency of taking measures.
Benefit of Early Remediation (BER) is an indicator to assess benefits obtained from early
implementation of measures against risks.
Characterization Uncertainty (CU) is an indicator to assess reliability or uncertainty in the risk
assessment model.

155

Table A3-1 Definition and score of FF
No.

Form

FF

1

Gas, liquid, watery sludge* and aggregated particles*

1

2

Other sludge

1/10 = 0.1

3

Powder and removable contaminants (surface
contamination, etc.)*

1/10 = 0.1

4Adhesive* or penetrating contaminants (surface
penetrating contamination)*

1/100 = 0.01

5Fragile and easily decomposable solid (porous MCCI
(Molten Core Concrete Interaction), etc.)*

1/10,000 = 1E-4

6Discrete solid (transportable size and weight byhuman power such as pellets, etc.)

1/100,000 = 1E-5

7Large monolithic solid, activated component

1/1,000,000 = 1E-6

* : Form which is added to the NDA definition to enhance applicability to the case of the Fukushima
Daiichi NPS
Table A3-2 Definition and score of CF

Category

No.

Time allowance available before any risk is realized

CF

1

Hours

1

2

Days

10

3

Weeks

100

4

Months

1,000

5

Years

10,0006

Decades

100,000

Table A3-3 Criteria and score of FD
Criteria(NDA definition is modified to enhance applicability to the caseof the Fukushima Daiichi NPS)

NDF
Score

1

No component for diffusion control function exists. Therefore, noassessment for containment function is available.

100

2“Safety assessment criteria*2” are not satisfied at “the time ofassessment*1” caused by the accident effects, etc. The component for
diffusion control function is single.

91

3“Safety assessment criteria” are not satisfied at “the time of assessment”
caused by the accident effects, etc.
The component for diffusion control function is multiple.

74

4“Safety assessment criteria” are not satisfied until “the time of work
(such as transfer, treatment, recovery, etc.) *3” for the risk source
contained in the component for diffusion control function.
The component or diffusion control function satisfying “safetyassessment criteria” exists at “the time of assessment”.

52

156

5

Integrity of diffusion control function has been assessed and “safetyassessment criteria” are satisfied until “the time of work (such astransfer, treatment, recovery, etc.)” for the risk source.
Frequency of occurrence of “contingency*4” is high, and whencontingency occurs countermeasures preventing diffusion of the risk
source contained in the component are not sufficient.
The component for diffusion control function is single.

29

6“Safety assessment criteria” is satisfied until “the time of work (such astransfer, treatment, recovery, etc.)” for the risk source.
Frequency of occurrence of “contingency” is high, and countermeasures
preventing diffusion of the risk source contained in the component are
not sufficient.
The component for diffusion control function is multiple.

15

7“Safety assessment criteria” are satisfied until “the time of work (such astransfer, treatment, recovery, etc.)” for the risk source.
Facilities dissatisfying “safety assessment criteria” exist in thesurrounding area, and the potentiality is high to make (receive) the
diffusion impact*5 of the risk source to (from) these adjacent facilities.
The component for diffusion control function is single.

88“Safety assessment criteria” are satisfied until “the time of work (such astransfer, treatment, recovery, etc.)” for the risk source.
The potentiality is high to make (receive) the diffusion impact of the risk
source to (from) these adjacent facilities.
The component for diffusion control function is multiple.

5

9“Safety assessment criteria” are satisfied until “the time of work (such astransfer, treatment, recovery, etc.)” for the risk source.
The potentiality is low to make (receive) the diffusion impact of the risk
source to (from) these adjacent facilities.
The component for diffusion control function is single.

3

10“Safety assessment criteria” are satisfied until “the time of work (such astransfer, treatment, collection, etc.)” for the risk source.
The potentiality is low to make (receive) the diffusion impact of the risk
source to (from) these adjacent facilities.
The component for diffusion control function is multiple.

2*1. This refers to “at the time” of study on SED score, i.e., “at the present time” of assessment.
*2. “Safety assessment criteria” described in this sentence refer to “the matters for whichmeasures should be taken” or “securing of diffusion control function within the scope of designbasis event”.
*3. This refers to the time of “recovery” of the risk source for disposition and carrying out for which
SED score shall be studied.
*4. External events (natural disasters, etc.) are postulated as contingencies.
*5. The potentiality of diffusion of the risk source exists to (from) adjacent facilities when facilities
receive external impact caused by contingencies or impact caused by any events (fire, etc.),
etc.

Category

1

Table A3-4 Criteria and score of WUD
Criteria(NDA definition is modified to enhance applicability to the caseof the Fukushima Daiichi NPS)
The material is fuel (which contains fissile material) and active*1.
Necessary information (existent amount, existent location, radioactivity,etc.) for work including treatment, recovery, etc. is insufficient (cannot

157

NDF
Score
100

be confirmed or estimated), and control and surveillance with
monitoring, etc. are unavailable.
Handling is impracticable for the current form or condition because ofreasons where the form is not proper for handling, or that it is not stored
in a special container.

2

The material is fuel and active (which has fissile properties).
Necessary information for work including treatment, recovery, etc. isinsufficient, and control and surveillance are unavailable.
Handling is practicable for the current form or condition because ofreasons where the form is proper for handling or that it is stored in a
special container.

90

3

Although the material is active, it is not fuel (but waste).
Necessary information for work including treatment, recovery, etc. isinsufficient.

74

4

The material is fuel and active (which has fissile properties).
Necessary information for work including treatment, recovery, etc. is
obtained (can be confirmed or estimated), and control and surveillancewith monitoring, etc. are available.
Handling is impracticable for the current form or condition.

50

5

The material is fuel and active (which has fissile properties).
Necessary information for work including treatment, recovery, etc. is
obtained, and control and surveillance are available.
Handling is practicable for the current form or condition.

30

6Although the material is active, it is not fuel (but waste).
Necessary information for work including treatment, recovery, etc.

177Although the material is inactive*2, it has physical or geometricalinstability.
Handling is impracticable for the current form or condition.

98

Although the material is inactive, it has physical or geometricalinstability.
Handling is practicable for the current form or condition.

5

9

The material is inactive and has no physical or geometrical instability or
has sufficiently low level of instability.
Handling is impracticable for the current form or condition.

3

10

The material is inactive and has no physical or geometrical instability or
has sufficiently low level of instability.
Handling is practicable for the current form or condition.

2*1 “Active” refers to possession of activity defined by CF at such a significant level as that activity
affects control and work.
*2 “Inactive” refers to non-possession of activity or possession of sufficiently low level of activity.

158

Table A3-5 Definition and score of SSR, BER and CU
Indicator
SSR

BER

Score

Criteria

25

Risks may be realized within 5 years.

5

Risks may be realized within 40 years.

1

40 years or over (There is very little possibility that risks are realized.)

20

Implementation of measures can reduce risks by 2 or more orders ofmagnitude or can facilitate control stepwise.

4

Implementation of measures can reduce risks by 1 or more order ofmagnitude, but cannot facilitate control.

1

Implementation of measures can only bring negligible risk reduction effects,
and cannot facilitate control, either.

20(1)+(2)=
5 to 6points4(1)+(2)=3 to 4points1(1)+(2)=
2 points

CU(1) Assessment for the present state
1 point: Major nuclear types and diffusion pathways aremonitored.
2 points: Monitored, but insufficient data for construction ofassessment model
3 points: Not monitored(2) Assessment on future prediction
1 point: Sufficient site characteristics are obtained forconstruction of assessment model.
2 points: Major characteristics representing the site are
obtained.
3 points: There is no model usable for future prediction

159

Attachment4 Risk sources that are not explicitly addressed in the major risk sources

Major risk sources are listed in the Table 1 in Chapter 2 of the body part. Looking ahead to thedecommissioning of the entire Fukushima Daiichi NPS, it is necessary to focus on risk sources thatare not explicitly addressed in the major risk sources. Table A4-1 focuses on waste existed before
the accident and radioactive materials with low concentration diffused by the accident, and issummarized with reference to Measures for Mid- term Risk Reduction at TEPCO’s Fukushima
Daiichi NPS (Risk Map)” provided by the NRA.102.
Table A4- 1
Issue

Liquidradioactive
materials

Spent fuelRisk sources that are not explicitly addressed in the major risk sources (1/2)
Risk source

Descriptions

Underground watertank

The residual water in all the underground water tanks werecompletely recovered 103. Dismantling and removal policies are underconsideration.

Accumulated wateron siteExtracted by the comprehensive risk inspection performed in 2015104.
Since then, the concentration of radioactive materials and volume ofwater are being checked accordingly 105。

Drainage

In drainage A, Cs-137: lowered to ND ~ 23 Bq/L 106. In drainage K,
the contamination source on the roof of the Unit 2 Reactor building
was removed, and the contamination level fell to 67 Bq/L. In addition,
purification materials were installed107, and measures such as
operation of discriminating-type PSF monitors were taken108.

Sludge on the floor
in the buildings

The floor surface of turbine buildings and radioactive waste disposal
buildings of Units 1 to 4, waste process building and Unit 4 reactor
building remain exposed, and radioactivity of sludge after the
exposure was 1.9×1013Bq109. For reactor buildings of Units 1 to 3,
process main building and high temperature incinerator building,stagnant water processing is underway.

Spent control rods,
etc.

Spent control rods, etc.: 24,030. Shroud fragments, etc.: 193 m3110.
The major nuclide is Co-60.

102 NRA, Measures for Mid-term Risk Reduction at TEPCO’s Fukushima Daiichi NPS(Risk Map)”, (March 2021
Edition)
103 The 44th Meeting of the Secretariat of the Team for Countermeasures for Decommissioning and Contaminated
Water Treatment “Reference 3-6: On-site Monitoring Status (Conditions of Water Discharge Channels in Units 1
to 3 and Underground Water Storage Tanks)”
104 Comprehensive Risk Inspection of Fukushima Daiichi NPS that impacts outside the Site Boundary - Review
Results - (April 28, 2015) Tokyo Electric Power Co., Inc.
105 The 88th Meeting of the Secretariat of the Team for Countermeasures for Decommissioning and Contaminated
Water Treatment “Reference 1: Status of contaminated water and other accumulated water on the premises (asof March 18, 2021)”
106 The 32nd Study Group on Monitoring and Assessment of Specified Nuclear Facilities “Reference 2: Status of
measures for reducing the concentration of waste water in drainage K”
107 The 63rd Study Group on Monitoring and Assessment of Specified Nuclear Facilities “Reference 2: Measures
for rainwater inflow control (Progress status of installing purification materials for rainwater drainage in turbine
buildings)”
108 The 74th Meeting of the Secretariat of the Team for Countermeasures for Decommissioning and Contaminated
Water Treatment “Reference 3-6: Starting of operation of PSF monitor in the drainage K”
109 The 87th Study Group on Monitoring and Assessment of Specified Nuclear Facilities “Reference 3-5 : Progressof the treatment of stagnant water in buildings, etc.”
110 NRA, Material of interview with the licensee “Solid Waste at Fukushima Daiichi NPS” September 21, 2018,
Tokyo Electric Power Company Holdings, Inc.

160

In-pool waterSolidradioactive
materials

Fuel in Units 5/6
SFP

Unit 5 : 1,374, Unit 6 : 1,456111

Waste before
the earthquake

185,816 drums are stored112. The major nuclide is Co-60.

Contaminatedsoil

As a result of the topsoil analysis, more than half of the samples are
in excess of the designated standards (8,000 Bq/kg) based on the
Act on Special Measures Concerning the Handling of Environmental
Pollution by Radioactive Materials116

Rubbles aroundbuildings

Dismantling of rubbles scattered on the roof floor of the buildings due
to hydrogen explosions is now in operation and planned. The amountof rubbles has not been confirmed.

Exhaust stack

Countermeasures
to external
events, etc.

Importantissues toprogress

decommissioning

Salt removal in Units 2 to 4 was completed in 2013.

Megafloat

Exhaust stack of Units 1/2 : dismantlement work was carried out
since August 2019, and the upper part of 61 m out of the total heightof 120 m was divided into 23 blocks in total for dismantling. On May
1, 2020, a lid was installed on a barrel 59 meters above the ground toprevent rainwater inflow, and dismantling was completed113.
Exhaust stack of Units 3/4 : Measured 3mSv/h at the base114.
The work of bottoming and internal filling was completed115. Revetment
maintenance and embankment work are underway.

Dust inoperating floor

Below the target value of release control (1×107 Bq/h). Gradually
declining116.

Rainwater inflow
into buildings

Rubble on the roof was removed and waterproofing was newly
provided. Purification materials were installed in the gutters. Checkvalves were installed in the drain pipes. The roof drain was repairedand closed117. Facing of the Elevation T.P.2m, T.P.6m and T.P.8.5m
was completed118.

Radiation sourceon the 3rd and 4th
floors of Unit 3 R/B

On the 3rd floor, beams at several locations were damaged. A
maximum of 45 mSv/h was measured. On the 4th floor, 104 mSv/h
was observed119.

111 The 89th Meeting of the Secretariat of the Team for Countermeasures for Decommissioning and Contaminated

Water Treatment “Reference 4-2 : Storage status of spent fuel , etc.”

112 NRA, Material of Interview with Licensee “Restoration Status of Exhaust Radiation Monitor at Auxiliary

Common Facilities for Common Spent Fuel Storage Pool and Ventilation & Air Conditioning System at Fuel
Storage Area of Fukushima Daiichi NPS” September 21, 2018, Tokyo Electric Power Company Holdings, Inc.
113 “Completion of Dismantling of Exhaust Stack of Units 1/2 at Fukushima Daiichi NPS” (May 1, 2020), Tokyo
Electric Power Company Holdings, Inc.
114 The 19th Committee on Accident Analysis of the Fukushima Daiichi NPS “Reference 4 : Interim report on theinvestigation and analysis of the accident at TEPCO’s Fukushima Daiichi NPS (proposal)”」
115 The 81st Secretariat Meeting of the Team for Countermeasures for Decommissioning and Contaminated Water
Treatment, “Reference 3-1:Progress Status of Megafloat Project at Fukushima Daiichi NPS to Reduce Tsunami
Risks
116 Daily Analysis Results of Radioactive Materials at Fukushima Daiichi NPS, (Website), Tokyo Electric Power
Company Holdings, Inc.
117 The 78th Secretariat Meeting of the Team for Countermeasures for Decommissioning and Contaminated Water
Treatment, “Reference 3-1: Progress Status in Rooftop Rainwater Measures”
118 The 84th Study Group on Monitoring and Assessment of Specified Nuclear Facilities “Reference 1-3: Progressand status of studies on measures to control the generation of contaminated water, amount of groundwaterand rainwater inflow per building”
119 The 14th Study Committee on Accident Analysis of the Fukushima Daiichi NPS “Reference 3: Progress of onsite Investigation”

161

Attachment5 Change in risk over time

Overview of the concept of risk management in the UK is shown in Fig. A5-1. Even if the current
risk level is plotted in the white region of the graph, it does not mean such risk level can always be
accepted over time, but the time will come when such risk level cannot be accepted in the future(yellow region). In addition, as time passes, the risk level may increase caused by degradation offacilities and risk sources (represented by the dotted line). On the other hand, when risk reduction
measures are taken, the risk level can be reduced so that it may not reach the unacceptable region(red region) with careful preparation and thorough management, although it may be temporarily
increased. In this way the risk level shall be targeted to be sufficiently reduced (represented by thesolid line) so that it may not reach into the unacceptable or intolerable region.

Fig. A5-1 Change in risk over time120

120 V. Roberts, G. Jonsson and P. Hallington, “Collaborative Working Is Driving Progress in Hazard and Risk

Reduction Delivery at Sellafield” 16387, WM2016 Conference, March 6-10, 2016. M. Weightman, “TheRegulation of Decommissioning and Associated Waste Management” 1st International Forum on theDecommissioning of the Fukushima Daiichi Nuclear Power Station (April 2016)..

162

Attachment6 Coverage of fuel debris retrieval

In the Mid-and-Long-term Roadmap issued on December 21, 2011, fuel debris is described as“material in which fuel and its cladding tubes, etc. have melted and re-solidified”, namely, fuel debris
is “fuel assembly, control rod and structures inside reactor have melted and solidified together”
according to the report by IAEA121,122.
The condition inside PCV is as shown in Fig. A6-11, as the comprehensive estimations from theinside investigation of reactor, the past accidents including TMI-2 or ChNPP-4, and the result of
the simulation test. It does not show any of specific unit. For more detail, as shown in the Fig A61, fuel debris can be classified by form such as damaged pellets, debris, crust, etc.

Fig. A6-1 Estimated inside of the PCV of the Fukushima Daiichi NPS
121 International Atomic Energy Agency Experiences and Lessons Learned Worldwide in the Cleanup andDecommissioning of Nuclear Facilities in the Aftermath of Accidents, IAEA Nuclear Energy Series No. NW-T2.7, Vienna (2014)
122 Managing the Unexpected in Decommissioning, IAEA Nuclear Energy Series No. NW-T-2.8, Vienna (2016)163

Since nuclear fuel material requires considerations to prevent criticality, it is rational that objects
which exist inside PCV should be broadly sorted into two from the viewpoint of retrieval,containment, transfer and storage. The one includes nuclear fuel material and the others. The onethat does not include nuclear fuel material is to be treated as a radioactive waste in case radioactive
cesium or cobalt are contained or adhered.
Based on this, an example of fuel debris concept as a retrieval target of fuel debris is as shown in
Fig. A6-22. Objects generated by core damage have been classified depending on necessity ofcriticality measures and the content of fuel, in spite that a lot of names are used according to thecontent of fuel component or form in appearance.

Fig. A6-2 An example of organized concept of fuel debris as fuel debris retrieval target at the
Fukushima Daiichi NPS【Glossaries and Terms】
FCM : Fuel Containing Materials. It refers broadly that molten fuel component comes to solidify inconjunction with structural materials. It is also called lava-like FCM due to its appearance.
Corium : A substance that mainly fuel assembly and component of control rod as core component
have molten and solidified.
Crust : A hard outer layer or shell on the surface. When molten fuel is solidified, it may become ahard solid state of shell because of higher cooling speed on the surface layer.
MCCI product : A product generated by Molten Core Concrete Interaction, that includes calcium,silicone, etc. which are concrete component.
Material with fuel : A substance that molten fuel has adhered to material that does not include fuelcomponent originally, like CRD housing, grating and s, then solidified. It is
possible to confirm fuel adhesion state by sight.
Material contaminated by fuel : A substance that adhering molten fuel cannot be confirmed bysight, but fuel component can be detected with α ray detector. It is
impossible to locate fuel component other than using by electron
microscope because particle of adhered fuel component isextremely small and whit.

164

Attachment7 Concept of Safeguards

Safeguards are verification activities undertaken to ensure that nuclear materials are used only
for peaceful purposes and not diverted to nuclear weapons, etc.
Japan has concluded the Safeguards Agreement between Japan and the International Atomic
Energy Agency (IAEA) pursuant to the “Treaty on the Non-Proliferation of Nuclear Weapons” (NPT).
In accordance with this Agreement, relevant domestic laws (the Law for the Regulations of Nuclear
Source Materials, Nuclear Fuel Materials and Reactors (the Nuclear Reactor Regulation Law) andothers) have been developed to establish a domestic safeguards system and accept the IAEA’s
safeguards.
Specifically, the Safeguards Office of the Nuclear Regulatory Agency (JSGO) has confirmed that
all nuclear materials in Japan have not been diverted to nuclear weapons, etc., by conducting thefollowing activities, and the IAEA has acknowledged this authorization through inspections, etc.
(1) Material accountancy by nuclear operators
A nuclear operator shall determine the places where nuclear materials are handled at its
facilities, shall perform a strict and accurate material accountancy including the amount of
nuclear materials to be entered or transferred from its premises and the periodic inventory of
nuclear materials, and shall report to the NRA. The JSGO will compile these reports received
from operators and submit them to the IAEA.
(2) Containment/Monitoring
To ensure that nuclear materials have not been secretly transferred, the JSGO and IAEA haveattached a “seal” on the lids of containers in which nuclear materials are stored and on thegateway for fuel at nuclear power plants. At nuclear power plants, etc., “surveillance cameras”
are installed to constantly monitor the movement of nuclear materials.
(3) Inspection
Inspectors from the JSGO and the IAEA will visit the nuclear facilities in person and conduct
inspection activities such as verifying the consistency of reports and records, checking thequantity and volume of nuclear materials on site, collecting samples for analysis, andassessing the containment/monitoring data.

165

*1 : Except for complementary access occurred during regular domestic inspection*2 : Based on Nuclear Power Plant Regulation, specify Nuclear Material Control Center (authorized foundation)
as “Specified safeguards inspection Agency” and “Specified information processing Agency”

Fig. A7-1 Safeguards implementation system in Japan
Sited from : “Safeguards” in NRA website, https://www.nsr.go.jp/activity/hoshousochi/index.html

The safeguards shall be applied to the Fukushima Daiichi NPS without exception. TEPCO, asa nuclear operator, will consult with the JSGO and IAEA to determine the method of implementing
the safeguards (material accountancy, containment/monitoring, inspection) to be applied to
Fukushima Daiichi NPS, and practically apply them on site.

166

Attachment8 Terms related to radioactive waste management

IAEA Safety Requirements GSR-Part 5123explains that predisposal of radioactive waste
encompass all stages of radioactive waste management from generation to disposal, including
processing, storage and transportation. Terms related to the management of radioactive waste as
defined in the IAEA glossary are shown in Fig A8-1. Within the pre-disposal management,
processing of radioactive waste is classified into pretreatment, treatment and conditioning.
Processing is carried out to be in the form of waste suitable for selected or anticipated disposal
options. Radioactive waste may also be stored in for its management, therefore it is thought to benecessary that the form is suitable for transportation and storage.
Processing: a generic term used to refer to thefollowing steps that change waste
characteristics
Pretreatment: a process before treatment(Treatment)
Treatment: a process intended to benefit safety
and economics efficiency by changing waste
characteristics
Conditioning: a process of manufacturing awaste package suitable for handling,transportation, storage and / or disposal

Fig. A8-1

Terms related to radioactive waste management (IAEA)124 and their translation

examples (For the Japanese translation example, refer to the materials of the Japan Atomic
Energy125,126)

123 IAEA, Predisposal Management of Radioactive Waste, IAEA Safety Standards Series No. GSR Part 5, (2009).

(NSRA, IAEA Safety Standard/Predisposal of Radioactive Waste/General Safety Requirement 5, No. GSRPart5, July, 2012)
124 IAEA, IAEA Safety Glossary Terminology Used in Nuclear Safety and Radiation Protection 2007 Edition, p.216,(2007).
125.AESJ, The Report of 2013, - Organizing information of radioactive waste and matters to be considered forsolving the issues (p.7), March 2014, the Expert Committee, “Processing /disposal of radioactive waste
generated by the accident of Fukushima Daiichi NPS”
126 AESJ, Seiya Nagao and Masafumi Yamamoto, “Introduction to radioactive waste - Management of radioactive
waste from operation and decommissioning of nuclear and other facilities (!) Perspective of radioactive waste
management, the 56 of (9) of Journal of the Atomic Energy Society of Japan, p.593, (2014).

167

Attachment9 Disposal of radioactive waste127,128,129

1. International classification of radioactive waste
Radioactive waste contaminated with radioactive materials is generated through operation anddismantling of nuclear power plants and the use of radioisotopes in medical and industrial
applications. Radioactive waste shall be classified appropriately according to the radioactivity leveland properties of waste, types of radioactive materials, etc., and strictly controlled, and then shallbe reasonably processed and disposed of so as not to affect the human living environment.
The IAEA's Specific Safety Requirements SSR-5 “Disposal of Radioactive Waste” (2011) 130
specifies that a preferred strategy for the management of radioactive waste that is internationallyagreed is to contain the waste and isolate it from the living environment, while minimizing thegeneration of radioactive waste. The required isolation and containment depend on the magnitudeof the hazards of the waste and the time, thereby a disposal option (design and depth of facilities)
being selected accordingly.
The IAEA’s General Safety Guide GSG-1 “Classification of Radioactive Waste”131 indicates therelationship between the classification of radioactive waste and disposal options depending on themagnitude of the hazards (amount of radioactivity) and the duration (the half-life) of the radioactive
waste, as shown in Fig. A9-1. Each classification is also shown in Table A9-1.

Fig. A9-1 Conceptual diagram of waste classification
127 Osamu Tochiyama, Principles and Basics of Radioactive Waste Disposal, Radioactive Waste Management

Funding and Research Center (a Public Interest Incorporated Foundation) (2016)

128 https://www.enecho.meti.go.jp/category/electricity_and_gas/nuclear/rw/
129 https://www.fepc.or.jp/nuclear/haikibutsu/index.html

130 IAEA SSR-5 "Disposal of Radioactive Waste"(2011)

131 IAEA GSG-1" Classification of Radioactive Waste" (2009)168

Table A9-1

Classification of radioactive waste in GSG-1

Classification

Description of Classification

Exempted waste (EW)

Waste satisfying the criteria for clearance, exclusion and exemption
from regulatory control for radiation protection purposes

Very short-lived waste(VSLW)

Waste that is decay-stored for a limited period of time up to several
years and then exempted from regulatory control, as approved by
the regulatory body.

Very low level waste(VLLW)

Waste that does not necessarily satisfy EW standards but does not
require high-level containment and isolation. Suitable for disposal inshallow landfills where regulatory control is limited.

Low level waste (LLW)

Waste that exceeds clearance levels but has a limited amount oflong-lived nuclides. Rigid isolation and containment are required for
periods of up to several 100 years and are suitable for disposal inengineering facilities in shallow soils.

Intermediate level waste(ILW)

Waste that requires higher-level containment and isolation than thenear surface disposal because of the nuclides it contains, especiallylong-lived nuclides. However, considerations on heat removal are
hardly required. Because ILW may contain concentrations of longlived nuclides (especially α-nuclide) that are not manageable innear surface disposal, a depth of tens to hundreds of meters are
required for disposal.

High Level waste (HLW)

Waste with a large amount of heat generation at high activity
concentration levels or waste containing large amounts of long-lived
nuclides for which a design equivalent to a disposal facility for such
waste needs to be considered. Generally, waste is disposed of in astable stratum at the depth of several hundred meters or more fromthe ground surface. In some countries, spent fuel is classified as
HLW.

2. Classification and disposal in Japan
In Japan, radioactive waste is broadly divided into “low-level radioactive waste” (equivalent to
VLLW to ILW in GSG-1), which is generated through the operation of nuclear power plants, and“high-level radioactive waste” (equivalent to HLW of GSG-1), which is generated through thereprocessing of spent fuel that is generated through the operation of nuclear power plants and isvitrified with a high level of radioactivity. When disposed of, waste shall be classified appropriately
according to its radioactivity level and properties, types of radioactive materials, etc., and shall bestrictly controlled, and reasonably processed and disposed of under the principle that
responsibilities lie with those who have generated the waste.
“High-level radioactive waste” is a vitrified waste liquid with a high radioactivity level that is
produced in the process of reprocessing spent fuel generated through the operation of nuclear
power plants. In Japan, the act (the Designated Radioactive Waste Final Disposal Act (the Final
Disposal Act)) stipulates that radioactive waste shall be disposed of in strata more than 300 meters
deep underground.

169

The term “low-level radioactive waste” refers to all types of radioactive waste other than “highlevel radioactive waste”, and is further divided into several categories depending on where it isgenerated and the level of radioactivity.
The types of radioactive waste generated by the operation of nuclear power plants and thedisposal methods assumed are shown in Table A9-2.
Of these, only waste with relatively low-level radioactivity generated through the operation of
nuclear power plants has been subject to disposal in pits since 1992 at the Rokkasho Low-level
Radioactive Waste Disposal Center of Japan Nuclear Fuel Limited in Rokkasho Village, Aomori
Prefecture. Including the existing facilities, approximately 1 million drums of waste contained in
200-liter drums are planned to be buried, and eventually the scale will be enlarged to approximately
3 million drums using 200-liter drums.
Table A9-2

Types of Radioactive Waste Generated by the Operation of Nuclear Power Plants

Fig. A9-2

Japan Nuclear Fuel Ltd. Low-level Radioactive Waste Disposal Center
170

Attachment10

Overall image of Storage Management Plan for the Fukushima Daiichi NPS 132

132 TEPCO, Solid Waste Storage Management Plan for Fukushima Daichi Nuclear Power Station of Tokyo ElectricPower Company Holdings, Inc., July 2021 edition (issued on July 29, 2021)

171

(a) Present storage condition of “rubble, etc.” and “secondary waste generated by water
treatment”

Future storage condition of “rubbles, etc.” and “secondary waste generated by water
Fig. A10-1 Present and future storage conditions of “rubble, etc.” and “secondary waste
generated by water treatment” on site of the Fukushima Daiichi NPS

172

Attachment11

Prospects of processing/disposal method and technology related to its safety

Table of contents

1. Overview ................................................................................................................................... 174
2. Policy for developing Technical Prospects .............................................................................. 176
3. Example cases from other countries for volume reduction of dismantled waste .................... 177
3.1 Case examples in the UK ................................................................................................... 177
3.1.1 Background .................................................................................................................. 177
3.1.2 LLW waste management policy based on the waste hierarchy concept .................... 177
3.2 Case examples in US ......................................................................................................... 181
3.2.1 Background .................................................................................................................. 181
3.2.1 DOE's initiatives to reduce radioactive waste volume ................................................. 181
3.2.2 Initiatives to develop regulatory systems for dismantled waste from private nuclear
power plants ................................................................................................................ 182
3.3 Case examples in France ................................................................................................... 182
3.3.1 Background .................................................................................................................. 182
3.3.2 Movement toward law amendment for metal recycling ............................................... 183
3.4 Summary............................................................................................................................. 184
4. Prospects of processing/disposal method and technology related to its safe ......................... 185
4.1 Approach for volume reduction .......................................................................................... 185
4.2 Development of analytical/evaluation methods for efficient characterization .................... 186
4.2.1 Development of analytical methods for simple/quick data acquisition ........................ 186
4.2.2 Development of characterization methods using statistical methods ......................... 190
4.2.3 Summary ...................................................................................................................... 193
4.3 Development of methods for reasonably selecting safe processing/disposal methods .... 193
4.3.1 Clarification of issues and measures for storage and management ........................... 195
4.3.2 Clarification of processing technology with disposal in mind ...................................... 201
4.3.3 Establishment of methods to reasonably extract candidate technologies applicable topreceding processing .................................................................................................. 211
4.3.4 Consideration of disposal concept and development of safety assessment method . 216
4.3.5 Prospects of establishing processing/disposal strategies for solid waste for which
processing technology with consideration for disposal has not been clarified ........... 230
5. Issues based on Technical Prospects and technical strategies to achieve them ................... 236
5.1 Issues based on Technical Prospects ................................................................................ 236
5.2 Technology strategy by category ....................................................................................... 237
5.2.1 Characterization ........................................................................................................... 237
5.2.2 Storage ......................................................................................................................... 237
5.2.3 Processing/disposal ..................................................................................................... 238

173

1. Overview
Waste management is a long-term effort that must attain the prospect of the implementation of
final disposal, while reducing risks in every stage. Since a large amount of solid waste with differing
properties is generated in association with the decommissioning of the Fukushima Daiichi NPS133,
efforts are being carried out based on the basic concept of solid waste management summarized
in the Medium-and-Long-term roadmap for decommissioning of the Fukushima Daiichi Nuclear
Power Station of Tokyo Electric Power Company Holdings Inc. (hereinafter referred to as the“Medium-and-Long-term roadmap”). TEPCO is required to thoroughly implement safe andreasonable storage of the generated solid waste. With NDF taking the lead, the concernedorganizations are examining integrated management of solid waste from characterization toprocessing/disposal in a professional manner based on their respective roles. In addition to
improving analytical abilities for characterization, a flexible and reasonable waste stream(integrated flow of the measures from characterization to processing/disposal) is underdevelopment. Regarding processing/disposal strategies among these activities, the “TechnicalStrategic Plan 2020 for Decommissioning of the Fukushima Daiichi Nuclear Power Station of Tokyo
Electric Power Company Holdings, Inc.” (hereinafter referred to as the “Technical Strategic Plan”)
specifies that the prospects of processing/disposal method and technology for solid waste related
to its safety are to be provided by around FY 2021, and the summary results are shown below.

<Key points of "basic concepts on solid waste”>(1) Thorough containment and isolation
Thoroughly contain and isolate radioactive materials to prevent human access to them, inorder not to cause harmful radiation exposure.
(2) Reduction of solid waste volume
Reduce the amount of solid waste generated by decommissioning as much as possible.
(3) Promotion of characterization
Have appropriate characterization for addressing the increase in the number of analysis
samples for the purpose of advancing studies on processing/disposal methods of solid
waste.
(4) Thorough storage and management
Store generated solid waste safely and reasonably according to characteristics of the solid
waste.
Secure storage capacity to ensure that the waste can be stored within the site of the
Fukushima Daiichi NPS.
(5) Establishment of method for selecting the preceding processing methods in consideration of
disposal
Before the technical requirements of disposal are established, establish a method for

133 There is a possibility that some rubble generated after the accident may not be classified as waste orradioactive waste due to reuse on site. However, such rubble, secondary waste generated by water treatment,
and radioactive solid waste stored at the Fukushima Daiichi NPS since before the accident are hereinafterreferred to as “solid waste”.

174selecting processing for stabilization and immobilization (preceding processing) and thenselect preceding processing methods.
(6) Promotion of effective R&D with a comprehensive view of overall solid waste management
Work with each R&D field in characterization and processing/disposal to form an overview ofthe overall management of solid waste and then identify the required R&D tasks.
(7) Development of continuous operational framework
In order to continue safe and steady solid waste management, establish a continuous
operational framework that includes development of relevant facilities and human resources.
(8) Measures for reducing radiation exposure of workers
Thoroughly implement radiation exposure control, and health and safety management basedon the relevant laws/regulations.

175

2. Policy for developing Technical Prospects
The Medium-and-Long-term roadmap states that Technical Prospects will be provided in the strategicplan by around FY 2021. Specifically, it will “present how to proceed with volume reduction of solid waste”,“develop analytical/evaluation methods for efficient characterization”, and “establish a method forreasonably selecting safe processing/disposal methods for solid waste once necessary information
including characterization is obtained”.
For “volume reduction”, the basic policy for solid waste specifies that TEPCO should reduce the burdenof overall waste management (for the entire process from generation, storage, processing to disposal) tothe extent possible. Strategic plan 2020 considered the possibility of further efforts based on precedents
in other countries. Then, example cases for volume reduction of dismantled waste in other countries were
investigated (Chapter 3).
“Developing analytical/evaluation methods for efficient characterization” and “establishing a method forreasonably selecting safe processing/disposal methods for solid waste once the necessary information
including characterization is obtained” means to develop the methods necessary for disposing of materials
that become waste even after undergoing volume reduction. NDF has organized and examined thefollowing specific goals in strategic plan 2018. Of the wide variety of solid wastes, secondary waste
generated by water treatment, which has high mobility and for which there is no precedent for processing
or disposing of it in Japan, has been selected as the main object for examination.
(1) Establish a safe and reasonable disposal concept based on the characteristics and volume of thesolid waste generated in the Fukushima Daiichi NPS with its applicable processing technology, anddevelop safety assessment methods that apply the features of the disposal concept, withconsideration of examples from other countries.
(2) Clarify radiological analysis and evaluation methods for characterization.
(3) Clarify processing technology for which practical application can be expected for stabilization andimmobilization, with consideration of the disposal of several important waste streams such as
secondary waste generated by water treatment.
(4) Before the technical requirements for disposal are determined (i.e., preceding processing), establish
a method for reasonably selecting processing technology for stabilizing and immobilizing waste basedon the above methodology.
(5) Have the prospect of setting processing/disposal strategies for solid waste for which the processing
technology considering disposal is not clarified, using a series of methods to be developed by around
FY 2021.
(6) Clarify issues and measures concerning storage of solid waste until it is conditioned.
Based on these results, Chapter 4 provides Technical Prospects.

176

3. Example cases from other countries for volume reduction of
dismantled waste
In countries that have been promoting the use of nuclear power from early on, decommissioning offacilities and environmental remediation of sites are in progress for reasons such as aging and economic
efficiency. Along with dismantling and remediation work, a large volume of low-level waste has been or will
be generated in the near future.
A variety of management policies for radioactive waste are being implemented in various countries
according to their circumstances. This chapter provides case examples from three countries, the UK, US
and France, where waste management including volume reduction has been promoted for diverse andlarge-volume low-level radioactive waste (hereinafter referred to as “LLW”) generated by decommissioning
and environmental remediation activities especially at legacy sites related to nuclear development. The
scope of this chapter does not include high and medium-level radioactive waste whose generation islimited compared with LLW.

3.1 Case examples in the UK
3.1.1 BackgroundIn the UK, against a backdrop of issues such as an increase in the estimated volume of dismantled
waste associated with decommissioning of government-owned legacy sites and tight disposal capacity at
an existing low-level radioactive waste repository (hereinafter referred to as “LLWR”) located near Drigg,
Cumbria County, the government reviewed their previous policy and announced their long-term
management policy of solid low-level radioactive waste134 (hereinafter referred to as “LLW management
policy”) in 2007. With this reviewed policy, the government presented their basic approach to properly
account for the nature of the subject LLW, and allow flexibility in selecting cost-effective LLW management
policies that not only ensure safety but that are also environmentally acceptable.
In response to that, the Nuclear Decommissioning Authority (hereinafter referred to as the "NDA"), whichis responsible for the decommissioning of legacy sites and the management of radioactive waste,developed the low-level radioactive waste management strategy135 (hereinafter referred to as the “LLW
management strategy”) in 2010 to present strategic initiatives for considering possible flexible options forvolume reduction of waste and LLW management in accordance with the waste hierarchy and has
implemented specific measures.
3.1.2 LLW waste management policy based on the waste hierarchy concept
The LLW management strategy provides the management policy for LLW based on the concept of thewaste hierarchy (refer to Fig. A11-6) as described below.

134 Policy for the Long-Term Management of Solid Low Level Radioactive Waste in the United Kingdom, By Defra, DTI andthe Devolved Administrations, 26 March 2007.
135 UK Strategy for the Management of Solid Low Level Radioactive Waste from the Nuclear Industry, NDA, August 2010.

177

3.1.2.1 Reduction of waste generation
The LLW management strategy places highest priority on minimizing the generation of radioactive waste
in the waste hierarchy with the ultimate goal of reducing the volume of radioactive waste. For example, forfacilities, the strategy points out the necessity of considering the reduction in the volume of potential
radioactive waste through the lifecycle from designing to decommissioning. For existing contaminatedmaterials generated as a result of past activities, implementing consistent management in accordance withthe radioactive waste management plan, which was developed to manage the process up to the disposalof each type of radioactive waste at each site, is required to reduce the secondary waste that may begenerated in the course of management.
3.1.2.2 Minimization of waste volume
For radioactive waste whose generation cannot be avoided despite efforts to reduce it, the amount ofradioactivity and physical volume should be reduced to as low as reasonably practicable. Minimization of
the waste volume is promoted by three activities: classification and segregation according to theradioactive waste properties, etc.; reduction of radioactivity by decontamination of facilities before
decommissioning and of contaminated materials before transportation as radioactive waste; andcharacterization of radioactive waste for appropriate classification.
3.1.2.3 Re-use
In the LLW management strategy, some materials can be reused after maintenance is performed as
required, such as temporary storage containers of radioactive waste and lead-shielding bricks. Moreover,facilities, systems and buildings may still be worth reusing for other uses even after serving their initial
purpose, and therefore radioactive waste producers should explore utilizing such opportunities to theextent practicable.
The NDA operates a website for assisting the identification of potential users of facilities and systems
no longer in use for their initial purpose and the transfer of such facilities and systems to them, and thisinitiative has resulted in the reuse of a considerable number of NDA assets that are no longer needed, andhas even contributed to cost reduction.
Soil and rubble can also be reused as materials for site ground leveling and back-filling of voids if
appropriate permission by the regulatory authorities is obtained upon evaluation of the site in accordancewith the guidance of the environmental regulatory authorities described below.
3.1.2.4 Recycling
Further use of materials through recycling is seen as an important opportunity to reduce the volume ofradioactive waste. Metal is the main object of recycling, but other materials such as concrete and rubble
are also being recycled. Processing technologies such as surface decontamination to decontaminate thetarget objects, separation of radioactive materials and impurities by melting metals, crushing, andsegregation are used for recycling. Characterization, monitoring, and evaluation of substances are also
carried out so that they can be recycled effectively.

178

In the UK, metal recycling is an option that should be adopted from the perspective of resource
conservation, optimum utilization of disposal sites and environmental protection, etc. Blast
decontamination of surface contamination is employed in facilities in the UK, and melting processing is
employed in other countries (Sweden, US, Germany).
The cumulative volume of metal treated since the launch of the National Waste Program (hereinafterreferred to as the “NWP), which was established to implement the LLW management strategy, is 32,000
m3 as of the end of March, 2021 (Fig. A11-1).
3.1.2.5 Waste volume reduction
Volume reduction of radioactive waste is not positioned as a formal stage of the waste hierarchy, but inthe LLW management strategy, volume reduction and reforming of LLW, for which there is no option except
disposal, by compression or heat processing (incineration or other heat processing) plays an importantrole in the optimum utilization of disposal sites.
As of the end of March 2021, the cumulative volume of waste incinerated since the launch of the NWP
exceeds 27,000 m3 (Fig. A11-1).
3.1.2.6 Waste disposal
One of the objectives of the LLW management policy/strategy in the UK is to make LLWR usable into
the future. For LLW that can be subject to management policies other than disposal at an LLWR, it
encourages consideration of management options that account for risks, and takes the policy of avoiding
disposing of waste at LLWR to the extent possible.
As one of the measures, the government has established the category of very low level waste(hereinafter referred to as “VLLW”), and allowed disposal by controlled burial in industrial disposal facilities
if the environmental regulatory authorities approve it based on an environmental safety case evaluation.
Similar to VLLW, waste in the LLW category whose radioactive concentration is below a certain level (200Bq/g) can be disposed of if the environmental regulatory authorities approve it based on an environmental
safety case evaluation. At present, three private facilities dispose of VLLW, etc. As of the end of March,
2021 the cumulative volume of VLLW disposed of since the launch of the NWP exceeds 72,000 m3 (Fig.
A11-1).

Fig. A11-1 Cumulative volume of waste processed and disposed of VLLW since the launch of NWP
pursuant to the LLW management strategy136

136 March 2021 Waste Metrics Dashboard, Period 12: 21st February to 31st March FY20/21, LLW Repository Ltd.

179

In addition to exploring the possibilities of constructing a new LLW disposal site in or near the licensed
nuclear sites in the future, in the LLW management policy, a management policy in which contaminatedunderground structures and soil in licensed nuclear sites are not collected and instead consideration is
given to in-situ disposal (in-situ disposal) with permission by the environmental regulatory authorities basedon an evaluation is listed as a possible option.
In response to this, in 2018 the environmental regulatory authorities published a guidance document on
the management of radioactive waste generated by decommissioning and environmental remediation of
nuclear sites. The document calls for the establishment of safety cases covering the entire site, including
on-site disposal facilities, in-situ disposal, backfilling of underground cavities by contaminated rubble andcontaminated soil, etc., and the waste management planning which serves as the basis. It also specifies
that the safety cases and waste management plan are to be evaluated to determine whether or not toapprove the release of the site from the environmental regulations (Fig. ).
At three sites (Winfrith, Trawsfynydd, Dounreay), the NDA has been engaged in preliminary studies on
the possibility of in-situ disposal and backfilling, etc., from the viewpoint of controlling the generation of
new waste and promoting environmental remediation with economic reasonability while ensuring safety137.
Promoting such flexible policies for LLW management has significantly reduced the volume of LLW
disposed of in LLWR. As a result, there is the prospect that LLW to be generated from decommissioningof legacy sites can be managed with the remaining available capacity in LLWR.

Fig. A11-2 Possible disposal strategy options in a licensed nuclear site138 (left), and safety case concept
covering the entire site139 (right)

137 Strategy Effective from March 2021, NDA, March 2021.

138 Nuclear Decommissioning, Consultation on the Regulation of Nuclear Sites in the Final Stages of Decommissioning

and Clean-Up, BEIS, May 2018.

139 Management of radioactive waste from decommissioning of nuclear sites: Guidance on Requirements for Release fromRadioactive Substances Regulation, Version 1.0, SEPA, EA, NRW, July 2018.

180

3.2 Case examples in US
3.2.1 Background
In US, there are a number of national defense-related legacy sites that were established and operated
during the Cold War. The US Department of Energy (hereinafter referred to as the "DOE") is responsible
for the decommissioning, environmental remediation of these sites, and the resulting waste. It provides
site management in accordance with its regulations enforced by DOE orders under the agreement with
the US Environmental Protection Agency and state regulatory authorities.
However, in nuclear power plants owned and operated by private electric utilities, their operators arepromoting decommissioning in their sites that have ceased operation due to aging, etc. LLW generated isdisposed of at disposal sites under the responsibility of the state subject to the regulations developed by
the US Nuclear Regulatory Commission (hereinafter referred to as the “NRC”).
3.2.2 DOE's initiatives to reduce radioactive waste volume
In promoting the decommissioning and environmental remediation of legacy sites since 1989, the DOE
has provided waste management that includes reducing a large volume of radioactive waste generated ateach site. Especially for the volume of contaminated metal scrap generated from dismantling uranium
enrichment facilities at the Oak Ridge site, the DOE has supported and promoted metal recycling by privatecompanies. As a result, at its peak in 1995, 13,600 tons of contaminated carbon steel and stainless steel
per year were melted and recycled to the open market. In 2000, its release to the open market wassuspended, but metal recycling for effective use within DOE sites continues140.
At the Bear Creek processing facilities of EnergySolutions, one of the major metal processing facilities,
steel, lead, and other metals received from home and abroad are melted and recycled into shielding blocks
for use at DOE sites.

Fig. A11-3 Melting processing at EnergySolutions141, shielding blocks, and total volume of accepting
metal processing by country142

140 Waste Management for Decommissioning of Nuclear Power plants: An EPRI Decommissioning Program Report, EPRI,

Symposium on Recycling of Metals Arising from Operation and Decommissioning of Nuclear Facilities, April 2014.

141 DOE Promotes Internal Recycle and Reuse of Its Metals in a Big Way!, 1995InSite, Vol.26 GTS DURATEK.

142 Recycling and Reuse of Materials Arising from the Decommissioning of Nuclear Facilities, OECD/NEA, 2017.

181

3.2.3 Initiatives to develop regulatory systems for dismantled waste from private nuclear power
plantsIn US, there are considered to be limitations to the economic benefits of metal recycling LLW generated
by decommissioning private nuclear power plants because disposal costs are relatively low; large
equipment such as steam generator, etc. can be buried in shallow ground disposal sites for LLW; and aclearance system has not been introduced.
Since there is no category of radioactive waste in US that corresponds to VLLW, it is necessary todispose of waste equivalent to VLLW as Class A waste, which is one of the categories of LLW, at LLW
disposal sites. However, in terms of risk-informed regulations aimed at the flexible application ofregulations according to the degree of risk, the NRC permits VLLW-equivalent waste to be disposed of atstate facilities for hazardous waste disposal on a case-by-case basis143,144.
With the outlook of an increasing volume of low-level dismantled waste generated from planned
decommissioning of nuclear power plants (Fig. A11-4), the NRC is discussing introducing a new category
for VLLW and a clearance system in terms of safety for transportation as well as economic efficiency.

Fig. A11-4 Estimated LLW volume and cost at nuclear power plants in US(waste generated from decommissioning and operation) 145

3.3 Case examples in France
3.3.1 Background
In France, radioactive waste is classified according to the place of generation in nuclear facilities basedon the concept of zoning, and this concept is also applied to dismantled waste, etc. generated bydecommissioning and environmental remediation in nuclear sites. Since France does not have a clearance
143 10 CFR 20.2002 Method for obtaining approval of proposed disposal procedures.

144 Backgrounder on Disposals of Very Low-Level Waste Under 10 CFR 20.2002, NRC Website in US

145 EPRI Research Summary: Very Low Level Waste, US NRC’s 30th Annual Regulatory Information Conference, March

2018.

182

system, radioactive waste categorized as VLLW is disposed of at the VLLW disposal site established inMorvilliers in 2003 (hereinafter referred to as “CIRES”) (Refer to Fig. A11-5).

Fig. A11-5 Conceptual drawing of disposal at CIRES146 (left), and bird-eye view of the facility147 (right)
The French national radioactive waste management agency (hereinafter referred to as “ANDRA”), whichis responsible for the disposal of radioactive waste, estimates the volume of VLLW generated fromdecommissioning the existing nuclear facilities in France to be 2,200,000 m3. As of the end of FY 2014,
280,000 m3 of VLLW are placed in CIRES, which is approximately 43% of the permitted disposal capacity(650,000 m3), had been placed at CIRES, and it is necessary to establish a new VLLW disposal site or
expand its permitted disposal capacity in 2025148.
3.3.2 Movement toward law amendment for metal recycling
Under these circumstances described above, in “the national plan on management of radioactive
materials and waste (PNGMDR)” developed triennially, the Nuclear Safety Authority (hereinafter referred
to as “ASN”) as the nuclear regulatory authority requires waste producers such as ORANO and Électricitéde France (EDF) to examine conditions for recycling metal and rubble in light of volume reduction of VLLW
disposal, and to report technical/safety options for facilities for metal melting processing and management
policies.
In France, while the necessity of constructing a second VLLW disposal site (centralized or distributed)
is being considered, law amendment is under discussion for a proposal to allow recycling of some metals
contained in VLLW by clearance as an exception, if such recycling is considered to be appropriate.
Discussions are in progress toward ASN’s approval of using rubble as a filler for voids at the CIRES
disposal site.

146 The surface disposal concept for VLL waste, Andra.
147 Source : Google Map

148 French National Plan for the Management of Radioactive Materials and Waste (PNGMDR) 2016-2018, ASN.

183

3.4 Summary
As described above, handling of VLLW generated in large volume even from decommissioning of normal
nuclear power plants has become a common issue for the nations. While ensuring safety, various
measures have been implemented according to the situation in each country, such as the nature ofradioactive waste generators and disposal entities (public/private), prospects for securing disposal sites,and differences in regulatory systems concerning radioactive waste disposal.
Especially for recycling by metal processing, the UK and the US have a track record. Therefore, various
implementation conditions for processing characteristics, etc. should be investigated/examined.
In light of the differences in systems and approaches, and based on the waste hierarchy rules/principles,
feasible measures for decommissioning the Fukushima Daiichi NPS and policies that includes regulation
to realize it should be materialized.

184

4. Prospects of processing/disposal method and technology related
to its safety
4.1 Approach for volume reduction
If a large volume of solid waste exists, not only does segregation and analysis take time, but the
number of storage containers and the scale of storage facilities increase, and so does the load ofsolid waste management. Therefore, it is extremely important to reduce the physical volume of solid
waste as much as possible.
At the Fukushima Daiichi NPS, it is important to permeate initiatives on volume reduction, aimed
at decreasing the burden of overall solid waste management, into the overall decommissioning
activities by referring to examples from other countries that have implemented the waste hierarchy
concept as show in Chapter 3.
Specifically, the priorities for measures to be taken as waste management are (1) prevention ofwaste generation, (2) minimization of waste volume, (3) reuse, (4) recycling, and (5) disposal. In
solid waste management, it is important to prioritize (1) as much as possible, and consider (5)disposal as the last option (Fig. A11-6) for volume reduction of solid waste to be stored, processed,and disposed of.
In terms of reducing waste generation, it is important to consider the reduction of the volume ofmaterials used in the design and construction plan. It is also important not to bring in substancesthat affect processing/disposal as much as possible. To minimize physical volume, considerationon

preventing

contamination

throughstrictsegregation,

maintenance/managementofmanufactures, extension of product life, and waste volume reduction is important. Reuse should
be promoted after contamination checks, decontamination, repair and parts replacement, and it isuseful to consider ease of reuse from the design stage. Considering alternative uses is also
beneficial. In recycling, it is important to consider the state of contamination of contaminatedvaluable sources, separate and process recyclable materials, and use them as new materials andproducts.
As shown in Fig. A11-6, TEPCO has also been implementing initiatives corresponding to thisconcept. As new measures to be implemented, of the rubble accumulated outdoors (surface
radiation dose rate ≦ 0.1 mSv/h), the reuse/recycling of metals, concrete, and scrapped flanged
tanks with extremely low surface radiation dose rates is being considered. As part of this activity,
decontamination methods for recycling metals are being examined. In promoting safe andreasonable waste management, it is important to examine further possibilities based on thecharacteristics of solid waste at the Fukushima Daiichi NPS in reference to precedent cases inother countries.

185

Fig. A11-6 Summary of waste hierarchy at the NDA, UK, and countermeasures at the Fukushima
Daiichi NPS

4.2 Development of analytical/evaluation methods for implementing efficientcharacterization
Efficient characterization is necessary as solid waste at the Fukushima Daiichi NPS ischaracterized by diverse nuclide compositions and activity concentrations, and a large physicalvolume. With the specific goal, “(2) Clarify radiological analysis and evaluation methods forcharacterization” described in Chapter 2, analytical methods for simplified and speed-up dataacquisition, as well as characterization methods using statistical methods have been developed in
the Project of Decommissioning and Contaminated Water Management149 and other programs. For
the latter, a statistical inventory estimation method was developed by incorporating a method for
efficiently identifying inventory, which combines the nuclide relocation models based on thecontamination mechanism with the analytical data, with a method for quantifying uncertainties in
evaluated values by using statistical methods.
4.2.1 Development of analytical methods for simplified and speed-up data acquisition
The “basic concept for solid waste management” indicates that appropriate characterization
responding to an increase in the number of analytical samples should be promoted to proceed withexamination for solid waste processing/disposal. Therefore, the subject waste and nuclides to beevaluated in the characterization were clarified, and efficient analytical methods with the required

149 IRID, supplementary budget in FY 2018, “Subsidies for the Project of Decommissioning and Contaminated Water

Management (R&D for Treatment and Disposal of Solid Waste), FY 2019 results, December
2020https://irid.or.jp/wp-content/uploads/2021/01/2019011kotaihaikibutsu_02.pdf
IRID, supplementary budget in FY 2018, “Subsidies for the Project of Decommissioning and Contaminated Water
Management (R&D for Treatment and Disposal of Solid Waste), FY 2020 results (in preparation for publication)186

measurement accuracy and the associated manual were developed. Then, the possible issues
were examined to incorporate them into analysis in the Radioactive Material Analysis and Research
Facility Building #1 (hereinafter referred to as “Okuma Building #1).
Solid waste generated from the decommissioning of the Fukushima Daiichi NPS undergoes
primary, secondary, and tertiary classification (Fig. A11-7), and is temporarily stored, stored, andconditioned according to its classification. In characterization, the physical and chemical properties(particle size distribution, viscosity, quality of materials, surface radiation dose rate, nuclide
concentration, etc.) of waste required for proper classification of generated waste are identified.

Fig. A11-7 Characterization according to waste classification at the Fukushima Daiichi NPS149
In terms of reducing risks in decommissioning the Fukushima Daiichi NPS, the types of waste
subject to characterization were selected from rubble, incineration ash, soil, waste generated fromfuel debris retrieval (hereinafter referred to as “waste from debris retrieval”), and secondary waste
generated by water treatment and other substances with different properties based on the physicalvolume, radiation dose, nuclide composition and disposal characterization, etc. As subjects(nuclides to be evaluated) for analyzing their nuclide concentration of waste properties, short halflife nuclides have been excluded from those important for radiation protection, and those importantfor domestic disposal safety assessment have been selected. For nuclides whose concentration
can be estimated based on fuel combustion calculation in terms of isotopes and chemical similarity,
nuclides to be examined and analyzed have been summarized, taking the importance of safety indisposing of waste generated from the accident into account.
The progress of characterization of the target waste is as follows. Of the secondary waste
generated by water treatment, characterization of high-risk sludge generated by decontamination
devices and concentrated waste liquid with the aim of secondary classification for volume reduction,
stabilization and storage has been completed. The sludge generated by decontamination devices
is being examined with the aim of retrieval from the process building and storage based on the187

information obtained from characterization. Characterization of ALPS slurry is underway forsecondary classification. For rubble, analysis of the nuclide composition of samples (111 samples
as of March end, 2020) taken for tertiary classification as well as examination of statistical inventory
evaluation methods are being carried out. Upon the completion of the Okuma Building #1, the data
will be accumulated in order to develop disposal validation methods for the tertiary classification,and the application of a scaling factor method will be considered.

Fig.A11-8 Examination of nuclide composition of rubble (reactor building) (correlation with Cs-137
concentration)149

4.2.1.1 Technology development for automating and accelerating characterization
4.2.1.1.1 Development of reasonable sampling technology for analytical samples
In order to efficiently collect a small number of analytical samples from rubble sampled on sitewhile ensuring representativeness, technology was developed that evaluates the contamination
distribution of the collected rubble (sample mapping) and collects a small number of analyticalsamples from the positions selected according to the contamination distribution (sampling), and its
mock-up machine was manufactured. Based on a performance evaluation test, areas for
improvement with a view to practical application were summarized.
4.2.1.1.2 Development of analytical methods using ICP-MS/MSIn determining the quantity of nuclides by measuring β-rays and characteristic X-rays, a timeconsuming and precise chemical separation process is required in advance. It takes a long time
to measure long half-life nuclides with low specific activity. However, analysis using ICP-MS/MS(triple quadrupole inductively coupled plasma mass spectrometry) enables the quantity of multiple
nuclides to be determined simultaneously because the substances can be sorted by mass. As
separation is possible by using reactant gas even if nuclides (isobar) of the same weight coexist ina sample, the separation process can be economically streamlined. In order to apply ICP-MS/MS
to the analysis of rubble in the Okuma Building #1, automation technology was developed for aseparation process using ICP-MS/MS and analytical samples.
The separation process using ICP-MS/MS was developed for difficult-to-measure nuclides (Zr93, Mo-93, Pd-107, Sn-126) contained in rubble. An automatic solid-phase extraction device was188developed, which removes the sample matrix from an analytical sample, and separates isobars by
one-step extraction chromatography (Fig. A11-9a). With the use of NH3 gas as reactant gas formeasuring by ICP-MS/MS, it was confirmed that Zr-93 (Fig. A11-9b) can be efficiently separated
from the isobars (Nb-93, Mo-93), Ag-107 from the isobar (Pd-107), and Te-126 from the isobar (Sn126).
The workload of analytical technicians is high for chemical separation of analytical samples,causing errors in the recovery rate of the nuclides depending on the analytical technician. Therefore,
an automatic solid-phase extraction device with small errors in the recovery rate of the nuclides
was developed for solid-phase extraction separation of Zr-93, Mo-93, Pd-107, Sn-126. The
separation process of Zr-93 and Mo-93, which are difficult-to-measure nuclides, was applied to theautomatic solid-phase extraction device. As a result of optimizing the dipping rate in each step, therecovery rate of Zr and Mo, and the mix rate of the isobar (Nb-93) were almost the same as thoseof analytical technicians. In addition, the working time using the automated solid-phase extraction
system was reduced to 60 to 70 minutes (120 min by analytical technician).

Fig. A11-9 Examination of analytical methods using ICP-MS/MS149
4.2.1.1.3 Development of a chemically reasonable separation process for radiometry
For β-rays and characteristic X-rays (Cl-36, Ca-41, Ni-63, Sr-90), for which an analytical method
using ICP-MS/MS is not appropriate in terms of detection sensitivity, etc., radiometry is used foranalysis. Among these nuclides, a process for Ca-41, Ni-63 and Sr-90, which have the potential toeconomically streamline a series of separation processes, was developed. As a result of applying
this process to the mock rubble, a good recovery rate was obtained. It has high separation
performance for Cs-137 and Sr-90, which coexist in large quantities in solid waste, and the working
time is 30 hours (half the conventional separation process). It was also found that the amount ofhighly corrosive HCl used was greatly reduced, resulting in reducing the load on the system.

189

4.2.1.1.4 Development of a standard analytical method
Analytical samples from solid waste vary in shape and chemical composition, requiring validationof analytical methods for multiple nuclides for each analytical sample according to the nuclide.
Therefore, requirements by analytical step (dissolution, separation and measurement) wereclarified, and then a standard analytical method was developed for validating all the analyticalmethods that can handle different nuclides by verifying each analytical step. The validation
conditions of the standard analysis method using ICP-MS/MS were examined for mock concrete toprovide the specific evaluation items (e.g., repeatability, influence of different sample properties,
evaluation of decontamination factors for interfering radionuclides, influence of measurement
sample matrix, etc.) by analytical procedure. As a result of validating the analytical flows for Zr93/Mo-93, Pd-107/Sn-126 and Ca-41/Ni-63/Sr-90 analyzed by radiometry, it was confirmed that theconditions qualified for the standard analysis method were satisfied. In addition, as a result of
evaluating the lower detection limit in a method using mock concrete in each analysis flow, it was
confirmed that the value was considerably lower than the upper limit of the concentration in thecase of trench disposal. Based on the above results, a quality assurance manual (draft) was
developed that takes into account validation of the standard analytical method using ICP-MS/MS
and radiometry in accordance with the quality management system.
4.2.2 Development of characterization using statistical methods
4.2.2.1 Establishment of statistical inventory estimation methods
4.2.2.1.1 Examination of a waste contamination mechanism at the Fukushima Daiichi NPS(1F)
With an attempt to estimate the inventory necessary for waste classification, the focus wasplaced on air and contaminated water as nuclide migration media, and to divide into source term,
migration process and immobilization of contamination (Fig. A11-10). Then, in terms of importanceof waste classification, the contamination mechanism was examined for Unit 2 for which there is alarge amount of analytical data on dismantled waste in its reactor building.
As a result of evaluating the transport ratio (the concentration ratio to Cs -137 was standardized
by fuel composition), where the elements with similar chemical properties show similar values, it
was estimated that the source term released through the air was H-3 and C-14 for damaged
cladding tubes, other FP and TRU nuclides for damaged fuel, and Co-60 and Ni-63 for activatedproducts. Since the radiation dose rate near the shield plug was high and contamination was
significant, and the Cs-137 concentration in the ceiling was more than 2 orders of magnitude higher
than that on the floor, it was presumed that the migration process through the air was diffusion fromthe shield plug which forms the boundary with the PCV. The transport ratios of major nuclides in
the ceiling, wall, and rooftop were similar, and the nuclide composition did not change during spatial
migration. In addition, the concentration of rubble on the east side was higher than that on the west
and north sides of the rooftop. Therefore, the implication is the diffusion of the contaminated air
due to the falling of the east-wall blowout panel. Imaging plate measurement confirmed that190

contaminants that traveled through the air and affixed to waste was spot contamination which would
not be likely to penetrate into concrete or metal.

Fig. A11-10 Outline of examining a contamination mechanism of waste at 1F149
4.2.2.1.2 Examination of reducing uncertainty in nuclide migration parameters
With in-core fuel and structures in the buildings (activation products) as a source term based on
the contamination mechanism examined, a nuclide relocation model was developed to calculate
the migration process in which nuclides released into the air inside the buildings affix to waste
through the air, and the other nuclide migration process in which nuclides migrate from fuel debris
and affix to sorption vessels through the stagnant water (Fig. A11-11). For nuclide migration
parameters such as the rate of release to outside the buildings, the distribution of the transport
ratios was analyzed by Bayesian statistics based on the analytical data of migration to the stagnant
water and to inside/outside the buildings. Then, by expressing the uncertainty of the parameters(mean and standard deviation) quantitatively, the migration ratio was set by the probability
distribution (lognormal distribution). The accuracy of the parameters was improved by accumulating
analytical data. Some analytical data on ALPS-sorbent was obtained and it demonstrated a good
match with the values estimated by the statistical inventory method.

191

Fig. A11-11 Radionuclide migration model to calculate migration to waste149

4.2.2.2 Consideration of analysis planning using statistical methods
In reference to the Data Quality Objectives (hereinafter referred to as the "DQO") process150
used by the National Nuclear Laboratory in the UK, analytical planning methods (combination of
the DQO process and Bayesian estimation method), which evaluates the number of analyticalsamples required for the goal established by statistical methods according to waste properties,
were tested for sludge generated by decontamination devices and dismantled waste (concrete). Ina trial run for the sludge generated by decontamination devices, it was assumed that the sludge
was disposed of after dehydration processing, and sampled during this processing. As a result of
calculating the probability of remaining below the standard concentration assumed at the time ofdisposal for Co-60, Sr-90, and Cs-137, it was presumed that the assumed criteria were met at aprobability of 90% or higher by analysis of four or more samples.
As described above, based on the statistical inventory estimates and the safety assessment on
disposal, the objective of analysis could be specified by waste to obtain the number of analysis
points, and the effectiveness of the analysis planning method using statistical methods was
confirmed.
4.2.2.3 Establishment of an analysis database management system
An analysis database management system called FRAnDLi (Fukushima Daiichi Radwaste
Analytical Data Library) has been developed, which contains and discloses the analytical data on
solid waste obtained from the Project of Decommissioning and Contaminated Water Management
150 DQO process: Process developed by the US Environmental Protection Agency (US EPA) for planning thesampling of analytical samples for decision-making. A sampling plan will be developed in seven steps.
Planning and analysis are repeated as necessary to achieve the objectives. This process can be applied tovarious problems that can be solved by analysis.

192

as well as the information on waste-related analytical data provided by TEPCO (sample information(type, sampling location, date/time, etc.), analytical values of radioactivity concentration)). The
system is utilized to review analysis strategies, and is widely used with over 1000 monthly accesses.
4.2.3 Summary
The following results were obtained from activities aimed at “clarifying radiological analysis andevaluation methods for characterization” as a specific goal for providing Technical Prospects.
An efficient analytical method with the required analytical accuracy was established. The
established analysis method is expected to be made into a manual in FY 2021.



The technical applicability of the analysis methods for automation/acceleration (samplingtechnique for analytical samples, analytical method using ICP-MS/MS, automation technology
for chemical separation, separation process for radiation measurement) was confirmed, andR&D results will be incorporated into Okuma Building #1.

FRAnDLi that contains information on solid waste properties has been developed, leading toestablishment of a statistical inventory estimation method using the analytical data collected.
In addition, examination of an analysis plan according to the collected waste properties, etc.
became possible.
As described above, analytical and evaluation methods have been developed for efficientcharacterization, including the development of methods for segregating and reasonably analyzing
substances by mass, the rationalization and automation of separation processes, and thedevelopment of statistical methods for estimating the inventory of solid waste with less analytical
data. In the future, the inventory estimation of solid waste necessary for disposal safety assessment,
etc. will be promoted by the statistical inventory estimation method, while reducing the uncertainty
in nuclide migration parameters, etc. by sample analysis using the developed analytical methods.
For solid waste that can be adequately sampled and evaluated, the statistical inventory estimates
will be replaced with the inventory calculated from analytical data.

4.3 Establishment of methods to reasonably select safe processing/disposal
methods for solid waste
In selecting processing/disposal methods in a reasonable manner, based on the waste properties,
an appropriate combination of processing (waste form) and disposal (disposal facility) methods
should be clarified so that the risk of buried solid waste to the public and environment can be
maintained sufficiently low in the future.
In the case of solid waste from a normal reactor, its properties can be estimated to some extent
by the previous findings (data) and analytical methods. Accordingly, the appropriate combination ofprocessing (waste form) and disposal (disposal facility) methods can sufficiently reduce the risk to
avoid a significant impact on the public and surrounding environment.
Even in the case of solid waste from the Fukushima Daiichi NPS, molten nuclear fuel is a major
source of contamination, and the radioactivity concentration does not exceed that of spent fuel.

193

Therefore, the risk can be sufficiently reduced by understanding the overall picture of the target
solid waste (properties such as nuclide composition, activity concentration by waste, waste volume),
and selecting a proper combination of processing (waste form) and disposal (disposal facility)
methods, while utilizing the experience and knowledge on radioactive waste processing/disposal
accumulated in domestic and overseas.
However, the overall picture of solid waste to be disposed of, including that which will be
generated, will gradually become clear as the progress and plans for fuel debris retrieval,contaminated water management, and other decommissioning work are clarified. Therefore, it isnecessary to repeatedly examine processing/disposal methods and safety assessments, startingfrom the waste for which properties have been clarified; to give consideration to making
processing/disposal methods more appropriate; and to accumulate knowledge to consider safe andreasonable processing/disposal methods for diverse solid waste collectively. Aiming for safer andnot extremely conservative storage of waste with high mobility such as slurry waste, processing(preceding processing) for stabilization/immobilization may be required before determining thedisposal method (disposal facility). Reprocessing would be necessary if the specifications of thewaste form even after preceding processing did not conform to those required by the disposal
method (disposal facility) to be determined. Therefore, in order to minimize such a possibility, aselection method for the preceding processing method with disposal in mind is needed.
As mentioned below, study on an appropriate combination of processing and disposal methods,
or preceding processing methods, is considered for the waste for which properties have been
identified to some extent.


Establish several feasible disposal methods suitable for waste characteristics (without
specifying the feature of facilities such as their locations and sizes)



In parallel, establish several processing methods suitable for waste characteristics to beconsidered, and set the specifications of the waste form after applying each processing
method.



Evaluate the safety of several selected disposal methods based on the specifications of thewaste form after processing to verify whether risk to the public and environment is sufficiently
low, and to consider more effective processing/disposal methods based on the evaluation
results.

The above examination steps are repeated to narrow down disposal methods and specifications
for the waste form after processing. Clarifying the overall characteristics of solid waste concurrently
with characterization helps identifying an appropriate combination of processing/disposal. When
preceding processing becomes necessary, candidate processing methods will be selected inconsideration of the status of examination and open issues at that point.
It is also important to consider the period during which pre-disposal management is to be
implemented, taking into account the risk reduction during that period, and examining necessary
and feasible technologies. Since storage is important to provide flexibility to respond to the progressof processing/disposal, and to reduce radiation exposure of workers due to the decay of the194radioactive materials, it is important to consider storage strategies as part of this examination
process.
A series of these studies is represented as a flowchart shown in Fig. A11-12. In establishing themethods as a series of methods for selecting processing/disposal methods in a reasonable manner,technical knowledge and evaluation methods necessary for these studies (establishment ofprocessing technology and conditioned waste specifications appropriate for waste, safe,
reasonable and feasible disposal methods, disposal safety assessment) have been developed
through research/development (verification of applicability of processing methods using
engineering-scale test equipment conducted mainly on the secondary waste generated by water
treatment, establishment of safe, reasonable and feasible disposal methods based on the waste
properties and applicable processing technologies, and development of safety assessment
methods) by the Project of Decommissioning and Contaminated Water Management149.
The results of these studies are shown below.

Generation ofwaste

Pretreatment andtreatment

Near-term storageand management

・ Improve accuracy of data on
properties, for disposal safetyassessment
・ Acquire additional data, review
migration models
・ Quantify factors interfering with
waste conditioning

・ Based on processing
schedules, update plans forrequired timing, capacity, etc.
of storage facilities
・ Update waste classifications
・ Change combinations of waste
to be treated

Disposal

Waste conditioning

・ Review waste specification
proposals to meet disposal
requirement proposals
・ Identify technologies that can beapplied・ Improve quality control
processes

・ Review disposal methods(design of disposal facilities) andassessment methods(scenarios, models, assessment
parameters)

Disposalrequirement
proposals

Policy fornarrowingdown

・ Specify disposal
technology suitable for
waste
・ Designate waste
specifications

Disposal・ Specify reasonable andfeasible disposal methods
・ Conduct disposal safetyassessments

Proposals forprocessing/disposal options

・ Manage risk through monitoringand surveillance・ Consider risk reduction
measures as necessary

Processing

Waste
specification proposals

Storage

Stablestorage・ Set inventories

Property
data

Characterization

Reflect in TEPCO and solid
waste storage management
plans

Fig.A11-12 Development of methods for reasonably selecting safe processing/disposal
methods of solid waste

4.3.1 Clarification of issues and measures for storage and management
As some solid waste materials are high-level waste that contain water or to which water isadhered, the hydrogen generation by radiolysis is of concern during storage. Therefore, regarding
the storage of high-level waste for which consideration of hydrogen management is required, issues
and measures with respect to the approach to safety were examined in the Project ofDecommissioning and Contaminated Water Management and other programs.
High-level waste includes waste generated by fuel debris retrieval (hereinafter referred to as“waste from fuel debris retrieval”) and secondary waste generated by water treatment. The

195

secondary waste generated by water treatment is temporarily stored in box culverts, etc., but
storage facilities will be established according to TEPCO’s storage plan to eliminate temporarystorage as much as possible. For this purpose, the Project mainly targets the waste from fuel debris
retrieval for conducting the following studies on storage. The waste from fuel debris retrieval was
temporarily divided into removed objects and other waste (that retains its original shape and has
no possibility of criticality). The removed objects were further divided into those above the fuelloading position and the external structures of the pedestal.
4.3.1.1 Safety function requirements during storage of high-level waste
As a result of examining hazards from major accidents and safety function requirements
anticipated during storage of high-level waste (containment, transfer and storage), it wasdetermined that the most important safety function requirement was prevention of dispersion ofradioactive materials (ensuring airtightness) and hydrogen release (ensuring air permeability).
Survey results of domestic and overseas storage cases of high-level waste showed that Zion and
TMI-2 nuclear power plants in US used dry storage, Paks-2 nuclear power plants in Hungary andthose in Japan (shroud replacement) used wet storage. The TMI-2 and Paks-2 nuclear power
plants used storage containers with filter vents. Since wet storage requires securing a pool forstorage and pool water management, dry storage was selected for high-level waste and examined
for storage containers with filter vents.
4.3.1.2 Examination of high-level waste storage methods
4.3.1.2.1 Concept of high-level waste storage
It is assumed that the high-level waste removed from the reactor buildings is stored in innercontainers in the reactor buildings, storage containers in storage container handling cells in the
expanded building, and then in shipping casks in delivery cells (Fig. A11-13). The assumption isthat the shipping casks are transferred to the temporary storage facility on site, and the waste istaken out from the shipping casks for drying in the drying area, and then brought into a dry storage
condition in the storage area.

Fig. A11-13 Example flow of handling high-level waste149
Assuming the timing of implementing hydrogen management for high-level waste, i.e., during onsite transportation and at the time of temporary storage, consideration was given to hydrogen
management for each case. Possible hydrogen management during on-site transportation include

196

installation of vent filters in storage containers, sealing of shipping casks and installation of catalysts
inside shipping casks. The vent filter discharges hydrogen generated in the storage container tothe shipping cask and maintains the hydrogen concentration inside the shipping cask below thelower limit of explosion (< 4 vol%) to prevent dispersion of radioactive materials. The shipping cask
is provided with a sufficient void space to maintain the hydrogen concentration below the lower limitof explosion during on-site transportation.
Hydrogen management during temporary storage includes vent filters installed in storage
containers during transportation and a ventilation system in temporary storage facilities. The ventfilter discharges hydrogen generated in the storage container to the temporary storage facility andmaintains the hydrogen concentration inside the storage container below the lower limit of
explosion to prevent dispersion of radioactive materials. The hydrogen concentration inside thetemporary storage facility should be maintained below the lower limit of explosion with theventilation system. As a measure for preventing corrosion of storage containers, drying treatment
is applied to free water adhered to the surface of high-level waste before temporary storage.
4.3.1.3 Examination of container and storage facility requirements
4.3.1.3.1 Preconditions for the properties and storage of high-level waste
Although different types of high-level waste to be retrieved along with debris retrieval are taken
into consideration, it is assumed that the maximum radiation dose of high-level waste to be
retrieved eventually is 400 Sv/h for more than 1,500 tons of contaminated metal and concrete, andits surface density (in the reactor well) is 1.2 × 108 Bq/cm2. It is assumed that high-level waste,
except for some large-sized waste, is cut and stored in inner containers. It is assumed that storage
containers, each containing an inner container, are stored in shipping casks, and transported totemporary storage facilities on site. It is also assumed that the storage containers are then removed
and stored in temporary storage facilities.
The surface radiation dose of the shipping cask during on-site transportation is assumed to be <30 mSv/h in reference to on-site transportation at the Fukushima Daiichi NPS, and the shipping
cask is assumed to be spill-resistant in terms of prevention of spread of α-contamination. As theconditions for acceptance in the temporary storage facilities, it is assumed that the surface radiationdose of the storage container is set to < 10 Sv/h and the weight (including shielding and contents)
is set to < 7.5 tons in reference to the solid waste storage facility building #9. Based on theproperties of high-level waste and conditions for on-site transportation and storage in temporarystorage facilities, the provisional dose classification of high-level waste, requirements ofacceptance in temporary storage facilities and shielding of the storage container are assumed as
the basis for considering storage methods.
4.3.1.3.2 Examination of reasonable container shape
The shapes of the inner container, storage container (storing an inner container) and shipping
cask were examined based on the required functions for high-level waste containers (shape,

197

material, containment and shielding performance, hydrogen management, operability, structural
strength). It is assumed that the inner container will have a mesh structure for draining the removed
objects. Depending on the debris retrieval method, large and small inner containers will be prepared.
It is also assumed that one large inner container or several small inner containers can be stored in
one storage container.
Storage containers (1.42 x 1.42 x 1.00 mm) were examined from the perspective of handlingsuch as its surface radiation dose and weight including waste. According to the radiation dose levelof the high-level waste, four types of shielding thicknesses were considered so that the surface
radiation dose rate of the storage container could be maintained at < 10 Sv/h. The total weight wasconsidered so as not to exceed the weight limit of the forklift.
The shipping cask is assumed to have a sealed structure to suppress the release of radioactive
materials during transportation.
4.3.1.3.3 Examination of hydrogen management
Vent filters should be installed in storage containers to maintain the hydrogen concentration in
the storage container below 4 vol%. The shipping cask should be spill-resistant to prevent the
spread of contamination during on-site transportation, and its shape will be determined from thevoid ratio, etc. for hydrogen management. The essential functions required for vent filters can bedivided into the aspects of safety and handleability. In terms of safety, the functions required forvent filters include those related to containment, criticality, shielding and hydrogen explosion. In
terms of handleability, however, those related to remote operability, drying processing andcondensation during storage are required. The essential functions for vent filters and theirrequirements were summarized in terms of safety and handleability. In terms of safety, the functions
required for vent filters include those related to containment, criticality, shielding and hydrogen
explosion. The essential functions for vent filters in terms of handleability are remote operability,
drying treatment, and dew condensation during storage. At WIPP disposal sites in the US, the air
flow rate, collection efficiency and hydrogen diffusion performance are set for each container as
the minimum specifications for the vent filter as requirements for accepting waste containers.
4.3.1.3.4 Examination of drying treatment
Drying treatment conditions (water adhered to waste, drying area) and drying treatment methods,
and nozzle installation and air supply/exhaust methods were examined. Based on the survey
results of existing drying technology, requirements of drying treatment systems (applicability,
treatment for temporary storage, applicability to the retrieval flow), influence during drying treatment,
technical requirements for drying, container-by-container drying treatment and operating indexes
were examined and summarized.
In order to prevent contamination in temporary storage facilities, storage container lids shall not
be opened. Therefore, a drying treatment method for high-level waste in which a nozzle is installedon the storage container and connected to a drying device was assumed. After completion of the198drying processing, the nozzle portion may be contaminated, and thus its decontamination isrequired. Drying processing methods include the convection (static) method and conduction (static)
method. With the convection (static) method with connected nozzle, a blower and a heater are
installed on the air supply side. Hot air (dehumidified air or dry nitrogen) is supplied to heat thehigh-level waste, and exhausted by the blower on the exhaust side. With the conduction (static)
method with connected nozzle, the bottom and side surface of the storage container are heated bya heater, and exhaust is exhausted out by a blower on the exhaust side or vacuum pump. A hot air
supply nozzle was installed on the body of the storage container, and the blow-out end was installedon the lower part of the container. The hot air supply provides dry air or dry nitrogen, and exhaust
methods include normal pressure exhaust and depressurized exhaust.
4.3.1.4 Examination of measurement methods during high-level waste storage
Measurement and inspection items necessary during high-level waste storage were examined.
Confirmation items that may be measured and inspected during storage of high-level waste(hereinafter referred to as "measurement/inspection items") depend on the timing in the handling
flow of high-level waste.
Measurement/inspection items during storage in inner containers are assumed to be the basic
properties of waste, physical properties of waste, radiochemical properties, calorific properties, andhydrogen generation characteristics. Measurement/inspection items during storage in storage
containers are assumed to be appearance, weight, filling rate, surface radiation dose rate, surface
contamination, surface temperature, hydrogen generation rate, and nuclide release volume.
Measurement/inspection items during storage in shipping casks are assumed to be appearance,
weight, surface radiation dose rate, surface contamination, surface temperature, and sealing.
The priority of the measurement/inspection items was examined in view of the purpose ofmeasurement, difficulty of measurement, estimability from measured values, avoidability byoperation, evaluation and prior investigation. The priority measurement and inspection items are
considered to be weight, surface contamination, and surface radiation dose. The priority
measurement/inspection items when estimation by the surface radiation dose is not feasible areconsidered to be calorific values, surface temperature, radioactivity, amount of nuclear materials,
and hydrogen generation rate. Measurement/inspection items and timing in the handling flow ofhigh-level waste were examined (Fig. A11-14). It is assumed that high-level waste is separated
from fuel debris when it is removed from reactor buildings.

199

Fig. A11-14 Examination example on measurement/inspection items and timing in the handling
flow of high-level waste149

4.3.1.5 Examination on long-term storage of high-level waste
Defects and countermeasures assumed during prolonged temporary storage of high-level waste,other than hydrogen management, were examined in reference to countermeasures in Japan andabroad. The following defects are assumed during prolonged storage: A. Container breach due tocorrosion, B. Prolonged generation of hydrogen, C. Occurrence of external events, and D.
Deterioration of consumables. Possible measures include use of corrosion-resistant materials,
drying of waste, temperature and humidity control for A; solidification process, gas venting,installation of hydrogen storage materials, etc., and monitoring of gas generation for B; protectionof structures, fall-prevention, and use of high-strength containers for C; and monitoring/inspection,
reduction of consumables, and securing of containment performance for D. Since waste other thanhigh-level waste is stored for a long period of time, further investigation of the actual problems andcountermeasures is planned in FY 2021.
During prolonged storage of high-level waste in storage containers, it is anticipated that thehydrogen diffusion performance will not be satisfied due to corrosion or deterioration of the ventfilter, and that the contents will tend to scatter. As a countermeasure, it is necessary to verify the
progression of corrosion by changing to corrosion-resistant materials and using test pieces. It isnecessary to maintain the vent filter functions for prolonged storage of high-level waste, therefore,
degradation factors of the vent filter will be investigated in FY 2021, and methods for verifying thevent filter functions will be examined..
4.3.1.6 Summary
As a result of the above activities, clarification has been achieved as described below on theissues and measures concerning storage and management of solid waste before being conditioned.
 By examining storage methods, the handling flow up to temporary storage of high-level waste
and the timing for implementing hydrogen management have been examined. Hereafter, the200handling flow from temporary storage to disposal should be considered. It has become possible
to consider hydrogen management, but it is still necessary to examine the evaluation conditionsof the hydrogen generation rate and validate the evaluation results.
 By examining the requirements of containers and storage facilities, the containers and drying
processing systems required for hydrogen management were considered. Verification of thedrying processing system and its drying performance is required. In accordance withexamination of the fuel debris retrieval method, it is necessary to review the shape of thecontainer, shielding properties, site for applying drying treatment, measurement items andtiming, etc.
 The possibility of the occurrence of an event of concern other than hydrogen management,such as leakage and dispersion due to corrosion of storage containers, has been evaluatedand measures have been proposed. Since it is important to maintain the vent filter functions forthe prolonged storage of high-level waste, the degradation factors of the vent filter will be
investigated in FY 2021, and methods for verifying corrosion and degradation of the vent filter
will be examined.
4.3.2 Clarification of processing technology with disposal in mind
The following efforts have been made with the goal of “clarifying processing technology which
practical application could be expected for stabilization and immobilization, considering disposal
for several important waste streams such as secondary waste generated by water treatment”.
Of the solid waste, processing technology with the prospect of practical application will be
extracted at least for ALPS slurry and sludge generated by decontamination devices.



The prospect of not only the initial performance of waste packages but also the long-term
stability after disposal based on the mechanism will be shown.
Then, processing technologies with possible applicability were investigated, taking into accountthe types and properties of secondary waste generated by water treatment. Of the technologies
already in practical use in Japan and abroad, for those that have reached the level of practicalapplication and those under research/development, the applicability of low-temperaturesolidification processing technology (cement, AAM (alkali-activated materials)) and hightemperature solidification processing technology (CCIM (cold crucible induction melting), In-Can,
GeoMelt ICV) to secondary waste generated by water treatment was examined (Table A11-1).

201

Table A11-1 Processing technology overview149,151,152,153
Classification

Technology Name

Overview

Low-temperatureprocessing

Cementsolidification

--

OPC

Solidification of waste by utilizing the setting properties of cement hydration reactions. Includes in-drum and out-drum
methods.

AAM
solidification--

AAM solidification

In AAM, a high-concentration alkali solution (mixture of NaOH, KOH, and sodium silicate) is added to an inorganic powder(metakaolin, etc.) containing an alumina silicate, and is solidified by the dissolution of raw materials and polycondensation
reactions. The solidified waste has an amorphous network structure.

High-temperature processing

Vitrification

Downflow Cold crucible induction Glass frit is melted by ignition metals, and waste and glass raw materials are heated by high frequency induction. The molten
methodmelting (CCIM)
materials flow downward and are solidified by cooling them in vessels. The mixture of molten materials is accelerated bymethods such as bubbling. By water-cooling the furnace wall, its corrosion is suppressed by a skull layer formed between themolten materials and the furnace wall. Organic waste is thermally decomposed by supplying oxygen from the upper part of
the furnace.
InCan
methodJoule heating(GeoMelt)

Waste and glass raw materials are supplied into a container and an electric current is applied between graphite electrodes togradually melt the materials in the vicinity of the electrodes. After melting, they are cooled in containers and solidified. The
molten materials are agitated by thermal convection.

InCan
methodExternal heating(InCan)Waste and glass raw materials are supplied into a container, which is heated by an external heater. After the materials aremelted, they are cooled in melting vessels and solidified. The molten materials are agitated by thermal convection generated
in vertically-divided heating regions.

OPC: Ordinary Portland Cement

4.3.2.1 Examination of applicability of low-temperature solidification processing tosecondary waste generated by water treatment
4.3.2.1.1 Examination of applicability of cement/AAM solidification processing technology
The applicability of low-temperature solidification processing technology to secondary waste
generated by water treatment was examined in small-scale testing (Table A11-2).
Table A11-2 Applicability of low-temperature solidification processing149

Next, solidified cement and solidified AAM were prepared for carbonate slurry as ALPS slurry(solidified waste content (WL) of 30 wt%) and iron coprecipitated slurry (solidified waste WL of 20%)
to obtain their characteristic data (fluidity, caking, compressive strength, void ratio, etc.) and

151 IHI Corporation, The Summary results of subsidy program for the Project of Decommissioning and Contaminated

Water Management, “R&D for processing/disposal of solid waste” (FY 2019 to 2020) in the FY2018
Supplementary Budget, being ready for release in https://dccc-program.jp/category/result
152 Orano ATOX D&D SOLUTIONS, The Summary results of subsidy program for the Project of Decommissioning
and Contaminated Water Management, “(R&D for processing/disposal of solid waste) Applicability evaluation of
the In-can vitrification process to Fukushima waste (FY 2019 to 2020) in the FY2018 Supplementary Budget,being ready for release in https://dccc-program.jp/category/result
153 Kurion Japan K.K, Subsidy program for the Project of Decommissioning and Contaminated Water Management,“(R&D for processing/disposal of solid waste) Project of HOLISTIC EVALUATION OF GEOMELT ICVTM FOR
TREATMENT OF 1F WATER TREATMENT SECONDARY WASTES” in the FY2018 Supplementary Budget,being ready for release in https://dccc-program.jp/category/result）

202

examine the applicability of low-temperature solidification processing technology based onsolidified package performance and operability as evaluation axes (Fig. A11-15).
In FY 2021, application testing of low-temperature solidification processing technology using
dehydrated ALPS slurry to homolytic/homogeneous solidification and filling/solidification was
performed on an engineering scale (200 L), and its practical application will be verified.

Green border: minimum range
Red border: preferred range

Applicability of low-temperature solidification processing technology to carbo
Fig. A11-15 nate slurry149

4.3.2.1.2 Investigation of the possibility of expanding applicability by using special cement
Cement with a high volume of blast furnace slag (CB90, normal cement : blast furnace slag =90% : 10%) and alumina cement containing fly ash (AF20, alumina cement : fly ash = 80% : 20%)
were selected as special cements with potential applicability in terms of high fluidity duringsolidification (feasibility of solidification process and suppression of hydrogen generation) and highheat resistance (increased radioactive inventory and WL) in accordance with the characteristics ofthe secondary waste generated by water treatment that contains chemical components (sodium
carbonate, boric acid, sodium chloride) which adversely affect cement solidification. While CB canbe applied to sodium carbonate mixed waste, AF may have superior heat resistance and fluidity.
The influence of their chemical composition is different from that of OPC.
The solidification characteristics of CB and AF according to their chemical composition were
tested. Then, the applicability of ALPS slurry was examined based on fluidity, setting time, breeding
and compressive strength as evaluation criteria. The maximum WL for carbonate slurry was 25%
for CB90 and 35% for AF20, and the WL for iron coprecipitated slurry was 25% for both CB90 and
AF20. The effectiveness of CB90 was confirmed for sodium carbonate and boric acid which
affected the setting time.
4.3.2.1.3 Examination of inspection methods on solidification possibilities by lowtemperature solidification processing
Physical properties (shape and size, particle size distribution, density, bulk density, water content,chemical composition, radiation dose, properties when liquid is added) that affect the feasibility of
low-temperature solidification processing (influence of processes and solidified waste) weresummarized and mock materials with these properties (silica sand, fine silica powder, calcium203carbonate, sodium carbonate, magnesium sulfate, iron oxyhydroxide, barium sulfate, metallic
aluminum, coal ash, montmorillonite) were selected. Using these mock materials, a two-stepscreening method for inspecting the potential of waste solidification in a small scale before lowtemperature solidification processing of the secondary waste generated by water treatment wasdeveloped. In the primary screening, the waste properties are measured with a 50 mL centrifuge
tube to evaluate the feasibility of the process as brief assessment. In the secondary screening, theproperties during solidification processing are measured with a 2 cm square sample to evaluate
the performance of the kneaded/mixed products and solidified waste. This screening method wastested using the screening method developed by using mock ALPS slurry, and it was confirmedthat solidification by low-temperature solidification processing technology was possible.

2044.3.2.2 Examination of applicability of high-temperature solidification processing
technology
4.3.2.2.1 Examination of vitrification of secondary waste generated by water treatment by
crucible testing
WL, for which high-temperature solidification processing of the secondary waste generated bywater treatment is possible, was examined through crucible testing. (Table A11-3). In order tosuppress volatilization of Cs generated during high-temperature processing of secondary waste
generated by water treatment, examination was also made on WL for which solidification
processing is possible at a low melting temperature. WL applicable to engineering-scale testing of
high-temperature solidification processing technology was examined in crucible testing, and it was
confirmed that each processing technology enables the secondary waste generated by water
treatment to be melted and solidified at the set melting temperature (Table A11-4).
Table A11-3 High-temperature solidification processing test in crucibles

Table A11-4 Filling rate of waste in engineering-scale testing (WL) 151,152,153( ) : Melting temperature during engineering scale test205

4.3.2.2.2 Engineering-scale testing of high-temperature solidification processing
technology
Fig. A11-16 shows the condition in the melting furnace during engineering-scale testing withsingle and mixed waste in CCIM. Characteristics data of the light-emitting, circular section at thecenter and the periphery of the molten metal surface was obtained using an engineering-scale
system (CCIM, In-Can, GeoMelt ICV) for the high-temperature solidification processing technology,
to produce solidified waste through high-temperature processing of the secondary waste generated
by water treatment. Due to bubbling (the molten metal is stirred by air supplied to the inside), thecirculation of molten metal is active in this section. The light-colored section between circular lightemitting sections at the center and the periphery is a calcinated layer.
Testing using In-Can was conducted152 to produce solidified waste. Mock waste and glass
additives were placed in a can before melting, and other mock waste and glass additives wereadded while heating at 1100°C. Fig. A11-17 shows the cross-sectional photo of the solidified waste
produced153 by In-Can engineering-scale testing.
A test using GeoMelt ICV was conducted to produce solidified waste. Mock waste was placed ina melting vessel before melting, and heated at approx. 1400°C from the bottom. After the mockwaste was left in the vessel for approx. 1.5 to 15 hours, top-off frit (low-melting-point glass material
that does not contain mock waste) was added for approx. 0.5 hour. Fig. A11-18 shows the GeoMelt
ICV engineering-scale testing condition.
In FY 2021, engineering-scale testing using dehydrated carbonate slurry, examination of supply
methods, promotion of high-filling of carbonate slurry, suppression volatilization of Cs, and test
restarting of the furnace after shutdown are planned. The temperature distribution in the furnaceand the Cs volatilization mechanism will also be studied.

Fig. A11-16 Engineering-scale testing condition in CCIM (4.5 hours after placing in service)

Fig. A11-17 Condition of solidified waste produced by In-Can engineering-scale testing206

Fig. A11-18 GeoMelt ICV engineering-scale testing and condition of solidified waste
4.3.2.3 Examination on stability of waste packages
In order to verify stability at disposal of the solidified waste produced by the selected
processing technology, leachability of the solidified waste, deterioration of the solidified waste
after low-temperature solidification processing caused by temperature, and the influence of thedeterioration of the solidified waste over the long term were examined149.
4.3.2.3.1 Examination of leachability of solidified waste
The leaching rate of the mock nuclide in the solidified waste obtained by low-temperaturesolidification processing of the ALPS slurry (evaluated with the leaching rate of all the mock nuclides
leached as 100%) was determined from the concentration of ion species in the solidified waste andthe amount of ion species leached from the solidified waste after soaking the solidified waste witha curing period of 28 days in pure water for 91 days (Table A11-5).
Table A11-5 Leaching rate of mock nuclides in solidified waste after low-temperature solidification
processing149
Processing
Technology

Solidified
Material

Leaching Rate (%)

Waste Loading
Rate (wt%)

Cs

Sr

Sn

Ce

CarbonIron

CarbonIron

CarbonIron

CarbonIron

CarbonIron

Cement

OPC

30

20

95

80

5

15

<0.1

0.1

--

--

AAM

M

30

20

23

12

0.5

<0.1

3-5

1-3

<1

<1

AAM

MB20

30

20

18

15

0.5

<0.1

3-5

1-3

<1

<1

AAM

MB40

30

20

27

20

0.5

<0.1

3-5

1-3

<1

<1

Carbon: Solidified carbonate slurry, iron: Solidified iron coprecipitated slurry
OPC: Ordinary Portland Cement, M: Metakaolin
MB20: Metakaolin + 20% of blast furnace slag, MB40: Metakaolin + 40% of blast furnace slag“- -”: No leaching

The MCC-1 test was conducted on the solidified waste produced by engineering-scale testing of
the high-temperature solidification processing technology to calculate the leaching rate of thesolidified waste (Table A11-6).

207

Table A11-6 Leaching rate of solidified waste produced by engineering-scale testing (MCC-1 test)

4.3.2.3.2 Examination on deterioration by temperature of solidified waste after lowtemperature solidification processing
Temperature increase due to the decay heat of the nuclides contained in the solidified waste was
examined149. The highest temperature of solidified cement (1 × 105 - 1 × 1010 Bq/cm3) was evaluatedin cylindrical and square containers, and by radiation transfer and thermal analyses. The result of
calculating the calorific value using the measured value of the solidified AAM showed that thecalorific value of the solidified AAM was lower than that of the solidified cement. The calorific value
was calculated by several layout models, in which solidified waste containers were aggregated, toderive a relational expression between activity concentration and maximum temperature. The
upper limit of activity concentration for the temperature limit of cement deterioration (60°C) was 6.8
× 199 Bq/cm3. Evaluation of changes in temperature of the solidified carbonate slurry cement (6m3 of square container) at the maximum WL (30 wt%) using a single layout model based on theinventory (1 × 107 Bq/cm3 : analyzed value) contained in it showed that the temperature rise wasabout 3°C and that the possibility of the performance of the solidified waste decreasing was low.
Next, the deterioration of solidified waste by changes in temperature of storage facilities was
examined. Heating and drying load testing was conducted on the solidified waste after lowtemperature solidification processing to assess the compressive strength and strain amount of thesolidified waste. The mass depletion rate, shrinkage strain and compressive strength of thesolidified waste by drying and heating were also assessed (Fig. A11-19). As a result, the solidified
cement had higher drying/heat resistance than the solidified AAM, while the strain amount of thesolidified AAM was extremely high. Although the solidified cement and solidified AAM satisfied thestandard value of compressive strength (1.47 MPa), it was proven that their performance drasticallydecreased under high-temperature and drying conditions.

Fig. A11-19 Heating and drying load testing on solidified carbonate slurry149

208

4.3.2.3.3 Examination of influencing factors on long-lasting deterioration of the solidified
waste after low-temperature solidification processing
For deterioration of the solidified carbonate slurry cement, the change of its mineral phase due
to WL or temperature was evaluated by thermodynamic equilibrium calculation149. Based on thechemical composition of cement hydrate and solidified waste, carbonate minerals and chloride
minerals to be generated were extracted from a thermodynamic database. The calculated results
could basically reproduce the changes in liquid-phase chemical composition, pH and mineralphase composition in experiments. However, hydrotalcite, the Mg-AI mineral that was not identified
in experiments, was formed at all liquid-solid ratios in calculations. It was assumed that thecalculated AI concentration of the liquid phase might become much lower than the experimental
results due to the formation of hydrotalcite. From these results, it was found that the long-lasting
deterioration behavior of solidified cement could be analytically captured by thermodynamic
equilibrium calculation. It is necessary to expand the thermodynamic database in order toaccurately grasp the long-term behavior of diverse types of solidified waste. Currently, it is difficult
to evaluate the long-lasting deterioration behavior of solidified AAM by thermodynamic equilibrium
calculation. Therefore, in order to examine the applicability of the solid phase of the solidified AAM
to the thermodynamic equilibrium calculation, the basic testing of AAM reaction with waste wasconducted. The solidified ALPS slurry AAM before/after immersion was analyzed by SEM and TEM
to obtain phase-change data of peripheral waste components. From SEM observation, the joint
between slurry and AAM matrix components was identified, and then TEM observation was
conducted. As a result of TEM observation of the interface between the waste and AAM matrix on
the section determined by SEM observation, no reaction was observed at the interface between
slurry and base materials. Deterioration due to the migration of free components or traces ofcrystallization of the AAM base materials was observed. From this fact, it was found that thepossibility of the mineral phase formation by the interaction between base AAM materials and slurrywould be low at this time. It was found that consideration of the mineral phase formation by theinteraction between slurry and base materials would not be required in thermodynamic equilibrium
calculation.
In order to understand the long-lasting deterioration behavior of solidified waste, it is necessary
to select an appropriate acceleration test method and evaluate the long-lasting deterioration
behavior by acceleration testing of solidified waste.
4.3.2.4 Summary
The following results were obtained for low-temperature (cement and AAM solidification) andhigh-temperature (CCIM, In-Can, GeoMelt ICV) solidification processing technologies.
It was confirmed through small-scale, low-temperature solidification processing testing andhigh-temperature solidification processing testing using crucibles that the selected processing
technology would be able to solidify secondary waste generated by water treatment. A

209

screening technique for determining the possibility of solidifying ALPS slurry with lowtemperature solidification processing technology was developed.
The applicability of ALPS slurry by low-temperature solidification processing was given. It will
be necessary to verify the applicability of the low-temperature solidification processing tosludge generated by decontamination devices containing cyanide.

Through engineering-scale testing in high-temperature solidification processing facilities(CCIM, In-Can, GeoMelt ICV), the prospect of the practical application of the high-temperaturesolidification processing technology to ALPS slurry and sludge generated by decontamination
devices was verified. As for the low-temperature solidification technology (cement and AAM
solidification), engineering-scale testing using ALPS slurry will be performed in FY 2021 toverify the prospect of its practical application.



The leaching ratio of mock nuclides contained in the solidified waste after low-temperaturesolidification processing and the leaching rate of elements as main components of somesolidified waste after high-temperature solidification processing were given. Deterioration of
the solidified waste after low-temperature solidification processing affected by temperature andlong-lasting deterioration of the solidified waste was also examined. Going forward, it will benecessary to examine the leaching rate of the solidified waste associated with elements andunder testing conditions with disposal in mind as well as evaluate the long-term deterioration
behavior of the solidified waste by acceleration testing.
These things have enabled clarification of the processing technologies for which practicalapplication can be expected for stabilizing and immobilizing several important waste streams suchas secondary waste generated by water treatment.

210

4.3.3 Establishment of methods to reasonably extract candidate technologies applicable topreceding processing
The following was examined in an attempt to “establish methods of reasonably selecting
processing technology to stabilize and immobilize waste based on the expected processing
technology before the technical requirements for disposal are determined (i.e., preceding
processing)”:Provide waste package specifications based on the identified candidate processing
technologies

Enable judgment on whether the candidate processing technologies can be extracted taking
into account uncertainties in characterization of the waste streamClarify the applicability of processing technologies being studied

4.3.3.1 Examination on the influence of waste composition on performance of solidified
waste
4.3.3.1.1 Examination of waste composition during low-temperature solidification
processing
The waste composition (water/solid ratio) during low-temperature solidification processing wasevaluated from the test result of fluidity, caking and compressive strength149. The water/solid ratio(W/P) of the waste composition in cement solidification processing was 54 wt% in the selectedcarbonate slurry (WL of 30 wt%) and 48 wt% in the selected iron coprecipitated slurry (WL of 20wt%). Here, W/P is the weight ratio of water to cement + waste. It was found that the compressive
strength of solidified carbonate slurry cement decreased as the filling ratio and W/P increased. W/P,
the waste composition in AAM solidification processing was 86 wt% in the carbonate slurry (WL:
30 wt%, blast furnace slag: 20 wt%) and 93 wt% in the iron coprecipitated slurry (WL: 20 wt%, blast
furnace slag: 20 wt%). Here, W/P is the weight ratio of alkali mixed water to metakaolin + blast
furnace slag + alkali mixed water. It was found that the compressive strength of the solidified
carbonate slurry AAM decreased with increasing WL and W/P, and decreased with a decreasing
Na/Si ratio by adding blast furnace slag to the waste composition.
4.3.3.1.2 Examination of waste composition during high-temperature solidification
processing
A glass property model (PNNL-DB) for determining a composition that enables vitrification wasused to evaluate the waste composition during single and mixing (combination of nine types ofwaste) processing of carbonate slurry, zeolite, silicone-titanate, etc., which satisfy the limiting
conditions of the melting furnace (melting temperature, viscosity, electrical conductivity, etc.) 149.
The maximum WL in high-temperature processing of the carbonate slurry was 12 wt% for GeoMelt
ICV, CCIM, and In-Can, and the maximum WL in high-temperature processing of the ironcoprecipitated slurry was 20 wt% for GeoMelt ICV, CCIM, and 15 wt% for In-Can. It was found thatthe applicability of the waste composition in high-temperature solidification processing was

211expanded in mixing processing in which easy-to-vitrify zeolite and iron coprecipitated slurry, etc.
were combined.
4.3.3.1.3 Influence of changes in waste composition on high-temperature solidification
processing
Due to the large uncertainty in the analysis results of waste, the possibility of solidification by
high-temperature solidification processing (CCIM, melting temperature at 1100°C) in case ofchanges in waste composition was examined149. For CaO, MgO and Na2O, which are the main
components of carbonate slurry (WL of 30 wt%), and Fe2O3 + CoO + TiO2 + ZnO, SiO2, and Al2O3,
which are the main components of iron coprecipitated slurry (WL of 40 wt%), the influence on hightemperature solidification processing in case of changes in the composition of the main
components was verified through crucible testing to provide the fluctuation range of solidifiablewaste composition (Fig. A11-20).

Fig. A11-20 Fluctuation range of waste composition eligible for high-temperature solidification
processing149
4.3.3.2 Examination on suppressing Cs volatilization during high-temperaturesolidification processing
Volatilization suppression measures of Cs during high-temperature solidification processing were
investigated to examine Cs volatilization characteristics and Cs volatilization suppression
mechanism based on processing methods and operating conditions. Items affecting volatilizationof Cs include cold cap (cold portion consisting of the waste supplied on the upper part of the molten
section, etc.), top-off frit, bubbling and boron addition.
The influence of cold cap, top-off frit and bubbling were evaluated by crucible testing149. As aresult, it was found that when a cold cap was formed at a surface temperature of 800°C or lower,
the volatilization volume of Cs was reduced by more than 2 orders of magnitude. For top-off frit, it
was found that the volatilization volume of Cs decreased, but the effect was limited when themelting progresses quickly. As for bubbling, the Cs volatilization volume almost doubled comparedwith the case without bubbling but the increase was small compared to that in the bubbling area ofthe molten surface.
In an attempt to investigate the mechanism of Cs volatilization suppression with a cold-cap whensolid waste is solidified at high temperatures in a GeoMelt ICV, crucible testing was used to212calculate the TG/DTA curves of zeolite + sludge generated by decontamination devices in the caseof undergoing high-temperature solidification processing. The test results showed that weight loss
ended at 730°C, and then vitrification was triggered. Since there was no significant Cs migration
around the calcinated layer during melting, it was confirmed that no Cs diffusion that could causesignificant compositional changes would not occur.
Regarding the addition of boron, Cs-sorbed zeolite or silicic titanium acid + zeolite was heatedwith a vitrifying agent, and the weight was measured in each temperature range to verify the Cs
volatilization volume for different temperatures. As a result, the Cs volatilization volume increased
with the temperature rise at 950°C or higher149. Mass spectrometry of the volatile components andmass balance of samples before/after heating showed that Cs and B volatilized at a molar ratio of
1:1. It was found that the concentration of B in the vitrifying agent might affect the Cs volatilization
volume.
4.3.3.3 Examination on the waste package performance produced by processing
technology
Solidified waste produced by low-temperature solidification processing is either a uniform
solidified waste package or a container-filled and solidified waste package stored in a 200 L drum
can or a large rectangular-shaped container. A solidified waste container produced by hightemperature solidification processing is either one of two types of 200 L containers for CCIM, a
universal canister (UC, used for waste package for high-level waste in France) or a cylindricalshielding container for In-Can, or an ICV container (square-shaped container used during melting
in GeoMelt ICV) for GeoMelt ICV (Table A11-7). It is assumed that a solidified waste container
produced by high-temperature solidification processing is used with a waste package container as
it is or is contained in another waste package container.
Table 4.3-1 Solidified waste container produced by high-temperature solidification processing
technology155,156,157

UC: Universal canister

Next, the specifications (density, compressive strength, and G-value) of the solidified waste
package produced by low-temperature solidification processing of carbonate slurry and iron213

coprecipitated slurry were examined (Table A11-8). The specifications (WL, density, compressive
strength, and composition) of the solidified waste package produced by engineering-scale testingof the high-temperature processing technology were also examined (Table A11-9).
Table A11-2 Specifications of the solidified waste package produced by low-temperaturesolidification processing of carbonate slurry149

Table A11-9 Specifications of the solidified waste package produced by high-temperaturesolidification processing of secondary waste generated by water treatment151,152,153

4.3.3.4 Examination on practical-scale systems for solidification processing technology
A practical-scale system concept for processing technology was examined to obtain economic
performance data, such as system configuration and processing efficiency, details of maintenance,
types of consumables and replacement frequency, and secondary waste volume.
The low-temperature processing technology has in-drum and out-drum mixing methods. In thein-drum mixing method, waste and cement are kneaded and mixed inside the container. In the outdrum mixing method, waste and cement are kneaded and mixed in a kneader outside the container.
In the case of solid waste, cement that is kneaded/mixed outside the container is fed into thecontainer containing waste149.
For high-temperature processing technology, it is assumed that the existing secondary waste
generated by water treatment will solidify in a decade. For CCIM, consideration is underway to
install two melting furnaces (to produce 200 kg of glass), and use a 200 L container depending on
the waste inventory to be treated as a container for solidification (single container or double
container) to which molten materials flow down. For In-Can, it is planned to install a Cs-rich unit
and a Sr-rich unit according to the waste properties, and to set up two Dem & Melt units each (toproduce 300 kg of glass). For GeoMelt ICV, assuming that the existing KURION and SARRY

214

sorbent will be solidified in a year and the ALPS slurry in 4 years, it is planned to install two GeoMelt
ICV melting furnaces (to produce 10 tons of glass).
4.3.3.5 Summary
As a result of examination for establishing methods to reasonably extract a set of candidate
processing technologies that can be applied to the preceding processing, the following outcome
was obtained:

The waste composition that can solidify an ALPS slurry at low temperature was given In
addition to the method for inspecting the possibility of solidification shown in 4.3.2.1.3, it
became possible to determine whether low-temperature processing can be applied to ALPS
slurry. Going forward, it is necessary to verify the applicability to sludge generated bydecontamination devices containing cyanide.

The carbonate slurry is produced in a large amount, but the WL during solidification processing
is small. For such waste, it is necessary not only to aim at improving WL, but also to examine
the measures required for optimizing all entire waste such as by processing several types ofwaste together.

The waste composition in the case of single and mixing processing of the secondary waste
generated by water treatment by applying each high-temperature solidification processing
technology was evaluated and provided in a glass property model. Comparative verification
with engineering-scale test results will be needed in the future.



The range of waste composition of ALPS slurry to which high-temperature solidification
processing can be applied was given.



The Cs volatilization volume by suppression during high-temperature solidification processing
was evaluated. The suppression mechanism of Cs volatilization with a cold-cap was also
investigated, demonstrating the effect of Cs volatilization suppression. Additional tests will be
conducted in FY 2021 to examine in detail the effect of high-temperature solidification
processing technology on the reduction of Cs volatilization.



The specifications (density, compressive strength, G-value, and composition) of the solidified
waste container and the solidified waste produced by each processing technology were given.

Examination was made on full-scale configuration of low-temperature and high-temperaturesolidification processing systems, details of maintenance, processing speed, amount of
secondary waste generation, and economic performance.
As a result, it has become possible to extract the candidate processing technologies based on

the uncertainty of the waste composition in characterization by providing the applicability of thesolidifiable waste composition. Waste package specifications after solidification were also provided
based on the identified candidate processing technologies applicable for preliminary processing.

215

4.3.4 Consideration of disposal concept and development of safety assessment method
4.3.4.1 Solid waste characteristics and development of disposal concept
4.3.4.1.1 Solid waste characteristics, and needs for safe and reasonable disposal
In order to develop technology for the safe and reasonable disposal of such waste, there is aneed for a research approach to disposal that takes into account the characteristics of waste.
This solid waste is characterized by a wide variety of contamination conditions and chemical
substances as well as a large physical volume (large amounts). There are also uncertainties interms of contamination conditions and selection of processing/segregation/decontamination
options. These characteristics raise the need for the diverse acceptability of waste, capacity
enhancement, and measures to address the risk of uncertainty depending on the disposal concept.
As an approach to research on the disposal of such waste characterized by diversity, large
amounts, and uncertainty, it is not appropriate to simply apply the existing disposal concept fornormal radioactive waste in Japan to solid waste because this concept was not developed basedon a sufficient understanding of these characteristics and the needs arising therefrom. Given this,a research approach was adopted that establishes a safe and reasonable disposal concept (a basic
concept including depth of disposal, barrier structure, and waste package) for the characteristics ofsolid waste, and that develops safety assessment methods (scenario development, evaluation
model development, and evaluation parameter setting) for such disposal concept, while also
investigating promising technology, reference cases, and research results in Japan and overseas
and incorporating the knowledge gained.
4.3.4.1.2 Needs based on the characteristics of solid waste, and reference cases identified
Fig. A11-21 links overseas reference cases with the needs for diverse acceptability of waste,capacity enhancement, and measures to address the risk of uncertainty raised by diverse, largevolume, and uncertain characteristics of waste. The reference cases were selected with respect towhether they comply with environmental conditions, including the geography and geology of Japan;
comply with regulations and are related to processing/disposal concepts in Japan; have a proventrack record; have high economic performance; and have low enough levels of uncertainty toachieve the processing/disposal concepts in detail.

216

Fig. A11-21 Needs arising from solid waste characteristics, and matching with overseas
reference cases149
The disposal concepts were examined based on the reference cases (numbering in Fig. A1121) corresponding to the needs. Then, as potential disposal options, the disposal concepts
equivalent to trench disposal, pit disposal, and mid-depth disposal were clarified to the extent thatwould allow model and parameter setting for safety assessment (Fig. A11-22)149. The status ofincorporation of the reference cases into the disposal concept are shown in association with thedisposal concept in Fig. A11-22, with reference cases shown and numbered in Fig. A11-21.For the
disposal category equivalent to trench disposal (L3), in reference to VLLW disposal sites in UK, thebarrier structure is likely to have clay-covered liners with a low-permeability layer in the upper partof the disposal area surrounded by the low-permeable floor and sidewalls. For the disposal
category equivalent to pit disposal (shallow ground disposal site (L2)), in reference to LLWR in UK,
the barrier structure is likely to be such that the disposal area of a reinforced concrete vault
consisting of a floor covered with low-permeability clay and walls with bentonite on the surface isbackfilled with cement grout, and water-shielding barriers using geomembrane, etc. are installed in
the upper part. For the disposal category equivalent to mid-depth disposal (mid-depth disposal
facility (L1)), in reference to the silo type disposal site for short-life, low- to medium-level radioactive
waste in Sweden, the barrier structure is likely to be such that the disposal area of the reinforced
concrete pit is placed inside a tunnel pit several dozen meters below ground or deeper, with theoutside covered with low-permeability liners and the void space backfilled with cement grout andclosed with plugs.

217

(1)(3)(4)(6)(7)(8)(1)(3)(4)(6)(7)(8)(11)(1)(3)(4)(6)(7)(8)(14)Fig. A11-22 Disposal concepts equivalent to trench disposal, pit disposal, and mid-depth disposal
clarified as potential disposal options based on reference cases149

218

4.3.4.2 Safety assessment technique for solid waste disposal
In order to identify priority issues to be discussed, parameters for assessment were defined
based on the disposal concepts assumed above, and safety assessment was performed149. An
outline is provided below.
4.3.4.2.1 Selection of representative solid waste
In order to provide Technical Prospects for a safety assessment technique and methods for
examining disposal concepts for solid waste, the following types of waste, (1), (2), and (3), whichhave “characteristics that make safety assessment difficult,” were selected for further examination.
One or more types of solid waste were selected from three categories: rubble, secondary waste
generated by water treatment, and dismantled waste consisting of solid waste.
(1) Large-volume: The feasibility of disposal is greatly affected because the capacity required for
disposal facilities also changes and the (economical) reasonability varies greatly dependingon the selection of disposal options. This makes it difficult to eliminate excessive conservatismin accordance with the ALARA concept154.
(2) Diversity and uncertainty of the types, concentrations and chemical substances contained incontaminated nuclides: These factors increase the uncertainty in determining the nuclide
inventory in disposal. It is difficult to determine the nuclide inventory given this uncertainty. As
the contained chemical substances may have a direct/indirect influence on the nuclide
relocation behavior, it is also difficult to properly incorporate its influence into safetyassessment.
(3) Uncertainty of processing/segregation/decontamination options: The selection of processing
options changes the nuclide release behavior of the waste package, and it is difficult toestablish a source term model according to the waste package characteristics.
○ For rubble, Rubble 1, Rubble 2, and waste inside buildings (concrete) were selected in termsof Characteristics (1) and (2).
○ For secondary waste generated by water treatment, KURION and carbonate slurry wereselected based on the above Characteristics (1), (2) and (3), and Decontamination device
sludge from Characteristics (2) and (3).
○ For dismantled waste, RPV waste and waste inside PCVs were selected in terms of the above
Characteristic (2) different from rubble (contamination overlap caused by radioactive materials
from activated materials and those from fuel).

154 The principle of optimizing radiological protection advocated by the International Commission on RadiologicalProtection (ICRP), and an abbreviation for “as low as reasonably achievable”. That is, the magnitude ofindividual doses, the number of people exposed, and the likelihood of incurring exposures “should all be kept as
low as reasonably achievable, economic and social factors being taken into account” (Japan Radioisotope
Association, 1990 Recommendations of the International Commission on Radiological Protection, (1991)).

219

4.3.4.2.2 Setting of evaluation parameters(1) Example of hydraulic parameter setting for a standard geological environment and engineered
barriers
The evaluation parameters were defined based on the disposal concept clarified in Fig. A11-22.
Table A11-10 shows parameter setting examples based on the standard geological environmental
conditions, functions/characteristics of the disposal system, and their changes. In setting theevaluation parameters, in light of uncertainties, a pessimistic case with pessimistic parameters, anda basic case with realistic parameters are provided.
Table A11-10 Parameter setting examples based on the standard geological environmental
conditions, functions/characteristics of the disposal system, and their changes149
Radiation dose evaluation parameters
Item (unit)
Conservative
Realistic
setting value
settingvalue-5
Natural
3.0×10
1.0×10-5
barrierpermeability
coefficient,surfacelayer (m/s)
Natural
1.0×10-6
1.0×10-7
barrierpermeability
coefficient,shallow
layer (m/s)
Natural
1.0×10-7
1.0×10-8
barrierpermeability
coefficient,mediumdepth(m/s)
Bentonite
1.0×10-8
1.0×10-10
mixed soil,
L2(m/s)
Bentonite
L1 (m/s)

1.0×10-10

1.0×10-11

River flowrate
(m3/y)

2.0×10-7

1.0×10-8(1) Distribution
form
Lognormal
Site-specificParameter characteristics(2) Sensitivity to(3) Magnitude ofradiation dose
impact to radiation(linear/nonlinear)dose (sensitivity)
Exponential
Large impact on
short
half-life
nuclidesEvaluation
Importance
Action policy(comprehensive
evaluation)
High impact andConservativesite dependency
setting
Multiplecondition
setting

Lognormal
Site-specificExponential

Large impact on
short
half-life
nuclidesHigh impact andsite dependency

Conservative
setting
Multiplecondition
setting

Lognormal
Site-specificExponential

Large impact on
medium and long
half-life nuclides

High impact,with site
dependency

Conservative
setting
Multiplecondition
setting

Lognormal
Depends oncomposition,
construction,
deterioration,etc.
Lognormal
Depends oncomposition,
construction,
deterioration,etc.
Site-specificNational
distribution
close tolognormal

Linear against
seepage water
volumeLarge impact also
on medium andlong
half-life
nuclidesHigh impact,
depends ondesign andconstruction

Conservative
setting
Multiplecondition
setting

Linear against
seepage water
volumeLarge impact on
medium and long
half-life nuclides
as wellHigh impact,
depends ondesign andconstruction

Conservative
setting
Multiplecondition
setting

LinearInverse proportion

High impact andsite dependency

Conservative
setting
Multiplecondition
setting(2) Incorporation as evaluation parameters with time variation in the degradation process ofengineered barriers
The degradation process of engineered barriers was incorporated as evaluation parameters withtime variation. Specifically, parameters with time variance were defined with the assumption of
degradation in engineered barriers such as geomembrane, concrete, and metal containers. With
the disposal concept equivalent to pit disposal, referencing LLWR in the UK, evaluation was
conducted under the assumption that the water permeability coefficient of water-shielding barriersconsisting of clay and a geomembrane would rise linearly from 10-12 m/s to 10- 9 m/s from 150 to
1,000 years after installation due to the degradation of the geomembrane. For the disposal option

220equivalent to mid-depth disposal, the assumption for the evaluation was that the water permeability
coefficient of the disposal facility would change linearly with time from 10-11 m/s equivalent toreinforced concrete to 10-6 m/s over 5,000 years due to loss of barrier functions.
Regarding the containment performance of disposal containers, according to research results fromthe UK, evaluation was performed based on the assumption that loss of containment performanceof ISO containers, stainless steel containers, and HICs would occur 10, 300, and 1,000 years after
their placement, respectively, and then the containment performance would be lost completely overthe same period.
(3) Impact of chemical substances contained in solid waste
First, the chemical substances contained in solid waste that may affect the nuclide relocation
behavior were extracted. Organic substances, seawater components, and boric acid solution,
which are common to most types of solid waste, were selected. Although they are not very similar,
ferrocyanide, sulfates, and carbonates were also selected as components, which are contained in
the priority secondary waste generated by water treatment and whose influence on nuclide
migration is a concern. The influence of these chemical substances shall be evaluated as a sorption
reduction factor (coefficient indicating the decrease of sorption distribution coefficient, which is aparameter of nuclide migration. Hereinafter referred to as “SRF”). Barrier materials (cement-basedmaterials and bentonite) were selected as substances to be sorbed, as they are common to all
disposal concepts, placed in spatial proximity to waste packages, and considered to be greatly
affected by chemical substances. The nuclides to be evaluated were determined from those for the
existing disposal concept and a grouping concept based on chemical similarities. Table A11-11
shows the substances of influence (framed in red) and the waste that may contain them.

221

Table A11-11 Substances of influence and waste in which they are contained149



Dismantled waste (other)

Dismantled waste (metal)Dismantled waste (concrete)Used protective clothing



Soil



Felled treesDismantled waste

Rubble (other)

Rubble (metal)

Filter

Rubble (concrete)

Resin-based adsorbent (column)



Chelate resin

Ferrocyanide

Raw waste componentsCarbonateMagnesium hydroxideSulfate

Activated carbon

Organic material



Titanium-based material









Ferrocyanide























Concrete

Coexisting materials

Impacting materials

Rubble / felled trees, etc.



Iron hydroxide

Metal

Titanium oxide

Titanate

Silica-based material

Ag rooted activated carbon

Carbonate slurry

Iron coprecipitation slurry

Secondary cesium adsorption vessels (SARRY)

ZeoliteRaw waste

Waste sludge

Cesium adsorption vessels

Secondary waste generated from contaminated water











Seawater components































Boric acid solution































Oil content































Organic material





































Silica-based material



























Material with fuel〇 Selected 8 types of wastes and the classification of the original waste in the table
are as follows.
• Rubble1 and Rubble 2 : rubble (concrete), rubble (metal) and rubble (other)• Waste concrete in the buildings : part of dismantled waste (concrete) [including all
waste in the PCVs and in the buildings, because dismantled waste is the entire
waste generated by future dismantling operations other than the rubbles currently
generated]• KURION : secondary cesium sorption vessels• Carbonate slurry : carbonate slurry• Iron coprecipitation slurry : iron coprecipitation slurry• Sludge generated by decontamination devices : sludge generated bydecontamination devices• RPV waste and waste inside PCVs : part of dismantled waste

222

With the objective of preparing data on the influence of the substances affecting disposal on
nuclide migration and other data necessary for evaluation, how to deal with the effects ofdomestic/overseas disposal methods and substances of influence on nuclide migration wasinvestigated. As a result, the following information was collected and organized: handling examplesof organic substances in Sweden, Switzerland and Japan; domestic and overseas examples ofsetting SRFs for isosaccharinic acid (hereinafter referred to as “ISA”), which has the greatest
negative impact on the sorption disposition coefficients of actinoid, etc.; and the same examples
for seawater components, etc.
While utilizing the above knowledge collected and organized from sources inside and outside
Japan, the method shown in Fig. A11-23 was used to assess the direct influence of chemical
substances, and the indirect influence of changes in pH in gap water between barriers and theconcentration of chemical species due to thermal (T), hydrological (H), mechanical (M), andchemical (C) changes (hereinafter referred to as “THMC”) on the environmental side.
After experimentally obtaining the data that had been lacking in the domestic and overseas
findings, together with the thresholds, the SRFs by substances affecting disposal were indicatedby element family grouped on the basis of chemical similarities (Table A11-12). The columns
highlighted in yellow show where, despite the experiment, the SRFs cannot be defined with
sufficient accuracy because the equilibrium concentration of dissolved and precipitated nuclides isvery low. In response to this, an attempt will be made to obtain data and understand the extent of
the impact through retesting and theoretical studies in FY 2021. Through these studies, it isanticipated that the influence of chemical substances on the sorption of nuclides can be evaluatedby the end of FY 2021.
Types and amounts of impacting materials of waste components
and their coexisting materials(organic material, boric acid, salt, cement, etc.)

A) Method for evaluating direct
effects:
Direct effects such as increases in solubility or
decreases in sorption, due to complex formationof impacting material components with nuclides

Reference conditions(barrier material, pore waterchemistry)B) Method for evaluating indirect
effects:
Changes in properties such as pH and salt
concentration of barrier material interstitialwater, due to changes in THMC (particularly
chemical) caused by impacting materials

Threshold value at which
effects occur+
Sorption reduction factor[Methods (1), (2), (3)]

Method for evaluating effects ofenvironmental conditions ofproperties such as pH and salt
concentration[Condition conversion method andsorption model]

Kd + uncertainty

Kd + uncertainty

Reference Kd +uncertainty

A + B) direct + indirect effects:
Consideration for the effects of both A and B according to a
combination of their evaluations

Fig. A11-23 Method to quantitatively evaluate the influence of substances affecting disposal on
nuclide sorption149

223

Table A11-12 Sorption reduction factor (SRF) based on the combination of substances affecting
disposal with elements that group nuclides according to their chemical properties149

Bentonite-based

Cement-based

Element Group

Represent
ative
Element

Alkali metals
Alkali earth metals
Divalent transition metals
Quadrivalent transition
metalsPentavalent transition
metalsTrivalent actinides
Quadrivalent actinides
Pentavalent actinides
Hexavalent actinides
Halogens

Cs
Sr
Ni

Anionic species

Se

Impact-causing Concentrations (Threshold Value) and Sorption Reduction Factors (SRF)
Organic material (ISA)
ThresholdSRF
1
1×10-2 M*1
10Seawater
SRF*3

Boric acid
SRF
Threshold1
1-

Ferrocyanide
Threshold
SRF

-

1

2
8
1

-

1

-

Sn

1×10-4 M

100

10

-

1

Issues exist

Issues exist

Nb

1×10-4 M

100

10Issues exist

Issues exist

Issues exist

Issues exist

Am/Eu
Th
Np
U
I

1×10-4 M
1×10-4 M
1×10-4 M
5×10-4 M
––*2

10
100
10
10
1

10
10
10
10
––*2

Issues exist
1×10-2 M
1×10-4 M

1Issues exist
1
500
1.3

Issues exist
1×10-3 M
1×10-3 M-

1Issues exist
3
3
1

––*2

1

––*2

Issues exist

Issues exist

Issues exist

Issues exist

Element Group

Representative
Element

Alkali metals
Alkali earth metals

Cs
Sr

Divalent transition metals
Quadrivalent transition metals

Ni
Sn

Pentavalent transition metals
Trivalent actinides
Quadrivalent actinides

Nb
Am/Eu
Th

Pentavalent actinides
Hexavalent actinides
Halogens

Np
U
I

Anionic species

Se

Impact-causing Concentrations (Threshold Value) and Sorption Reduction Factors (SRF)
Ferrocyanide
Boric acidThreshold
-

SRF
1
1

1×10-2 M
1×10-4 M
1×10-4 M

1
100
2.6
1.3

Issues exist
Issues exist
Issues exist

SRF
1
1

Threshold
Issues exist

Issues exist

1×10-3 M

2.7
1
1

Issues exist
11Issues exist
1×10-3 M
1×10-3 M

Issues exist
42Issues exist
Issues exist

1×10-4 M

1
2.0

*

1
1
1

*1 M: mol/L, *2 Set

Kd = 0 (m3/kg), *3
SRF between
precipitation
conditions andseawaterconditions

M: mol/L, ** Set Kd = 0 (m3/kg).

* The direct impacts of sulfates and carbonates
were evaluated to be insignificant based on pastfindings on complex formation.
:
Specified based on resultsof past informative investigations.
:
Specified based on dataobtained through sorption tests.

:
Data was acquired but
partition coefficients could not be
obtained with sufficient accuracy(however, those effects are estimated to
be insignificant).

4.3.4.3 Safety assessment for disposal
4.3.4.3.1 Evaluation system for nuclide migration models in a groundwater scenario
Fig. A11-24 shows an evaluation system for nuclide migration models in a groundwater scenario.
In order from top to bottom, this figure provides concepts equivalent to trench disposal, pit disposal,
and mid-depth disposal. The left side of the figure shows the case without engineered barriers, andwith engineered barriers on the right side. The disposal facility is shown in blue and the naturalbarrier in orange. Regarding the Darcy flow rate of the natural barrier and the underground flux
from the disposal facility to the natural barrier, the pessimistic cases are written in black, the basic
cases in red, and the cases common to both in blue. Using these evaluation systems, nuclide
release from waste, sorption distribution to cement-based materials, and sorption distribution tobuffer materials were calculated for the disposal facility area, and sorption distribution to andadvection/dispersion in the natural barrier were calculated for the natural barrier. The right end ofeach evaluation system is connected to the river.

224

Trench Model with Engineered BarrierX

Trench Model with Engineered Barrier

Burial facility
Y

Natural barrierStandard case: 1.0 m3/y/m 2
Pessimistic case:1.0 m3/y/m2

RiverStandard case: 9.5 m/yPessimistic case: 2.8x101 m/yMovement distanceZ

X Burial facility

Pit Model with Engineered BarrierStandard case : 1.6×10-1 m/yPessimistic case : 1.6 m/y
Burial facility

Burial facility

River

X(m)
Y(m)
Z(m)

Pit

100
5
100

100
5
100

River

Burial facility

Standard case: 3.2×10-4 m3/y/m 2
Pessimistic case: 3.2×10-2 m3/y/m2

Standard case : 1.6×10-2 m/yPessimistic case : 1.6× 10-1 m/yTrenchNatural barrierStandard case: 1.6×10-1m/y
Pessimistic case: 1.6 m/yNatural barrier (rock)Pit Model with Engineered BarrierStandard case: 3.2×10-3 m3/y/m2
Pessimistic case: 3.2×10-2 m3/y/m2

Medium-depth Model with Engineered BarrierStandard case : 1.6×10-2 m3/y/m2
Pessimistic case: 1.6× 10-1 m3/y/m2

Movement distanceZ

Standard case: 9.5 m/yPessimistic case: 2.8x101 m/yNatural barrier
River

Y

Natural barrier River

Standard case: 3.2×10-1 m3/y/m2
Pessimistic case: 3.2×10-1 m3/y/m2

Burial facility
Standard case: 0.1 m3/y/m2
Pessimistic case:0.1 m3/y/m 2

BentoniteStandard case: 1.6×10-2m/y
Pessimistic case: 1.6x10-1 m/yMediumdepth

100
10
10

Medium-depth Model with Engineered Barrier
Natural barrier

River

Movement distance: 100m・Trench: 100m・Pit・Medium-depth: 300m

Fig. A11-24 Evaluation system for nuclide migration models149
4.3.4.3.2 Examples of processing effects (improvement in leaching rate) based on safetyassessment at the time of disposal
Based on the disposal concepts shown in Fig. A11-24, radiation dose evaluation models were
established to set evaluation parameters using the method shown in 4.3.4.2.2 and to perform
nuclide migration analysis149. Table A11-13 shows an evaluation example of the basic cases. The
leftmost columns of this table show the disposal categories based on the assessment of theisolation failure scenario after 400 years. The leaching rates of each waste package is reduced
from instantaneous release, release at 10-3/y, 10-4/y and 10-5/y (set close to the standard values,
which are defined as the leaching rate of vitrified waste in nuclide migration analysis of high-level
radioactive waste) from left to right to evaluate the radiation dose in the groundwater scenario, andshow the disposal categories based on the radiation dose assessment. Table A11-13 are dominatedby the groundwater scenario. Looking into the waste types with letters highlighted in red in thistable (the rows with red letters), the groundwater scenario is dominant on the left side of the row,where the leaching rate of the waste package is high. On the right side of the row, however, theleaching rate decreases and so does the radiation dose in the groundwater scenario, i.e., disposal
as a shallower disposal category becomes possible. Then, the radiation dose is eventually reversed
from that in the isolation failure scenario, which is irrelevant to the leaching rate, and the red lettersno longer exist, making the isolation failure scenario dominant in determining the disposal category.
Such examination makes it possible to quantitatively verify the effect of the improvement in theleaching rate by processing upon disposal for each waste type. However, the range of the columns
with red letters is narrow for the waste types evaluated this time. This indicates that, for the waste
types for which the inventory was defined based on the available knowledge, the isolation failure
scenario plays a greater role in determining the disposal category than the groundwater scenario,
and that the extent to which the improvement of leaching characteristics by processing is effectivein enhancing safety in disposal or relaxing the disposal category is limited. Whether the225groundwater scenario or the isolation failure scenario will dominate depends on the nuclide
composition of the waste, i.e., how the nuclide inventory is given, and it is not deterministic at thispoint.
Table A11-13 Example effect (improvement in leaching rate) of processing by safety assessment
on disposal (isolation failure and groundwater scenarios) (Basic case; Conservative
parameter values are used for the groundwater scenarios)149
Isolation
failurescenario

Waste

Basic case of the groundwater scenario
Leaching rate of 1.0 × 10-3/y(L2, L3: 400 y,L1: 1 ×105 y)-

SRF forSRF forSRF for facilities/
naturalfacilities naturalbarriers
barriers

Leaching rate of 1.0 × 10-3/y-

SRF fornaturalbarriers

SRF forfacilities

SRF forfacilities/
naturalbarriers

Leaching rate of 1.0 × 10-3/y-

SRF fornaturalbarriers

SRF forSRF for facilities/
facilities naturalbarriers

Leaching rate of 1.0 × 10-3/y-

SRF forSRF forSRF for facilities/
naturalfacilities naturalbarriers
barriers

KURION

L1

L3/L3

L2/L3

L3/L3

L2/L2

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

Waste sludge

L1

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

Carbonateslurry

L1

L2/L2

L2/L2

L2/L2

L2/L2

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

Rubble 1

L3

L2/L2

L2/L2

L2/L2

L2/L2

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

Rubble 2

L3

L2/L2

L2/L2

L2/L2

L2/L2

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L1/L1

L1/L1

L1/L1

L1/L1

L1/L1

L1/L1

L1/L1

L1/L2

L1/L2

L1/L2

L1/L2

L2/L2

L2/L2

L2/L2

L2/L2

L1/L2

L2/L2

L1/L2

L2/L2

L2/L2

L2/L2

L1/L2

L2/L2

L2/L2

L2/L2

L1/L2

L3/L3

L2/L3

L3/L3

L2/L3

L2/L2

L2/L2

L2/L2

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

L3/L3

Geological
L1/L1
layer
Waste inside Geological
L2/L2
layer
PCV

RPV waste

Waste inbuildings(concrete)

L2

L2/L2

*In this assessment, the rows with red letters are where the groundwater scenario evaluation
results are dominant in determining the disposal category.
4.3.4.3.3 Relationship between processing/decontamination/segregation options anddisposal concept based on safety assessment (disposal category)Fig. A11-25 illustrates the treatment/decontamination/segregation options for Kurion, and thecorresponding disposal categories149. It may be possible to conduct disposal equivalent to L1 by
adopting Option C (segregate sorbents and temporary storage containers, and vitrify sorbents).
As with the Kurion case above, the disposal categories of the selected eight types of waste were
evaluated for each processing, decontamination, and segregation option. As a result, it wasdetermined to classify the RPV waste and the waste in PCVs as equivalent to geological disposal,
regardless of the processing/decontamination/segregation options. However, it was also
determined that, except for these two, other types of waste including secondary waste generated
by water treatment by Kurion could fall into the disposal category equivalent to L1 depending on
the selected processing, decontamination, and segregation options.

226

Temporary

Option storage
A

Option B

Option C

Option D

container

KURIONRaw waste

Removal and separation ofadsorbent

Removal and separation ofadsorbent

Temporarystoragecontainer

Adsorbent

Adsorbent

Treatment

Removal and separation ofadsorbent

Melting

Melting

Grout filling inshielded highdurability container

Grout filling in
SS container

Disposal / reuse
Changes inradioactivity andphysical amountVitrification

Slag

Slag

Adsorbent
packageSlagpackage

Metal(CL level)

99%.・
198%

1%・
3%

0%・
96%

Slag
SS
Container
packaging

Grout filling in SS
container

Adsorbent
Slag package
package99%.・
60%

DryingMelting

Dehydration
compression

Grout fillingin SS
container

Temporarystoragecontainer

Adsorbent

1%・
3%

Grout filling in SS
container

Metal(CL level)

Adsorbent
packageSlagpackage

0%・
96%

99%.・
36%

1%・
3%

Metal
(CL level)
0%・
96%

Dried raw
waste
100%・
100%

Note)

DisposalConcept Option

A
Cement solidification and packaging
in SS containers

Processing Option
B
C
Dehydration compression andVitrification and packaging in SS
packaging in high-durability
containers
containers
TemporaryTemporary
Adsorbent
storage
Adsorbent
storagecontainer
container

D
Drying incontainers

E
Raw waste
disposal,no packaging

Adsorbent

Temporarystorage container

L3

Well users
River users

Site residents

Well users
River users

Site residents

Site residents

Site residents

Well users
River users

Well users
River users

L2

Well users
River users

Site residents

Well users
River users

Site residents

Site residents

Site residents

Well users
River users

Well users
River users

L1

Well users

3%

Well users
River users

3%

36%

3%

Well users
River users

Well users
River users::

In all scenarios, the dose is below the dose reference value, and burial is expected to be possible as a result of dose evaluations. Figures indicate
amounts based on raw waste.
In even one scenario or more, the dose is above the dose reference value, and burial is expected to be impossible as a result of dose evaluations. The
main reasons are also described.

Fig. A11-1 Treatment/decontamination/segregation options for the secondary waste generated bywater treatment by Kurion, and corresponding disposal categories149
As with the Kurion case above, the disposal categories of the selected eight types of waste were
evaluated for each processing, decontamination, and segregation option. As a result, it wasdetermined to classify the RPV waste and the waste in PCVs as equivalent to geological disposal,
regardless of the processing/decontamination/segregation options. However, it was also
determined that, except for these two, other types of waste including secondary waste generated
by water treatment by Kurion could fall into the disposal category equivalent to L1 depending on
the selected processing, decontamination, and segregation options.

227

4.3.4.3.4 Establishment of methods for reasonably selecting processing methods forstabilization and immobilization before technical requirements of disposal are
established (preceding processing)
The associated requirements for the methods for appropriately selecting processing methods forstabilization and immobilization before technical requirements of disposal are established(preceding processing) are (1) to demonstrate safety using safety assessment models
corresponding to the conditioned waste specifications and disposal concepts, and (2) to determine
whether the candidate processing methods can be selected in consideration of waste properties
and uncertainties in characterization. As shown in Table A11-13, it can be said that the goal of (1)has been achieved if a technique is available that compares the models with the prescribedradiation dose constraint based on the dose assessment in isolation failure and groundwater
scenarios to determine the disposal category.
However, with regard to (2), upon understanding the impact to safety assessment of uncertainty
in characterization, it is necessary to extract the treatment, decontamination, and segregation
options as shown in Fig. A11-25, and establish a method for selecting the applicable candidate
technologies in the waste management process by using evaluation techniques and results todetermine the disposal category according to the radiation dose assessment for each option, as
shown in Fig. A11-26. According to this method, and in association with the applicable candidate
processing technologies, the consistency of the waste package specifications with the disposal
system and their safety in case of disposal are evaluated. As a result, an evaluation report will
indicate “Results of disposal safety assessment” and the resultant “Requirements on waste
package specifications”, or “Requirements on disposal options, etc.” if it is more desirable toperform handling on the disposal side than on the processing side.
Disposal safety assessment
Disposal concept based on the characteristics of 1F waste

Disposal concept
Realistic models, scenarios, and parameters

Waste specifications
Classification proposal in response to disposal
concept

Waste proposal from wasteconditioning supervisor
Selection of waste to be considered for processing

Consistency assessment(waste specifications,disposal system concept,
safety)Examination of waste specification
proposalsNO

NO

Disposal safety assessment
Assessment report
Selection of processing technology

Fig. A11-2 Method for selecting applicable processing technologies and disposal options in thewaste management process

2284.3.4.4 Summary
As a result of examination for establishing safe and reasonable disposal concepts and safetyassessment methods based on the characteristics of solid waste, the following outcome wasobtained:Based on the characteristics of solid waste and the need for safe and reasonable disposal,
and in reference to Japan and overseas reference cases, examples of disposal concepts(basic concepts such as disposal methods, barrier structure, and conditioned waste)
appropriate for solid waste were provided.



Representative types of solid waste were selected, and parameters for safety assessment
were defined based on domestic/overseas cases and the disposal concepts shown in 4.3.1.1.3.
In this process, examination was performed to incorporate the influence of chemical
substances contained in solid waste. The target waste and radionuclides and thecorresponding sorption reduction factors were given to evaluate their influence on the radiation
dose. In addition, realistic parameters incorporating degradation behavior were defined,accounting for functions such as the low-permeability property of concrete, the corrosion
resistance of disposal containers (physical containment performance), and water-shielding
barriers (capping) in safety assessment. The influence of parameter changes by processing
on the disposal categories, and the relationship between the processing/decontamination/
segregation options and the disposal categories were exemplified.
Based on the above, safe and reasonable disposal concepts were given, and safety assessment

methods incorporating their characteristics could be developed.
Moreover, a series of methods for reasonably selecting processing methods for stabilization andimmobilization before technical requirements of disposal were established (preceding processing)as described below.


Safety can be provided using safety assessment models corresponding to waste package
specifications and disposal concepts.



As a result of 4.3.3, the candidate processing technologies can be extracted in light of theuncertainty in characterization, and associated waste package specifications be provided.



Further, according to the method shown in Fig. A11-26, the consistency of the waste package
specifications with the disposal system and their safety in case of disposal will be evaluated.
If inappropriate, “Requirements on waste package specifications” will be indicated, and if it ismore desirable to perform handling on the disposal side instead of the processing side,“Requirements on disposal options, etc.” will be indicated.

1.

Now that these requirements have been incorporated and proper measures have been
implemented in the waste management process in light of solid waste characterization and its
uncertainties, it becomes possible to select applicable candidate processing technologies.

229

4.3.5 Prospects of establishing processing/disposal strategies for solid waste for which
processing technology with consideration for disposal has not been clarified
As shown in 4.3.2, for the secondary waste generated by water treatment, including ALPS slurry,
it has become possible to clarify the processing technology for which there are expectations forpractical application with consideration for disposal. The following describes the establishment of
examination methods to present the prospects for defining a strategy for each step from generation
to disposal for other types of solid waste as well149.
4.3.5.1 Method for examining the integration of R&D results
An input chart was developed to summarize the information, issues and research results
necessary for examining each step from waste generation to disposal. Using this chart, the entire
waste stream was examined in a comprehensive manner, and mutual feedback was provided on
research results in order to develop methods for efficient examination.
A general overview of the method for examining the waste stream is shown in Fig. A11-27. The
figure shows input information, integration of examination and research results, and output, fromleft to right.
A processing flow with multiple options can be developed based on information such as theresearch results on inventory evaluation by characterization, the waste list and waste management
sheet. Providing feedback from R&D results to the input chart enables the information and issues
required to be examined in each step in the waste stream to be summarized. The timing at whichthese should be considered can be summarized as a timeline chart in reference to the Mediumand-Long-term roadmap and TEPCO’s storage plan for solid waste.
Based on the above, matters to be considered and their details, status of efforts, and necessary
actions can be summarized in light of the waste stream examination as a hub. Also, for a processing
flow with many options, the flow that is currently proposed was given, and it was refined by theconcept of narrowing down the flow in accordance with the required technical requirements.

230

Fig. A11-3 Method for examining integration of R&D results149
4.3.5.2 Refinement of several flows of the waste stream
Fig. A11-28 takes ALPS slurry (multi-nuclide removal equipment slurry) as an example to show
the concept of refining multiple processing flows.

Fig. A11-4 Example concept of refining multiple processing flows (ALPS slurry) 149
Currently, the carbonate slurry and iron coprecipitated slurry are generated from ALPS andtemporarily stored in high-performance container HIC. For the next steps of pretreatment, treatment,
and reprocessing, several processing flows from (1) to (5) can be used. “Continue current situationof storage” corresponds to (1), “Prevent spread of leakage” to (2), “Separate container packaging
by segregation” to (3), “Moisture reduction” to (4), and “Store after additional solidification” to (5).
TEPCO’s solid waste storage plan already provides a plan for dehydrating solid waste using a filter

231press to reduce risks such as leakage, and to contain the dehydrated materials in containers andstore them in storage facilities. Here, a method for examining the validity of the concept of narrowing
down from several flows to (4) “Moisture reduction” based on the basic policy for solid waste
management and the required technical requirements is developed, leading to contributing tomaterializing the concept of such narrowing down.
4.3.5.3 Basic policy for solid waste management and required technical requirements
The technical requirements required by the International Atomic Energy Agency (hereinafterreferred to as the “IAEA”.) Safety Standards, GSR-Part 5, “Predisposal Management of Radioactive
Waste,” were summarized (Fig. A11-29).
According to Requirements 9, “Characterization and classification of radioactive waste,”characterization is cited as a technical requirement. Therefore, the determination of whether actions
at different stages of predisposal management are implemented should be made based on thecharacterization research results.
“Radiation safety for the public” and “radiation safety for workers” were derived from
Requirement 4, “Responsibilities of the operator”, and the corresponding technical requirements of“radiation protection and shielding” and “containment of radioactive materials” were developed,
respectively. “Preventing the spread of accidents involving the release of large amounts ofradioactive materials” and “maintaining containment” were derived from Requirement 11, “Storageof radioactive waste”, and the corresponding technical requirements were developed as “measuresagainst waste-specific accidents/events such as explosions, fires, leakage, and diffusion” and“securing boundaries,” respectively.
“Understand the impact on later stages, and have prospects for satisfying the acceptance criteria”
was also defined from Requirement 6 “Interdependences”, Requirement 10, “Processing ofradioactive waste” and Requirement 12 “Radioactive waste acceptance criteria.” Based on the
input chart and timeline chart, the interdependence between all stages in predisposal management
should be duly considered, as well as the influence of the anticipated disposal options.

232

IAEA Safety Standards GSR-Part 5“Pre-disposal Management ofRadioactive Waste”
Requirement 9
Characterization and classification ofradioactive waste

Requirement 4

Requirement 11

Responsibility ofoperators

Storage ofradioactive waste

Radiation safety
for the publicRadiation safety
for workers

Preventing the
spread of accidents
involving the releaseof large amounts ofradioactive materials

Maintaining
containment

Radiation
protection andshieldingContainment ofradioactive
materials

Measures against
waste-specific accidents
and events, such as
explosions, fires, andleakage or scattering

Securingboundaries(multiplexing throughrisk management)

Items that should be considered in individual designs (evaluation of exposure doses, methods for
detecting/controlling accidents, economic efficiency, etc.), common issues for 1F (natural disastercountermeasures, site planning)

Fig. A11-5 Summarization of required technical requirements149
4.3.5.4 Concept for narrowing down to storage of ALPS slurry
As ALPS slurry is contained in high-performance container HIC, and then stored at temporarystorage facilities, the risk of dispersion/leakage is kept sufficiently low. There was an event in thepast where radiolysis of water in a container containing high radiation-dose ALPS slurry causedhydrogen generation but it was stagnating without being diffused, resulting in volume expansion of
the slurry, and leakage of the supernatant water from the upper lid of the container. As acountermeasure, the flooding of supernatant water is controlled by lowering the water level in thehigh-performance container HIC. However, some slurry may still contain hydrogen, and even ifhydrogen is degassed, the risk of recurrence of events may remain in high-radiation dose slurry.
Therefore, further reduction of risk during storage will be considered in this study.
Fig. A11-28 shows an evaluation example where the technical requirements are applied toseveral processing flows to storage of ALPS slurry as shown in Table A11-14. Especially from theperspective of “Prevention of hydrogen retention in waste,” there would still be the risk thathydrogen would be contained in the current storage conditions. The evaluation showed that (4)“Moisture reduction” and (5) “Moisture reduction and solidification” would be promising as
countermeasures. Of these, the technical requirement for (5) “Understand the influence on laterstages, and have prospects for satisfying the acceptance criteria” is not satisfied, and thus (5) is
excluded because the elemental technology for solidification related to (5) is still in the stage of
research/development, and there are no prospects yet. Therefore, the options can be narrowed
down to (4) “Moisture reduction and storage in storage facilities.”
As described above, it was confirmed that it is possible to narrow down from several flows by
clarifying the required technical requirements.

233

The scope of this study is until storage, and development of a concept for the entire waste
stream including disposal is planned to be completed by the end of FY 2021.
Table A11-3 Processing options until storage, and evaluation examples149
No.

Processing options up
until storage

(1)
(3)

Prevention of hydrogen retention
in waste

ShieldingContainment

Securing boundaries

Continuation of current
storage conditions /Packaging and storage inseparate containers〇〇△ Slurry assimilates hydrogen△ (Temporary storage)(2)

Prevention of leakage
expansion(storage in storage facilities)〇〇△ Slurry assimilates hydrogen〇 Containers and storage
facilities(4)

Moisture reduction andstorage in storage facilities〇〇〇 Hydrogen is released from voidsof dehydrated material〇 Containers and storage
facilities(5)

Moisture reduction,solidification, and storage instorage facilities〇〇〇 Generally stable, but depends on
the solidification method〇 Waste materials themselves,containers, and storage
facilities

4.3.5.5 Action status by waste stream from generation to disposal, and prospects formeasures to be taken
Table A11-15 shows the items to consider from generation to disposal in each waste stream,
and the status of action. Currently, characterization is mainly promoted as shown on the left side of
the table. For some types of secondary waste generated by water treatment, examination is being
performed on low- and high-temperature processing as part of the elemental technology researchfor processing, and on the disposal strategies shown on the right side of the table. The area in graywill be addressed by acquiring data on waste properties.
As described in 4.3, the overall picture of solid waste to be disposed of, including waste that will
be generated in the future, will gradually become clear with the clarification of the progress andplans for fuel debris retrieval, contaminated water management, and other decommissioning works.
Therefore, it is necessary to repeatedly examine processing/disposal methods and safetyassessments, starting from waste for which properties have been clarified; to give consideration to
making processing/disposal methods more appropriate; and to accumulate knowledge to consider
safe and reasonable processing/disposal strategies for diverse solid waste collectively.
Based on these characteristics, characterization is insufficient for some waste streams, which
need the overall picture to be captured and examination to continue. Therefore, as a method forreasonably selecting a safe processing/disposal method for solid waste, examination will be
performed with the waste stream as a hub as shown in Fig. A11-12.

234

Table A11-154 Action status by waste stream from generation to disposal149

Waste acceptance criteria

Waste verification methods and reuse verification
methods

Waste verification methods

Disposal management

Disposal

Storag e before disposal

Material and radioactivity balances

Basic processes and specifications of equipment(including considerations for safety)

Specifications of delivered waste (including
methods of ensuring performance)

Feasibility of essential technology

Specifications of received waste

Waste conditioning treatment

Material and radioactivity balances

Basic processes and specifications of equipment(including considerations for safety)

Specifications of delivered waste (including
methods of ensuring performance)

Feasibility of essential technology

Specifications of received waste

Degree of segregation and container
specifications

Pretreatment, treatment, reprocessing

Physical properties

Chemical properties (including materials
impacting disposal)

Radioactive inventory

Amount of waste (including the possibility ofgeneration)

Waste

Stream Number

Waste

Dismantled waste
Pressure vessels
Metal from containment vessels

S1
S2

Concrete from containment vessels

S3

Metal in buildings

S4

Concrete in buildings

S5

RubbleMetallic rubble
Concrete rubble
Combustible materials (felled trees, protective clothing, etc.)

S6

Generation ongoing

S7

Generation ongoing

S8

Generation ongoing

Incineration

Secondary waste generated by water treatment
Adsorption vessels (1) (SARRY, KURION)
Adsorption vessels (2) (mobile purification equipment)

Generation ongoing

S9

Multi-nuclide removal equipment (1) (slurry)
Multi-nuclide removal equipment (2) (adsorbents)

S10

Multi-nuclide removal equipment (3) (treatment columns)

Low- and high-temperature treatment

Generation completed

Generation ongoing

Dehydration

Low-temperature treatment

Generation ongoing
Generation ongoing

Sludge generated by decontamination devices

S11

Generation completed

Filters

S12

Generation ongoing

Concentrated waste liquid

S13

Generation completed

Low-temperature treatment

Waste associated with debris retrieval
Top retrieval
Side retrieval
Contaminated soil

S14
S15

Waste generated during operation
Granular used resin and sludge
Metal (used control rods, etc.)
Solidified substances (pellets, etc.)

Generation completed

S16

Generation completed
Generation completed

Investigation completed (from generation to disposal)
Investigation partially completed (provide design specifications for some processes such as storage, treatment, and disposal)
Research in progress (design specifications have not been provided)
Not investigated (cannot be investigated due to a lack of input information, such as data on properties)
Investigation not started (investigation has not been started due to considerations for priority, etc.)

4.3.5.6 Summary
The following outcome was obtained as a result of the above examination:
The entire waste stream was examined in a comprehensive manner, and mutual feedback wasprovided on research results, using the input chart to develop methods for efficient examinationof the waste stream.

In an attempt to narrow down several flows, the technical requirements in the IAEA Safety
Standards, GSR-Part 5, “Predisposal Management of Radioactive Waste,” were summarized
and applied to narrow down the flow until storage through evaluation using it. Then, it was
confirmed that this could serve as one of the approaches for narrowing down.

As the scope of this study up to FY 2020 is until storage, development of a concept for
narrowing down for the entire waste stream including processing/disposal is planned to becompleted by the end of FY 2021.



For waste streams with insufficient characterization, the prospects for waste stream
development have been obtained by future characterization data, continuous R&D activities,
and in accordance with the examination method to integrate the outcomes.
As a result, with regard to waste streams for which processing technology has not yet beenclarified for practical application with consideration for disposal due to insufficient characterization,that it was possible to offer prospects for setting processing/disposal strategies using the series ofmethods developed could be offered.

235

5. Issues based on Technical Prospects and technical strategies toachieve themAfter providing for future issues based on the prospects of processing/disposal methods andtechnology related to their safety, the technical strategies by category to realize them are show
below.

5.1 Issues based on Technical Prospects
Volume reduction is extremely important for the safe and reasonable management of solid waste
according to the progress of decommissioning work in the future, so the measures in progressshould be continued steadily. Since solid waste continues to be generated, it is important tocontinuously examine further possibilities by referring to advanced cases of overseas’ for morevolume reduction. It is recommended to realize volume reduction in consideration of the expected
outcome and feasibility.
For the development of analytical/evaluation methods for efficient characterization, it isnecessary to improve evaluation methods and continuously incorporate them into solid waste
management, including processing/disposal, while accumulating analytical data using efficient
analytical methods established through achievements in research/development. In this case,efforts should be made for low-activity waste such as rubble, as well as high-activity waste such as
secondary waste generated by water treatment and waste generated from fuel debris retrieval,according to the characteristics of each type of waste.
To establish methods for selecting safe processing/disposal methods in a reasonable manner,
the methods shown in Fig. A11-12 shown in 4.3 should be used to proceed with examination toward
determining waste form specifications and manufacturing methods for Phase 3, as specified in theMedium-and-Long-term roadmap. Specifically, through these methods, the trial examples of
optimization/rationalization of processing/disposal methods will be accumulated by waste stream
according to the progress of characterization and under the assumption that safety is ensured towidely acquire findings on optimization by waste stream. Moreover, consideration will be given to
specify strategies for optimization/rationalization of the overall picture covering the entire waste
stream, allowing clarification of approaches toward such purposes. In doing so, it is important toconsider the most appropriate measures, taking the actual use and economic feasibility into
account by reflecting to the latest findings and applying the Best Available Techniques concepts.
As the examination progresses and the processing/disposal measures for the overall picture ofwaste are finalized, it will be important to share the examination process for optimization, such as
by sharing the awareness of problems with local communities and society.

236

5.2 Technical strategy by sectors
5.2.1 Characterization
For low-activity waste such as rubble, the analysis work itself is not so challenging, but animmense amount of time is required to measure entire quantity because of the enormous volume
of waste, and so there are needs for volume reduction and a corresponding efficient analysis
strategy. To that end, it is important to take an approach that efficiently ensures the required
accuracy. In order to achieve this, promote efficient analyses by making them simplified/speed-up,
and establish inventory evaluation methods that combine the DQO process with statistical methods.
For high-activity waste, there are difficulties inherent in sampling and analysis, and the amountof analysis data to be obtained is limited, which makes inventory assessment based on thetransition model more important. It is necessary to obtain actual sample data, such as by ongoing
efforts for sampling from cesium sorption vessels and its analysis, which is currently in progress.
The application of inventory evaluation methods, which combine the DQO process with statisticalmethods, and the priority of data to be collected should also be considered to enhance the accuracyof the transition model.
Following the phase of analyzing samples that are easy to collect, characterization is now in thephase of collecting/analyzing samples that are important for waste management. Going forward, it
is important to develop a medium-to-long-term analysis strategy that defines the solid waste to beanalyzed, its priority, and quantitative targets for analysis, etc., and to proceed withanalysis/evaluation accordingly. It is useful to accumulate trial results and verify their validity inorder to establish the flow from the development of a medium-to-long-term analysis plan using
statistical methods; analysis and data acquisition; the incorporation of the acquired data into
examination of processing/disposal methods and evaluation of the outcome; to the development of
the next medium-to-long-term analysis plan based on the evaluation results.
As for facilities for analysis, in addition to the existing facilities in the JAEA’s Ibaraki area, etc., itis planned to establish Radioactive Material Analysis and Research Facilities currently underconstruction, as well as facilities for analysis by TEPCO, allowing characterization of a variety ofsolid waste in parallel. Since the target nuclides, analysis items, accuracy, and the number ofsamples for analysis depend on the target solid waste are different, a structure should beestablished based on the appropriate division of roles and according to the characteristics offacilities.
5.2.2 Storage
For storage of all waste, it is important to reconsider measurement items and timing, etc., interms of diverse information for characterization, while acquiring necessary information throughcontinuous monitoring and surveillance of the storage status commensurate with the risks involved.
With regard to high-activity waste, such as waste generated from fuel debris retrieval, the issues
and countermeasures assuming the further expansion of the fuel debris retrieval scale have been237

clarified according to the results of research/development as of FY 2021. Going forward, reviews
should be performed along with the examination of the fuel debris retrieval methods. Reliablemeasures should be taken to ensure storage of the solid waste that is expected to be generated
during fuel debris retrieval (trial retrieval and gradual expansion of the retrieval scale) before fullscale retrieval.
The site also has solid waste stored since before the accident, and a large volume of dismantled
waste is expected to be generated after the completion of fuel debris retrieval. If the site only
increases the storage capacity for solid waste, it will eventually reach the limit, so efforts should bemade to reduce the volume of solid waste to be generated as much as possible.
In considering further possibilities of volume reduction, with the aim to reuse/recycle metals withan extremely low surface radiation dose, chemical decontamination (decontamination byphosphoric acid), physical (mechanical) decontamination (steel blasting), and decontamination bymelting (decontamination by melting slag) are under consideration as metal decontamination
methods for recycling.
As metal recycling with decontamination by melting slag has already been used in many Westerncountries, it is considered a promising candidate technology. Thus, it is important to focus on theareas where the conditions are different between such Western countries and Fukushima Daiichi
NPS (target nuclides, etc.), and to evaluate the applicability of the method.
5.2.3 Processing/disposal
The objective is to establish safe and reasonable processing/disposal methods for all solid waste
in which diverse waste streams exist, and to widely obtain knowledge for optimizing each individual
stream. Therefore, it is necessary to continue the development/research of processing/disposal
technologies required for the series of studies shown in Fig. A11-12.
Regarding the processing technology, outstanding issues in low- and high-temperatureprocessing technology, for which research/development is promoted, should be addressed. Waste
streams for which the possibility of solidification has not been investigated will be evaluated asnecessary, and performance such as leachability for solidified substances to be produced will be
evaluated. As for low-temperature processing technology, consideration is given to thetransformation of solidified substances as well as inspection methods to verify the possibility ofsolidification. In the case of high-temperature processing technology, the feasibility of the wholeprocessing system, including supply and exhaust systems, is an issue, in addition to thesolidification process, and therefore it is necessary to carry out examination in a timely manneraccording to the start time of processing. Furthermore, in order to expand technological options, it
is important to examine the possibility of low-temperature solidification after interim treatment suchas steam reforming.
Regarding disposal technologies, in order to establish reliable safety assessment techniques,
important issues specific to solid waste at the Fukushima Daiichi NPS will be explored and identified
based on the understanding of the sensitivity structure of parameters to radiation dose and the238long-term transition behavior of disposal facilities. Then, priorities will be examined andincorporated into research plans. Development and improvement of the proposed disposal options
will also be promoted in reference to case examples in Japan and abroad. Furthermore, the targetof waste streams on which disposal safety assessments will be performed will be expanded, groupof the disposal options and categories from a bird-eye-viewpoint of all solid waste at Fukushima
Daiichi NPS will be examined, and contributions will be made to considering appropriate measuresfor overall waste management in coordination with areas other than disposal, such as presenting
targets for waste form performance and the accuracy required for characterization.
The main technical issues and plans in the future described above are summarized as shownin Fig. A11-30.
The Mid-and-Long-Term Roadmap states that properties of solid waste will be analyzed andspecifications of waste form and their production method of conditioned waste will be determinedin the third phase. As a systematic approach toward this goal, further examination will be conductedto present appropriate measures for overall management of solid waste in the (1) of third phase.
Specifically, the first step is to develop processing/disposal options by examining theunimplemented issues related to processing technology, interim treatment, and disposal options.
Then, the options will be compared and evaluated using the property data and other data that arebecoming clear, and the examination will be conducted to identify waste streams suitable for thecharacteristics of solid waste.

239

Fiscal year

2021

2022

2023

2026

2031 (Year)

On-site work
Technical studies for on-site construction

Related Milestones

Research and development

Processing/disposal methods
and technical prospects
regarding their safety

Outdoor storage will be eliminated completely

1. Waste characterization(1) Acquisition/managementof analysis data

Sampling
Development of sampling technology forhigh activity samples
Analyzing

Development/establishment of waste confirmation method
Analyses and measurements
for waste confirmation(2) Enhancement of efficiency
on waste characterization

Investigating the applicability of statistical inventory estimation methods and improving their accuracy
Study on analysis plan (Developing mid-to-long-term/yearly analysis plan, study on nuclides to be analyzed,
lower limit of quantification and number of samples to be collected, etc.)(3) Improvement of analysis
capacity

Development of Radioactive Material
Analysis and Research Facilities
Human resources development for analysis

Storage

2.Storage(1) Storage Management Plan

(2) Study and evaluation ofstorage methods(3) Reduction in volume

Development and review of the Storage Management Plan

Review according to fuel debris retrieval (Evaluation of type, volume, etc.)

Reduction of carry-in, re-use, recycling, volume reduction treatment, etc.

Development of technology for volume
reduction/recycling

3. Processing/disposal(1) Processing technology

Development of processing technology, Evaluation of long-term stability of solidification

Development of interim treatment technology(2) Disposal technology

Creating disposal concept suitable for solid waste

Improving reliability of safety assessment technology(3) Study on waste stream

Identifying waste stream
suitable for solid waste
Building a methodlogy for safetyassurance

Study on appropriate measures for entire waste management

Fig. A11-30 Main technical issues and future plans on waste management (progress schedule)

240

Attachment 12 Efforts toward discharging ALPS-treated water into the ocean

Table of Contents

1. Introduction ............................................................................................................................... 242
2. Significance of the policy for disposing of ALPS-treated water ............................................... 243
3. Discharge of ALPS-treated water into the ocean..................................................................... 246
3.1 Background of announcing the disposal policy .................................................................. 246
3.2 Impact of tritium contained in the water discharged to the ocean on the human body ..... 248
3.3 Case examples of discharge in Japan and overseas, and regulatory standards .............. 251
3.4 System and operation for offshore discharge .................................................................... 256
4. Future efforts ............................................................................................................................ 275

241

1. Introduction
On April 13, 2021, under the overriding principle of Balancing between Reconstruction andDecommissioning, and on the premise of ensuring safety and implementing comprehensive
measures to prevent reputational damage, the Government announced the basic policy fordischarging ALPS-treated water into the ocean from the Fukushima Daiichi Nuclear Station (1F)
after comprehensive discussions at expert meetings for more than six years155,156. In addition, theGovernment requested TEPCO to proceed with preparations, including the installation of a specificdischarge system, aiming to start the discharge of ALPS-treated water into the ocean in about two
years157.
In response to this policy, on April 16, 2021, TEPCO indicated their approach to ensuring safety
through further efforts and in compliance with regulatory standards pursuant to laws andregulations; minimizing reputational damage; providing compensation in the event of reputational
damage; and addressing issues for the future158. TEPCO also working to provide a briefing forstakeholders, and to obtain permission for the implementation plan.
This Attachment describes the significance of the policy for disposing of ALPS-treated water,and summarizes past discussions and studies from the viewpoints of the background of announcing
the disposal policy; the effects of tritium on the human body; examples of discharge cases andregulatory standards in Japan and overseas; and systems and operation associated with discharge
into the ocean. It also provides technical points for achieving offshore discharge.

155 Tritiated Water Taskforce Report, June 3, 2016
156 Report by the subcommittee dealing with water treated with multi-nuclide removal equipment, February 10, 2020
157 Inter-Ministerial Council for Contaminated Water, Treated Water and Decommissioning Issues (5th meeting),

Material 1, “Basic policy for disposing of treated water by multi-nuclide removal equipment at the TEPCO
Fukushima Daiichi Nuclear Power Station (draft)”, April 13, 2021
158 TEPCO, Attachment 1. Actions by TEPCO in response to the Government’s basic policy for disposing of treatedwater by multi-nuclide removal equipment, press release, April 16, 2021

242

2. Significance of the policy for disposing of ALPS-treated water
Waste liquid generated from nuclear reactor facilities in Japan is allowed to be discharged to theenvironment in a controlled manner “by reducing the concentration of radioactive materials in the
discharged water as much as possible by filtration, evaporation, sorption by the ion-exchange resin
method, etc., decay of radioactivity with time, and dilution with a large amount of water, and by
preventing its concentration from exceeding the concentration limit established by the NRA” 159,160.
This operation is in accordance with the concept of “Dilution and Dispersion” as specified in
Publication 81161, the International Commission on Radiological Protection (ICRP) and the General
Safety Requirement, Part 5, International Atomic Energy Agency (IAEA)162163, and follows the guideof the IAEA General Safety Requirements, WS- GSG-9164. These guidelines require monitoring inorder to ensure that discharged water concentrations meet the regulatory limits and that there are
no unanticipated discharged. This controlled discharge avoids storing large amounts of waste liquidin the facility, enabling safe operation and management of the facility.
At the Fukushima Daiichi NPS, on the other hand, most of the radioactive materials contained in
the contaminated water generated by the accident is removed by multi-nuclide removal equipment,
and then the treated water is stored in tanks located on higher ground as ALPS-treated watercontaining large-volume tritium (ALPS-treated water165, water under treatment166). As most of thecontaminated water is stored in welded tanks, the possibility of leakage is very low, but the risk is
not zero. At present, however, the number of these tanks exceeds 1000, which makes the spaceon the plant site tight and poses a serious problem on the sustainability of decommissioning work
ahead. In particular, the following issues must be solved.
● Contaminated water is continuously generated167 by various factors for the time being, andeventually it becomes difficult to secure a transfer destination, which hinders decommissioningwork.
159 Ministerial Ordinance for Commercial Nuclear Power Reactors concerning the Installation, Operation, etc., Article

90, Ordinance of the Ministry of International Trade and Industry 77 of 1978

160 Article 16, Rules on Safety and Physical Protection of Specific Nuclear Fuel Materials in Reactor Facilities at theTEPCO Fukushima Daiichi Nuclear Power Station, Ordinance 2 of the Nuclear Regulation Authority, 2013

161 ICRP, Radiation protection recommendations as applied to the disposal of long-lived solid radioactive waste. ICRP

Publication 81. Ann. ICRP 28 (1998).

162 IAEA, Predisposal Management of Radioactive Waste, IAEA Safety Standards, No. GSR-Part5, Vienna (2011)
163 Compliance with the concept of “Dilution and Dispersion” means verifying that the concentration of radionuclides

is at a level that does not have a significant health impact, and then discharging ALPS-treated water into theenvironment where dilution is expected to occur
164 IAEA, Regulatory Control of Radioactive Discharges to the Environment, IAEA Safety Standards, No. GSG-9,
Vienna (2018).
165 Water that is treated with multi-nuclide removal equipment until the level of radioactive materials other than tritium
falls below the regulatory limit on safety, without fail (the sum of ratios of concentrations required by law, except
for tritium, is less than 1).
166 Water that is treated with multi-nuclide removal equipment, and does not meet the regulatory limit on safety (thesum of ratios of concentrations required by law, except for tritium, is 1 or higher).
167 Various factors include inflow of rainwater/groundwater into buildings, transfer from the ground at T.P.+2.5 m,
injection of chemicals during purification by ALPS, transfer along with decommissioning work, and emergency
transfer. The target of reducing the amount of contaminated water generated in 2025 is 100 m3/day or less. Referto “3.3 Contaminated and treated water management” in the main document for details.

243

● The risk of leakage from tanks due to aging or natural disasters increases with long-term
storage.
● It is impossible to secure enough space for systems and facilities necessary for stable,
temporary storage of spent fuel as well as safe fuel debris retrieval.
As such, simply increasing the number of tanks does not solve the fundamental problem Note).
Therefore, the Government's basic policy has the following significance.
● A risk of not being able to secure a transfer destination for contaminated water can beeliminated permanently, and the sustainability of decommissioning work can be ensured.
● Presenting an outlook for the disposal of ALPS-treated water can result in reducing the risk
caused by long-term storage of more than 1,000 tanks.
● Along with more effective use of the entire site, resources spent for tank inspection andmaintenance management due to the prolonged storage of ALPS-treated water can beconcentrated on more risky decommissioning work.
The announcement of the basic policy for the disposal of ALPS-treated water ensures thesustainability of the overall decommissioning work and contributes to its steady progress.
Though the basic policy for 1F decommissioning is “to continually and quickly reduce the risks
associated with the radioactive materials caused by the accident and that do not exist in normal
nuclear power plants”, this disposal policy intends to bring the risk level of the water stored inwelded tanks, including ALPS-treated water, from the “Sufficiently stable management” region into
the level that does not require storage itself as a permanent measure.

244

Note) Options for waste liquid disposal
Article 16, “Rules on Safety and Physical Protection of Specific Nuclear Fuel Materials in
Reactor Facilities at the TEPCO Fukushima Daiichi Nuclear Power Station” specifies as follows:
6

Radioactive waste liquid shall be disposed of by any of the following methods:(a) Discharge by drainage facilities.
(b) Store and dispose of in waste liquid tanks with radiation hazard prevention.
(c) Enclose in containers or solidify with containers, and then store and dispose of instorage/disposal facilities with radiation hazard prevention.
(d) Incinerate in incineration facilities with radiation hazard prevention.
(e) Solidify in solidification systems with radiation hazard prevention.

7

In the case of disposal by the method described in (a) of the preceding item, theconcentration of radioactive materials in the discharged water shall be reduced as much as
possible by filtration, evaporation, sorption by the ion-exchange resin method, etc., decay ofradioactivity with time, and dilution with a large amount of water in drainage facilities. In thiscase, the concentration of radioactive materials in the discharged water at drainage outlets andin discharged water monitoring systems shall not exceed the concentration limit specified by the
NRA.
On the other hand, contaminated water is stored in tanks as ALPS-treated water after theconcentration of radioactive materials is reduced by filtration, and sorption by the ion-exchange
resin method, etc., in a water treatment system. Of the above, (b) means transferring liquid fromtanks to waste liquid tanks, and “underground burial” compared/evaluated by the Tritiated Water
Taskforce falls under this case. (c) and (e) can reduce the risk of leakage by solidification, butconstructing additional storage facilities is required. (d) is a measure for organic solvents, etc.
Based on the above, a remaining option for ALPS-treated water management is (a). However,when implementing (a), Article 16, Item 7 of the aforementioned Rule should be observed.

245

3. Discharge of ALPS-treated water into the ocean
3.1 Background of announcing the disposal policy
For the purpose of evaluating various options for the handling of ALPS-treated water, the“Tritiated Water Task Force155” was established under the Committee on Countermeasures for
Contaminated Water Treatment, and started examination from December 25, 2013. This Task
Force consolidated scientific information on tritium as basic data for determining the long-term
handling of ALPS-treated water, and reviewed basic requirements (regulatory and technologicalfeasibility) and possible constraints (period, cost, scale, secondary waste, work exposure, etc.) for
the five disposal methods of geosphere injection, offshore discharge, vapor discharge, hydrogen
discharge and underground burial). Concerning the tritium separation technique, a confirmation
test project on the tritium separation technique was implemented in FY 2015. The conclusion was,“Taking the amount and concentration of ALPS-treated water into consideration, there was notechnique that could be put to practical application immediately.” These results were submitted ina report dated on June 3, 2016.
Based on the findings compiled in the Task Force Report, the Committee on Countermeasures
for Contaminated Water Treatment decided to establish “The subcommittee on handling of ALPStreated water (ALPS Subcommittee)156” with the aim of comprehensive review of the handling of
ALPS-treated water from a social perspective, including reputational damage, and started
examination from November 2016. In August 2018, public briefing/hearing sessions were held in
Fukushima Prefecture and Tokyo to hear opinions on how to dispose of ALPS-treated water, as
well as concerns about disposal. Then, discussions were held on the issues raised, while verifying
the facts from a scientific perspective. Subcommittee meetings were held 17 times over a period of
about three years and, on February 10, 2020, a report was compiled as a reference for theGovernment to evaluate the ALPS-treated water disposal method.
In this report, the proven methods of “vapor discharge” and “offshore discharge” were identified
as realistic options. Despite the difference in scale, there is a precedent for “vapor discharge” froman accident reactor. Vapor containing radioactive materials is discharged in a controlled manner
even from normal reactors during ventilation. For the purpose of disposing of radioactive waste
liquid, however, it was pointed out that there was no case in Japan, where waste liquid wasevaporated to gas and discharged as vapor. As for “offshore discharge”, on the other hand, it was
concluded that this method would be more reliable for reasons that radioactive waste liquidcontaining tritium has been diluted with seawater used for cooling and discharged into the ocean
from nuclear facilities in Japan and abroad, and in terms of track records in normal reactors, easeof handling the discharge system and monitoring methods. However, it was noted that thequantitative relationship between the amount of discharged water and that of tritium discharge
would not be equivalent to that before the accident at the Fukushima Daiichi NPS. In addition, froma social point of view, the offshore discharge could have a negative impact on the fishery and246tourism industries in Fukushima Prefecture and the surrounding sea areas. Especially, the fish
catch from the trial fishing in Fukushima Prefecture has not even recovered to 20% of the level
before the Earthquake, and it is necessary to consider countermeasures based on this situation.
Concerning the tritium separation technique, while keeping the conclusion that there is notechnique that could be put to practical application immediately, studies on new techniques areunderway. Thus, it has been determined that continued attention should be paid to technologicaltrends. The International Atomic Energy Agency (IAEA) has evaluated that the conclusion of the
ALPS Subcommittee mentioned above is based on scientific and technical grounds168.
While receiving recommendations from the ALPS Subcommittee, the Government then selected“offshore discharge” as the basic policy, based on opinions obtained through exchanges with local
governments and agricultural, forestry and fishery operators, public hearings, and written requests
for public comment, in terms of past results of offshore discharge in Japan, and potential of reliable
and stable monitoring. This conclusion is supported by the IAEA, as “The offshore discharge istechnically feasible and consistent with international practices” and “The offshore discharge in acontrolled way is constantly performed in countries operating nuclear power plants around theworld”169.
As described above, “offshore discharge” was announced by the Government as the disposal
policy after comprehensive discussions at expert meetings and based on the opinions of a wide
range of people. The IAEA, as a third party, has evaluated the policy as reasonable169.
On the other hand, it is also a fact that there have been concerns about reputational damage dueto the discharge of ALPS-treated water into the ocean. Therefore, efforts should be continued toobtain understanding to eliminate such concerns. In addition, the reliability of TEPCO has declined
due to inappropriate incidents in terms of physical protection of nuclear materials at the
Kashiwazaki-Kariwa Nuclear Power Station and insufficient provision of information during
earthquakes at the Fukushima Daiichi Nuclear Power Station. TEPCO needs to take this reality
seriously and respond more carefully than before. With this matter, for example, the lack of twoway communication may be one of the reasons for insufficient understanding from local
stakeholders. One of the possible factors of the concern from commercial distributors about
reputational damage is that information to suppress a negative impact has not been sufficiently
provided. In addition, South Korea and China have expressed concern and criticism of Japan’s
policy, we need to continue our efforts to disseminate accurate information based on scientificevidence to the international community, such as clearly communicating the difference between
contaminated water and ALPS-treated water. Therefore, greater transparency is required, forexample, by repeatedly providing explanations in an easy-to-understand and careful manner,

168 IAEA Review Report (translation for reference), MOFA website,

https://www.mofa.go.jp/mofaj/dns/inec/page24_001364.html, April 2020
169 Statement by IAEA Director General Grossi, MOFA website,

https://www.mofa.go.jp/mofaj/dns/inec/page24_001364.html, April 2021

247

mainly by TEPCO, in order to increase understanding of the basics for safe offshore discharge, i.e.,(1) an operation plan for offshore discharge; (2) the effects of tritium contained in the water to bedischarged to the ocean on the human body; and (3) the method for verifying the operation status,and by verifying safety through reliable third parties such as IAEA in cooperation with organizations
concerned, and by delivering accurate information.

3.2 Impact of tritium contained in the water discharged to the ocean on thehuman body
Tritium (3H, T) is a radioisotope of hydrogen whose nucleus consists of 1 proton and 2 neutrons,
and decays to 3He with a half-life of 12.3 years. At that time, it emits faint β-rays (18.6 keV at
maximum, 5.7 keV on average), but the maximum range is very short at 5 mm in air and 6 μm in
water. On earth, 70 quadrillion Bq of tritium is produced annually by nuclear reaction of neutrons
produced by cosmic rays with nitrogen and oxygen in the atmosphere170. Tritium exists primarily aswater molecules, and is contained in any water, including water vapor in the atmosphere, rainwater,seawater, tap water, and even in the human body. Water molecules containing tritium (tritiatedwater) have the same chemical properties as ordinary water molecules, and tritium is notconcentrated in specific living organisms or organs. The human body also contains tens of Bq oftritium due to ingestion of drinking water.
In nuclear power plants and nuclear fuel reprocessing plants, tritium is produced by ternary
fission of 235U in fuel, and by reaction of deuterium (2H, D) and lithium contained in cooling wateras water molecules, and boron and neutron in control rods. Since tritium exists as water, unlike
other radioactive materials, it is technically difficult to recover and contain, and the effect of β-rays
from tritium on the human body is very small compared with other radioactive materials. Therefore,
tritium is discharged into the ocean, rivers, lakes/reservoirs, and the atmosphere according to theregulations of each country.
Fig. A12-1 shows the result of evaluating the tritium concentration in surface seawater based on
the “Long-term assessment model of radionuclides in the sea (LAMER) 171 ” 172 . As a result of
atmospheric nuclear tests conducted in the 1950s - 1960s, it is estimated that (1.8 - 2.4) × 1020 Bq
was discharged into the environment170. In 1963, the tritium concentration in surface seawater isestimated to have risen to a level exceeding 1 Bq/L. Subsequently, the tritium concentration causedby nuclear tests decreased year by year due to the decay of the radioactive materials and transfer
to seawater, and it is considered that the tritium concentration has recently fallen below 0.1 Bq/L.
The concentration of naturally occurring tritium is estimated to be about 0.02 Bq/L.

170 Hideki Kakiuchi, Tritium in the environment and its evaluation methods, Journal of the Atomic Energy Society ofJapan, Vol. 60, No. 9, pp.537-541 (2018)

171 Masanao Nakano, LAMER; Long-term assessment model of radionuclides in the ocean, JAEA-Data/Code 2007-

024(2008)
172 Masanao Nakano et al., Tritium Concentration and Diffusion in Seawater Discharged from Tokai Reprocessing

Plant, Japanese Journal of Health Physics, 44(1), pp.60 - 65 (2009)

248

(Source : Masanao Nakano, et al., Concentration and diffusion of tritium in seawater discharged
from Tokai Reprocessing Plant, Physics in Health)Fig. A12-1 Changes in tritium concentration in surface seawater172
UNSCEAR 2016 Report, Annex C - Biological effects of selected internal emitters - Tritium
complies a wealth of systematic information on the effects of tritium on the human body. As a modelfor exposure assessment, this report indicated a metabolism model that takes into account the factthat a portion of tritiated water changes its form to organically bound tritium (OBT) when ingested
into the human body173. Tritiated water has an average biological half-life of about 10 days and iseliminated from the body, but about 3 - 6% of tritiated water changes to OBT in the body. The
biological half-life of OBT is estimated to be about 40 days for the short half-life component andabout 350 days for the long half-life component. Therefore, the committed effective radiation doseof OBT is estimated to be about 2 - 5 times that of tritiated water.
As described above, in the tritium exposure assessment model, tritiated water and OBT derived
from tritiated water discharged from nuclear facilities are evaluated as not concentrating in specificliving organisms or organs, although the period until they are eliminated from the body is longer1773.
In the UK, tritium concentrations in bivalves and flounders have been found to be several 1000
times higher than in seawater174. However, subsequent investigations have revealed that this isdue to artificially synthesized tritiated organic compounds contained in the discharged water froma nearby chemical plant175 that produces research medicines containing radioactive materials176,177,
tritium discharged from nuclear power plants is not the cause.
173 UNSCEAR 2016 Report, Annex C - Biological effects of selected internal emitters-Tritium

174 Environment Agency, Food Standards Agency, Northern Ireland Environment Agency and Scottish Environment

Protection Agency, Radioactivity in Food and the Environment 2002, RIFE-8 (2003)
175 In this chemical plant, various tritiated organic compounds such as amino acids, fatty acids, and carbohydrates

labeled with tritium were produced for medical research, etc.

176 McCubbin D. et al., Incorporation of organic tritium (3H) by marine organisms and sediment in the Severn

Estuary/Bristol Channel (UK) Mar. Poll. Bull. 42 852–63 (2001)
177 A. Hunt et al., Enhancement of tritium concentrations on uptake by marine biota: experience from UK coastal

waters, J. Radiol. Prot. 30, 73–83(2010)249

A correlation between tritium discharges and newborn infant mortality was reported178 in local
governments near a nuclear power plant in Ontario, Canada, from 1974 to 1985. However, theCanadian nuclear regulatory authorities subsequently expanded the data from 1971 to 1988 foranalysis, and found no correlation between tritium discharges and newborn infant mortality179.
Moreover, there have been no reports of effects on the human body that could be caused by tritiumin the vicinity of nuclear facilities that discharge tritium156.
In France, the regulatory authorities issued the Tritium White Paper in coordination with the Local
Information Committee (CLI: Commission Local d'Information), including fishery workers, and byincorporating the views of various stakeholders, including operators, local residents, environmentalgroups, and congress members180. The White Paper confirms that tritium has little impact on thehuman body, while providing an action plan on tritium concentration measurement, discharge
control, and environmental monitoring, which is constantly checked by parties concerned including
above members181. In France, these activities by CLI have continued for over 40 years, and played
an important role in helping local residents understand the effects of tritium on the human body andthe environment.
Table A12-1 shows a list of reports on the effects of tritium on the human body. The Tritium
Taskforce155 and ALPS Subcommittee156 refer to updated status in the world. In addition, opinions
received at public briefings/hearings on the handling of treated water by multi-nuclide removal
equipment are discussed based on scientific information, and their findings are summarized156156.
Examples of typical questions and answers on the effects of tritium on the human body are shownin Table A12-2.
Using the exposure assessment model presented in UNSCEAR 2016182, the effects of radiation
were estimated in the case of continued disposal of all ALPS-treated water stored in tanks every
year. The result showed that the radiation exposure dose associated with offshore discharge would
be approximately 0.071 - 0.81 μSv per year, which is less than 1/1,000 of the natural exposure of
2.1 mSv (2,100 μSv) per year183.

178 McArthur, D., Fatal Birth Defects, Newborn infant fatalities and tritium emissions in the town of Pickering,

Ontario: A Preliminary Examination. A study by David McArthur, Toronto, Ontario, November 30 (1988)
179 Tritium releases from the Pickering Nuclear Generation Station and birth defects and infant mortality in nearby

communities 1971-1988, Atomic Energy Control Board, Canada (1991)

180 ASN, The Tritium White Paper (2010)

181 Tritiated Water Taskforce Report (7th meeting), meeting minutes, April 9, 2014
182 UNSCEAR 2016 Report, SOURCES, EFFECTS AND RISKS OF IONIZING RADIATION

183 17th ALPS Subcommittee meeting, Material 3-2, Explanation on the UNSCEAR 2016 model, January 31, 2020250

Table A12-1 Reports on the effects of tritium on the human bodyTime ofpublication

Publishing organization

Document name2007

International Commission on
Radiological Protection
ICRP Publication 103(ICRP)

2007

Health Protection Agency(HPA)

Reviews of risks from Tritium

2010

Canadian Nuclear Safety
Commission (CNSC)

CNSC 2010 Report

2010

2016

French Nuclear Safety
Authority
The Tritium White Paper(ASN)
United Nations Scientific
UNSCEAR 2016 Report
Committee on the Effects ofAnnex C. Biological effects ofAtomic Radiation
selected internal emitters - Tritium(UNSCEAR)

Objective
To update radiation and tissue weighting factors at
equivalent and effective doses based on the latest
scientific information in biology and physics onradiation dose exposure, and thereby update
information on radiation damage.
To provide scientific opinions on internal radiation
exposure doses and risks due to tritium, and to accountfor diverse point of views.
To perform an independent review of the scientific
literature to assess the health risks to workers and the
public caused by tritium exposure, to evaluate Canadianand international dosimetry on tritium intake, and toreview approaches to limit tritium exposure.
To provide detailed analysis of evaluation methods forthe environmental dynamics of tritium and biologicaleffects of tritium on the human body.
To underpin radiation risk assessment and internationalradiation protection standards with the scientificknowledge compiled by UNSCEAR.

Table A12-2 Example questions and answers on the effects of tritium on the human body156
Q. Does tritium bioaccumulate?
A. Tritium does not bioaccumulate because it has the same properties as ordinary water.
Q. If tritiated water changes to organically bound tritium (OBT), does it become a harmful substance?
A. At about 3 - 6%, the rate of change to OBT in the body is small, but as the biological half-life becomes longer in the body, the healtheffects are likely to be two to five times greater. However, the health effects of tritium are originally as small as 0.000000019 per Bq, and it
cannot be said that the health effects are especially large compared with other radioactive materials. This is about 1/300 of the effect on
health from cesium.
Q. If the tritium atoms that compose genes change into helium atoms, will the genes be damaged?
A. Genes are constantly damaged and repaired daily by repair enzymes. The damage received by radiation of about 2 mSv per year is verysmall, less than 1 million times the frequency of the damage from the ultraviolet rays of the sun.
Q. Does tritium intake increase carcinogenicity?
A. In experimental carcinogenesis on mice, it has been reported that the incidence of cancer is within the range of the spontaneous incidence
rate even if mice continue to drink tritiated water at a concentration of approximately 140 million Bq/L, and therefore the effect is small.
Q. Are there any examples of damage to health that could be caused by tritium in the vicinity of nuclear facilities that discharge tritium?
A. No examples of possible tritium-induced effects have been found.

(Source : METI)

3.3 Case examples of discharge in Japan and overseas, and regulatory
standards(1) Current situation of discharging radioactive waste liquid containing tritium in Japan andabroadRadioactive waste liquid generated from nuclear facilities is allowed to be discharged to theenvironment in a controlled manner by reducing the concentration of radioactive materials as muchas possible by removing and decay of the radioactive materials and dilution with a large amount ofwater, and by satisfying the regulatory standards in various countries. There are many caseexamples of discharge in Japan and abroad159,164.
Fig. A12-2 shows the annual discharge of tritium from major nuclear power plants andreprocessing facilities in Japan and abroad. Based on the characteristics of each facility, the annual
discharge tends to be: Reprocessing plants > Heavy water reactors/advanced gas reactors >251Pressurized water reactors > Boiling water reactors. In nuclear power plants in Japan, about 18 83 trillion Bq/year of tritium is discharge d into the ocean from pressurized water reactors, andabout 31.6 billion - 1.9 trillion Bq/year from boiling water reactors. Since deuterium and boron arecontained in the core cooling water of heavy water reactors and pressurized water reactors, theamount of tritium generated is larger than that of boiling water reactors. At 1F, before the accident(FY 2010), about 2.2 trillion Bq of tritium was discharged into the ocean annually.
In foreign countries, tritium is discharged from nuclear facilities as in Japan. From the La Hague
Reprocessing Plant in France, about 10 quadrillion Bq of tritium is discharged into the ocean
annually. Fig. A12-3 shows examples of liquid tritium discharge from neighboring Asian countries
and regions. Some sites discharge around 100 trillion Bq per year184.

Fig. A12-2 Annual discharge of tritium from nuclear facilities in Japan and abroad185

184 Detailed explanatory material on ALPS-treated water, available on METI’s website

https://www.meti.go.jp/earthquake/nuclear/hairo_osensui/alps.html, April 2021

185 The 1st Expert Meeting for marine monitoring on ALPS-treated water, Reference 2, June 18, 2021

252

Fig. A12-3 Case examples of annual discharge of tritium in neighboring Asian countries/regions184
These nuclear facilities discharge tritium by means appropriate to the characteristics of eachfacility in order to minimize the radiological impact on the surrounding environment as much as
possible. Fig. A12-4 shows the treatment flow up to discharging waste liquid containing tritium innuclear power plants in Japan. Waste liquid generated in a power plant is [purified] to reduce theconcentration of radioactive materials in the waste water as much as possible by filtration,
evaporation, sorption by the ion-exchange resin method, etc., and decay of radioactivity with time.
The purified liquid is collected in a sample tank where the concentration of the radioactive materialis [analyzed and verified] by monitoring and sampling. Subsequently, waste liquid containing tritiumis mixed with seawater used for cooling and then discharged to the environment.
As such, [purification] → [analysis/verification] → [discharge] is the basic process, and thedischarge is in compliance with the regulatory standards on the concentration of radioactive
materials. In nuclear power plants in Japan, radiation monitors are installed to measure γ-rays
before radioactive materials are discharged into the environment to monitor concentration.

［ Purification ］［Analysis
Verification］［Release］

Environmental
ReleaseWaste liquidstorage tanks

Purification
devices

Sample tanks

Seawaterfor cooling

Source : Prepared by NDF with reference to materials on domestic nuclear power plants

Fig.A12-4 Treatment flow up to discharging waste liquid containing tritium in nuclear facilities

253

(2) discharging pumped water from sub-drains at 1F into the ocean
At 1F after the accident, in order to reduce the amount of groundwater flowing into the reactor
buildings and to reduce the amount of contaminated water generated, the groundwater pumped up
from the sub-drain installed in front of the buildings, and the groundwater pumped up from thegroundwater drain installed in the revetment area to prevent overflow of the groundwater by seaside impermeable walls, are purified before being discharged into the ocean (Fig. A12-5). At thattime, the groundwater pumped up from the sub-drain and groundwater drain is purified, and its
concentration is measured to confirm that the concentration is less than the target concentration in
all tanks before being discharged. As in (1), the process of [purification] → [analysis/verification] →[discharge] is applied.

(TEPCO material edited by NDF)

Fig. A12-5 Flow from pumping from sub-drains to discharge 186(3) Regulation and management of tritium in nuclear power plantsa.

Regulation on concentration
Regulations on radioactive materials in Japan are provided in accordance with therecommendations of the International Commission on Radiological Protection (ICRP) (basis forinternational laws and regulations). According to the ICRP, 1990 Recommendations, “the public

186 Overview of wells (sub-drains) near the buildings, TEPCO website.

https://www.tepco.co.jp/decommission/progress/water management/subdrain/index-j.html

254dose during normal times should be less than 1 mSv per year”. The regulatory standards for nuclear
power plants in Japan specify the concentration limits (concentration limits required by law) ofradioactive materials contained in liquid and gas to be released to the environment according to
the type. The concentration limit in water required by law is the concentration at which the averageradiation dose rate reaches 1 mSv per year when a person continues to drink approximately 2 litersof water at the discharge outlet every day from birth until age 70. The concentration limit in air
required by law is the concentration at which the average radiation dose rate reaches 1 mSv per
year when a person continues to breathe air at the site boundary every day from birth until age 70.
Based on these standards, the concentration limits of tritium required by law are 60,000 Bq/L inwater and 5 Bq/L in air. However, the target concentration of tritium in the water generated bypurifying the groundwater pumped up from sub-drains at 1F is set at 1,500 Bq/L, one-fortieth of theconcentration limit required by law (60,000 Bq/L), which is sufficiently low.
In the US, the liquid concentration is defined at 37,000 Bq/L and the air concentration (water
vapor) at 3.7 Bq/L, both of which are evaluated so that the radiation exposure dose does not exceed
1 mSv/year, taking into account the daily habits and other factors in the US. Similar values (in water:
40,000 Bq/L, in air: 3 Bq/L) have been established in South Korea. France, the UK, and Canada,
on the other hand, have not established uniform concentration limits.
International standards for drinking-water, including those of the World Health Organization, areshown in Table A12-3170. For example, the WHO Guidelines for Drinking-Water Quality specifies
the concentration of 10,000 Bq/L as the guidance level, which would result in the 0.1 mSv annual
dose if ingested for a year (730 L/year: Equivalent to 2 L/day). There are no specific regulatory
standards for tritium in drinking water and food in Japan (The regulatory limit is specified for theconcentration of tritium at the time of discharge for control). The ICRP recommends an annual
radiation dose limit of 1 mSv for public exposure under normal conditions. Based on thisrecommendation, each country has established the limits with different margins170. The EU has
defined a tritium concentration of 100 Bq/L as a screening value, requiring to survey the presenceof artificial radionuclides other than tritium, if exceeded.
Table A12-3 Concentration limits of tritium in drinking-water and exposure radiation doses if
ingested for a year170
Country/Organization*

Tritium concentration limits (Bq/L)

Exposure dose (mSv/yr)

USA

740

0.01

Canada

7,000

0.09

Russia7,700

0.1

Switzerland

10,0000.13

WHO10,0000.13

Finland30,0000.4

Australia

76,103

1* The screening value of 100Bq/L is used as an indicator to determine the need for additional investigation in EU.
(Atomic Energy Society of Japan material edited by NDF)

255

b.

Total volume management
Prior to the accident at 1F, “the Discharge Control Target” for tritium for discharge into the ocean

was 22 trillion Bq per year. “The Regulatory Guide for the Annual Dose Target for the Public in theVicinity of Light Water Nuclear Power Reactor Facilities” (Decision of the Atomic Energy
Commission of 1975) defines the target value for the discharge control within the range that enables
it to meet the Target Radiation Dose of 0.05 mSv/year, which is a goal to maintain the low level ofthe radiation dose to the surrounding public due to the release of radioactive materials to theenvironment during normal operation. The guide also requires efforts to not exceed this Discharge
Control Target. The Target Radiation Dose does not substitute for the regulatory concentration limit,
and failure to achieve it should not be construed as a safety hazard. If 22 trillion Bq of tritium were
to be discharged into the ocean from 1F, the radiation effects would be conservatively estimated at
0.00001 mSv/year (sufficiently lower than 0.05 mSv/year)156. After the accident at 1F, no discharge
control targets or limits have been specified.
Though the discharge control target does not substitute for the regulatory limit, it is the target
value that should not be exceeded under the concept of As Low As Reasonably Achievable without
being satisfied that it is enough if it falls below the legal regulatory limit. While respecting thisphilosophy, at 1F, it is necessary to provide management using appropriate methods according to
the facility conditions.

3.4 System and operation for offshore discharge(1) Properties of ALPS-treated water
ALPS-treated water, etc. (ALPS-treated water, water under treatment) is the water after purifying
the contaminated water in buildings with purification systems such as ALPS. Though its properties
greatly differ from those of the contaminated water in buildings, it contains tritium that cannot be
removed physically or chemically by purification devices. ALPS is a device that performs purificationof radioactive materials other than tritium to bring them below the concentration limit required bylaw. About 70% of the ALPS-treated water currently stored in tanks contains water (water under
treatment) exceeding the legal limit (the sum of concentration ratios required by law is less than 1)
for discharging radioactive materials, except for tritium, into the environment.
This is due to the fact that in FY 2013 during the initial period of ALPS operation before
performance was improved, there were cases where the legal limit for discharge was exceeded.
Another reason was that the effective radiation dose (additional radiation exposure dose) at thesite boundary exceeded 1 mSv per year due to highly contaminated water, etc., generated andstored in tanks after the accident, and thus priority was given to reducing the additional radiation
exposure dose by reducing the frequency of ALPS sorbent replacement and increasing the amountof treatment, instead of satisfying the effluent standard. Specifically, from 2013 to the end of 2015,priority was given to reducing the additional exposure dose to less than 1 mSv/year as soon as
possible, and the frequency of ALPS sorbent replacement was reduced to increase the amount of256

treatment. In addition, although the additional radiation exposure dose of less than 1 mSv/year has
been achieved since FY 2017, priority was given to treating the strontium-treated water stored inflanged tanks with a high risk of leakage, by the end of 2018, and the amount of treated water wasincreased by reducing the frequency of sorbent replacement. As a result, the water under treatment
stored in tanks exceeds the limit for discharging into the environment.
Therefore, the concentration of the ALPS-treated water stored in tanks varies depending on theoperation of ALPS (frequency of sorbent replacement, etc.) and the time of treatment. Fig. A12-6
shows the storage volume of treated water by ratios of concentrations required by law. Approx.
30% of the entire storage volume has the ratios of concentrations required by law less than 1, and
1 or higher for approx. 70%. For water with a ratio of 1 or higher, it is planned to undergo secondary
treatment to reduce it to less than 1. Detailed data on such ALPS-treated water are available on
the TEPCO website and are updated regularly.

(Source : TEPCO)

Fig. A12-6 Storage volume of treated water by ratios of concentrations required by law(as of March 31, 2021)187
In addition, Fig. A12-7 shows changes in tritium concentration in stagnant water in buildings(results of concentrated water analysis of desalination unit RO). The tritium concentration, which
was high immediately after the accident, decreased monotonously until the middle of 2016 due tothe injection of reactor cooling water and the inflow of rainwater and groundwater. Later, as thewater level in the buildings decreased, the concentration increased due to the pumping up of
stagnant water containing highly-concentrated tritium that existed at the bottom of the buildings.
However, the tritium concentration peaked in March 2018 and has been declining again. As
described above, the ALPS-treated water contains tritium of different concentrations depending on
the time of treatment.
187 TEPCO, Treated Water Portal Site, website, https://www.tepco.co.jp/decommission/progress/watertreatment/

257

Tritium concentration in stagnant water in buildings (Bq/L)

1E+7

1E+6

1E+5

1E+4

(Source : TEPCO)

Fig. A12-7 Changes in tritium concentration in stagnant water in buildings(results of concentrated water analysis of desalination unit RO)188(2) Outline of the discharge system for offshore discharge
Fig. A12- 8 and (Source : TEPCO)
Fig. A12- 9 show conceptual drawings of the discharge system used for offshore discharge andthe overall system image, announced by TEPCO, respectively189,190. Of the ALPS-treated water
stored in reservoir tanks on site, the water under treatment that does not satisfy the legal limit fordischarge (the sum of concentration ratios of nuclides required by law is less than 1) undergo
secondary treatment and purification as many times as necessary until it becomes ALPS-treatedwater that satisfies the legal limit for discharge. Such ALPS-treated water is transferred to themeasurement/verification system (approx. 10,000 m3/tank group x 3) and stirred there, theconcentrations of radioactive materials, specifically, tritium, the 62 nuclides (nuclides removed by
ALPS) and carbon 14 are measured in order to verify whether the concentrations of these materials,
except for tritium, satisfy the legal limit. Then, the ALPS-treated water is transferred to the header
pipe, and mixed with seawater taken separately from the outside of the port so that the tritium
concentration after dilution becomes less than 1,500 Bq/L. The mixed and diluted water isdischarged into the ocean through a drainage pit equipped with a partition through an underwatertunnel about 1 km long. Emergency shut-off valves are installed to suspend discharge in the event

188TEPCO, Results of daily analysis of radioactive materials at the Fukushima Daiichi NPS, https://www.tepco.co.jp/

decommission/data/daily analysis/

189TEPCO, Attachment 2: TEPCO Holdings’ Action in Response to the Government’s Policy on the Handling of

ALPS-Treated Water [Digest version], press release, April 16, 2021
190TEPCO, Status of Review Regarding the Handling of ALPS-Treated Water at the Fukushima Daiichi Nuclear

Power Station, press release, August 25, 2021

258

of any system failure of the seawater transfer pumps, etc. or any abnormal value is confirmed bymarine monitoring.
[Conceptual diagram of the offshore releasing system]
Secondary treatment

Site use planning

Conduct secondary treatment, if necessary, toensure that it is below the regulatory limits forreleases to the environment*.

To ensure the stable release of ALPS-treated water andproceed with the construction of facilities necessary fordecommissioning, consideration will be given tomeasurement/verification system, replacement tanks anddismantling of storage tanks.

*Total notified concentration ratio is less than 1

Sr-treated water, etc.
(Water before ALPStreatment)

Multi-nuclide
removalequipment[ALPS]

Total notified concentration ratio
except for tritium [1 or greater]

Storage tanks•on premises

Diluting

Treated water analysis
The results of measurement and evaluation of theconcentrations of tritium, 62 radionuclides (to be removed by
ALPS), and carbon-14 radioactive in the ALPS-treated water
will be disclosed as needed, and assessments by third parties
and public disclosure will also be conducted.

Secondarytreatment
equipment

Waste

Since it will be diluted with a large
amount of seawater (more than 100
times), the total reported
concentration ratios of nuclides otherthan tritium in the discharged water
after dilution will be less than 0.01.

Total notified concentration ratio
except for tritium [less than 1]

Measurement/
verification system

Tritium concentration in released water

Measures in case of emergency
The release will be cancelled, if the facility is
unable to perform its planned functions due tomalfunction or power failure, or if abnormal
values are detected by marine monitoring.

The tritium concentration in the released water
should be less than 1,500Bq/L, which is 1/40th of theregulatory limit (60,000Bq/L), and is evaluated by thetritium concentration before discharged and theamount of dilution water.

Shut-off
valve

Released volume
For the time being, the volume will be within
the range of 22T Bq/year, which was the preaccident emission control target of Fukushima
Daiichi, and will be reviewed as appropriate
according to the progress of decommissioning.

Mix with seawater anddilute thoroughly

Seawater transfer Pump

Release into the sea

Drainage pit(Monitoring before release)

Underwater tunnel(approx.1km)(TEPCO material edited by NDF)

Fig. A12- 8 Conceptual diagram of the discharge system for offshore discharge 190(Source : TEPCO)

Fig. A12- 9 Overall image of the discharge system for offshore discharge 190

259

This system used for offshore discharge consists of the process from [purification] →[analysis/verification] → [dilution] → [discharge]. Technically, the utilization of technology andsystems with past results enables the establishment a highly reliable system. After the discharge,
marine [monitoring] is planned to oversee whether the ALPS-treated water is discharged into theocean in a safe way.
These systems will be approved by the NRA. The following shows the details and essential
technical points of the secondary treatment, analysis, dilution and discharge, and emergency
measures of this discharge system for offshore discharge.
a. Secondary treatment of ALPS-treated water
As described in the section on the properties of ALPS-treated water, since the water under
treatment contains 62 nuclides and carbon 14, whose sum of ratios of concentrations required bylaw is 1 or higher, except for tritium, purification by secondary treatment is required. As shown in
Fig. A-12-6Fig. A12-6, in particular, there are some tanks whose sum of ratios of concentrations
required by law exceeds 10,000 at the highest. For these tanks, it is planned to repeat secondary
treatment by purification by ALPS or reverse osmosis membrane device until the ratio becomesless than 1. In order to confirm the feasibility of the secondary treatment, purification testing(performance confirmation test for secondary treatment) is underway, using the water stored in thetank groups whose sum of ratios of concentrations required by law is high191. As part of this test, of
the water under treatment, performance confirmation testing for the secondary treatment using the
expanded ALPS has been carried out since September 2020, for the tank group (Group J1-C) of
comparatively high concentration and the tank group (Group J1-G) of comparatively low
concentration. The analysis of 62 nuclides, carbon 14, and tritium has also been performed for the

191 The 85th Secretariat Meeting of the Team for Countermeasures for Decommissioning and Contaminated Water

Treatment, Material 3-1, Status of performance verification testing on secondary treatment of treated water bymulti-nuclide removal equipment, December 24, 2020260

waterafter

secondarytreatment.

 J1-C group (62 nuclides + C-14 + H-3)
Before secondary treatment(System inlet)*1
Concentration limits
required by law[Bq/L]
Cs-134Seven
major
nuclidesAnalysis results[Bq/L]

After secondary treatment(Sample tank)*2

Concentration
ratios requiredby law*3

Analysis results[Bq/L]

Concentration
ratios requiredby law*3

0.49<7.60E-02

0.001360

2.93E+01

Cs-13790

5.99E+02

6.7

1.85E-01

0.0021

Co-60

200

3.63E+01

0.18

3.33E-01

0.0017

Ru-106100<5.00E+00

0.050

1.43E+00

0.014

Sb-125800

8.30E+01

0.10

2.26E-01

0.00028

Sr-90

30

6.46E+04

2,155

3.57E-02

0.0012I-129

9

2.99E+01

3.3

1.16E+00

0.13

1.76E+01

0.0088

C-14

2,000

1.53E+01

0.0076H-3

60,0008.51E+05

14.2

Sum of the concentration
ratios of the seven major
nuclides required by law

2,165

13.7 Of the 0.35, the sum of theconcentration ratios of thenuclides (51 nuclides)
After secondary treatment2
required by law, which(Sample tank)*
analysis and evaluation
confirmed was below the0.15 lower detection limit, is 0.19.

Sum of the concentration
ratios of the 62 nuclides*4 +C-14 required by law

2,406

0.35

Before secondary treatment(System inlet)*1

8.22E+05*1: Composite analysis was performed on the samples collected on September 19, 20 and 21.
*2: Analysis was performed on the samples collected on September 27.
*3: The value of the lower detection limit was used for calculating nuclides whose analysis results are below thelower limit of detection.
*4: See Reference for details of analysis results and the concentration limits required by law.

Numerically expressed, ○.○○E ±XX
means ○.○○ × 10±XX.

(Source : TEPCO)
Table A12-5 and A12-5 show the results of the performance confirmation testing on secondary
treatment. It is confirmed that the sum of ratios of concentrations required by law decreases from2406 to 0.35 in the J1-C group tank water and from 387 to 0.22 in the J1-G group tank water by asingle secondary treatment, and that the sum of ratios of concentrations required by law can be
reduced to less than 1.
In addition, in order to verify the reliability of the analysis, a third-party organization has
conducted analysis of the water generated by secondary treatment of the J1-C group tank water.
As a result, the sum of ratios of concentrations required by law was 0.28 compared to 0.35 reported
by TEPCO, and the outcomes were almost equivalent192. In this way, third-party analysis also
confirmed that the sum of ratios of concentrations required by law became well below 1 after thesecondary treatment.
As described above, the removal technology has already been verified for water under treatment
whose sum of concentration ratios required by law is 100 or higher. It is concluded that it is possible
to implement secondary treatment to sufficiently lower the concentration ratios to below theregulatory limit on safety, except for tritium.

192 The91st Secretariat Meeting of the Team for Countermeasures for Decommissioning and Contaminated
Water/Treated Water Treatment, Material 3-1, Results of performance verification testing on secondary
treatment of treated water by multi-nuclide removal equipment (third-party organization), June 24, 2021

261

Table A12-4 Secondary treatment test results of J1-C tank group191
 J1-C group (62 nuclides + C-14 + H-3)
Before secondary treatment(System inlet)*1
Concentration limits
required by law[Bq/L]
Cs-134Seven
major
nuclidesAnalysis results[Bq/L]

After secondary treatment(Sample tank)*2

Concentration
ratios requiredby law*3

Analysis results[Bq/L]

Concentration
ratios requiredby law*3

0.49<7.60E-02

0.001360

2.93E+01

Cs-13790

5.99E+02

6.7

1.85E-01

0.0021

Co-60

200

3.63E+01

0.18

3.33E-01

0.0017

Ru-106100<5.00E+00

0.050

1.43E+00

0.014

Sb-125800

8.30E+01

0.10

2.26E-01

0.00028

Sr-90

30

6.46E+04

2,155

3.57E-02

0.0012I-129

9

2.99E+01

3.3

1.16E+00

0.13

1.76E+01

0.0088

C-14

2,000

1.53E+01

0.0076H-3

60,0008.51E+05

14.2

Sum of the concentration
ratios of the seven major
nuclides required by law

2,165

13.7 Of the 0.35, the sum of theconcentration ratios of thenuclides (51 nuclides)
After secondary treatment2
required by law, which(Sample tank)*
analysis and evaluation
confirmed was below the0.15 lower detection limit, is 0.19.

Sum of the concentration
ratios of the 62 nuclides*4 +C-14 required by law

2,406

0.35

Before secondary treatment(System inlet)*1

8.22E+05*1: Composite analysis was performed on the samples collected on September 19, 20 and 21.
*2: Analysis was performed on the samples collected on September 27.
*3: The value of the lower detection limit was used for calculating nuclides whose analysis results are below thelower limit of detection.
*4: See Reference for details of analysis results and the concentration limits required by law.

Numerically expressed, ○.○○E ±XX
means ○.○○ × 10±XX.

(Source : TEPCO)

Table A12-5 Secondary treatment test results of J1-G tank group191
 J1-G group (62 nuclides + C-14 + H-3)
Before secondary treatment(System inlet)*1
Concentration limits
required by law[Bq/L]

Seven
major
nuclidesAnalysis results[Bq/L]

After secondary treatment(Sample tank)*2

Concentration
ratios requiredby law*3

Analysis results[Bq/L]

Concentration
ratios requiredby law*3

Cs-13460

5.94E+00

0.099<6.65E-02

0.0011

Cs-13790

1.18E+02

1.3

3.29E-01

0.0037Co-60

200

1.31E+01

0.065

2.33E-01

0.0012

Ru-106100<2.27E+00

0.023

4.83E-01

0.0048

Sb-125800

3.23E+01

0.040<1.37E-01

0.00017

Sr-90

30

1.04E+04347

3.18E-02

0.0011I-129

92.79E+00

0.31

3.28E-01

0.036

C-14

2,000

1.26E+01

0.0063

1.56E+01

0.0078H-3

60,0002.73E+05

4.6

Sum of the concentration
ratios of the seven major
nuclides required by law349

Of the 0.22, the sum of theconcentration ratios of thenuclides (53 nuclides)
After secondary treatment
required by law, which(Sample tank)*2
analysis and evaluation
confirmed was below the0.048 lower detection limit, is 0.17.

Sum of the concentration
ratios of the 62 nuclides*4 +C-14 required by law

387

0.22

Before secondary treatment(System inlet)*1

2.72E+05*1: Composite analysis was performed on the samples collected on October 5, 6 and 7.
*2: Analysis was performed on the samples collected on October 13.
*3: The value of the lower detection limit was used for calculating nuclides whose analysis results are below thelower limit of detection.
*4: See Reference for details of analysis results and the concentration limits required by law.

4.5

Numerically expressed, ○.○○E ±XX
means ○.○○ × 10±XX.

(Source : TEPCO)

262

b. Analysis
ALPS-treated water that satisfies the regulatory limit, except for tritium, is transferred to themeasurement/verification system, and 64 nuclides (tritium, 62 nuclides (nuclides removed by
ALPS) and carbon 14) are always analyzed before dilution and discharge.
Note) Treatment of the lower detection limit in analysis (ND value)
In nuclide analysis, there is a case where no significant concentration is detected, andthe limiting value is referred to as the lower detection limit (ND (Not Detected) value)
The lower detection limit varies depending on the detection accuracy of the instrument,
the measurement time, the amount of sample, the natural background, etc. When distillationor chemical separation is involved, an advanced analytical technique is also required.
Therefore, the lower detection limit is not a constant value, and fluctuates depending on theinstitution and facility for analysis. When the measurement result is below the lower
detection limit, TEPCO does not evaluate that the nuclide of concern is not contained (doesnot evaluate that the concentration is zero), but conservatively evaluates that theconcentration of the lower detection limit exists.
For a nuclide whose concentration is considered to be low, the lower detection limit
should be lowered in advance. In particular, when analyzing a large number of nuclides andevaluating the sum of ratios of concentrations required by law from nuclide concentrations,
it is important to select appropriate analytical conditions such as detection limit setting sothat sufficient accuracy can be obtained for each nuclide analysis.

Table A12-6 shows nuclide measurement and evaluation methods employed by TEPCO. Table
A12-7 gives details of the analytical methods used in the above secondary treatment tests.
Of the total 64 nuclides, 45 nuclides are analyzed/evaluated by γ-ray nuclide analysis, and 10
nuclides are by total-α activity. The other 9 nuclides are analyzed by β-ray measurement withpretreatment and inductively coupled plasma mass spectrometry (ICP-MS). Analysis takes time,
especially if distillation or chemical separation is required (For example, in the case of tritium, it
takes about half a day to a day.)。
Fig. A12-10 shows a standard analytical flow for tritium concentration193, but water samples at
1F are generally contaminated with salt, and must first be desalted by distillation. Then,
pretreatment using a concentration device is required according to the target lower detection limit.
Even after the completion of the pretreatment, mixing with the emulsion scintillator andhomogenization also take time. Similarly, analysis of carbon 14, nickel 63, cadmium 113m, etc.,
require a large amount of time for pretreatment.
In the performance confirmation test for secondary treatment, it took a long time, 2 months, toanalyze nickel 63 and cadmium 113 m in particular. In order to secure this period and achieve an
193 Tritium Analysis Method revised in 2002, Office of Emergency Planning and Environmental Radioactivity,

Nuclear Safety Division, Science and Technology Policy Bureau, MEXT

263

efficient discharge in future operations, it is planned that three tank groups of about 10,000 m3 will
be prepared. As shown in Fig. A12-11, with this method the tank groups have 3 roles, receiving,measurement/evaluation and discharge are used every two months by rotation. For analysis,
ALPS-treated water is circulated and stirred in the tank group to make it uniform, and then thesolution is sampled.
At present, it takes about two months for the analysis, which is one of the conditions limiting theamount of water that can be discharged into the ocean. For this reason, in order to achieve morestable discharge into the ocean, it is necessary to rationalize the pretreatment and analysis
procedures, and to further reduce the time required for the analysis. However, when demonstratingthat the sum of concentration ratios required by law for 63 nuclides, except for tritium, is less than
1, it is necessary to conservatively evaluate nuclides, whose analysis results are less than the lower
detection limit, using the lower detection limit. Therefore, it is important to select appropriate
analysis conditions in consideration of factors affecting the lower detection limit Note). The results
obtained from these analyses should also be disclosed each time before dilution and discharge,
and measurement/evaluation by third-party analytical institutions be considered. For third-party
involvement, it is important to develop a quality assurance system because the results may varydepending on the institution and facility performing analysis due to fluctuation of the lower detection
limit caused by the detection accuracy of measuring instruments or analysis techniques. For
example, it is effective to analyze common samples in an integrated way before discharge to see
the fluctuation range of the results, and verify and disclose the quality of each analysis institution
beforehand. It is also necessary to consider a structure for verifying the operation status of thedeveloped system.
Note) Treatment of the lower detection limit in analysis (ND value)
In nuclide analysis, there is a case where no significant concentration is detected, andthe limiting value is referred to as the lower detection limit (ND (Not Detected) value)
The lower detection limit varies depending on the detection accuracy of the instrument,
the measurement time, the amount of sample, the natural background, etc. When distillationor chemical separation is involved, an advanced analytical technique is also required.
Therefore, the lower detection limit is not a constant value, and fluctuates depending on theinstitution and facility for analysis. When the measurement result is below the lower
detection limit, TEPCO does not evaluate that the nuclide of concern is not contained (doesnot evaluate that the concentration is zero), but conservatively evaluates that theconcentration of the lower detection limit exists.
For a nuclide whose concentration is considered to be low, the lower detection limit
should be lowered in advance. In particular, when analyzing a large number of nuclides andevaluating the sum of ratios of concentrations required by law from nuclide concentrations,
it is important to select appropriate analytical conditions such as detection limit setting sothat sufficient accuracy can be obtained for each nuclide analysis.

264

Table A12-6 Measurement and evaluation methods for the concentration of 62 nuclides, tritiumand carbon 14 removed by ALPS191● Nuclides to be quantified and evaluated based on the results of● Nuclides to be quantified and evaluated based on the results of γ-ray nuclide analysis using a Ge semiconductor detector
total-α activity measurement
NuclidesNuclide measurement or evaluation method
NuclidesNuclide measurement or evaluation method
NuclidesNuclide measurement or evaluation method
1 Rb-86
γ-ray nuclide analysis
24 Cs-137
γ-ray nuclide analysis
46 Pu-238
Total-α activity
2 Y-91γ-ray nuclide analysis
25 Ba-137m
Cs-137 and radioactive equilibrium
47 Pu-239
Total-α activity
3 Nb-95
γ-ray nuclide analysis
26 Ba-140
γ-ray nuclide analysis
48 Pu-240
Total-α activity
4 Ru-103
γ-ray nuclide analysis
27 Ce-141
γ-ray nuclide analysis
49 Pu-241
Evaluated value from Pu-238
5 Ru-106γ-ray nuclide analysis
28 Ce-144
γ-ray nuclide analysis
50 Am-241
Total-α activity
6 Rh-103m
Ru-103 and radioactive equilibrium
29 Pr-144
Ce-144 and radioactive equilibrium
51 Am-242m Evaluated value from Am-241
7 Rh-106
Ru-106 and radioactive equilibrium
30 Pr-144m
Ce-144 and radioactive equilibrium
52 Am-243
Total-α activity
8 Ag-110m
γ-ray nuclide analysis
31 Pm-146
γ-ray nuclide analysis
53 Cm-242
Total-α activity
9 Cd-115m γ-ray nuclide analysis
32 Pm-147
Evaluation from Eu-154
54 Cm-243
Total-α activity
10 Sn-119m
Evaluation from Sn-123
33 Pm-148
γ-ray nuclide analysis
55 Cm-244
Total-α activity
11 Sn-123
γ-ray nuclide analysis
34 Pm-148m
γ-ray nuclide analysis● Nuclides to be quantified and evaluated by other methods
12 Sn-126
γ-ray nuclide analysis
35 Sm-151
Evaluation from Eu-154
NuclidesNuclide measurement or evaluation method
13 Sb-124
γ-ray nuclide analysis
36 Eu-152
γ-ray nuclide analysis
56 H-3β-ray measurement after separation by distillation
14 Sb-125γ-ray nuclide analysis
37 Eu-154
γ-ray nuclide analysis
57 C-14β-ray measurement after chemical separation
15 Te-123m
γ-ray nuclide analysis
38 Eu-155
γ-ray nuclide analysis
58 Sr-90β-ray measurement after chemical separation
16 Te-125m
Sb-125 and radioactive equilibrium
39 Gd-153
γ-ray nuclide analysis
59 Sr-89
β-ray measurement after chemical separation
17 Te-127
γ-ray nuclide analysis
40 Tb-160
γ-ray nuclide analysis
60 Y-90
Sr-90 and radioactive equilibrium
18 Te-127m
Evaluation from Te-127
41 Mn-54
γ-ray nuclide analysis
61 Tc-99
ICP-MS measurement
19 Te-129
γ-ray nuclide analysis
42 Fe-59
γ-ray nuclide analysis
62 Cd-113m β-ray measurement after chemical separation
20 Te-129m
γ-ray nuclide analysis
43 Co-58
γ-ray nuclide analysis
63 I-129
ICP-MS measurement
21 Cs-134
γ-ray nuclide analysis
44 Co-60
γ-ray nuclide analysis
64 Ni-63β-ray measurement after chemical separation
22 Cs-135
Evaluation from Cs-137
45 Zn-65
γ-ray nuclide analysis
23 Cs-136
γ-ray nuclide analysis(Source : TEPCO)

265

Table A12-7 Analytical method used for secondary treatment testing191
NuclidesTarget lower detection
limit (Bq/L)

Analysis method

γ-ray emitting nuclide

Collect a sample in a 5-liter marinelli beaker, and measure with a Ge
semiconductor detector.

0.07 (Cs-137)*1

Sr-90, Sr-89

After purifying Sr with Sr resin, measure the substance precipitated and collected
as carbonate with a beta spectrum analyzer.

0.04 (Sr-90)*2

I-129

Measure using ICP-MS after adding hypochlorous acid to a sample and adjusting
it to iodate ion.

0.2

H-3

Measure with a liquid scintillation counter after mixing a sample, from which
impurities have been removed by distillation, with a scintillator.

30

C-14

Add concentrated nitric acid and potassium persulfate to a sample, heat it, collectthe generated CO2 with adsorbent, mix it with a scintillator, and then measurewith a liquid scintillation counter.

10

Tc-99

Dilute a sample with nitric acid, and then measure it using ICP-MS.

2

Total-α activity

Coprecipitate α-nuclides with iron hydroxide,remove the iron by extraction,
evaporation to dryness and baking on a stainless steel dish, and then measurewith a ZnS scintillation counter.

0.04

Cd-113m

Purify and collect Cd by ion exchange, mix with a scintillator, and measure with aliquid scintillation counter.

0.2

Ni-63

Purify and collect Ni by Ni resin, mix with a scintillator, and measure with a liquid
scintillation counter.

20*1: The target lower detection limit of other nuclides varies depending on the baseline, interfering nuclides, background, and γ-ray
emission rates.
*2: The target lower detection limit of Sr-89 varies depending on the Sr-90 concentration.

(Source : TEPCO)

266

Distillation method
Sample water, 100 ml

Alkaline solution electrolytic
enrichment method
Sample water, 600 ml

SPE electrolytic
enrichment method
Sample water, 1000 ml

Potassium permanganate
Sodium peroxide
Distillation (normal pressure, reduced pressure)

Distilled
Electrolysis cell sample
About two weeks
500 ml(1300 AH)
Na2O2
2.53 gSPE automaticenrichment device
About three days
Automatic shutdown
by water levelsensorDistilled
sample1000 ml

Final liquid volume, 55 ml
Lead chloride
Distillation

Distilled sample, 40 - 50 ml
Emulsification scintillator, 50 - 60 ml
Measurement container (Teflon vials, etc.)

Mixing (Gelation requires warming)

Homogenization
Leave in a measuring container
for one day to one week
Liquid scintillation measurement(Source : MEXT)

Fig. A12-10 General analytical flow for tritium concentrations in water samples192(Source : TEPCO)

Fig. A12-11 Tank management during analysis194

194 The 91st meeting, Study group on monitoring and assessment of specified nuclear facilities, Material 2, Statusof examining the discharge system for offshore discharge concerning treated water by multi-nuclide removal
equipment, June 7, 2021

267

c. Dilution and discharge
After analysis and concentration verification, the ALPS-treated water is diluted with seawater andthen discharged. Fig. A12-12 shows the method of diluting ALPS-treated water with seawater. For
discharge, ALPS-treated water with an upper limit of 500 m3/day is mixed and diluted in the
discharge pipe (header pipe) with seawater transferred from the seawater transfer pump (up to
three pumps at 170,000 m3/day per pump) so that the tritium concentration after dilution is lessthan 1,500 Bq/L. Then, the diluted water is discharged into the ocean through a drainage pit andan underwater tunnel about 1 km long. Therefore, the dilution ratio is about 340 times even in thecase of the minimum ratio (assuming an injection volume of ALPS-treated water of 500 m3/day andoperation of one seawater transfer pump). After this dilution process, the sum of concentration
ratios required by law, except for tritium, becomes less than 0.003, resulting in a smaller effect on
the human population and the natural environment.

500 m 3/day at maximum
Sample water to beanalyzed after
homogenization by
circulation/stirring,and final analysis tobe performed before
release.
Duration: Approx. 2 months

ALPS-treated watertransfer pump

Evaluate the dilution ratio fromthe flow ratio.

170,000 m3/day
FT
340,000 m3/day
510,000 m3/day Seawater

Measurement/
verification
system
(K4 tank group)

FT
ALPS-treatedwater
flowmeter

flowmeter

ALPS-treated water
discharge control valve

Discharge pipe

Seawater transfer
pump

Drainage pit

Undersea tunnel(TEPCO material edited by NDF)
Fig. A12-12 Diluting method of ALPS-treated water with seawater190
As described above, since it takes from half a day to a day to measure the tritium concentration,
it is difficult to measure and verify the results in real time at discharge. Therefore, the tritium
concentration after seawater dilution will be checked as follows.
(1) Evaluation by the tritium concentration and flow rate of ALPS-treated water
TEPCO intends to guarantee that the tritium concentration after seawater dilution is below 1,500
Bq/L based on the tritium concentration obtained by analysis in the measurement/verification
system, and the flow ratio between ALPS-treated water and seawater. The specific concentration
can be calculated using the following equation:
Tritium concentration
after dilution with seawater =Tritium concentration of ALPS-treated waterx ALPS-treated water flow rate (control by discharge control valves)
ALPS-treated water flow rate (control by discharge control valves)+ seawater flow rate

In addition to the above, in order to confirm that the tritium concentration at the end of a discharge
outlet is below 1,500 Bq/L after mixing and diluting as designed, the following two measurement
methods (2) and (3) are planned.
(2) Daily measurement of tritium concentration in a drainage pit

268As part of the marine monitoring in the port (see Table A12-8 below), the water in a drainage pit
is sampled daily to check the tritium concentration. Although the analysis results will be disclosed
promptly, a time lag of about one day is expected between sampling and announcement of theresults.
(3) Immediate pre- discharge measurement of tritium concentration after dilution using a drainagepit
The drainage procedure is shown in Fig. A12-13. Once the pit (approx. 2,000 m3) is drained,
dilution is performed by running a small amount of ALPS-treated water (20 m3 or less) while
operating one seawater transfer pump for about 10 minutes. Then, the pump is shut down before
the pit becomes full, and the water in the pit is sampled in order to confirm that the calculated tritium
concentration and the measured tritium concentration are equivalent or less than 1,500 Bq/L (about
2 days). After verification, the water in the pit and the remaining water (approx. 10,000 m3/tankgroup) measured in the measurement/ verification system are discharged constantly orintermittently.

B. Dilute by running a small amount of ALPS
-treated water, while accumulating seawat
er in the pit

A. Drain the pit once

ALPS-treated waterTransfer tothe downstream side

SeawaterRise in
water
level

To the tunnel
Drainage pit

Drainage pit

C. Shut down the pump before
the pit becomes full, and sample andmeasure the water in the pit

D. After checking the tritium concentr
ation, pump seawater again and release the water

Water sampling/measurement

Release the remaining
measured water (approx.
10,000 m3/tank group)
continuously or intermittently
SeawaterTo the tunnel
Drainage pit

Drainage pit(Source : TEPCO)

Fig. A12-13 Method for verifying tritium concentration using a drainage pit after dilution withseawater190
Discharge will be carried out in the manner described above. The start of discharge will becarefully initiated from a small amount. After that, the amount of tritium will be evaluated for eachwater sample (approx. 10,000 m3/tank group) measured in the measurement/verification system
to control cumulative values and confirm that the tritium level is below 22 trillion Bq per year.

269

d. Emergency measures
If there is a possibility that the dilution rate of the ALPS-treated water will become abnormal due
to system failure (seawater pump shutdown, decrease of seawater flow rate, increase of treatedwater flow rate, failure of flowmeter), or if the properties of the ALPS-treated water are abnormal(radiation monitor operation or failure), the plan is to immediately close the two emergency shut-off
valves, shut down the ALPS-treated water transfer pump, and suspend discharge. Fig. A12-14
shows an emergency shut-off valve layout. It is planned that one emergency shut-off valve will be
installed in the vicinity of the seawater transfer piping to minimize the amount of ALPS-treated water
discharge in the event of an abnormality, and the other emergency shut-off valve will be installedinside the tide embankment in anticipation of submergence due to a tsunami.
If any abnormal value is confirmed by marine monitoring, discharge will also be suspended. As
described later, however, the marine monitoring results will be verified after discharge.
Measurement/
verification system(K4 tank)

Radiation monitor

In the eventa radiation monitor detects
any abnormality or fails

Emergency shut-off valve layout

RT
EL 33.5 m
ALPS-treatedwater transfer pump

Safety considerations
should be taken to
ensure that emergency
shut-off valves can beclosed even in the eventof a loss of power

ALPS-treatedwater flowmeter

FT

In the event
a flowmeter shows
any abnormal indication
or fails

Emergency shut-off valve (1)EL 11.5 m

Emergency shut-off valve (1)Placed inside the seawall in light of tsunami measures
Emergency shut-off valve (2)Placed at the point before the treated water is mixed
with diluted seawater in light of minimizing the amountof release

Diluted seawaterflowmeter
Emergency shut-off valve (2)Seawall

FT
Seawater transfer
pump trip

EL 2.5 m
Seawater transfer pump(Source : TEPCO)

Fig. A12-14 Emergency shut-off valve layout and emergency measures190(3) Marine monitoringa. Simulation of tritium diffusion in the ocean
Fig. A12-15 shows the simulation results of tritium diffusion in the ocean performed by TEPCO
using the model verified by the measured data of cesium 137 after the accident. The simulation
conditions are as follows:- Target sea area: About 500 km north-south and about 600 km offshore centered in Fukushima
Prefecture- Discharge method: Planned to discharge ALPS-treated water evenly within a 1 km x 1 km area ina horizontal direction adjacent to the 1F site.
- Resolution: 1 km mesh horizontally, 30 layers vertically against water depth (to a depth of 1 km)- Weather conditions, etc.: Wind velocity, atmospheric pressure, temperature, humidity, andprecipitation from January to December 2014 are used (including flow conditions (Black Current,
mesoscale vortex) off the coast of Fukushima Prefecture)

270

(Source : TEPCO)

Fig. A12-15 Range of tritium diffusion195
The areas where the tritium concentration exceeds the background level (0.1 - 1 Bq/L) arelimited within the area near the Fukushima Daiichi NPS for which the common fishery right is not
specified. The result is sufficiently small compared with the WHO drinking water standard (10,000
Bq/L).
Although there are limitations in the prediction accuracy of these results, general models andappropriate publicly available weather and ocean current data for input were used. The results areassessed that they are consistent with the trend of observation values and simulation results
already disclosed in published articles196.
However, considering that the simulation results of diffusion in the ocean include uncertainties
due to differences in models, external forces, and ocean fluctuations197, it is recommended thatsimulations that incorporate specific discharge conditions of ALPS-treated water currently planned
are conducted, and that the effects on the human population and the environment are evaluated.
b. Marine monitoring plan
According to the simulation results described above, although the range of 1 Bq/L198 or higher islimited, TEPCO has proposed measures to strengthen the marine monitoring in order to understandthe diffusion situation (Table A12-8). As mentioned in (2) c. Dilution and discharge, daily
measurements of tritium concentration and increasing sampling points in a drainage pit are planned.
195 TEPCO Draft Study Responding to the Subcommittee Report on Handling ALPS-Treated Water, press release,

March 24, 2020
196 e-GOV portal, Public Comment, Results of a written request for public comments on the handling of treatedwater by multi-nuclide removal equipment, page 8, April 13, 2021

197 Yukio Masumoto, et al.,: Near Coastal Ocean Variability Affecting Radionuclide Dispersion Simulation and Their

Reproducibility in Numerical Models, Bulletin on Coastal Oceanography, Vol. 54, No. 2, 151 - 157, 2017
tritium concentration of tap water in Fukushima Prefecture is about 1 Bq/L. The WHO drinking water
standard is 10,000 Bq/L.

198 The271

This marine monitoring will start about one year before the scheduled discharge (around fromthe spring of 2022), and TEPCO plans to ask agricultural, forestry and fishery operators, local
government officials, etc. to participate in monitoring and inspection when collecting samples andmeasuring radioactivity. In case any abnormal value is detected by marine monitoring after
discharging, discharge will be suspended and a condition survey will be conducted.
In addition, the task force on monitoring and measuring the marine environment has been held
under the government-sponsored Monitoring Coordination Council since June 2021. The Ministryof the Environment, NRA, The Agency for Natural Resources and Energy Agency, Fukushima
Prefecture, and TEPCO are participating in the study on strengthening and expanding themonitoring199. In the future, related organizations will collaborate to ensure the implementation of
the marine monitoring.
Table A12-8 Summary of marine monitoring plan190(Source : TEPCO)(4) Marine organism feeding trial
In order to foster understanding of the discharge of ALPS-treated water into the ocean and to
reduce reputational damage, a marine organism feeding trial is planned as shown in Fig. A12-16.
Before offshore discharge of ALPS-treated water, this trial intends to observe the growth of marineorganisms in both environments, in seawater and in ALPS-treated water diluted with seawater. Afterthe start of offshore discharge, their growth will be observed in the water diluted with seawater andactually discharged to the ocean. These trials will see any health abnormalities in fish, etc.,
compare the concentration of radioactivity including tritium in breeding water and in vivo, and
199 The 13th Monitoring Coordination Council, Reference 5, Establishment of the Task Force of Monitoring and

Measuring the Marine Environment, April 27, 2021
https://www.env.go.jp/water/post_110.html

272

observe hatching rate of eggs, survival rate of adult fish, etc., while streaming of breeding
conditions is planned.
Prior to discharge of ALPS
treated water into the sea

After initiating the discharge of
ALPS treated water into the sea
Confirm the status of its development
under an environment where water isdiluted with seawater and actuallydischarged into the environment.

Confirm the status of its development in seawater (test tank 1) and
ALPS treated water diluted using seawater (test tank 2)

CompareTest tank 1: Seawater around thepower station (Tritium
concentration approx. 1Bq/liter)

Test tank 2: ALPS treated water diluted using
seawater around the power station(Tritium
concentration approx. 1,500Bq/liter)

Test tank: water discharged into theenvironment(Tritium concentration < 1,500Bq/liter)

Subject to be reared [Both before and after initiating discharge of ALPS
treated water into the sea]
Fish: flounder (size of specimen should be around 30cm –40cm or below)
Shellfish: details currently under review
Seaweed: details currently under review(Source : TEPCO)

Fig. A12-16 Summary of marine organism feeding trial190(5) Technical study on tritium separation
TEPCO plans to publicly solicit technical proposals from third-party organizations to identify new
trends in tritium separation technologies. They will solicit technologies that are expected to reduce
tritium concentration to 1/1,000 or less and to expand operating capacity to 50 - 500 m3/day in the
future. Primary evaluation of the technologies sought from the public will be conducted by a thirdparty organization, followed by a secondary evaluation by TEPCO. As a result, if it is confirmed thata technology is realistically practicable for ALPS-treated water, etc., it is planned that a specificdesign will be examined and technology demonstration testing will be conducted to establish thetechnology.
(6) Simple estimation of the volume reduction of tritium due to the discharge of ALPS-treatedwater into the ocean
As mentioned above, the discharge system used for offshore discharge consists of the processfrom [purification] → [analysis/verification] → [dilution] → [discharge]. Technology and systems withpast results are used, and thus strict compliance with the regulatory standards and operation
procedures enables safe discharge of the ALPS-treated water.
Fig. A12-17 shows the results of a simple estimation of the change in the amount of tritium
stored in tanks if the ALPS-treated water is discharged into the ocean. In this estimation, it was
assumed that as of 2020, 780 trillion Bq existed in the tanks, and that 300,000 Bq/L (based on the
recent tritium concentration shown in Fig. A12-7) of stagnant water in buildings would be newly
generated every year by 51,000 m3 (140 m3/day x 365 days). Tritium is assumed to decay
spontaneously (half-life of 12.3 years), and the offshore discharge is assumed to start in 2022.

273

In this estimation, the amount of tritium stored in tanks became zero after 2045 by discharging
22 trillion Bq of tritium per year into the ocean.
A plan has been developed to maintain the total amount of tritium to be discharged per year
below the discharge control target (22 trillion Bq per year) before the accident at 1F, and the plan
is to be revised as necessary according to the progress of decommissioning.

Fig. A12-17 Simple estimation results of the amount of tritium stored in tanks274

4. Future efforts
TEPCO's planned discharging system, if operated reliably in accordance with the implementation
plan approved in the review by the NRA, will have no adverse effects on humans and theenvironment, including other radionuclides, and therefore it is an important issue to operate thesystem "reliably" “as planned”.
Going forward, TEPCO will need to proceed with the following preparations in order to realize
the discharge of the ALPS-treated water into the ocean in a safe way. During actual operation, it isnecessary to ensure the implementation of the plan (system, operation, information distribution,etc.) established in the preparation stage, to perform checks and reviews, and to review and expand
the plan as needed.
 In the operational phase, develop a series of operation plans including system operation,
analysis of ALPS-treated water, flow control of the treated/diluted water, marine monitoring,
maintenance, and troubleshooting, and then develop a system plan that minimizes risks
and eliminates social concerns
 Perform a radiological impact assessment on the human population and the naturalenvironment, and disclose evaluation results based on the specific discharge dischargeplan Verify safety by experts from the International Atomic Energy Agency (IAEA) and otheragencies Develop a plan to strengthen marine monitoring, and perform marine monitoring before
the discharge Education and training on system operation and analysis, etc., for parties concerned
including contractors (TEPCO)
 Development of strategies to provide accurate and understandable information
domestically and internationally without causing anxiety from a social perspective, andtimely dissemination of the status of preparations
 Ensuring implementation of measures to prevent reputational damage as set forth in the
Government’s basic policy announced in April 2021
NDF will provide technical and professional support for TEPCO in designing the discharge
system and considering discharging methods, while promoting distribution of accurate informationand increasing understanding through various opportunities in Japan and abroad in line with theinterests of those who will receive the information. NDF will also ensure that TEPCO implements
measures to minimize reputational damage, and that TEPCO takes action with adequate andsufficient compensation in the event of reputational damage.

275

Attachment 13

Basic direction of 6 Essential R&D Themes
December 12, 2017
Nuclear Damage Compensation and Decommissioning Facilitation Corporation (NDF)
Basic direction of 6 Essential R&D Themes

The Mid-and-Long-Term Roadmap towards the Decommissioning of TEPCO’s Fukushima Daiichi Nuclear
Power Station (September 26, 2017) specifies enhancement of the activities for matching the R&D required fordecommissioning (Needs) with the basic and fundamental R&D (Seeds) and for human resource development. It
also specifies enhancement of the functions of Japan Atomic Energy Agency’s Collaborative Laboratories for
Advanced Decommissioning Science (JAEA/CLADS) and promotion of joint researches with domestic andinternational universities and researching institutions to establish the international decommissioning research centerwith concentrated wisdom.
Following the above, the Ministry of Education, Culture, Sports, Science and Technology (MEXT) states in its
budget request for FY2018 that they will reform the Center of World Intelligence Project for Nuclear Science andTechnology and Human Resource Development (“World Intelligence Project”) into a subsidy program intended for
JAEA/CLADS and the program will be implemented under the system centered by JAEA/CLADS from the newly
adopted proposals from FY2018.
On the occasion of this reform, MEXT showed NDF their intention to discuss how to proceed with the future
R&D of the World Intelligence Project. This includes selection of the theme for the call for proposal considering the
Essential R&D selected by the task force from the viewpoint of promoting basic and fundamental researches withsatisfactory understanding of the Needs.
Therefore, concerning the 6 Essential R&D Themes which are described in the interim report of the Task Forceon Research Collaboration (November 30, 2016), the Basic Direction of the 6 Essential R&D Themes was compiled
including the background of the issues, the problem consciousness, and the expected research image, with
consulting the discussions in the working group for each theme.
Theme“Descriptions /Background issues”
on the interim report

Basic direction(1) To identity process of characteristic change in fuel debris over time
The fuel debris retrieval is scheduled for 2021 onward, 10 years after the fuel debris
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. In addition, the retrieveddebris must be stored safely. Choosing the best possible methods ofretrieve/transmission/storage of fuel debris requires predictions of characteristic changes
of fuel debris over time.
In relation to the accident of Chernobyl nuclear power plant, detection of particulatesof the micron order which contain fuel components from around the fuel debris has beenreported and the national report of the Ukrainian government showed concerns about theincrease risk of radioactive dust emergence over time through self-decay. One of thepossible reasons is that the fuel debris with high radioactivity exposed to rich humidity
caused rapid progress of aging which quite slowly proceeds with normal uranium mineralin the geological environment. It results in oxidation activated by radioactive dissolution
and generate hexavalent uranium compound. On the other hand, because the PCV of
the Fukushima Daiichi NPS (hereinafter referred to as “1F”) is currently in the nitrogen
atmosphere under subtle positive pressure, oxidation is unlikely to proceed immediately.
In the future, such an event similar to the above may occur because the air that contains
oxygen may flow into the PCV when a negative pressure control is applied to retrieve thefuel debris. Since the radiation level is about one order higher in 1F than that of the fueldebris in the case of the Three Mile Island Unit 2 (TMI-2) accident (occurred in a shorttime after the operation started), it is under unexperienced condition. In addition, it should
be noted that it will take a longer period to complete the retrieval of the fuel debris fromthe time of accident than in the case of TMI-2.
Various factors are involved in the aging of such fuel debris in addition to the oxidation
described above. Roughly classifying, those factors may include the chemical mechanism(oxidation-reduction, leaching of included components, changes in the chemical form or
the phase state, etc.), physical mechanism (structural or characteristic changes by heat
cycle etc., irradiation damage by α-ray), and coupled actions of these factors.
Since decay or leaching of fuel debris due to aging lead to emission of FP particles
or gas, or effluent of particles that contain α-nuclides that are confined in the fuel debris,they have significant impact on the system design and procedures. It includes the retrieval
mechanism, the cooling and circulating system, the containment function, the criticality
monitoring system, the PCV gas control system, exposure evaluation, containing,transferring and storing, and processing and disposal. In particular, as for the Mid-andLong-term Roadmap, while the processing and disposal method of the fuel debris will bedecided in the third period (from 2022) after the retrieval of the fuel debris is started,276

obtaining the aging information of the fuel debris is an urgent issue. While taking into
consideration the permission and authorization about the safety regulation, to provide
sufficient prediction and explanation of the risk changes resulted from the aging of thefuel debris, it is required to clarify the real situation of what are expected to have critical
impacts on the decommissioning works preferentially.
Therefore, it is necessary to build a fundamental theory of the aging model byclarifying the aging process while using the current knowledge of the actinoid chemistry.
To do so, demonstration test should be performed using real uranium according to thematrix pattern of parameters (temperature, pH, etc.) to collect basic data, and it needs toestablish the prediction method of aging. In this case, it is important to maintain thefoundations for advancement of the actinoid chemistry, which provides the basis for
examination of the physical property of fuel debris. In addition, heat analysis for 1F should
be included in the basis of investigation since the temperature distribution of the fueldebris has to be understood by calculating the heat distribution and the impact of regionaltemperature rise due to the decay heat should also be required to be examined.
Theme(2) To elucidate corrosion mechanisms under unusual/extreme circumstances“Descriptions /Background issues”
on the interim report

It is required to collect data on corrosion under a variety of circumstances withconsideration of the circumstances specific to 1F decommissioning such as high radiationlevels and unsteady routes of cooling water in order to prepare for potential corrosion
during decommissioning.
A boiling water reactor (BWR) consists of various metallic material. While stainlesssteel, which is corrosion resistant, is used inside reactor where it is high temperature andhigh oxidizing environment, carbon steel, which is not corrosion resistant, is used for the
PCV that is the confinement boundary and is assumed to be used in the normal
atmosphere. On the other hand, substantial knowledge has been collected so far about
corrosion of structure and piping for commercial electric generation reactors, andespecially, data has been collected being focused on the corrosion data in theenvironment of high radiation, high temperature, and deionized water for the operation of
BWR.
However, after the accident, 1F has been in a special environment with highradiation, room temperature, suspended solids and deposited materials. The knowledge
about such environment is not sufficient. Since water has been injected into PCV to cool
the fuel debris, carbon steel is dipped in the water. In addition, it is known that chemical
species of oxidation nature such as hydrogen peroxide and various radical species havebeen generated through radiolysis of water. Currently, since nitrogen has been injected
into PCV to prevent hydrogen explosion and the oxygen density has been decreased, thedensities of the oxygen and the hydrogen peroxide are considered to be decreased in thewater and corrosion of PCV is also considered to be suppressed in some degree. In the
future, since the air containing oxygen flows into PCV when a negative pressure control
is applied to retrieve the fuel debris, it is important to maintain soundness of the structure
and pipes that form the boundary for confinement of radioactive materials and it is also
important to prepare countermeasures based on the knowledge on corrosion in suchenvironment.
Since the corrosion is essentially a kind of battery reaction, it is likely to happen if theelectric conductivity of water raises, pH falls, and the electric potential raises under thecondition of declining of surrounding water quality. Although corrosion has been
suppressed by nitrogen injection in general, it is still under the condition of potential
corrosion. Regional changes of the environmental condition may lead to an increase of
corrosion speed at the part. It is quite a special environment surrounded by various factorsthat promotes corrosion such as a formation of liquid film of dew, a humid environmentthat repeats wet and dry conditions near the water surface, an irregular flow of coolingwater, convection flow, or backwater due to irregular paths created by gaps between
various shapes of fallen objects or deposited materials, a progress of corrosion on theanode side between different kinds of metals touching each other, a progress of acidbase reaction by microbes, or any other potential factors. In the future, further changes
may occur in the internal environment when the air that contains oxygen flows into the
PCV when a negative pressure control is applied to retrieve the fuel debris. Since thecorrosion progresses over time during the decommissioning works under the special
environmental conditions, estimation of corrosion phenomenon and investigation ofcountermeasures are required based on the consideration on the environmental changes
resulted from the progress of the decommissioning process.
Therefore, it is necessary to collect basic data related to the progress of corrosion
phenomena and systematically clarify and understand the phenomena in order to provide
satisfactory prediction and explanation of the risk changes that follow corrosion of

Basic direction

277

structures taking the permission and authorization of the safety regulation into accountwith giving priority to the factors of higher needs that may have critical impacts on thedecommissioning works from the viewpoint of the probability of occurrence such as thefactors above and the impacts on the functionality (parts and severity), the scale and thetiming. In this case, in order to examine various approaches including not only the use ofexisting anti-corrosive agents but also electric protection, it is important to accumulate
and maintain the knowledge related to the corrosion phenomena in addition to theelectronic state of materials in a special environment through principle analysis andclarification of the corrosion progression mechanism.
Theme(3) Radiation measurement technologies adopting innovative approaches“Descriptions /Background issues”
on the interim report

The radiation levels are still extremely high inside the 1F reactors/buildings due to theaccident and the existing measurement devices do not meet the capability/functional
requirements to provide accurate figures. It is vital to develop an innovational device
adopting brand-new ideas/principles based on 1F needs.
Currently, radiation measurement can be performed following a predefined operation
procedure without detail knowledge of measurement since a number of radiation
measurement products using various principles or materials including ionizationchambers, counting tubes, semiconductor detectors, and scintillation detectors are
already offered. However, it is very important to develop the human resources formeasurement since it is necessary to understand the principles of the equipment in order
to interpret the measurement data and address possible troubles such as the case of thedisorder (inversion of data value) between the all β-radioactivity value and the value of
Sr-90 because miss-counting has not been taken into account for the resolution time in
the analysis of the sampled water at the under-ground water observation hole on 1F.
In addition, general radiation measurement products are not able to offer satisfactory
performance and functionality to inspect the conditions inside reactors and buildings at
the decommissioning site of 1F. The decommissioning works on 1F must be performedby remote operation since the radiation level is extremely higher than the one in the work
environment of existing nuclear facilities. It is necessary to develop highly radiation
resistant and small sized measurement sensor, electronic circuits, and systems in order
to be remotely operated. In addition, it is necessary to research on the basic mechanism
related to radiation damages of materials in order to develop highly radiation resistantsensors and circuits. As for the specific examples of sensor development, it is necessary
to develop measurement devices of neutron from the viewpoint of criticality prevention,
real time measurement of α-ray from the viewpoint of identification of the fuel debris, andɤ-ray measurement with high energy resolution for nuclide estimation under the
background of high gamma radiation, those what satisfy various needs: radiation
resistance, noise resistance, size (small size), counting rate and responsiveness, highradiation resistance, energy discrimination, space resolution (identification of radiation
source position), ease of operation, and maintainability. As for the composition of themeasurement targets, development of the technology so-called “on-site analysis” isrequired since there are needs of functions that can be used to analyze the target withouttransferring a sample to other facility or equipment and obtain rough results used topromptly judge if the target is debris or not, and if the target is debris, the function to judge
co-existence of reactor internals and neutron poison is required.
In addition, effective support tools for the decommissioning can be provided bydeveloping the technology to visualize the radiation field and the contamination situationand clarify the profile of the fuel debris based on the information of the strength and the
direction of the radioactive sources.
It is necessary to develop the generic technologies for innovative measurement ofradiation using new ideas and principles by considering on-site measurement
requirements.

Basic direction

Theme“Descriptions /Background issues”
on the interim report(4) To clarify behavior of radioactive airborne particle generated duringdecommissioning (incl. α-dust treatment)
As thermal cutting of the fuel debris via machine or Laser may produce a large amount of
α-dust, it requires safety measures and dust confinement solutions. It is necessary tounderstand physical/chemical properties of α-dust, to predict the amount of dust to be
produced for each method, and to consider how to seal the dust according to the resultsin order to make sure the retrieval will be conducted in a safe and effective manner.

278

Basic direction

As the fuel debris retrieval work will start, cutting the fuel debris will create a lot ofradioactive airborne particles (α-dust) that contain α-nuclide and they will be dispersed
within the boundary. When retrieving the fuel debris, since the work will be performed in
the confinement boundaries which are the broken buildings, it is important to understandthe property of the α-dust for studying how to secure the confinement capability,
designing the filtering system, and performing the exposure evaluation of the surrounding
environment and workers including the time of accident.
With regard to the data about the scattering rate when α-dusts are generated, thereis the data obtained at the decommissioning of JAEA’s JPDR and the dismantling of glove
box of JAEA’s Nuclear Fuel Cycle Engineering Laboratories. However, the data has notbeen collected just for the nuclear fuels but collected for the objects polluted by the
nuclear fuels, and the data has been collected for the amount of the radioactive materials
and their densities from the viewpoint of radiation exposure control and it is not
systematically organized.
On the other hand, the radioactive airborne particles in 1F will be generated directlyfrom the fuel debris when retrieving the fuel debris and from the polluted objects in thedecommissioning process. The types of radioactive materials are α-nuclides and β (ɤ)nuclides. While the α-nuclide of which typical element is plutonium is important from theviewpoint of internal exposure, the β(ɤ)-nuclide such as cesium should be wellconsidered as well from the viewpoint of the total exposure evaluation.
In order to study on collection of radioactive airborne particles, efficient filtering andpurification, criticality prevention, etc., it is necessary to grasp the amount of generated
particles, distribution of particle diameters, radioactive diameters, and the physical andchemical property of particles according to the differences of the target objects and themethod of cutting. It is also important to understand the behavior of the generated
particles in the gas phase, at the air-liquid interface, and in the liquid phase duringtransportation or transition. For example, it is important to understand the growth ofparticles through coagulation in the gas phase, evaluation of mist generation from the airliquid interface, leaching behavior of the components into the water of the liquid phase,transportation behavior such as settling of particles in the water or filtering, etc.
With regard to the exposure evaluation of radioactive airborne particles, it isimportant to evaluate the impact of exposure to radioactive materials derived from thefuel debris, especially the one of α-nuclides. In this case, it is important to decide whether
the conventional exposure evaluation methods can be applied by judging if the chemical
form and the particle diameter of the radioactive airborne particles represented byplutonium is consistent with the ones that have been used as the criteria of internal
exposure evaluation for plutonium.

Theme(5) To understand fundamentally mechanisms of radioactive contamination“Descriptions /Background issues”
on the interim report

To figure out the mechanism of radioactive contamination towards effective
decontamination; it is critical to implement effective approaches of decontamination
based on the mechanism of the contamination to radiation sources, and to decrease thevolume of radioactive wastes as well.
With regard to reduction of the radiation in the buildings, the target object of
decontamination are pipes and ducts, metals in the equipment, resins in cables and soon, paints, and concretes of the walls or floors. The contamination source includes molten
high temperature fuels at the time of accident, steams that contain radioactive materials
such as Cs leaked in consequence of hydrogen explosion and so on, dusts andcontaminated water that contain radioactive materials. Currently, as for radiation
reduction in the buildings on 1F, decontamination of floors and walls has only limited
effects since there are other contamination sources such as objects remained in pipes,
and hidden side behind the pipes located at high inaccessible positions. In the future,when considering each step in the long decommissioning process, it is expected thatmany situations that require decontamination will arise, therefore, effective and efficientdecontamination is considered highly necessary. With regard to decontamination, notonly reduction of radiation but also reduction of wastes should be taken into consideration.
For decontamination, while engineering approaches are required, including physicalmethods such as dry ice blast, chemical methods such as chemical decontamination
using chemicals such as acid or alkali, and decontamination methods using parting
agents, it is indispensable to understand the contamination mechanism of target objects
in order to perform such decontamination activities effectively.
In the field of researches for clarifying the contamination mechanism, there are a
sufficient number of existing researches on the metal materials, which are used in pipes,

Basic direction279

tanks and so on to confine radioactive materials. However, there is almost no researchon the concrete of structures or radiation shields that does not directly touch radioactive
materials.
The inside the buildings of 1F is widely contaminated by the radioactive materials
emitted by the accident. Since most parts of the buildings consist of concrete, it isimportant to clarify the contamination mechanism of both concrete and radioactive
materials in principle in order to reasonably and effectively manage concrete wastes
resulting from the decontamination and the process of decommissioning. Therefore, thecontamination mechanism must be clarified in principle on the concrete exposed to theaccident and the subsequent environment and the process of decommissioning by
obtaining the basic data about sorption, penetration, and leaching of the typical nuclide(Cs, Sr, U, Pu, etc.) that should be well considered. In addition, from the mid-and-longterm viewpoint, it is necessary to establish the evaluation method based on theunderstanding of the contamination mechanism including the changes over time in thecontamination state and the penetration behavior in the concrete.
Even though a number of researches have been conducted on removal ofcontamination source in pipes during nuclear fuel reprocessing regarding thecontamination mechanism of the metal of the piping and equipment by radioactive
materials, few number of researches are found on the contamination mechanism of themetal of the piping and equipment in the environment of the 1F. While it is considerednecessary to clarify the contamination mechanism inside PCV or RPV exposed to thehigh temperature environment at the time of accident, it is not necessary to take into
consideration a special contamination mechanism such as penetrate into metals outside
PCV. As for the decontamination mechanism of resins and paints of cables, it isconsidered that it not necessary to conduct a special research on decontamination since
they can be replaced or removed.

Theme“Descriptions /Background issues”
on the interim report
Basic direction(6) Environmental fate studies of radioactive materials generated duringdecommissioning
It is essential to clarify the behavior of radioactive materials such as sorption, dispersion,
moving along with groundwater flow in shallow underground in order to conduct
environmental fate studies to ensure they will not affect the environment.
In order to properly evaluate and reduce the risk of future environmental impact
caused by radioactive materials in the Fukushima Daiichi NPS site, it is necessary toprovide proper evaluation and estimation of the environmental fate of radioactive
materials around the site via the shallow underground water and the surface water, or the
ports, the marine, and the air, and to provide appropriate environmental
countermeasures.
Targeted radioactive materials are (1) the radioactive materials that exist in the
ground or on the surface of the ground through the contaminated water leaked just after
the accident (137Cs, 90Sr, 3H, etc.), (2) the radioactive materials that poured into the ports
in past and accumulated on the bottom of the sea (137Cs, 90Sr, etc.), and (3) theradioactive materials that are contained in the contaminated water that will be generated
as the result of retrieval of the fuel debris or decommissioning and dismantling of thebuildings (including ion such as actinide and suspended solids) that can be the futuresource term impacting environment.
In order to estimate the impact of radioactive materials on the surrounding
environment, it is indispensable to understand the existence form and the transport
behavior of the radioactive materials as the required basic knowledge. Specifically, thetargets include the existence form of the radioactive materials in the underground water,
the distribution in the soil, the advection and diffusion behavior in the underground water,
the existence form and the advection and diffusion in the surface layer, the existence form
and the molten and diffusion behavior of the radioactive materials in the seawater in theport and on the bottom of the sea, and the transportation behavior to the surrounding
environment through marine or air.
Although all of those depend on the characteristic of the intermedium such as theproperty of soil and the geological condition, since the measurement work on 1F is limited,
it is necessary to aim at establishing the evaluation method in a similar environment.
In addition, in order to provide the accurate future estimation of the environmental
fate, it is also necessary to develop the monitoring technology to identify the accuratecontamination condition and the analysis technology to simulate the transportation
behavior of the radioactive materials. As for the monitoring technology, the technology forlong term and continuous remote measurement and the mapping and behavior
identification technology using the big data are expected. As for the simulation
technology, the creation model that can be used to analyze the behavior (influence of280unsaturated layer, kinetic evidence, etc.) specific to shallow underground and theestimation technology using the code are expected.
It is also important to aim at reducing the risk of the radioactive materials as
environmental countermeasure. While a number of technologies can be developed
including control of the amount of underwater, soil improvement, stabilizing agent, sorbent
for purification of contaminated materials, and the permeable reaction wall, it is necessary
to examine the factors that have priority since they may have critical influence on thedecommissioning works.
In order to provide reasonable environmental fate studies for the radioactive
materials, it is important to proceed with considerations on the risk of environmental
influence so the development of the evaluation method related to the risk of environmental
influence has to be taken into account from this viewpoint.

281

Attachment 14 R&D medium-to-long-term plan
R&D Mid-and-Long-term Plan
- Major Processes and Approach to Major R&D Activities for Decommissioning [Fuel Debris Retrieval] of the Fukushima Daiichi NPS of TEPCO FY

2020

2021

2022

2023

2026

Phase 2
Key
Milestones[A] Trial retrieval/
Gradual expansion of fuel debris retrieval

2032

Phase 3-(1)Start of debris
retrieval at the first
implementing unit

Further expansion of fueldebris retrieval scale

Gradual expansion of retrieval scale[1] Experimental retrieval at Unit 2

Ledgends :Improve the environment inside the building

Trial retrieval
and internal
investigations *1: Reveraging the Naraha Remote Technology Development Center

Manufactue, test and install retrieval device, etc. *1,
Operational change of the safety systems, Removal of
obstacles in openings, etc.

Main Processes 1)
Main R&D activities
Key Technical issues
1) Descrubed based on TEPCO DAP 2021

To fuel debris characterization

Development of technology for detailed PCV internal[2] Gradual expansion of retrieval scale at Unit 2: Where R&D results are applied: Linkage between processes

Improve the environment
Fuel debris retrieval facility/safety system/fuel debris temporary storage facility/maintenance facility
Installation

Design, manufacturing and testing (Reveraging of the Naraha Remote Technology

Fuel debris retrieval (Gradual expansion of scale)Operation of temporary storage facility for fuel debris
Operation of maintenance equipment

Technological development for gradual expansion of retrieval scaleTechnical issues for gradual expansion of fuel debris retrieval(1) Providing retrieval system (improving device durability, verifying passing with transport truck)(2) Manufacturing equipment for gradual expansion of fuel debris retrieval(3) Preparing the safety systems (gas system, water treatment system, criticality control system)(4) Preparing fuel debris and waste storage facilities

To fuel debris characterization[1] Environmental improvement of inside and outside the buildings
Environmental improvement
inside and outside the Unit 1 buildings

Inside the buildings: Radiation dose reduction/Removal of obstacles, etc.
Outside the buildings: Removal of exhaust stacks at Units 1/2, removal of transformers, etc. => Secure the site *2

Environmental improvement
inside and outside the Unit 3 buildings

Inside the buildings: PCV water level lowering/Radiation dose reduction, etc.
Outside the buildings: Removal of exhaust stacks and removal of transformers at Units 3/4,
Improvement of the basement floor, etc. => Secure the site *2

Development of technology for environmental improvement inside/outside reactor buildings
Development of assessment methods for deterioration status, etc., of building frames and equipment*2: Added buildings, maintenance buildings, storage buildings

Lowering of PCV waterlevel was achieved[2] PCV/RPV internal investigations, Characterization of fuel debris
PCV internal
investigations at Unit 3

Development of technology for PCV internal investigations at Unit 1

Development of RPV internal investigation technology/Advancement of RPV internal investigation[B] Further expansion of fuel debris retrieval

Development of technology for RPV quickinternal investigation

To RPV internal investigation design

To retrieval facility design
Characterization of fuel debris

Development of Analytical and Estimation Technologies
for Fuel Debris Characterization

Demonstration of analytical technology(Utilization of radioactive material analysis and research facilities)[3] Retrieval equipment, safety systems & maintenance equipment, and storage facility *3

Conceptual study*3: Main processes for proceeding with the study on Unit 3 as a preceding case

Reflecting Findings obtained
after the start of a gradual scale
expansion

Reflecting the results of Trial
retrieval and• Internal investigation*4: Installationp process for side-access retrieval
Design of retrieval installations *4

Verification

Design of safety system

Study of safety systems

Manufacturing, installation and testing of retrieval equipment and training
facility *4 and fuel debris retrieval
Production, installation and testing

Promotion of development for verification of
actual site applicability
Development of technologies toward further expansion of fuel debris retrieval scale
Side-access retrieval method: Examination and enhancement of elemental technologies, verification of theactual site applicability
Top access method: Conceptual study, examination and enhancement of elemental technologies
Technical issues toward further expansion of fuel debris retrieval(1) Development of fuel debris retrieval method (for side access/top access/common in side and top accesses)
Development of installation method for access equipment, development of dismantling technology, development of technology for concept delivery of large structure removal,
development of processing/recover technologies, and others(2) Development of safety systems(Liquid systems / gas systems, Criticality control technology)
Improving removal function of radioactive material, development of boron control technology including boron recovery, behavior evaluation/recovery of dust(3) Enhancement and advancement of containment technology(4) Development of safety assessment technology
Development of containment technology of radioactive materials by passive control
Data acquisition on debris scattering rate, Analysis method for exposure assessment
Design of facilities, manufacturing of equipment, and construction of facilities

Development of maintenance policy, Study of maintenance methods

Start of operation

Development of remote device support technologies
Basic design of fuel debris storage facilities

Equipment design, Equipment manufacturing

Trial operation, Start of operation

Development of technology for collection, transfer and storage of fuel debris
Development of support technology for 1F decommissioning integrated management[1] Characterization[C] Waste Management

Characterization(1)

Conducting analysis, effective characterization, etc.

(2)Technical development for sampling of high radiation dose sample[2] Storage
Storage(1)

Study on Storage method[3] Processing/disposal
Development of processing/disposal
concept and safety asessment methods

(2)
(1)
(2)
(3)

Reducing of amount of waste
Processing technology (Development of processing technology and temporary processing technology)
Disposal technology (Creating disposal concept suitable for solid wast)
Study on waste stream

In conducting the study on R&D medium-to-long-term plan, TEPCO, Agency for Natural Resources and Energy and NDF consider the required results, timing needed and implementing structures while sharing
information with MEXT through the R&D plan meetings. Center of the World Intelligence project and the Government-led R&D Program on Decommissioning and Contaminated Water Management will further
strengthen their ties by sharing the results, and the outcome of basic and fundamental research will also be appropriately reflected in the decommissioing site according to the progress of the research.

282

Attachment 15 Current progress of the Project of Decommissioning and Contaminated Water
Management200

1. Grasping state inside reactor, characterizing of fuel debris, and internal investigation
1-(1) Advancement of comprehensive grasping of state inside reactor (FY 2016 - 2017)(Related projects) Advancement of accident progression analysis technology for assessing
conditions inside reactor (FY 2011)
Assessing conditions inside reactor by advancement of accident progression
analysis technology (FY 2012 - 2013)
Assessing conditions inside reactor through application of severe accident
analysis code (FY 2014)
Advancement of grasping conditions inside reactor by accident progression
analysis and actual data, etc. (FY 2015)
1-(2) Development of analytical and estimation techniques for characterization of fuel debris (FY2019 - 2021), Development of analytical and estimation techniques for characterization offuel debris (Development of estimation technologies for variation character across the ages)(FY 2019 - 2021), Development of analytical and estimation techniques for characterizationof fuel debris (Development of technologies for improving analytical accuracy and estimationof thermal behavior of fuel debris) (FY2020 – FY2021)(Related projects) Characterization of fuel debris using mock-up debris and development of fueldebris processing technologies (FY 2011 - 2014)
Construction of material accountancy method related to fuel debris (FY 2011 2013)
Property analysis of actual debris (FY 2014)
Grasping properties of fuel debris (FY 2015 - 2016)
Development of techniques for characterizing and analyzing fuel debris (FY2017 - 2018)
1-(3) Development of technologies for in-depth investigation of PCV inside (FY 2019 - 2021)(Related projects) Development of investigation technologies of inside of PCV (FY 2011 - 2013)
Development of investigation technologies of inside of PCV (FY 2014 - 2015)
Development of investigation technologies of inside of PCV (FY 2016 - 2017)
Development of technologies for in-depth investigation of PCV inside (FY 2017
- 2018)
1-(4) Development of investigation technology inside RPV (FY 2016 - 2021)(Related projects) Development of investigation technologies inside RPV (FY 2013 - 2015)
1-(5) Development of technologies for the detection of fuel debris inside reactors (using muon) (FY
2014 - 2015)
200 Information Portal for the Research and Development for the Fukushima Daiichi Decommissioning(http://www.drd-portal.jp/)283

2. Retrieval of fuel debris
2-(1) Development of technologies for retrieving fuel debris to be gradually expanded in scale (FY2019 - 2021)(Related projects) Development of sampling technologies for retrieving fuel debris and internal
structures (FY 2017 – 2019)*Change in the project name: With the publication of “Decommissioning R&D Plan 2020 of The
75th meeting of the Team for Countermeasures for Decommissioning and Contaminated Water
Treatment/Secretariat held on February 27, 2020, Material 4: Others
2-(2) Development of retrieval method of fuel debris (FY 2019 - 2021), and safety system (FY 2019
- 2021)(Related projects) Development of technologies for retrieving fuel debris and internal structures(FY 2014)
Advancement of retrieval method and system of fuel debris and internal
structures (FY 2015 - 2018)
Development of technologies toward further expansion in scale for retrieving
fuel debris and internal structures (development of technologies for dust
collecting system) (FY2019 - 2020)*Change in the project name: With the publication of “Decommissioning R&D Plan 2020 of The
75th meeting of the Team for Countermeasures for Decommissioning and Contaminated Water
Treatment/Secretariat held on February 27, 2020, Material 4: Others
2-(3) Advancement of fundamental technologies for retrieving fuel debris and internal structures(FY 2017 - 2018)(Related projects) Development of fundamental technologies for retrieving fuel debris and internal
structure (FY 2015 - 2016)
2-(4) Development of PCV closed water circulation systems (FY 2018 - 2019)
2-(5)-1 Development of repair methods for leak spots in PCV (FY 2016 - 2017)(Related projects) Development of identification technology of leaks in PCVs (FY 2011 - 2013)
Development of repair method for PCVs (FY 2011 - 2013)
Development of repair (water stoppage) technology toward water filling in
PCV (FY 2014 - 2015)
2-(5)-2 Full-scale test of repair methods for leak spots in PCV (FY 2016 - 2017)(Related projects) Full-scale test of repair methods for leak spots in PCV (FY 2014 - 2015)
2-(6) Development of evaluation methods of seismic performances of RPV and PCV and theimpacts of the damages (FY 2016 - 2017)(Related projects) Development of evaluation methods for the structural integrity of RPV andPCV (FY 2011 - 2013)
Development of evaluation methods for the structural integrity of RPV andPCV (FY 2014 - 2015)
2-(7) Development of corrosion inhibition technology for RPV and PCV (FY 2016)284

(Related projects) Full-scale test of repair methods for leak spots in PCV (FY 2014 - 2015)
2-(8) Development of criticality control technologies of fuel debris (FY 2012 - 2016)
2-(9) Development of technologies for non-destructive detection of radioactive material depositedin S/C, etc. (FY 2014)
2-(10) Development of remote decontamination technology in the reactor building (FY 2014 - 2015)(Related projects) Development of remote decontamination technology in the reactor building (FY2011 - 2013)
2-(11) Formulation of comprehensive radiation dose reduction plan (FY 2012 - 2013)
2-(12) Development of technologies for containing, transferring, and storing fuel debris (FY 2016 2021)(Related projects) Development of containing, transferring, and storing technologies of fuel debris(FY 2014 - 2015)
2-(13) Development of technologies for environmental improvement inside reactor buildings(FY2020 - 2021)
2-(14) Development of technologies for maintaining remote equipment (FY2021)
2-(15) Development of support technologies for integrated management of Fukushima Daiichi NPS
decommissioning (Development of continuous monitoring system inside PCV) (FY2021)
3. Waste management
3. Research and development of processing and disposal of solid waste (FY 2019 - 2021)(Related projects) Development of technologies for processing/disposal of secondary waste bytreatment of contaminated water (FY 2012)
Development of technologies for processing/disposal of radioactive waste (FY2012)
Research and development of processing/disposal of solid waste (FY 2013 2014)
Research and development of processing/disposal of solid waste (FY 2015 2016)
Research and development of processing/disposal of solid waste (FY 2017 2018)
Research and development of preceding processing methods and analyticaltechniques (FY 2018)
4. Spent fuel management
4-(1) Evaluation of long-term integrity of fuel assembly removed from SFPs (FY 2015- 2016)(Related projects) Evaluation of long-term structural integrity of fuel assemblies removed from
SFPs (FY 2012 - 2014)
4-(2) Investigation of method for processing damaged fuel, etc. removed from SFPs (FY 2013 2014)285

5. Contaminated water management
5-(1) Verification tests of tritium separation technologies (FY 2014 - 2015)
5-(2) Verification of technologies for contaminated water treatment (FY 2014)
5-(3) Large-scale verification of impermeable walls (frozen wall) (FY 2014)
5-(4) Development and verification of high-performance multi-nuclide removal equipment (highperformance ALPS) (FY 2014)286

Attachment 16

Major activities related to enhancing international collaboration

Table A16- 1 Intergovernmental Framework between Japan and other countries
Framework
Annual Japan-UK Nuclear
Dialogue

Japan-France Nuclear Energy
Committee
Japan-US Decommissioning
and Environmental
Management Working Group

Japan-Russia Nuclear
Working Group

Descriptions
This dialogue is held based on the appendix to the jointstatement of the Japan-UK top level meeting in April 2012,“Japan-UK Framework on Civil Nuclear Energy
Cooperation” (Since February 2012).
It was established under the joint statement of Japan–
France top-level meeting in October 2012 (Since February2012).
After the Fukushima Daiichi NPS accident in March 2011,
the establishment of the US-Japan Bilateral Commission
on Civil Nuclear Cooperation (the Bilateral Commission)
was announced in April 2012 based on the relationshipbetween Japan and the US to further reinforce bilateral
cooperation. Under this commission, “theDecommissioning and Environmental Management
Working Group (DEMWG)” was established (Since
December 2012).
The Nuclear Working Group was established after
confirming that Energy is one of the eight areas ofcooperation plan approved at the Japan-Russia top-levelmeeting in September 2016, (Since September 2016).

Table A16- 2 Inter-organizational Cooperation Agreement
Domestic

International

Descriptions

NDF

NDA

Exchange of information for various technical knowledge on
decommissioning, etc. and personal exchange are provided.
(Concluded in February 2015)

NDF

CEA

Exchange of information for various technical knowledge on
decommissioning, etc. and personal exchange is provided.
(Concluded in February 2015)

TEPCO

DOE

Umbrella Contract was made and information is exchanged
as needed. (Concluded in September 2013)

TEPCO

Sellafield, Ltd.

Information Exchange Agreement for site’s operation, etc.
was concluded. (September 2014)

TEPCO

CEA

Information Exchange Agreement on for decommissioning
was concluded. (September 2015)

JAEA

NNL

Comprehensive Agreement for advanced technology on
nuclear R&D, advanced fuel cycles, fast reactor, radioactive
waste

JAEA

CEA

Cooperation Agreement for specific technical issues onmolten core-concrete interaction, etc.

287

JAEA

Belgium Nuclear
Research Center

Agreement of Cooperation for Nuclear R&D and Researchon the accident of the Fukushima Daiichi

JAEA

Nuclear Safety
Research Center(Ukraine)

Memorandum for decommissioning research, etc. of the
Fukushima Daiichi NPS and Chernobyl was concluded.

JAEA

IAEA

Research Agreement on characterization of fuel debris

Table A16- 3 Dissemination of information to the world (Holding or attending International
Conference (from April 2020 to August 2021))
Conference Name

PeriodOrganization

The 64th IAEA Conference Side event

September,
2020

NDF
METI
TEPCO

ICRP/JAEA International Conference on Recovery After
Nuclear Accidents

December,
2020

NDF
METI
TEPCO
JAEA

Japan-UK Nuclear Dialogue

December,
2020

METI

IRPA 15, International Commission on RadiologicalProtection, Fukushima Special Session

January, 2020

TEPCO

OECD/NEA Fukushima Daiichi Accident 10years Report,
Web seminar

March, 2021

NDF

IAEA Fukushima Webinar

March, 2021

NDF

US Waste Management 2021

March, 2021

TEPCO
IRID

WTO/SPS Committee

March, 2021

METI

Japan-Russia Nuclear Working Group

March, 2021

METI

The 5th UK-Japan Nuclear Industrial Forum

June, 2021

NDF
METI
TEPCO
IRID
JAEA

The 27th International Conference on Nuclear
Engineering (ICONE27)

August, 2021

TEPCO
IRID

Fukushima Research ConferenceYear round

JAEA

288

Table16- 4 Dissemination of information to the world (on web (in English)
Site

Organization

Mid-and-long-term Roadmap towards the Decommissioning of TEPCO’s
Fukushima Daiichi Nuclear Power Station Units 1 to 4
（http://www.meti.go.jp/english/earthquake/nuclear/decommissioning/）

METI

Monthly report to the embassies concerning discharging and seawater
monitoring from the Fukushima Daiichi NPS

METI,
MOFA

Nuclear Damage Compensation and Decommissioning Facilitation
Corporation’s website（http://www.dd.ndf.go.jp/eindex.html）

NDF

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

NDF

Activities for Decommissioning（https://fukushima.jaea.go.jp/english/）

JAEA

IRID website (http://irid.or.jp/en/)

IRID

Fukushima Daiichi Decommissioning Project（http://www7.tepco.co.jp/responsibility/decommissioning/index-e.html）

TEPCO

Providing English version of Press release to foreign media

TEPCO

Management Office for the Project of Decommissioning and Contaminated
Water Management(https://en.dccc-program.jp/)

MRI(Businessconsignee)

Table A-16 - 5 Major collaborative projects with foreign organizations
Contents/Period of project

Participating
Organization

・Knowledge and experience obtained from the efforts onchallenges of decommissioning and recovery of damaged
nuclear power facilities (regulations, technologies, systems,
and strategies) are shared among the relevant countries.
・Project period : 2015 to 2017

NDF

Project
IAEA Project

DAROD

OECD/NEA Project

BSAF

・Researching institutions and governmental organizations
from eleven countries joined to conduct benchmark study
using severe accident analysis codes developed by theseorganizations to find out how the accident in the Fukushima
Daiichi NPS progressed and how the fuel debris and FPs
spread inside the reactors. Knowledge and findings related
to the modeling of phenomenological issues obtained bymember countries’ organizations are being utilized.
・Data measured during the accident and information
database regarding the post-accident radiation levels areshared.

289

IRID
JAEA
TEPCO

・Project period : 2015 to 2018

ARC-F

・In succession to the BSAF project, researching institutions
and governmental organizations from twelve countries joined
to investigate the situation of the accident in more detail andutilize it for further researches to improve safety of lightwater reactor
・Project period : 2019 to 2021 (scheduled)

NRA
IR
JAEA

PreADES・Sharing characteristics information that helps tounderstand properties of fuel debris such as its phase stateand composition.
・Enhancing “Fuel debris Analytical Chart” that summarizes
needs and priority of fuel debris analysis.
・Maintenance of tasks after analysis and analysis facility
information
・Project period : 2018 to 2020 (scheduled)

METI
NRA
IR
JAEA
IRID
NDF
TEPCO

TCOFF

・In reference to the accident progression of the Fukushima
Daiichi NPS, (1F) advancing molten core and molten fuelmodels, FP migration behavior model and thermodynamic
database as their basis. Based on the material scientificknowhow, evaluating details of molten core and fuel on
condition of 1F accident, and characteristics of fuel debris
and its producing mechanism. Then, providing materialscientific knowhow and result of detail evaluations tointernational cooperation project including PreADES, ARC-F,
TAF-ID, and domestic decommissioning project like IRID.
・Project budget was contributed from MEXT.
・Project period : 2017 to 2019

MEXT
JAEA
IR
Tokyo
Institute ofTechnology

EGCUL

・Discussing on characterization method for waste fromunknown derivation

290

METI
NDF
JAEA
TEPCO