Technical Strategic Plan 2018 for the Decommissioningof the Fukushima Daiichi Nuclear Power Station ofTokyo Electric Power Company Holdings, Inc.

October 2, 2018
Nuclear Damage Compensation and Decommissioning
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.

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Table of Contents
1. Introduction ................................................................................................................................... 6
1.1 Enhancement of structure and system toward appropriate and steady
implementation of decommissioning ..................................................................................... 6
1.2 Strategic Plan ......................................................................................................................... 8
1.2.1 Positioning of the Strategic Plan ...................................................................................... 8
1.2.2 Overall Structure of Strategic Plan 2018 ......................................................................... 9
2. Decommissioning of the Fukushima Daiichi NPS as risk reduction strategies ......................... 12
2.1 Basic concept of the decommissioning of the Fukushima Daiichi NPS ............................... 12
2.2 Progress status of the decommissioning of the Fukushima Daiichi NPS ............................ 12
2.3 Basic concept of reducing risk of radioactive materials ....................................................... 15
2.3.1 Quantitative grasping of risk .......................................................................................... 15
2.3.2 Identification of Risk Source and Risk assessment ...................................................... 15
2.3.3 Risk Reduction Strategy ................................................................................................ 18
3. Technological strategies toward decommissioning of the Fukushima Daiichi NPS .................. 22
3.1 Fuel debris retrieval work ..................................................................................................... 22
3.1.1 Sectoral target................................................................................................................ 22
3.1.2 Sectoral strategies ......................................................................................................... 22
3.1.3 Technical issues for promoting sectoral strategies and future plans ............................ 27
3.2 Waste management.............................................................................................................. 57
3.2.1 Sectoral target................................................................................................................ 57
3.2.2 Sectoral strategy ............................................................................................................ 57
3.2.3 Technical issues and plans for promoting sectoral strategies ...................................... 63
3.3 Contaminated water management ....................................................................................... 67
3.3.1 Sectoral target................................................................................................................ 67
3.3.2 Sectoral strategies ......................................................................................................... 67
3.3.3 Technical issues and plans for promoting strategies .................................................... 70
3.4 Fuel removal from spent fuel pools ...................................................................................... 74
3.4.1 Sectoral target................................................................................................................ 74
3.4.2 Sectoral strategies ......................................................................................................... 74
3.4.3 Technical issues and future actions for prompt strategies ............................................ 77
3.5 Other specific measures ....................................................................................................... 82
3.5.1 Sustaining of reactor cold shutdown status ................................................................... 82
3.5.2 Radiation dose reduction and contamination expansion prevention all over the power
station ......................................................................................................................... 82
3.5.3 Plan for decommissioning measures for nuclear reactor facilities ................................ 89
3.5.4 Concrete efforts toward securing safety ........................................................................ 89
3.6 Comprehensive efforts for the decommissioning project of the Fukushima Daiichi
NPS ..................................................................................................................................... 92
4. Handling critical enablers for smooth operation of the project................................................... 94
4.1 Actions toward improvement of working environment and conditions ................................. 94
4.2 Concept of securing safety and promotion of cooperation................................................... 96
4.3 Enforced management structure for steady mid-and-long-term decommissioning ............. 97
4.4 Developing and securing human resources ......................................................................... 99
4.4.1 Developing and securing operators and engineers ..................................................... 100
4.4.2 Fostering the next generation to handle the decommissioning of the Fukushima Daiichi
NPS ........................................................................................................................... 101
5. R&D initiatives .......................................................................................................................... 103
5.1 Basic policy for R&D ........................................................................................................... 103
5.1.1 Basic policy .................................................................................................................. 103
5.1.2 Entire perspective of R&D ........................................................................................... 103
5.2 R&D of decommissioning required for on-site work/engineering ....................................... 106

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5.3 Enhancement of basic study and R&D infrastructure for the success of thedecommissioning project ................................................................................................... 107
5.3.1 Essential R&D Themes and its strategic promotion derived from the needs .............. 107
5.3.2 Construction of basic research centers and R&D infrastructure for mid-and-long-term
prospects .................................................................................................................. 108
6. Enhancement of international cooperation .............................................................................. 110
6.1 Significance of international cooperation ........................................................................... 110
6.2 Facilitation of international cooperation activities ............................................................... 111
6.2.1 Enhancement of partnership with overseas decommissioning agencies ................... 111
6.2.2 Integrating and utilizing wisdom in the world ............................................................... 113
6.2.3 Dissemination of information to the global society ...................................................... 114
6.3 Close cooperation with relevant domestic organizations ................................................... 116
7. Local community engagement and further enhancement of communication .......................... 117
7.1 Approaches for local community engagement and further enhancement ofcommunication .................................................................................................................. 117
7.2 Actual effort for better communication................................................................................ 118
7.3 Further spread of communication and measures to Reputational damage ....................... 119
List of Acronyms/Glossaries ......................................................................................................... 121
List of Attachment ......................................................................................................................... 124

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Table of Figures
Fig. 1 Division of roles of related organizations in the decommissioning of the Fukushima
Daiichi NPS

8Fig. 2 Positioning of the Strategic Plan based on the system of the Reserve Fund9

Fig. 3 Three principles and measures

13

Fig. 4 Reduction of risks contained in the Fukushima Daiichi NPS

17

Fig. 5 Examples of Risk Levels of Major Risk Sources at the Fukushima Daiichi NPS

18

Fig. 6 Investigating flow toward determination of fuel debris retrieval method (draft plan)

25

Fig. 7 Estimation of fuel debris distribution of Units 1 to 3, access Route and condition ofsurrounding structures

32

Fig. 8 Further internal investigation schedule and image of investigation equipment

34

Fig. 9 Example of containment function (gas phase) by negative pressure control

36

Fig. 10 Build example of containment function (liquid phase)

40

Fig. 11 PCV lower area water sealing (Examination example)

42

Fig. 12 Fundamental technologies for retrieval of fuel debris and internal structures

52

Fig. 13 Technical issues and further plan on fuel debris retrieval

56

Fig. 14 Summary of Waste Hierarchy at the NDA, UK and Countermeasures at the
Fukushima Daiichi NPS

60

Fig. 15 Technical issues and further plan on waste management

66

Fig. 16 Changes in the amount of radioactive materials in the stagnant water in the buildings
69
Fig. 17 Transitions of the generated amount of contaminated water and the inflow amount ofgroundwater/storm sewage, etc. into the buildings

71

Fig. 18 Technical issues and further plan on contaminated water management (Process
Chart)

73

Fig. 19 Storage situation of spent fuel etc.(as of Jun 28, 2018)

77

Fig. 20 Technical issues and further plan on removing fuel from the spent fuel pool81

Fig. 21 Radioactive material concentration in seawater in the port

83

Fig. 22 Evaluation of annual radiation exposure dose at site boundaries due to radioactive
materials (CS-134 and Cs-137) emitted from Units 1 to 4 reactor buildings

85

Fig. 23 Effective dose evaluation at site boundaries

87

Fig. 24 Need for a mid- and-long-term viewpoint overlooking the entire activities

93

Fig. 25 Changes in each worker’s monthly exposure dose (Monthly average dose)

95

Fig. 26 NDF and TEPCO Organizational System for Project Control

98

Fig. 27. Whole picture of R&D structure of the decommissioning of the Fukushima Daiichi
NPS (As of FY 2018)

105

Fig. 28 A conceptual picture of the division of roles among main R&D institutions for thedecommissioning of the Fukushima Daiichi NPS

4106

Fig. 29 Structural drawing inside Reactor building

123

Fig. 30 Structural drawing inside Reactor Pressure Vessel(RPV)

123

Table 1 Major risk sources in the current risk map

16

Table 2 Image of fuel debris retrieval by a step-by-step approach (example of Unit 2)

29

Table 3 Solid Waste Storage Management Status

61

Table 4 Description of the Projects Included in the Withdrawal Plan

99

Table 5 Cooperative relationship among the institutions for the decommissioning ofFukushima Daiichi

111

Table 6 Approach for dissemination of information to the world

5115

1. Introduction
The overall approach to the decommissioning of the Fukushima Daiichi Nuclear Power Stationof Tokyo Electric Power Company Holdings, Inc. (“Fukushima Daiichi NPS”) started under the Midand-Long-term Roadmap towards the Decommissioning of TEPCO’s Fukushima Daiichi NPS Units
1-4 (“Roadmap”), released by the Japanese Government in December 2011.
Urgent issues, such as treating contaminated water and removing fuel from the spent fuelpools, have been given top priority in this effort. However, to complete the decommissioning, longterm measures are required such as fuel debris retrieval work, and so it is essential to prepare amid- and long-term decommissioning strategy. On August 18, 2014, the former Nuclear Damage
Compensation Facilitation Corporation was reorganized into the Nuclear Damage Compensation
and Decommissioning Facilitation Corporation (NDF), a new organization responsible for technicalstudies needed to proceed with the decommissioning properly and steadily from the mid- to longterm 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.
Seven years have passed since the Fukushima Daiichi NPS accident, and prospects for shortterm measures are becoming clearer, progress has been made at the accident site in the treatmentof contaminated water including the construction of the land-side impermeable wall and in removing
fuel from the spent fuel pools. In addition, the site work environment has improved. Regarding
measures over the mid- and long-term, investigation and R&D have advanced in the areas of fueldebris retrieval and waste management. The Roadmap has also been revised to establish thepolicy for fuel debris retrieval and the basic concept of solid waste management.
Strategic Plan 2018 provides updates on these developments in decommissioning and the roleof NDF and analyzes them closely, and so some areas of the document have been reorganized.

1.1 Enhancement of structure and system toward appropriate and steady
implementation of decommissioning
As the decommissioning project is shifting focus to mid- and-long-term challenges, thestructures and systems for the decommissioning are being reinforced to ensure that thedecommissioning project continues and that mid- and-long-term issues are properly addressed.
To ensure the implementation of the decommissioning financially, a law to partially revise the
Nuclear Damage Compensation Facilitation Corporation Act was established in May 2017 andcame into effect in October of the same year. Under the Revised NDF Act, NDF has been assigned
the new duty of managing the reserve fund for decommissioning. Tokyo Electric Power Company
Holdings, Inc. (TEPCO) is required to reserve an amount of funds at NDF every year that NDF
determines to be necessary for the appropriate and steady implementation of the decommissioning
and is authorized by the competent minister (the Minister of Economy, Trade and Industry). TEPCO

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will proceed with the decommissioning while withdrawing the reserved funds in accordance with
“Withdrawal plan for reserve fund” (“Withdrawal Plan”), which will be prepared jointly by NDF andTEPCO and approved by the competent minister. This ensures that the necessary amount of funds
for the appropriate and steady implementation of the decommissioning will be reserved, building aconsistent decommissioning structure that works irrespective of TEPCO’s earnings or other factors.
In this decommissioning reserve fund system, NDF will play even greater roles and responsibilities
than before as the major supervisor and administrator of the decommissioning project conductedby TEPCO. These include (1) appropriate management of the funds for decommissioning, (2)maintenance of an appropriate system for executing the decommissioning, and (3) steady work
management based on the decommissioning reserve fund system.
Likewise, TEPCO, as the operator of the project, is reinforcing its project management
structure with the goal of implementing the ongoing activities steadily while responding to mid- andlong-term issues in a strategic manner. TEPCO and NDF have just launched this decommissioning
reserve fund system under the project management framework formed to optimize the overall
decommissioning plan.
Specifically, prior to the preparation of the Withdrawal Plan, NDF will present to TEPCO thework targets and major tasks to be included in the Withdrawal Plan by dividing them into major
projects in accordance with “The Policy for Preparation of Withdrawal Plan for Reserve Fund for
Decommissioning” (“The Policy for Preparation of Withdrawal Plan”). NDF will also support theappropriate and steady implementation of the decommissioning through activities such as
evaluating the adequacy of TEPCO’s efforts in the course of jointly preparing the Withdrawal Planin terms of project execution (including the aspect of collaboration and communication with local
communities).
NDF presented the first of the Policy for Preparation of Withdrawal Plan in January 2018. The
first Withdrawal Plan, prepared based on this Policy, was submitted by NDF and TEPCO for
approval by the Minister of Economy, Trade and Industry in March the same year. The Plan was
successfully approved in April.1
The division of roles among the organizations directly involved in the decommissioning of the
Fukushima Daiichi NPS (the Japanese government, NDF and TEPCO), as well as organizations
specializing in R&D (the International Research Institute for Nuclear Decommissioning [IRID] andthe Japan Atomic Energy Agency [JAEA]), is shown in Fig. 1 , which also indicates how theabovementioned systems are implemented. Among these roles, R&D are discussed in Chapter 5,and dialogue with local residents and communities is described in Chapter 7.

1“Withdrawal Plan for Reserve Fund for Decommissioning FY 2017” approved on April 11, 2018, NDF
http://www.ndf.go.jp/oshirase/sonota_20180411.html

7Nuclear Emergency Response Headquarters (Director: Prime Minister)Council for the Decommissioning of TEPCO's Fukushima Daiichi Nuclear Power Station

Nuclear Regulation Authority(Chairperson: Chief Cabinet Secretary) Approval of implementation
plan, pre-service inspection,
welding inspection, etc.

Mid-and-Long-Term Roadmap

The Team for Countermeasures for Decommissioning and Contaminated Water Treatment(Head of Team: Minister of Economy, Trade and Industry)

Progressmanagement

Granting of
project budget

R&D organization

Apply forapproval

Approve

Nuclear Damage Compensation and Decommissioning Facilitation Corporation

 Development of a mid-to-long-term strategy
Withdrawal plan Tech. support for progress mgmt of important issues (in collaboration)
 R&D planning and progress management
Strategic  Strengthen international partnershipReserve Fund forplan  Management of Reserve Fund for
Decommissioning
Decommissioning

International Research Institute for
Nuclear Decommissioning (IRID)*
Share progress
Japan Atomic Energy Agency (JAEA), and challengesetc.
 R&D implementation
ReportAdvice, guidance, recommendations,supervision and support
Reserverelated to project management, etc.

Monitoring, examination

ReportReport on theoutcome

Minister of Economy, Trade and Industry

Presentation ofimportant issues

Apply for report Progress management of each countermeasure according toMid-and-Long-Term Roadmap

ReportBidirectional dialogue

Local residents and communities

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

Disburse-mentTokyo Electric Power Company Holdings, Inc. (Fukushima Daiichi Decontamination and Decommissioning Engineering Company)
 Contaminated water management, fuel removal from spent fuel pools, fuel debris retrieval, waste disposal,
improvement of power plant premises and working environment, management of Units 5 and 6, etc.
 Through TEPCO's management reform, the funds required for decommissioning should be reserved
responsibly (Implementation of new comprehensive special project plan)

Implementation
plan* 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 in the decommissioning of the
Fukushima Daiichi NPS

1.2 Strategic Plan
1.2.1

Positioning of the Strategic Plan

NDF has issued a Technical Strategic Plan for Decommissioning of the Fukushima Daiichi
NPS of Tokyo Electric Power Company (“Strategic Plan”) every year since 2015 with the goal ofproviding reliable technological grounds for the government’s Roadmap and contributing to its
smooth and steady implementation (Attachment 1).
Strategic Plan 2015, the first edition, provided specific discussions on key decommissioning
issues expected over the mid and long term, which had not been sufficiently considered before,
and established a policy for the risk reduction strategy. Subsequently, Strategic Plan 2016 wasissued as a revision to the first edition to reflect the progress made in field work, technologicaldevelopment and other areas, providing input to the determination of the policy on fuel debris
retrieval and the establishment of the basic policies on solid waste management, which are among
the milestones (key target processes) set in the Roadmap. Strategic Plan 2017 was released andcontributed to these two milestones by presenting strategic proposals based on technologicalgrounds. These strategic proposals were incorporated into the September 2017 revision of the
Roadmap, leading to the achievement of the two milestones: the determination of the policy for fueldebris retrieval and the establishment of the basic policies for solid waste.
Thus, the previous Strategic Plans focused on the two issues that NDF should address as partof its decommissioning strategy over the medium and long term—fuel debris retrieval and waste
management—and continued deliberations to reach the two milestones. To take concrete action
for fuel debris retrieval, future discussions should consider, in addition to these issues, the8relationship and consistency with other activities, such as the treatment of contaminated water andthe removal of fuel from the spent fuel pools. To address these needs, it was decided that future
editions of the Strategic Plan should additionally include the topics of contaminated water
management, fuel removal from the spent fuel pools and other associated issues and should
propose future directions from mid- and long-term perspectives that examine the overall efforts
toward the decommissioning of the Fukushima Daiichi NPS. Any issues that are identified throughthese studies as requiring current attention will be reflected in the Policy for Preparation ofWithdrawal Plan for Reserve Fund for Decommissioning that NDF will present to TEPCO (seeFig.2).
Government*) Continually revise based on the progress of programs

Mid-and-Long-Term Roadmap*)

Next R&D Plan

(Project of Decommissioning andContaminated water management)

Proposals for Revise

Strategic Plan (annual)

NDF

The Policy for Preparation ofWithdrawal Plan for Reserve Fundfor Decommissioning (annual)

Withdrawal Plan for Reserve Fundfor Decommissioning (annual)

Feedback

TEPCO

Jointly created with NDF and approved by METI
• Contaminated water management PG• Pooled fuel retrieval PG• Fuel debris retrieval PG• Waste management PG• Site area /Work environment improvements PG• Management of Unit 5 and 6 PG
• Non-program work/Operational management

On-site work
Engineering

Planning from
Mid-and-long-term aspect

Feedback

Technological
Development

Promotion of Decommissioning project

Fig. 2 Positioning of the Strategic Plan based on the system of the Reserve Fund
1.2.2

Overall Structure of Strategic Plan 2018

Strategic Plan 2018 consists of seven chapters.
Chapter 1 ( Introduction) states that the decommissioning of the Fukushima Daiichi NPS ismoving to a phase where issues over the medium and long term should be addressed from wider
perspectives that consider the relationship and consistency among different projects, and that the
introduction of the decommissioning reserve fund system has placed greater responsibilities androles on NDF than before. The chapter also notes that against this background, Strategic Plan 2018
addresses issues that had not been sufficiently covered by previous Strategic Plans, such as
contaminated water management and fuel removal from the spent fuel pools, to provide a strategy
that would facilitate the smooth implementation of the overall decommissioning project.
Chapter 2 ( Decommissioning of the Fukushima Daiichi NPS as risk reduction strategies )9presents a basic policy for the decommissioning of the Fukushima Daiichi NPS as a strategy to
reduce risks, along with risk reduction tools that should be used in carrying out this policy, such as
near-term targets, the basic concept of risk reduction, the approach to priority, and the stance on
responding to a temporary increase in the risk level.
Chapter 3 ( Technological strategies toward decommissioning of the Fukushima Daiichi
NPS ) sets targets in each of the four areas: fuel debris retrieval, waste management, contaminatedwater management, and fuel removal from the spent fuel pools. Area-specific strategies to achieve
the targets are described, along with technological challenges in implementing these strategies,
and a plan for future actions to address them. This chapter also discusses the need to establish an
overall plan that helps maintain consistency throughout the decommissioning project bycoordinating these concurrent efforts with other specific measures, yet pursues optimized entire
planning while supposing interim targets to the extent possible.
In accordance with the fuel debris retrieval policy, Section 3.1 (Fuel debris retrieval ) of

Chapter 3 states how to proceed with studies to determine the fuel debris retrieval method for the
first reactor in FY 2019 and describes the directions and progress of efforts to intensify andaccelerate preparatory engineering and technological development for this purpose. The section
also proposes example of fuel debris retrieval image and technological issues to be considered ina step-by-step approach.
Section 3.2 ( Waste management) presents, in accordance with the basic policies onprocessing and disposing of solid waste, specific targets to gain a technological vision for waste
processing and disposal for around FY 2021, along with a plan for R&D needed to reach thesetargets.
Section 3.3 ( Contaminated water management) proposes, in view of the progress being madeto complete the treatment of the stagnant water in the buildings by the end of 2020, the directionof efforts toward treatment of contaminated water in the reactor buildings after the start of fueldebris retrieval work.
Section 3.4 ( Fuel removal from spent fuel pools ) proposes the direction of the efforts toward
the appropriate storage of spent fuel throughout the Fukushima Daiichi NPS, including Units 5 and6, and the efforts to determine the future processing and storage methods for spent fuel (e.g.
evaluation of the long-term integrity of such fuel), in addition to how fuel removal work should be
carried out in each unit according to plan.
Chapter 4 ( Handling critical enablers for smooth operation of the project ) provides
discussions on key issues contributing to the smooth progress of the overall project in wider fields
that go beyond the topics addressed in previous editions, such as R&D, international cooperation,and communication with local communities, in addition to technological considerations describedin Chapter 3.
Chapter 5 (R&D initiatives ) focuses on R&D efforts expected in the future. These include
performing even more needs-oriented R&D based on the results of preparatory engineering andincorporating practices such as progress management and implementation evaluations under the10

project management system. The chapter also discusses, from the mid- and long-term point ofview, the establishment of a center for basic research and the construction of R&D infrastructure,
as well as the importance of fundamental R&D.
Chapter 6 (Enhancement of international cooperation ) describes the need to strengtheninternational cooperation, such as partnerships with institutions involved in decommissioning
projects in countries with legacy nuclear sites, in order to gather wisdom and knowledge fromaround the world, and proposes efforts that should be made in this regard.
Chapter 7 (Local community engagement and further enhancement of communication ) notesthat as fuel debris retrieval work and other key activities move into full gear, the organizations
involved should consider and implement ways to ensure closer communication among them. The
chapter describes how these organizations should work together toward that end.

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2. Decommissioning of the Fukushima Daiichi NPS as risk
reduction strategies
2.1 Basic concept of the decommissioning of the Fukushima Daiichi NPS<Basic concept of the Decommissioning of the Fukushima Daiichi NPS>
To continuously and quickly reduce the radioactive risks caused by the accident that do notexist in the usual NPS

The Fukushima Daiichi NPS has the necessary safety measures in place that are required by
the NRA in “the matters for which measures should be taken” and is maintained in a state with acertain level of stability.
However, the Fukushima Daiichi NPS is considered to be at great risk because fuel debris andspent fuels still remain in reactor buildings damaged by the accident, some of the status of the NPS
are not sufficiently grasped, and the site has radioactive contaminated water and enormous
amounts of extraordinary radioactive wastes. If left unaddressed, these risks may increase due tothe 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 “toreduce continually and quickly the risks associated with the radioactive materials that resulted fromthe accident and do not exist in normal nuclear power plants” by taking measures specificallydesigned for risk reduction. In general, actions that are effective for reducing the risks at nuclear
facilities that have suffered an accident are: (i) improving the containment function of the damaged
facilities, (ii) changing the properties and form of the contained radioactive material to more stableones, and (iii) strengthening monitoring and control over the equipment to better prevent or mitigate
the occurrence or propagation of abnormalities. To enable these actions comprehensively, (iv)removing radioactive materials from the damaged facilities or insufficient containment status andplacing them in sound storage is also effective.
These diverse risk reduction measures have been continued based on careful preparations
aimed at preventing accidents and exposure of workers to radioactivity (refer to Attachment2).

2.2 Progress status of the decommissioning of the Fukushima Daiichi NPS
Contaminated water management
Contaminated water has been addressed in accordance with the three principles (“Removing”
contamination sources, “Redirecting” fresh water from contamination sources, and “Retaining”
contaminated water from leakage) (Fig. 3).
As a measure of “Removing” contamination sources, radioactive materials are removed fromcontaminated water with multi-nuclide removal equipment (advanced liquid processing system:

12

ALPS). As a measure of “Redirecting” fresh water from contaminated sources, freezing operationof the land-side impermeable walls were proceeded step by step with the approval of NRA, andfreezing in all of the areas started by August 2017. In March 2018, the temperature declined to 0℃
or below in almost all areas of the wall, and then water level difference of 4 - 5m was formed
between inside and outside of the wall in the mountain side. These indicated the installation of thewall completed except for deep areas. Consequently, the completion of the land-side impermeable
wall and reinforcement of the sub-drain system led to the suppression of the groundwater inflow
into the buildings, and the sharp decrease in the amount of the water transferred from thegroundwater-drains to the buildings. Because of this kind of preventive and multi-layered measures,
total generation amount of contaminated water has been decreased to approximately 220m3 per
day in FY2017 from approximately 400m3 per day in FY2016. In addition, the effect of the land-side
impermeable wall is apparent in the decrease the pumping amount both in the sub-drains andgroundwater drains. As a result of “Retaining” the contaminated water from leakage, theconcentration of radioactive materials in the surrounding sea area is constantly low.
Principle 1.
REMOVING contamination sources

Water treatment facility special for
Subdrain & Groundwater drain①Treatment of contaminated water through theMulti-nuclide Removal System(ALPS)②Removal of contaminated water inside Seawater
Piping Trench

Temporary storage tanks

Contaminated
Water Treatment

SubdrainLand-side
Impermeable Wall
(Frozen Soil wall)

Principle 2.
REDIRECTING fresh water from contamination sources③Groundwater pump-up by Groundwater Bypass
System④Groundwater pump-up by Subdrain⑤Land-side Impermeable Wall (Frozen soil wall)⑥Wide-area coating (Waterproofing of ground surface)

Groundwater
BypassWater collecting tanks

Seawater Piping
TrenchPrinciple 3.
RETAINING contaminated water from leakage

Groundwater Drain⑦Foundation improvement using water glass⑧Sea-side Impermeable Wall
⑨Groundwater pump-up by Groundwater Drain⑩Installation of welded tanks

Sea-side
Impermeable Wall

(Courtesy of TEPCO)

Fig. 3 Three principles and measures
In treating stagnant water in the buildings, the amount of stored water has been decreasing
steadily by lowering the water level within turbine buildings with the goal of completing the treatment
in 2020. The level of stagnant water in the Unit 1 Turbine Building was reduced to the lowest floor
level in March 2017. In the Unit 2 - 4 turbine building, the level of the stagnant water lowered to thelowest floor’s intermediate ceiling, which was exposed in December 2017. For the condensers in
Unit 1 – 3, where the highly radioactive stagnant water was stored at the time of accident, draining
completed by December 2017 and so forth, thereby radioactive materials contained in the stagnant
water sharply decreased (approximately a half compared to the end of FY2014). A new circulation
system for the contaminated water management started the operation in February 2018 to reducethe concentration of the radioactive materials in the stagnant water. Decreasing both the amountof the stagnant water and the concentration of the radioactive materials accelerate the reduction of13

inventory of the radioactive materials in the stagnant water.
Moreover, the water treated by the multi-nuclide removal equipment is stably stored andmanaged in the welded type tanks in order. Comprehensive discussion including social standpoints
such as mitigation of the reputational damages have been made in the government’s subcommittee
on the water treated by the multi-nuclide removal equipment.
Progress of removing fuel from the spent fuel pool
In Unit 1, to remove fuel from spent fuel pools at the operating floor, rubble removal in thenorth of the operating floor was started in January 2018.
In Unit 2, as a part of preparation of fuel removal from spent fuel pool, opening to provide
access to the operating floor was made, then investigation of the dose and the contamination statein the operating floor was started in using remote robots and heavy machine, etc. from July 2018.
In Unit 3, installation of a cover for fuel removal was completed in the middle of February 2018.
Once started the trial operation for fuel handling equipment in March 2018, several troubles haveoccurred. TEPCO intends to examine and review the start of fuel removal in the spent fuel poolwhich was scheduled in November 2018.
Fuel debris retrieval
In Unit 2, internal investigation of Primary Containment Vessel (“PCV”) was conducted in
January 2018, continued from January and February 2017. It was confirmed that deposits, which
seemed to be fuel debris, are accumulated in the bottom of pedestal, according to the result ofanalysis of the images obtained.
In Unit 3, based on results of investigation in July 2017 using a remotely operated underwater
survey vehicle (“ROV”), three D shape reconstruction was conducted, to grasp whole image within
the pedestal. It was found out that swivel platform came off the track, and partly buried in the deposit,
showing relative position of structures.
Waste management
In February 2018, operation of solid waste storage building No. 9, which has about 40% ofstorage capacity of all the existing solid waste storages (building No.1 – 8), was started. In thestorage facility with shielding function, it became possible to store the highly radioactive rubbles
arisen from removal in the operating floor of Unit 1 and dismantling of upper part of reactor buildingof Unit 2. TEPCO formulated Storage Management Plan to store and manage these solid waste
properly, it was revised in June 2018 based on the results of prediction of generation of solid waste.
(Attachment 9 ） In addition, sampling and analysis has been progressing for the purpose ofcharacterization.
Other specific measures
The surrounding road of Unit 1 to 4 buildings and part of east side of turbine buildings were14

determined as “ordinary clothing area”, where the workers are allowed to enter and work in general
workwear or in on-site safety workwear with disposable dust-proof mask. As a result, ordinary
clothing area expanded to approximately 96% of the Fukushima Daiichi NPS site.

2.3 Basic concept of reducing risk of radioactive materials
2.3.1

Quantitative grasping of risk

The term “risk” may have various meanings depending on the field or scene of use. In general,in the context of appropriate risk management, “risk” can be understood as an expectation valueof the negative impact of an event. In other words, the magnitude of a risk (risk level) posed by asubject (risk source) can be expressed as the product of the level of impact and the likelihood of
occurrence of the event.
In the Strategic Plan, the method based on the Safety and Environmental Detriment score(SED), which accounts for the public impact developed by Nuclear Decommissioning Authority(NDA) is used to study the reduction of the risk level of radioactive materials.
Here the risk level is defined as follows:
Risk Level = Impact × Likelihood= Hazard Potential × Safety Management
Hazard Potential here is an index of impact of the event, namely, the impact of internal
exposure in the event of human intake of the radioactive material contained in the risk source. It
can be expressed as the product of Inventory, which is the radioactivity of the risk source (toxicityof the radioactive material), and factors that depend on the form of the risk source and the timeallowable until the manifestation of the risk. Safety Management is an index of the likelihood thatan event will occur. It is determined by factors that depend on the integrity and other aspects ofthe facility and on the packaging and monitoring status of the risk source (see Attachment 3)
2.3.2 Identification of Risk Source and Risk assessment
For the development of risk reduction measures, the major risk sources of the Fukushima
Daiichi NPS are indicated in Table 1. The sum of these risk sources represents the overall risks for
the Fukushima Daiichi NPS, and the overall risk levels are shown in Fig. 4. Continuous efforts havebeen made to reduce these risks through various measures, including those described in Section
2.2. Moreover, the risk levels of each risk source expressed using Hazard Potential and Safety
Management are shown in Fig. 5.
In the Roadmap, management of these risk sources is classified into the three broad
categories: 1. Highly contaminated water and fuels in the spent fuel pools, etc., relatively high risks
given high priority, 2. Fuel debris, etc., immediate risks unlikely, risks may grow in case of dealing

15

with haste, 3. Solid waste, etc. increased risk unlikely in the future, but appropriate
decommissioning efforts necessary. Their priorities are set, and appropriate measures are
implemented. Risk reduction strategies for each of these sources will be described in Section 3.
Table 1 Major risk sources in the current risk map
Fuel debris
Spent fuel

Fuel debris in the RPVs/PCVs in Units 1-3
Fuel in SFPs

Fuel assemblies stored in the spent fuel pools (SFPs) in
Units 1-3

Fuel in the common

Fuel assemblies stored in the common pool

pool
Fuel in dry casks

Fuel assemblies stored in dry casks

Contaminated Stagnant waterContaminated water accumulated in the reactor

water etc.

buildings and turbine buildings in Units 1-4, mainin the buildings

process building, and high temperature incinerator
building.
Stored water

Strontium-treated water, residual concentrated salt

in flanged tanks

waterStored water

Strontium-treated water, treated water by ALPSin welded tanks
Secondary

Waste adsorption

Storing adsorbent used in the cesium adsorption

waste fromcolumns

apparatus, the second cesium adsorption apparatus,

water

high-performance ALPS, mobile-type strontium removal

treatment

equipment, second mobile-type strontium removal

systems

equipment and mobile-type treatment equipment, etc.
HIC slurry

Slurry produced during the treatment by the multinuclide removal equipment stored in high integrity
containers (HIC)

Waste sludge

Precipitation from the decontamination instruments

Concentrated liquid

Concentrated liquid waste generated by evaporative

waste, etc.

concentration of concentrated salt water and carbonate
slurry collected from concentrated liquid waste

RubbleSolid waste storage

Rubble with high-dose (30mSv/h and above) stored in

etc.

facility

the solid waste storage facility

Soil covered temporary

Rubble stored in Soil covered temporary storage facility,

storage, etc.

Temporary storage tent and Outdoor container storage(1～30mSv/h), and Fallen tree stored in Temporarystorage pool

Outdoor storage, etc.

Rubble stored in Outdoor seat covered storage (0.1～
1mSv/h), Rubble stored in Outdoor storage (Under
0.1mSv/h), and Fallen tree stored in Outdoor storage

Contaminated structures, etc. in the

Structures, pipes, components, and other items insidebuildings

the reactor buildings, PCVs or RPVs that arecontaminated with radioactive materials dispersed dueto the accident; and activated materials from operation
before the accident

16

Risk level(logarithmic scale)
0

1

2
3
4
5
Year after accident

6

Risk level (linear scale)

Risk level (linear scale)

Fuel removal from
Unit 4 SFP started

7Mid-and-Long-Term
Roadmap Phase 1

Mid-and-Long-Term
Roadmap Phase 2

Risk level in the rightgraph is multiplied by 10

0.5

Fuel debris

1
1.5
2Year after accident
Fuels in SFPs

2.5

3

3.5

Contaminated water

4

Fuel debris

4.5
5
5.5
Year after accident
Fuels in SFPs

6

6.5

7

Contaminated water※1

The risk level was high due to fuel debris right after the accident, however, it became largely lower
because Hazard Potential has been largely decreased by attenuation of radioactive materials inside fueldebris over one year after the accident.
※2 The risk level increased from four to four and a half years after the accident because Safety Managementof fuel debris became high by dismantling the building cover of Unit 1 in order to prepare spent fuel removalfrom the pool such as removal of rubbles in the upper area of operating floor. From the viewpoint of risk
reduction of the entire site, it was one of the measures needed for spent fuel removal from the pool. When
dismantled the cover, measures against dispersion of radioactive material was conducted.

Fig. 4 Reduction of risks contained in the Fukushima Daiichi NPS

17

Hazard Potential (logarithmic scale)

Region of sufficiently
stable management(March, 2018)

Fuels incommonpool

Fuel debris

ContaminatedUnit 1structures in bldgs. Unit 3
Unit 2
Fuels in
HIC
Unit 3
SFPs
Waste
slurry Uint 2
Unit 1absorption
columns
Waste
Stagnant water
Concentrated sludgein bldgs.
liquid waste, etc.

Fuels indry casks

Stored water inwelded tanks

Rubble instoragefacility

Stored water inflanged tanks

Soil covered
temporarystorage, etc.
Outdoorstorage, etc.

Safety Management (logarithmic scale)Fig. 5 Examples of Risk Levels of Major Risk Sources at the Fukushima Daiichi NPS
2.3.3

Risk Reduction Strategy
Interim Targets of the Risk Reduction Strategy

The risk levels of the risk sources shown in Fig. 5 are higher toward the upper right corner.
There are two risk reduction strategies which are reduction in Hazard Potential or reduction in
Safety Management.
The examples of the former strategy include decrease in inventory and decay heat associatedwith radioactive decay, and changing the form of liquid and gas into the form that is hard to move.
To process contaminated water to change into the secondary waste is an example of form change.
The examples of the latter one include transfer of fuel in the pools to the common pool, andplacement of rubbles stored outside to the storage. Of the various risk reduction measures, thereduction in Safety Management is considered generally to be easily achieved. Therefore, thedecommissioning of the Fukushima Daiichi NPS, which is implemented under the basic policy of“reducing continually and quickly the risks associated with the radioactive materials that resulted
from the accident and do not exist in normal nuclear power plants” (Section 2.1), should first focus
on steadily managing risk sources by keeping them in higher-integrity facilities to lower their Safety
Management levels. The interim goal of risk reduction strategies is to bring the risk levels into the“Region of sufficiently stable management” (pale blue area) indicated in Fig. 5.

18

Basic approach to risk reduction
The decommissioning of the Fukushima Daiichi NPS is a project with inherent considerable
uncertainty. To date, internal status of PCV in the Unit 1 to 3 is presumed to some extent by
simulation of the accident progression using various measurements, estimation of location of fueldebris by measurement of muon-base, projection of investigation equipment into PCV, dose
measurement in the buildings and others. However, since the radioactive environment inside thereactors is still too severe for workers to easily access, the properties of some radioactive materials
and the degree of damage to the field equipment and structures left there remain unknown, creating
uncertainty.
While it is desirable to eliminate uncertainty by grasping all the above unknowns that are
difficult to know, many resources, especially a considerably long time are required. To realize
prompt risk reduction, it is necessary to make integrated decisions taking flexible and prompt
approach, based on the directions determined with previously obtained experience and knowledge
and with experiment- and analysis-based simulation, placing safety at the top priority, even thougha certain extent of uncertainty remains. At this time, it is not necessary to take the same steps
among the different Units, and it is important to flexibly address the issues with experiences,reflecting the internal information, technical feasibility and other information obtained in advance ina certain Unit on the subsequent works and the works in other Units. As the viewpoint to makeintegrated decisions like this, which are not easy to do, NDF summarizes the following five guiding
principles.
(Five guiding principles)

Safe

Reduction of the risks posed by radioactive materials and ensuring
work safety(Examples of considerations: containment of radioactive materials[environmental impact], workers’ radiation exposure, risk reduction
effects)ProvenHighly reliable and flexible technologies(Examples of considerations: conformity with the requirements, flexible
stance to uncertainty)

Efficient

Effective utilization of resources (e.g. human, physical, financial andspace)(Examples of considerations: restricting the generation of waste, cost,
securing the work area and space)TimelyAwareness of time axis
(Examples of considerations: early start of fuel debris retrieval, durationneeded for fuel debris retrieval)

Field-oriented

Thorough application of the “Three Actuals” (actual field, actual things

19and actual situation)(Examples of considerations: workability [environment, accessibility,
operability], serviceability [maintenance, troubleshooting])
These guiding principles are the viewpoint necessary to discuss priority of approaches andtotal optimization.
Order of priority
It is important to be aware of position of each project and their mutual relationship in theirrespective fields in managing progress of overall project based on the five guiding principles. In
other words, although addressing the highest risks first may seem effective from a near-term
perspective, such an approach may not always be optimal when viewed from a long-term
perspective that considers various factors, such as the feasibility of the relevant technologies andthe time required for preparation.
For example, although the vent stack for Units 1/2 located near Unit 2 is contaminated withradioactive materials, mainly cesium, that were released by the accident and adhered to its innersurface, its risk associated with radioactive materials is rated as lower than that of the fuels in the
spent fuel pool. However, considering fractures and deformation were found at the junctions of the
diagonal bracing of the steel tower that supports the vent stack, it is planned to dismantle the upper
part of the vent stack with remotely controlled equipment as an environmental improvement aheadof fuel removal from the pool.
In the decommissioning of the Fukushima Daiichi NPS that aspires to a continuous and prompt
risk reducing, the traditional approach of proceeding with project tasks by resolving imminent issues
one by one would not be effective. Instead, it is important to aim at choosing the best option among
the various possible options, with the integrated viewpoint of looking at the entire site on a longterm basis and considering time axis. This idea of pursuing risk reduction optimized for the entire
project is discussed in Section 3.6. Taking into account such point of view, TEPCO and NDF havejust introduced project management mechanism, as described in Section 4.3.
Addressing temporary increase of risk level associated with the decommissioning
operations
Implementation of the decommissioning work aims at prompt risk reduction from the mid- andlong-term viewpoint. However, the possibility that conducting the work may temporarily change therisk levels and increase the radiation exposure of workers must be carefully considered. This isbecause carrying out the decommissioning work involves acting on the current situation of the NPS,
which is kept in a state with a certain level of stability despite some risks. Such risks may materialize,
depending on the way actions are taken. For example, accessing the inside of the reactor to retrievefuel debris will affect the current containment status, and special operations and maintenanceperformed for the retrieval will increase the exposure of workers involved in these activities.

20

This possibility of a temporary increase in the risk level and a rise in workers’ exposure arisingfrom the decommissioning work must be addressed by ensuring measures to prevent and restrict
them. It is imperative to limit the increase in the risk level during the decommissioning work within
the permissible range by thorough preparations for the work. In particular, the radiation safety (i.e.
restricting radiation exposure) of workers should be ensured in accordance with the concept of
ALARA2.
Note that the basic stance of promptly implementing the decommissioning must be firmly
maintained because if the work is delayed excessively, existing large risks will remain over the longterm and their risk levels may gradually rise as the buildings and facilities deteriorate over time. For
this reason, the selection of the method of work, the design and production of equipment and safetysystems, and the development of work plans for the decommissioning work should many
constraints such as consider time, cost and worker’s exposure needed for associated preparations
and work, while giving priority to limiting the risks involved in the decommissioning work. Based onthese considerations, cautious and comprehensive decision-making are required for the early
implementation of the decommissioning (see Attachment 4).
It must also be strongly emphasized that the decommissioning of the Fukushima Daiichi NPS
as a risk reduction project should be promoted, not by the understanding among the limited working
parties, but by gaining the broad understanding and support by wide range of public including local
residents. It is crucial to foster public understanding of the overall risk reduction efforts, including
measures to prevent a temporary increase in the risk associated with various decommissioning
work and thus gain understanding of the decommissioning project. It is particularly important to
inform local residents, in an easily understandable manner, of issues such as what kind of risk
reduction strategy has been established for the decommissioning work, how the safety of thedecommissioning work is ensured, and how the overall risks at the NPS have been continuously
reduced through the decommissioning work. For this purpose, it is important to implement an easyto-understand mechanism of monitoring of risks for local residents.

2

ALALA is the principle of optimization for radiological protection recommended by International Commission on
Radiological Protection (ICRP), and is the abbreviation of “as low as reasonably achievable”. It is advocatedthat individual dose, number of the exposed people and likelihood of exposure must be kept as low asreasonable achievable in consideration of economic and social factors. (The 1990 Recommendation by ICRP,
Japan Radioisotope Association (issued in 1991),

21

3. Technological strategies toward decommissioning of the
Fukushima Daiichi NPS
3.1 Fuel debris retrieval work
3.1.1

Sectoral target

The immediate objectives of fuel debris retrieval are as follows (For the objective of fuel debris
retrieval refer to Attachment 5).
(1) Retrieve fuel debris safely after thorough preparations including safety measures, and bring itto the state of stable storage that is fully managed.
(2) Toward determination of fuel debris retrieval method for the first implementing unit in FY 2019,and start of fuel debris retrieval work for the first implementing unit within 2021, necessary
approaches will be taken, according to policy on fuel debris retrieval.

3.1.2

Sectoral strategies
Approach to risk reduction in fuel debris retrieval work

It is hard to consider that the fuel debris immediately appears as a risk, but it is a risk sourcethat risk may grow in case of dealing with haste. (refer to section 2.3). Currently, it is maintained ina state with a certain level of stability. However, on a long-term, the possibility of changes in theform and physical property due to aging degradation is considered. Namely, two types of risk exist
in fuel debris. One is a risk from medium-term perspective that adverse effects on the environment
can occur due to the development of criticality and cooling problems though its possibility is low as
far as it is properly managed in a state with a certain level of stability. The other is a risk from along-term perspective that environmental pollution can occur resulting from any leaks of included
nuclear fuel material due to deterioration of the buildings in future. Therefore, as indicated in (1) of
Sectoral goal, fuel debris should be retrieved safely after thorough preparations such as safetymeasures and bring it to the state of stable storage that is fully managed as soon as possible toachieve a low risk level that is largely acceptable to society. Technical investigations are on torealize this.
Until now, in the Strategic Plan, securing of safety during fuel debris retrieval work andtechnical requirements that are defined including feasibility of fuel debris retrieval method andstable storage of fuel debris have been considered while organizing the issues logically. In the
future, consideration should be made according to these technical requirements to implement fueldebris retrieval work safely when expanding the scale with a step by step approach in accordancewith policy on fuel debris retrieval as described in the next chapter.
Furthermore, based on the current risk assessment for each unit, there would be a difference22

in terms of Safety Management across the units. Unit 1 has no upper reactor building and Unit 3
has a cover for fuel removal instead of the upper reactor building. On the other hand, the reactor
building of Unit 2 is in a sound state, and the damage level of Reactor Pressure Vessel (“RPV”) is
considered small because most of fuel debris are presumed to stay inside RPV. Regarding theshape of fuel debris influencing Hazard Potential, there is a possibility of various states such as astate close to fine particles to a solid state, but the specification of the shape has not been done
currently. The calculation in Fig. 5 has been estimated from the findings obtained so far. Particularly
with respect to Unit 2, many of fuel debris are presumed to stay in the RPV, and Hazard Potential
becomes relatively low because it is considered to be a stable form due to the low ratio of Molten
Core Concrete Interaction (“MCCI”) products compared to Units 1 and 3.
Policy on fuel debris retrieval
The Strategic Plan 2015 and 2016 examined the fuel debris retrieval method option by
combining the PCV water level (Submersion method, Partial submersion method) and the fueldebris access direction (Top access, Side access and bottom access). After that the threeconstruction methods were selected (1) The submersion - Top access method, (2) Partial
submersion - Top access method, (3) Partial submersion - Side access method) and the study onthese methods have been continued.
In the Strategic Plan 2017, with respect to these three fuel debris retrieval methods, in addition
to nine technical requirements that need to be satisfied for safe retrieval of fuel debris (1.
Containment functions, 2. Cooling functions, 3. Criticality control, 4. Structural integrity, 5.
Reduction of radiation exposure, 6. Work safety, 7. Access routes, 8. Development of devices andequipment, 9. System facility and area construction), feasibility evaluation was conducted for eachof the three technical requirements (1. Collecting, transferring and storing, 2. Treatment ofradioactive waste during fuel debris retrieval work, 3. Safeguard measures) for safe and stablestorage of fuel debris. Then, the comprehensive evaluation was implemented based on five guiding
principles, and made a strategic proposal (toward the decision and efforts after the decision) for
determining the fuel debris retrieval policy. Based on the contents of this strategic proposal in the
Roadmap revised in September 2017, the policy on fuel debris retrieval has been determined asfollows.
＜Policy on fuel debris retrieval＞① Step-by-step approach
We will adopt a step-by-step approach wherein we will set the method of fuel debris
retrieval to be started first in order to reduce risks at an early stage and will adjust our
direction flexibly based on information that comes out as retrieval proceeds.
Fuel debris retrieval, the investigation inside the primary containment vessel, and theinvestigation inside the reactor pressure vessel will be performed in a coordinated,
integrated manner. The fuel debris retrieval starts from a small-scale task and the scale ofretrieval will be stepped up as we review our operations flexibly based on new findings
obtained from the nature of the fuel debris and working experience.

② Optimization of entire decommissioning work23

We will examine fuel debris retrieval as a comprehensive plan aimed at total optimization,from preparation to cleanup through retrieval work, discharge, processing and storage,
including coordination with other construction work at the site.

③ Combination of multiple methods
We will combine optimum retrieval methods suitable for the part of each unit where fueldebris is expected to be present, instead of making an assumption that all the fuel debris is
to be taken out using a single method.
At present we will examine how to combine methods from an accessibility standpoint,
assuming that access is made to the bottom of the primary containment vessel from theside and that access is made to the inside of the reactor pressure vessel from the upper
part of the vessel.

④ Approach focused on partial submersion method
Given the technical difficulty of stopping leaks at the upper of the primary containment
vessel and expected radiation doses during work, the full submersion method is technicallydifficult at present, so we will base our efforts on the partial submersion method that is morefeasible.
However, given the advantages of the total submersion method, such as being effective inproviding shielding against radiation, we might consider adopting the full submersion
method in the future depending on the progress of R&D.

⑤ Prioritizing fuel debris retrieval by access to the bottom of the primary containment
vessel from the side
According to an analysis, fuel debris is expected to be present in both the bottom of the
primary containment vessel and the inside of the reactor pressure vessel of each unit,
although their distribution varies among the units. In view of rapidly mitigating risks fromfuel debris while minimizing any increase in risks that might be caused by retrieval, we willprioritize retrieval of fuel debris in the bottom of a primary containment vessel by accessfrom the side by taking the following into account:○The bottom of the primary containment vessel is most accessible and a certain amount ofknowledge about it has already been accumulated through the investigation inside theprimary containment vessel;○There is a possibility that fuel debris retrieval could be started earlier;○Fuel debris retrieval could be performed at the same time as spent fuel removal

How to proceed discussion toward determination of fuel debris retrieval methods
for the first implementation unit
In the Roadmap, it is stated that “regarding the method of fuel debris retrieval for the first
implementation unit to begin the operation, the fuel debris retrieval for the first implementing unit
will start within 2021 by determining the method of containing, transfer and storage (by FY2019)
after due consideration of the results of the preliminary engineering work and R&D”. In choosing
this first implementation unit, it is premised that for each of the units 1 to 3, the medium-to-longterm engineering schedule for implementing fuel debris retrieval work are already established. It isonly reasonable that the first implementation unit and its retrieval method would be determined
after this above is incorporated in the overall plan to ensure consistency. Figure 6 shows theinvestigating flow (draft plan) to determine the fuel debris retrieval method for the first implementing
unit, based on the R&D achievements in the decommissioning and contaminated water
management project, results of internal investigation of PCVs, state of the work environment
adjustment, and the use plan of the entire site including the plan for fuel removal from the pool and24

contaminated water management etc. 1: It is necessary to prepare a scenario (operation processplan) during the preliminary engineering work3 implemented by TEPCO (see next section), basedon the consideration on the concept of fuel debris retrieval system for each unit and on theevaluation of its applicability on site. 2: After this, the first implementing unit and the retrieval method
will be determined, by considering multiple whole scenarios combining the scenarios of each unit
and the plan for surrounding areas and by identifying the whole scenario which is considered to becomprehensively the most reasonable from the viewpoint of the total optimization of the entire site
as described in Section 3.6 such as time, safety, consistency in the project as a whole, etc.
At this time, from the viewpoint of evaluation of the applicability of operations to the actual site,
ensuring safety during fuel debris retrieval work (containment functions, cooling functions,structural integrity, criticality control, reduction of radiation exposure during work, etc.), feasibility offuel debris retrieval method (construction of access route, development of devices and equipment,
system facility and area construction) and stable storage of fuel debris (examination of collecting,transferring and storing, and safeguard measures) are compatible with the technical requirements
described in Section 3.1.3.
In particular, to select the first implementing unit, consideration should be made for theparticularity of unprecedented undertaking of fuel debris retrieval work in the environment withmany uncertainties, and the benefit of getting experiences and information of handling of fuel debris
early, also taking into account the viewpoints of reliability of internal information, work environment
such as involvement of necessary preparatory works, the results of the risk assessment of eachcurrent unit shown in Fig. 6 and Section 3.1.2.1, and the risk level at the time of assumed fueldebris retrieval work, etc.
Internal
Investigation of
PCVs
Incorporate

Needs

Improvement of workplaceenvironment
・ Reduction of radiation dose rate
Incorporate

Study onIncorporate concept of

retrieval

R&D
Needs

⁃Plan forincremental scale
expansion of
retrieval⁃Transferring andstoring method⁃Concept on how tosecure safety, etc.

Needs

Evaluation ofApplicability
at site⁃Safety andreliability⁃Schedule⁃Economical
efficiency
etc.

Establishmentof Fuel debris
retrieval
scenario foreach unit

Study on fuel retrieval
method considering
optimization of site layout
and retrieval scenario of
each unit

Determination of fuel debris
retrieval method to be
applied to the first
implementation unit

Apply the obtained information and study
results appropriately

Site-wide planning• Layout planning• Contaminated water management• Fuel removal from the spent fuel pool, etc.

Fig. 6 Investigating flow toward determination of fuel debris retrieval method (draft plan)3

Preliminary implementation of engineering related study work to determine feasibility of construction, prior to thebasic design to be performed at the time of construction work. Considering on-site applicability of technologies
including technologies, devices and equipment including maintenance performance, placement, flow lines etc.,
fully in consideration with the site situation, will minimize any rework that maybe required post basic design.
Therefore, through the results of preliminary engineering work, the construction method will be reviewed asnecessary25Proceeding method of preliminary engineering and creation of concrete plan ofeach task
In the preliminary engineering work conducted by TEPCO, by confirming the applicability ofoperations on site in light of the situation of each unit about operations and system concept foreach step of fuel debris retrieval proceeded by a step-by-step approach, and it will be specified asa scenario for fuel debris retrieval work for each unit. In particular, details of preliminary engineering
work should be fully planned and investigated to obtain the necessary information, aiming at
milestones, which is the determination of fuel debris retrieval method of the first implementing unit
within FY 2019. From this point of view, considerations meeting the following items are expected
to be conducted in the preliminary engineering work: The whole related operations ranging from internal investigation in the preliminary stage offuel debris retrieval work and preparations to the improvement in the surrounding
environment to fuel debris storage should be included in the scenario.
 In each step, the information that should be gained in advance should be organized in order
to secure safety and engineering reliability on the equipment for retrieval.
 Clarification of prerequisites as well as evaluation on its uncertainties and forecast
concerning a development of the scenario should be conducted Sufficient safety evaluation should be implemented for the major troubles and others
assumed at present.
As a result of these considerations, the following items are expected to be obtained, as theachievement of preliminary engineering work: Image of operation process to fuel debris retrieval for each unit and identification oftechnical issues to solve
 Engineering schedule incorporating the period of solving the technical issues (scenario
for each unit)
As a result, it will be possible to create the overall scenario, combining the obtained scenario
for each unit and fuel debris retrieval from the pools and contaminated water management. And
the candidate of the first implementing unit will be selected according to this overall scenario.
Continued internal investigation and accelerated/prioritized R&D
It is necessary to draw up a roadmap for solving the technical issues extracted so far andidentified in the process of implementing preliminary engineering work, through further internal
investigation and accelerated/prioritized R&D.
As the PCV internal investigation conducted so far has obtained various information, further
detailed internal investigation is planned to collect data such as the distribution of deposits and fueldebris inside the PCV. Most of these kinds of internal investigations have been conducted till now
as field validation of instruments which was developed in decommissioning and contaminated26

water management project, with only partial of environmental conditions inside PCV being unclear.
In the future internal investigation, it should be thoroughly identified what kind of data is necessary
as part of assembling the whole project in preparing for retrieving fuel debris, which is progressing
with a step by step approach. Then to implement the scenario for the goal of each phase is needed.
In addition, according to the fuel debris retrieval policy based on the partial submersion method,by advancing the side access to the bottom of the PCV, R&D tasks that newly needs have become
clear are to be proceeded to extract through preliminary engineering work, as well as accelerating
and prioritizing R&D including establishing a containment function (gas phase) on the assumptionof the existence of α-nuclides and technology to manage the water level in the PCV. It is also
important to take a step-by-step approach to flexibly adjust the direction including the method of
retrieving fuel debris, depending on the outcome of R&D and the situation on site to which thescenario is applied.
3.1.3

Technical issues for promoting sectoral strategies and future plans

Fuel debris retrieval work should proceed with a step by step approach according to the fueldebris retrieval policy. After presenting the step-by-step concept in 3.1.3.1, the main issues aredescribed in Section 3.1.3.2 to 3.1.3.5 including the approach to comprehensively understandinternal reactor condition, the technical requirements related to safe assurance shown in theStrategic Plan, the technical requirements concerning the fuel debris retrieval method, and thetechnical requirements on the stable storage of fuel debris, together with further considerations
required as the scale grows.
Concept of step-by-step approach
In the fuel debris retrieval policy, internal investigation and fuel debris retrieval work will be
performed in a coordinated, integrated manner. Starting with small-scale fuel debris retrieval, thescale of retrieval will be gradually expanded, reviewing the work flexibly, based on the nature offuel debris and new findings obtained from work experience etc.,.
Details of work elements pertaining to fuel debris retrieval for each unit are under consideration(see below and Table 2). Note that these do not necessarily apply to all of Units 1 to 3 across theboard.
Internal investigation (Investigation of internal situation and understanding of properties offuel debris etc.)
For the condition inside of the PCV and damages of the internal structure, investigation andobservation are to be made provided that the internal environment has not changed. This will
provide information for checking the distribution and accessibility of fuel debris at the bottom of the
PCV and information for judging the safety of fuel debris retrieval work, which are used for
examining the fuel debris retrieval method.
Furthermore, the nature of fuel debris inside of the PCV, including the shape, existing state,27

composition, and mechanical/chemical properties, etc. are identified through sampling and analysis.
As fuel debris sampling inevitably involves moving fuel debris, it aims to improve the accuracyof determining the feasibility of the fuel debris retrieval method and to improve the reliability of theprotective measures for ensuring safety, by collecting important data that assists evaluation of
ensuring safety in the subsequent stage such as information pertaining to the transportation for
temporarily storage of retrieved fuel debris and stable storage, information pertaining to thetransition state of fuel debris to the gas/liquid phase and potential criticality, etc.
Fuel debris retrieval
In the initial stage of fuel debris retrieval, fuel debris will be taken out to the extent that thecondition of the site is not significantly changed. In order to improve the efficiency of workthroughout the fuel debris retrieval period, various kinds of information to identify the work andequipment used for large-scale retrieval scheduled thereafter should be collected and verified,including to confirm the effectiveness and evaluate the work efficiency of the fuel debris retrieval
work and equipment, to evaluate the effect on safety when extending the scale of fuel debris
retrieval and to check in advance the conditions for collecting, transferring and storing.
Based on the data obtained from the operation until the small-scale fuel debris retrieval,installing equipment is to be made that can handle the target amount of fuel debris taken out per
day and more efficient fuel debris retrieval will be performed after studying the fuel debris retrieval
equipment and the system for ensuring safety.
While entering this large-scale retrieval stage, checking the conditions of the site regularly, thefirst step should start with a small amount of fuel debris retrieval work. It is necessary to pay
attention to the step-by-step approach as the scale gradually expands.

28

Objectives

Contents

29

Sampling (ex.)• Transport the collected fuel debris • Removal of obstacles in PCV shall be limited tothe minimum necessity within accessible area.
to the analysis facility
Continue the small-scale retrieval work using the• Carefully collect samples while
methods such as suction, gripping, scooping,monitoring not to scatter the dust
small scale cutting, core boring and so on.
much, Using such methods assuction, gripping, scooping, cutting • Based on the results of these work, evaluate theeffectiveness of the equipment, the influence on( or core boring)safety assurance against expansion of the retrieval• Basically removal of large
scale, and the work efficiency etc.
obstacles in PCV is not planned.
• Collected debris is to be stored in a temporalstorage facility in the building and carry out
preliminary confirmation concerning the upgrade
to large-scale retrieval, transfer and storage• As a preliminary stage for large-scale retrieval,retrieve fuel debris using a small-scale equipment
without changing the current states to a great
extent, obtain further various information for
retrieving fuel debris and verify the plan of largescale retrieval method.

Small-scale retrieval (ex.)

Small-scale retrieval image

Turning
axis
Tilt
axis
Telescopic

Pan
axis

RPV

PCV

Arm expansion situation

Inside of PCV

CRD rail

Pedestal
opening part

Grip and suction tool
Pedestal
PCV

Access equipment

outside of PCV

CRD rail

PedestalPedestal
side

Tilt
axis

Access device(Example ofrobot arm)
Move backand forth

PCV

Transport
container

Airtight cellfor retrieval

Pedestal

Arm device

Large-scale retrieval image of side-access method (ex.)

PCV

Rail device

RPV• The portion around the X-6 penetration is to
be expanded so as to set up an arm with rails
etc. for large-scale work.

• In addition to the retrieval work listed left, sludge-likedebris and massive debris (collected by large-scalecutting) etc are to be retrieved.
• Retrieval of debris attached to the structure (cutting
the structure)• Removal of interference substances in PCV• Storage facilities of collected fuel debris are to beprepared for long term storage.

Large-scale retrieval (side‐access method)(ex.)• Based on information obtained through small-scale
retrieval work, review the fuel debris retrieval device,safety related systems, etc. to improve reliability,
perform more efficient debris removal.
• Also, depending on the part where fuel debris is present,
select an appropriate access and retrieval method.
• Retrieved fuel debris shall be stored in an appropriate
storage facility

Fuel debris retrieval• Use an arm type access device of a size small enough not to expand the X-6 penetration .
• Equipment attached to the tip is selected appropriately for internal investigation, sampling and debris retrieval.

• Distribution of fuel debris at thebottom of PCV• Accessibility information for fueldebris retrieval• Information that contributes to theevaluation of ensuring safety fueldebris retrieval work,• Further possibly including• Information on Pedestal damage• Information on the possibility offalling objects from the bottom of
the RPV during work at the bottomof the PCV• To enforce the previously acquired • Collect fuel debris at various
locations in PCV and analyse their
PCV internal information andproperties (shape, existing state,analysis results, etc., obtainand mechanical/chemical
additional PCV internal information
properties).
that is the basis for studying fuel• Analyse of the effects induced bydebris retrieval methods.
the sampling work on the• By contact with fuel debris, confirm
surrounding environment andwhether any debris can be easily
facilities.
sampled .

Internal investigation (ex.)

Internal investigation etc.

Table 2 Image of fuel debris retrieval by a step-by-step approach (example of Unit 2)

Image diagram is excerpted from Material 2-1 of 60th Study group on Monitoring and assessment of specified nuclear facilities, TEPCO (May 18, 2018), Courtesy of IIRID

Retrieval Device/Equipment

Comprehensive understanding of reactor conditions by continued internal
investigation, etc.
The comprehensive analysis and evaluation results on the distribution of fuel debris of the Unit
1 to 3 and the access route to fuel debris and the situation of the surrounding structure are basedon the measured value of plant parameters which are obtained at the time of accident, severe
accident progression analysis, information on the actual situation of the site by PCV internal
investigation muon measurement, and scientific knowledge obtained by examination etc.
(1) Internal condition of Unit 1
It is presumed that most of the fuel debris is at the bottom of the PCV. It is presumed that thefuel debris at the bottom of the PCV has spread to the pedestal inside floor line and has reactedwith the concrete to form the MCCI product. There is a possibility that a part of the fuel debris has
spread to the outside of the pedestal through the worker entrance.
Regarding access routes, it is confirmed that access from the top of the steel grating to thebottom of the PCV outside of the pedestal is possible as per the PCV Internal investigation of 2015.
Regarding the situation of the surrounding structure, no large damage was detected on theouter wall surface of the pedestal on the top of the steel grating by the PCV Internal investigationof 2015.In the PCV Internal investigation of March 2017 too, no large damage was detected tothe structure.
(2) Internal condition of Unit 2
Fuel debris is presumed to be mostly present at the bottom of RPV and from muon
measurement, it is calculated that the fuel is present in RPV. Based on evaluation of core energy
it is estimated that an equivalent quantity of non-melting fuel pellet is also present. In the PCVinternal investigation of 2018, the deposit which appears to be pebbly-like/clay-like shape wasobserved in most of the bottom area of the pedestal. Since a part of the fuel assembly had fallen
to the bottom of the pedestal, the deposit observed around it is presumed to be fuel debris. However,damage to the concrete wall in the pedestal, melting of cable tray and steel pillar of the CRD4
handling machine etc. are not to be observed, so the fuel debris is considered to be containing a
lot of metal components such as internal structural material. There may have several drop paths offuel debris because there are other deposits that are accumulated higher than the surroundings.
Also, although the worker entrance of the basement floor of the pedestal has been confirmed, it
has not been able to observe outflow of fuel debris outside the pedestal.
Regarding the access route, in the PCV internal investigation of 2018, it was observed fromPCV X-6 penetration that access was possible from the worker entrance to the bottom of thepedestal via the CRD rail.
Regarding the condition of the surrounding structure, in the PCV internal investigation of 2017and 2018 it was observed that part of the steel grating of the pedestal inside the platform had fallen,
4

Control Rod Drive30

but no significant damage was found on the pedestal inner wall surface and the existing structuresin the pedestal. It was also confirmed that a part of the fuel assembly had fallen to the bottom of
the pedestal, but no large damage was observed in CRD housing support within a range of theinvestigation.
(3) Internal condition of Unit 3
It is presumed that there is more fuel debris at the bottom of the PCV of Unit 3 than that of Unit2, although some fuel debris may be left at the bottom of the RPV. In the muon measurement, it isalso shown that a range of the original core in the RPV does not have any large lump of fuel debris.
The evaluation of core energy shows the possibility of a large massive fuel solidifying after melting
once existing at the bottom of the RPV. In the PCV internal investigation of July 2017, it wasobserved that the molten material had solidified in the pedestal. In addition, since the sandy or
massive deposits were observed at the bottom of the RPV as well as structures presumed to be
the internal structures have been observed, it is presumed that the surrounding deposits may
include fuel debris. The worker entrance of the basement floor of the pedestal could not be
observed. Since the deposit was observed in the vicinity, the possibility that fuel debris has spreadoutside the pedestal cannot be ruled out. On the other hand, cooling water was poured into thedrywell (“D/W”) at the time of the accident, which may have restricted the spread of the fuel debris.
Regarding the access route, it was observed that it is possible to access the bottom of thepedestal via the worker entrance of the pedestal in the PCV internal investigation by the underwater
ROV in July 2017.
Regarding the condition of the surrounding structure, in the PCV internal investigation in July
2017, damage to the structure and fallen substances in the pedestal, missing portions anddeformation of CRD housing support were observed. Also, in a range of the investigation, the steel
grating was not observed on the platform and it was observed that the steel grating had fallen in
the lower part of the pedestal. No large damage was detected in the pedestal inner wall surface.
This information is summarized as in Fig. 7.

31

Unit 2

Unit 1

Unit 3Spent fuelSpent fuelSpent fuel

Core

Core
Water Levelin PCVapprox. 2m

Reflection from survey
in future because of
non-observation

Water Level in
PCV approx.
30 cm

Fuel Debris

RPV Lower
Head

Water Leakage(visual observation)

A small amount of fuel debris is present
A small amount of fuel debris is present
in the inside and on the outer surface of
the CRD housing
Large amount of fuel debris is present
A small amount of fuel debris is present in the inside and on the outer surface of
the CRD housing

Most of the fuel debris is present

A small amount of fuel debris is present

Fuel debris may have spread on thepedestal outside through the personalentrance

The possibility of fuel debris spreadingon the pedestal outside through thepersonal entrance is low.

Unit 1X-6 Penetration

CRD rail

Pedestal

Information of
the structurearound fueldebris

X-6 Penetration
Pedestal Inside

CRD Housing

It is possible to access to the D/Wbottom from the upper side.
State around the CRD rail connecting to
the pedestal from the X-6 penetration is
not observed.
There is no large damage on the surfaceof the pedestal outer wall above thesteel grating.

Little fuel debris remains
Fuel debris remains on the RPV lower
head party
A small amount of fuel debris is present
in the inside and on the outer surface of
the CRD housing.
Amount of fuel debris in Unit 3 is morethan that in Unit 2.
Fuel debris may have spread on thepedestal outside through the personalentrance.
CRD Housing

X-6 Penetration

Pedestal Inside
DepositCRD rail

Pedestal

X-53 Penetration

Steel grating

CRD rail

CRD Guide tube

Pedestal
InsideSteel grating
Steel Grating
Upper tie plate

Water Leakage(visual observation)

Unit 3

CRD Housing

Unconfirmed, to reflect
results of planned survey





Unit
Unit22
Pedestal

Steel grating

Information of
the accessroute for fueldebris*Fuel Debris

Little fuel debris remains(Stub-shaped fuels might exist in
peripheral region)

Fuel Debris

Personal
EntranceLittle fuel debris remains

High dose rate

Pedestal
OutsideWater Leakage(visual observation)

Pedestal Inside 

Water Levelin PCVapprox. 6 m

Personal
Entrance

Personal
Entrance

Core Region

Core









Supporting pillar
Cable tray

There is no large obstacle on the CRD
rail and around an entrance to thepedestal inside.
It is possible to access to the pedestal
inside through an entrance.
However a part of fuel assembles fall,no large damage is observed on CRD
housing support.
No large damage is observed on innerwall and structures of the pedestal.







It is possible to access to the pedestal
inside through an entrance.
Some damaged structures and the fallof the CRD housing support wereobserved in the pedestal inside.
No large damage is observed on innerwall of the pedestal.

It is thought that a route to the pedestal inside from the X-6 penetration is important for fuel debris retrieval from asmall-scale task by side access method. The contents observed by previous internal investigation are mentioned as
information to judge whether troubles are caused by falling objects on the route.
In the access route for the fuel debris retrieval in the PCV (large-scale task), an access route through an equipment
hatch is planning in the project of retrieval method and system.
Due to high dose rate around the X-6 penetration of Unit 1, it may use the same access route of the large-scale task, incase it is difficult to improve work environment. Next internal investigation of Unit 1 is scheduled to develop by
accessing from X-2 penetration.

•••(Prepared by Achievement Report 2017 and June 2018 of “Decommissioning and Contaminated Water Management Fund (Advancement ofcomprehensive grasping of inner reactor)” provided by IRID, The Institute of Applied Energy)

Fig. 7 Estimation of fuel debris distribution of Units 1 to 3, access Route and condition ofsurrounding structures
Also, as mentioned in Section 3.1.2.5, for future internal investigations, the kind of information
required as a part for assembling the whole project is thoroughly examined in the process of stepby-step approach for fuel debris retrieval, and the information that needs to be obtained should beclearly defined as the goal at each stage. In particular, in order to utilize it for creating the scenario
of fuel debris retrieval, the following investigation and study5 for each unit should be considered“Study status on RPV internal investigation, sampling and analysis” by TEPCO, Handout 3-3, 56th Secretariat
meeting/Decommissioning and Contaminated water management team meeting, July 26, 2018532

steadily. (Fig. 8)【Unit 1】Grasping distribution of structures and deposits outside the pedestal (including sampling)〔scheduled in the first half of FY2019〕【Unit 2】
Grasping mobility of deposits at the bottom of the pedestal when force is applied to deposits〔scheduled in the second half of FY2018〕



Grasping distribution of structures and deposits inside the pedestal (including sampling)〔scheduled in the second half of FY2019〕

Considering to increase sampling volume〔scheduled in FY2020〕【Unit 3】
Considering availability of internal investigation technology developed and demonstrated
by the decommissioning and contaminated water management project, in parallel to study
water level reducing in PCV

Considering necessity of further investigation in utilizing underwater ROV used in theprevious investigation

Prior to implementing further detailed internal investigation, giving due consideration will be
required for continuous ensuring safety, including maintaining containment function and prompt
recovery of containment function in the abnormal event as well as measures against exposure anddust management, in consideration of maximum use of existing penetration into internal PCV thathas been used in the previous investigations as larger equipment will be used.

33

FY2018

1st

2nd

1st

Approx.
1m 1m

Pan-tilt
camera2nd

Grasping distribution of
structures or deposits inside thepedestal (including sampling)

Thruster Approx.
0.3m
0.3m

Lighting

Investigation Unit
(Ultrasonic range finder, etc.)

Investigationequipment by boat-type access with
潜水機能付ボート型アクセス・調査装置
submerging
function (Current image)（現時点のイメージ図）

Investigationequipment by arm-type access
アーム型アクセス・調査装置(Current image)（現時点のイメージ図）

Unit 2

Grasping mobility of deposits at the bottomof the pedestal when force is applied
PCV
Φ110mm(X-6

FY2021)

Consider to increasesampling volume

Image of internal PCV investigation using guidepipe

Unit 3

Considering availability of internal investigation technology that has been demonstrated in parallel tostudy water level reducing in PCVConsidering necessity of further investigations in utilizing ROV that was used in the previous investigation

Start of fuel debris retrieval for the first implementation unit

Grasping distribution of structures ordeposits outside the pedestal (including
sampling)

FY2020

Determine the method of fuel debris retrieval
for the first implementation unit

Unit 1FY2019Drawings in the balloons were cited from the material from IRID, TEPCO5

Fig. 8 Further internal investigation schedule and image of investigation equipment

Technical issues for ensuring the safety
3.1.3.3.1 Concept for ensuring safety in fuel debris retrieval work
Generally, when it comes to deal with securing safety at nuclear facilities, it is required toestablish the concept of securing safety of nuclear facilities, which clarifies the necessary safetyfunctions and corresponding protective measures based on the scope of typical accident events,
in order to avoid unnecessary exposure to radiation or excessive meaningless protective measures.
However, in the decommissioning operation of the Fukushima Daiichi NPS, it may be difficult toapply the idea of a stereotyped and standardized concept of securing safety like a normal nuclear
power plant. Therefore, it may be necessary to establish the concept of securing safety according
to the severe actual circumstances at the site. For this reason, NDF has been working on organizing
the idea of securing safety at the site (see Section 4.2) in order to foster a shared understanding
among related organizations including regulatory authorities that share the purpose of promoting
safe and smooth decommissioning.
In operations involving unprecedented uncertainty such as the fuel debris retrieval, it may benecessary to incorporate reasonable protective measures at each conceptual design phase for thefuel debris retrieval steps while taking continuous actions to reduce the uncertainties of the PCVs’internal conditions (see Fig. 6).

34

As the concept of ensuring the safety during the fuel debris retrieval work is being constructed,
NDF is focusing on the studies on specifying the technical requirements for ensuring safety during
fuel debris retrieval work, as shown in 3.1.3.3.2 to 3.1.3.3.7. Specific measures for securing safety
not limited to fuel debris retrieval work are described in Section 3.5.4.
3.1.3.3.2 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 maintaining under negative pressure) and pressure in a PCV is maintained
to be equal to that inside of a reactor building (passive containment function). On the other hand,
the reactor buildings, PCVs, etc. of the Fukushima Daiichi NPS are partially damaged by thehydrogen explosion and their containment function has been deteriorated. Due to this,establishment of an active containment function by controlling pressure in negative is being
considered during fuel debris retrieval work. Moreover, from the viewpoint of prevention ofhydrogen explosion due to steadily generated hydrogen by the process of radiolysis of water andof corrosion (inactivation) of structural material due to the presence of oxygen, nitrogen is currently
injected into the PCV to maintain it in a nitrogen atmosphere and under slightly positive pressurestate. The dispersion of radioactive material has been preventing by the PCV exhaust gas control
system which was furnished with filters to remove radioactive material and radioactivity
measurement system.6
When retrieving the fuel debris in future, the dispersion of radioactive fine particles (α-dust)
containing α-nuclides that has a large contribution to dosage during internal exposure derived
from fuel debris may occur due to the works such as fuel debris cutting and drilling, and it may
create a concern for increasing the concentration of radioactivity in the PCVs gas phase. Therefore,a function for containing the gas phase with protective measures should be provided that suppress
the dose impact of workers and public within the allowable value, by minimizing the disperse of α
dust from PCV as much as possible taking evaluated dust condition at normal operation as well as
abnormal condition into account.
Accordingly, in order to build a reasonable containment function at the each stage as describedin Section 3.1.3.1, it is reasonable to upgrade the scale of the retrieval work while verifying thevalidity of the containment function built in the subsequent stage by grasping the tendency of dust
dispersion at the each stage of scale expansion. For example, internal PCV investigations and thesmall-scale retrieval planned in the near future, α-dust emission can be controlled by reducing theamount to be retrieved, and selecting a retrieval method that does not involve cutting and drilling,
while maintaining the current inert atmosphere (slightly positive pressure state due to nitrogen gas
charge) inside the PCVs. Based on both the results of monitoring changes in the state of α-dustemission from the retrieval work and the evaluation of effects on the surroundings, it will likely be
6

Evaluated radiation dose value at the site boundary due to the release of airborne effluent is 0.03mSv/year, on
the other hand, the evaluation of additional emission amount from the reactor building is approximately 2.9 x 10 4mSv/year. (Evaluation value at the time of July, 2018) (Refer to Section 3.5.2.2)

35

shifted to the larger size of the fuel debris retrieval work gradually, using such as the cutting anddrilling equipment. If it is evaluated that the expansion of the retrieval work may induce the effects
on the surrounding environment, further study is required for the means to strengthen thecontainment function by equalization or negative pressurization inside the PCVs as well as
constructing a secondary containment function outside the PCVs.
R/B and Container

Primary Containment Boundary(PCV Repair + Work Cell)

Cell (Top Access)* PCV closed w/top head in the case of side accessSecondary Containment Boundary(R/B + Cover or Container)* Necessity will be considered
Negative Pressure
Control System

Air Phase
Sectionabove PCV
water level(Air Phase Section)

PCV Repair(Negative Pressure Maintained)

Cell
Land Surface

Inner PCV water level should
be kept at/below this level

PCV Repair Water Shielding (Liquid–phase)
Stagnant Water Level

Ground Water level

Fig. 9 Example of containment function (gas phase) by negative pressure control
For building this containment function (gas phase), technical issues to be addressed immediately
are as follows.
(1) Understanding of dispersion rate of α dust, etc.
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 gas
phase as much as possible based on the collected data.
In order to collect data such as dispersion rate of α dust, it is necessary to plan the dispersion
rate measurement during work state of sampling and small-scale of fuel debris retrieval work.
Moreover, in order to progress with the technology investigation and R&D activity for the fuel debris
retrieval method/system under the condition where such actual data cannot be obtained,
approximately understanding on the general behavior of α-dust dispersion. R&D for verification ofthis approximate understanding using simulated debris is currently underway.7
In order to suppress the transition of α-dust to the gas phase, it is desirable to submerge thefuel debris and to perform retrieving of it as much as possible under water. However, the waterlevel in the PCV is to be adjusted with other technical requirements such as the building of the7

Refer to "(4) Elucidation of Behavior of Radioactive Suspended Particle Behavior Generated in theDecommissioning Process" (including α-dust Measures) in Attachment 12

36

containment function of the liquid phase described in the next section. Therefore, not all fuel debris
can be retrieved under water, and transition of α-dust to gas phase is considered to be mitigatedby splashing water on the fuel debris that is not submerged.
(2) Feasibility of negative pressure control in the PCVA.

Technical feasibility of maintaining negative pressure based on site conditions
In order to maintain pressure in the PCV in negative, enough capacity of gas exhaust system

is required considering PCV damage condition. Although damaged part have not been identified
yet, and the exhaust capacity is currently evaluated based on the relationship between actual
nitrogen supply volume and actual PCV internal pressure. At this time, it is necessary to maintainsufficient pressure difference to respond to the internal pressure rise of the PCV due to abnormal
events such as an internal temperature rise or stoppage of the gas exhaust system. In order toachieve these, repair of the damaged parts of at the upper part of the PCV will be considered asnecessary, but there are some difficulties such as remote work or exposure of workers due to work
under high 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 as well as the information obtained at each stage
described in Section 3.1.3.1.
B.

Influence of air flow into the PCV on negative pressure control
Air flows into the PCV when maintaining pressure in the PCV in negative. Therefore, it isnecessary to examine measures of maintaining inactivated condition in the PCV by increasing
nitrogen gas supply into the PCV as necessary based on evaluation on occurrence of accidents
such as fire and Hydrogen explosion using accumulated information regarding to volume ofhydrogen generated by radiolysis of water in the PCV and inflow volume of air (oxygen) into the
PCV. Evaluation results on influence of air (oxygen) inflow on corrosion of structural material andits preventive measure are discussed in Section 3.1.3.3.6 for details.
C. Necessity of secondary containment function
As illustrated in Fig. 9, fuel debris is planned to retrieved and collected into the canister whichis to be packed into the transfer cask in the working cell which is installed to be connected to the
PCV maintained pressure under negative. The PCV and this working cell have a primarycontainment function to prevent of α-dust to the exterior.
In addition to this, in order to respond to the event that radioactive material is dispersed fromthe containment boundary caused by loss of primary containment function by fail of maintaining
negative pressure in the PCV and connected working cell, necessity of the secondary containment
function has been investigated. The secondary containment function is established by maintaining
negative pressure inside of the Reactor Building and a connected cover or cell which is installed
close to the Reactor Building. However, large capacity of gas exhaust system is considered to be37

required to maintain negative pressure in the secondary containment boundary since the reactor
building has a large volume and the its leak tightness may have been deteriorated due to theaccident. Therefore, based on the accumulated results of the tendency of dust dispersion obtainedat each stage of the fuel debris retrieval scale expansion, it is necessary to identify the required
functions to establish the secondary containment boundary and to perform the necessary
technology development.
D. Control of deterioration of PCV containment function
In order to maintain pressure in the PCV in negative during the fuel debris retrieval work, it isnecessary to consider how to respond deterioration of the primary containment function consistedof the PCV and the attached cell due to earthquakes and aging. This is discussed in detail in Section
3.1.3.3.6.
Study on exhaust gas management
In managing gas exhaust system for maintaining pressure in negative, radioactive material
derived from fuel debris in exhaust gas shall be controlled under the level of dose rate limit for
public around the site by measuring and controlling release concentration and release amount of
the radioactive material. In addition, concentration or amount of the nuclides derived from the fueldebris which releases α or β (γ) rays in the exhaust gas needed to be regularly measured so as
to evaluate their normal fluctuation range in advance during fuel debris retrieval work. By using
such data, it is necessary to develop a system to detect abnormal events such as nuclides leakage
in advance and to mitigate the impact of the released nuclides on environment and workers.
Reliability or accuracy of mechanical property and chemical composition of the fuel debris
needs to be improved because these are essential information for designing the decontaminationequipment for efficient collection of radioactive dust.
3.1.3.3.3 Establishing the containment functions (Liquid-phase)
As discussed in the previous section, to lower the dispersion rate of generated α-dust and to
mitigate the transition to the gas phase, the fuel debris retrieval work would be performed underwater condition by submerging the fuel debris or by pouring water over the fuel debris. In suchcases, a great number of α-dust will be mixed in cooling water (liquid phase). To prevent the
impact of α-dust in the cooling water on the environment, it may be important to establish a coolingwater circulation/purification system as well as a containment function of liquid phase.
The Roadmap requests to investigate feasibility of the PCV cooling water circulation line forcooling water taken from PCV and injected to the RPV which is necessary at the time of fuel debris
retrieval work. As described in Section 3.3.2.3, this PCV cooling water circulation and purificationline is effective to prevent α dust dispersion during the fuel debris retrieval work because it has aneffect of preventing contamination of stagnant water in the building. At this time, to ensure a morereliable containment function, water sealing methods by repairing the bottom of the PCV pouring

38

the grout materials have been studied. However, based on the results of previous experimental
studies, it is becoming clearer that water sealing by repairing the bottom of the PCV is very difficult.
Thus, in consideration of the PCV repair technology and the results8 of the actual scale test, it may
be important to explore a retrieval method to minimize leakage with the realization of suitable
circulation and purification systems of cooling water (See Attachment 6).
To establish a reasonable containment function of liquid phase at each stage as scale grows
described in Section 3.1.3.1, it is rational to monitor the radioactive concentration of the coolingwater and verify the validity of the containment function built in the subsequent stage. For example,
if the PCVs internal investigations and small-scale retrieval work are performed using the current
water circulation system, it may be possible to suppress the increase of the radioactive
concentration of cooling water by reducing the retrieved volume and selecting a retrieval methodthat does not require cutting and drilling, together with the containment function (gas phase). Fromthe viewpoint of investigating the methods to detect the effects induced by the retrieval work on theliquid phase, the circulating water system would be monitored, and based on the confirmation ofmonitoring results ofα-nuclides, the fuel debris retrieval method may gradually be changed to alarger retrieval method using cutting and drilling equipment. If an abnormal event such as a large
volume of cooling water outflow from a PCV to the torus room of a reactor building occurs,
measures should be taken such that the water level in the torus room be maintained lower than the
groundwater level to prevent the outflow of cooling water to the surrounding ground. For this reason,
it is necessary to establish an appropriate PCV water level and a PCV water level control system.
In establishing this containment function (liquid phase), technical issues to be addressed
immediately are as follows.

IRID, “FY2016/2017 Report of the results on Development of PCV leakage repair technology , Subsidy for
Decommissioning and Contaminated water management project by the supplementary budget FY2015”, releasedin July 2018, http://irid.or.jp/wp-content/uploads/2018/06/20170000_08.pdf ,IRID, “FY2015 Report of the result (Final Report) on The actual scale test for PCV repair technology, Subsidy for
Decommissioning and Contaminated water management project by the supplementary budget FY2015”, releasedin March 2018, http://irid.or.jp/wp-content/uploads/2018/06/20170000_07.pdf
839

R/B and Container

Primary Containment Boundary(PCV Repair + Work Cell)

Cell (Top Access)* PCV closed w/top head in the case of side accessSecondary Containment Boundary(R/B + Cover or Container)* Necessity will be considered

Circulating
Water Cooling
SystemWater Level
Management

PCV Repair (Liquid-phase))(Liquid Phase Section)

CellLand Surface

Liquid Phase
Sectionbelow PCV
water level

Inner PCV water level should
be kept at/below this level
PCV Repair Water Shielding (Liquid–phase)

Stagnant Water Level

Ground Water level

Fig. 10 Build example of containment function (liquid phase)(1) Suppression of radioactive concentration increase in the cooling water due to fuel debris
retrieval work
It is necessary to plan to investigate the impact of the sampling or small scale of the fuel debris
retrieval work on radioactive concentration in the circulating cooling water.
For suppressing an increase of radioactive concentration in cooling water in the PCV, dust
generated by fuel debris cutting is planned to be collected through PCV circulating cooling water
system for mitigate dispersion of dust. Demonstration and confirmation of its performance are
required at the stage of small scale of fuel debris retrieval.
(2) Feasibility of PCV water sealing
With an aim of improving the containment function of liquid phase at the bottom of the PCV by
suppressing leakage of cooling water into the torus room, various PCV water sealing methods
including blockage of the vent tubes, burying of the tip of the down-comers and burying the pipe
strainers on the S/C wall by concrete have been studied. (Refer to Fig. 10)
Workability and performance of the water sealing technology of blockage of the vent tubes
were examined through an actual scale test, using self-filling concrete for blockage of the venttubes. Drilling holes through 1st floor of the Reactor Building, wall of the S/C and the vent tubes
prior to concrete pouring into the vent tubes as well as treatment of these drilled holes after watersealing work completion are also necessary to be performed. Countermeasures for the worker's
exposure and reduction of the radiation at the work area are also required for these works. In
addition, method/technology with high repeatability for blockage of the vent tubes shall be
developed because all sealing work of eight vent tubes need to be completed at the same time.
However, current method/technology has not been reached sufficient reliable level.

40

Rise in concrete temperature during its curing, height of placed concrete need to be studied
and repair material for these cracks is required to be developed because some cracks on the placed
concrete surface were observed at the full scale test of water sealing by injecting concrete into theSuppression Chamber (S/C) for burying of the tip of the down-comers and the pipe strainers on the
S/C wall. The feasibility of S/C guide pipe installation work necessary for concrete injection into the
S/C has been experimentally examined. These experiment results show that improvement of theremote construction technology for the guide pipe connecting part, development of the remoteexamination technology including nondestructive examination at the connection part compiled with
Regulatory Requirement and its maintenance method need to be continued. These activities havebeen carried out as tasks in the Government-led Project of Decommissioning and Contaminated
Water Management.
In addition for establishing the water tightness at the penetration of piping through PCV shelland torus room wall, material for water sealing and the remote placing equipment have beendeveloped, and the feasibility of the water sealing work using them has been confirmed. However,
it is difficult to access to such place of actual units for water sealing work because of high radiation
field. It is also presumed that there are water leakage through unknown paths such as small cracks
or gaps were observed in unit 5 and 6 even after the water sealing work. Therefore, complete watersealing is considered to be difficult at units 1 to 3.
As mentioned above, technical issues still remain to be continuously resolved in order toimprove the performance of PCV water sealing, as well as current level of water sealing technology
is found not to be able to achieve complete water sealing. Even in case of practical water sealing
work of blockage of the vent tubes, it is necessary to install the PCV cooling water circulation
system for controlling the water level in the PCV. In addition, due to the low seismic margin of the
S/C supports and water leakage paths from the PCV other than from S/C, control of water level in
the PCV and torus room has been studied low by cooling water circulation system. In the future,additional study will be carried out, as necessary, to resolve the technical issues identified throughthe examination on the applicability of current water sealing technologies to each unit which is being
performed as a part of the preliminary engineering work. Finally, the applicability of these watersealing technologies to actual unit should be judged based on these results.

41

Reactor Building

Vent tube water seal

Vent tube

PCV
RPV

Strainer Water Seal

Strainer
Downcomer Water seal
（Immersion of Downcomer ）Downcomer
S/C
Torus Room
Enlarged views in the right are material from IRID8 that was partly modified.

Fig. 11 PCV lower area water sealing (Examination example)(3) Setting of water level in PCV
Because the seismic margin of the S/C supports is low water level in the PCV is heavily
depends on the applicability/usability of water sealing technologies to each unit.
In the case that flow of cooling water from the Dry Well (D/W) into the S/C can be suppressedby the blockage of the vent tubes, the water in the PCV (D/W) can be maintained at relatively highlevel. If not, it shall be kept at low level. In such case, cooling of fuel debris, radioactive dust
dispersion and location of taking water from the D/W need to be examined, as well as the waterlevel in the PCV, S/C, torus room and groundwater are required to be monitored and controlledduring fuel retrieval work.
Attachment 7 shows the water level in the PCV of each unit for the three cases of water sealing
work application. First is the cases of blockage of the vent tubes, second is burying of the tip of thedown-comers and the pipe strainers on the S/C wall by high-flow concrete and last is no watersealing work.
3.1.3.3.4 Maintaining the cooling functions
The decay heat of the fuel debris has been decreased dramatically since the core melt
accidents. However, it may be necessary to keep the cooling function to prevent nuclides fromshifting from the liquid phase to the gas phase due to thermal energy during the fuel debris retrieval
work. At present, the cold shutdown state is maintained with keeping the temperature well below
100℃ using cooling water. In addition, during the fuel debris retrieval work, it may be necessary42to keep the temperature below the level at which the fuel debris retrieval device can continue towork without any problems for a long period of time. Further, it may be required to consider
alternative water injection functions including mobile equipment, in preparation for an event thepermanent equipment cannot maintain cooling due to a large earthquake, tsunami or other events.
At this time, to maintain the cooling function at each stage of retrieval scale expansion ddescribed in Section 3.1.3.1, the consistency with other systems should be checked, such as thewater level control system within the PCVs for confining the liquid phase and the circulating watercooling and purification system. Based on the data obtained for each stage of retrieval scale
expansion, it may be necessary to examine the circulation cooling system to be required in thesubsequent stage while evaluating the reasonable circulation flow rate necessary for cooling.
However, it should be considered that the injection of cooling water may become redundant in the
future due to further decrease in the decay heat, as the fuel debris retrieval progresses.
In maintaining this cooling function, the technical issues to be addressed for the time being are
as follows.
(1) Establishment of a temperature target during operation and countermeasures to an abnormal
conditions
It is necessary to establish the PCV internal temperature target at which each operation of fueldebris retrieval work is able to continue. Also, countermeasures against the unanticipated abnormal
condition need to be studied. An essential countermeasure would be to continue cooling water
circulation by early recovery of the cooling water circulation system or by the recovery action bymobile equipment, as well as it is necessary to establish countermeasures and their implementation
process/step such as salvage of equipment taking change of condition in the PCV including time
margin until reaching the abnormal condition.
Also, monitoring items and their criterion need to be studied and prepared through thepreliminary engineering work, in order to carefully proceed fuel debris retrieval work with observing
how this work will affect the existing cooling water circulation and purification system as well as its
cooling function.
3.1.3.3.5 Criticality control
At present, the monitoring of the concentration of rare gas (Xe-135), which is a fission product("FP") of short half-life, has shown no sign of criticality as it does not exceed the criticality criterionof 1 Bq/cm3. In addition, the possibility of re-criticality of the fuel debris at the Fukushima Daiichi
NPS is presumed to be extremely low from the engineering viewpoint, because the alternation ofmolten fuel assemblies is not likely to reach criticality due to the abundance ratio with water andthe mixture of impurities such as internal structures. Further, the fuel debris is presumed to scatterin a wide range, instead of remaining in the core, as results of the accident progression. Even if
criticality should occur, it may be pointed out that the effect of re-criticality is presumed to be small
because the fuel debris widely scatters.

43

To keep the subcritical condition more reliably, it is important to store the retrieved fuel debris
stably, controlling the shape and size of retrieved fuel debris, for instance, by storing the debris instorage cans. To prevent fuel debris from reaching criticality during the fuel debris retrieval work, it
is necessary to investigate what kind of conditions during the work would lead to criticality, including
the shape of the fuel debris or the water volume changes, and establish an appropriate
management method with combined functions of prevention, detection, and termination of criticality.
To realize the management method to avoid re-criticality, it is necessary to carefully expand
the scale of fuel debris retrieval while inspecting the character/composition etc. of fuel debris ateach stage described in Section 3.1.3.1. In the early stage of the retrieval work, the fuel debris
should be retrieved in a way that does not significantly change the shape of fuel debris. Then, thecomposition and characteristics of fuel debris and the fluctuation of the neutron signal in the vicinityof fuel debris handled during the work should be checked to evaluate the possibility of the criticalityof the fuel debris. The volume of fuel debris retrieval at one time can be increased based on themeasurement of subcritical conditions or taking measures such as placing neutron absorbers.
However, in case the confidence level is not enough by just limiting the retrieval amount of fueldebris additional measures such as measuring the degree of criticality before retrieval operation
and/or inputting neutron absorbing materials may be considered.
For this criticality control, technical issues to be addressed for the time being are as follows.
Establishment of criticality evaluation method
As described in Section 3.1.3.1, it is necessary to refine the information on the conditions forreaching re-criticality of the fuel debris based on obtained information in each step of the fuel debris
retrieval work. In this regard, the evaluation method have been developed to estimate theconditions for sub-criticality and the degree of influence of criticality if it should happen. In
conducting these evaluations, it is necessary to plan so that the information on the criticalparameters can be obtained through the process of progressing internal investigations.
Local neutron measurement around retrieval point
As existing neutron detectors, there are various kinds according to the application such as
fission chamber, B-10 proportional counter tube, semiconductor detector, etc. It is important toselect a neutron detector taking advantage of each feature. The key specifications for the neutron
detectors required for the criticality monitoring are, (1) ability to survive accumulated dose (Gy)according to the operation period and (2) installability in assumed equipment (size/weight, cablediameter). Accordingly, it is important to select an optimal detector based on the information on
PCV dose rate obtained by internal investigation and the progress of equipment development foreach unit.
For the preparation of the actual operations, it is necessary to examine the installation locationof neutron detectors and also to set up criteria for stopping retrieval operations and for determining
the injection rate of boric acid, which is a neutron absorber, if fluctuations in neutron flux should

44

arise.
Feasibility study of measuring the degree of sub-criticality
When measuring the degree of sub-criticality, in addition to the required specification of (2), it
is necessary to select a detecting device with high fidelity and response speed in order to measurea weak neutron signal, capturing a neutron fluctuation in a very short time under a gamma rayenvironment. From the examination to date, the equipment mountability (size/weight) of detectors
is a key issue 9 mainly due to the necessity of shielding the lead under a high gamma rayenvironment (assuming 1000 Gy/h). In the future, the neutron detector has been selected andoptimized, reflecting the constraints introduced by the study results of the project of the fuel debris
retrieval method and system. Meanwhile, efforts have to be put into ascertaining the gamma raydose rate and neutron count rate around the fuel debris and downsizing the detector as well.
In addition, in order to verify the applicability of the measurement system of sub-criticality forthe actual retrieval work of the fuel debris, where various mix of the compositions andcharacteristics of the fuel debris are expected, it is necessary to formulate a plan for verification of
the technology reflecting the actual circumstances.
Feasibility study of neutron absorber
Based on the information obtained at each stage described in Section 3.1.3.1, the feasibilityof constant use of sodium pentaborate as a neutron absorber has been studied, investigating thenecessary boron concentration and the installability of the equipment, in case it turns out that thecriticality of fuel debris is relatively high9. While examining the influence on PCV circulation coolingsystem and concrete operation to maintain boron concentration, based on the fuel debris
composition obtained by the retrieval operation, it is necessary to judge whether the injection of
sodium pentaborate would be required during normal retrieval operation.
In addition, if criticality should occur, urgent sodium pentaborate injection would be used toachieve sub-critical state, but after the suspension, it is necessary to select a method of maintaining
sub-criticality, which may be the decrease in water level or the maintenance of boron concentration.
In addition, the development of a non-soluble absorbing material which can locally restrain theinfluence on the PCV circulation cooling system, has been performed as well. B 4C metal sinteredmaterial or B/Gd-containing glass or Gd2O3 particles and water glass /Gd2O3 granulated powder
have been proposed as candidates so far by conducting fundamental physical property test andradiation resistance performance test. For the candidate materials, further studies are underway
on methods of obtaining data by long-term exposure test for the storage of the fuel debris, andmethods of dispersion to fuel debris and confirming the effect after spraying as adaptability

IRID, Final Report of “Development of Criticality management technology for fuel debris” Subsidy for
Decommissioning and Contaminated water management project by the supplementary budget FY2015, releasedin March 2018, http://irid.or.jp/wp-content/uploads/2018/06/20170000_05.pdf
945

corresponding to debris processing9.
Detection of criticality by PCV gas management installations
The current PCV gas management equipment is to be maintained as criticality monitoring inorder to detect re-criticality due to the fall of fuel debris and/or accumulation of debris of fineparticles in location other than the cutting area.
By measuring Kr - 87/88 in addition to Xe - 135 that has already been measured, it has beenfound that criticality detection can be accelerated and that the level of subcriticality of the entire
PCV can be presumed9. However, the applicability to the actual plants is to be examined in the
future.
3.1.3.3.6 Securing structural integrity (earthquake resistance) of the PCVs and reactor
buildings
The reactor buildings, PCVs, and RPVs experienced the hydrogen explosions when theaccident occurred. Further, the exposure to the high-temperature environment and the corrosion
due to seawater injection may have affected the integrity of the structures. During the period of thefuel debris retrieval work, it is necessary to secure the integrity for the important structures like
PCVs and RPVs and to suppress the deterioration of containment functions of PCVs and reactor
buildings, taking into account the effects mentioned above against the possible occurrence of asevere earthquake. In addition, the deterioration of containment function of PCVs and reactor
buildings due to corrosion have to be suppressed during the retrieval work. Furthermore, it may berequired to prepare for the mitigation measures in advance, evaluating the impact on human beingsand the environment induced by the postulated damage to important functions due to a large
earthquake.
In maintaining the structural integrity and safety functions of the PCVs and the buildings, thetechnical issues to be addressed for the time being are as follows.
(1) Evaluation of earthquake resistanceA.

Evaluation of earthquake resistance based on the impact of accidents and secular changes
In the previous studies, the reactor buildings and major components of PCV, RPV and RPV

pedestal of each unit, have been evaluated such that they have relatively large seismic safetymargin against a design basis seismic ground motion Ss (600Gal), taking into account the damage
caused by the accident, the degradation over 40 years and the increased weight of new facilities
required for the fuel debris retrieval work.
In these evaluations, while taking into account major damage caused by hydrogen explosions
in the reactor buildings of Unit 1, 3 and 4, on-site damage investigations such as observation of
cracks, have been performed by engineers and/or robots. Particularly, in the reactor building ofUnit 4 that was heavily damaged by the hydrogen explosion but with low dose, the investigation of
cracks and concrete strength of major structural parts have been done periodically. According to46

the results of the investigations, it was identified that there has been no development of harmful
crack and the concrete strength of major structural parts exceed the design strength as well10.
However, the periodic investigations of the buildings have to be implemented continuously over theperiod of the fuel debris retrieval work and a maintenance plan for the buildings shall be preparedto suppress the deterioration, in which remote repairing methods are to be included in case ofharmful cracks are found.
For the evaluation of pedestals, conservative conditions have been set taking into account the
influences of exposure to the high-temperature environment11, based on the results of accident
progress analyses and the information obtained by PCV internal investigation. Since there might
be a possibility of erosion by the fuel debris, case studies of seismic margin of pedestals have been
performed by analyses with the variation of hypothesized eroded zones of the pedestals, taking
into account the material test results of both concrete and rebars on the effects exposed to hightemperature environment.
The S/C supports have relatively small seismic margins and may be greatly affected by theway of lower PCV repair. Therefore, the evaluation has been performed by a detailed analyticalmodels. The results shows that the S/C supports may have a certain level of safety margin against
the design basis earthquake ground motion11 in the case of the water seals at the bent tubes andthe strainers or down-comers inside S/C with relatively small berried depth of grout.
It is necessary to evaluate the seismic safety margin of R/Bs, PCVs and RPVs in more detailand to implement the countermeasures, according to the progress of the further PCV internal
investigation and the studies of the fuel debris retrieval methods.
B.

Evaluation of the impact induced by postulated damage due to a large earthquake and the
preparation of the countermeasures
Preparatory studies of the countermeasures have been done for the postulated damage tomajor structural components such as a pedestal during a large earthquake. In the studies,countermeasures using mobile equipment are also included such as a countermeasure using
mobile pumps for the drainage of the torus room in case major leakage from a PCV occurs due tolarge cracks induced by the damage to the pedestal. As future investigations improve theunderstanding of the PCV internal situations, it may be necessary to reevaluate thecountermeasures and to reflect the results into the concept of the fuel debris retrieval.
Countermeasures to suppress the deterioration during the fuel retrieval work
As described in Section 3.1.3.3.2, when maintaining negative pressure in the PCV, there may
be a concern that the concentration of oxygen rises and corrosion progresses due to inflow of

10 e.g. Contents from Measurement to the Fukushima Daiichi NPS accident/, Reservation and status of
implementation on interviews with regulated person, etc. dated on July 24th, 2014, “The result of periodic
inspection (the 9th) for confirming integrity of the reactor buildings of Unit 4 of the Fukushima Daiichi NPS”,
released in July 2014 on NRA web, http://www.nsr.go.jp/data/000054381.pdf
11 IRID, “Development status of integrity assessment technology for PCV/RPV” Subsidy for Decommissioning
and Contaminated Water Management project by the supplementary budget FY2013, released in November
2015, http://irid.or.jp/_pdf/201509to10_09.pdf

47

atmospheric air during fuel debris retrieval work. The evaluations of earthquake resistancedescribed in (1) have been done considering the impact of corrosion progress. It has been
evaluated such that they have relatively large seismic safety margin against a design-basis seismic
ground motion Ss(600Gal) taking into account of thinning by material corrosion for 40 years as
described above. On the other hand, in order to maintain the current conditions of PCVs, RPVs
and major piping, it may be necessary to suppress the progress of the corrosion during the long
period of the fuel debris retrieval work. In one of the subsidized R&D projects of theDecommissioning and Contaminated Water Management, countermeasures against further
suppression of corrosion, have been studied to select suitable corrosion inhibitors and toinvestigate the practical application of corrosion control measures, which would be able to get intoalignment with the current liquid process system.
The effectiveness of steel corrosion inhibitors for the PCVs, RPVs, piping, etc. have beentested, taking into account the influence of radiation environment and seawater injection, andcorrosion inhibitor options effective for not only overall corrosion but also localized corrosion are
extracted12. On the other hand, in order to mitigate the influence of the corrosion inhibitors on theexisting water cooling circulation and purification system, it has been shown that it is necessary to
reduce the corrosion inhibitors concentration considerably before the preliminary stage of the water
purification system12. Henceforth, to realize a PCV circulation cooling system for the fuel debris
retrieval, it may be required to come up with the countermeasures that satisfy both the corrosion
control and other required functions such as the water purification, in an integrated manner.
3.1.3.3.7 Reduction of radiation exposure during work
The main work areas of fuel debris retrieval are high dose areas such as inside the reactor
buildings. Also, in the new stage of fuel debris retrieval, there comes a need handling of nuclear
fuel materials containing α-nuclides derived from fuel debris with a large internal dose impact.
Accordingly, the strict control of not only for external exposure but also for internal exposure isessential.
Specifically, it is important to prevent excessive exposure to workers through appropriate
radiation protection schemes depending on the working environment (target nuclide, dose
equivalent ratio, air concentration, surface density) and working style (direct operation or remote).
Regarding protection from external radiation exposure, the radiation exposure dose is evaluatedconsidering the radiation sources and the dose rate in the work area. Then, based on the threeprinciples, namely “time, distance, and shielding" (to shorten the exposure time, distance from theradiation source, shield if possible), it is needed to take measures to keep the radiation exposure
dose as low as reasonably achievable.
Therefore, an appropriate combination of exposure reduction measures such as
IRID, “The report of the result on Development of corrosion inhibition technology for PCV/RPV”, Subsidy for
Decommissioning and Contaminated Water Management project, released in July 2017, http://irid.or.jp/wpcontent/uploads/2017/06/20160000_10.pdf
1248

decontamination, shielding, remote technology etc. is to be selected, with the following ideas inmind.
Consider first of all the reduction of exposure to radiation by 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 contaminated areas such as inside the PCV and torus rooms, work shouldbe pursued by remotely controlled machines, etc. to avoid engaging personnel inside.

In the areas inside the R/B buildings other than the above, the best combination of
decontamination, shielding, removal of unnecessary objects, remote technologies andreduction of on-site time, with considerations for the balance between the exposure during
the decontamination tasks and the exposure during the PCV repair tasks, to minimize thecumulative dose for overall work.

Where remote technologies are employed, additional work will be required, such as theinstallation of systems, maintenance and technical troubleshooting, which must be taken
into consideration in the above evaluation and planning.



As for the decontamination tasks, the judgment between remote technologies and personnel
employment must be made based on factors such as the dose rate in the target areas, typeof contamination, space for work, frequency of use, applicability and development situationsof remote technologies, schedule and cost, etc.



A priority must be placed upon areas where work requirements are clearly identified.
Considerations must not be pursued if task requirements are unclear, or in a non-specific“betterment-oriented” manner such as to aim for an overall reduction of radiation dose.
Regarding the protection from internal exposure, measures such as suppressing dispersion ofradioactive dust and prevention of contamination expansion are being taken and appropriate
protective measures are to be selected depending on the target nuclides, concentration in air andsurface contamination density in the work area, to prevent inhalation ingestion and body pollution.
In addition, it is essential to prepare for countermeasures against internal exposure cases inadvance and establish an emergency exposure medical care system including exposure evaluation
by bioassay etc.
Meanwhile, for the reduction of radiation exposure in the long-term decommissioning, it isnecessary to have an overall perspective and comprehensive, shared measures of exposure
reduction. It is important that actual on-site performance and lessons learned from them provide
feedback to the planning of subsequent steps to improve the accuracy of the plans, and to prevent
recurrence of problems such as delays in work procedures. The above approach must be adopted
to develop systems of cross-sectional exposure reduction management and know-how/technology
transfer.
In particular, concerning the reduction of exposure of workers in the reactor building, it isimportant to set up a dosage reduction plan in consideration of the contamination condition in order
to secure the necessary operation environment for the operation target area by the appropriate

49

dosage reduction technique (decontamination, removal, shielding). The target dose rate of theoperation target area is set to be lower than the exposure dose limit (50 mSv/year and 100 mSv/5
years) specified by laws and regulations, it is necessary to consider and set the operation methods,
working hours and the number of workers by considering them in a single group. Further, it isimportant to appropriately update the information on dosage reduction technologies and utilize it
on-site.
Since various nuclides are mixed at the Fukushima Daiichi NPS, for exposure management,
it is necessary to select nuclide (attention nuclide) for which an appropriately estimation of the
exposure dose for each curve font can be rationally conducted. In other words, based on the nuclide
information in each operation environment, select the nuclide (curve font) for external exposure
management and internal exposure management. In the external exposure management, it shallreflect on the dose equivalent rate management and the basis for wearing the protection equipment.
In the internal exposure management, it shall reflect on the management standard of radioactive
material concentration in the operation environment air and the basis for wearing the operator's
respiratory protection equipment etc., and also for the installation of air purification equipment.
Therefore, appropriate protection measures should be taken. In particular, it is necessary to study
the measurement management considering the mixture of alpha-nuclides from both hardware andsoftware sides and to prepare for the application to the fuel debris retrieval work.
Technical issues related to fuel debris retrieval methods
3.1.3.4.1 Securing access route
For transporting, installing and unloading the equipment and devices used for fuel debris
retrieval work, and transporting fuel debris and waste, an access route should be established byremoving obstacles and reducing the dose to the level at which such tasks can be done. Whenestablishing a new opening in the PCV or the like to construct an access route to fuel debris, fromthe perspective of a containment (the air phase) function described in Section 3.1.3.3.2, it should
be kept in mind to suppressing the release of radioactive materials from the PCV and RPV andmaintaining the integrity of existing structures.
According to the fuel debris retrieval policy based on the partial submersion method that
extends the side access to the bottom of the PCV, TEPCO is now conducting the preliminary
engineering work. Based on the results of R&D of decommissioning and contaminated water
management, an access route from the PCV side opening to the fuel debris is to be constructed.
Then, it will be drawn up a plan in which a side wall opening is made on the reactor building andthe side wall opening on the PCV is enlarged as required. In this case, in the side access method,
it is a problem that there is a risk of exceeding the load bearing capacity of the reactor building floor
at the time of installation of the cells and the like. For this reason, a comparative study of a methodof releasing the load to the outside of the building is being conducted, such as suspension bridge
system and access tunnel system.

50

In the future, based on the above-mentioned tasks, it is necessary to define the access route
clearly to be built at the next stage from the data obtained at each stage of the internal investigation
etc. described in Section 3.1.3.1.
According to the fuel debris retrieval policy, the optimum retrieval method should be selected
depending on where the fuel debris exists for each reactor unit. At present, it is considered on thepremise of accessing from the top to the inside of the RPV.
3.1.3.4.2 Development of devices and equipment
In order to retrieve fuel debris safely, reliably and efficiently, it is necessary to develop fueldebris retrieval devices and equipment that meet the site requirements and have the necessaryfunctions. To flexibly respond to the situations inside the RPV and the bottom of the PCV wherefuel debris is predominantly present, this devices/equipment should necessarily meet specificationsof radiation resistance, dust resistance, waterproofness and protection mechanisms such as thetemperature target described in Section 3.1.3.3.4, remote investigation/maintainability, remote
operability, securing visual field, earthquake-resistance, protection mechanism for collision
avoidance or automatic stop in case of abnormality, high reliability, appropriate redundancy, arescue mechanism that does not disturb the subsequent work when trouble occurs, such as
efficiency of retrieving fuel debris (payload).
Based on these issues, it is necessary to clearly define the fuel debris retrieval
devices/equipment to be used at the next stage from the data obtained at each stage described in
Section 3.1.3.1. As the functions to be specifically implemented, the development of a recovery
system that can handle various conditions of fuel debris (fragments, sludge, or fine powder), a fueldebris cutting system (laser, boring, or crushing, etc.), and a dust collecting system, is underway.
Furthermore, a technique for installing retrieval equipment is required. The technology used forremote operation, including installing the working cell for establishing a containment function (gasphase) and the device for removing obstacles to provide an access route, are now underdevelopment. (Refer to Fig.12)
Although these technological developments are currently under way in the project fordecommissioning/ contaminated water, in the future, the developed devices and equipment will be
combined and undergo remote mock-up tests to verify that the performance mentioned above isdemonstrated on the actual site.
This remote mockup test needs to be implemented by a facility simulating the site in order toverify applicability of technically developed remote equipment to the actual environment andoperability/maintenance ability of entire remote system under the harsh environmental conditions
including uncertainties. It is therefore important to rationally implement according to purpose, suchas utilization in units 5, 6. Therefore, in cooperation with related organizations, NDF is working onorganizing and examining how to proceed with the remote mockup test plan, the way of the test
plan review, the scope of the mockup facility to be maintained, the necessary time, operation
management, etc.

51

Example of removing interference
objects technology (side access)
Example of removing interference objects
technology (top access)

Example of working cell installation
technology(Inflate seal)

Example of working cell installation
technology (biological shielding wall drilling)

Example of fuel debris collection system

Examples of contamination spread prevention
technology

Example of cutting system element technology(chisel, ultrasonic wave)(Images provided by IRID)

Fig. 12 Fundamental technologies for retrieval of fuel debris and internal structures
3.1.3.4.3 Establishment of system equipment and working areas
When retrieving fuel debris, it is necessary to install system equipment (Container/ Working
cell/ including devices and equipment) to establish safety functions and operate them appropriately.
Also, it should be prepared that sufficient spaces for installation, operation, and maintenance, andfor installing shields for reducing radioactive exposure to operators, so that the required
environmental conditions are satisfied.
The system equipment includes a negative pressure control system required for establishinga containment function of the gas-phase described in Section 3.1.3.3.2, a circulation watercooling/purification system required for maintaining the containment function of the liquid phase
described in Section 3.1.3.3.3 and cooling function described in Section 3.1.3.3.4, and a criticality
control system required for controlling criticality described in Section 3.1.3.3.5. The installation planfor the respective unit is being examined13. Because it is essential to monitor the internal situation
when retrieving fuel debris, the development of measurement systems (visualization, pressure,temperature, radiation, criticality (rare gas concentration, etc.) hydrogen concentration, etc.) is an
important future issue to be addressed. An investigation on how to implement the systemequipment integrating these systems is under way.
The working area required for installing the fuel debris retrieval equipment/related devices andsystem equipment is now being calculated. Considering the handling of the high-dose area in thereactor buildings and interference with other tasks, study of setting up the systems, including

IRID, “The report of the result FY2017, “Advancement of retrieval method and system for fuel debris andstructure inside reactor” Subsidy for Decommissioning and Contaminated Water Management project by the
supplementary budget FY2016, released in April 2018, http://irid.or.jp/wpcontent/uploads/2018/06/20170000_10.pdf
1352

outside of the existing buildings, is underway. From now on, in order to carry out fuel debris retrieval
work and to implement detailed examination of the layout of the area for installation and operationof the facilities constituting each system, proceed with consideration of the place for temporary
placement/treatment of the removed equipment, on-site plot plan for storing the retrieved fuel debris,
etc.
Technical issues related to safe and stable storage of fuel debris
3.1.3.5.1 Handling of fuel debris (collecting, transferring and storing)
The fuel debris retrieval work requires to establish a comprehensive system consisting of aseries of steps from collecting, transferring to storing of retrieved fuel debris furnished with safetyfunctions such as maintaining subcritical condition, containment function, countermeasures against
hydrogen generation, and cooling. Specifically, the basic specifications of the canister such as thetotal length and inner diameter has been studied taking easiness of its handling, the storing
efficiency and maintaining of the subcritical condition into account14. The allowable transportation
period has been investigated in accordance with the estimated volume of hydrogen generation fromcanister. Based on these results, details of equipment/systems for collecting, transferring, andstoring of the retrieved fuel debris need to be discussed. Storing facility considering to thesafeguards requirements is also necessary to be studied based on the estimated volume of theretrieved fuel debris. It is also important to plan mock-up tests for confirmation of handling of thecanister combined with a fuel debris retrieval devices.
Proper handling method/devices for retrieved fuel debris in the subsequent stage need to bestudied and improved taking the data obtained from work experience in previous stage describedin the Section 3.1.3.1 as well as above mentioned technical issued into account. It is currently
considered to be appropriate that the retrieved fuel debris at the step of the fuel debris samplingmust be collected in the shipping container for transporting to the radioactive material analysis andresearch facility outside of the site, and retrieved fuel debris during the small-scale fuel debris
retrieval step should be stored appropriately at a temporary storing facility until completion of theactual storing facility because volume of the retrieved fuel debris at this step will be small.
The Roadmap states that the processing and disposal method of the retrieved fuel debris areinvestigated and fixed during the third phase after starting the fuel debris retrieval work.
3.1.3.5.2 Treatment of radioactive waste during fuel debris retrieval
During fuel debris retrieval work, various radioactive waste such as replaced or disassembled
components or parts other than retrieved fuel debris are to be generated from fuel removal
device/facility inside/outside of the PCV, at each stage of fuel debris retrieval work including its
IRID, “The report of the implementation FY2017, “Development of technology for Collection, Transferring andStorage of fuel debris” Subsidy for Decommissioning and Contaminated Water Management project by the
supplementary budget FY2015, released in June 2018, http://irid.or.jp/wpcontent/uploads/2018/06/20170000_03.pdf
1453

preparation and withdrawal work. It should be properly classified and stored under safe condition
same as fuel debris.
It is practically difficult to accumulate and organize necessary and sufficient information suchas distribution and characteristics of the fuel debris and the radioactive wastes such as coreinternals in PCV prior to its retrieval work which are essential for investigating classification
standard and storing method, even though it is continued in operation with collecting and organizing
such information. Also, since the essence of the fuel debris is considered to be the existence of
nuclear fuel material in it, it is desirable that the retrieved material is classified into fuel debris orradioactive waste based on the concentration of nuclear fuel material in it. However, it is currently
considered as highly-difficult technical challenges to accurately measure or estimate theconcentration of nuclear fuel material in the retrieved material itself or in it collected into the canister.
It is important to develop a classification standard in order to judge appropriately such material
into fuel debris or radioactive waste during retrieval work, even if only limited information on the
retrieved material is available in advance. From above reason, detailed studies on treatment of
retrieved materials including classification standard needs to be continued based on the findings
and information obtained from the investigation inside of the PCV.
3.1.3.5.3 Study on Safeguards Measures
Nuclear facilities in Japan need to observe the duties concluded with the International Atomic
Energy Agency (“IAEA”) of comprehensive safeguards agreement and its additional protocol, andare expected to show that nuclear materials are utilized for peace purpose only as declared by
implementing material accountancy as a part of the domestic safeguards system, and by accepting
containment/surveillance of nuclear materials, and inspection and supplemental access by IAEA.
In the Fukushima Daiichi NPS, as above, it had been confirmed via the IAEA safeguards
activities that the nuclear material was used as declared. After the nuclear accident, however, thebasic information of the facilities and the situation for utilization and storage of the nuclear materials,
provided in advance for the execution of the safeguards, greatly changed. In a sound nuclear power
plant, the nuclear material is clearly counted as an item one by one with a fuel assembly of a certain
defined shape as one unit, but implementation of conventional material accountancy andsafeguards is deemed difficult for the units 1 to 3 of the Fukushima Daiichi NPS, since the fuelassemblies would no longer maintain their shape as an item due to melting, the damaged facilities
would not allow necessary containment/surveillance schemes for key measuring points, and highradiation would restrict the entry of human beings into the areas and prevent normal investigation
activities, and so forth. Additional safeguards operations are currently applied under this situation
to the units 1 to 3 as an alternative measure to ensure no undeclared relocation of the nuclear
materials, and so forth.
Given the fuel debris retrieval (FDR) operations in future intrinsically move nuclear materials
from the inside of PCV to the outside of PCV in the units 1 to 3 of the Fukushima Daiichi NPS,
corresponding safeguards operations will be required. However, in addition to the difficulty of item

54

control, fuel debris is highly likely to contain structural materials other than nuclear materials andso forth, and the confirmation of the composition and quantitative evaluation of contained nuclear
materials are believed quite hard to technically achieve. Accordingly, material accountancy andsafeguards to be applied to the units 1 to 3 of the Fukushima Daiichi NPS would pose an essential
question on how realistic solution could be created to replace the conventional scheme.
Towards the solutions for these issues, discussion started in the Fukushima Task Forceestablished between the IAEA and the Japanese Nuclear Regulation Authority about thesafeguards options being reflected to the probable FDR schemes. Japan is providing updated
information of the units 1 to 3 of the Fukushima Daiichi NPS from the early stage to the stakeholders
including the IAEA, aiming at the prompt and smooth application of the safeguards in the FDR
operations. These efforts are consistent with the idea of the IAEA’s “Safeguards by Design(considerations to be paid to the safeguards from the design phase)”. Realistic and transparent
enough material accountancy and safeguards, which would not impose excessive load to the site
work, must be proposed proactively by Japan to reach the global agreement among thestakeholders including the IAEA.
Summary of key Technical issues
The main technical subjects and plans described in this section are summarized as shown in
Fig. 13.

55

56

Access Route &Equipment for Survey

Technologies of Fuel Debris

2019

2020

Determination of fuel debris retrieval methods
for the first implementing unit

2021
Start of fuel debris retrieval
at the first implementing unit

2022

Continued as necessary→

Continue as necessary

Measurement for ensuring safety during large-scale fuelretrieval and study on safer handling of fuel debris

Retrieval method (layout plan)
Collecting/transferring plan
Storage plan
Ensuring of Safety(confinement, water level control,

Small-scale retrieval &Preparation of large-scale retrieval

Retrieval equipment
Facilities for collecting, transferringand temporary storage
Radiation protection for safety(modification of existing facility)

Preliminary engineering / Detailed engineering

Retrieval method (Layout plan etc.)
Collecting/transfer plan
Temporary storage plan
Securing Safety

Research and development

Technical studies for on-site work

On-site work

Transfer method to outside the site
Adaption plan for Safeguards
Additional plan for internal investigation

Preparation of small-scale retrieval

Continue as necessary

Analysis technology
for fuel debris

Access Route &Equipment for Survey

Preparation of investigation/ internal investigation/ Characterization of fuel debris (from retrieved samples etc.)

Retrieval method,
system & concept design

2018

Fig. 13 Technical issues and further plan on fuel debris retrieval

inside PCV- Development of collection, transport, and Storage- Development of closed circulation systems for water

Retrieval and Internal Structure of Fuel Debris

Fuel Debris- Advancement of Fundamental Technologies forStudy for subsidized project on dicommissioning andcontaminated water management- Advancement of Retrieval Method and System ofEngineering involved with fuel debris retrieval

Retrieval of fuel debris

Engineerng for internal investigation- Development of Investigation Technology of Inside RPV- Development of Technologies for Grasping andAnalyzing Properties of Fuel Debris

to Retrieving Fuel Debris and Internal Structure- Development of Sampling Technologies- Develoment of Technologies for In-depth Investigationof PCV Inside

Study for subsidized project on dicommissioning andcontaminated water management

Internal Investigation

Milestone

Fiscal Year3.2 Waste management
3.2.1

Sectoral target

Near-future targets in waste management are as follows:(1) As the approaches of solid waste storage, the Solid Waste Storage Management Plan(“Storage Management Plan”) is appropriately developed, revised and implemented including
waste prevention, volume reduction and monitoring, with updating the estimated amount ofsolid waste to be generated in the next ten years periodically.
(2) As the approaches for processing/disposal, countermeasures integrated from characterization
to processing/disposal of solid waste will be proceeded from the expert point of view, and theprospects of a processing/disposal method and technology related to its safety should bemade clear by around FY2021.
3.2.2

Sectoral strategy
The concept of risk reduction in waste management

Waste management is a long-term effort that needs to attain the prospect of implementing final
disposal, while reducing risks in every stages. For the terms regard to radioactive waste
management, the definitions of terms provided by IAEA is shown in Attachment 8.
The solid waste, such as concentrated wastewater, waste sludge, slurry in the high integrity
container (HIC), “rubbles, etc.”, and contaminated building structures, is unlikely to increase in riskin the future, but they are still a source of risk which should make its countermeasures properly inthe decommissioning process. (refer to Section 2.3) Compared with other major risk sources, they
are in a state where the risk level is generally lower than other major risk sources, and it isconsidered that a constant risk level can be continuously maintained by appropriate maintenanceand management in the future.. Solid waste is, however, highly diverse of characteristics, including
waste contaminated by radioactive materials adhering to the surface, waste in the form of sludge,
and waste with high dose that generates hydrogen. For this, risk reduction measures are to be
implemented systematically by continuously conducting research on risk reduction for safer
management, by seeing throughout disposal and by ensuring safety as a precondition.
The basic policies on solid waste
The Strategic Plan 2017 explains the properties of solid waste estimated based on the accident
conditions, post-accident decommissioning efforts, and the results of the characterization of waste
so far. Based on these findings, the plan proposes a policy of solid waste management, along withstrategic proposals that include specific tasks to be performed at present (i.e. statement for the
preparation of the basic concept for the processing/disposal of solid waste).

57

The Roadmap revised in September 2017 summarizes the basic concept regarding solid waste,based on the content of the strategic proposals.

＜Basic policies on solid waste＞① Thorough containment and isolation
Solid waste management should be implemented thoroughly, with containment and isolation ofradioactive materials to prevent their dispersion/leakage and human access to them, in order not cause
harmful radiation exposure.
② Reduction of solid waste volume
The amount of solid waste generated by decommissioning is reduced as much as possible in orderto ease the burden of solid waste management.
③ Promotion of characterization
To proceed with study on processing and disposal method of solid waste, characterization of solid
waste such as nuclide composition and radioactive concentration is needed. In addition to the fact thatsolid waste of the Fukushima Daiichi NPS is large in volume, and have varied nuclide compositions, itis necessary to address an increase in the number of analysis samples and proceed theircharacterization properly.
④ Thorough storage
To dispose of solid waste, it is essential to understand the volumes and characteristics of the solid
waste, and to establish specifications of disposal facilities and technical requirements for waste
packages (technical requirements for disposal). However, the volumes and characteristics of solid
waste will become clear step by step, with the future clarification of progress and plan ofdecommissioning. Therefore, the solid waste generated should be stored safely and reasonably
according to characteristics of solid waste. 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
In order to safely store solid waste, the system for selecting the method of processing for stabilizationand immobilization (preceding processing) will be established, and selecting the method of thepreceding processing, before the technical requirements of disposal are established.

⑥ Promotion of effective R&D with a bird’s-eye-view of overall solid waste management
To efficiently proceed with R&D concerning solid waste management, close cooperation should be
realized between R&D fields such as waste characterization, processing and disposal. Issues anddiscussions on R&D should be shared between parties, and necessary planning made with a bird’seye-view of overall solid waste management, should be progressed collectively.
⑦

Efficient implementation of R&D projects from the perspective of overall solid waste management
In order to continue safe and steady solid waste management, the continuous operational frameworksystem including development of adequate facilities and human resources, which are concerned withsolid waste management, must be undertaken.

⑧ Measures to reduce the exposure of workers to radiation
To steadily proceed with solid waste management, it is important to ensure the safety and health ofworkers. Therefore, radiation exposure control, safety management and healthcare programs shouldbe implemented thoroughly based on the relevant laws/regulations.
(Note) Numbered and titled by NDF for each item.

The challenge on solid waste generated by decommissioning of the Fukushima Daiichi NPS
is the existence of a large volume of waste with various characteristics compared to the waste
generated from the decommissioning of normal nuclear power plants. While the characteristics ofthe solid waste are known to some extent so far, it is necessary to improve capabilities of analysis(i.e. efficiency and capacity) to further clarify such characteristics, to foresee the final disposal and58

to develop a flexible and reasonable waste stream (integrated measures from the characterization
to processing/disposal of the waste).
Specifically, the relevant organizations should proceed their efforts based on each role in linewith the basic policies, while the technical study on the integrated countermeasures, from thecharacterization to processing/disposal of solid waste, should be proceeded with following policies
at the initiative of NDF.
Storage
3.2.2.3.1 Storage at present
The fundamentals of managing solid waste are to contain not to scatter or leak, by placing it
into container or immobilizing, and so on, as need. Also, it should be kept isolated in a properly
storage place, and managed appropriately by monitoring and so on.
From the view point of minimizing the volume of solid waste to be generated and to bedisposed of in order to lower environmental impact, “Waste hierarchy” (1.Waste Prevention,
2.Waste Minimization, 3.Re-use of Materials, 4. Recycling, 5. Disposal, in the order of preferredapproach) is shared as a priority of measures to be taken in the U.S. and the UK. It has succeededin suppressing the final volume of disposal by implementing waste management according to thisconcept. For practically developing Waste Hierarchy, it has been pointed out that waste
management departments should be involved in it from the stage of developing construction work
plan for decommissioning.
In the Fukushima Daiichi NPS, such efforts are being made in line with the “Waste Hierarchy”
shown in Fig. 14. It is important to instill the concept of the waste hierarchy, and raise awarenesson reducing the volume of solid waste to be generated.

59

Fig. 14 Summary of Waste Hierarchy at the NDA, UK15 and Countermeasures at the Fukushima
Daiichi NPS
Among the solid waste that has been generated by April, 2018, “rubbles, etc.” have beenclassified into rubble, fallen tree, used protective clothing and so on, and about 400,000 m3 of thewaste is stored in temporary storage areas according to the surface dose rates. Among thesecondary waste generated by water treatment, about 4,000 used vessels, which have undergone
additional measures such as adding shield for increasing water tightness, are stored in usedadsorption column storage facilities.
Table 3 shows the status of storage of the solid waste.
To store 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 suchas on building waste management facilities to be required based on the volume.
The Storage Management Plan states that the volume of “rubbles, etc.” should be reduced as
far as possible, and will be carried to store in the solid waste storage building around FY 2028,
excluding reusable rubble. Regarding secondary waste generated by water treatment, large, heavy
waste such as used vessel is planned to be stored in a building that will be built to accommodate
large waste. These measures will consolidate waste storage areas and supersede many of thecurrently scattered temporary storage areas (Refer to Attachment 9).
In the Storage Management Plan, however, the future generation of some items is left
uncalculated. It will be reflected to the estimated amount of generation when the plan is
implemented and the possibility of activity such as dismantlement become known in the next 10
years. Since the estimated amount of generation fluctuates depending on the progress of thedecommissioning work and plans in the future, it is necessary to revise the estimated amount once
15 NDA : Nuclear Decommissioning Authority, Edited based on the “Summary of the Waste Hierarchy” on Figure 7of p. 60 of the "NDA Strategy Effective from April 2016," 2016.

60

a year and update the Storage Management Plan as appropriate, as TEPCO expresses to do.
Table 3 Solid Waste Storage Management Status(a) Management status of rubble, fallen tree, used protective clothing, etc. (as of April 30,
2018)
Rubbles
Surface dose rate
(mSv/h)Storage method≤ 0.1

Outdoor storage
Outdoor sheet covered
storage
Soil-covered temporarystorage facility,
Temporary storage tent,
Outdoor container storage
Containers (in Solid waste
storage building)

0.1～1

1 to 30

> 30
Total

----

Classification

Storage method

Root

Outdoor storages

Branch/leaves

Temporary storage pool

Total

----

Stored volume (m3) / Storage
capacity (m3)(Percentage)
174,800 / 250,700(70%)
36,600 / 71,000(51%)
21,900 / 27,700(79%)
8,900 / 45,600(20%)
242,200 / 395,000(61%)

Fallen tree
Stored volume (m3) / Storage
capacity (m3)(Percentage)
96,600 / 134,000(72%)
37,300 / 41,600(90%)
133,900 / 175,600(76%)

Used protective clothing, etc.
Stored volume (m3) / Storage
capacity (m3)(Percentage)
56,000 / 71,200(79%)

Storage method
Container(b) Management status of secondary waste generated by water treatment (as of May 3, 2018)

Storage
Place

Type of Used Vessels

Outdoor Temporary
Storage area of used
Vessels

Used vessels

Cesium adsorption apparatus
2nd Cesium adsorption apparatus
HICs from multiple radio-nuclides
removal system
HICs from improved multiple radionuclides removal system
Used vessel from high-performance
multiple radio-nuclides removal system
Used column from multiple radionuclides removal system

61

Storage
NumberStorage Number/ Capacity(Percentage)

767
198
1,470
1,263
74
11

3,983 / 6,368(63%)

Used vessels and filters from mobiletype strontium system

200

Waste Sludge
Sludge storage facility (Indoor)

597

m3

597 / 700
(85%)

9,364

m3

9,364 / 10,700(88%)

Concentrated waste liquid
Concentrated waste liquid storage tanks(Outdoor)
3.2.2.3.2 Further safety improvement in storage
The secondary waste generated by water treatment, which has high fluidity (such as slurry
generated from multi-nuclide removal equipment and waste sludge generated from thedecontamination system) poses a relatively large risk in the storage because of the fluidity, and it
should be stored in a more stable and reasonable way after risk reduced by performing a certain
processing. In general, it is ideal to conduct waste processing based on the technical requirementsof disposal once it is established, if the processing is conducted prior to disposal. However, there
may be cases where processing for stabilization and immobilization are required although thetechnical requirements for disposal are not determined (i.e. preceding processing). Therefore,
study will be continued on how to select the preceding processing method with disposal in mind.
The possibility that the specifications of solid waste conducted preceding processing areincompatible with the technical requirements for disposal should be minimized. Therefore, safety
evaluation should be conducted for each specification of processed solid waste, using scenarios,
models, data and so on, for multiple disposal methods which are rational and feasible to implement,
and also suitable for features of the waste without specifying disposal facility’s place, scale and soon. The selection of preceding processing method should be studied based on the result.
Study on the processing/disposal measures
The Roadmap specifies that the prospects of a processing/disposal method and technology
related to its safety should be made clear by around FY2021, including waste which does not needpreceding processing. The overall picture of solid waste becomes clear step by step according to
the progress of the effort. Keeping it in mind that it will still remain in a stage of accumulatingnecessary information on its characteristics around FY2021, the concrete targets for technicalperspective are listed as follows: Establish safe and rational 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 method reflecting features of the disposal
concept, with considering examples of foreign countries.
 Clarify radiological analysis and evaluation method for characterization.

62 Clarify processing technology that could be expected to introduce actual equipment forstabilization and immobilization considering disposal for several important waste streams
such as secondary waste generated by water treatment.
 Establish method of rationally selecting processing technology to stabilize and immobilize
waste based on the above methodology although the technical requirements for disposal
are not determined (i.e. preceding processing).
 Have prospect of setting processing/disposal measure for solid waste of which theprocessing technology considering disposal is not clarified, using a series of method to be
developed by 2021.
 Clarify issues and measures concerning storage of solid waste until conditioning
It is described in the Roadmap that, in accord with these efforts, specifications and production
methods of the waste packages should be determined in the third phase after the start of fuel debris
retrieval. Then, a processing system should be installed in the Fukushima Daiichi NPS. Afterestablishing the prospects of disposal, production of waste packages should be started, and thencarried out.
3.2.3

Technical issues and plans for promoting sectoral strategies
Promotion of characterization

Under the circumstance that the Building 1 of JAEA Okuma Analysis and Research Center isscheduled to commence operation at the end of FY2020; it is important to improve the accuracy ofmodels to obtain evaluation data based on the limited radiological analysis data. The study will beproceeded on the method of reflecting dispersion of radiological analysis data to inventory
evaluation using analytical method, and concept of setting and revising radioactive inventory basedon comprehensive evaluation of radiological analysis data and analytical value.
The radiological analysis has been conducted for characterization so far, but target ofradiological analysis will be mainly on the predisposal management and review targeted nuclides
in near future. The study on radiological analysis to make it simple and rapid is proceeded together
to develop efficient radiological analysis method. While reducing exposure by introducing remotesampling method, high dose sampling will be conducted. Through these efforts, it will be achallenge to build an environment where system, facilities/equipment, and technologies for highly
accurate characterization of solid waste will be established by the end of FY 2020, and necessary
radiological analysis data will be acquired for some solid waste.
Further safety improvement in Storage
The secondary waste generated by water treatment is proceeded to be dehydrated, forstabilization and extracted and transferred from temporary storage facilities to storage facilities ona hill as risk reduction measures for the time being.
On the technologies for stabilization, immobilization and conditioning of the secondary waste63generated by water treatment, some technologies such as the high temperature processing
technology and cement improvement technology has been considered. Challenges for introducing
actual equipment should be dealt, and gathering and evaluation of data on technical requirements
should be proceeded by using engineering scale test equipment and so on, from the point of view
whether it may contributes to establish selection method for preceding processing method.
Processing technology expected to be applicable for actual processing should be identified, andspecification of waste package should be determined.
Regarding the methods of storing high-dose solid waste generated as a result of fuel debris
retrieval, study should be proceeded on the items such as the way of sorting fuel debris/waste, thetype of waste, the evaluation of waste volume, and the flow of handling waste, and narrow down
candidate methods for storage. For other solid waste, study should be proceeded on generation ofhydrogen during storage, together with the timing and content for the case of further measuresthose are required in order to secure safety, and reflects to the Storage Management Plan asneeded.
Establishing processing/disposal concept and development of safety evaluation
methodIn order to select candidate technologies for preceding processing methods, it is necessary to
conduct safety assessment with the specifications of the waste package that is developed by therespective candidate technologies, as a target. The selection of reasonable and feasible candidate
technologies and the development of suitable safety assessment methods will be proceeded until
the end of FY 2021. In particular, the research on those such as domestic and international waste
acceptance criteria, disposal concepts and safety evaluation methods are conducted; then, thestudy of disposal concept based on the characteristics of solid waste of the Fukushima Daiichi NPS
is proceeded; subsequently, multiple disposal methods are to be presented; and suitable safety
evaluation methods will be developed.
The precision of the selection results for preceding processing methods are then to be
improved using the specifications of conditioned waste including more precise data on theproperties.
OthersAs solid waste to be generated with the retrieval of fuel debris in future, structures such ascore internals and outside of reactors to be dismantled and removed and second waste such asfilters generated during fuel debris retrieval related works are expected. It should be noted that thiskind of waste may include α-nuclides that derived from fuel debris, it is necessary to proceed
study on the storage method for the solid waste, together with the study on the method of fuelremoval.
As a long-term challenge, TEPCO is scheduled to develop the decommissioning plan in thethird term after the start of fuel debris retrieval, as later described in Subsection 3.5.3. At that time,64

measures against the waste generated from dismantling buildings and so on should be establishedbased on the progress of decommissioning, future expectations, the conditions of reactor and otherbuildings, and the latest R&D and so on.
In addition, as measures for promoting efficient decommissioning, the characterization using
quick measurement (in situ analysis and on-site analysis) by means of electromagnetic methods,
and so on, should be examined in addition to radiochemical analysis, and R&D required for thispurpose is expected.
Summary of key Technical issues
In regard to the technical issues described in this Chapter, entire management of solid waste
is overlooked and studies are proceeded on close cooperation between the efforts to respective
issue as shown in Fig.15.

65

Fiscal Year

2015

2016

2017

１．Reduction of solid waste

2018

2019

2020

2021

2022 and later

Carry-in control, reuse, recycling and volume reduction

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

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

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

Establishment of Database system

Development of radioactive material analysis and research facilities(2) Improvement of analysis
Human resource development for analysis personnel

capacity(3) Improvement in efficiency ofDevelopment of characterization method with complimentarily use ofdata evaluated from nuclide migration model

waste characterization
Optimization of the number of analytical samples, review of nuclides
to be analyzed, simplified and speed-up analysis methods, etc.

３． Storage

Storage(1) Storage Management Plan

Development and review of the Storage Management Plan

(2) Studyand evaluation of

Study on Storage measures of secondary waste from water treatment

Study on safety
improvement
measures

Storage method
Study on storage method, storage containers,
and the way of putting waste into container(3) Study and evaluation of storage
methods of solid waste generated

Design/prepare
Storage containers

during fuel debris retrieval

４． Development of

Establishment of a selection system of preceding procesing

process/disposal concept
Survey and setting of disposal method
based on characteristics of solid waste

and development of safetyassessment method

Survey and development of safetyassessment method for each disposal method
Survey on conditioning
technology

Technical development of processinhg methods related to stabilization and immobilization

Compilation of basic concept ofprocessing/disposal

Clarification of prospects on processing/disposal
method and safety technology

Study of processing/disposal

Fig. 15 Technical issues and further plan on waste management66

Use of a
selectedsystem3.3 Contaminated water management
3.3.1

Sectoral target

The near term objectives of measures against contaminated water are as follows.
(1) Under the three basic principles concerning contaminated water issues (“Removing”
contamination sources, “Redirecting” fresh water from contamination sources, and “Retaining”
contaminated water from leakage)16, the reinforcement and optimum operation of the waterlevel control system should be continued. The multilayered measures should be implemented
to complete the processing the stagnant water in the buildings by 202017.
(2) Considering the total decommissioning process including the full-scale fuel debris retrieval
beginning in near future, the long term strategy should be examined for the measures of thecontaminated water.
3.3.2

Sectoral strategies
Concept of risk reduction in contaminated water management

For the contaminated water management at the Fukushima Daiichi NPS, as stated in Section
2.2, various measures based on the three principles has been implemented both to thecontaminated water in the buildings and trench/pits and uncontaminated water such as
groundwater/storm sewage.
From the viewpoint of measures to reduce the risk due to the radioactive material, safetymeasures for the stagnant water in the building are required as soon as possible. Since thestagnant water is generated from groundwater/storm sewage mixed with the cooling watercontacted with the fuel debris, it contains a considerable amount of the dissolved radioactive
materials. The storage conditions of the stagnant water are not inherently and include uncertainties.
Thus the Safety Management of the stagnant water in the buildings is relatively high. The stagnant
water is recovered and treated with cesium adsorption apparatus (KURION and SARRY) etc., andthe radioactive materials are transferred to the water treatment secondary waste such as
adsorption vessel with lower Safety Management. As the risk level of the stagnant water in thebuildings, the Hazard Potential is reduced.
As measures against risk reduction so far, drainage of stored water in condenser in Unit 1
turbine building and removal of the stagnant water in the lowest floor area of Unit 1 turbine building
16

Nuclear Emergency Response Headquarters "Tokyo Electric Power Company, Incorporated Basic Policy on
Contaminated Water Problems at the Fukushima Daiichi Nuclear Power Plant On September 3, 2013, Nuclear
Emergency Response Headquarters "Tokyo Electric Power Company, Incorporated Additional measures for the
problem of contaminated water at the Fukushima Daiichi NPS" December 20, 2013
17 Exposing the floor line for buildings other than the reactor buildings and lowering the water level of the reactor
building to T.P.-1740mm (O.P. - 300 mm or less) (Cyclic water injection cooling is carried out in the reactor
building, and stagnant water continues to exist).

67

(completed in March 2017) have been carried out, henceforth, the quantity of radioactive materials
will be reduced continually from the stagnant water in the turbine building where the radioactive
material concentration is high.
Steady execution of contaminated water management indicated in the Roadmap
Since the Fukushima Daiichi Nuclear Power Plant is located in an area with abundant
groundwater, the groundwater flows into the reactor buildings and mixes with the cooling water
injected to the fuel debris. This is the main cause of generation of the contaminated water in thesite. Treating the contaminated water and the ground water was urgent issue. Currently, as a resultof the multilayered measures based on the three principles, the site situation has shifted from that
requiring urgent measures immediately after the accident to certain stable situation that can
examine a medium/long term plan. The multilayered measures are; pumping up high-concentration
contaminated water in the underground tunnel (trench), blockage of the trench, purification by multinuclide removal equipment (ALPS), paving (facing) of the site to prevent the storm sewage
infiltration to reduce the amount of groundwater, groundwater bypassing, pumping up of thegroundwater in wells (sub drains) near the reactor buildings, installation of a land-side impermeable
wall to reduce the groundwater inflow into the area of the reactor buildings, installation of the seaside impermeable wall for suppressing the outflow of groundwater into the ocean, and soil
improvement with the water glass in the seawall area of the ocean-side of the buildings.
The milestones mentioned in the Roadmap are as follows: (1) reduction of the contaminatedwater generation to about 150 m3/day (by 2020), (2) storing all the water treated by multi-nuclide
removal equipment in the welded type tanks (FY 2018) (3) by lowering the level of stagnant waterin the buildings, separation of the penetrations between Units 1 and 2 and between Units 3 and 4,
respectively (by 201818), (4) reduction of the radioactive materials in the stagnant water in thebuildings up to approximately one tenth of the amount at the end of FY 2014 (FY 2018), and (5)completion of the treatment of stagnant water in the buildings (by 2020). It is expected that thesepreventive and multilayered drastic measures will be steadily implemented to achieve themilestones.

18

Completed in September 2018.

68

2015
FY2015

Item/ Year

2016
FY20162017
FY20172018
FY20182019
FY20192020
FY2020Current

Water level of Groundwater/building

Building stagnant water

Groundwater level

Processing complete

1 - 4 Building water level
Current groundwater level
Current building water level
Building stagnant water storage

T.P.-36

T.P.460
Approx.T.P.460

No. 1 T/B
1 onlyT/B Low waterlevel
End of 2014 Almost86,000m
86,000 m3 3
2014

LessT.P.-1740
thanT.P.-1740
Units 1 to 3 performing circulating water~injection
The lowest floor surface exposure of the building otherthan the reactor building

Almost53,000m
53,000 m

33

1-4 Buildings and integrated
water processing buildings※

Rw/B
Unit 1 Rw/B
floor exposure

1.0E15 Bq
Building residence water radioactive substance amountThe amount of radioactive material in thestagnant※water in the buildings is
calculated from the radioactivity
concentration measurement value and
Cs137
Cs134
the storage
amount Sr90of the representative
nuclide (Cs 134, Cs 137, Sr 90). For thisreason, the radioactivity concentration of
the stagnant water in the buildings
fluctuates due to the influence of thestagnant water and the like having a highradioactive concentration locally, so thatthe amount of radioactive substancesometimes increases or decreases due
to evaluation. As for the R/B stagnant
3
R/B
water of Unit 3 where high radioactivity
concentration was confirmed, it will bereflected after checking the concentration
distribution etc.

8

6

Units
T/B lowest floor middle
2 42 to 4T/B
exposure

6,000m3 m3Less than about 6,000

2 24 to 4 Rw/B,
Rw/BT/BT/B
Units
floor exposure
exposure

2018
1/10

2014

Milestone of the Roadmap (Reduce the amount of radioactive substances in building residence water to
one tenth as of the end of FY 2014 within FY 2018).

Estimated trend of the amount of radioactive substances remaining in buildings.

1 1 T/B
T/B
Unit
floor exposure
Building stagnant water
Actual value

55

※

4

55th Monitoring and Evaluation Study Meeting
Reservoir in condenser
Actual value

Half-value
2014 at the end of fiscal 2014

Approx.2.2*22.2

2

55

55th Monitoring/Evaluation Study Meeting

0.4
Approx.0.4*2

1/10of the end of fiscal
2014 * 12014
1/10

Approx.0.2*
0.22

2020
End
of 2020

0.06
Approx.0.06*20[Note] T/B
︓ nuclear reactor building,T/B︓ turbine building, R/BR/B
Rw/B Rw/B
︓ Waste treatment building

Fig. 16 Changes in the amount of radioactive materials in the stagnant water in the buildings19

Contaminated water management based on the relationship with fuel debris
retrieval
As mentioned above, the contaminated water management at the Fukushima Daiichi NPS,
although out of a state where urgent countermeasures are required and it still needs further efforts,
it seems that it has shifted to a certain stable condition where the medium to long-term plan can be
examined. Because the full-scale decommissioning work including the fuel debris retrieval will start
after a while, it is necessary to discuss the optimal control of the contaminated water and thegroundwater together at each stage of the decommissioning process. Furthermore, it should
examine a future images of the contaminated water management considering the inherent
measures and the issues to be addressed with a long-term perspective until the completion of thefuel debris retrieval.
In the present situation, cooling water injected into RPV flows to PCV, then the water leaks
from PCV to the torus room in the basement floor of the reactor building. A circulation water coolingand purifying system has been adopted in which the stagnant water is recovered in a turbine
building through penetration and purified with cesium adsorbent equipment (KURION and SARRY)
and then reused as cooling water. At this time, by controlling the water level of the stagnant waterin the reactor buildings lower than that of the groundwater around the buildings to let the
groundwater flow into the building (in-leak), the function of the buildings for the containment can be
secured by preventing outflow of the radioactive materials to the outside of the building. (SeeTEPCO, “Progress on the stagnant water in the buildings” Handout 2-3 of 61st The Committee on Supervision
and Evaluation of the Specified Nuclear Facilities”, Dated on July 6th, 2018
19

69

Attachment 6).
As described in the preceding paragraph, according to the target process of the Roadmap, it
is considered that the processing of the stagnant water in the buildings other than the reactor
buildings of unit 1 - 3 will be completed by 2020. By that time, the circulation water cooling andpurifying system needs to be modified to the system that the stagnant water is recovered in thereactor buildings (instead of turbine buildings) and purified to use as the cooling water.
Furthermore, the studies including feasibility of the PCV circulation cooling system for fueldebris retrieval is under way, and water shielding by repairing the bottom of PCVs has been
examined to secure the multiple boundaries. However, it has been found that complete water
shielding through repairing the bottom of PCVs is very difficult, so the circulation cooling systemshould prepare for the inflow of α-particles from inside of the PCVs to the stagnant water in thereactor buildings. Even if the water shielding is applicable, it is necessary to consider the setting ofa sufficient difference in water level between stagnant water in the reactor building and thegroundwater, in case a large amount of cooling water leaks from the PCV into the reactor building.
The details of the containment function of this liquid phase have been described in 3.1.3.3.3.
In addition to this, requirements from both sides of the fuel retrieval work and the contaminatedwater management should be clarified and coordinate them in case of items to be reconciled suchas the increase of the radioactive materials concentration arose as the fuel debris retrieval work
proceeds.
Also, as the fuel debris retrieval proceeds, the water injection for cooling the fuel debris will be
no longer necessary. Then, there is no water stagnant in the lowest floor of the reactor buildings.
In such a situation, the groundwater level should be adjusted to a level below the lowest floor of
the reactor buildings, so it can be expected that there is no water inflow into the reactor buildings.
In such a case, it is important to plan to develop a system that can stably control the groundwaterlevel for a long period by studying, the combination of the dynamic equipment and the passive
equipment with lower potential of machine troubles.
3.3.3

Technical issues and plans for promoting strategies
Steady execution of contaminated water management in the Roadmap

Since there are penetrations between the buildings of each unit, the water levels of thestagnant water in the buildings are almost the same in all buildings. In order to prevent the stagnant
water in the building from leaking to the outside, it is necessary to make the water level lower than
the groundwater level (keep the water level difference). Stable management of groundwater aroundthe buildings has been made by strengthening the function of sub-drains and completing the landside impermeable wall, etc.
By proceeding with these preventive and multilayered measures, the amount of contaminatedwater has been reduced to approximately 220m3 per day in FY2017 from approximately 400m3 per
day in FY 2016 (Fig.17). Also, the amount of the groundwater pumped-up from sub-drains and the70groundwater drains in the seawall area have been reduced as well20.
Start operation of sub-drain
サブドレン稼働開始
Start closing land-side impermeable wall
陸側遮水壁閉合開始

Start operation of groundwater bypass
地下水バイパス稼働開始

Completefreezing land-side
陸側遮水壁(海側)凍結完了
impermeable wall (sea side)

日平均量
Daily Inflow
AverageComplete closing sea-side
impermeable wall
海側遮水壁閉合完了

800

600

Amount of rainfall of Fukushima Daiichi
福島第一降雨量
Generated amount of
汚染水発生量contaminated water
Inflow amount of groundwater
建屋への地下水・雨水等流入量/storm sewage into the buildings

Approx.400

400

Approx.350

Approx.220

Approx.270

200

50

40 日
平
均
降
30 雨
量
（
福
島
20 第
一
）

Approx.490

Approx.470

mm/daymm/日(Fukushima Daiichi)

1000

Average Daily Rainfall

mm3/日
/day

10

Approx.200
Approx.140

0FY2014FY2015FY2016FY2017

Jul
7月

May
5月

Jan
1月

Mar
3月

Sep
9月

Nov
11月

May
5月

Jul
7月

Mar
3月

Jan
1月

Nov
11月

Jul
7月

Sep
9月

May
5月

Jan
1月

Mar
3月

Sep
9月

Nov
11月

May
5月

Jul
7月

Mar
3月

Jan
1月

Nov
11月

Jul
7月

Sep
9月

May
5月

0

FY2018(Courtesy of TEPCO)

Fig. 17 Transitions of the generated amount of contaminated water and the inflow amount ofgroundwater/storm sewage, etc. into the buildings21
In this way, the condition of contaminated water is being controlled. From now on, further
measures should be taken to complete the treatment of the stagnant water while addressing themeasures for contaminated water directed by the government's contaminated water management
committee22, and focusing on the problems that have not been clarified yet.
(1) Reduction of occurrence of stagnant water in buildings, including countermeasures against
storm sewage inflow
In order to reduce the generation of the contaminated water to 150 m3/day or less in 2020,
which is indicated as a milestone in the Roadmap, it has to reduce 70 m3 per day of thecontaminated water that is occurring in FY 2017 by approximately 220 m3 per day. In the future,
further effort to reduce the generation of the contaminated water should be continued by“Evaluation of Frozen Soil Wall and further Contaminated Water Management”, Committee on Contaminated
Water Treatment, issued on March 7, 2018
http://www.meti.go.jp/earthquake/nuclear/osensuitaisaku/committtee/osensuisyori/2018/pdf/020_s04_00.pdf
21 Handout 2 of “Overview of Decommissioning and Contaminated Water Treatment”, The Secretariat Meeting(57th) of Team meeting on Decommissioning and Contaminated Water Treatment, issued on September 7, 2018
http://www.meti.go.jp/earthquake/nuclear/decommissioning/committee/osensuitaisakuteam/2018/09/2-1.pdf
22 Quoted from Page 9 of “Evaluation of Frozen Soil Wall and further Contaminated Water Treatment”,
Committee on Contaminated Water Treatment, issued on March 7, 2018, as follows;
4. Further contaminated water treatment(1) The generated amount of contaminated water should be reduced even more by continuing multilayered
measurements including treatment of the stagnant water in the buildings, reduction of sub-drain water level, storm
sewage control, etc.
(2) During heavy rain such as typhoon, generation of contaminated water temporarily increased due to inflow ofstorm sewage from damaged part of roof of the buildings and unpaved area around the buildings. In response tothis, further measurements should be implemented systematically.
(3) Moreover, concerning the structures that water inflows from outside of frozen soil wall such as K-drainage, it ishighly likely that it becomes the water supply path to the inside of frozen soil wall. For this reason, survey should
be continued and necessary measures should be implemented.
20

71implementing the multilayered countermeasures such as stagnant water treatment in the buildings,
lowering the sub-drain water level, measures against inflow of storm sewage into the buildings.
Furthermore, the countermeasures against inflow of rainwater including the measures fordamaging the building roof by remote equipment are necessary.
(2) Operation in response to lowering of the water level in the building
For areas where the presence of oil on the surface of stagnant water was found, it is necessary
to recover the oil in order to avoid the performance deterioration of the contaminated water
processing system before exposure of the floor line. Also, after the floor line exposure, since floor
line sludge etc. dries, it is feared that dust is developed, so dust countermeasures will be
implemented.
Furthermore, installing the pump to the floor drain sump and transferring the residual water in
the isolated area will be needed for the floor line exposure. It is necessary to implement themeasures of dose reduction for the workers engaging the floor line exposure.
(3) Enhanced contaminated water management installations
It is essential to appropriately maintain and reinforce the established water level management
system and the purification equipment including improving reliability of sub-drain.
Study on contaminated water management based on the relationship with fueldebris retrieval
As described in Section 3.3.2.3, establishment of a PCV circulation cooling system as a system
for fuel debris retrieval is being studied, and substances generated from fuel debris containing αparticle is mixed to the circulation water. Therefore, it is necessary to properly remove alphaparticles in the PCV circulation cooling system. Since the groundwater flows into the buildings
continuously, a part of the water treated by the PCV circulation system will be sent to the existing
circulation water cooling and purification system in order to maintain the water balance. It is
presumed that the water is treated with equipment such as multi-nuclide removal equipment.
Therefore, it is necessary to monitor the alpha-nuclides concentration of the stagnant water in thereactor buildings and to set the conditions for receiving a part of water treated by the PCV circulation
cooling system in the existing circulation water cooling and purification system.
Summary of key Technical issues
The main technical subjects and plans described in this section are summarized as shown in
Fig. 18.

72

Fiscal Year

2018

2019

2020

2021

2022

On-site work
Technical studies for on-site construction"REMOVING " contamination sources

Treatment with purification Systems

Additional effective doses at thesite boundary .

Reducing the total amount of contraminated water to about 150m3 per day"REDIRECTING " fresh waterfrom contamination sources
Reducing the total amount ofcontraminated water to about
150m3 per day by rainwater of
average rainfall

Controle by groundwater by-pass, subdrainage and land-side impermeable walls

Paving the site, Removal of rubble from the roof, Waterproofing
To store all purified water by purificationequipments in welded type tank

"RETAINING" contaminated water

Switch frange type tank towelded type tank

To be responded properly according to
generation of contaminated water, etc.

from leakage

Securing Tank capacity
Ground stabilization, Maintenance of land-side impernmeable wall , Monitoring of groundwater/inside harbor

Treatment for the stagnant water

Reduction of radioactive materials instagnant water up to one tenthCompletion of stagnant water in the buildingsin the buildings
Lowering the level of groundwater/
stagnant water in the building

Maintaining emptied condition of turbinbuildings

Separation of the penetrations between
Units 1 and 2 and between Units 3 and 4

Determination of fuel debris retrieval methods
for the first implementing unit

Start of fuel debris retrieval at the first
implementing unit

Taking required actions in accordancewith the phase of fuel debris retrival

Contaminated water management
in consideration of fuel debris retrieval

Study for consistency between existing circulation cooling system
and PCV circulation cooling systems and monitoring method

Fig. 18 Technical issues and further plan on contaminated water management (Process Chart)73

3.4 Fuel removal from spent fuel pools
3.4.1

Sectoral target

The immediate objectives of fuel removal from the spent fuel pool are as follows.
(1) To start removing the fuel from the pool by around
1) FY 2023 for Unit 1
2) FY 2023 for Unit 23) Mid FY 2018 for Unit 323
under the rigorous risk assessment and management and taking measures for safety andsecurity including preventing the disperse of radioactive materials.
(2) By transferring the fuel stored in the Common Spent Fuel Storage Pool to the dry cask at theTemporary Cask Custody Area, the fuel in the spent fuel pools of Units 1 to 4 is to be stored in
the Common Spent Fuel Storage Pool appropriately.
(3) Based on the assessment of the long-term integrity and investigation for future treatment of theremoved fuel, the storage and future treatment methods of them will be fixed around 2020.
3.4.2

Sectoral strategies
Risk reduction concept and concrete plan for removing fuel from the pools

Fuel assemblies (fuel in the pool) stored in the spent fuel pool of the reactor buildings of Units
1 to 3, some of which were damaged by hydrogen explosion etc. are relatively high risk and highpriority ranked risk sources (see section 2.3). From the viewpoint of reducing the risk due toradioactive material, the fuel is kept in a solid state not to diffuse as contained in the cladding tube.
Furthermore, even though the decay heat of spent fuel is decreasing and coolant supply stops, thetime until the pool water evaporates and the fuel assemblies starts to be exposed become relatively
longer. Also, since it is a nuclear fuel material with a substantial amount of radioactive material(inventory), its hazard potential impact is relatively high. In addition, due to damage to the coolingfacility due to the accident and damage to the reactor building at Units 1 and 3, the same functions
as the conventional containment function and management function are not completely secured
and the safety management importance is also relatively high. For this reason, it is to be transferredto the Common Spent Fuel Storage Pool needs lower safety management than the pool of thereactor buildings in accordance with an appropriate and concrete work plan developed depending

23

In Unit 3, once started the trial operation for fuel handling equipment in March 2018, several troubles haveoccurred. After improved quality control issues which are recognized as a common factor, trial operation will berestarted, as well as investigation of the causes and countermeasures are taken. Making the safety as a toppriority, TEPCO intends to examine and review the start of fuel removal in the spent fuel pool which wasscheduled in November 2018, too. (Reference: “Trouble on FHM for Unit 3 of the Fukushima Daiichi NPS,TEPCO”, Handout 3-2 of Secretariat Meeting (57th)/Decommissioning and Contaminated Water Management
team meeting, released on September 6, 2018.
http://www.meti.go.jp/earthquake/nuclear/decommissioning/committee/osensuitaisakuteam/2018/09/3-2-7.pdf

74

on the situation of each unit, as soon as possible.
In addition, since the number of spent fuel assets held by each unit is different, there are slightdifferences in the degree of hazard potential impact. Depending on the rubble condition of the spent
fuel pool and on the operating floor, damage condition of the reactor building, units 1 and 3 areinferior to Unit 2 in terms of diffusion inhibition function and further it is judged that there is an impact
on retrieval since there is damaged fuel in Unit 1, so these differ in terms of safety management
importance (Fig. 5). Based on these points, it is necessary to respond by designing an appropriate
and concrete work plan for each unit.
In light of the decline in the decay heat of spent fuel, the evaluation of the increase in watertemperature at the time of stopping the cooling was reviewed by confirming the condition at thetime of stoppage of the spent fuel pool cyclic cooling system. This system was implemented inrecent years and it has been observed that excessive maintainability was lost compared to theconventional evaluation, and an evaluation formula close to actual temperature was obtained24.
According to this evaluation formula, it has been confirmed that even if the cooling of the spent fuelpool is stopped for about 1 to 2 months, the water temperature will only rise by about 10 to 20 °C
by considering natural heat dissipation.
Specific plan for fuel removal from the pools
Work schedule for fuel removal in Unit 1 to 3 was already specified in the Road map, andTEPCO has been proceeding with their tasks to meet this schedule.
Unit 1, the building cover had been removed, wind breaking fence had been installed,, andthen removal of some fallen rubbles generated by Hydrogen explosion were initiated andcountermeasures against falling into the spent fuel pool, etc. were initiated under taking measuresfor prevention of radioactive dust dispersion. The time to start removing fuel in the pool is targeted
for FY 2023.
Unit 2 is not affected by hydrogen explosion and maintains the soundness of the building, but
it is planned to dismantle the upper part of the reactor building in order to install fuel removal
installations. For this reason, Making of opening to have access to the operating floor and a
installing a front chamber to prevent dispersion of radioactive material have been implemented.
High dosage has been confirmed in the operating floor and surveillance of the operating floor andcountermeasures are under way. The time to start fuel removal from the pool is targeted for fiscal
2023.
In Unit 3, due to the influence of hydrogen explosion, instrument and things such as rubbles
dropped from upper part of the building and existing fuel handling machine were scattered on theoperating floor. Because it was also contaminated, the environmental improvements of operating
floor was performed and completed in June 2016. In February 2018, a cover for fuel removal fromthe pool was installed. Once started the trial operation for fuel handling equipment in March 2018,“Change of evaluation formula for water temperature in spent fuel pool, TEPCO” Handout 3-5 of Secretariat
Meeting (50th)/Decommissioning and Contaminated Water Management team meeting, released on February 1,
2018.
24

75

several troubles have occurred. TEPCO intends to examine and review the start of fuel removal in
the spent fuel pool which was scheduled in November 201823.
Unit 4 was under periodic inspection at the time of the accident, so all fuels were kept in the
spent fuel pool without suffering significant damage. Although the rubbles had fallen into the spent
fuel pool due to the influence of hydrogen explosion, after carefully installing a cover for fuelremoval in the pool, installation of a frame for supporting the crane was performed and the in-pool
fuel removal work started prior to other units and was completed in December 2014.
The fuel is also being stored in the pool of Units 5 and 6 reactor building under stable conditions
same as a normal nuclear power plant. The Roadmap states that the fuel of units 5 and 6 should
be properly stored in each spent fuel pool for the time being and will be removed them from eachpool at when such removal work becomes not to affect fuel removal work, etc. of Units 1 to 3. With
regard to some of the new fuels, there is a plan to begin exporting them to fuel processing
manufacturers. Also, with regard to the spent fuel, it should be taken out at an appropriate time inview of the capacity of the shared pool and dry cask temporary storage installation.
While conducting the above-mentioned safety measures properly, fuel removal operation fromthe pool should be steadily pursued. Although these efforts are first examined according to thecondition of each machine, as a whole, operation space (yard) adjustment and resource
management, etc. between interfering operations are necessary. Also, when removing fuelhandling equipment, it is necessary to reuse or treat waste. Based on the above and on thetechnical considerations described in Section 3.4.3, detailed build plans should be considered fromthe viewpoint of global optimization described in Section 3.6.
Storage plan for removed fuelFor removal of fuel from the pool, appropriate capacity should be reserved both in Common
Spent Fuel Storage pool and in Temporary Cask Custody Area in each unit. As of now, because
there is insufficient free capacity to transfer fuel in the pool of Units 1 to 3, 5 and 6 as shown in Fig.
19, it is necessary to advance the expansion of dry cask temporary storage installation etc. and tosecure the spare capacity by transferring the fuel out of the premise. Incidentally, in terms of design,
dry casks store fuels that satisfy conditions such as fuel type and cooling duration. Among the fuels
currently stored in the shared pool, those that satisfy the conditions and have a relatively longcooling period are to be stored.

76

Numbers of spent fuel etc.

For fuel removal from SFPs of unit1-3,unit5,6 (4999),
it is necessary
to ensure storage capacity of the Temporary Cask Custody Area,
to concider exporting fuel to off-site.

8000

Common Spent Pool,
Temporary Cask Custody Area
Available：1591

6000
unit1-3,unit5,6
All：4999

4000
storage6381

unit6：
1884

2000

storage
1757

New fuel：360（Plan）

Temporary Cask
Custody Area

Export to
off-site(nuclear fuelmanufacture)

unit3：566
unit2：615

0

unit5：
1542

unit1：392

units1-3

units5,6

Common Spent
Pool(Data provided by Material 3-8 of the 52nd Decommissioning and Contaminated Water Treatment Team
meeting/ The Secretariat Meeting, March 29, 2018 and Material 3- 2 of the 55th Decommissioning andContaminated Water Treatment Team meeting/The Secretariat Meeting, June 28, 2018)

Fig. 19 Storage situation of spent fuel etc.(as of Jun 28, 2018)
Decision on storage and future treatment methods
The spent fuel pools is storing many kinds of spent fuel such as flawless spent fuel, damaged
one before the accident by embrittlement of the cladding tube and falling of the fuel body, and that
may be damaged by the fallen rubbles into the spent fuel pool. Corrosion due to seawater sprayedand injected into the spent fuel pool of Units 2, 3, and 4 at the accident is another concern. It isnecessary to investigate that additional attention for the handling/storing of such fuels is requiredcompared with that for normal spent fuel.
Storage and future treatment methods for removed fuels from the pools will be decided around2020 based on the investigation results on long-term integrity assessment and treatment method
for them.
3.4.3

Technical issues and future actions for prompt strategies
Removal of fuel from the pools(1) Items common to each Unit
Operating floor dosage reduction measures are necessary in order to limit operation exposure
during the installation work of fuel removal installations (fuel removal cover, fuel handlingequipment, etc.), manned work during removal work, maintenance check of installations and soforth. Meanwhile, depending on the dosage reduction condition, it is necessary to reflect shielding
to instrument design related to removal and introduction of remote devices, etc., based on77experience such as decontamination in other units, it is necessary to determine the final operational
dosage at an appropriate time.
Also as stated in Section 3.4.2.2, there are a number of things to be prepared in parallel, suchas fuel removals in the pools of both Units 1 and 2 that start for FY 2023, preparation work on fueldebris retrieval at the time of removal of fuel in the pool of Unit 3. Therefore, it is necessary to adjust
yards (securing flow line) with interference works and resources management and to prepare adetailed construction plan. As described before, fuel of Unit No. 5 and 6 should be removed inappropriate time. In case such operations use shared pool, adjustment is required as such removal
work becomes not to affect fuel removal work, etc. of Units 1 to 3.
(2) Removing fuel from the pool of Unit 1
In Unit 1, a hydrogen explosion occurred at the upper part of the building when the accident
occurred, so the upper building collapsed and the roof slabs and other rubbles were scattered on
the operating floor. Also, existing fuel handling machine and overhead cranes are damaged andexist in a form that they cover over the spent fuel pool. Therefore, careful examination of timberingthat has sufficient structural strength was conducted in order to prevent falling as it is necessary tocure the spent fuel pool then remove them as soon as possible.
Also, from the viewpoint of the influence on the surrounding environment, it is necessary totake measures against dust dispersion at the time of removal of the rubbles. It is confirmed that thewell plug is shifted from the predetermined position, the dosage has been confirmed in dosage ofabout 200 mSv/h on the plug and it cannot be readily approached, there is a concern of skyshine(the effect that the radiation upward from the radiation source falls on the ground surface due todiffusion in the air and the dosage near the surface rises). Therefore, after taking safety measures
such as prevention of radioactive dust dispersion and dose monitoring, it is necessary to proceed
treatment for the same.
Regarding the 67 fuel bodies damaged by the cladding stored before the earthquake, although
the radioactive material concentration in pool water before the earthquake was also sufficiently low,
its influence is considered small, but appropriate response is required for handling when removal.
(3) Unit 2 fuel removal in poolIn the Roadmap, plan (Plan (1)) that installing a common container for both fuel removal andfuel debris retrieval, or Plan (Plan (2)) that a cover exclusively for fuel removal only on the operating
floor are selected for consideration at an appropriate time. NDF has shown the necessity of
adopting plan (1) which is superior in terms of earlier fuel debris retrieval time, lower worker

78

exposure dosage, radioactive material dispersion amount, and waste generation amount, etc.2526,
TEPCO is now being consideration. Study on Plans (1), sufficiently consideration of thepreconditions for the fuel debris retrieval plan for top access should be made such as shielding at
the time of operation, uploading, installations weight (structural integrity, etc.), there is an issue thatit requires suitable design for use over the fuel debris retrieval period. In either case, it is necessary
to judge the plan selection by an appropriate time based on the relation with the fuel debris retrieval
work and the removal timing of the fuel in the pool.
As a result of the past investigation, a high dosage of 880 mSv/h at most has been confirmedon the operating floor when dismantling the upper part of the reactor building, which is carried out
before the installation of the container or the cover. Since contamination by alpha-nuclides has also
been confirmed, it is necessary to select a safe dismantling method by remote operation after taking
measures such as prevention of dust dispersion.
As mentioned in Section 2.3.3.3, there is an exhaust gas stack of Unit 1/2 around Unit 2,
radioactive materials centered on cesium released by the accident may be adhering inside of thestack, and fractures and deformation were found at the junctions of the diagonal bracing of thesteel tower that supports the vent stack, therefore, the diffusion inhibition function is low and it isevaluated that the Safety Management is low. If this collapses, there is a risk of influencing the fuelremoval process in the pool, so it is planned to disassemble the upper part of the exhaust pipe bya remote equipment prior to fuel removal in the pool.
(4) Unit 3 fuel removal in pool
The fall of rubbles to the spent fuel pool has been confirmed and it is clear that there is adeformed fuel in the upper handle. Fuel in a spent fuel pool need to be taken out combined withremoval of the fallen rubble on top of the spent fuel. Accordingly, procedure for removing fuel needs
to be developed considering how to remove rubble, as well as it will be required to examine thecorresponding method in case fuel assemblies cannot be lifted due to rubbles at the time of fuelremoval.
Proper storage of removed fuelFor the deliberate transportation and storage of spent fuel and fresh fuel possessed by theentire site, a fuel transport plan need to be developed taking fuel stored in the Units 5/6 into account.
As well, the storage capacity need to be secured for additional facilities according to that plan.
In particular, when retrieval of fuel in the pool of Units 5, 6 precedes other units, it is necessary
to take measures such as adding dry provisional temporary storage installations by the appropriate“Evaluation and proposal for decision of fuel removal plan of Unit 1and 2 of the Fukushima Daiichi NPS, NDF”,
Handout 3-5 of Secretariat Meeting (11th)/Decommissioning and Contaminated Water Management teammeeting, released on October 30, 2014.
http://www.meti.go.jp/earthquake/nuclear/pdf/141030/141030_01_044.pdf
26 “Evaluation and proposal for dismantle and remodeling area of the upper part of the operating floor in thereactor building of Unit 2 of the Fukushima Daiichi NPS, NDF”, Handout 3-2 of Secretariat Meeting(24th)/Decommissioning and Contaminated Water Management team meeting, released on November 26, 2015.
25

79

time so as not to put pressure on the capacity for transferring unit 1, 2’s pool fuel to the common
spent pool. As mentioned earlier, some of the new fuel is planned to be moved to the fuelprocessing company and it is important to secure the capacity of the common spent pool by these
efforts.
Decision of future treatment and storage methods
As described in Section 3.4.2.4, if there are technical matters necessary for treating theremoval fuel as equivalent to ordinary spent fuel for long-term soundness assessment andprocessing of the extracted fuel, it is necessary to organize and identify the same.
In the decommissioning and contaminated water management project, R&D on the long-term
integrity of the fuel in the pool which has contacted with seawater or fallen rubble has indicated thatremoved fuel can be stored safely under the environment of Common Spent Fuel Storage Pool fora long period. Also, it was confirmed that the effect of seawater and scratches by fallen rubble on
the fuel is limited for integrity of stored fuel in the dry cask. In addition, proposals have also been
made regarding a fuel inspection method for dry storage.
Furthermore, another R&D showed a technical perspective that the impact to fuel by
earthquake disaster including contamination of chloride ions and concrete is very limited.
In the future, necessity of further study on possibility of its long term storage and processing
need to be evaluated based on the investigation results of the fuel taken out from Unit 3, which isexperienced the severe explosion caused by the accident and may be damaged by the fallen rubble.
Summary of key Technical issues
The major technical issues and future plans described in this section are summarized as shownin Fig. 20.

80

Fiscal year

2018

2019

2020

2021

2022

2023

2024

2025

On-site work
Technical studies for on-site construction

Unit 1

Rubble removal, etc.

Unit 2

Investigation of operating floor, etc.

Cover installation, etc.

Fuel removal

Dismantlement and modificationof upper part of building, etc.

Plan①

Container installation,

Plan②

Cover installation, etc

Fuel removal

Disassembly of Unit No.1 and No.2 exhaust pipe

Environment
around Unit 2

Measures to prevent sea contamination, etc.
Preparatory work
Unit 3
Fuel removal
Cover installation, etc

Toward construction to place
covers/containers
Selection of retrieval

Study onSelecting plan

Submit/Approval request to implementation plan
/Preparatory work

plan for Unit 2

Procurement of dry cask

Proper storage ofremoved fuelTransfer from Common Spent Fuel Storage pool to Temporary Cask Custody Area

Fuel removal from Unit 5 and 6

Implemented at the extent of rendering no impact
to the works on Unit 1 to 3
Adding dry cask temporary storage

Determination of future treatment and stroring
method of removed spent fuel (around FY2020)
Study forFuture treatment・Study on long-term integrity inconsideration of fuel of Unit 3

storing method
Study for method oftreatment and storing

Fig. 20 Technical issues and further plan on removing fuel from the spent fuel pool81

3.5 Other specific measures
3.5.1

Sustaining of reactor cold shutdown status

Currently, the plant conditions of Units 1, 2 and 3 are judged to be in a stable cold shut-down
status based on the internal PCV plant data monitored continuously since the accident, including
radiation dose, temperature, hydrogen concentration, pressure, and radioactive material
concentration, etc..27 As an example, the concentration of Xe-135, a continuously monitored shorthalf-life fission product, does not exceed the criticality level of 1 Bq/cm 3, showing no signs ofcriticality. Because fuel debris generates decay heat, it should be continued to monitor parametersin PCVs and seal nitrogen to reduce a risk of a hydrogen explosion, while maintaining andimproving reliability through maintenance and management, to maintain the stable state in the
future.
From the standpoint of lightening the contaminated water management load and reducing
criticality risks, cooling the fuel debris by air is one possible alternative in the future; however, it isconsidered there is not enough information on the plant interior for examining this alternative atpresent.28 Water supply to each unit was reduced in stages from 4.5 to 3.0 m 3/h after December
2016, and there was no significant temperature rise. Such data will be continuously appraised.
3.5.2

Radiation dose reduction and contamination expansion prevention all over the
power station
Prevention of sea contamination expansion

Immediately after the accident, radioactive materials flowed out to the port, as can be seenfrom the fact that high-concentration contaminated water in the turbine buildings flowed out throughthe underground trenches. To tackle this problem, emergency measures were taken, such as soil
improvement in contaminated areas, pumping up of underground water, and removal of highconcentration contaminated water in the trenches. Fundamental measures were also taken, suchas installation of the sea-side impermeable wall and covering of the seabed soil including
radioactive materials. As a result, the radioactive material concentration in the port was lowered toabout 10 Bq/L, although it rises slightly when rain falls (in the open ditch, Cs-137) (Fig.21). It islower than the limit concentrations specified by the dose regulations announcement for the area
outside the “warning area” (average three-month concentration: Cs-134, 60 Bq/L; Cs-137, 90 Bq/L).

27 For example, refer to the in-core condition data and diagrams (at the portal website for Decommissioning R&D

Information) at http://www.drd-portal.jp/assets/files/current_data_jp.pdf. The website posts the in-core condition
data released by TEPCO regarding the Fukushima Daiichi NPS (each unit's in-core temperature, dose rate,
feed water rates, Xe concentration, etc.).
28 The possibility of cooling fuel debris by air is conservatively evaluated by the Strategic Plan 2016, Appendix
4.17, “Overview of the analysis evaluation for air cooling in the reactor,” which states: “The results of the Unit 1
analysis that assumed a completely dry engineering method at the start of removal of fuel debris indicated thatthe maximum temperature of the fuel debris accumulated inside the RPV pedestal was approximately 480°C
and the maximum surface temperature of fuel debris, approximately 350°C.”82

In addition, the port entrance seawater radiation monitor indicated that the Cs-137 concentration in
the seawater had lowered to about 0.1 Bq/L29.
Sea-side impermeable wall closure completion (Oct. 26, 2015)
In front of shallow-draft wharf

On the south side of Units 1 to 4 water
intakes (in front of impermeable wall)(Courtesy of TEPCO)

Fig. 21 Radioactive material concentration in seawater in the port
Because of these various measures, the ingress of radioactive materials into the port water
has been strictly controlled; however, even today, the concentration of radioactive material in theport rises when it rains. It is therefore important to suppress the ingress of contaminated surface
water by way of drainages. The radioactive material concentration in drainage K, which runs near
the buildings, is higher than that in drainages A, B and C, and the concentration in drainage K
apparently rises when it rains. Thus, the radioactive materials eluted from the top of the buildings
or from the laid sand, the suspension flow of soil particles from contaminated soil or rubble on theTokyo Peil (T.P.) 8.5-meter bed, and eluted radioactive materials from the bed, are considered toreach the water in the port. Based on this supply mechanism, measures for reducing theconcentration of radioactive material in the drainages channel that flow into the port need to becontinued to further reduce the concentration, such as measures against rainwater ingress fromthe top of the buildings and paving around the buildings.
“Concentration of radioactive material in ground water and sea water at the east side of turbine buildings,
TEPCO”, Handout 3-6 of Secretariat Meeting (57th)/Decommissioning and Contaminated Water Management
team meeting, released on September 7, 2018.
http://www.meti.go.jp/earthquake/nuclear/decommissioning/committee/osensuitaisakuteam/2018/09/3-6-2.pdf
2983

Because soil near the port has contaminants in the shallow layers, the long-term environmental
impact evaluation involves technical challenges; namely, unlike the relocation analysis in the deep
underground disposal of high-level radioactive waste, knowledge of which been obtained, theinfluence of unsaturated layers and the effect of adsorption reaction rates need to be considered.
Regarding contamination of the soil in the vicinity of the harbor, assessing the impact on the ocean
from a long-term perspective and future environmental restoration, it is important to continuously
proceed with R&D such as to clarify the near-surface nuclide migration mechanism and refine theanalytical models30.
Management of gas and liquid waste
The Roadmap specifies that monitoring of gas and liquid waste should be continued and its
emission should be closely controlled to ensure that the concentration limits defined in the
Notification is strictly observed. With this view, proper countermeasures should be taken as theirconcentrations are made as low as possible based on a reasonable methods.
For the control of air release, centralized control by the monitoring system of exhaust stacks
cannot work because of the impact of the accident, the PCVs of Units 1 to 3 are continuously
monitored with gas control equipment (dust monitors and gas monitors), while the reactor buildingsof Units 1 to 4 are continuously monitored with equipment (dust monitors) installed in the upper
part of the building or nearby. The emission control target from Units 1 to 4 is set to a total Cs-134
and Cs-137 of 1.0×107 Bq/h, and it is ensured that the dust monitors near the boundaries of thepremises indicate values lower than the limit concentrations for the warning area, and the dose
evaluation is conducted once a month. When the emissions from Units 1 to 4 are 1.0×107 Bq/h, theeffective dose evaluation value near the boundaries of the site should be 3×10-2 mSv/year. The
effective dose evaluation based on the actual measurements of July 2018 was about 2.9×10-4
mSv/year, indicating that air release is controlled at a sufficiently low level.

Refer to “(6) Environmental fate studies of radioactive materials generated during decommissioning” includedin Attachment 12
30

84

Jul.
Sep.
Nov.
Jan.
Mar.
May
Jul.
Sep.
Nov.
Jan.
Mar.
May
Jul.
Sep.
Nov.
Jan.
Mar.
May
Jul.
Sep.
Nov.
Jan.
Mar.
May
Jul.
Sep.
Nov.
Jan.
Mar.
May
Jul.
Sep.
Nov.
Jan.
Mar.
May
Jul.
Sep.
Nov.
Jan.

Radiation exposure dose (mSv/year)

2011

2012

2013

2014

2015

2016

2017

(Reference)* Gas-phase concentration limits outside the surrounding monitored area
[Cs-134] 2×10-5 Bq/cm3,[Cs-137] 3×10-5 Bq/cm3* Monitoring post (MP1 to MP8) data
The monitoring posts (MP), which measure the dose rate in the air around the siteboundary indicated 0.404 to 1.757 μSV/h (Feb. 29 to Mar. 28, 2018; 10-mn. value)
To more accurately measure the fluctuations of the MP2–MP8 dose rates in theair, environmental improvement measures have been taken (cutting of surrounding
trees, removal of surface soil, and shield installation).

2018(Note) The dose evaluation previously differed between the Facility Management Plan and the monthly reports. The method was standardized in
September 2012.
Unit 4 was added to the evaluation targets in preparation for the removal of fuel in the spent fuel pools in November 2013.
In FY 2015, the evaluation method was changed to consider the values of continuously operating dust monitors. After this change, the results are
released the following month.

(Courtesy of TEPCO)

Fig. 22 Evaluation of annual radiation exposure dose at site boundaries due to radioactive
materials (CS-134 and Cs-137) emitted from Units 1 to 4 reactor buildings
Regarding drainage control of the water purified after being pumped up from the sub-drains
and underground water drains, four major radionuclides: Cs-134, Cs-137, all beta emitters (Sr-90),
and H-3, plus 44 other radionuclides have been selected for control based on the core inventory
and other related information, as significant target radionuclides for radiation exposure evaluation.


Before water drainage, it should be checked that the ratio to the concentration limit of thefour major radionuclides is 0.15 or lower.



A monthly analysis with a lowered detection limit concentration should indicate no significant
increase for the six indicators, which are the four major nuclides, Sr-90, and all alphaemitters.



It should be checked quarterly that the sum of the ratios to the concentration limits of all thetarget radionuclides is 0.21 or smaller.
As described above, drainage is strictly controlled. Radionuclides having a limit concentration

rate of 0.001 or smaller are removed from the checklist. After the application of April 2017, thenumber of target radionuclides was reduced to 41, including the four major ones.
The underground bypassed water pumped up on the mountain side before entering the reactor
buildings is temporarily stored in tanks, and is released after checking that it satisfies theoperational target, namely, the limit concentration rate of 0.22.
Dose reduction through the site decontamination
Of primary concern immediately after the accident was the workers’ exposure to radiation in
the Fukushima Daiichi NPS. Possible causes were fallout contamination across the entire site anddirect radiation from the plant. In March 2014, TEPCO formulated “Implementation Policy of Dose
Reduction on the Site of the Fukushima Daiichi NPS"31 and set target dose rates in stages for each
31 TEPCO, "Efforts for Dose Reduction inside the Fukushima Daiichi NPS ," presented as Handout 3-2 at the 19th85

area in the sites to reduce the dose. Decontamination progressed through the removal of highly
contaminated rubble, tree cutting, and removal of surface soil, and shielding was provided by suchmeans as facing. As a result of these efforts, the target dose equivalent rate of 5 μSv/h wasachieved at the end of FY 2015 in the area that lots of workers work other than the zone surrounding
Units 1 to 4 and the waste storage area.

In the vicinities of Units 1 to 4, as a result of thecontamination zone classification for more detailed contamination control such as by theidentification of radiation sources by means of directional monitoring, currently the area in which itis permitted to work with general work clothes has increased to 96% of the entire sites.
Improving the radioactivity environment and labor environment in the site is essential for
protecting workers’ health and safety (refer to Section 4.1). TEPCO should continuously maintainan average level of 5 µSv/h or under as specified in the Roadmap and should gradually reduce thetarget dose equivalent rate to ultimately reach the pre-accident level.
Reduction of environmental impact
By the end of FY 2015, the dose evaluation at the boundary of the site (“effective dose”),
encompassing additional emissions, attained the target dose of 1 mSv/year or less, as a result of
the previously described efforts to reduce the radiation dose and prevent the expansion ofcontamination, namely the purification of high-concentration contaminated water described in
Section 3.3, adequate Storage of solid waste specified in Section 3.2, and the entire risk inspection
including the inspections described below. Effective dose reduction efforts were continuously made,and at the end of FY 2017, it was 0.90 mSv/year (Fig. 23).

Meeting of The Committee on Supervision and Evaluation of the
http://www.nsr.go.jp/data/000051045.pdf.

86

Specified Nuclear Facilities (March 31, 2014)

Site boundary dose “evaluated value” (mSv/year)

0.96 mSv/year

0.92 mSv/year

0.90 mSv/year

Dose attributed to gas waste emissions

Dose attributed to liquid waste drainage

Dose attributed to sprayed water in thepremises

Direct radiation and mid-air reflected radiation fromindividual facilities

Fig. 23 Effective dose evaluation at site boundaries32
Besides high-concentration contaminated water, other water with radioactive materials has
been thoroughly inspected as a risk that may have an impact on areas outside the site, and efforts
such as contaminant removal and drainage cleaning are being made. Solid waste such as rubble
is reduced in volume where possible and is placed indoors to eliminate temporary outdoor storage
areas. The effective dose at the boundary of the site contributes not only to the direct doses fromradiation sources on the premises but also to the radiation emitted upward from sources, dispersedin the air, and radiated down to the surface soil. It is therefore important to adequately control andshield the radiation sources. Risk reduction outside the site of the Fukushima Daiichi NPS is
continued through such efforts, to retain the effective dose of under 1 mSv/year.
The representative individuals in the dose evaluation for public protection should be selectedfrom among those with relatively high exposure, in consideration of actual lifestyle habits andenvironmental conditions in the vicinity of the site. Intermediate radioactive waste storage facilities
are placed near the Fukushima Daiichi Power Station, and the representative individuals should be
realistically selected from this perspective. Setting control targets and evaluation conditions isdesirable. Discussion is desirable from the viewpoint of the possible integration of the 2007
Recommendations 33 of the International Commission on Radiological Protection (“ICRP”) into
32 TEPCO, "Dose Conditions in the Fukushima Daiichi NPS Buildings," presented as Handout 3-7 at the 53rd

Meeting of the Secretariat of the Team for Countermeasures for Decommissioning and Contaminated Water
Treatment (April 26, 2018).
http://www.meti.go.jp/earthquake/nuclear/decommissioning/committee/osensuitaisakuteam/2018/05/3-06-03.pdf
33 The International Commission on Radiological Protection (ICRP), 2007 Recommendations of the International
Commission on Radiological Protection, ICRP Publication 103, Annals of the ICRP, 37 (2-4), 2007.

87

Japanese laws.
Comprehensive risk review
TEPCO conducted a comprehensive inspection of risk sources that may affect the outside of
the site, and systematically established the following six categories of additional measures mainly
for leakage passages (liquids containing radioactive materials) and for work (dust), and announced
them in April 2015: (1) Research required; (2) Measures required; (3) Measures being taken; (4)Conditions being observed after implementation of measures; and (5) Measures not presently
required.
As a result, for the radiation sources requiring additional measures, specific measures werediscussed while taking priorities into consideration. They were reviewed as appropriate reflecting
environmental changes, and have been explained and announced at places such as the local
adjustment meeting for decommissioning and contaminated water management.
Also, the Nuclear Regulation Authority created a target map34 for reducing the mid-term riskof the Fukushima Daiichi NPS in February 2015. This Mid-term risk reduction target map, which
has been updated from time to time, is characterized by a risk reduction work process of about
three years emphasizing the presentation of residual risk. In regard to this Target map for reducing
the mid-term risk, TEPCO reported the current approach, issues and responding status according
to the further schedule as needed after May 2018.35 In the reports, as an example, a very large
floating structure (VLFS) can become a drifting object and damage nearby equipment if hit by atsunami, so the mid-term risk reduction map treats it as a risk that requires action in the event ofan earthquake or tsunami. TEPCO plans to move this VLFS to the open ditch of the water intake
channel for Units 1 to 4 and position it in contact with the sea bottom, to effectively use it as arevetment and shallow-draft wharf for the preparation of a newly built port yard.
Regard to the other countermeasures against earthquake and tsunami, building of tideembankment is being studied at the sea side of Unit 1 to 4 buildings36, TEPCO should continue
this kind of efforts for the purpose of avoiding increase of stagnant water as a result of water inflow
into the building due to tsunami and reducing damage of essential facilities in preparation of
Chishima Trench Tsunami which was identified as a highly imminent possibility.
It is important to continue to reduce and comprehensively grasp the risk sources like this, while
taking the positioning and priority in the entire decommissioning project into consideration for
implementing each measure.

Refer to “Target Map for Reducing the mid-term risk” in NRA web:
http://www.nsr.go.jp/disclosure/committee/kettei/04/04_01.html (Category: the Fukushima Daiichi NPS related
topics)35 TEPCO, "Measures Taken in Response to Risk Map Appraisal Instructions," presented as Handout 1 at the60th Meeting of The Committee on Supervision and Evaluation of the Specified Nuclear Facilities (May 18,2018). http://www.nsr.go.jp/data/000230852.pdf
36 TEPCO, “Progress on countermeasures against earthquake and tsunami”, presented as Handout 3 at the 63th
Meeting of The Committee on Supervision and Evaluation of the Specified Nuclear Facilities (September 14,2018). http://www.nsr.go.jp/data/000245445.pdf
34

88

3.5.3

Plan for decommissioning measures for nuclear reactor facilities

In the Roadmap, TEPCO should formulate the decommissioning plan of the Fukushima Daiichi
NPS in phase 3 after commencing fuel debris retrieval, aiming at the completion of
decommissioning in 30 to 40 years, depending on the progress of tasks such as fuel debris retrieval,and R&D. NDF should provide multifaceted and expert advice and guidance based on the progress
and forecast of the decommissioning, the situation of the reactor buildings, and the trends of R&D
with wisdom and knowledge from around the world.
As for Units 5 and 6, spent fuel removal should be carried out with reference to progress ofworks in Units 1 to 4, and then, decommissioning plans should be formulated for Unit 5 and 637.
TEPCO intends to use Units 5 and 6 as full-scale test mockups for remote-controlleddecontamination in the reactor buildings, internal PCV research, and fuel debris retrieval systems.
After their use as mockups, their effective use should be discussed among the relevant parties, tothe extent that later decommissioning work will not be influenced.
3.5.4

Concrete efforts toward securing safety
Efforts to ensure work safety

In FY 2017, the average number of workers per day was consistently 5,000 to 6,000, while thenumber of accidents (number of workers involved in an accident) was reduced from 20 in FY2016
to 11 in FY2017, thanks to thorough compliance with plant rules (22 standardized safety rules,
TBM-KY [tool-box meetings (talks) and kiken yochi or danger forecasting] training, and thorough5S methodology (sort, set in order, shine, standardize and sustain), introduction of accident cases
to other sections, reinforcement of safety control mechanisms and organizational supervision.
However, while the number of workers involved in an accident was reduced, the causes of manyof the accidents were similar (falling and stumbling), as indicated by an analysis on personnel,
equipment and management. To further reduce the number of accidents, a safety policy has been
established from the viewpoint of awareness raising, skills improvement and management, tocontinuously improve the labor environment.38
Similar to fuel debris retrieval operation, for the work plans that require workers to intervene ina high-dose environment and that complete remote-control system operation is difficult, it is“The Act for partial amendment of the Act on Regulations to Nuclear source materials, nuclear fuel materials
and reactor to reinforce safety measures for nuclear use” has been enacted and promulgated in April, 2017. It
was imposed that power reactor licensees must develop and publish its measurement in conjunction withdecommissioning of power reactor (Implementation policy for decommissioning) (will be enforced in October
2018). The nuclear facilities installed in the Fukushima Daiichi NPS have been designated as a specific nuclear
facility, its application to Unit 1 to 4 has been exempted by the government ordinance, because “Implementation
Plan for preservation on specific nuclear facility and protection of specific nuclear fuel material” (ImplementationPlan) was formulated and smooth transition to decommissioning are being performed. TEPCO is proceedingstudies for Implementation policy for decommissioning for Unit 5 and 6.
37

38 TEPCO, "FiY2017 Accident Occurrence Conditions in the Fukushima Daiichi NPS" and "FY 2018 Safety Action

Program," presented as Handout 3-7 at the 53rd Meeting of the Secretariat of the Team for Decommissioning
and Contaminated Water Treatment (April 26, 2018).
http://www.meti.go.jp/earthquake/nuclear/decommissioning/committee/osensuitaisakuteam/2018/05/3-07-03.pdf

89

important to evaluate the environment according to input resources from the viewpoint of"justification and optimization" aiming to ensure the safety of the working environment as well as
suppressing personal dose in appraising multifaceted approaches. Particularly when the work is“for the first time,” “changed,” or “for the first time in a long time,” it is essential to fully implement
work training using a mockup in order to design, implement and verify effective work procedures
and test methods. To prevent labor accidents, it is important to verify the procedures and at thesame time repeat preliminary training using a virtual reality (“VR”) system, and to map out a safe,
reliable work plan with clear hold points. Therefore, VR system reinforcement is effective, by
updating the method information and site conditions.
On-site work including the preliminary work for retrieval of fuel debris is expected to include:
decontamination in reactor buildings; investigation of PCV leakage positions; repair of PCV lower
and upper sections; construction of a network system; preparations for the installation of fuel debris
retrieval devices and systems; fuel debris retrieval operation; and fuel debris collecting, transferringand storing. Therefore, for each work stage, a detailed work plan should be prepared, and in thecase of an accident or problem, prevention measures should be taken based on the proper
preliminary risk evaluation and measures. It is necessary to examine how to handle an unexpected
accident: in the event of an accident or problem, a maintenance work area should be provided todeal with it quickly. In the future, it is necessary to review the dose reduction work in the reactor
buildings conducted so far and the PCV internal investigation, and use the information for
preliminary measures, such as preparations, planning, and training, for other work.
In line with the Roadmap, it is stated that measures for industrial accident prevention (joint
operation of the labor safety and health control system by TEPCO and the original contractors; risk
assessment by TEPCO, etc.; thorough communication and coordination between tasks;
experience-based education and training facilities for improvement of new employees’ risk
expectations capability, etc.) will be taken and reviewed continuously, medical preparedness will
be planned in anticipation of industrial accidents, and measures will be taken to reduce
occupational risk exposure as much as possible. It is important to ensure a perfect system of worksafety by continuing these efforts.
Efforts for facility safety
In the Fukushima Daiichi NPS where there are various kinds of work and safety facilities,
special attention to the safety of facilities is also essential. TEPCO has been preparing a complete
set of information, such as databases, maintenance plans for individual tools and devices, anddrawings.39 Measures for retaining and improving reliability will be taken in the future to ensure
long-term use, based on the maintenance plans for individual tools and devices, by steadily
performing periodic investigations, updating equipment at the proper timing, and making equipment
39 TEPCO, "Development of Database and Maintenance Plan for equipment, etc.," presented as Handout 1-6 at

the 23rd Decommissioning and Contaminated Water Treatment On-Site Coordination Meeting (July 27, 2015).
http://www.meti.go.jp/earthquake/nuclear/osensuitaisaku/committee/genchicyousei/2015/pdf/0727_01k.pdf

90permanent.
Particularly important safety assurance equipment, such as circulation system to cool fueldebris, nitrogen gas separation system, and PCV gas control system, are operated, inspected,
maintained, remote-controlled, monitored, and otherwise properly managed by TEPCO, based on
maintenance plans and other schedules. it is important to continue thorough measures that will betaken to prevent their important function from stopping, not just from the standpoint of equipment
servicing but from management and operation standpoints as well.
When installed new equipment or facility, it is critical to avoid defects on site as much as
possible, and steady quality assurance should be conducted through design reviews and testingand examinations.
Security enhancement
At decommissioned nuclear power plants, appropriate fuel removal is conducted, and, when
nuclear material protection concerns are removed, dismantling and other decommissioning work is
performed. Because a great quantity of nuclear fuel material is stored in the Fukushima Daiichi
NPS, it requires particular attention to be paid for its security same as a normal nuclear plant,measures to confirm the reliability of each individual, enhance nuclear security training, prevent
unauthorized intrusion into the sites, etc. are being implemented.
Along with these measures to be continued, it is necessary to implement appropriate measuresin operation to allow accepting visits of inspectors including locals to see the progress of thedecommissioning work as it is, at the Fukushima Daiichi NPS on site, as described in Section 7.2,
because showing the progress with their own eyes is extremely valuable for building the public’s
understanding of the decommissioning operation.

913.6 Comprehensive efforts for the decommissioning project of the Fukushima
Daiichi NPS
As mentioned in Chapter 2, in the decommissioning project of the Fukushima Daiichi NPS in
the future, it must be addressed more complex, uncertain challenges over the long term, such asfuel debris retrieval. Therefore, regarding the efforts of each field listed in Sections 3.1 to 3.5, amore proactive, systematic approach will be needed to solve the challenges rather than fulfilling
tasks in a reactive manner such as those mentioned above.
The Fukushima Daiichi NPS decommissioning involves many uncertain factors, and thedecommissioning plan, in addition to the information and expectations obtained at the time ofplanning, should be flexibly reviewed according to the various knowledge and conditions that will
emerge with the progress of the project. If the information and expectations at the time of planning
are significantly different from the information obtained with the progress of the project, it may bedifficult to perform the work as scheduled, and thus it is necessary to keep confronting such projectrisks when the project is proceeded. Moreover, the delay resulting from the appearance of such arisk may influence the ensuing tasks, and may further influence the progress of the entire project.
Furthermore, it is important to optimize resource allocation and scheduling while ensuring theconsistency and feasibility of the entire efforts to be promoted concurrently while having
interrelations, as shown in Fig.24. In order to perform full-scale fuel removal from spent fuel pooland fuel debris retrieval beginning in near future safely and smoothly, it is a growing issue tocoordinate the use of the site by ensuring area which can be stably used with sufficient room.
Namely, while this complicated and multilayer project should be controlled based on individualtasks of appropriate scale and control size, the entire decommissioning project should makeprogress comprehensively in consideration of the mutual relationship among the smaller projects.
The overall plan of the decommissioning project should be formulated and discussed while
presuming intermediate targets to the extent possible from a facility-wide mid-and-long-term
perspective.
Regard to the project risks to sustain stably for entire decommissioning project as this, it isessential to take appropriate responses from the proper implementation of progress management
and fund management of work, while learning from overseas precedent cases.

92

Present● Fuel debris retrieval

Determine fuel debris▼ retrieval policy

Establish the fuel debris retrieval▽ method for the first reactor

Discuss fuel debris retrieval method and develop necessary technologies
Start of fuel debris retrieval

Specific activities to prepare
for retrieval (preliminary
engineering, etc.)● Waste management▽ at the first implementing unit

Preparatory work & retrieval work

Consider consistency
with other activities

Waste from fueldebris retrieval

Store and manage solid waste
Secondary waste fromwater treatment, etc.

Rubble, etc.
on operating floor● Contaminated water management

Complete the treatment of▽ stagnant water in buildings

Operation methodof circulatory coolingsystem during fueldebris retrieval

Implement preventive and redundant measures under the basic principles of "remove,“"keep away" and "prevent leakage."

Prevent leakage of stagnant
water in reactor buildings● Fuel removal from spent fuel pools

Secure space for
concurrent work

Assess the status of the operating floor in eachreactor unit and discuss removal method▼ Completed removal from
Preparatory work
Removal work

Unit 4 (December 2014)▽ Start removal from Unit 3

Start removal from

▽ Units 1 and 2

Preparatory/removal work in each reactor unit

Fig. 24 Need for a mid- and-long-term viewpoint overlooking the entire activities934. Handling critical enablers for smooth operation of the project4.1 Actions toward improvement of working environment and conditions
The working environment of the Fukushima Daiichi NPS has been steadily improved. Needlessto say, this improvement is essential for protecting and respecting the rights of workers. For the
Fukushima Daiichi NPS decommissioning project, which will continue for a long period of time after
the end of the emergency status, a good working environment is the basis for ensuring safe andsteady implementation on a healthy foundation. To name a few typical efforts, TEPCO is improving
the working environment infrastructure by consolidating and removing existing rest stations andsetting up alternative rest stations, a cafeteria, convenience store, and shower booths, which are
infrastructure for improving the working environment. The company conducted a questionnaire
survey on the working environment for those working in the buildings, and is using the results toimprove the working environment.40
A variety of measures are taken to improve working safety and health, according to the“Guidelines for Safety and Health Control Measures for the TEPCO Fukushima Daiichi NPS”41
established in August 2015 by the Japanese Ministry of Health, Labour and Welfare. Reinforcementof the safety and health control system that integrates TEPCO and the original contractors is
implemented, by designating necessary positions such as the safety and health supervisor and by
establishing a health cooperation organization, while proceeding with steady efforts, such as risk
assessments regarding dangers and hazards caused by on-site materials, working behavior orother duties; implementation of measures based on the risk assessment results; implementation ofhealth education for new staff members; centralized radiation exposure dose control; reinforcementof entry/exit control of those working in the station facilities; and integration of effective radiation
exposure reduction measures into the project plan at the stage of placing purchase orders.
As for heat attack countermeasures, wet-bulb globe temperature (WBGT) indicators (whichindicate the three factors that have a great impact on the heat balance of human bodies: humidity,
radiation heat, and ambient temperature) have been installed at seven places, and thereinforcement of acclimation to heat, and checking of heat attack records and health checks (forearly discovery of health problems) are performed, to name a few. In FY 2017, there were no heatattacks that resulted in a worker having to leave, the same as in FY 2016, but the number of workers
who suffered from a heat attack increased slightly from four in FY 2016 to six in FY 2017.
Accordingly, the elimination of heat attack incidents has been determined as an important target
for the FY 2018 safety policies, and safety actions are being continued.38
40 TEPCO, "Results of Questionnaire for a Better Labor Environment (8th Meeting) and Future Improvement

Plans," presented as Handout 3-7 at the 49th Meeting of the Secretariat of the Team for Countermeasures for
Decommissioning and Contaminated Water Treatment (December 21, 2017).
http://www.meti.go.jp/earthquake/nuclear/decommissioning/committee/osensuitaisakuteam/2017/12/3-07-02.pdf
41 “Reinforcing safety and health control measures for the TEPCO Fukushima Daiichi NPS”, Press release data
announced by MHLW, Aug.26, 2015, http://www.mhlw.go.jp/stf/houdou/0000095466.html

94

As radiation exposure control, the monthly average radiation exposure dose in fiscal years
2014 to 2017 has been about 1 mSv. The FY 2017 monthly average radiation exposure dose wasabout 0.36 mSv, which is below the target radiation exposure dose of 1.7 mSv/month (Fig. ).
Monitoring the workers’ eye crystalline lens dose limit was started in FY 2018 independently basedon the statement of ICRP42, intending to reduce the dose from 150 to 50 mSv/year. Currently control
methods are examined to apply the limit dose of 100 mSv for five years from FY 2019.43 With theprogress of the project, work in a high-dose environment is expected to become more frequent. To
comply with the dose limits of individual workers, a limit should be designated that will be lower
than the legislative limit. When a worker’s exposure is at or above or may be above the limit, such
worker is to be excluded from engaging in radiation work or requested that the worker prepares adose control plan where the dose control methods will be determined in detail and work according
to the control plan to ensure that the dose does not exceed the legislative limit dose.

Exposure dose (monthly average dose)

TEPCO employee

Partner companies

Feb. 2018 average
0.38 mSv (provisional value)(Courtesy of TEPCO 21)

Fig. 25 Changes in each worker’s monthly exposure dose (Monthly average dose)
As measures for dose reduction in the premises as described in Subsection 3.5.2.3, high-dose
rubble removal, surface soil removal, and decontamination by facing and shielding have been
performed, and as a result, the target dose equivalent rate of 5 μSv/h was achieved by the end of
FY 2015 (excluding the vicinities of Units 1 to 4 and waste storage area). The area where workers
can wear ordinary uniforms or similar attire has extended to 96% of the entire premises. As it is nolonger necessary to wear protective masks and protective clothes, improvement has been achievedin terms of narrowed sight, longer moving time due to the load on the body, and the possibility of aheat attack. This has helped improve safety and health control measures not associated withradiation.

42 ICRP, ICRP Statement on Tissue Reactions and Early and Late Effects of Radiation in Normal Tissues andOrgans – Threshold Doses for Tissue Reactions in a Radiation Protection Context, ICRP Publication 118,
Annals of the ICRP, 41(1-2), (2012).
43 TEPCO, "Changes in Control Methods for Eye Crystalline Lens Equivalent Dose Limit Change," presented as
Handout 3-7 at the 52nd Meeting of the Secretariat of the Team for Countermeasures for Decommissioning andContaminated Water Treatment (March 29, 2018).
http://www.meti.go.jp/earthquake/nuclear/decommissioning/committee/osensuitaisakuteam/2018/04/3-07-02.pdf

95

In the future, while continuing the actions specified in the Roadmap for improvement in thework environment and conditions,44 it is needed to specifically design internal radiation exposure
prevention measures in preparation for the fuel debris retrieval that will generate alpha-dust, which
contributes significantly to radiation dose in the event of internal radiation exposure. Specifically, it
is necessary to control signs through appropriate measurement and monitoring, in addition tocontainment in specific areas using contamination expansion prevention methods, prompt
contamination removal, and wearing of protective gear. Especially, at the entrance and exit control(control of surface density), measures against the contamination of protective gear that have directcontact with the body are important, because adhering radioactive materials may indirectly
contaminate the skin and may become airborne particles that could cause internal radiation
exposure. TEPCO has an agreement with Japan Nuclear Fuel Limited regarding the dose
evaluation of internal radiation exposure due to α-nuclides derived from fuel debris, (internalradiation dose evaluation using bioassay analysis), and a system has been prepared to enable
quick action.

4.2 Concept of securing safety and promotion of cooperation
This is the first attempt in Japan at decommissioning damaged reactors. There are no
established or standardized regulations associated with the risk reduction of accident-hit nuclear
power facilities. As represented by fuel debris retrieval, the decommissioning of damaged reactors
involves reducing existing potential risks. The concept of safety is essentially different from theoperation at normal nuclear facilities. Namely, as described in Subsection 2.3.3, it is required toachieve relocation to a storage area with a higher control level as early as possible in considerationof temporal changes in the risk levels; safety assurance in and outside the premises may be
influenced if retrieval is started with insufficient preparations and measures. If excessive protective
measures are selected, retrieval may be difficult. Therefore, it is a challenge to realize balanced
44

The following is a summary from Chapter 5. (2), "Efforts for improving working environment and conditions" ofRoadmap:
A. Expansion in the areas where staff can wear ordinary uniforms, and improvement of the working environmentby the provision of a restaurant, larger rest stations, partner company buildings, etc., and the use of an appropriate
working plan.
B. Establishment and implementation of a working safety and health control system run integrally by TEPCO andthe original contractors, and the provision of worker safety with the participation of partner companies
C. Effective radiation exposure dose reduction measures, integrated management of radiation exposure dose
data by TEPCO, and radiation control by providing original contractors and subcontractors with information,
guidance and advice.
D. Improvement in working safety health level through implementation of risk assessment, the use of experiencebased education and training facilities, patrolling, and close coordination and communication among workers
engaged in different work.
E. Continued follow-up after physical checkup, heat attack prevention measures and prevention measures against
influenza and other infectious diseases, promotion of the use of health support counseling services, andimplementation of emergency medical system.
F. Working conditions knowledge expansion activities such as holding of seminars, and provision of proper
working conditions by continued counseling services.
G. Working contract with all workers and registration with the appropriate social welfare services, checking withthe original contractors where necessary.

96

risk reduction. Referring to safety efforts from overseas damaged reactors and legacy sites (refer
to Chapter 7), when conducting tasks for which no preceding experience exists in the world,principle of securing safety should be established according to the conditions and circumstancesof the Fukushima Daiichi NPS.
Because it needs to discuss specific tasks and regulations to be observed simultaneously,
NDF, TEPCO, the Agency for Natural Resources and Energy, and the organizations involved
should mutually cooperate, and positively communicate with the Nuclear Regulation Authority. It isalso necessary to make appropriate efforts, such as presenting policies and observation data
related to ensuring safety at an early stage. For the organizations involved, including regulation
authorities, sharing the goal of safe and smooth decommissioning and performing their own roles
regarding the reduction of on-site hazards and risks, it is important to make positive efforts while
sharing purposes and challenges, and for this safety assurance concept to be effective, it isdesirable to promote cooperation among the organizations involved.
As an example, one year after the accident at the United States Three Mile Island Nuclear
Power Plant Unit 2 (“TMI-2”), four parties, namely, General Public Utilities (GPU), the ElectricPower Research Institute (EPRI), the Nuclear Regulatory Commission (NRC) and the United States
Department of Energy (DOE), signed a cooperation agreement for fuel debris retrieval. Named
after the initials of the parties, GEND was a unified team that shared technological challenges
beyond the boundaries of the organizations, as various reports indicate.45,46 At the Sellafield site,
U.K., the G6 group was established to carry out decommissioning, with the stakeholders having
the common purpose of accelerating the reduction of hazards and risks at the site. G6 included
Sellafield Ltd., NDA, the Office for Nuclear Regulation (ONR), the Environment Agency (EA), UK
Government Investments (UKGI) and the Department for Business, Energy and Industrial Strategy(BEIS). The group has discussed the order of priority for risk reduction, effective resource use, andhow to realize the balance when risks may temporarily grow to reduce the risk in the long run. As
an achievement of this commitment, acceleration of the hazard and risk reduction has been
observed at the legacy facilities at the Sellafield site.

4.3 Enforced management
decommissioning

structureforsteadymid-and-long-term

To date, TEPCO has used the project management method to carry out large projects such as
the new or additional installation of a reactor: The company, as an example, prepared a structural
diagram of the Work Breakdown Structure (WBS), in which the project is disassembled and layered
based on units such as tasks. Regarding the decommissioning, the company has partially prepareda project structure for various work relating to decommissioning. However, line operation

45 EPRI, The Cleanup of Three Mile Island Unit 2, a Technical History: 1979 to 1990, EPRI,

NP-6931 (1990).

46 NRC, Three Mile Island Accident of 1979 Knowledge Management Digest, Recovery and Cleanup,

NUREG/KM-0001, Supplement 1 (2016).

97

management and project operation coexist in actual management, leading to slow decision-making,there is a challenge that it is not necessarily efficient as a whole. It is necessary to shift theorganization management to project-oriented management and reinforce the organization andfunctions to coordinate the projects.
Therefore, at TEPCO, in addition to the conventionally located Program Manager (PGM) and
Project Manager (PJM), a Program Management Office (PMO) responsible for project control has
been set up. NDF, as the supervisory and supporting organization, recently established a Program
Supervision & Support Office (PSO). These project operation sections in TEPCO are to conduct
the project safely and steadily under the supervision of safety and quality control division (Fig. 26).

TEPCO Holdings
東京電力ホールディングス

Nuclear Damage Compensation and Decommissioning
原子力損害賠償・廃炉等支援機構
Facilitation Corporation

Fukushima Daiichi D&D Engineering
福島第一廃炉推進カンパニー
Company(General director: CDO)
（総括責任者：CDO）

廃炉支援部門Decommissioning
Office(General director: President)（総括責任者：理事長）

PMO

Coordination between
組織間連携
organizations[PG1]【PG1】汚染水
Contaminated
対策PG
Water
Solution PG[PG2] PG for Fuel
【PG2】プール燃
Removal from the
料取り出しPG
Spent Fuel Pools【PG3】燃料デブ[PG3] Fuel Debris
Retrieval PG
リ取り出しPG【PG4】廃棄物[PG4] Waste
Management
対策PGPG[PG5] PG for【PG5】発電所
Improvement in Power
敷地・労働環境
Plant Site and Labor
Environment
改善PG

[PG6]
Program for【PG6】5/6号
Units 5 & 6
機対応PG

Nuclear Damage
賠償支援
Compensation
Office部門

PSO

Decommissioning
廃炉総括ＧManagement G

International
国際G
Affairs G

Technological
技術GG
Strategy* CDO, Chief Decommissioning Officer, in charge of decommissioning and contaminated water management; PG, program; and G, group.

Fig. 26 NDF and TEPCO Organizational System for Project Control
These projects that TEPCO will perform need to be positioned clearly in the Reserve Fund forthe Decommissioning System. TEPCO is the licensed decommissioning company in the Policy for
Preparation of Withdrawal Plan. Based on the planning of the project, NDF has recently shown
TEPCO the following: 1) the status quo of the decommissioning project and operational targets forthe mid-and-long-term perspective; and 2) the tasks to be implemented and their purposes, as wellas the targets and major tasks for the three years to come, as activities to be incorporated in theWithdrawal Plan. TEPCO is to submit the “Decommissioning Implementation Status,
Decommissioning Implementation Plan, and Other Items Specified by the Ordinances of the
Minister in Charge” every year via NDF to the Minister of Economy, Trade and Industry, who is incharge of decommissioning. TEPCO and NDF jointly prepared the Withdrawal Plan and obtained
approved from the Minister of Economy, Trade and Industry to withdraw from the fund. These
documents have been created according to the planning of the project (Table 4).

98

Table 4 Description of the Projects Included in the Withdrawal Plan
ProgramPurposeThree-year targets

Major tasks (for the coming three years)① Contaminatedwater solution
program・ Implementation of fundamental
measures against contaminatedwater
・ Removal of contaminants that
may have an impact on the
Fukushima Daiichi NPS
boundaries

・ The effective dose at the boundaries of thepremises should be maintained at a level below
1 mSv/year.
・ The total volume of contaminated water underan average rainfall amount should be controlled(should be no greater than 150 m3/day in total).
・ Tank capacities should be prepared according
to a plan.
・ While the difference in level between stagnant
water in the buildings and underground water isretained, the water level in the buildings should
be lowered (create a structure that prevents the
movement of stagnant water from the reactor
buildings to other buildings).

・ Subdrain reinforcement and land-side
impermeable wall related tasks
・ Measures against rainwater ingress into
buildings
・ Welded tank preparation, flanged tank removal,etc.
・ Transportation of stagnant water in thebuildings, installation of purification equipment,etc.
・ Purification treatment of stagnant water in thebuildings
・ Stabilized storage measures for secondary
waste from water treatment (sludge) resulting
from the decontamination system
・ Appraisal and implementation of other risk
reduction measures② Program for fuelremoval from thespent fuel pools

・ Spent fuel removal from Units 1
to 3
・ Storage in shared pool, etc. instable condition

・ Unit 1 fuel removal cover installation and related ・ Unit 1 operating floor rubble removal
work can be started in FY 2021.
・ Unit 2 investigation inside the reactor building・ Unit 2 buildings upper part dismantling andoperating floor and building upper part
related work should be able to be completed in
dismantling
FY 2022.
・ Unit 3 fuel removal cover installation, fuel・ Fuel removal from Unit 3 spent fuel pool isremoval and stable storage
completed.
・ Dismantling of Units 1 and 3 shared exhaust
・ Environment improvement in the vicinity of Unit
stack upper part2 is completed.

③ Fuel debris
retrieval program・ Units 1 to 3 fuel debris retrieval
method system construction
・ Units 1 to 3 fuel debris retrieval,containment, transportation andstorage・ Determination of Unit 1 fuel debris retrieval
method and implementation of the engineeringthat enables it. In addition, preparations for thestart of retrieval should be making progress.

④ Wastemanagement
program・ Establishment of waste storage
management plan andappropriate waste storage

・ Solid waste is safely and reasonably stored and ・ Waste storage and volume reduction treatment
controlled, and the required storage capacity
equipment (incinerators, etc.)
should be provided.
・ Establishment of mid- and-long-term storage・ Substances that otherwise would become waste
plan (including the storage of adsorption towers
should be recycled, reused and treated forand concentrated liquid waste and slurry)steady volume reduction.
・ Technology development, mainly by TEPCO HD⑤ Program for
improvement in
power station
premises andworkingenvironment

・ Use based on scheduled
premises utilization schedule
・ Preparation of working
environment infrastructure

・ According to the premises utilization plan,
・ Retention of buildings in good ordervarious measures related to decommissioning
・ Provision of decommissioning infrastructuresare steadily carried out.
・ Provision of buildings and rest facilities
・ To retain and improve the working environment
in the station, infrastructure should be prepared.

⑥ Program for Units ・ Units 5 and 6 equipment retention ・ Spent fuels are stably kept cooled.
5 and 6and control
・ Decommissioning plan planningand implementation

・ Appraisal of fuel debris retrieval method foreach unit (steady implementation ofengineering, etc.)
・ Detailed condition investigation inside thecontainment vessel・ Preparations related to fuel debris retrieval・ Technology development, mainly by TEPCO HD

・ Retention of cooling equipment
・ Decommissioning planning and proposal
creation・ Relocation of mega-float

To make that the mechanism of the described project management functions effectively, it isnecessary to enable TEPCO to control its engineering task and make it function effectively, basedon a deep understanding of the technical work of individual tasks under appropriate operation
control. This means that the power utility company, whose major operation is equipment control,
must have a wide range of experience and use technological expertise that such a company
conventionally does not have. TEPCO should commit itself to both technological areas and humanresources/organizational control, in collaboration with not only other power companies, but also
manufacturers, general contractors, and engineering companies, while utilizing the expertise of
external specialists including international ones. Through such collaboration, the company should
accumulate technology and know-how including engineering capacity, and in collaboration with
suppliers including related companies and partner companies, reinforce the supply chain
management including cost management.

4.4 Developing and securing human resources99

4.4.1Developing and securing operators and engineers

The decommissioning project of the Fukushima Daiichi NPS requires completely different skills
from the technologies related to the construction and operation of NPSs that TEPCO has
accumulated to date as well. Furthermore, the technical challenges posed by the existence ofenormous amounts of groundwater and others, the uncertainty of the internal site, and highradiation due to α- nuclides and fission products, etc. compose a very different condition from thedecommissioning of normal reactors.
Therefore, the Decommissioning R&D Partnership Council has prepared a draft technology
map to grasp the overview of necessary core technologies and personnel required for thedecommissioning of the Fukushima Daiichi NPS. (Attachment 10) On this map, some humanresource fields seems to be able to obtain by training people in the technological talent pools of theexisting nuclear industry such as analysis engineers that are currently known for its necessity. Also,
there are fields where technological personnel should be obtained from a wider variety of fields not
limited to the existing nuclear industry or should be developed in a planned manner. It is expected
for related institutions to utilize this map for developing and securing human resources in ways
such as (1) to clearly grasp the overall picture of the technologies required for the decommissioningof the Fukushima Daiichi NPS; (2) to clearly understand the strength of their own personnel; (3) toprepare training programs; and (4) to plan recruiting strategy aiming wider areas not limited to theconventional nuclear industry.
In handling a complex, large-scale project like the decommissioning of the Fukushima Daiichi
NPS that involves many related factors, it requires special personnel who are not only able todemonstrate expertise in their assigned area, but also have the ability to manage projects from acomprehensive perspective, including the consideration of relationship between the projects basedon the overview of the entire decommissioning process. A subgroup of Professional Engineer, in
the Council for Science and Technology proposed that an elective subject “Nuclear Reactor
Systems and Facilities” in the second stage of the qualifying exam of Professional Engineers (PE)for the fields of nuclear power/radiation should include “Reactor Decommissioning including
Handling after Severe Accident”, and another elective subject “Nuclear Fuel Cycle and Radioactive
WasteManagement

andDisposal“

shouldinclude“Decommissioning

andNuclear

Fuel/Radioactive Waste Management and Disposal after Severe Accident”. Following this,qualifications of the major positions such as PEs are asked in the public offerings for the Project ofDecommissioning and Contaminated Water Management beginning from March 2017. It is
expected that companies will continue to make effort to expand the employees’ capabilities byencouraging them to acquire related qualifications such as PE, Chief Engineer of Reactors, orRadiation Protection Supervisor, etc.
The forecast of the number of workers expected to be required for the upcoming three years
is shown in the Roadmap, however, as newer tasks will be implemented such as fuel debris retrieval
required number of workers may change. Therefore, it is necessary to grasp the amount of required100

personnel of each field on a timeline basis by imaging a long-term prospect for the project ofdecommissioning of the Fukuoka Daiichi NPS. According to this, the required number of personnel
having sufficient skills should be steadily and systematically developed and recruited. Such longterm prospect is to be examined through the entire decommissioning plan described in Section 3.6.
4.4.2

Fostering the next generation to handle the decommissioning of the Fukushima
Daiichi NPS

To carry out R&D activities sustainably for a long period, industry, academia, and thegovernment related to nuclear power should steadily continue efforts for human resources such as
developing and securing of future researchers and engineers.
Specifically, the industry and educational institutions should cooperatively continue theactivities to nurture students' understanding of the nuclear power industry and show theattractiveness of the industry. They should also convey students that the decommissioning of the
Fukushima Daiichi NPS is an extreme technological challenge unprecedented in the world, as wellas building and showing a variety of career paths for researchers and engineers to take part.
To this end, the Ministry of Education, Culture, Sports, Science and Technology (MEXT)’s theCenter of the World Intelligence Project for Nuclear S&T and Human Resource Development(“World Intelligence Project”) has been proactively facilitating HRD via research activities among
universities and colleges. Not only the courses related to decommissioning prepared under thisprogram are counted in the units required for graduation, but also the program provides students
with the opportunity to study decommissioning through their graduation research and to foresee
future career paths. “Conference for R&D Initiative on Nuclear Decommissioning Technology by
the Next Generation (NDEC),” intended for students, and “Creative Robot Contest for
Decommissioning,” intended for college students are held. Students exchange opinions withresearchers and engineers involved in the Project of Decommissioning of the Fukushima Daiichi
NPS, and those who with excellent results receive awards in the events. In addition, for humanresources retention and expansion through the entire industry, various initiatives such as “Nuclear
Dojo” providing nuclear education programs in collaboration among 16 universities in Japan and“Nuclear Facilities Tour for College/High School Students towards Brighter Future” have been
carried out.
For the decommissioning of the Fukushima Daiichi NPS is difficult operation that has neverbeen experienced, accordingly, knowledge in a variety of field is required. It is essential to cultivate
human resource through R&D activities including wide range of areas not only nuclear field but
mechanical, chemical, civil or material engineering, etc. Furthermore, in long-term and large-scale
projects such as the decommissioning of the Fukushima Daiichi NPS, it is essential to develop corepersonnel for research and development who can perform scientific and engineering investigationfrom an academic perspective and personnel with a panoramic perspective (system integrators)
who can integrate individual technology seeds into a system with practical functionality. This activity
is being implemented by R&D themes described next in Chapter 5.

101

We should not think these activities for fostering next-generation decommissioning personnel
just as an effort to supply researchers and engineers. Approaching to a wider layer of people should
be considered to enable the large cyclic rotation of personnel, from a long-term viewpoint. The
previously described programs for developing decommissioning R&D personnel has brought insolid results, such as graduates joined the nuclear power industry engaged in decommissioning,
while many students decided to study further at graduate schools, joined regulating authorities, andwere admitted to local municipalities. It is expected that their experience in decommissioning as a
student and a wide range of experience in society may provide a new perspective to the project ofDecommissioning of the Fukushima Daiichi NPS and drives the project forward.

102

5. R&D initiatives
5.1 Basic policy for R&D
5.1.1 Basic policy
Decommissioning of the Fukushima Daiichi NPS is technically extremely difficult and involves
many challenging issues never dealt with before. In order to implement various measures steadilyin accordance with the Roadmap, development of new technologies to solve such issues and ofreliable technologies aiming at on-site application is indispensable. Therefore, R&D projects aiming
at practical application, construction of research centers and facilities by JAEA, basic and genericresearch as well as applied research at researching institutions including universities have been
conducted under the governmental subsidized or commissioned project and facility buildings of
R&D center. At the same time, TEPCO has been working by itself.
NDF has determined the R&D duties implementation policy 47 based on the NDF Act.
According to this policy, NDF has been gathering expertise of R&D institutions and managing avariety of R&D project, such as clarifying the inner reactor conditions, feasibility study of retrieval
method and so on, that support steady implementation of the Roadmap.
Since the policy for fuel debris retrieval has been determined, such R&D processes are toenter a new phase in the future soon ahead. As detailed processes towards the decommissioning
becomes clearer, the role of each R&D player should also be more clarified. Here, it is necessary
to allocate the division of roles between the government and the enterprise in an appropriatemanner in order to implement R&D results on the decommissioning site steadily. To back up the
implementation of the mid-and-long-term decommissioning project, it will be further expected thatthe government and the relevant researching institutions should establish centers of basic
research/research infrastructure based on a mid-and-long-term perspective. Research institutes
are expected to enhance the technologies required for decommissioning through considerations
on the status of the project and the fundamental R&D activities according to scientific andtechnological issues (needs) regarding the decommissioning.
For this challenge entering into an the pathless field, it is important to further enhance many
approaches from different angles through the collective effort of Japan such as realizing effective
R&D, cooperation between the related organizations including those overseas, utilization ofresearch facilities, and human resources development.
5.1.2 Entire perspective of R&D
Various issues involving difficult R&D elements exist in the decommissioning of the Fukushima
Daiichi NPS. R&D activities for solving these problems are being conducted by variety of industrialacademic-governmental institutions engaged with the Fukushima Daiichi NPS decommissioning
47

The R&D Duties Implementation Policy, http://www.dd.ndf.go.jp/jp/committee/policy/index.html

103

R&D projects through the areas of basic/fundamental research, applied research, anddevelopment/utilization (Fig. 27 and 28). To organically link these activities and efficiently solve theon-site problems through the R&D, it is important to understand and share of the information of
R&D activities conducted by the related organizations and to cooperate and coordinate between
the decommissioning site and research site. Therefore, NDF regularly holds meetings of theDecommissioning R&D Partnership Council according to the decisions of the Team forCountermeasures for decommissioning and contaminated water management. The members of
the Council are NDF, JAEA, TEPCO, IRID, plant manufactures, the experts concerned and theministries concerned. The main task of the Council is to tackle with the issues such as sharing
information on R&D needs and seeds, coordinating R&D based on the needs of decommissioning
work, and promoting cooperation in R&D and human resource development in manners ofcollaborating with related organizations, gathering international expertise and promoting
comprehensively and systematically. Through these activities, NDF is working on overall
optimization to promote the R&D activities of related organizations effectively and efficiently.

104

105

Administering

National R&D Agency
Plant makers
Electric power Companies, etc.

IRID

※８
※９

※７

※３
※４
※５
※６

※１
※２

※８

Commission

Information
sharing※９

Sector Fukushima R&D

Exchange of
researchneeds/seeds

Participation

Project
Management
Operation

Platform of Basic
Research for
Decommissioning

Fukushima Environmental
Safety Center

Okuma Analysis andResearch Center

Request for study
and Support

TEPCO（Member）
NDF,JAEA,TEPCO,
IRID, Plant Makers,
Related Experts, METI,
MEXT

Naraha Remote Technology
Development Center

CLADS

JAEA

Decommissioning
R&D PartnershipCouncil

MEXT

Adopteduniversities, etc .

Acceleration
Program for Reactor
Decommissioning

Domestic universities,
Research institutes, etc.

Adopteduniversities, etc .

Human Resource
Development andResearch Program for
Decommissioning ofFukushima Daiichi NPS

Foreign
Universities/
Joint
ResearchResearch
Institutes

Foreign
Governmental
Agencies

Cooperation

Administering

Commissioned Research Institutes

Contract/
collaborative
Research

Project
Management
OfficeAssignment

Cooperative
relationship※7

Funding

Steering committee

Center of World Intelligence Project for Nuclear S&T
and Human Resource Development

NDF

Management
Expenses Grant

SupportDecommissioning
Strategy Committee

Sharing/planning
R&D information

SupportSupportFig. 27. Whole picture of R&D structure of the decommissioning of the Fukushima Daiichi NPS (As of FY 2018)

Decommissioning R&D Partnership was established in NDF according to a decision of Team for Countermeasures for Decommissioning and Contaminated Water Treatment.
Bold solid arrow mean supply of R&D and management expenses (except for facilities expense).Light solid arrow means cooperati on. Dotted line arrow means participation in
Decommissioning R&D Partnership Council.
Some institutions including JAEA are at the multiple positions.
Each institution has their own cooperation with foreign institution based on MOU, etc.
R&D activities by other institutions including Central Research Institute of Electric Power Industry are abbreviated.
Among the promotion projects of nuclear energy science technology/human resource development that require collective know -how, the adopted projects up to 2017
were entrusted by MEXT to the trustee, which are not shown in this figure.
Subsidies for the promotion projects of nuclear energy science technology/human resource development that require collective know-how are delivered to JAEA, but expressed
as being delivered to CLADS for the sake of clarity here.
For Decommissioning and Contaminated water treatment subsidy project, NDF drafts the next R&D plan and METI determines it based on the policies on Roadmap and the Strategic Plan and R&D progress.
NDF participates in the steering committee of human resource development promotion project of Nuclear Science Technology that gathers wisdom and intelligence, as a constituent.

Domestic / Foreign Enterprises

Othermakers, etc.
(including
foreignenterprises)

Subsidized Project Operating Entities

Grant

Project Management
OfficeAssignment

Fund Establishment Organization
Fund for Decommissioning andContaminated Water Management

Funding

Project of Decommissioning andContaminated Water Management

METI

Needs ofthe site

R&D
organizations,e.g. IRID

Decommissioningof 1F
Development andpractical
application

Tokyo Electric Power
Company Holdings, Inc.

Applied research
Basic researchorganizations,e.g., universities

Basic research

Collaborative Laboratory for
Advanced Decommissioning Science,
Japan Atomic Energy Agency

Fundamental research (development of common
fundamental technology, improvement of infrastructure
and management of basic data)

Fig. 28 A conceptual picture of the division of roles among main R&D institutions for thedecommissioning of the Fukushima Daiichi NPS

5.2 R&D of decommissioning required for on-site work/engineering
There are two types of R&D activities towards their practical use for the accomplishment of thedecommissioning of the Fukushima Daiichi NPS: the engineering activities including technologicaldevelopment implemented by TEPCO and the Project of Decommissioning and Contaminated
Water Management carried out by selected subsidiary companies (Attachment 11).
In both cases, aiming at the overall optimization of the decommissioning project as a whole as
described in Section 3.6, project-oriented R&D management will be required. R&D tasks that areraised as necessary through the engineering considerations conducted by TEPCO should beshared with NDF. An appropriate implementing institution takes over the implementation of such
R&D tasks in a timely and accurate manner under the coordination of NDF. Then, the result of the
R&D will be accurately provided to the site in a timely manner. As NDF was designated to oversee(manage and supervise) the project management of TEPCO under the Decommissioning Reserved
Fund System described in Section 4.3, it is important to realize effective R&D management throughthe system.
Specifically, for the time being, R&D tasks are going to be identified according to the progressof the preliminary engineering implemented by TEPCO as described in Section 3.1.2, and then, it
is presumable that the timing to implement such R&D tasks will be determined throughconsiderations through the project management. To realize an R&D management system undersuch project-based schedule, appropriate information sharing on the situation of R&D
implementation by TEPCO and on the R&D tasks of which the necessity became clear byengineering considerations should be carried out under the project management system enforced
jointly by NDF and TEPCO as described in Section 4.3. To this end, TEPCO and NDF should
marshal the R&D tasks currently in progress, and R&D tasks required in the future under the project
management system in relation to the engineering schedule, that is, clarifying when the problem
must be solved and which project needs it.

106

In this regard, even including implementing the tasks in the Government-led Project ofDecommissioning and Contaminated Water Management, implementation of R&D tasks should beconsidered in line with the basic concept of an appropriate division of roles between the government
and TEPCO according to the substances of the tasks. Specifically, R&Ds that are more difficult areconsidered to need supports by the government.
As actual processes for the decommissioning become apparent according to the progress ofthe project management, R&D tasks that are intended to apply to the site become clearer.
Therefore, it will be required for TEPCO to make efforts to raise the ratio of technologicaldevelopment to improve the safety and efficiency of the decommissioning work. In addition, taking
into consideration the fact that the United States Department of Energy (DOE) has received arecommendation from the task force to elevate the science and technology budget among thebudgets for decommissioning of the legacy sites48 and the DOE has declared a policy to increase
the budget49. For the Fukushima Daiichi NPS as well, it is important to steadily figure out andimplement the required technological development under the Reserve Fund system.

5.3 Enhancement of basic study and R&D infrastructure for the success of thedecommissioning project
For decommissioning of the Fukushima Daiichi NPS, which is a large-scale project involving
opacity and uncertainty, predicting what would be the critical technical challenges in the future is avery difficult task. Based on the available information at the moment, precedents and experts’knowledge and experience, we have to extract the R&D tasks which might become critical in the
future. At the same time, we also have to maintain and expand the infrastructures of personnel,
organization and facilities on a long-term basis so that forthcoming cross-sectional technologicalchallenges can be managed. These are the major elements to secure smooth execution of the
project.
5.3.1 Essential R&D Themes and its strategic promotion derived from the needs
To facilitate the decommissioning project of the Fukushima Daiichi NPS in a safely steady andeffective manner, it is essential to develop mid-and-long-term R&D strategies including scientificand technological investigation based on understandings of the principles and the theories. For thispurpose, NDF has built a task force on research collaboration and specified the 6 Essential R&D
Themes that should be preferentially and strategically targeted based on the discussions at theDecommissioning R&D Partnership Council. Moreover, the Platform of Basic Research for
Decommissioning 50 was investigated and compiled R&D strategies for the Themes, based on
Secretary of Energy Advisory Board, “Report of the Task Force on Technology Development for Environmental
Management”, U.S. Department of Energy, (2014).
49 U.S. Department of Energy, “Departmental Response: SEAB Task Force Recommendations on Technology
Development for Environmental Management”, (2015).
50 Promotion council of basic and generic research jointly managed by JAEA/CLADS and MEXT the Center of the48

107which the R&D to be executed immediately was initiated in FY2017.
Furthermore, on the occasion of transition of the World Intelligence Project into a subsidy
program described in the following section, the Ministry of Education, Culture, Sports, Science andTechnology (MEXT) showed NDF their intention to discuss how to proceed with the future R&D of
the World Intelligence Project. This included selection of the theme for the call for proposal
considering the Essential R&D Themes from the perspective of propulsion of basic and genericresearch based thoroughly on necessity. Therefore, NDF has compiled the Basic Future Directionof the six Essential R&D Themes drawing on the discussion of the subcommittee meetings of the
Platform and presented it to MEXT (Attachment 12). The Basic Future Direction includes
background of the issues, awareness of the problems by the consumer-side and a conceivable
image of the research. It is adopted in the call for proposal of the World Intelligence Project started
in May 2018 by JAEA.
5.3.2 Construction of basic research centers and R&D infrastructure for mid-and-long-term
prospects
In order to make the long-term decommissioning project of the Fukushima Daiichi NPS
proceed steadier in technical aspects, it is essential to work on developing R&D infrastructure andaccumulating technological knowledge, developing generic technologies and collecting basic data,
including the Essential R&D Themes described in the preceding section, building up research
centers, facilities and equipment, and human resource development. Decommissioning of the
Fukushima Daiichi NPS is an opportunity for trial of state-of-art science and technology and suchthe accumulation of such activity is expected to become a source of innovation.
The building for International Research Collaboration of the Collaborative Laboratories for
Advanced Decommissioning Science of JAEA (JAEA/CLADS) has opened in April 2017 in
Tomioka-machi, Fukushima prefecture as a location for universities, research institutes andindustries inside and outside the country to create a network to integrally promote R&D and humanresources development. TEPCO and some universities have moved into the building to undertake
research and so on. It is expected that, in the future, interaction between various talents inuniversities, research institutes, industries inside and outside the country will make JAEA/CLADS
a central organization acting as a hub of such activities. Following this idea, the Ministry of
Education, Culture, Sports, Science and Technology (MEXT) reformed the Center of WorldIntelligence Project for Nuclear S&T and Human Resource Development into a subsidy project
intended for JAEA according to a proposition by NDF. The project is currently implemented by thesystem centered on JAEA/CLAD from the newly adopted tasks in FY2018.
It is also important to develop research infrastructure of hardware. JAEA put the Naraha
Remote Technology Development Center in full service in April 2016 in Naraha-machi, Fukushima
Prefecture. It is a facility with a variety of equipment for development and experiment of remotelyWorld Intelligence Project: Decommissioning Research and Human Resource Development Promotion Program
selection agency. See https://fukushima.jaea.go.jp/initiatives/cat05/haishi05.html.

108controlled devices and installations. Specifically, prior to sending the devices into the harshenvironment inaccessible for people, testing in a real-scale mockup is indispensable for
performance verification as well as for training and establishment of the operating procedure. It is
desirable that the enterprises and so on utilize the mockup actively.
Moreover, Fukushima Prefectural Centre for Environmental Creation where Fukushima
Prefecture, JAEA, and the National Institute for Environmental Studies have their offices (Miharumachi, Fukushima Prefecture) opened in July 2016, serving as a comprehensive center ofcollaboration of the three agencies. In addition, JAEA opened the facility management building of
the Okuma Analysis and Research Center (radioactive material analysis and research facility)(Okuma-machi, Fukushima Prefecture) in March 2018 as well. The 1st Building that is to handle
low-dose samples such as rubble and the 2nd Building that is to handle high radiation samples suchas fuel debris are respectively under construction and designed.
As seen above, R&D infrastructures related to decommissioning, contaminated water
management and environmental decontamination measures are being set up mainly in Fukushima
Prefecture. These research facilities are formulating a global center of decommissioning R&D
aiming at the mid-and-long-term vision.

109

6. Enhancement of international cooperation
6.1 Significance of international cooperation
In recent years, nuclear reactors and nuclear fuel cycle facilities, constructed in the dawn ofnuclear power, are coming to the end of operating life and these decommissioning have alreadybeen progressing in the world. And, there are three nuclear reactors experienced severe accidents:
Winscale Unit 1 in UK, Three Mile Island Unit 2 in the US, Chernobyl Unit 4 in Ukraine. These
reactors have taken stabilization and safety measures for a long time. Furthermore, there issignificant uncertainty in the management of various radioactive substances in past nuclear
development facilities called “legacy sites” which are located in the UK, US and France etc. In thesesites, decommissioning and environmental remediation have been implemented for a long-term.
For this reason, these countries have been challenging the issues such as technological difficulties
called “unknown unknowns” (unknown issues that even uncertainty level is not known), project
management over several decades, and securing of large amount of funds.
In order to make steady progress on the decommissioning of the Fukushima Daiichi NPS asa risk reduction strategy described above, it is important to learn from the measures taken for thereactors which had accident and from the decommissioning activities of the legacy sites being
undertaken in the world while utilizing technologies and human resources at the world’s highest
level. In addition, to secure and maintain understanding and support from international communities,
it is also important to engage in decommissioning open to international societies.
Moreover, the decommissioning of the Fukushima Daiichi NPS is expected to offer theopportunity of development for cutting-edge science and technology and to become a source ofinnovation. For instance, the multicopter developed in the U.K. which enables three-dimensionaldose assessment (RISER) was tested for applicability to dose assessment at the Fukushima
Daiichi NPS in February and April of 2017. As the result, it could be measured the dose distribution
inside the Unit 1 and 3 buildings. This test result at the Fukushima Daiichi NPS in a harshenvironment is expected to utilize for development of drones and so on in the future. In this way,technologies and knowledge acquired through decommissioning activities of the Fukushima Daiichi
NPS can be beneficial for the decommissioning engaged in each country.
The international institutions such as IAEA, Organization for Economic Co-operation andDevelopment/Nuclear Energy Agency (“OECD/NEA”) have a role of contribution to design theinternational standard, share the knowledge and experience among nations, and form theinternational common perceptions for decommissioning. To proceed with the decommissioning of
the Fukushima Daiichi NPS in an open manner, it is important for Japan to participate in theactivities of these institutions. At the same time, it is expected to fulfill a part of Japan’s responsibility
to the international society by sharing the Fukushima experience with memberships of theinstitutions through participating in the discussion for designing the international standard.

1106.2 Facilitation of international cooperation activities
6.2.1 Enhancement of partnership with overseas decommissioning agencies
Concerning the long-term process of the Fukushima Daiichi NPS decommissioning,cooperation with overseas decommissioning agencies should be continuous partnership nottransient one. Especially, the decommissioning of legacy sites would serve, as a model for
approach preceding to the Fukushima Daiichi NPS decommissioning, in many knowledge inaccordance with technology and management.
Each country has established its own public decommissioning organization to promote
decommissioning of legacy sites because the decommissioning requires the expertise, concept,
and new technologies which are different from the operation/maintenance for nuclear reactor andnuclear fuel cycle facility. Therefore, it is important for NDF to reinforce a continuous partnershipwith these institutions such as Nuclear Decommissioning Authority (NDA), Commissariat al’energie atomique (CEA), United States Department of Energy, Office of Environmental
Management (DOE/EM) under government-level framework. TEPCO should also establish a longterm partnership with overseas decommissioning operators, and these should be broadcooperation-based.
Also in the past, NDF has concluded memorandums of cooperation with the NDA and the CEA,
while TEPCO has concluded agreements with Sellafield Ltd. in the U.K. and so on, to construct aframework of constant exchange of opinions as shown in table 5. In the future, it is also requested
to learn a lesson from various countries to utilize it for the decommissioning of the Fukushima
Daiichi NPS. Moreover, it is important for our country to share the technology and experience
derived from the decommissioning of the Fukushima Daiichi NPS to develop an interactive
cooperative relationship.
Table 5 Cooperative relationship among the institutions for the decommissioning of the
Fukushima Daiichi
Intergovernmental Framework between Japan and other countries
Framework

Descriptions

Annual Japan-UK Nuclear

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”(1st meeting: February 2012).

Dialogue

Japan-France Nuclear
Energy Committee
Japan-US
Decommissioning andIt was established under the joint statement of Japan–France
top-level meeting in October 2012 (1st meeting: February2012).
After the Fukushima Daiichi NPS accident in March 2011, theestablishment of the US-Japan Bilateral Commission on Civil

111

Nuclear Cooperation (the Bilateral Commission) wasannounced in April 2012 based on the relationship between
Japan and the US to further reinforce bilateral cooperation.
Under this commission, the Decommissioning andEnvironmental Management Working Group (DEMWG) wasestablished (1st meeting: December 2012).

Environmental
Management Working
Group

The Nuclear Working Group was established after confirmingthat Energy is one of the eight areas of cooperation plan
approved at the Japan-Russia top-level meeting in September
2016, (1st meeting: September 2016).

Japan-Russia Nuclear
Working GroupInter-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 asneeded. (Concluded in September 2013)

TEPCO

Sellafield,
Ltd.

Information Exchange Agreement for site’s operation, etc. wasconcluded. (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 on molten
core-concrete interaction, etc.

JAEA

Belgium
NuclearResearchCenterAgreement of Cooperation for Nuclear R&D and Research on
the accident of the Fukushima Daiichi

JAEA

Nuclear
SafetyResearchCenter
(Ukraine)

Memorandum for decommissioning research, etc. of the
Fukushima Daiichi NPS and Chernobyl was concluded.

JAEA

IAEA

Research Agreement on characterization of fuel debris1126.2.2 Integrating and utilizing wisdom in the world
In regard with the wisdom in the world that should be obtained for the decommissioning of the
Fukushima Daiichi NPS, there are a variety of approaches not only in technical aspect, but also
operational aspect such as system/policy, providing strategy, project plan/operation, safety,
regional communications, and so on. Up until now, there are opportunities to support thedecommissioning of the Fukushima Daiichi NPS by international society, and several kinds ofsupports have been given by overseas governmental institutions and experts including DAROD
project led by IAEA and collaboration projects promoted by OECD/NEA. It is deeply appreciated tothese supports.
As for system and policy among matters regarding management, information exchange withvarious countries by means of intergovernmental meetings, commissioned research on the system
structure to overseas organizations has been conducted as shown in Table 5. As for formulation ofstrategies, planning and management of projects, we learned from opinion exchanges with the
NDA, case studies, and the efforts of overseas engineering firms. These are utilized for the
introduction of project management techniques. Concerning the effort to communicate with thelocal residents, as described in Chapter 7, the decommissioning of the Fukushima Daiichi NPS
should be implemented with broad understanding and support of local residents. Based on thisidea, the efforts of local societies in various countries have been learning reference to overseas
practices.
Moreover, as for safety, we have been studied the information on international regulations
obtained through activities in international organizations and the regulatory activities related to theoverseas reactors suffered from accidents and the legacy sites. We will have used them as areference to formulate the institution for decommissioning in our country and the idea of ensuringsafety described in Section 4.2.
In the technical aspect, while learning the use situation of technologies in the decommissioningof legacy sites and so on, many suggestions from joint research by national projects anduniversities have been obtained. For example, as described in Section 5.3.1, referring also to theinformation that fuel debris in Chernobyl has been pulverized due to aging, the “To identify processof characteristic changes in fuel debris over time” project has just been initiated as the Essential
R&D Themes (Attachment 12). In addition, an international tendering system was applied to theGovernment-led R&D Program on Decommissioning and Contaminated Water Management of the
Ministry of Economy, Trade and Industry (METI) and several foreign companies are participatingin the project as subsidized operators. Moreover, our country participates in the international joint
research PreADES project led by the OECD/NEA from the viewpoint of characterizing fuel debris
and so on. As with the World Intelligence Project by MEXT, international joint research withoverseas organizations in the U.K., the U.S.A. and so on are implemented as well.
We have to keep on proceeding with the decommissioning of the Fukushima Daiichi NPS whilegathering and utilizing global knowhow. For this purpose, while encouraging participation in the113

activities of international organizations and intergovernmental meetings and promotinginternational joint research, experts from overseas are to be invited to receive advice andevaluations and gather know-how and experience with regard to decommissioning and so on
overseas (Attachment 13). Recently, OECD/NEA has established a committee named CDLM
(Committee on Decommissioning of Nuclear Installation and Legacy Management) to exchange
opinions concerning the decommissioning in April 2018, Japan should participate in this kind ofopportunity actively.
Also in the future, as described in Section 5.3.2, by developing international R&D centersorganized around JAEA/CLADS where various talents from Japan and overseas interact, it isessential to develop an environment to accumulate global know-how on decommissioning. It is also
important to promote international joint research continuously to obtain new technical know-how
for decommissioning effectively and efficiently and, in doing so, it is also necessary to pay attention
to overseas demand for the research utilizing fuel debris and so on. In addition, as described in
Section 5.1.1, the country and the operator must divide the roles appropriately to proceed steadily
to on-site application of the R&D results and, for this purpose, it is also important to refer to theprinciples of the R&D system for decommissioning overseas.
Moreover, regardless of whether it is inside or outside the country, decommissioning is
implemented under a contract between many companies and the decommissioning operator forwhich the global market is widely expanding. Under the circumstance that decommissioning isbeing conducted across by many companies across countries, it should understand the latest
status to the utilization of global-level technologies and human resources.
6.2.3 Dissemination of information to the global society
As Japan’s responsibility to the international global society for having caused the accident at
the Fukushima Daiichi NPS, also from the viewpoints of keeping concerns of human resource whowill be able to lead decommissioning and contributing to gather wisdom and knowledge from theworld, it is important to proceed with the decommissioning in a manner open to the international
community. It is also necessary to prevent reputational damage as described in Section 7.3;
therefore, the dissemination of clear information should be strengthened to formulate a precise
understanding in the international community.
Accordingly, NDF takes dissemination of information about the situation of the Fukushima
Daiichi NPS decommissioning globally through holding side events of the IAEA General
Conference, by taking the platform at major international conferences such as OECD/NEA Steering
Committee, Nuclear Safety Regulatory Information Conference (RIC) sponsored by the U.S.
Nuclear Regulatory Commission’s (NRC’s) and Decommissioning Forum in France, and throughpresentations at intergovernmental meetings. Moreover, in order to clearly report the situation of
the decommissioning of the Fukushima Daiichi NPS to the world, and to work on dialogue forsymbiosis with the local community, NDF has held the “International Forum on theDecommissioning of the Fukushima Daiichi NPS” every year since 2016. This Forum is a trial for114

global dissemination of the way to communicate about decommissioning through dialogue with thelocal residents, which is evaluated as an important event internationally.
Furthermore, it is also important to make the latest information on the decommissioning of the
Fukushima Daiichi NPS available throughout the world through publicity in foreign languages,
English websites, mailing lists and other means; therefore, this efforts must be continued to provide
clear information utilizing videos and so on. (Refer to Table 6, Approach for dissemination of
information to the world)
Table 6 Approach for dissemination of information to the world
Holding or attending International Conference (FY2017）
Conference Name
PeriodInternational Congress on Advances in Nuclear Power Plants
April, 2017
(ICAPP) 2017
Fukushima Research Conference (FRC)
Jun, Jul, Sep andNov. 2017 andMar. 2018
Japan-US Nuclear R&D Collaboration Symposium
June, 2017nd
The 2 International Forum on the Decommissioning of theJuly, 2017
Fukushima Daiichi NPS
Seminar in Ulsan Institute of Science and Technology (Korea)
September, 2017
IAEA side event
September, 2017
MIT-TIT Workshop
October, 2017
OECD/NEA Steering Committee
October, 2017
OECD/NEA Working Party on Decommissioning andOctober, 2017
Dismantling
Nuclear decommissioning forum (In France)
December, 2017
Uk-Japan Nuclear Industrial Forum
January, 2018
NRC Regulatory Information Conference (RIC)WM Symposia 2018
ASEM The 5th Nuclear Safety Seminar
Announcement in other conferences (FY2017)
Releasing the Strategic Plan to the embassies of EU countries
Dissemination of information on web (in English)
Site

JAEA
NDF, TEPCO
NDF
METI,
IRID
NDF
METI

March, 2018
March, 2018
March, 2018

NDF,JAEA
METI,NDF,
TEPCO
NDF
TEPCO
METI

November, 2017

NDF

Mid-and-long-term Roadmap towards the Decommissioning of TEPCO’s Fukushima
Daiichi Nuclear Power Station Units 1-4（http://www.meti.go.jp/english/earthquake/nuclear/decommissioning/）
Monthly report to the embassies concerning discharging and seawater monitoringfrom the Fukushima Daiichi NPS
Nuclear Damage Compensation and Decommissioning Facilitation Corporation’s
website（http://www.dd.ndf.go.jp/eindex.html）
Portal site of Decommissioning R&D Information（http://www.drd-portal.jp/en/）
Activities for Decommissioning（https://fukushima.jaea.go.jp/english/）
IRID website (http://irid.or.jp/en/)

115

Organization
NDF, JAEA, IRID,
TEPCO
JAEA

Organization
METI

METI, MOFA
NDF

NDF
JAEA
IRID

Fukushima Daiichi Decommissioning Project（http://www7.tepco.co.jp/responsibility/decommissioning/index-e.html）
Providing English version of Press release to foreign media
TEPCO CUUSOO（https://tepco.cuusoo.com/）
Management Office for the Project of Decommissioning and Contaminated Water
Management(https://en.dccc-program.jp/)

TEPCO
TEPCO
TEPCO
MRI

6.3 Close cooperation with relevant domestic organizations
As described in Section 6.2.1, the related domestic organizations are proceeding with efforts
to construct and strengthen partnerships with overseas related organizations in accordance withrespective roles. On the other hand, with regard to decommissioning R&D, as shown in Fig. 28,
cooperation such as joint research between research institutes and universities is in progress.
Since sharing know-how and personal relationships obtained through these activities is importantfor our country also in the context of ensuring consistency with international cooperative activities
and realizing effective international cooperation, close cooperation between related domesticorganizations must be promoted more than ever.

1167. Local community engagement and further enhancement ofcommunication
7.1 Approaches for local community engagement and further enhancement ofcommunication
For the continuous implementation of the decommissioning of the Fukushima Daiichi NPS,
symbiosis with the local community is the major premise. It must be developed awareness that thedecommissioning of the Fukushima Daiichi NPS is implemented primarily for the local community
and it should be aimed at decommissioning that contributes to reconstruction, making an effort tobuild a community-based, trusting relationship.
In the practice of communication which forms the basis of such trusting relationships with thelocal community, the starting point is to listen sincerely to the opinions of people of various positions
including local residents. In the session entitled “Thoughts on Decommissioning of 1F with the
Local community” of the International Forum on the Decommissioning of the Fukushima Daiichi
NPS51, many participants from inside and outside the prefecture including local residents attended
and expressed their feelings about the decommissioning and spoke with the experts. It wassuggested that a lot of people have questions and requests about the plans and activities of thedecommissioning, anxiety and doubt about safety and risk management of the decommissioning,
dissatisfaction and requests about the way to disclose information regarding the decommissioning.
In addition, the discussions were held on how dissemination of information can be improved andon how decommissioning should assist local reconstruction. For those concerned with thedecommissioning, it is important to proceed with information disclosure and communication,
accepting these feelings of the local residents and responding to them positively.
It is particularly important to respond to the questions about projects, plans anddecommissioning activities and to the anxiety about safety and risk management and, for thispurpose, while providing information appropriately on safety measures, work progress andradiation safety and keeping interactive communication, a common understanding with the local
residents must be formulated regarding the risk reduction policy to proceed with the futuredecommissioning. Moreover, as described in Section 2.3.3.4, in order to help the effort, it isimportant to create a mechanism of risk monitoring which indicates the progress of continuous andspeedy risk reduction, for example, by grasping the time variation of the risk.
Interactive communication is not satisfactory through simply sharing information between thetransmitter and receiver, what is needed is a mutual effort to reduce the gap in understanding bycontinuing adequate dialogue. By understanding interests, doubts and anxieties and by continuing
the process of the corresponding dialogue, based on mutual understanding between concerned
parties including the local residents and through the involvement of many stakeholders, it must be51

Day 1 of The 2nd International Forum on the Decommissioning of 1F, held in July 2, 2017, in Hirono-machi,
Fukushima prefecture, and Day 1 of the 3rd International Forum on the Decommissioning of 1F, held in August 5,
2018

117

aimed at reaching a more desirable decision-making process.
On the foundations of this kind of communication, it is necessary to initiate specific efforts that
decommissioning and the related various activities will be regionally-oriented industries through thecontribution to reconstruction and revitalization of the local community. Just as, for example, onsite infrastructure maintenance such as management of the canteen in the new main office building,operation and management of electric and air-conditioning equipment is commissioned to local
companies at the Fukushima Daiichi NPS, TEPCO keeps on promoting the purchase of locallyavailable goods and services and so on to proceed with the decommissioning work with thecontinuous cooperation of the local community. Moreover, a group of research facilities constructedin the Hamadori region to play a part in the Innovation Coast Scheme (see Section 5.3.2) where,by accumulating global know-how, R&D for decommissioning and human resources development
are being promoted are also important facilities for symbiosis with the local community. It is greatly
anticipated that expansion of these activities related to the decommissioning leading to newtechnologies, new industries and job creation and their acceptance by the local community will leadto reconstruction and revitalization of the Hamadori region.

7.2 Actual effort for better communication
In the future, when undertakings such as fuel debris retrieval will begin in earnest, referring to
the various overseas experiences and with proper cooperation between the related organizations
such as the government, TEPCO and NDF, it is necessary to consider and practice more carefulways to communicate.
The government will work on careful communication by providing information to andstrengthening communication with various people including local residents, by holding “The
Fukushima Advisory Board” as an occasion to discuss public relations, by preparing videos on thesituation of the decommissioning task and the brochure “Important Stories on Decommissioning,2018”52 as well as by active explanations to and conversations with local residents and officials of
related municipalities.
NDF will continuously hold the “International Forum on the Decommissioning of the Fukushima
Daiichi NPS”, provide clear information on the decommissioning of the Fukushima Daiichi NPS,and exchange opinions with the local residents, while making efforts to make it an opportunity forboth Japanese and foreign experts to widely share the latest progress and the technical results ofthe decommissioning. Moreover, NDF are vigorously conducting active explanations to andconversations with the local residents and the officials of the related municipalities. In addition, inthe viewpoint of being thorough with clear explanations and putting themselves in the listener’s
shoes, the brochure “Hairo no Iroha (ABCs of decommissioning)” 53 has just been prepared tocommunicate the story of decommissioning which tends to be technical in a simple way by
5253

http://www.meti.go.jp/earthquake/nuclear/pdf/reactorpamph_pdf_8MB.pdf
http://www.dd.ndf.go.jp/jp/about/pdf/hairo_iroha_2018_1.pdf

118

narrowing the focus, in an effort to disseminate accurate and clear information on thedecommissioning. Also in the future, it has been decided to continue these interactive
conversational activities more vigorously to sincerely accept the opinions of local residents.
TEPCO is continuously working on explanations to and conversations with local residents andthe officials of the related municipalities through management and the risk communicators at the“Prefectural Safety Assurance Conference for the Decommissioning of the Fukushima Nuclear
Power Station” sponsored by Fukushima Prefecture and so on, and on the explanations to andconversations with local representatives and so on regarding the progress of the Roadmap. In
November 2017, the Decommissioning Communication Center was established to enable theintegrated dissemination of information by adjusting the system of information dissemination on
decommissioning, which had been divided into departments and, at the same time, by disclosureof web contents and regular publication of the info-magazine on decommissioning of the Fukushima
Daiichi NPS, “Hairo Michi (the way to decommissioning)”, efforts will be made to further strengthen
the careful dissemination of information on the situation of decommissioning task and so on.
Moreover, to have people observe the real state of progress of the decommissioning task on thesite of the Fukushima Daiichi NPS is very effective for formulating a common understanding ofdecommissioning. TEPCO has been proactively accepting visitors to the Fukushima Daiichi NPS,
including evacuees, and 12,000 visitors were accepted in FY2017. In the future, setting theacceptance level at 20,000 visitors per year in FY2020 as a target, a further increase of theacceptance level and improving the visit (route and content) will be attempted54. In addition, in
March 2018, TEPCO released web-related content55 through which viewers can experience a real
sense of actually visiting the premises.

7.3 Further spread of communication and measures to Reputational damage
Reputational damage can be caused only with the presence of anxiety even if the risk is not
actualized. Moreover, it is pointed out that, even though seven years have passed since theaccident, the image immediately after the accident still remains without being erased and has an
influence. The delay in responding to the reputational damage, occurrence of troubles indecommissioning operations, exposure of operators and the cost increase may damage the social
evaluation of the decommissioning effort, which might lead to a vicious circle of further delays inrisk reduction activities due to radioactive materials. In order to avoid this vicious circle, while
making efforts in regard to proper safety management to prevent leakage of radioactive materials
and so on, it is more important than anything to quickly reduce the existing risk.
54

TEPCO, Approach to the Countermeasures for Decommissioning and Contaminated Water Management of
the Fukushima Daiichi NPS - Information Dissemination and Communication -, Handout 3-5 of the 16th
Fukushima Council for Countermeasures for Decommissioning and Contaminated Water Management, April 27,
2018,
http://www.meti.go.jp/earthquake/nuclear/decommissioning/committee/fukushimahyougikai/2018/pdf/0427_01f.pdf
55 “INSIDE FUKUSHIMA DAIICHI - A virtual tour of the site of decommissioning -“,http://www.tepco.co.jp/insidefukushimadaiichi/index-j.html

119Moreover, in order to prevent reputational damage, in addition to communicating with the local
residents as described above, approaching a wider range of people becomes necessary. Not tomention the local residents, press, market and distributors, efforts to widen the range ofcommunication to consumers including those overseas become necessary. In case of problems, asincere and careful explanation must be provided as much as possible on the major premises ofpreciseness and transparency, and also after any problems, it is needed to actively communicate
the execution status of safety measures thereafter and the improvement status of safety
management onsite. Furthermore, considering the case of internal investigation of Unit 2 executedin January and February 2017 when a high dosimetry value was measured, causing anoverreaction inside and outside the country, it is important to try to disseminate information carefullyeven during normal times.
In December 2017, the government formulated the “Strategies to Eliminate Negative
Reputation and to Strengthen Risk Communication”, and the concerned government agencies areworking on it in an integrated manner56. Moreover, TEPCO formulated and published the “Action
Plan to Combat Harmful Rumors” in January 2018 57 . Based on these, it is important for theconcerned organizations to continue their concerted efforts.

56

Reconstruction Agency, Strategies to Eliminate Negative Reputation and to Strengthen Risk Communication,
see the portal site (http://www.fukko-pr.reconstruction.go.jp/2017/senryaku/ )57 Development of Action Plan against Harmful Rumor, TEPCO, January 31, 2018
http://www.tepco.co.jp/press/release/2018/1475368_8707.html

120

List of Acronyms/Glossaries
Acronym

Official Name

CEA

Commissariat à l'énergie atomique et aux énergies alternatives

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

MCCI

Molten Core Concrete Interaction

NDA

Nuclear Decommissioning Authority

NDF

Nuclear Damage Compensation and Decommissioning Facilitation
Corporation

NRC

Nuclear Regulatory Commission

OECD/NEAOECD Nuclear Energy Agency

PCV

Primary Containment Vessel

RPV

Reactor Pressure Vessel

S/C

Suppression Chamber

SED

Safety and Environmental Detriment

TMI-2

Three Mile Island Nuclear Power Plant Unit 2

VR

Virtual Reality

X-6 penetration

PCV X-6 penetration

Center of the World The project that promotes nuclear science and technology and humanIntelligence project
resource development gathering wisdom and knowledge
Operating Floor

Operating Floor of the buildings

Underwater ROV

A remotely operated underwater survey vehicle (Remotely Operated
Vehicle)Strategic Plan

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

As a specific nuclear facility, “the matters for which measures should
Matters to be addressed be taken” that The Fukushima Daiichi NPS must have the necessary
safety measures in place that are required by the NRA.
Roadmap

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

TEPCO

Tokyo Electric Company Holdings, Inc.

121

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.

Glossary

Description

CRD housing

Housing that contains Control Rod Drive mechanism

MCCI ProductA product generated by high temperature molten core and concreteinteraction (MCCI)T.P.

A height from the average surface of the bay of Tokyo (Tokyo Pail) as astandard of elevation.
(cf. O.P: Onahama Port construction level (the lowest position of the water
surface)

Well 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 in operation)

Basis seismic
ground motion

A magnitude of shake associated with earthquake that may affect greatly
to nuclear facilities. Based on the latest scientific and engineeringknowledge, it is provided in accordance with the geological structure or
ground structure, etc.

Submersion method

A method to retrieve fuel debris by submerging all fuel debris in watering
up to upper part of Primary Containment Vessel

Partial submersion
methodA method to retrieve fuel debris in a state that a part of fuel debris is
exposed in the air without watering

Grating

A lattice-type scaffolding used for side ditch lid or work scaffold

SludgeMuddy substance, dirty mud

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

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

Platform

Footing for work installed under RPV inside pedestal

Pedestal

A cylindrical basement that supports a body of reactor

Measurement bymuon (fuel debris
detection technology
with muon)

A technology to grasp location or shape of fuel in using characteristics bychange of number or track of particle depending on the difference ofdensity, when muons (muonic atoms) arrive from the cosmos andatmospheric air and pass through a substance

Mock-up debris

An artificial fuel debris, its chemical composition and chemical form were
estimated based on the case of Unit 2 of Three mile Island Nuclear Power
Plant accident

Mock-up

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

Preliminary
engineering

Engineering operation to determine the feasibility of engineering work
preliminary in advance of basic design that is conducted at the beginningof general construction122

Well shield
ウェルシールド
プラグplug

Operating floor
オペレーティングフロア
使用済燃
Spent
Fuel
料プール
Pool
（SFP）

Component

機器ピット
Pit

4th4floor
階 (SFP)

Primary
原子炉格納容器Containment（PCV）
Vessel (PCV)
Reactor
原子炉圧力容器Pressure Vessel（RPV）(PRPV)

3rd3floor階

階
2nd2floor

Pedestal
ペデスタル
Vent Tube
ベント管

st
floor
11階

1st 地下basement
Fl.
1階

サプレッションSuppression
Chamberチェンバ（S/C）

トーラス室Torus room
Courtesy of IRID

（IRID 提供）

Fig. 29 Structural drawing inside Reactor building

Top cover of RPV
原子炉圧力容器上蓋

蒸気乾燥器
Steam dryer
RPV body
原子炉圧力容器本体

気水分離器
Steam separator
給水スパージャFeedwater
sparger

炉心スプレイ配管
Core
spray piping

Upper上部格子板
grid plate
Fuelassembly
燃料集合体

炉心シュラウドCoreshroudCore support
plate
炉心支持板

制御棒
Control rod

Neutron flux measuring guide

中性子束計測案内

ジェットポンプ
Jet pump

Differential pressure
detector/
差圧検出/
管
Piping of Boric acid solution
ホウ酸水注入系配管
injection system

制御棒駆動機構
Control Rod Drive Mechanism
Pedestal for RPV
原子炉圧力容器用
ペデスタル
Courtesy of IRID

（IRID 提供）

Fig. 30 Structural drawing inside Reactor Pressure Vessel(RPV)123

List of Attachment

Attachment 1

Revision of the Roadmap and the earlier published strategic plan............. 125

Attachment 2

Major risk reduction measures performed so far and future course of action ..
..................................................................................................................... 126

Attachment 3

Overview of SED indicator........................................................................... 131

Attachment 4

Change in risk over time .............................................................................. 140

Attachment 5

Coverage of fuel debris retrieval.................................................................. 141

Attachment 6 Reduction of liquid phase and contaminated water (Water balance around thereactor buildings) .......................................................................................... 144
Attachment 7

Study on the water level at the bottom of the PCV during fuel debris retrieval..................................................................................................................... 145

Attachment 8

Terms related to radioactive waste management ....................................... 149

Attachment 9

Overall image of Storage Management Plan for the Fukushima Daiichi NPS ..
..................................................................................................................... 150

Attachment 10 Draft of technical map toward human resource development fordecommissioning reactors ......................................................................... 152
Attachment 11 Current Progress of R&D in the Project of Decommissioning andContaminated Water Management, etc. .................................................... 153
Attachment 12 Basic Direction of 6 Essential R&D Themes ............................................. 184
Attachment 13

Record of collaborative activities with foreign organizations..................... 190

124

Attachment 1 Revision of the Roadmap and the earlier published strategic plan
[1st Edition of the 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 the spent fuel pools, 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).
[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.
[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 the spent fuel pool and fuel debris retrieval, and R&D Plan was reviewed
based on the above.
[Strategic Plan 2015 (April 30, 2015)]
 The 1st edition of the Strategic Plan was published to provide a verified technological basis to theRoadmap from the viewpoint of proper and steady implementation of
decommissioning of the
Fukushima Daiichi Nuclear Power Plant.
(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 Plant 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, Proven, Efficient, Timely, 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 submersiontop 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 treatment
method.
[Roadmap Revised 3rd Edition (June 12, 2013)]
 While much importance was placed on risk reduction, priority-setting for actions was performed so thatrisks 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), reduction of amount of radioactive materials contained in
the stagnant water in the buildings by half (FY2018), etc.
[Strategic Plan 2016 (July 13, 2016)]
 In response to the progress state of decommissioning after publication of the Strategic Plan 2015,
concrete concepts and methods were developed based on the concept and direction of the efforts of theStrategic Plan 2015 to achieve the target schedule specified in “Policy decision on fuel debris retrieval foreach Unit” which is expected to be completed by about summer 2017 defined in the Roadmap, “Compilingof the basic concept concerning treatment and disposal of radioactive waste” which is expected to becomplete in FY2017, etc.
[Strategic Plan 2017 (August 31, 2017)]
 Feasibility study was conducted on three priority methods for fuel debris retrieval. Recommendations forpolicy decision on fuel debris retrieval were made and efforts after policy decision including preliminary
engineering were recommended as strategic proposals.
 Recommendations were made for compiling the basic concept concerning solid waste treatment anddisposal.
[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 treatment and disposal were compiled.
 Individual work was defined based on the viewpoint of “Optimization of total decommissioning work”.

Attachment 2 Major risk reduction measures performed so far 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. A2-1. 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 the fuel pool, cooling of fuel debris with water injection by core spraysystem, nitrogen injection, etc. (in 2011) as well as the contribution of inventory and decay heatdecrease due to decay of radioactive materials.

Fuel removal from
Unit 4 SFP started

0

1

2
3
4
5
Year after accident

6

Risk level (linear scale)

Risk level (linear scale)

Risk level(logarithmic scale)

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) at the bottom left in the figure and the similar graph since 3 years after theaccident is given in the bottom right with the risk level multiplied by 10.These graphs demonstratethat a continuous risk reduction has been achieved.

7

Mid-and-Long-Term
Roadmap Phase 1

Mid-and-Long-Term
Roadmap Phase 2

Risk level in the rightgraph is multiplied by 10

0.5

Fuel debris

1
1.5
2Year after accident
Fuels in SFPs

2.5

Contaminated water

3

3.5

4

Fuel debris

4.5
5
5.5
Year after accident
Fuels in SFPs

6

6.5

7

Contaminated water

Fig. A2-1 Reduction of risks contained in the Fukushima Daiichi NPS (Same of Fig.4 of p.17)

Change in the risk level with further breakdown of major risk source over the time since 0.5 years
after the accident is shown in Fig. A2-2. In using with logarithmic scale, risk sources can beindicated that are too small to be displayed in the linear scale of Fig. A2-1. Fuel in the common
pool and fuel in dry casks are not shown which stay in the region of sufficiently stable management.

Total

Risk level (logarithmic scale)

Fuels in SFPs

0

Fuel debris
Stagnant water in
bldgs
Contaminatedwater in trenches
Stored water intanks
Secondary wastes

1

2

3
4
Year after accident

5

6

7

Contaminatedstructures in bldngs
Rubble, etc.

Fig. A2-2 Change in the risk level of every major risk source
Reasons of change in the risk level of every major risk source shown above are as follows, andthe summary of the risk reduction measures described below is shown in Table A2-1.
(1) Fuel in SFPs
Since one year after the accident, removing rubble and installing the fuel removal cover were
implemented as preparations of fuel removal in Unit 4. Fuel removal in SFP in Unit 4 was started
2.5 years later after the accident by transportation to the common pool with low Safety Management,
resulting in the risk level reduction (completion in 2014). Except reduction due to decay of theradioactive materials, it is this fuel removal from SFP in Unit 4 that contributes most so far to therisk level reduction of the total site.
In Unit 3, rubble removal, etc. have been conducted as preparations of fuel removal, and recently
the cover for fuel removal was installed (in 2018), resulting in reduction of risk level.
In the future fuel removal in SFP in Unit 3 is scheduled to start from about the middle of FY2018
and in Unit 1 and 2 by 2023 to reduce the risk level.
(2) Fuel debris
Although fuel debris was at an high risk level just after the accident because it was at molten
state, and in addition, the release risk of radioactive materials was realized, the risk level was
reduced, not only by decay of the radioactive materials, but also by reduction of Hazard Potential
and Safety Management because of restoration and strengthening of cooling function.
Although the risk level reduction effect was observed because of Safety Management reduction
caused by diffusion control function with the building cover in Unit 1 (in 2011), this effect currently

127

disappeared due to removal of the building cover (in 2015) to prepare for fuel removal in SFP.
(3) Stagnant water in the buildings
Stagnant water in the buildings is generated by cooling water for fuel debris and leakage ofgroundwater into the buildings, and its treatment has been progressed because of the starting ofoperation of the cesium adsorption apparatus (KURION and SARRY) (in 2011). In addition, due tothe effects of sub-drain systems and land-side impermeable walls, the groundwater level was
lowered and the inflow amount of groundwater into the buildings was reduced, resulting inachievement of lowering of the water level in the buildings. This also helped progress of treatmentof the stagnant water. Moreover, draining of the condensers has been completed (in 2017). This
stagnant water treatment in the buildings so far significantly contributes to risk level reduction of
the total site following contribution by fuel removal in SFP.
This treatment of the stagnant water in the buildings resulted in generation of secondary wastefrom water treatment systems described in (6) below to which most inventory was transferred andthe stored water in the tanks described in (5) below as the treated water.
Recently, an increase in radioactive material concentration in the stagnant water in the buildingof Unit 3 is observed and the risk level of total stagnant water in the buildings is rising. This isconsidered as an effect that isolated water, etc., which was not included in the risk assessment,
has leaked in, and investigation and study are required in the future.
(4) Contaminated water in trenches
Although the high concentration of contaminated water has been stagnated since just after theaccident in the seawater pipe trenches in Unit 2 to 4, treatment of the stagnant water has been
completed and trenches were blocked (in 2015).
Stagnant water in the seawater pipe trench in Unit 1 with lower concentration compared to thosein Unit 2 to 4 is now being continuously processed.
(5) Stored water in tanks
In each tank, there are several types of stored water with different radioactive material
concentration. Although concentrated liquid waste generated by operation of the evaporative
concentration system (in 2011) just after the accident had high concentration of radioactive
materials, the evaporative concentration system stopped after a short period of operation (in 2011)
and concentrated liquid waste is not generated at present. In addition, the concentrated liquid waste
slurry was separated from this concentrated liquid waste and has been regarded as secondary
waste from water treatment system (in 2014). The risk level of the residual concentrated liquid
waste was decreased because of reduced inventory and transfer o the welded tanks that is secured
more.
Processing of the concentrated salt water generated by the cesium adsorption apparatus has

128been completed in 2015 by operation of the Advanced Liquid Processing System(ALPS) (in 2013
with the existing system and in 2014 with the expanded system) and the high-performance ALPS(in 2014).
Reduction of risk level is devised by raising and duplicating the weir (completed in 2014 forexisting tanks) and replacing tanks from flanged type to welded type (will be completed in FY2018).
(6) Secondary waste from water treatment
Many radioactive materials have moved from contaminated water to secondary waste throughwater treatment. Secondary waste from water treatment system includes the waste sludge, waste
adsorption columns by operation of the cesium adsorption apparatus (KURION and SARRY) (in2011), HIC slurry by operation of the Advanced Liquid Processing System (in 2013), waste
adsorption columns by operation of the high-performance ALPS (in 2014), waste adsorption
columns by the mobile type treatment equipment that processed contaminated water in theseawater pipe trenches, etc. Although the contribution level of the waste sludge is higher as therisk level, currently the waste sludge is not newly generated. Therefore, the risk level of the total
secondary waste from water treatment system does not tend to increase.
Although the concentrated liquid waste slurry separated from the concentrated liquid waste (in2014) was stored in the welded type horizontal tanks without the weir and placed on the groundwithout the base, its risk level was reduced due to implementation of safety measures by installationof the reinforced-concrete base and the weir (in 2017).
(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. in the reactor buildings, PCVs
or RPVs that are contaminated by dispersed radioactive materials caused by the accident sincefuel debris retrieval is not started yet.
(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 thehighest risk level. In the past, the facilities with better management condition have been enhanced
by soil covered temporary storage facilities (in 2012), fallen tree temporary storage pool (in 2013),
expansion of solid waste storage facilities (in 2018), etc. Furthermore, outdoor temporary storageis planned to dissolve by increasing incinerator facility, volume reduction facility and solid waste
storage, etc. by 2028 in accordance with the solid waste storage management plan.

129

Table A2-1 Major efforts affected directly to the risk level
implemented up to today and future actions
Year
2011

Fuel debris Starting water
injection by corespray system
 Starting Nitrogen
injection
 Installation ofbuilding cover(Unit 1)

Contaminated water

2013

2014

 Removal ofbuilding cover(Unit 1)

 Removing large rubbles inpool (Unit 3)
 Investigation of internalpool (Unit 3)

2016
2017

Futureactions

 Receipt start of soil
covered temporarystorage facility
 Operation start of ALPS
 Starting transfer to weldedtank Installation of fuel removal
cover (Unit 3)
 Start of retrieval
from first
implementing unit
in 2021

 Start of Unit 3 removal in
the middle of FY2018
 Start of Unit 1 and 2
removal, aiming by
FY2023130

 Receipt start offallen tree
temporary storagepool Completion of raise andduplication of weir forexisting tanks
 Operation start of
expanded ALPS and highperformance ALPS Operation start of mobiletype strontium removal
equipment Separation ofconcentrated liquid waste
slurry from concentrated
liquid waste Adding Sr-removing
function to SARRY Adding Sr-removing
function to KURION
 Completion ofconcentrated salt water
treatment Completion of seawater
pipe trench blockage (Unit
2 to 4)
 Start of pumping up by
sub-drain system Closure of sea-side
impermeable walls Completion of condenser
drainage (Unit 1-3)

2018

Solid waste

 Operation start of KURION
and SARRY Operation start andshutdown of AREVA
Decontamination System Operation start andshutdown of evaporative
concentration system

 Investigation of internalpool（Unit 4）
 Removing rubbles inoperation floor（Unit 4）
 Installation of fuel removal
cover (Unit 4)
 Removing large rubbles inpool (Unit 4)
 Removing rubbles inoperation floor (Unit 3)
 Start of fuel removal (Unit4)
 Completion of fuelremoval (Unit 4)

2012

2015

Fuel in SFP
 Cooling function
restoration

 Closure of land-side
impermeable walls, mostly
completed Completion of transfer towelded tanks (FY2018)
 Completion of stagnant
water treatment in thebuildings in 2020

 Completion ofsafety measuresfor concentrated
liquid waste slurry Installation of new
solid waste storage
facility Storing of solid
waste in thestorage facility inaccordance withsolid waste storage
plan by FY2028

Attachment 3 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 indicator 58 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)4or
𝑆𝐸𝐷 = (𝑅𝐻𝑃 + 𝐶𝐻𝑃) × (𝑆𝑆𝑅 × 𝐵𝐸𝑅 × 𝐶𝑈)4
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 dose59. The STP is defined as the volume of waterrequired to dilute 1TBq of radioactive materials and corresponds to the dose coefficient. Ingestionof a certain amount of such diluted water throughout the year will result in an exposure dose of
1mSv. The SED indicator conservatively uses the larger dose coefficient between ingestion andinhalation.
𝐼𝑛𝑣𝑒𝑛𝑡𝑜𝑟𝑦(𝑚3 ) = 𝑅𝑎𝑑𝑖𝑜𝑎𝑐𝑡𝑖𝑣𝑖𝑡𝑦(𝑇𝐵𝑞) × 𝑆𝑇𝑃(𝑚3 ⁄𝑇𝐵𝑞 )58
59

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

131

Form Factor (FF), as shown in Table A 3-1, 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 A 3-1, 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 A 3-2, is an indicator representing time allowanceavailable before restoration when safety functions maintaining current stable state are lost. CF is
taking into account exothermicity, corrosivity, flammability, hydrogen generation, reactivity with airor water, criticality, etc. which are typical characteristics of risk sources. CF is the same as the one
defined by the NDA.

132

Table A 3-1 Definition and score of FF
No.

Form

FF

1

Gas, liquid, watery sludge* and aggregatedparticles*

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 caseof the Fukushima Daiichi NPS
Table A 3-2 Definition and score of CFNo.

Time allowance available before any risk is realized

CF
1

1

Hours

2

Days

3

Weeks

4

Months

5

Years

6

Decades

10
100
1,000
10,000100,000(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 based133

on 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 A 3-3 and Table A 3-4 respectively.

134

Table A 3-3 Criteria and score of FD
Category*

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, no assessment forcontainment function is available.

100

2“Safety assessment criteria*2” are not satisfied at “the time of assessment*1” causedby 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 astransfer, treatment, recovery, etc.) *3” for the risk source contained in the component
for diffusion control function.
The component or diffusion control function satisfying “safety assessment criteria”
exists at “the time of assessment”.

52

5

Integrity of diffusion control function has been assessed and “safety assessment
criteria” are satisfied until “the time of work (such as transfer, treatment, recovery,etc.)” for the risk source.
Frequency of occurrence of “contingency*4” is high, and when contingency 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 as transfer,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 as transfer,treatment, recovery, etc.)” for the risk source.
Facilities dissatisfying “safety assessment criteria” exist in the surrounding area, andthe 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 as transfer,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 as transfer,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 as transfer,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

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.),135

etc..

Table A 3-4 Criteria and score of WUD
Category*1*2

Criteria(NDA definition is modified to enhance applicability to the caseof the Fukushima Daiichi NPS)

NDF
Score

1

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

100

2

The material is fuel and active (which has fissile properties).
Necessary information for work including treatment, recovery, etc. is insufficient, andcontrol and surveillance are unavailable.
Handling is practicable for the current form or condition because of reasons 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. is insufficient.

74

4

The material is fuel and active (which has fissile properties).
Necessary information for work including treatment, recovery, etc. is obtained (canbe confirmed or estimated), and control and surveillance with monitoring, etc. areavailable.
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, andcontrol 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 geometrical instability.
Handling is impracticable for the current form or condition.

98

Although the material is inactive, it has physical or geometrical instability.
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“Active” refers to possession of activity defined by CF at such a significant level as that activity
affects control and work.
“Inactive” refers to non-possession of activity or possession of sufficiently low level of activity.

(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 A 3-5.
Speed to Significant Risk (SSR) is an indicator concerning the time until the public is affected136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.

137

Table A 3-5 Definition and score of SSR, BER and CU

Indicato
r

SSR

BER

CU

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 of magnitude or canfacilitate control stepwise.

4

Implementation of measures can reduce risks by 1 or more order of magnitude, 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 6 points

4(1)+(2)=3 to 4 points

1(1)+(2)=
2 points(1) Assessment for the present state
1 point:
Major nuclear types and diffusion pathways are monitored.
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 for constructionof assessment model.
2 points: Major characteristics representing the site are obtained.
3 points: There is no model usable for future prediction

Table A 3-6 shows the overview of containment function, safety equipment, the state of control
and surveillance, etc. of each risk source used when Safety Management was assessed..

138

Table A 3-6 Characteristics of each risk source concerning Safety Management
Risk source

Characteristics

Fuel debris

No significant damage has been observed in the PCV and redundancy of criticality
control, cooling system and hydrogen explosion prevention system is secured. In addition,
important parameters including Xe concentration, temperature, hydrogen concentration,etc. are being monitored.

Spent fuelThe spent fuel pool of each Unit is designed to maintain subcriticality, and redundancy ofthe cooling system is also secured. Some Units have experienced falling rubble andheavy weight objects, ceiling defects of the buildings, seawater injection, etc.
Common pool and dry casks, as well as the buildings are not damaged by the earthquake
and the tsunami.

Contaminated water,etc.

For stagnant water in the buildings, containments of the contaminated water are
maintained by management of the water level difference between the contaminated waterand ground water.
The water treated by the purified equipment is stored either in flanged tanks or weldedtanks, and flanged tanks are being replaced to welded tanks.

Secondary wastefrom water treatment
systems

The waste adsorption column is a shielded carbon-steel vessel storing Cs-adsorbed
zeolite, which is contained again in a shielded vessel, and is mounted on the box culvert
or the frame. Any control of decay heat removal, etc. is not required.
Waste sludge is stored in the agglomeration pit integrated in the process main building,where leak monitoring, decay heat removal, and hydrogen exhaust operations are
performed.
HIC slurry is stored in the polyethylene container, which is further contained in the SUS
reinforced vessel, and is stored in the box culvert. Although decay heat removal is not
required, measures against hydrogen generation is continuously taken. Concentrated
liquid waste is the liquid waste generated by concentrating the concentrated saltwater
through the evaporative concentration system and has a high concentration of radioactive
material and salinity. Its precipitation is extracted as a concentrated liquid waste slurry andstored in welded tanks within the weir.

Rubbles, etc.

Rubbles, etc. that is stored in the solid waste storage facility is waste which has a highconcentration of radioactive material, etc. and is contained in the container. Special
control is not required for this waste.
Temporally stored rubbles, etc. are waste which has various concentrations of radioactive
material and is stored outdoor in various forms, then monitoring and such is required forthis waste.

Contaminatedstructures, etc. in thebuildings

They are a variety of structures and the like in the buildings that were contaminated with apart of fission products like Cs which was released from fuel due to molten core at theaccident. The location and from of the contaminants are still investigation.

139

Attachment 4 Change in risk over time
Overview of the concept of risk management in the UK is shown in the conceptual diagram below.
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 increasecaused by degradation of facilities 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 notreach 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 the solid line) so that it may not reach into the unacceptable or intolerable
region.

Risk Level

Unacceptable area

Area where risk level
cannot be allowed
when taking a longview of the future.

Area where judgment ismade whether the risk levelis acceptable or not withclose examination ofdifficulty of risk reduction

Broadly cceptable area

Time
Reference: 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 Plant (April 2016).

Fig. A 4 Change in risk over time140

Attachment 5 Coverage of fuel debris retrieval
In the Roadmap issued in 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 “fuelassembly, control rod and structures inside reactor have melted and solidified together according
to the report by IAEA60,61.
The condition inside PCV is as shown in Fig.A5-1, as the comprehensive estimations from theinside investigation of reactor, the past accidents including TMI-2 or ChNPP-4, and the result of thesimulation test. It does not show any of specific unit. For more detail, as shown in the Fig. fueldebris can be classified by form such as damaged pellets, debris, crust, etc.

Fig. A5-1 Estimated inside of the PCV of the Fukushima Daiichi NPS

60

International Atomic Energy Agency Experiences and Lessons Learned Worldwide in the Clenup andDecommissioning of Nuclear Facilities in the Aftermath of Accidents, IAEA Nuclear Energy Series No. NW-T-2.7,
Vienna (2014)
61 Managing the Unexpected in Decommissioning, IAEA Nuclear Energy Series No. NW-T-2.8, Vienna (2016)141

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, collection,transfer and storage. The one includes nuclear fuel material and the others. The one that does notinclude 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 shownin Fig. A5-2. 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. A5-2 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 a hardsolid 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 fuel142

component originally, like CRD housing, grating and s, then solidified. It is possible to confirm fueladhesion state by sight.
Material contaminated by fuel : A substance that adhering molten fuel cannot be confirmed by sight,
but fuel component can be detected with alpha ray detector. It is impossible to locate fuelcomponent other than using by electron microscope because particle of adhered fuel component
is extremely small and whit.

143

Attachment 6
Reduction of liquid phase and contaminated water (Water balance around thereactor buildings)
The capability to seal in the liquid is required to contain radioactive material to restrict the releaseof the uncontrolled radioactive material from the reactor buildings with the objective of minimizing
the effect of exposure to the liquid the on the public as with the exposure to the gas.
However, since the primary containment vessels (PCVs) and buildings (torus rooms) are foundto be pierced and damaged by the earthquake, it is difficult to statically contain the liquid toperfection. For this reason, containing radioactive material is planned by maintaining the differencein level between the contaminated water and groundwater to prevent water containing radioactive
material from being released from the reactor buildings to soil outside the buildings.
Due to the difference in level between groundwater and the water inside the reactor buildings,
groundwater flows into the buildings. In the torus rooms, this groundwater is mixed with coolingwater coming out of the PCVs and then discharged, treated, and stored in a planned way. Still, theemergence of contaminated water should be desirably minimized to the extent possible. Fig. s 1
and 2 show the current mass balance of water around the reactor buildings and the mass balancethat will be achieved in the future if a water circulation system is built. From the viewpoint of watermass balance, it is possible to reduce the emergence of contaminated water possible by:(1) Reducing the inflow of the groundwater
It is important to take measures to prevent rainwater from flowing into the reactor
buildings and to set an appropriate groundwater level.
and/or(2) Reducing the emergence of contaminated waterThis method reduces the amount of the PCV cooling water that flows out into the torus
rooms to reduce the amount of contaminated water by setting the level of water inside the
PCVs (dry well sections) and building system for directly recovering PCV cooling waterwith a combination of the reduction of PCV cooling water and water stopping (whether it isfeasible or not must be determination).
In the future, in reviewing in details of the method for removing fuel debris in light of theengineering of the undertaker, it is necessary to identify how the water level inside the PCVs should
be set, whether or not it is possible to stop water, the system for directly recovering cooling waterfrom the PCVs (dry wells), and so on.
i1:Supply Cooling water
i2:Drain Cooling water
i3:Inflow due to waterlevel difference
i4:Storing after treated
i5:Rainwater Inflow

i5

Reactor Building
PCV
RPV

Cooling
Water

i1

i2

i3

Liquid Waste
treatment
equipment

Reactor Building

i5

PCV

RPV
Cooling/pu
rificationequipment

i1

i4

i6

Liquid Waste
treatment
equipmenti3

Torus Room ： i2＝i1＋i3＋i5
Storing Treated water ： i4＝i2 - i1

Sub-drainage
S/C

i2

Storing
Treated water

i1:Supply Cooling water
i2:Drain Cooling water
i3:Inflow due to waterlevel difference
i4:Storing after treated
i5:Rainwater Inflow
i6:Emission of surpluses

i4

Torus Room ： i2＝i1＋i3＋i5
Storing Treated water ： i4＝i6＝i2 - i1

Sub-drainage

Torus Room

S/C

Relationship with Water Balance surrounding the Reactor Building (Current status)

144

Torus Room

Relationship with Water Balance surrounding the Reactor Building(After established the system)

Fig. A6-1 Water balance around the reactor buildings

Storing
Treated water

Attachment 7 Study on the water level at the bottom of the PCV during fuel debris retrieval
Assuming that control of the water level can be performed by installation of underwater pumps,
state of the water level during fuel debris retrieval at the bottom of the PCV is illustrated in theconceptual diagram (Fig. A7-1) with grouping of the following cases. Each low of the diagram showsthe expected result on the water level in the following cases; the 1st low shows the case of present
state, the 2nd low shows the case when water stoppage is performed by injection of the water
stoppage material inside the vent pipe (vent pipe water stoppage), the 3rd low shows the casewhen water stoppage is performed by burial of the downcomer tip with filling of the highly workableconcrete inside the S/C (Suppression Chamber) (downcomer water stoppage), and the 4th low
shows the case when water stoppage is not performed. Since it is desirable to maintain the waterlevel inside the S/C as low as possible from the viewpoint of earthquake resistance during fueldebris retrieval, it was assumed that lowering of the water level inside of the PCV down to thelowest part of the vent pipe joint can prevent water from flowing inside the S/C when water stoppage
is not performed in each Unit.
Furthermore, development of water stoppage technology has just been started with installationof the water stoppage plate at the jet deflector and installation of the weir inside the D/W (Dry Well)in addition to the conventional vent pipe water stoppage, etc. as development of the water stoppage
technology conducted in the Government-led R&D Program on Decommissioning andContaminated Water Management. While considering these situations described above, it isnecessary to perform setting and control of the water level based on implementation or nonimplementation of water stoppage.
Progress of the study on each Unit is shown below.
A.

Study on the water level in the PCV in Unit 1
Current water level in the PCV is assumed to be about 2m and most part of fuel debris at thebottom of the PCV is considered to be submerged under water. In addition, leakage of the coolingwater in the PCV is found to the S/C through the vacuum break line (about 1.1m high from thebottom of the PCV) and to the torus room through the sand cushion drain pipe.
Therefore, it is necessary to lower the water level of the PCV to the level of the vacuum breakline or lower before applying the water stoppage technology between the PCV and the S/C (ventpipe water stoppage or downcomer waster stoppage) or the water level in the PCV is required tobe maintained at the level of the lowest part of the vent pipe joint or lower when water stoppage is
not performed. In addition, even in the case of performing either measure described above, it isconsidered to become necessary to take action against the cooling water flowing into the torus
room through the sand cushion drain pipe when retrieving fuel debris in the water or under thecooling water being kept flowing on the fuel debris. Furthermore, installation of the drain receiver145

by means of the remote technology at the sand cushion drain part is considered necessary.
When either vent pipe water stoppage or downcomer water stoppage is performed, as the waterlevel in the PCV is maintained to the level of the vacuum break line or lower as described above,
the amount of the cooling water stagnating in the PCV is not large. Therefore, even if the coolingwater in the PCV leaks out to the torus room supposedly in case of damage at the vent pipe joint
part, the water level in the torus room can be maintained at the groundwater level or lower. In
addition, as the measures to prepare against leakage of the cooling water to the S/C from inside of
the PCV supposedly in case of damage of the downcomer joint part, it is being studied to takemeasures by collecting the cooling water with pumps installed in the S/C.
When water stoppage is not performed, the water level in the PCV will be maintained at the levelof the lowest part of the vent pipe joint or lower and fuel debris retrieval will be performed in the air
while the poring cooling water is kept flowing on fuel debris. In this case, it is necessary to study inadvance on the decay heat of fuel debris, required amount of the cooling water, etc., because apart of fuel debris will be expected to be exposed in the air.
B.

Study on the water level in the PCV in Unit 2
Current water level in the PCV is assumed to be about 0.3m and most part of fuel debris at thebottom of the PCV is considered not to be submerged under water. In addition, occurrence ofleakage is assumed from the S/C to the torus room because both water levels in the S/C and thetorus room are almost the same.
When either vent pipe water stoppage or downcomer water stoppage is performed, the waterlevel in the PCV can be raised and fuel debris retrieval at the bottom of the PCV will be able to be
performed in the water. In this case, expected measures to be taken when the cooling water leaks
out from inside of the PCV are the same as ones taken in case of Unit 1.
When water stoppage is not performed, the water level in the PCV will be maintained at the samelevel of the current one and fuel debris retrieval will be performed in the air while the cooling water
is kept flowing on fuel debris. In this case, it is necessary to identify and repair the damaged partof the S/C to prevent leakage of the cooling water from inside of the PCV to the torus room via S/C.
C. Study on the water level in the PCV in Unit 3
The current water level in the PCV is assumed to be about 6.3m and is higher compared to thoseof Unit 1 and 2. Therefore, fuel debris at the bottom of the PCV is assumed to be submerged underwater. On the other hand, the water level in the PCV is required to be lowered for the side entry
method to be performed from the viewpoint of leakage prevention of the cooling water from theaccess part.
When either vent pipe water stoppage or downcomer water stoppage is performed, fuel debris
retrieval can be performed in the water, although the water level in the PCV is required to belowered to almost the same level (about 2m) of the current one in Unit 1 due to the reason described146

above. In this case, measures expected to be taken when the cooling water leaks out from insideof the PCV are the same as ones taken in case of Unit 1.
When water stoppage is not performed, the water level in the PCV will be maintained at the levelof the lowest part of the vent pipe joint or lower and fuel debris retrieval will be performed in the air
while the cooling water is kept flowing on fuel debris. In this case, it is necessary to study in advanceabout the decay heat of fuel debris, required amount of cooling water, etc., because a part of thefuel debris will be expected to be exposed.

147

Fig. A7-1 Water level at the bottom of the PCV and water stoppage in the PCV (Conceptual
diagram with assumed results)

Attachment 8 Terms related to radioactive waste management
IAEA Safety Requirements GSR-Part 5 62 explains 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. 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 terms used to refer to thefollowing steps that change waste characteristics
Pretreatment: a process before treatment (Treatment)
Treatment: a process intended to benefit safety andeconomics efficiency by changing waste characteristics
Conditioning: a process of manufacturing a waste
package suitable for handling, transportation, storage
and / or disposal

Fig. A8 Terms related to radioactive waste management (IAEA)63 and their translation examples(For the Japanese translation example, refer to the materials of the Japan Atomic Energy6465
62

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. GSR-Part5,
July, 2012)
63 IAEA, IAEA Safety Glossary Terminology Used in Nuclear Safety and Radiation Protection 2007 Edition, p.216,(2007).
64 A special committee of AESJ, “Processing and disposal of radioactive waste generated by the Fukushima
Daiichi NPS accident” 2013 Report, p.7 of Considerations for compiling waste information and solutions, March
2014
65 Seiya Nagao, Masafumi Yamamoto, 1 st Overview of Radioactive Waste Measurement, Radioactive Waste

149

Year

Current storage conditions

150

Temporary storage facility

Storage conditions after countermeasures are taken

BeforeCountermeasures

Absorption towerprotective
facility, etc.

Large-size waste storage

Absorption towerprotective
facility, etc.

Contaminated soil

Less than 0.005 mSv/h(outdoor accumulation)

0.005 to 0.1 mSv/h(outdoor accumulation)

Used protective
garment, etc.
1 to 30 mSv/h(soil-covered, etc.)
0.1 to 1 mSv/h(sheet curing)Cut down trees

Solid waste storage

Incinerator andvolume reduction facilities

Temporary storage facility
Start of operation of
expanded miscellaneous
solid waste incinerator*
Incineration of cut
down trees

Start of operation of
preprocessing equipment
for expanded incinerator*

Incineration of
flammable materials inrubbles▽Building 9 completed

Start of stabilization
processing of ALPS slurry

Study of disposal
method of water-treatedsecondary waste

Existing buildings

 Start of operation of large-size waste storage

Study of recycling and method ofrecycling each type of rubble ofless than 0.005 mSv/h▽Target of icompletion of storage ▽Buildings 10 to 13
completed one by one
only for contaminated soil

Start of operation of
volume reduction
facility

in FY 2023

Getting rid of
1 to 30 mSv/h (soilcovered, etc.)
in 2028

Getting rid of outdoor
accumulation of
0.005 to 0.1 mSv/h
in FY 2028

Getting rid of storage ofused protective garment
in FY 2028

Getting rid of contaminatedsoil targeted in FY 2026

Getting rid of
0.1 to 1 mSv/hsheet curing in 2024
in FY 2023

Getting rid of storageof cut down tree
in FY 2024

Reduction of volume of
metal and concrete0010,000 units

units

0
10,000 units
8,000
6,000
4,000
2,000
0
110,000

100,000

200,000

300,000

400,000

500,000

600,000

700,000 m3

0

100,000

200,000 m3

0

100,000
Incineration of
flammable materials
periodically generated

200,000

Contaminated soil

Incineration of used
protective garments,
etc.

300,000

Less than 0.005 mSv/h(outdoor accumulation)

400,000

0.005 to 0.1 mSv/h(outdoor accumulation)

Start ofoperation of
miscellaneous
solid waste
incinerator

600,000
500,000

700,000

・Amount of rubble to be generated when fuel debris is extracted is considered based on a decision on a policy of removing nuclear fuelfrom each reactor unit as indicated by the mid- to long-term roadmap.

Used protective
garment, etc.
1 to 30 mSv/h(soil-covered, etc.)

0.1 to 1 mSv/h(sheet curing)

800,000 m3

・Water treatment facilities including the nuclear reactors, turbine buildings, and tanks are excluded from prediction of amount of debris to be generated, on the assumption that they will remain.

Record of amount ofdebris generated(March 2017)
Approx. 345,000
m3
Expected amount ofdebris(through March 2029)
Approx. 409,000
m3

・Amount of debris to be generated is predicted, taking into consideration operation of miscellaneous solid waste incinerator (*).

Debris at upper part of reactor unit 1 (approx. 6,000 m3)
Debris at upper part of reactor unit 2 (approx. 6,000 m3)
Cut down trees (approx. 177,000 m3), used protective garment (approx. 2,000 m3)*, Other construction work (approx. 218,000 m3)

Unauthorized copy and transcription prohibited.
Tokyo Electric Power Company Holdings, Inc.

Cut down trees

Quantity of debris expected to generate
during decommissioning work・Future progress in decommissioning work and review of prediction of amount of debris to be generated will be reflected on thedecommissioning work as necessary.

・As much as possible, flammable materials are burned and metal and concrete are reduced in volume and stored in a building.

・Debris that has high influences on radiation at site boundaries is given a priority and stored in a building.

Attachment 9
Overall image of Storage Management Plan for the Fukushima Daiichi NPS66

After
Countermeasures

Storage conditions ofabsorption tower, etc.

Overview, The measurement of radioactive waste generated by operation and decommissioning of facilities,
AESJ 56(9), p.593, (2014).
66 TEPCO, Solid Waste Storage Management Plan for Fukushima Daichi Nuclear Power Station of Tokyo ElectricPower Company Holdings, Inc., June 2018 edition (issued in June 28, 2018)

Rubble Storage Area
Felled Tree Storage Area
Used protective clothing, etc.

Rubbles, etc.

Temporary Storage
for Used Cs absorption column
Waste Sludge Storage
Concentrated Liquid Waste (Completed)

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

Treatment Facility
Storage, etc.
Storage for reuse/recycle
Temporary Storage
for Used Cs absorption column
Waste Sludge Storage
Concentrated Liquid Waste (Completed)

Secondly Waste bywater treatment(b) Future storage condition of “rubbles, etc.” and “secondary waste generated by water treatment”
Fig.A9-1: Present and future storage conditions of “rubble, etc.” and “secondary waste generated
by water treatment” on site of the Fukushima Daiichi NPS

151

Attachment 10 Draft of technical map toward human resource development for decommissioning
reactors
Field Related to Electrical andMechanical Engineering

Stages of
NuclearIndustries

Nuclear Power
Station (fromPlanning andConstruction to
Operation)

Nuclear Fuel andWaste (from
Front-end to
Back-end of Fuel
Cycle)

Decommissioningof Nuclear
Facilities such as
Non-accidental
nuclear Power
Plant

Field Related to Plant Engineering

Field Related to Civil, Geotechnicaland Architectural Engineering

Field Related to Chemical andMaterial Engineering

Field Related to Nuclear Engineering(including Environment andRadiation)

Other Fields(Examples of Field)
Mechanical Engineering,Electrical and Electronic Engineering,Electronic Material,
Instrument Engineering and System
Control Engineering,
Signal and Image Processing,Information Engineering(Examples of Field)
Chemical Engineering,
Industrial Chemistry,
Chemical Plant,
Plant Engineering,
Process Engineering,
Water Chemistry(Examples of Field)
Civil Engineering,
Geotechnical Engineering,Architectural Engineering,
Groundwater Engineering,
Deep Geological Science,
Geohydrology, Soil Science,
Solid State Engineering (Concrete,etc.)(Examples of Field)
Material Engineering,
Solid State Engineering (Metal, etc.),
Metal Corrosion Engineering,
Nuclear Fuel Engineering,
Actinoid Chemistry,
Radiation Chemistry(Examples of Field)
Nuclear Reactor Physics,
Nuclear Reactor Design and
Structure,
Simulation Engineering,
Criticality Control,
Maintenance Engineering,
Radiation Science,
Radiation Effects,
Radiation Protection,
Environmental Radiation,
Kinetics theory(Examples of Field)
Project Management,
Legal Affairs,
Labor Management,
Financial Affairs, PA,
Risk Communication,
External Relations,
Information System,
Radiation Management

Features

Work Process(Examples of Basic Technology)
Electric System Design Engineering(Instrument Engineering and System
Control Engineering, Heavy Current
and Weak Current ElectricalEngineering, MechanicalEngineering),
Imaging,
Human Interface,
VR Technology,
Visualization Technology and
Recognition Support,
Mechanical Crafting (Cutting,
Shearing, etc.),
Remote-control Systems (Robotics),
Resistance to Radiation(Examples of Basic Technology)
Plant Designing,
Manufacturing,
Construction,
Assembly,
Piping,Plant Design,
Systems Designing,
Plant Instrumentation,
Boundary Building,
Negative Pressure Management,
Filter Design,
Engineering Flow

(Examples of Basic Technology)
Architectural Design,
Building and Construction Diffusion,
Demolishing Technology,
Non-destructive Inspection(Concrete, etc.),
Structural Soundness,
Boring,Underground Diffusion Control andEnvironment Restoration,
Property of Concrete(Examples of Basic Technology)
Welding, Corrosion Evaluation,
Metal Physics,
Nuclear Reactor Material,
Fuel Design, Radiolysis,
Rust Prevention,
Long-term Soundness Evaluation,
Volume Reduction and Stabilizationand Solidification,
Metal Recycling Technology,
Property Grasping(Examples of Basic Technology)
Designing, Operation,
Subcriticality Monitoring and Control,
Analysis Codes and Simulation,
Human Factor, Burn-up Analysis,
Nuclide Identification, Protection,
Environmental Impact Assessment,
Exposure Evaluation,
Radiation Evaluation,
Evaluation of Emission/Scatteringand Migration,
Future Estimation (Long-term
Dynamics),
Transfer Behavior(Examples of Basic Technology)
Project Management (Planning, Risk Evaluation andManagement, Resource Evaluation, EVM, etc.),
Laws and Ordinances,
Public Administration Affairs,
Accounting, Procurement, Contract, Workforce Estimation,
Labor Safety,
Management of Records, Documents and Knowledge,
Decision Making, Problem Solution, Strategic Thinking,
Leadership, Communication Capability,
Negotiation and Influence,
Presentation Capability, Coaching,Foreign Language Ability,
ICT and Systems Engineering,
Facility Maintenance and Management,
Other sorts of Back-office Operation,
Nuclear Material Accounting/Nuclear Material Control/
Safeguards and Inspection Affairs

Sustaining of
Criticality/Subcriticality,
Existence of Defense in Depth,
Remote Controlling in Some
Parts,High Temperature and High
Pressure,
Long-term Operation

Design and Construction,
Equipment Design,
Manufacturing and Assembly,
Power Generation (Turbine, etc.),
Operation and Maintenance,
Fuel Replacement and Storage

Electrical System (Heavy Current andWeak Current),
Instrumentation System,
Electricity Generators (Turbine, etc.),
Maintenance and Checkup (Electrical
Equipment, etc.),
Non-destructive Inspection andAnalysis (Devices)

Systems Designing (Process
Design),Layout Design,
Structure Design (Process
Equipment, Piping, etc.),
Thermal Hydraulics Design,Circulation System for Power,
Generation (Heat Exchanger, etc.),
Confinement System (without αnuclides),Air Conditioning Design

Geotechnical Evaluation,
Seismic Design (Structural
Soundness),
Large-size Equipment Assembly,
Maintenance and Inspection of
Buildings, etc.

Reactor Design (Materials),
Systems Designing (Materials),
Structural Soundness of Structures,
Maintenance and Checkup(Structures and Systems)

Core Design,
Safety Design,
Shielding Design,
Thermal Analysis and Cooling
Evaluation,
Plant Operation,
Response to Emergency,
Fuel Storage (Dry and Wet)

<Legal Affairs, Finance, and Back-office>Financial Management, Contract Administration,
Purchase Administration, Electricity/Gas/Water Supply,Intellectual Property Management,
Information System Development and Operation, Security,
Management of Records and Documents

Keep Criticality;
Existence of Defense in Depth;High Radiation Field;
Basically Remotely Controlled,
Long-term Operation;
Mechanical Crafting in Some
Parts;
Emission of Dusts Containing αnuclides;
Numerous Chemical Processes

Design and Construction,
Equipment Design,
Manufacturing and Assembly,
Operation and Maintenance,
Transportation of Nuclear Materials
and Fuels,
Fuel Processing and Manufacturing,
Fuel Storage (Dry and Wet),
Spent Fuel Reprocessing,
Waste Processing and Disposal

Electrical System (Heavy Current andWeak Current)
Instrumentation System
Nuclear Fuel and Spent Fuel Handing
Equipment
Remote Crafting and Measuring
Instruments
Cutting and Dismantlement (Spent
Fuel, etc.)
Waste Vacuuming and Removal
Equipment
Decontamination Technology (Device
Development)

Systems Designing (Process
Design),Layout Design,
Structure Design (Process
Equipment, Piping, etc.),
Confinement System (for α-nuclides)
Measures for Scattered Particle
Dusts (as System Development),
Air Conditioning Design,
Spent Fuel Reprocessing,
Decontamination of Systems (Using
Chemical Methods),
Waste Processing (Volume
Reduction and Stabilization),
Maintenance and Inspection of
Systems

Geotechnical Evaluation,
Seismic Design (Structural
Soundness),
Large-size Equipment Assembly,
Decontamination Technology(Concrete, etc.),
Cement and Asphalt Solidification,
Storage Facility and Disposal Site
Design (Thermal Design, etc.),
Maintenance and Inspection of
Buildings, etc.

Fuel Design (Material),
Fuel Processing and Manufacturing
Systems Designing (Material),
Maintenance and Inspection of
Systems,
Decontamination Technology of
Systems,
Material Design of Storage Cans,
Waste Storage Management(Chemical Stability and Long-term
Change Prediction)

Fuel Design (Heat, Neutron),
Safety Design,
Shielding Design,
Thermal Analysis and Cooling
Evaluation,
Criticality Control,
Radiation Measurement,
Radiation-resistant Design,
Nuclear Material Accounting(Criticality),
Measures for Scattered Particle
Dusts (Against Exposure) ,Heat Calculation of Waste,
Waste Shielding Storage andManagement (Criticality),
Radioactivity and Material BalanceManagement (Inventory Evaluation)

Known Internal Conditions,
Gradual Removal of Defense in
Depth,
Remote Controlling in Some
Parts,
A lot of Mechanical Crafting,
Emission of Dusts,
Long-term Project

Retrieval of Spent Fuel,
Decontamination of Inside of
Buildings,
Waste Sampling,
Dismantlement of Reactor Area
Dismantlement of Buildings, etc.,
Waste Processing and Disposal,
Environment Restoration (Soil
Restoration, etc.)

Remote Crafting and Measuring
Instruments,
Cutting and Dismantlement(Structures),
Waste Vacuuming and Removal
Equipment,
Decontamination Technology (Device
Development),
Non-destruction Inspection andAnalysis (Devices),
Waste Sampling Devices,
Dismantlement of Steel and Concrete
Structures

Confinement System (Without αnuclides),Decontamination Systems (Using
Chemical Methods),
Measures for Scattered Particle
Dusts (as System Development),
Waste Processing (Volume
Reduction and Stabilization)

Structural Soundness of Buildings,
Decontamination Technology(Concrete, etc.)
Recycling of Concrete, etc.,
Dismantlement of Buildings, etc.,
Storage Facility and Disposal Site
Design (Thermal Design, etc.),
Environment Restoration (Soil, etc.)

Decontamination Technology(Structures),
Property Grasping of Waste (Radiochemical Analysis),
Waste Storage and Management(Chemical Stability and Long-term
Change Prediction)

Radiation Measurement (Handling
Operations, etc.),
Radiological Protection,
Nuclide Identification,
Shielding Design,
Measures for Scattered Particle
Dusts (Exposure Prevention),
Heat Calculation of Waste,
Waste Storage Management(Criticality, Shielding),
Radioactivity and Material BalanceManagement (Inventory Evaluation),
Clearance Level Assessment<Labor, Organization Management, Safety Management>
Labor Safety,
Response to Emergency and Accident,
Fire and Explosion Prevention,
Failure Protection Measures,
Operation Risk Assessment,
Crisis Management, Assessment (Audit),
Organization Management,
Equipment Management (Operation),
Maintaining Motivation, Incentive Design,
Technology Succession, Education and Training, HR(human Resource Management and Development Plan)<Project Management>
Strategic Vision Formulation, Scenario Study,
Overall Plan Formulation, Planning and Application of
Decommissioning Plan of the Site,
Cost Evaluation and Management, Time Management,
Risk Assessment and Management, Process Control,
Material Amount Management, Material Procurement,
Expenditure Prioritization, QA, Land Planning,Work Space Management, Special Procurement (Planningand Implementation),
Redundancy Securing, Construction Management,
Quality Control<Engineering>
Process Planning, Construction Plan, Maintenance,
Operation Instruction, Planning and Training Simulator,(Remote-controlled) Equipment Operation,
High-place Work Plan
<Matters Common to Nuclear Power Facilities>
Site Planning and Regional Relations,
Environmental Impact Assessment,(Environmental) Monitoring,
Exposure Management and Assessment,
Radiation Management Equipment,
Anti-terrorism Measures and Nuclear Material Protection,
Designating and Canceling Radiation Controlled Areas,
Equipment Maintenance,
Planning, Managing and Implementing Fuel and Waste
Transportation,
Compliance with Safety Regulations and Permission andAuthorization,
Safeguards Administration<Research and Development Environment>
Precedent Case and Studies, Environment for Innovation,
Development and Maintenance of Foundations for R&D,
Strategy for Mock-up Facilities and Testing,
Collaboration with Researching Institutions<Social Relations>
Securing Consistency with Past Cases,
Relations with Stakeholders, Social Impact,
Public Relations, Treating Visitors

Decommissioningof Fukuoka
Daiichi NPS andOther Accidentdamaged NPS

Keep Criticality,
High Radiation Field,
Basically Remotely Controlled,
Loss of Defense in Depth,
Unclear Internal Conditions,High Uncertainty,
A lot of Mechanical Crafting,
Emission of Dusts Containing αnuclides,
Long-term Project,
Large-scale Project<Site Stabilization>
Countermeasures for Groundwaterand Contaminated Water,
Securing Cooling Function (Water
Circulation),
Retrieval of Spent Fuel
<Improvement of Operation
Environment>
Decontamination of Inside of
Buildings,
Structuring Inside PVC

Spent Fuel Handling Equipment

Measures for Cooling Water andContaminated Water (Circulation
System)Electrical System (Heavy Current and Decontamination of Systems (Using
Weak Current),
Chemical Methods)
Instrumentation System,
Decontamination Technology (Device
Development)<Investigation andRemote Crafting and Measuring
Preparation for Fuel Debris Instruments (Investigating
Retrieval>Instruments)
PCV/RPV Internal
Investigation,
Waste Sampling,
Fuel Debris Sampling

Systems Designing (Process
Design),Layout Design
Structure Design (Process
Equipment, Piping, etc.),
Confinement System (for α-nuclides)
Measures for Scattered Particle
Dusts (as a System Development)
Air Conditioning Design<Fuel Debris Retrieval>
Fuel Debris Retrieval,Transportation and Storageof Fuel Debris,
Processing and Disposal of
Fuel Debris

Radiation Measurement (System
Development),
On-site Sample Analysis
Non-destructive Inspection andAnalysis (Devices),
Remote Crafting and Measuring
Instruments (Retrieving Instruments),
Cutting and Dismantlement(Structures),
Sampling Devices for Fuel Debris,
Retrieving Devices for Fuel Debris,
Waste Sampling Devices,
Waste Vacuuming and Removal
Equipment

Fuel Debris Processing
Waste Processing (Volume
Reduction and Stabilization)<Dismantlement andEnvironment Restoration>
Reactor Area Dismantlement
Dismantlement of Buildings,
etc.,
Waste Processing andDisposal,
Environment Restoration(Soil Restoration, etc.)

Remote Crafting and Measuring
Instruments (Dismantling Devices)
Cutting and Dismantlement(Structures)
Waste Sampling Devices
Waste Vacuuming and Removal
Equipment

Waste Processing (Volume
Reduction and Stabilization)

Measures for Groundwater

Measures for Contaminated Water(Processing)

Thermal Analysis and Cooling
Evaluation

Decontamination Technology(Concrete, etc.)

Unravelling Contamination
Mechanism,
Decontamination Technology(Structures),
Investigating the Chemical Effect of
Seawater,Corrosion Prevention (PCV/RPV)

Making Contamination Maps

Systems Designing (Material)
Property Grasping of Fuel Debris(Chemical Analysis)
Material Design of Storage Cans

Severe Accident Progression
Analysis,
Property Grasping of Fuel Debris,
Thermal Analysis and Cooling
Evaluation,
Radiation-resistant Design,
Measures for Scattered Particle
Dusts (Against Exposure)

Radiation Measurement(Development of Components),
Aging Characteristics of Fuel Debris,
Property Grasping of Waste(Radiochemical Analysis),
Waste Storage and Management(Chemical Stability and Long-term
Change Prediction)

Radiation Measurement (Handling
Operations, etc.),
Radiological Protection,
Nuclide Identification,
Criticality Control,
Radioactivity and Material BalanceManagement (Inventory Evaluation),
Fuel Debris Storage andManagement (Criticality, Shielding),
Heat Calculation of Waste,
Waste Storage and Management(Criticality, Shielding)

Property Grasping of Waste(Radiochemical Analysis),
Waste Storage and Management(Chemical Stability and Long-term
Change Prediction)

Heat Calculation of Waste
Waste Storage and Management(Criticality, Shielding) ,Clearance Assessment

Structural Soundness of Buildings,
Structural Soundness ofContainment Vessels,
Water Sealing at Building Borders,
Water Sealing at PCV Borders

Dismantlement of Buildings, etc.,
Cement and Asphalt Solidification,
Storage Facility and Disposal Site
Design (Thermal Design, etc.),
Environment Restoration (Soil, etc.)Implementation of all the above actions, keeping in
mind Safe, Proven, Efficient, Timely, and Fieldoriented work under the environment anduncertainty at 1F that are different from normal
environment, with unknown shape, composition, andinternal condition (especially, Project Management)

Response to unprecedented regulations andSafeguards



Players well-versed in more than one field arenecessary, considering that necessary humanresources differ and processes change dependingon the period.

(Note) In the section of “Decommissioning of the Fukushima Daiichi NPS and other Accident-damaged NPS”, technical issues that have similar
entries in the other stages of nuclear industries are highlighted by yellow color, and other issues that have similar entries as well but havedifferent preconditions or require higher levels of response are highlighted by green color. Technical issues with no similar entries arehighlighted by red color. (The corresponding entries in the other stages of nuclear industries are also highlighted by yellow or green colors.)(Source: Material 2-4-2, 6th Meeting of Decommissioning R&D Partnership Council (December 12, 2017)152

Attachment 11 Current Progress of R&D in the Project of Decommissioning and Contaminated
Water Management, etc.
1. Grasping state inside reactor, grasping properties of fuel debris, investigation of reactor inside
1-(1) Upgrading level of grasping 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 technologies for grasping and analyzing properties of fuel debris (FY 2017 2018)(Related projects) Development of technologies for characterization and processing of fuel debris
using mock debris (FY 2011 - 2014)
Properties analysis of actual debris (FY 2014)
Grasping properties of fuel debris (FY 2015 - 2016)
Construction of material accountancy method related to fuel debris (FY 2011 2013)
1-(3) Development of technologies for in-depth investigation of PCV inside (FY 2017 - 2019)(Related projects) Development of investigation technology of inside of PCV (FY 2011 - 2013)
Development of investigation technology of inside of PCV (FY 2014 - 2015)
Development of investigation technology of inside of PCV (FY 2016 - 2017)
1-(4) Development of investigation technology of inside of RPV (FY 2016 - 2019)(Related projects) Development of investigation technology of inside of RPV (FY 2013 - 2015)
1-(5) Development of technologies for the detection of fuel debris inside reactors (using muon) (FY
2014 - 2015)
2. Retrieval of fuel debris
2-(1) Advancement of retrieval method and system of fuel debris and internal structures (FY 2015
- 2018)(Related projects) Development of technologies for retrieval of fuel debris and internal structures(FY 2014)
2-(2) Advancement of fundamental technologies for retrieval of fuel debris and internal structures(FY 2017 - 2018)(Related projects) Development of fundamental technologies for retrieval of fuel debris andinternal structures (FY 2015 - 2016)153

2-(3) Development of sampling technologies for retrieving fuel debris and internal structures (FY2017 - 2018)
2-(4) Development of closed circulation systems for water inside PCV (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 and PCV(FY 2011 - 2013)
Development of evaluation methods for the structural integrity of RPV and PCV(FY 2014 - 2015)
2-(7) Development of corrosion inhibition technology for RPV and PCV (FY 2016)
2-(8) Development of criticality control technologies of fuel debris (FY 2012 - 2017)
2-(9) Development of technologies for non-destructive detection of radioactive material
deposited in 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 dose rate reduction plan (FY 2012 - 2013)
2-(12) Development of technologies for containing, transportation and storage of fuel debris (FY2016 - 2019)(Related projects) Development of technologies for containing, transportation and storage of fueldebris (FY 2014 - 2015)
3. Waste management
3. Research and development of processing and disposal of solid waste (FY 2017 - 2018)(Related projects) Development of technologies for processing and disposal of secondary wasteby treatment of contaminated water (FY 2012)
Development of technologies for processing and disposal of radioactive waste(FY 2012)
Research and development of processing and disposal of solid waste (FY 2013
- 2014)
Research and development of processing and disposal of solid waste (FY 2015
- 2016)154

4. Spent fuel management
4-(1) Evaluation of long-term integrity of fuel assembly retrieved from spent fuel pool (FY 20152016)(Related projects) Evaluation of long-term integrity of fuel assembly, etc. retrieved from spent fuelpool (FY 2012 - 2014)
4-(2) Investigation of method for processing damaged fuel, etc. retrieved from spent fuel pool (FY
2013 - 2014)
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 (ice wall) (FY 2014)
5-(4) Development and verification of high-performance multi-nuclide removal equipment (highperformance ALPS) (FY 2014)

155

Project objectives
In the decommissioning process of the Fukushima Daiichi NPS, with the aim of confirming the stability of plants and determining methods toretrieve fuel debris, comprehensive analysis and evaluation were performed to precisely comprehend the state of fuel debris and fission
products (FPs) in reactors. This project collected and arranged through analysis and evaluation, the data and information required for theassessment of impacts on the integrity of structures, such as PCVs, criticality assessment, and assessment of FP behaviors at the time of fueldebris retrieval, as well as information about the location and distribution of fuel debris and FPs for each unit.
1. Project details and progress
The following describes the results this project produced so far based on results from the related projects described in Section 2 below.
(1) Comprehensive analysis/evaluation of conditions inside reactors
Information summary maps which summarizes the various information from the RPV, PCV, and R/B, were created using the assumed
conditions inside the RPV/PCV of all units. The information was then comprehensively analyzed and evaluated, and used to provide anestimation diagram of the fuel debris distribution (Fig. 1), an estimation of the FPs distribution, and an estimation of the radiation dose
distribution. To effectively advance these efforts toward, a database was developed by comprehensively collecting and arranging, datameasured with the Fukushima Daiichi NPS and the results of on-site investigations.
(2) Reduction of uncertainties using analysis methods
Sensitivity and other analysis taking into consideration the boundary conditions and an analysis model for events assumed to have occurred
within the reactor were conducted using the accident progression analysis codes (MAAP, SAMPSON), and which then provided expertise thatwill contribute to the more comprehensive analysis and evaluation. A simulated fuel assembly plasma melting tests (Fig. 2) were performed,
which provided expertise leading to reducing the uncertainties of events such as core meltdown and relocation in fuel assembly systems.
(3) Evaluation of FPs chemical properties
In evaluating the FPs chemical properties, and with a focus on Cs, which islargely contributing to the radiation dose level during the decommissioning, astudy has commenced upon the distribution of Cs and its chemical
properties, including identifying the chemical species that need to beconsidered in addition to standard chemical species such as CsI and CsOH,
and the insoluble Cs particles whose production has been confirmed in theenvironment around the Fukushima Daiichi NPS , and the possibility ofuneven Cs associated with the reactions in the upper structures of the RPVs.
In addition, on-site test samples have been analyzed, and a study of theFig. 2composition and spatial distributions of uranium and FPs conducted from the
aspect of identifying the conditions inside the reactor (Fig. 3).
(4) Utilization of domestic/overseas knowledge through international joint
research
The international Joint Research, OECD/NEA BSAF Phase 2 Project involves
the sharing of accident progress scenarios and plant information with
Fig. 1: Example of diagram of estimatedFig. 3
fuel debris distribution
overseas organizations utilizing the database, and the evaluation results of
the accident progression and plant/FP distribution by participating organizations being compared to actual measured values and the results ofon-site investigations. It was confirmed that the evaluation results of the FP release amounts was consistent with the amounts released into theenvironment. The estimated accuracy of the fuel debris distribution was improved via more through knowledge of the progression of theaccident. The expertise obtained through discussions with overseas organizations were utilized in the comprehensive analysis and evaluations.
Implemented byInternational Research Institute for Nuclear Decommissioning (IRID) (FY 2013 -), the Institute of Applied Energy
2. Related projects
The following describes the results of previous related projects.
○ Advancement of accident progression analysis technology for assessing conditions inside reactor (FY 2011)-Plans for future projects have been developed.
○ Assessing conditions inside reactor by advancement of accident progression analysis technology (FY 2012 - 2013)- To improve severe accident progression analysis codes, significant phenomena for the progression of severe accidents were identified and aphenomena identification and ranking table (PIRT) was complied. Sensitivity analysis results have also been revised.
- Each model of analysis codes (MAAP and SAMPON) has been improved, and the progression of severe accidents has been analyzed.
- Specifically, debris spreading behaviors on the floor of a PCV and the chugging phenomenon in a suppression chamber were evaluated.
- The OECD/NEA BSAF Project was operated by the Institute of Applied Energy as the operating agent. Common analysis conditions were
established. Meetings were held, and information was shared with participating organizations.
○ Assessing conditions inside reactor through application of severe accident analysis code (FY 2014)- To improve severe accident progression analysis codes, based on sensitivity analysis with MAAP, the PIRT was evaluated and thesignificance ranking was reset.
- Each model of analysis codes (MAAP and SAMPON) has been improved, and the progression of severe accidents was analyzed.
- Specifically, thermal hydraulic analysis related to the spreading of debris and molten core-concrete interaction (MCCI) model has been
upgraded.
- Phase 2 has been started after summarization of OECD/NEA BSAF Phase 1 Project.
○ Advancement of grasping conditions inside reactor by accident progression analysis and actual data, etc. (FY 2015)- To improve severe accident progression analysis codes, each model of analysis codes (MAAP and SAMPSON) has been improved, andsevere accident progression analysis and sensitivity analysis were performed.
- An analysis code for the evaluation of MCCI has been developed, and the behavior of fuel debris fallen in PCV was predicted.
- Specifically, melting test of penetrating tubes through RPV was performed at the Korea Atomic Energy Research Institute and data for theverification of analysis results was obtained.
- Comprehensive analysis and evaluations were performed using the results of severe accident progression analysis, internal investigations,and results of other R&Ds, and the distribution of fuel debris was predicted.
FY2011FY2012FY2013
FY2014FY2015FY2016FY2017
Assessing conditions inside
Advancement of grasping
reactor through application of conditions inside reactor by
severeaccident
analysiscode
accident progression analysis
Assessing conditions inside reactor by
Upgrading level of grasping
and actual data, etc.
advancement of accident progression
state inside reactor
analysis technology

Fig. A11-1-(1): Upgrading level of grasping state inside reactor

156

Project objectives
In the decommissioning of the Fukushima Daiichi NPS, the properties of fuel debris will be analyzed and evaluated to obtain data to contributeto the comprehensive analysis and evaluation of state inside reactors, the determination of methods to retrieve fuel debris and internal
structures, and the study of containing, transportation and storage of fuel debris. Therefore, tests will be performed using simulated debris, andtechnologies required for the analysis and measurement of fuel debris will be developed for the fuel debris retrieval.
1. Project details and progress
The following describes the results this project produced so far based on results from the related projects described in Section 2 below.
(1) Assumption of fuel debris properties
Molten core-concrete interaction (MCCI) was simulated, andthe results of the analysis of the MCCI products, previous
Upper partsevere accidents at nuclear facilities, and information from
Upper partcrustoxide layer
related documents were surveyed. The information has been
Side bordercompiled in the Fuel Debris Characteristics List, which is
Lower part
updated.
oxide layer(2) Characterization via use of simulated debris
Fundal border(metal)In international cooperation with the CEA in France, a largescale MCCI test was performed in view of the condition of the
In the case ofsample madeFukushima Daiichi NPS, and information of the elementfrom oxide mainly,
distribution of products (Fig. 1), crystal structures and
U-Ze oxide andhardness was obtained. Re-hydration of deteriorated cement
Cr-rich oxide, etc.
at high temperature, drying behavior of radioactive dust andare deposited in
the release behavior of medium volatile fission products
Si-rich matrix(FPs) that are thought to release in drying of fuel debris are(containing Ca,
Metal layer is made from
evaluated.
Al).
iron-rich alloy mainly(3) Development of elemental technologies for fuel debris
Fig. 1: Element distribution of products by large-scale MCCI test
analysis
The following technologies are under study and development: a multi-element simultaneous analysis method using inductively coupled plasma
atomic emission spectrometry (ICP-AES), a quantitative porosity assessment method using computerized transverse axial tomography (CT),and a multi-nuclide rationalized analysis method using inductively coupled plasma mass spectrometry (ICP-MS). Study on the transportation offuel debris samples is underway.
(4) Estimation of aging properties of fuel debris
To study the change by aging in behaviors of fuel debris, testing methods to evaluate the change by aging, and the possibility of chemical andphysical changes are being studied.
(5) Estimation of behaviors of fuel debris fine particles (relocation characteristics in air/under water)
Fundamental data on physical/chemical characteristics and behaviors of radioactive fine particles generated from fuel debris is being
surveyed and collected.
Implemented by(Comprehensive proposal) International Research Institute for Nuclear Decommissioning (IRID) (FY 2013 -)(Partial proposal) Estimation of aging properties of fuel debris: TENEX (FY 2017 -)2. Related projects
The following describes the results of previous related projects.
○ Development of technologies for characterization and processing of fuel debris using simulated debris (FY 2011 - 2014)- Observations of microstructure and measurement of hardness were carried out using TMI-2 debris stored in the JAEA.
- The applicability of the alkaline solution was verified to evaluate analysis technologies.
- In cooperation with the NNC in Kazakhstan, solidified melt bulk was prepared from metals/ceramics with uranium oxide.
- The hydration/drying characteristics were evaluated using porous ceramics.
○ Properties analysis of actual debris (FY 2014)- Using simulated debris, solution, quantitative analysis, simplified quantitative analysis and modification of analyzers (SEM, EDX, and WDX)
were examined.
- The applicability of high-sensitive active neutron detection technology was studied as a technology to detect nuclear fuel materials.
- The applicability to nuclide analysis and handling required to accept fuel debris, samples splitting, pre-treatment, analysis and posttreatment were evaluated.
○ Grasping properties of fuel debris (FY 2015 - 2016)- The properties of fuel debris required to evaluation was identified. Micro-properties, macro-properties and cross-sections of MCCI products
were added in the Fuel Debris Characteristics List.
- Data on physical properties required for designing equipment for retrieval, containing, and storage of fuel debris was obtained. Drying
properties and changes in characteristics during oxidation reaction were evaluated.
- A multi-element simultaneous analysis using ICP-AES and a multi-nuclide rationalized analysis using ICP-MS were examined. The
development of a quantitative porosity evaluation by X-rays CT was continued.
- Data on the hardness and particle size distribution of the solidified melt bulk prepared from metal/ceramics in Kazakhstan was obtained.
○ Construction of material accountancy method related to fuel debris (FY 2011 - 2013)- The amount of fuel materials in the TMI-2, nuclear fuel material measurement technologies used in the Chernobyl NPS and material
accountancy procedures were collected and compiled.
- Nuclear fuel materials measurement technologies applicable to the Fukushima Daiichi NPS were evaluated.
- Under the safeguards cooperation agreement between the JAEA and the DOE, a cooperative framework with U.S. national research
institutes has been established.
FY2011FY2012FY2013
FY2014FY2015FY2016FY2017FY2018Development of technologies for characterization and processing of fuel debris
using simulated debris
Properties analysis of actual debris
Construction of material accountancy method related to fuel debris

Grasping properties of fuel debris

Development technologies for grasping
and analyzing properties of fuel debris

Fig. A11-1-(2): Development technologies for grasping and analyzing properties of fuel debris

157

Project objectives
To contribute to the determination of methods to remove fuel debris, equipment and more sophisticated investigation technologies will be
developed and verified to more accurately and more widely investigate the distribution of fuel debris in PCVs and circumstances in or outsidepedestals.
1. Project details and progress
The following describes the results this project produced so far based on results from the related projects described in Section 2 below.
(1) Establishment of investigation plans and development plans
Detailed investigation plans and development plans for individual reactors have been established, based on the results of the latest
investigations of the inside the PCVs of Units 1 to 3.
Penetration(2) Development of access/investigation equipment and elemental technologies
Isolation valvei) Establishing an access route into PCV through X-6 penetration,access/investigation equipment
Equipment related to the establishment of an access route to the inside of the
PCV through X-6 penetration was studied, and detailed design is beingArmconducted. Then, prototypes will be fabricated and verified in plants (Fig. 1). In
X-6 penetration connection structure
addition, detailed design is being conducted for arm-type access equipment to
Fig. 1: Connection structure of X-6 penetration
access the inside of the PCV through X-6 penetration and to identify the range offuel debris based on the shapes of debris and distribution of gamma rays. Then,
Gratinga prototype will be fabricated and verified in plants.
Inner doorii) Establishing an access route into PCV through X-2 penetration,access/investigation equipment
rail
Equipment related to the establishment of an access route to the inside of the
PCV through X-2 penetration is being studied. Detailed design and prototypes
Hydraulic cylinder for bending pipe
are being fabricated, and they are under verification in plants (Fig. 2). In addition,
Mobile carriage
underwater ROV(remotely operated vehicle) that can access the inside of the
PCV through X-2 penetration and has measuring technologies suitable for the
Fig. 2: Boring device of X-2 penetration
purpose of the investigation, which takes into account the deposits over a wide
inner door, grating, etc.
area on the basement floor (distribution and thickness of deposits, and distribution of fuel debris under the deposits, etc.), is being designed indetail, and a prototype will be fabricated and verified in plants.
iii) Verifying the applicability of elemental technologies
In association with the update of investigation plans, measuring technologies are under review, and detailed design of measuring equipment
and the fabrication of a prototype are underway. The applicability of access/investigation equipment will be verified in combination.
(3) Mock-up testing, training, and on-site verification
Mock-up testing, training, and on-site verification will be carried out (FY 2019) with the aim of conducting detailed investigations by accessing
PCVs through X-6 (Unit 2) penetration and X-2 penetration (Unit 1).
Implemented byInternational Research Institute for Nuclear Decommissioning (IRID)

2. Related projects
The following describes the results of previous related projects.
○ Development of investigation technology of inside of PCV (FY 2011 - 2013)- Targets and items to be investigated and technical challenges were identified, and access routes for the implementation of the investigations
were studied.
-In FY 2012, the inside of the Unit 1 PCV was accessed through X-100B penetration and an investigation was carried out with a camera.
Based on the results of the investigation, investigation equipment has been developed for prior investigation of the inside of PCVs (collecting
videos and data on radiation dose and temperature).
- In FY 2012 and 2013, the inside of the Unit 2 PCV was accessed through X-53 penetration, and investigations were carried out with camera.
Since the radiation level around the X-6 penetration was much higher than the estimate, it was necessary to take measures to reduce theradiation level. Therefore, investigations through the X-6 penetration were postponed.
- Design and fabrication of prototypes of equipment to remove Unit 2 X-6 shielding blocks and equipment for investigations through X-6
penetration have been started.
○ Development of investigation technology of inside of PCV (FY 2014 - 2015)(1) Technologies to access the inside of pedestals
X-100B- Verification tests at a plant for investigation
penetration
Validation point of A2 investigationequipment for the Unit 2 A2 Investigation
Gratingequipment (inside of Unit 2 pedestal)(investigation of the situation on the platform in
the pedestal) have been completed.
- For further investigation of the inside of thepedestal (A3 Investigation), prototypes of
X-6 penetrationequipment to open the hatch of X-6 penetration
were made and tested.
(2) Technologies to remove shielding blocks
Pedestal- Shielding block removal equipment was: Access routedeveloped, and during the period from June
Validation point of shielding block
through July of 2015, the removal of shielding
Validation point of B1 investigation
removal equipment (front of Unit 2
blocks in front of the Unit 2 X-6 penetration was
equipment (outside of Unit 1 pedestal)X-6 penetration)completed (on-site verification).
Equipment body(3) Technologies to access the outside of theGrip partpedestal-Verification tests at a plant for investigation
Simulated block
equipment for Unit 1 B1 Investigation(investigation, through X-100B penetration, of
circumstances on the grating on the first floor in
the PCV) have been completed, and on-site
The figure of upper-right showed the PCV inside of Unit 1, but in validation A2 investigation wasverification was performed in April 2015.
performed in Unit 2, and B1 investigation was performed in Unit 1.

158

It was confirmed that there was no significant damage to the existing equipment in the PCV, and data on radiation levels and temperatures in arange about three-fourths of an orbit above the grating on the first floor was obtained.
- Methods for further investigation (B2 Investigation) was drafted.
(4) Fuel debris measuring technologies- Methods to upgrade measuring technologies were studied for the fuel debris investigations that will start after FY 2017.
○ Development of investigation technology of inside of PCV (FY 2016 - 2017)(1) Developing access/investigation equipment and systems for particular parts- Access/investigation equipment for Unit 1 B2 Investigation (investigation, through X-100 B penetration, of the basement floor outside thepedestal) was developed, and on-site verification was performed in March 2017. It was confirmed that there was no significant damage to theinside wall of the PCV and to existing equipment and that there were deposits on the bottom of the PCV. The radiation level increased as the
distance from the surface of the deposits decreased.
- Access/investigation equipment for Unit 2 A2 Investigation (investigation, through X-6 penetration, of circumstances on the platform in thepedestal) and A2’ Investigation (investigation, through X-6 penetration, of circumstances below the platform in the pedestal) has beendeveloped, and A2 Investigation was carried out in January 2017 and A2’ Investigation in January 2018. Pieces of gratings that had fallen were
found on the platform, and there was a part of fuel assemblies on the bottom of the pedestal. There were also deposits around them, which
seemed to be fuel debris.
- Underwater ROV has been developed to access the inside of the PCV through Unit 3 X-53 penetration and to investigate the circumstances
inside the pedestal. Then, on-site verification was performed in July 2017. It was confirmed that several structures were damaged and that
there were melted objects solidified and clump deposits.
- A concept for how to establish access routes for the next phase of investigations was examined, and elemental tests were performed. As
access equipment, conceptual studies and elemental tests were carried out for underwater ROV, floor traveling-type equipment, and arm-typeaccess equipment.
(2) Formulating and updating investigation plans and development plans for next phase investigations- Conceptual studies and elemental tests were performed for the Unit 1 investigation using underwater ROV that accesses the inside of the
PCV through X-2 penetration and for Unit 2 investigation using arm-type access equipment that accesses the inside of the PCV through X-6
penetration.
- Conceptual studies and elemental tests were performed for the following technologies that can be applied in combination with the above
access/investigation equipment: technology to measure dimensions, distances, and shapes; visual investigation technologies; and radiationlevel measuring technologies (measurement of gamma ray and measurement of radiation released from debris).
: Access route
Appearance of B2 investigationequipment (outside of Unit 1 pedestal)Appearance of A2 investigationequipment (inside of Unit 2 pedestal)

X-100B
penetration

X-6
penetration

Drive thruster

Structure
Pedestal

Front camera
Pedestalunderwater
Fallen grating part
Underwater ROV [2]CR guide tube
(assumed)
A2 investigation: image processing
results based on acquired data [1]Fuel assembly component(upper tie plate)

Dosimeter/
thermometer

CRD index tube
(assumed)
Near CRD housing
External
lighting

Clump deposits

Camera

FY2011A2’ investigation: part ofA2’ investigation: tip
image data acquired underpart of investigation
platform [1]device [2]Lower part of pedestal
Underwater ROV investigation: part ofacquired image data

Results of investigation inside the Unit 2
pedestal

Results of investigation inside the Unit 3
pedestal

FY2012FY2013

FY2014FY2015FY2016FY2017FY2018Development of investigation technology of inside of PCVDevelopment of investigation
technology of inside of PCVDevelopment of investigation
technology of inside of PCVDevelopment of technologies for indepth investigation of PCV inside
Development of technologies for in-depth
investigation of PCV inside (on-site verification)

Fig. A11-1-(3): Development of technologies for in-depth investigation of PCV inside159

Project objectives
To contribute to the studies on the retrieval of fuel debris in RPVs, investigation technologies to identify the state of fuel debris in RPVs will be
developed.
Project details and progress
The following describes the results this project produced so far based on results from the related projects described in Section 2 below.
(1) Development of equipment to access the core(i) Developing equipment to access from the top (top-hole-drilling method)- To establish an access route for investigations from the top of the RPV to the core, the position of holes was studied and a work plan wasdeveloped. To avoid interference with structures, the removal method of RPV spare top head nozzle was verified through elemental tests.
- The specifications of the work cell in which a negative pressure environment is maintained in order to prevent the expansion ofcontamination were studied. In addition, the sealing and installation performance of the guide pipe at the connection part wth the PCV head
were checked through elemental tests.
Equipment carry-in entrance

Investigation method to
For investigation For drilling
access from the top
devicereactor internals
Work cell- Studies were performed on a processing method to
Air lock cell Work cellmake holes in the reactor internals remotely and inTool box
Tool box cart
narrow space and on its equipment specifications.
Access device
Well coverThen, prototypes of an abrasive water jet (AWJ) tool(stored in tool box)
head and access equipment were made, and their
RPV head drilling Guide pipe
PCV headremote operability was checked through elemental
device (Cutting
Boundary function maintenance device
and removal of(PCV head connection Guide pipe
tests.
RPV headspare nozzle)
part)(ii) Developing equipment to access from the side
Seal
(side-hole-drilling method)
Simulated PCV- The access route to the core from the side washeaddetermined, and the concept of the method was
PCVestablished, including the selection of tools
RPV
appropriate to the drilling and sealing work and major
Reactor internals
operation steps. In addition, overall specifications ofdrilling device
For full-scale For(AWJ tool head)
Work cellthe equipment were summarized, including theinvestigation preliminary
confirmation
design of maintenance systems for the equipment.
(2) Development and selection of investigation
Biological shield wall
method to the core
PCV shield wall
- Specifications of equipment for prior-checking ofDevice for investigation
Investigation method to access from the sideof reactor insideaccess routes and investigation of the core werestudied by each investigation step. The visibility, radiation resistance, and operability were checked through elemental tests.
(3) Designing the overall system of investigation equipment and drafting a construction method plan
- Based on the assessment of radiation exposure due to dust dispersion, it was decided to maintain the inside of PCV under slightly positive
pressure (the current status) during the investigation from the side, but under negative pressure during the investigation from the top to preventthe expansion of contamination.
- In addition, the equipment and functions required to ensure safety were summarized in construction method plans.
(4) Problems to be solved and future plans- The investigation method from the top has been verified as feasible, and the investigation from the side as probably feasible. Therefore,
focusing on both investigation methods, prior on-site investigations will be carried out in the early stages toward the designing of equipment
applicable to on-site investigations, and information on the conditions of connection parts and environmental information will be collected. In
addition, elemental tests will be performed if necessary, and the test results will be reflected in equipment design.
- Study will be conducted on the applicability of ancillary systems necessary for the investigation, and on the determination of the conditions
required for the designing of equipment for actual reactors, including coordination with other related construction work.
Implemented byInternational Research Institute for Nuclear Decommissioning (IRID)
2. Related projects
The following describes the results of previous related projects.
○ Development of investigation technology of inside of RPV (FY 2013 - 2015)- To collect information on temperatures, radiation doses in RPVs and the positions of fuel debris in RPVs, and the damage status of reactor
internals, the method of accessing the target of the investigation, the investigation method, and the method to collect samples were studied.
Investigation technologies under a high radiation environment (estimated as 1000Gy/h) in RPVs were also sorted out. Then, a technology
development plan for the investigation of the inside of RPVs was established.
- The needs of the investigation derived from the field and related projects, such as the fuel debris retrieval project, were sorted out and thefeasibility of investigations were assessed.
- To access the inside of RPVs in the early stages, study was conducted on access technologies through existing routes (such as pipes) andthrough new routes that are to be drilled, as well as the fundamental design of investigation technologies and elemental technologies.
- Elemental tests were conducted to verify the feasibility of RPV top hole drilling method which is one of major technologies for theinvestigation of the inside of RPVs, systems to maintain the boundary function, and technologies to collect fuel debris samples from the side of
PCVs.
FY2011FY2012FY2013
FY2014FY2015FY2016FY2017FY2018Development of investigation technology of inside of RPVDevelopment of investigation technology of inside of RPVDevelopment of investigation technology of inside of RPV

Fig. A11-1-(4): Development of investigation technology of inside of RPV

160

Project objectives
In the decommissioning of the Fukushima Daiichi NPS, two technologies have been developed as technologies
to detect the distribution of fuel debris in reactors which use cosmic rays, muons: a transmission method and ascattering method. Measurement and evaluation for the transmission method have been completed at Unit 1 of
the Fukushima Daiichi NPS, while as for the scattering method, performance tests of a large muon paths
detector with a sensitive area of a 7m x 7m have been completed. The applicability of both methods to nuclear
power reactor system under high dose level environment has been verified.
Project details and progress(1) Implementation of small-scale verification tests (resolution of approx. 1 m)
In Unit 1, measurement was performed using a detecting system (transmission method) at three points in total
over about 90 days from February to September 2015. Based on the fuel in the spent fuel pool, it was verified
that the system had a resolution of 1 m, approximately (Fig. 1). According to the measurement results, there is
probably not fuel around the core. By improvement of the muon detector, a small device has been developedthat is one-fourth the size of the original with no loss in analysis performance (Fig.2).
(2) Design and manufacturing of detection system (resolution of approx. 30 cm)
Items for improvement to solve the displacement of the detector were identified, and noise reduction
technology was added to the identification algorithm. Then, it was verified through simulations that theerroneous detection rate will not drop even if the measuring period is shortened to one-fourth of the original
period (Fig. 3).
Transmission tests were performed for reactor components using a large muon paths detector with a sensitive
area of 7m x 7m, and it was verified that it could produce images of lead (high-density materials) without being
obstructed by concrete.

Transmissivity

Reactor
core

Fig. 2: Small device for transmission method
Measuring period
1/4 (23days)

Measuring period
1/4 (23days)

Erroneous detection
rate = 44.97%

Erroneous detection
rate = 15.18%

Horizontal direction (m)Fig. 1: Result of measurement and evaluation usingtransmission method in Unit 1

Implemented byInternational Research Institute for Nuclear Decommissioning (IRID)
FY2011F2012

FY2013

FY2014FY2015

Fig. 3: Simulation result of improvement of erroneous
detection rate

FY2016FY2017FY2018Development of technologies for the detection offuel debris inside reactors (using muon)

Fig. A11-1-(5): Development of technologies for the detection of fuel debris inside reactors (using
muon)

161

Project objectives
Various fuel debris retrieval methods have been studied while focusing on the retrieval directions (top and side) and retrieval environment(submersion and partial submersion). In addition, concepts of systems to ensure safety that conform to each method have been studied. In
accordance with the results of these studies and the fuel retrieval policies, technical investigations, elemental tests, and analysis will be
performed for issues that have been identified in developing required elemental technologies. In addition, this information and knowledge will
be used in the consideration of construction methods/systems while focusing mainly on the in-air/side-entry access method.
1. Project details and progress
The following describes the results this project produced so far based on results from the related projects described in Section 2 below.
(1) Development of technologies related to containment function：本事業にて要素技術の開発を計画する範囲
Gaseoussystem液体系システム
気体系システム
Liquidsystem- To develop elemental technologies to ensure a containment：臨界管理ＰＪでの開発結果を反映する範囲※(全体イメージ図)(全体イメージ図)(Overallimage)※(Overallimage)
function by maintaining negative pressure, it has been decided
as development policy to carry out analysis in combination with
負圧管理による閉じ込め技術
elemental tests. Analysis and elemental tests will be performed
continuously to develop technologies to ensure a containment
溶解性核種
モニタリングfunction by maintaining negative pressure.
除去設備
設備(2) As regards the following technologies, existing technologies
will be investigated and comparisons will be made to selectoutstanding technologies and to enrich our knowledge throughelemental tests : technologies for the collection and removal ofdust generated from fuel debris, technologies to collect and
前処理
粒子除去
粒子除去
設備
設備
設備
モニタリング
粒子除去設備remove dust in gas systems (Fig. 1, right), and technologies to
設備(HEPAフィルタ)
collect and remove dust and soluble nuclides in liquid systems
Fig. 1: Scope of element technology development for system formulation(Fig. 1, reft).
(3) Study on Alpha-nuclide monitoring system required for the retrieval of fuel debris- We are organizing and identifying the necessity and objective of thenuclide monitoring technology (Fig. 1) in the retrieval of fuel debris andthe required measuring range in gas system monitoring. In addition, we
are investigating existing technologies for nuclide monitoring in gas
systems and sorting out challenges toward the application to fuel debris
retrieval.
(4) Study on optimization of ensuring safety of methods and systems- As regards construction methods whose applicability to each reactor
has been studied with a focus on partial submersion-side access,Fig. 2: Installation image of fuel debris retrieval cell
design conditions are being organized and methods to install cells arebeing studied (Fig. 2).
- Safety and functional conditions required for the retrieval of debris are being reorganized and brushed up to review the necessary systems(Fig. 1).
- The results of the public exposure evaluations at the time of fuel retrieval is under review, and worker exposure, which is important, is under
examination.
Implemented by International Research Institute for Nuclear Decommissioning (IRID)
Ｂ

ホウ酸水
ホウ酸水タンク
注入ポンプ

注水ポンプ

監視計器略号
Ｌ：水位計
Ｔ：温度計
Ｐ：圧力計
Ｆ：流量計
ｎ：中性子モニター
R：線量モニター【凡例】
青字:常用の対処設備
赤字:非常用の対処設備
緑字:監視設備

注水水源※但し、PCV補修技術は他PJにて開発

Ｂ

ホウ酸水
ホウ酸水タンク
補給ポンプ

排気口

補給水源

Ｆ Ｆ

R

Ｐ

粒子除去 前処理
フィルタ 設備

水素 ダスト 臨界濃度計 モニタ モニタ

Ｔ

非常用
バッファタンク
水処理設備へ
水処理設備へ

Ｔ
Ｌ
Ｌ Ｌ

粒子除去
設備

ｎ

ホウ酸
調整装置

Ｐ

排気ファン

水処理設備より

補給ポンプ

粒子除去
設備

粒子除去
設備

水処理設備へ

冷却器

トーラス室取水ポンプ D/W取水ポンプ S/C取水ポンプ（常用／非常用） （常用／非常用）（常用／非常用）

粒子除去 前処理
フィルタ 設備

Ｔ

窒素供給設備循環ポンプ

バッファタンク

バッファタンクへ

排気ファン

排気口

トーラス室より

排気ファン

冷却器

粒子除去 前処理
フィルタ 設備

PCV負圧によるインリーク

窒素供給設備水素 ダスト 臨界濃度計 モニタ モニタ※本図は、機能要求を満足するための実施形態の一例であり、設備の設置位置（R/Bの内外）や機器の型式を特定するものではない。

2. Related projects
The following describes the results of previous related projects.
○ Development of technologies for retrieval of fuel debris and internal structures(FY 2014)(1) To determine the fuel debris retrieval method, plant data and information about
the achievements of other development projects were collected and sorted out.
Then, three major retrieval methods that are considered to be highly feasible(submersion-top access, partial submersion-top access, and partial submersionside access) were selected, and issues were identified and summarized.
(2) As regards the identified issues, existing technologies were investigated,countermeasures were considered, and plans for necessary development were
established.
(3) Elemental tests were performed for technologies that are considered to berequired for the retrieval of fuel debris.
○ Upgrading of approach and system for retrieval of fuel debris and internal
structures (FY 2015-2016)(1) Study on the feasibility of the three major methods
Process flows and operation step charts for the three major methods were
examined and developed. Required specifications have been determined andissues have been identified.
(2) Study on the concepts for systems to ensure safety
Safety and functional conditions required for the retrieval ofdebris were recapped, and configurations of systems toensure safety were studied.
Fig. 3: Concept of partial submersion - top
access method (example)
In addition, the exposure dose was evaluated roughly with
Fig. 2: Concept of fuel debris retrieval equipment
the systems that have been studied.
in partial submersion - side access method(3) Study on the concept of fuel debris retrieval equipment(example)
The concept of fuel debris retrieval equipment that will be applied to the three major methods (such as the one shown in Fig. 4) was studied.
FY2011FY2012FY2013
FY2014FY2015FY2016FY2017FY2018Development of technologies for retrieval of fuel debris and internal
structures
Upgrading of approach and system for retrieval of fuel debris andinternal structures

Fig. A11-2-(1): Advancement of retrieval method and system of fuel debris and internal structures

162

Project objective
Toward the decommissioning of the Fukushima Daiichi NPS, in view of the results of studies that have been performed so far regarding theretrieval method of fuel debris and internal structures, elemental technologies required for retrieval (technologies to prevent fuel debris fromscattering, elemental technologies for installing retrieval equipment, technologies to perform remote maintenance of retrieval equipment, andmonitoring technologies during retrieval) will be developed, and the feasibility of devices and equipment will be assessed.
1. Project details and progress
The following describes the results of this project produced including theresults from the related projects described in Section 2 below.
(1) Development of technologies to prevent scattering of fuel debris
Fuel debris cutting/dust collecting systems (Figs. 1 and 7) andrecovery/transfer systems are under development, and study is underway
Fig. 1: State of chisel
processing preliminary teston the concept of the scattering prevention method, a technology to retain
Fig. 2: Inflatable seal element test
fuel debris in the operation area.
(2) Development of elemental technologies for the installation of fueldebris retrieval equipment
Elemental technologies for the installation of operation cells where
retrieval equipment will be stored (sealing mechanism (Fig. 2) and remote
Fig. 4: Element test of biologicalwelding) and technologies to remove materials on access routes that will
Fig. 3: Element test of removal
shield wall removal
interfere with the retrieval work (Figs. 3, 4, and 5) are under development.
of interfering objects in pedestal(3) Development of technologies for remote maintenance of fuel debris
retrieval equipment
Maintenance of retrieval equipment installed in a high-radiation area
Fig. 5: Combined element test image
should be performed remotely in principle, so concepts about how toof Robotic arm and access rail
remotely maintain devices and equipment handling fuel debris have beenlaid out. In accordance with these concepts, studies on maintenancemethods, feasibility assessment, and identification of issues are
Fig. 7: Prototype of laser head
(ONET Technologies CN)
underway.
(4) Development of monitoring technologies during the retrieval offuel debris
Fig. 6: Prototype of imaging tube
To monitor the retrieval of fuel debris, an image tube camera (Fig. 6) and a small
and camera system
neutron detector (Fig. 8) that can be used under a high radiation environment areunder development.
Implemented by
Fig. 8: Prototype of neutron(Comprehensive proposal) International Research Institute for Nuclear Decommissioning (IRID)
detector (RosRAO, FSUE)(Partial proposal) Development of fuel debris cutting/dust collecting system: ONET Technologies CN, Taisei Corporation (FY 2015 - 2016)
Development of monitoring technology during the retrieval of fuel debris: Hamamatsu Photonics K.K.
Development of small neutron detector: International Research Institute for Nuclear Decommissioning (IRID), RosRAO,
FSUE, ONET Technologies CN (COMEX NUCLEAR)
2. Related projects
The following describes the results of previous related projects.
○ Development of fundamental technologies for retrieval of fuel debris and internal structures (FY2015 - 2016)
Fig. 9: Seal inside RPV and Fig. 10: Test of access
Focusing on the three access methods (the submersion method, the partial submersion-top access equipment, lower seal test equipment to RPV
method, and the partial submersion-side access method), plans for elemental tests to identify theirfeasibility have been established and the following elemental tests were implemented.
(1) Technology to prevent expansion of contamination during the retrieval of large structures
Fig. 11: Test of access- Scale model tests by operation step
equipment to pedestal(2) Technology to prevent expansion of contamination during the retrieval of fuel debris from RPVs
- Testing of access equipment to RPVs using the partial submersion-top access method (Fig. 9)(3) Access technology to fuel debris- Testing of water hydraulic manipulator
Fig. 12: Example of flexibly Fig. 13: Performance test
- Testing of access equipment to the inside of RPVs using the submersion method (Fig. 10)
structured arm for remote of cutting fuel debris
operation- Testing of access equipment to the inside of pedestal using partial submersion-side
access method (Fig. 11)(4) Remotely operated technologies required for the retrieval of fuel debris- Testing of flexibly structured arm (Fig. 12)- Testing of container handling equipment(5) Technology to prevent expansion of contamination during the retrieval of fuel debris
Fig. 16:
Prototype of- Testing of platform/cell for the submersion method
Fig. 14: Boring robot Fig. 15:
tube- Welding test for the remote sealing of cells for the partial submersion-side access method (Taisei Corporation) Underwater cutting imaging(Hamamatsu
head (COMEX
Photonics
K.K.)(6) Technology to reduce the exposure dose of workers during fuel debris retrieval work
NUCLEAR)- Testing of shape-following lightweight shields to be used in the top-access method(7) Cutting/dust collecting technology required for the retrieval of fuel debris- Testing of fuel debris cutting/dust collecting performances (Figs. 13, 14, and 15)(8) Monitoring and measurement technologies during the retrieval of fuel debris-development high radiation tolerant image camera tube (Fig. 16)
FY2011FY2012FY2013
FY2014FY2015FY2016FY2017FY2018Development of fundamental technologies for retrieval of fuel debris and internal structures
Advancement of fundamental technologies for retrieval of fuel debris and internal structures

Fig. A11-2-(2): Advancement of fundamental technologies for retrieval of fuel debris and internal
structures

163

Project objectives
Toward the decommissioning of the Fukushima Daiichi NPS, in view of the results of studies that have been performed so far regarding theretrieval method of fuel debris and internal structures, elemental technologies required for retrieval (prevention scattering of fuel debris,
installing retrieval equipment, remote maintenance of retrieval equipment, and monitoring during retrieval) will be developed, and the feasibilityof devices and equipment will be assessed.
Project details and progress
To contribute to criticality control during fuel debris retrieval and also to the design of fuel debris retrieval equipment and the rationalization offuel debris retrieval work procedures, scenarios for fuel debris sampling are being prepared and sampling equipment including access devices
are being developed. At the same time, component testing is being performed for related equipment such as arm-type access devices andattachments that would enable the sampling of pebble shape debris and sandy debris while taking advantage of existing safety systems. The
activities conducted so far and the results obtained from them are described below.
(1) Study and development of fuel debris sampling scenarios
An overall scenario for fuel debris sampling has been developedin consideration of needs and the results of surveys on
conditions inside PCVs, and development has begun on plansconcerning the development of required techniques. In addition,
the impacts of sampling activities on safety are also being
evaluated.
(2) Design and trial production of systems and equipment for the
sampling of fuel debris inside PCVs
Sampling concept for debris in PCV① Basic design of the fuel debris sampling systems
The requirements to be fulfilled are being determined and summarized by sampling systems fordebris of different shapes (pebble shape debris, sandy debris, larger cylindrically cut-out debris,etc.) from the viewpoints of minimizing workers' exposure dose and ensuring safety through theprevention of criticality as information to be referenced later on in the basic design phase.
② Design and trial production of devices accessing to fuel debris
Study is being performed of the feasibility of reusing thearm-type access devices designed for the detailed
Cuttingmechanism
survey of conditions inside PCVs and the specifications
切削機構
of improved access devices.
③ Design and trial production of fuel debris sample
Post-cut powder
collection
mechanism
collection devices
切削粉回収機構
Component testing has been performed on attachments,
Grating
グレーチングfor the collection of pebble shape and sandy fuel debris
Sealingmechanism around grating
グレーチング周囲の密封機構
samples, and their conceptual design is complete. The
next step is trial production.
Example of cutting and collection tool ofdebrisExample of access device for sampling

Cutting
切削試験
test

Sealing
密封試験
test
Cutting
切削刃
blade

Sealing
密封機構
mechanism
Element test of cutting and collection
tool of debris(3) Conceptual study of sampling systems for fuel debris inside RPVs
Based on the status of the development of the internal investigation of the RPV, work is being performed with the concept of a system thatwill enable access to debris inside an RPV either from the top or from the side, and attempts are being made to establish plans concerningcomponent testing, etc.
Implemented byInternational Research Institute for Nuclear Decommissioning (IRID)
FY2011F2012

FY2013

FY2014FY2015FY2016FY2017FY2018Development of sampling technologies for
retrieving fuel debris and internal structures

Fig. A11-2-(3): Development of sampling technologies for retrieving fuel debris and internal
structures

164Project objectives
In order to ensure the safety of fuel debris retrieval work, develop technologies to access and connect to PCV while securing the PCV
confinement function which is issue in constructing the necessary closed circulation system for water inside PCV.
Project details and progress(1) Formulating technical specifications of repair and Drafting work and development plan for access and connection to the inside of the PCV①To ensure safety in fuel debris retrieval work, intaking water from the dry well (hereinafter referred to as "D/W"), a suppression chamber(hereinafter referred to as "S/C"), and the torus room is being investigated for a water circulation system being studied or developed under the
project of“Advancement of Retrieval Method and System of Fuel Debris and Internal Structures". To intake water from the D/W and the S/C, it
is necessary to establish a new access route to the inside of the PCV and water intaking system, while securing the confinement function. In
order to realize this, it is necessary to establish a construction technology and a work plan taking into account severe on-site environmental
conditions, such as high doses and narrow paths, reliability, inspectability, earthquake resistance, long-term soundness, remote
maintainability, and other factors. The required technical specifications and a system establishment work procedure should be investigated anddevelopment challenges should be identified and a development plan should be drafted. Required technology development challenges for thefollowing items should be identified and the development plan should be drafted.
:
i. Formulating technical specifications considering the on-site environmentii. Investigating plans for access route construction work and its maintenanceiii. Identifying development challenges and drafting a development plan
② Regarding PCV repair technology affecting the construction of closed circulation systems, we also extract technical development
challenges corresponding to the situation at the on-site work and extract technical issues corresponding to situations in the field, and draft
development plan.
(2) Developing and verifying element technologies for access and connection to PCVBased on the development plan formulated under (1) above, we will develop and verify each elemental technology required to access andconnect to the PCV. Below are examples of possible elemental technologies :- Remote construction technology for connection- Remote operation technology for access route inspection during service and construction- Remote repair technology for connections during service and construction(3) Verifying access and connection technology on a full-scale
Based on the development results of each elemental technology, we will conduct trial design on access and connection to the inside of the
PCV and carry out to verify the design on full-scale. In particular, we will ready full-scale test equipment, for example of Naraha Center for
Remote Control Technology Development, to verify the workability on a full-scale, clarify the work requirements for actual construction, andidentify technical challenges. :- Confirming the workability by remote control on full-scale, and identifying challenges- Securing confinement function for actual construction work and the measures to reduce workers' exposure and identifying challenges
Investigation of the test body after connection work

Implemented byInternational Research Institute for Nuclear Decommissioning (IRID)
FY2011F2012

FY2013

FY2014FY2015FY2016FY2017FY2018Development of closed circulation systems for water inside PCV

Fig. A11-2-(4): Development of closed circulation systems for water inside PCV

165

Project objectives
Repair methods for the bottom parts of PCVs have been developed aiming for the establishment of the containment capability of PCVs in orderto maintain the prevention of the scattering of radioactive materials, the radiation shielding, the cooling function, and the containment of thecontaminated water bearing alpha-emitting nuclides during fuel debris retrieval. In addition, the process of filling the PCV with water has beenstudied, as well as the concept of environmental improvement methods to enable the practical repair works.
1. Project details and progress
The following describes the results of this project produced so far based on results from therelated projects described in Section 2 below.
(1) Study and planning of the process up to the completion of PCV water filling
For each PCV water level that is considered achievable, the target sealing capability to beachieved by leak repair techniques was defined.
(2) Development of repair methods for the bottom parts of PCVs
① Reinforcement method for suppression chamber (S/C) supports
Component testing was performed under conditions resembling actual plant conditions toverify the effectiveness of reinforcement based on the evaluation of the strength and fluidity.
② Sealing method of pouring grout material into S/C
The feasibility of implementation was evaluated by the series of the tests, namely the grout
Vent pipe 1/1 scale test
test with the PCV model with reinforcement rings, the long-distance pumped delivery tests,
the component testing under conditions resembling the actual plant conditions, as well as the functional tests of S/C guide pipe
implementation.
③ Sealing method of pouring grout material into the vent pipes
By performing an experiment using a full-scale (1/1) test set using self-compacting concrete as sealant, the implementation feasibility andsealing performance were verified. A heavy slurry based repair material was developed as well.
④ Sealing method by the plugging of the vacuum breaking line
By performing an experiment on a full-scale (1/1) test set, the implementation
Guide pipe
feasibility and sealing performance of a seal plug that had been improved for easier
Water stop
implementation were verified.
Vacuumplug insertion⑤ Repair method for the establishment of boundary for the pipes connected to thebreakage valve
device
PCVs
Bellows
Sealant was developed, as well as the core technologies for remote implementation
Water stop
equipment, and the feasibility of implementing the proposed methods were verified.
plug
Water stop(3) Development of repair methods for the upper parts of PCVs
plug
After studying how it might be possible to improve the leak repair device forDrawing
of
water
stop
plug
for
vacuumbreakageline
application to the equipment hatch sealing, it was concluded that it would probably
be possible to increase the ease of implementation.
(4) Sealing method for piping penetrations through the torus chamber walls, etc.
A type of sealant that would allow application by spraying was identified, and test were performed to verify its sealing performance.
(5) Studying the concept of environmental improvement to be achieved to enable the practical application of repair methods
The dose that workers will be exposed to while engaged in PCV bottom repair (sealing) works was assessed, the implementation techniquesthat may be employed to reduce the exposure dose were reviewed, and challenges were identified.
Implemented byInternational Research Institute for Nuclear Decommissioning (IRID) (FY 2013 -)2. Related projects
The following describes the results of previous related projects.
○ Development of identification technology of leaks in PCVs (FY 2011 - 2013)
Devices to be used for locating leaks in the PCVs and torus chamber walls were developed by performing examinations in a high dose,
congested, water-filled environment, and their performances were tested using mockups, etc. Some of the devices were put into practicalapplication (to examine the torus chamber rooms in Units 1 and 2, for example), which contributed to the discovery of a leak from the vacuum
breaking line bellows in Unit 1.
○ Development of repair method for PCVs (FY 2011 - 2013)
Detailed examination of repair techniques was performed as a step toward the design and fabrication of repair devices that may be used forthe establishment of boundary in the vent pipes, S/Cs, etc.
Basic testing was performed for sealing techniques that would contribute to the design and fabrication of devices that may be used in the
repair of components at the upper part of PCV, such as hatch flanges and penetration bellows that are likely to have been damaged,○ Development of repair (water stoppage) technology toward water filling in PCV (FY 2014 - 2015)(1) Development of PCV repair/sealing methods- Reinforcement method for S/C supports: The grout material for the reinforcement was improved, and the possibility of practical application
was determined through various tests.
- Sealing by grouting inside S/C: The composition of the sealant was determined, and its sealing performance was studied through various
tests.
- Sealing by grouting the vent pipes: Deployment testing was performed for the plugging assisting material and the screening of secondary
plugging assisting material, and challenges to practical application were identified.
- Sealing by the plugging of the vacuum breaking line: Plug insertion using a flexible guide pipe was tested, and the feasibility of a series oftechnical procedures was verified.
- Methods for the establishment of boundary for the pips connected to PCVs: The performance required from the sealant was defined, asealing test was performed, and issues were identified.
- Dry well repair technique: The sealing technique and the test to be performed to verify the applicability of sealant were studied, and theschematic design of the sealing device was completed.
(2) Determining the plan up to the completion of PCV water filling
The work steps up to the completion of PCV water filling at Units 1 and 2 were defined.
FY2011FY2012FY2013
FY2014FY2015FY2016FY2017Development of identification technology of leaks in PCVs
Development of repair method for PCVs

Development of repair (water stoppage)technology toward water filling in PCVDevelopment of repair methods for leak spots
in PCV

Fig. A11-2-(5)-1: Development of repair methods for leak spots in PCV

166

Project objectives
The repair methods of the bottom part of PCV from the R&D project entitled " Development of repair methods for leak spots in PCV " were
tested using a full-scale test set to verify the acceptability of procedures from the viewpoints of actual implementation, the feasibility of
implementation by remote control, and the sealing performance. In addition, the effectiveness of simulation using a virtual reality (VR) system
for simulator-based operator training for the remote control manipulator was verified.
1. Project details and progress
Work floor (simulated
The following describes the results of this project produced so far based
first floor of R/B)
on results from the related projects described in Section 2 below.
(1) Full-scale test of the repair methods of the bottom part of PCV, etc.
The following were tested using the full-scale test set:①
Reinforcement method for the suppression chamber (S/C) supports
Injection
Experiments were conducted with implementing reinforcement by the Simulated interfering object
equipmen
Injection
injection of highly fluid grout material to the S/C bottom, and the(2 stage of grating)t
hose
applicability of an implementation procedure developed in consideration
Simulated S/Cof actual plant conditions was verified, as was the feasibility of
implementation monitoring, such as the checking of injection height.
Simulated torus②
Sealing method by grouting the vent pipes
room
A test was conducted to check the feasibility of an implementation
Overview of test facility: full-scale 1/8 sector (simulated Unit 2)
procedure that would involve remote controlled maneuvers to remove(Test location: Naraha Center for Remote Control Technology Development)
obstacles and drill a hole through the vent pipe model, and the feasibility
Fig. 1: Overview of the reinforcement test for S/C supportsof accessing targets under conditions resembling the actual plant conditions was verified.
③
Sealing by grouting inside S/C (downcomer plugging)Experiments were conducted with stopping the leak by grouting highly fluid sealant into S/C, and the applicability of an implementation
procedure developed in consideration of actual plant conditions was verified, as was the feasibility of implementation monitoring, such as thechecking of injection height.
④ Preparation for testing, etc.
Maintenance services were administered for water feeding and draining facilities, and regular inspections were performed for this purpose.
(2) Preparation of VR data for preliminary simulation experiments
As efforts were being made to prepare an environment that would allow the performing of operator training on a VR system simulating the
motions of the manipulator used for plugging the vent pipes, the effectiveness of VR system in operator training was evaluated by capturing
manipulator motions using motion capture technology, etc., and comparing the motions simulated by VR system loaded with the motion data
and the actual motions made by the manipulator.
Implemented byInternational Research Institute for Nuclear Decommissioning (IRID)
2. Related projects
The following describes the results of previous related projects.
○Full-scale test of repair methods for leak spots in PCV (FY 2014 - 2015)(1) Testing of PCV bottom repair techniques using a full-scale test set, etc.
The designing and fabrication of a full-scale model reproducing a 1/8 sector portion of PCV bottom at Unit 2 was completed, as was theinstallation of water feeding and draining facilities required for testing, including the turbid water treatment system.
(2) Preparation of VR data for preliminary simulation experiments
Data on the motions of the remote operated device (manipulator) for loading into the VR system was prepared, and it was verified that the data
can be used for simulating motions on the VR system. In addition, research was completed on the knowledge of various institutions inside andoutside Japan regarding the functions of remotely operated devices and the systems used to verify operators' skills.
○Full-scale test of repair methods for leak spots in PCV (FY 2016)(1) Full-scale test of the repair methods of the bottom part of PCV etc.
① Reinforcement method for the suppression chamber (S/C) supports
An implementation procedure verification test was performed, and it was verified that works in a high dose environment, operation by remotecontrol, and the application of devices prepared in the course of PCV repair method
development may be accomplished without problems.
Suppression
Vent pipe② Sealing method by grouting the vent pipes
chamber
The implementation procedure verification test was initiated.
④
Preparation for testing, etc.
Pre-testing preparations were completed by transporting the test set, installing buffermaterials for S/C supports reinforcement, and ensuring readiness for filling the test set
DownTorus roomcomer
with water.
wall surface(2) Preparation of VR data for preliminary simulation experiments
Quencher
Based on information collected through hearing sessions with remote control systemedesigners and experienced operators, the issues that must be solved in order to achievenecessary improvement in accuracy were identified, and work was conducted toward theFig. 2: Full-scale test facility
realistic simulation of manipulator motions on the operation console and the improvementof remote control system functions.
FY2011FY2012FY2013
FY2014FY2015FY2016FY2017Full-scale test of repair methods for leak spots in PCV
Full-scale test of repair methods for leak spots in PCV

Fig. A11-2-(5)-2: Full-scale test of repair methods for leak spots in PCV

167

Project objectives
The seismic resistances of RPV, PCV, and other major plant components shall be evaluated through assessment in consideration of effects ofthe falling of highly heated fuel debris at the time of the accident, the aging degradation accompanying the corrosion of steel, and the waterleak repairs performed, and the facilities added prior to fuel debris retrieval. In addition, the effects in the event of damage shall be predicted,
and the measures that may be taken to prevent or mitigate the consequences shall be determined. Furthermore, the effectiveness shall also
be verified for the safety scenarios developed on the basis of the proposed prevention/mitigation measures.

Internal

External

1. Project details and progress
The following describes the results this project produced so far based on results from the related projects described in Section 2 below.
(1) Producing the scenarios of safety in the event of a severe earthquake
The facility improvement measures to be implemented were determined, as
Bonds
were the emergency action plans to be made ready before the beginning of fueldebris retrieval to address the consequences of damage to large componentsthat may be induced by a severe earthquake, and safety scenarios weredeveloped (each indicating the flow of a series of actions to be taken to
Burialheightmaintain safety functions or to restrict effects from the accident).
(2) Development of approaches to evaluating seismic resistance anddetermining impacts from damage in the process of producing safety scenarios
Fig. 1: Vent pipe-S/C system coupled analysis model (Unit 1)① Development of an approach to evaluating the seismic resistance of
CRD opening
suppression chamber (S/C) supports and determining impacts from damage
CRD opening
By performing the elastic time-history response analyse of vent pipes and S/C
CRD
Lower
system coupled model (Fig. 1), the seismic resistance of the structure after the
opening
AccessCRD
injection of sealant into S/C was assessed.
opening(2) Deficiency
opening- In addition, elastoplastic analyses (by the double gradient method) were also
range, inside(1)
Internal only
Central/outside (freedeficiency (element
performed for critical components, such as the column supports, to determine theopeningboundary)
reduction)
maximum allowable quantity of sealant that may be injected into S/C.
Model of debris
FEM analysis model
Cross-sectional
erosionpartsof
1F-Unit1drawing② Development of an approach to evaluating the seismic resistance ofFig. 2: 3-D FEM elastic-plastic analysis model of RPV pedestal
pedestals and determining impacts from damage
An assessment approach was developed and material data was collected as described below to enable prediction of the distribution oftemperature in the pedestal when it was exposed to high temperature and the impact of erosion by fuel debris:- Performing elastoplastic analyses by three-dimensional finite element method (FEM) (Fig. 2) and evaluating strength and rigidity using a fibermodel- Coupled response analysis method for evaluating how changes in pedestal strength or rigidity may impact the seismic resistance of large
components such as PCV and RPV- Degree of heat-induced corrosion and strength reduction of reinforcement bars that have a history of being exposed to high temperatures(3) Supporting the advancement of safety scenarios
In order to support the advancement of assessment approaches mentioned above, the approaches to verification by analyses, tests, etc., werereviewed, and detailed analyses and material tests were performed in connection with the following:- Assessment of S/C supports at Unit 1 by performing elastoplastic time-history response analyses- Collection of data from PCV material tests performed to determine the effects of the history of being exposed to high temperature at the timeof the accident
Implemented by International Research Institute for Nuclear Decommissioning (IRID) (FY 2013 -)2. Related projects
The following describes the results of previous related projects.
○Development of evaluation methods for the structural integrity of RPV and PCV (FY 2011 - 2013)(1) Evaluating the seismic resistance of major plant component such as RPV, PCV, and pedestal
Evaluation was conducted on the seismic margins of major plant components at present and at the time of fuel debris retrieval (considering theimpacts of thinning by corrosion, the additional load from the weight of fuel debris retrieval devices, etc.). It was concluded that major parts ofthe components may retain seismic resistances of a satisfactory level, but some components required more detailed analysis of seismic
resistance.
(2) Evaluation of the effects that hot debris may have had as they fell inside the pedestal
A literature search, etc., on the molten core-concrete interactions (MCCI) was used to prepare basic data that would contribute to theprediction of erosion suffered by the pedestal.
○ Development of evaluation methods for the structural integrity of RPV and PCV (FY 2014 - 2015)(1) Evaluating the feasibility of fuel debris retrieval by the subversion method considering the aseismic integrity of PCV/RPV
Based on the results of evaluating the seismic resistance of PCV, RPV and other major plant components under plant conditions at the time offuel debris retrieval, the feasibilities of fuel debris retrievals by the partial submersion method or the full submersion top access method werestudied, and detailed evaluation was performed for the seismic resistance of components and parts that demonstrated small seismic margin in
the earlier evaluation.
(2) Simplified evaluation of seismic resistance of components in consideration of PCV repairs and water level increase
The parameters (such as the D/W water level) that would impact seismic response analysis were identified or selected, and then seismic
response analysis was performed using different assumptions on the parameter values. Based on the results obtained, it was determined howchanges in parameter values would impact seismic response, and a simple method of seismic resistance evaluation was developed. Then, thevalidity of the simple evaluation method was demonstrated by comparing results from the simple evaluation method with results from themethod that is normally used, the dynamic analysis method.
(3) Estimating the impacts of erosion suffered by the pedestal
The knowledge required for estimating or discussing pedestal strength in real units was obtained by collecting concrete strength data from thetesting of cylindrical specimens, block specimens, and scaled pedestal specimens, etc., that were heated to high temperature and then left in
air or in water as well as reinforcement bar corrosion data.
FY2011FY2012FY2013
FY2014FY2015FY2016FY2017Development of evaluation methods for the structural integrity of RPV and PCVDevelopment of evaluation methods for the structural integrity of RPV and PCVDevelopment of evaluation methods of seismic performances of RPV and PCV and the impacts of the damages

Fig. A11-2-(6): Development of evaluation methods of seismic performances of RPV and PCV
and the impacts of the damages

168

Project objectives
To verify the corrosion protection measure and to evaluate its feasibility for practical application so as to maintain the current structural integrityof RPV and PCV through the period of fuel debris retrieval by protecting the progress of corrosion in the structural materials.
Crevice corrosion test piece (ER,CREV,
Flag type test piece
1. Project details and progressconstant potential crevice corrosion(ESP measurement)
test)
The following results were produced from ongoing projects described in
Liquiddescribed in Section 2 below.
level(1) Evaluation of performance of inhibitor for corrosion prevention and its side
effects (Figs. 1 and 2)① Evaluation of the local corrosion prevention performance by electrochemicalmeasurement
To evaluate the effectiveness in preventing the local corrosion of carbon steels
Fig. 1: Electrode composition of corrosion test pieceby the selected inhibitors through a screening process in the previous year, atest procedure was established for electrochemical measurements(repassivation potential for crevice corrosion, open circuit potential, crevice
corrosion test at constant electrode potential), and measuring of them was
performed under gamma ray irradiation.
② Evaluation of the impacts of phosphate-based inhibitor firmly stuck onto hightemperature surface areas
Batch type test and test with loop were conducted to evaluate the impacts ofa phosphate-based inhibitor firmly stuck onto high temperature surface areas.
Another experiment was also conducted to find out the impact on theFig. 2: Test condition under irradiation
effectiveness on the corrosion protection of carbon steel when phosphatebased inhibitor is used combined with sterilizing agent.
③ Evaluating impacts on water treatment systems
The impacts of corrosion inhibitor on water treatment
systems were experimentally evaluated. Undervarious combination of seawater dilution ratio andcorrosion inhibitor concentration, experiments wereconducted to examine the impacts of corrosion
inhibitor on water treatment systems.
(2) Conceptual design of the corrosion prevention
system (Fig. 3.)
The conceptual design of the corrosion prevention
system was developed to inject the corrosion inhibitor
Fig. 4: Example of conceptual design of corrosion control system
into the actual water circulation system.

Above actions were implemented by International Research Institute for Nuclear Decommissioning (IRID)
2. Related projects
The followings are the results of related previous projects.
Note: Activities from FY2011 to FY2015 were pursued as part of the project entitled "Development of RPV/PCV integrity assessment
techniques".
○Development of RPV/PCV integrity assessment techniques (FY 2011 - 2013)
PCV material (carbon steel)
50℃, Artificial seawater diluted 1,000 x (19ppmCl -)(1) establishment of the model for predicting long term corrosion rate
Test time: 500h
To develop corrosion rate prediction model, corrosion tests were performed for up
to 2,000 hours under various the chloride ion concentration and water temperature.
(2) Development of an corrosion protection measure
Corrosion test was performed using sodium nitrite or sodium tungstate as acorrosion inhibitor, and it was confirmed that corrosion may be protected by injecting
such corrosion inhibitors of equal or greater molar concentration to that of chloride
ions.
○Development of RPV/PCV integrity assessment techniques (FY 2014 - 2015)(1) improvement of the model for predicting long term corrosion rate
No corrosion
General corrosion
By performing additional corrosion testing with loop for the extended period (10,000
No rust-preventive agents Zinc/sodium molybdic acidhours), the model for predicting long term corrosion rate was improved. In addition,
mixing phosphate (2,000ppm)
study was conducted on influence of substances eluted from fuel debris or concreteFig. 4: Development status of anti-corrosion measureson corrosion protection effect of inhibitor.
(2) Establishment of the corrosion protection measure
Selected candidates of corrosion inhibitor in this project (sodium tungstate, sodium pentaborate, etc.) was experimentally confirmed theircorrosion protection effect for the carbon steel under anticipated condition including presence of radiation. Based on the test results,
applicable inhibitors to the actual units was narrowed down. In addition, experiments were performed to identify the impact of each corrosion
inhibitor on water treatment systems, and related technical issues were recognized.
FY2011FY2012FY2013

FY2014FY2015FY2016FY2017Development of RPV/PCV integrity assessment techniques
Development of RPV/PCV integrity assessment techniques
Development of corrosion protection technology for RPV and PCV

Fig. A11-2-(7): Development of corrosion protection technology for RPV and PCV

169

Project objectives
To protect on-site workers from radiation exposure and prevent impacts on the surrounding environment, even if re-criticality would occur whenfuel debris configurations and water levels may change during fuel debris retrieval work, it is necessary that an impact evaluation method foruse with criticality, criticality monitoring technology, including technology for criticality approach detection and re-criticality detection, andcriticality prevention technology.

keffective (simulated)

Project details and progress(1) Studying criticality control method- Based on the estimated information of each unit on the fueldebris distributions, the points to note in each of the steps up tocompleting the retrieval of fuel debris were clarified from theperspective of criticality control. These were updated appropriatelyfrom the information of evaluations and investigations inside thereactor, for example.
(b) Example of test reactor core
- A behavior evaluation method during criticality was developed,
and an evaluation method that will be utilized in studying the(a) Appearance of KUCA test
procedures to use to mitigate any impact in the case criticality
occurs and safety measures to use at facilities in the future.
- Basic policies about criticality control in different methods that
may be used for fuel debris retrieval were summarized, and the(d) Image of neutron
keffective (reference)
requirements for the fuel debris retrieval systems were compiled. (c)B-10 neutron detector
signals(e) keffective(2) Development of criticality control techniquesFig. 1: Criticality approach test at Kyoto University Criticality Assembly (KUCA)
laboratorydevice① Development of Criticality Approach Monitoring Method
Absorption material
水中の岩盤状- Using the reactor noise method that is based on neutron measurements,that spreads over
模擬デブリにdevelopment of subcriticality measurement technology has been
the surface of rock
広がる吸収材
shape simulated
promoted.
Debris with
By the measurement operational verification test in a high radiation
unevenness in theenvironment were performed, and by using a reactor core, which simulates
water
the various fuel debris conditions, was configured at KUCA laboratory, and
水中の粒状模擬the degree of subcriticality were evaluated.The feasibility of the criticality
Absorption material
デブリに流下す
approach monitoring technology was thus confirmed (Fig. 1).
flowing down toる吸収材
granular simulated② Development of Re-criticality Detection Technology
debris in water- The gas control system used to monitor slight amounts of FP gas
concentrations present in the PCV was improved for use in detecting any
recriticality in the early stages, and it was verified through a Kr-88
detection performance test at Unit 1 that the system is capable ofTable 1: Insoluble Neutron
Fig. 2: Example of Workability
Absorption Material
Verification Test of Water Glass
selectively detecting Kr-88 without interference from the presence of other(candidate material)
Type Absorption Materialradioactive nuclides.
③ Development of Criticality Prevention Technology- Nuclear characteristic verification tests at the KUCA, dissolution tests with long-term exposure, and performance workability tests took placeusing candidate materials selected by their fundamental physical properties and radiation resistance performance tests, and a prospectabsorbent material candidate obtained (Table 1).
- As for the soluble neutron absorbent material (sodium pentaborate), a nuclear characteristic verification test was performed at the KUCA, andthe basic specifications of the boron concentration maintenance facility were examined. Herewith, the prospect of boron concentration
evaluations and the technical feasibility of the necessary facilities, including the concentration maintenance facility, were obtained.
* Using results from past development activities, in the future, the results will be utilized in studying the optimal practical implementation of afuel debris retrieval device and relevant systems, and thus ensuring greater safety.
Implemented by
Hitachi-GE Nuclear Energy, Ltd., Toshiba Corporation, Mitsubishi Heavy Industries, Ltd., Japan Atomic Energy Agency (FY 2012)
International Research Institute for Nuclear Decommissioning (IRID) (FY 2013 -)FY2011F2012

FY2013

FY2014FY2015FY2016FY2017Development of criticality control technologies of fuel debris

Fig. A11-2-(8): Development of criticality control technologies of fuel debris

170

FY2018

Project objectives
For the planning of fuel debris retrieval as a step in the decommissioning of the Fukushima Daiichi NPS, in order to the repair and stop waterleakage in the suppression chamber (S/C) located in the PCV lower, it is necessary to obtain information about the presence of radioactive
materials (removal of radioactive materials required to repair works). Therefore, the condition of radioactive materials that may precipitate
inside the S/C, was predicted and a measurement method was developed.
Project details and progress(1) Formulation of development plan
Developments and work items required for the detection of radioactive materials believed to be indispensable were present extracted and adevelopment plan was formulated.
(2) Radioactive material migration scenarios
Consideration was given to scenarios involving the migration of radioactive materials into the S/C and torus room. It is thought that thelikelihood of an inflow of radioactive materials that exceeded the acceptable limit was low. It can be confirmed that there are no radioactive
materials that exceed the acceptable limit by measurements of the lower part of the S/C bottom and the region around the sand cushion drainpipe outlet where there is a relatively high probability of radioactive materials having been precipitated.
(3) Evaluating of the impacts of radioactive substances on water stoppage material
Among the influence factors arising from remaining radioactive materials, there is a concern that even the minimum sediment may heat
cement to a temperatures of 80°C. Cement deterioration may be caused by heat generated in water stoppage materials. Based onconservative estimates, the presence of uranium of 13 kg or more was calculated. Therefore, it was concluded that the presence of uranium of
10 kg or more should be detectable by nondestructive measurement.
(4) Development of a technology for detecting of radioactive materials
Nuclides originating from fuel (Eu-154, Cm-244, etc.), background nuclides, and nuclides in shielding material were determined using the
ORIGEN code. The mixture ratio of nuclides originating from fuel and structural materials has been set based on the results of a severe
accident analysis code (MAAP code). Neutron and gamma rays flux in the area around the S/C bottom were evaluated using a 3-dimensional
1/16 scale model of the S/C and torus room. An assessment of the effect of background gamma radiation (Cs-134 and Cs-137) from stagnant
water was conducted. To measure acceptable background radiation level and radio-sensitivity, a B-10 neutron detector and a CdTe gamma
rays detectors were selected as the best choice for the task andtheir responses evaluated at the locations in
the S/C and torus room.
Feasibility of non-destructive detection for the
S/C bottom with or without radioactive
Stagnant
water
トーラス室
level in the torus
materials was verified technically. Design and
滞留水水位
Filled with waterroom
production for measurement systems anddevices to access are will be decided based on
the results of developments of methods torepair and stop water leakage.
Cs-137 measurements
Eu-154 measurements
Cs-244 measurements
Implemented byInternational Research Institute for Nuclear Decommissioning
Fig. 2: Background radiation
Fig. 1: Distribution of radiation originating(Unit 1)from fuel around the S/C (Unit 1)(IRID)FY2011F2012

FY2013

FY2014FY2015FY2016FY2017FY2018Development of technologies for non-destructive
detection of radioactive material deposited in S/C, etc.

Fig. A11-2-(9): Development of technologies for non-destructive detection of radioactive material
deposited in S/C, etc.

171

Project objectives
To enable decontamination of the interior of reach reactor building contaminated by the scattering of radioactive materials without causing
excessive radiation exposure to workers, studies were performed on different decontamination methods for different decontamination targets(such as concrete, metal, and resin) and for different types of contamination (such as loose contamination, tightly adhering contamination, andpenetrating contamination) was examined. Then, apart from the remotely controlled decontamination machine for low locations surfaces that
had already demonstrated its performance in the field, development was performed for remote controlled decontamination machines that canbe applied to surfaces in high locations or on the second or upper floor levels as machines that are vitally needed to prepare a proper working
environment for decommissioning works.
Arm/head
1. Project details and progressArm(2 sets)(1) Development of a decontamination machine for surfaces in high locations
Arm
In order to use it for decontamination work at the Fukushima Daiichi NPS,
movement
decontamination machine for surfaces in high locations meeting the following
performance and functional requirements was developed and demonstration
Elevatingtests were conducted.
cart- Decontamination performance of the target dose rate (3 mSv/h or less forwork areas, 5 mSv/h or less for access areas)
Dry ice blast- Remote operability , travel performance, and arm operability
Cart (in moving)
High-pressure water jet
decontamination
Suction/blast decontami- Recovery function in case of failure, toppling prevention, etc.
decontamination equipment equipment
nation equipment(2) Development of a decontamination machine for upper floor levels
Fig. 1: Decontamination equipment for high places
In addition to the performance and functional requirements, a decontamination machine was developed and tested for use on the upper floors,which must meet the following design requirements for use on the upper floors.
- Capability to use a general-purpose elevator workbench for round trip with for the upper floor- Capability to round trip from upper floor levels as quickly as possible- Capability to can operate in the structure of the upper floor, equipment layout(3) Examining the concept of decontamination in the basement floor level
If the accumulated water level in the basement decreases because of the retrieval of fuel debris, theimplementation of measures concerning contaminated water, or any other reason, the air dose rate
may increase due to the scattering of dust and shielding effect by water. Therefore, technicalchallenges were examined and formulated for attempts toward lowering the air dose ratio in theFig. 2: Upper floor decontamination equipment
basement, including the following in the list of items that would require attention.
- Changes in plant conditions (changes in the air dose rate, accumulated water radioactivity concentration, and dust concentration)- Scenario for environmental improvement including the decontamination of the basement floor- Combination of different approaches such as decontamination, shielding, and dust collection
Implemented by International Research Institute for Nuclear Decommissioning (IRID) (FY 2013 -)2. Related projects
The following describes the results of previous related projects.
○ Development of remote decontamination technology in the reactor building (FY 2011 - 2013)(FY 2011)- Contamination survey: A field survey plan was produced that involved the determination of survey areas, theFig. 3: Investigation result by
preparation of ideas about sampling tools, the listing of analysis targets, and so on.
γ camera in the south side
- Dummy contamination test: A method for preparing a dummy contamination test piece such as a concreteof Unit 1
sample was examined.
- Decontamination machine: Technological options were studied by making a public call for technical catalog proposals.
(FY 2012)- Collection of basic data: By taking measurements on concrete core samples, it was found that there was no penetration of contamination into
the concrete interior, it was confirmed that tightly adhering contamination is in scars produced by the aging degradation of epoxy coating on
the surface.
- Verifying the appropriateness of decontamination techniques: Based on the analysis of the collected basic data and the results ofdecontamination testing performed on dummy contamination samples, it was found that it would be possible to successfully address different
types of contamination on different decontamination targets through high-pressure water jet decontamination, dry ice blast decontamination, or
blast suction decontamination, and therefore it was concluded that these decontamination techniques were appropriately selected.
- Demonstration of remote controlled decontamination: By in-factory mock-up testing and demonstration testing, it was concluded that it would
probably be possible to successfully put a remotely controlled decontamination machine to practical use, and challenges and necessary
improvements were identified.
(FY 2013)- Collecting basic data on contamination: Surveys were conducted on the dose rate at upper floor levels and in high location surfaces in reactor
buildings, the distribution of contamination, internal radiation sources, and the penetration of contamination.
- Organization of decontamination techniques and examining decontamination concepts: A basic policy was established for thedecontamination machine for surfaces in high locations and the decontamination machine for upper floor levels.
- Design and fabrication of remotely controlled decontamination machines and demonstrative decontamination testing: Design was performed
for the decontamination machine for surfaces in high locations and thedecontamination machine for upper floor levels. The decontamination
machines that had been fabricated in FY 2012 were modified and put todemonstration testing and performance evaluation.
Dry ice blast- Demonstration of the shielding system for practical application: The
High-pressure water
Suction/blast
decontamination
shielding installation plan for remote controlled was finalized, and thedecontamination
decontaminationequipment
equipment (Hitachi GE)
equipment (MHI)(Toshiba)
shielding system was put to in-factory demonstration testing and evaluation.
FY2011FY2012FY2013
FY2014FY2015FY2016FY2017Development of remote decontamination technology in the building
Development of remote decontamination technology in the reactor building

Fig. A11-2-(10): Development of remote decontamination technology in the reactor building172

Project objectives
In order to reduce the exposure of workers to radiation in the decommissioning of the Fukushima Daiichi NPS as they engage in work inside
reactor buildings, a comprehensive dose reduction strategy was developed combining various dose reduction techniques with remotely
controlled decontamination techniques.
Project details and progress
In FY2012,againt the first floor level of Units 1 through 3, the fifth floor level of Unit 2, and the work areas in Units 1 and 3 at floor levels
impacted by hydrogen explosions, in FY2013, against areas of relatively high dose rate inside reactor buildings, and the surfaces in highlocations at the first floor level in respective reactor buildings, the higher floor levels in respective reactor buildings, the southern part of the firstfloor level, and access areas in staircases, etc. were addressed. Work was conducted toward establishing a dose reduction strategy that would
enable the achievement of the target dose rate (3 mSv/h) using various dose reduction techniques like decontamination. In addition, study was
performed and information was collected on dose reduction techniques developed outside Japan, and their effectiveness was examined and apublic call was made for technical offers from overseas entities.
Ceiling surface(1) Analyzing dose distribution in buildings
Data from dose measurements performed in buildings was compiled as basic data to be used in theWall surface
development of a dose reduction strategy. The dose rate contributions of radiation sources, such as
contaminated floor and wall surfaces, were modeled from the distribution of the dose rate in work areas. In
Measurethis way, it was found that radiation from floor/wall/ceiling surfaces and also from the ducts and pipes close
Hot spotpoint
to the ceiling in spaces devoid of hot spots contributed significantly to the dose rate.
(2) Evaluating and selecting dose reduction techniques from among the decontamination, removal, andFloor faceshielding options
Distribution model of radiation dose rate
Conditions were clarified regarding the selection of decontamination, removal, or shielding options forradiation sources such as floor and wall surfaces. The effectiveness ofexisting techniques was evaluated, and issues were identified.
(3) Working toward the establishment of dose reduction strategy
Trying to determine the optimal dose reduction technique for floor and wall
surfaces, etc., at each reactor building, it was found that radiation from thehigh location in the building contributed significantly to the dose rate. Based
Case of approaching to theon this finding, a strategy was established to achieve the target dose rate by
Current distribution of facilities of ceiling direction
Case of Shielding
preparing machines that would enable the remote decontamination, removal,
radiation dose rateand decontaminating
partiallyor shielding of target components, particularly the components installed close to the ceiling. That strategy was established by additionallyperforming detailed surveys to determine the radiation source intensity of contaminants attaching to the target components.
(4) Employment of techniques imported from abroad
To employ techniques owned by overseas entities with sufficient experiencein working in high-dose or high-contamination environments, study was
performed and information was collected on dose reduction techniques heldby the entities mentioned below. It was concluded that these techniques
may be effectively applied to solve issues when they are improved or further(1) Remote measurement system(2) Remote sampling tool
developed to suit conditions at the Fukushima Daiichi NPS sites.
of radiation dose(AREVA)(Babcock)① AREVA (France): 3D contamination distribution capture by remote control in a high-dose environment and dose rate contribution evaluation② Babcock (Britain): Technique for remote contamination analysis in a high-dose environment③ CH2MHILL (U.S.): Technique for decontaminating in a high-dose environment (cable trays, ducts, gratings, power boards, instrumentation
racks, etc. in high location)
In FY 2013, the public offering was conducted for overseas entities on specific solutions that may contribute to the development of dose rate
reduction programs, and the proposal was summarized.
Implemented by ATOX Co., Ltd.
FY2011F2012

FY2013

FY2014FY2015FY2016FY2017

Formulation of comprehensive dose rate reduction plan

Fig. A11-2-(11): Formulation of comprehensive dose rate reduction plan173

FY2018

Project objectives
To develop the canister necessary for reliably and reasonably containing, transferring and storing the retrieved fuel debris taking its
characteristics into account, as well as to establish conceptual idea of the system for containing, transferring and storing of the retrieved fueldebris in conjunction with the ongoing effort by the project “Establishment of the fuel debris retrieval method”.
1. Project details and progress
The following results were produced from ongoing projects described in Section 2 below.
(1) Survey on transferring and storing of fuel debris and producing research plan
Knowledge was expanded and accumulated through a minute investigationon the licensing documents of the Waste Management Facilities of Japan
Nuclear Fuel Ltd. at Rokkasho and on gathered overseas information
regarding transferring and storing of the damaged fuel for contributing study
specified in (2) through (4) below.
(2) Study on safety requirements and specifications of the canister, andstoring system
A flow of a series of process of containing, transferring and storing of theretrieved fuel debris was determined and a primary assessment of thethroughput (processing amount of fuel debris, number of the canister,necessary area for storing etc.) of above mentioned process under thepredefined set of assumptions was performed. Some ideas for reducing a
number of the canister were also investigated. It is also confirmed thatconceptual design of dry or wet storing system for retrieved fuel debris neednot to be modified based on review of the outcome from related research
programs.

Micromanometer

Compoundgauge

Thermocouple
Piping for gas/
test water supply

Safety valve
Piping for gas
collection
Test container
Test water
Inner
Spent fuelcontainerpellet piece(1) Testing(2) Test equipment configuration
Fig. 1: Test measurement of amount of hydrogen generated using
spent fuelCanister(3) Development of safety assessment methods and verification of safety
Evaluation of convention flow state in
Various studies and tests were performed to verify the safety of the fuel debris
Fuel debris
canister (hatched part of the left figure)canister. For example, the applicability of water content restriction was studied,(1) Evaluated object(2) Evaluation of convention
as well as possibility of relaxation of the sub-criticality condition was examinedflow state (example)
for increasing inner diameter of the canister which could increase operability ofFig. 2: Evaluation of convention flow state in canister
fuel debris retrieval work. Lid structure of the canister was also studied in orderto satisfy safety requirements and aging phenomena of the canister material
was assessed taking environmental condition during fuel retrieval work in to account. The Hydrogen generation test was performed using
spent fuel (Fig.1) and convective flow inside of the canister was evaluated for determining the effectiveness of hydrogen recombination
catalyst (Fig.2). From such examination results, the issue on generated hydrogen gas was addressed and the knowledge required for theestablishment of an approach to safety assessment was expanded and accumulated.
(4) Studying on the design/specification of the canister for retrieved fuel debris
It is confirmed that the design/specification of the canister which had been established last year need not to be modified from the outcomesfrom safety requirements and safety assessment results mentioned in (2) and (3) above.
Above actions were implemented by International Research Institute for Nuclear Decommissioning (IRID)
2. Related projects
The followings are the results of related previous projects.
○ Development of containing, transferring and storing technology of fuel debris(1) Survey on the transferring and storing of the damaged fuel (FY 2014 - 2015)
The following overseas experiences that pertain to the transferring and storing damaged fuel were studied.
- Transferring of the fuel debris from TMI-2 (U.S.) to INL, the wet storage of fuel debris, and dry storage of it after drying- Design and safety assessment capability for canister and drying technology by PNNL (U.S.)- Study and evaluation of characteristics of MCCI products conducted at ANL (U.S.)- Transferring and storing of damaged fuel at Paks Unit 2, Hungary- Transferring of damaged fuel in France- Radioactive waste disposal technologies performed at Sellafield (Britain)(2) Study on fuel debris containing method into the canisters (FY 2013 - 2016)
The basic specification and geometry (incl. lid structure and structural buffer) of the canister were established through the definition of therequirement for the canister suitable for work form containing to storing of the fuel debris in coordination with other related projects.
Furthermore, a work flow up to transferring and storing of them was proposed and the basic specifications of the canister handlingequipment, etc. were established.
(3) Study on safety requirements, specifications for the canister and its storing systems (FY 2013 - 2016)
Preferred scenario was selected regarding the containing of fuel debris into the canisters, transferring, and storing from the viewpoints ofsafety assurance, work efficiency, etc., and it was reviewed repeatedly taking the latest inputs from related projects into consideration.
(4) Development of an approach to safety assessment and verifying safety (FY 2013 - 2016)
Literature search and survey of overseas cases were performed to investigate more about approaches to the safety assessment of thecanister for fuel debris (maintaining sub-criticality, structural integrity, and counter measure against hydrogen generation). Necessary data
for safety assessment of the canister was obtained, and a general outlook on safety assessment was confirmed through performing
experimental analyses and component tests based on assumed the fuel debris characteristics.
FY2011FY2012FY2013
FY2014FY2015FY2016FY2017FY2018Development of containing, transferring and storing technology of fuel debris
Development of containing, transferring and storing technology of fuel debris
Development of containing, transferring and storing technology of fuel debris
Development of containing, transferring and storing technology of fuel debris

Fig. A11-2-(12): Development of containing, transferring and storing technologies of fuel debris

174

Project objectives
Aiming to obtain the prospects of a processing/disposal method and technology related to its safety that should be made clear by around FY
2021, the efficiency of solid waste characterization shall be improved with considering features of waste generated by the accicent, andresearch and studies shall be conducted toward establishing processing technologies, disposal concepts, and its safety assessment methodon the basis of the characterization results. In addition, technologies necessary to reduce the risks associated with the storage of solid waste
shall also be developed.
1. Project details and progress
The following describes the results of this project so far reflecting results from the related projects described in Section 2 below.
(1) Characterization① Analysis data collection, management, etc.
○ Identifying the distribution of contamination- Analysis of “rubble, etc.”, secondary waste generated by water treatment, and contaminated water have been carried out continuously.
- As to “rubble, etc.”, analysis was conducted for those such as rubble from soil covered temporal storage, and material from the Unit 4 reactor
building. In addition, the distribution of contamination in parts of samples was also studied.
- As to secondary waste generate by water treatment, analysis was conducted for sludge from decontamination systems as well as slurry andadsorbers from multi-nuclide removal systems.
- As to contaminated water, stagnant water and treated water were analyzed. The concentration of Sr-90 appeared to be correlated with theconcentration of Cs-137 at a certain ratio over the range of about two orders of magnitude (Fig. 1). This finding will be used in characterizing
secondary contamination cause by contact with stagnant water, in terms of contributing nuclides.
- In order to study the nuclide migration pattern contributing to contamination, the concentrations of different nuclides were studied incomparison with Cs-137 using existing data. With every reactor unit, samples of rubble from the reactor building showed correlation between
the Sr-90 concentration and the Cs-137 concentration (Fig. 2). Similar migration patterns were also observed among other nuclides.
○ Development of sampling techniques, etc.
- For the sampling of zeolite from a cesium adsorption apparatus, a method of drilling at the top of apparatus was studied, and a test
equipment for trials was produced (Fig. 3). For remote sampling operation inside the reactor building (R/B), trial use was conducted with asmall core boring/sampling tool.

Fig. 1: Correlation of Sr-90 and Cs-137 in stagnant water

Fig. 2: Correlation of Sr-90 and Cs-137

Upperlayer of zeolite
ゼオライト上層
Simulated column
模擬カラム
Sampling サンプリングヘッド
headOutside 外形寸法：φ27×200
dimension: φ27x200mmmm
3
採取量：約20
Sampling amount:cmabout
20cm2
Outsidedimension: w800xl1000x2435mm
外形寸法：W800×L1000×H2435mm
ストローク：1000mm
Stroke:1000mmFig. 3: Sampling test equipment of adsorbent
from cesium adsorption apparatus

in rubble inside of R/B○ Improving the efficiency of waste characterization- The patterns of nuclide composition of contamination were broken down, and studied the classification of nuclide migration patterns ofcontamination. The selection of nuclides for analysis was reviewed, and it was decided to make supplementary use of fuel burnup calculationin the calculation of isotope concentration in order to improve the efficiency of analysis.
- A method to evaluate the representativeness of analysis data was developed, and progress
was achieved in the analysis of contamination mechanisms. The nuclides for analysis werereselected, and techniques were developed to simplify and speed up analysis on the selectednuclides, seeking to improve the efficiency and reasonability of waste characterization.
○ Management of analysis data
データ頻度
確率密度
- Disclosed analysis data was published as reports. For the convenience of users, work was（対数正規分布）
conducted toward enabling publication on the internet and at the same time toward enabling thesharing of data on the management of samples and analysis activities among relevant parties.
② Improving the accuracy of analytical assessment techniques- The nature of frequency distribution of contamination by radioactive nuclides was identified as
lognormal, and estimations of radioactive nuclides had been tried based on this finding (Fig. 4).
The inventory of radioactive nuclides of waste was provisionally estimated using an improved
assessment method.
- The accuracy of analytical inventory assessment was improved by taking account of variationsof analysis data in the estimation of inventory.
③ Compiling the results of comprehensive inventory assessments
Fig. 4: Examples of analysis value distribution (upper
The concept of a system was been established to determine and update the radiologicaldrawing) and inventory assumption (lower drawing)
inventory by conducting comprehensive assessment of the values from the analysis of actual
samples and the values estimated using an analytical assessment method.
④ Addressing the harmful substances for disposal, etc.
- Typical harmful substances may cause influence to the disposal system were identified by studying actual examples such as waste
acceptance criteria of overseas repositories for harmful substances. Based on the result, studies were conducted on the views with harmful
substances regarding allowable concentration of waste acceptance criteria and contained amount within waste. Studies were also conductedon the method used to analytically assess the impacts of harmful substances on the safety of predisposal management and disposal itself, andcorrection and compilation of data required for assessment will be conducted.
(2) Pre-disposal management① Study and assessment on the method of storage adequate for feature of solid waste○ Study of measures implemented in the storage of highly radioactive waste- Research was conducted and information was collected from a wide range of knowledge/experience at home and abroad concerning
measures may be taken to address the issue of hydrogen generated during the storage of highly radioactive solid waste such as about waste
specifications, container specifications, hydrogen generation evaluation methods, laws and regulations, technical requirements for facilities,
measures taken to address the issue of hydrogen generation. As a result, it was identified that, even with solid waste, it would be a valid
solution to use a vent-equipped container for storage, with minimizing the presence of free water, and defining a G value specific to theproperties of waste. In order to study the applicability of such method to solid waste of specific types such as HIC slurry, and identify
challenges, deeper research has being conducted into the approaches to hydrogen generation by conducting case studies, based on
regulations and actual performances of other countries.

175

- The method of storage has being studied for “rubble, etc.” to be generated from the process of fuel debris retrieval.
○ Technologies for the stabilization and solidification of secondary waste generated by water treatment- In-drum vitrification technology was focused as a stabilization technology required forpre-disposal management of secondary waste generated by water treatment, in the
Solid content
context of preventing the spread of contamination and the size of the equipment.
Aboveground floor: Decontaminant
Simulated solidified waste was produced with the process of simultaneously melting(AREVA) installation area
Actual collected sludge
zeolite used for contaminated water treatment together with other simulated secondary(10ml vial container)
waste generated by water treatment. Then, basic data of those solidified waste wascollected such as the results of performing chemical analysis and leaching tests, andcomposition data and conditions for producing solidified waste required for engineering
scale testing were presented.
- Conditions inside Tank D in the main process building, where sludge from the
The condition of sludge atdecontamination systems is present, were studied, and the actual sludge was sampled
the bottom of storage layerOverview of investigation inside D (Yellow)(Fig. 5). Through observation with a submerged camera, it was found that the sludge
storagelayer-D(Cross-section)
sedimentation layer in the tank had a thickness of about 40 cm, and the volume wasFig. 5: Sampling of sludge from decontamination equipment
estimated to be around 37 m3. To prepare for sludge removal, simulated sludge wasproduced and tested to obtain basic data on sludge fluidity.
② Selection of preceeding processing method adequate for feature of solid waste- In order to identify the processing technologies that may be applied to solid waste on the projected scale from those have been proventhrough actual implementation (high-temperature treatment and improved cement treatment), necessary data will be collected and assessment
will be performed using an engineering scale test equipment, etc. At that time, assessment will be conducted on how the processing
technologies may be impacted by the salt that waste is likely to contain, radiation, heat, and so on.
③ Contamination assessment techniques for segregation of solid waste- To enable the reliable segregation of waste, research and studies were performed on methods for the measure/assess contamination byalpha emitting nuclides, and so on, and techniques were developed with due attention to the implementation in the field.
(3) Study of disposal concepts and safety assessment methods adequate for the feature of solid waste
To collect necessary information toward the establishment of disposal concepts and safety assessment methods, in-depth study was
conducted on disposal sites abroad (Fig. 6). In addition to that, study was conducted on the conditions for the case of applying solid waste fromthe Fukushima Daiichi NPS based on various cases, with attention also to earlier examples of waste disposal in Japan and the features of solid
waste. Furthermore, the items of the specification may have an impact on the result of safety assessment for disposal will be identified for thesolid waste processed by introducing preceeding processing.
(4) Integration of R&D results and study of the waste stream
Regarding the waste stream, latest results derived from previous researches
are reflected to it, and establishment of methodology to integrate progress,
consistency of results, and challenges to be overcome, is made to start.
Implemented by(Comprehensive proposal) International Research Institute for Nuclear
Fig. 6: Detailed survey of overseas repositories
Decommissioning (IRID) (FY 2013 -)(Partial proposal) IHI Corporation (FY 2018 -), Orano ATOX D&D Solutions Co., Ltd. (FY 2018 -)2. Related projects
The following describes the results of previous related projects.
○ Development of technologies for processing/disposal of secondary waste generated by treatment of contaminated water (FY 2012)(1) Characterization of spent adsorbers, sludge, etc.
- Nuclides in contaminated water and treated water were analyzed, and progress was pursued in the analysis of nuclides that are difficult tomeasure.
- Work continued in estimating the distribution of cesium adsorbed into the adsorber inside an adsorption tower and in performing tests tocharacterize the properties, such as the thermal stability of spent zeolite. It was not possible to perform sampling and analysis according to theschedule due to the high dose rate from the sludge itself and in the surrounding area.
- Progress was made in the collection of information regarding the basic properties of secondary waste from newly employed contaminatedwater treatment systems (cesium adsorption system No. 2 and multi-nuclide removal system).
(2) Studying long-term storage- Corrosion testing was performed on sludge storage container material and cesium adsorption apparatus material to collect electrochemical
data, etc., pertaining to container corrosion that could contribute to the evaluation of long-term storage.
- Based on the results of spent zeolite property characterization, the hydrogen concentration and the temperature inside adsorption towers
were estimated as safety-related information.
- Heat flow analysis was performed on sludge under the current storage conditions, and the appropriateness of measures that had been taken
to address the issue of heat was verified.
- A literature search was begun on adsorbers used in the multi-nuclide removal system, etc.
(3) Research on the conditioning techniques- The results of research conducted were compiled on techniques used for conditioning of spent zeolite, sludge, etc.
- By conducting basic experiments on conditioning such as using the cement solidification technique, progress was made in the collection ofdata required for assessing the implementation feasibility of conditioning techniques.
- Information was obtained about secondary waste from the multi-nuclide removal system (types, predicted quantities, etc.).
○ Development of technologies for processing/disposal of radioactive waste (FY 2012)(1) Characterization of “rubble, etc.”- The radioactivity of rubble and felled trees was analyzed to collect analysis data required to know features on the contamination status ofwaste.
(2) Development of techniques to analyze nuclides that are difficult to measure- To address nuclides such as Zr-93 and Mo-93 that are difficult to measure and therefore require the development of analysis techniques,
literature searches were performed and analysis flows were studied.
(3) Study on the development of infrastructure for R&D on waste processing/disposal- The study was conducted on the components of R&D as well as those solutions necessary to obtain prospects of a processing/disposal
method and technology related to its safety. In addition to that, information was compiled such as needs of users, future possibility ofdeveloping a database, and conceptual design was implemented for its setting up.
(4) Establilshing an R&D program on waste processing/disposal techniques- Referring to a technical development program discussed within the special expert committee of the Atomic Energy Society of Japan, a draft
R&D program for waste processing/disposal techniques was produced.

176

○ Research and development of processing/disposal of solid waste (FY 2013 - 2014)(1) Characterization- Rubble, felled trees and secondary waste generated by water treatment were transported to JAEA for radioactivity analysis. A proportional
relationship was found between Cs-137 and Sr-90 in rubble and felled trees in terms of radioactivity concentration.
- The radioactivity (i.e. the inventory of radioactive nuclides) in secondary waste generated by water treatment (cesium adsorption apparatus,etc.) was estimated based on the concentration analysis data of treated water.
(2) Studying long-term storage- The conditions inside cesium adsorption apparatus were estimated based on data from simulation test. It was noticed that corrosion was
suppressed by the coexistence of zeolite.
- The technique to be used to stabilize slurry from the multi-nuclide removal system was selected, and test was conducted using simulated
samples.
(3) Studies on waste processing- Experiments were conducted with the solidification of slurry and spent adsorber from the multi-nuclide removal system using solidifiers ofvarious types.
- A catalog of existing waste processing and conditioning techniques was compiled.
(4) Studies on waste disposal- Assuming the application of existing disposal concepts to the disposal of waste generated by the accident, safety assessment methods(scenarios, etc.) in consideration of the features of waste was provisionally defined.
- In defining analysis cases for the postulated scenarios, provisional assessment was performed for the safety of disposal of different types ofwaste.
(5) Studying postulate of R&D- Waste generated by the accident was classified by property, contamination history, etc. And an example of handling flow consisting withstorage, processing, and disposal and those may realize safe disposal was defined for each classification of waste.
- The collection of analysis data was expanded or updated with newly obtained data. In addition, the relationships between major developmenttasks of processing/disposal techniques and various information items were defined to prepare for the development of information
management tools.
○ Research and development of processing/disposal of solid waste (FY 2015 - 2016)(1) Integration of R&D results- Regarding the waste stream, a technique was developed to narrow down streams that
presents multiple options, and case studies were conducted to identify challenges in applying
the technique.
- At a study meeting held by OECD/NEA expert group to address issues around waste
generated by the Fukushima Daiichi accident, contribution was made to compile a report(published in December 2016).
(2) Characterization- Rubble, secondary waste generated by water treatment, contaminated water, and so on, weretransported to an offsite facility for radioactivity analysis. Data from the analysis of rubble(concrete) from the Unit 1 reactor building (R/B) was found to be consistent with existing datafrom the analysis of rubble (Fig. 7).
- Work was conducted toward reducing uncertainty in the analytical estimation of inventory, andFig. 7: The ratio of radioactive concentration
inventory data sets were produced using the improved estimation technique and madeof Sr-90/Cs-137 in rubble
available for use in safety assessment for disposal.
(3) Studies on waste processing and long-term storage- A conceptual design was studied for equipment to be used in the stabilization of slurryfrom the multi-nuclide removal system based on the results of studies on operational aspectof introducing it to the field and verification tests.
- For secondary waste generated by water treatment with which the solidification had notbeen attempted in the past, basic tests were performed to experiment with solidification
through existing techniques (Fig. 8) to determine the feasibility of solidification and collectdata on confirmation of the integrity of solidified waste. Comparing the data with the
Glass solidification
Cement solidification
requirements defined for technical assessment, applicable candidate techniques for eachtype of waste were assessed.
- Accelerated test was performed for residual water evaporation from cesium adsorption
apparatus due to heat generation and the evaluation of increase in salt concentration, and amethod for the verification of the salt concentration evaluation results was developed.
(4) Studies on waste disposal- Data was compiled including information on the examples of disposal concepts from
Sintered solidification
Compressed solidification
abroad that may be referenced in the development of disposal concepts, and the features ofFig. 8: Various type of solidification of simulated waste
existing disposal concepts were organized.
- A safety assessment method was developed to evaluate disposal categories with consideration of uncertainty.
FY2011FY2012FY2013
FY2014FY2015FY2016FY2017FY2018Development of technologies for processing/disposal of secondary waste generated by treatment of contaminated water
Development of technologies for processing/disposal of radioactive waste
Research and development of processing/disposal of solid waste
Research and development of processing/disposal of solid waste
Research and development of processing/disposal of solid waste

Fig. A11-3: Research and development of processing/disposal of solid waste

177

Project objectives
In the decommissioning of the Fukushima Daiichi NPS, to safely store the fuel assemblies removed from the spent fuel pool that was affectedby the hydrogen explosion and seawater injection in the common pool over the long term as well as to examine the feasibility of future dry
storage, technologies have been developed to obtain the necessary data for the examination of decisions on the necessity of the evaluationand management methods for long-term integrity assessment and long-term wet and dry storage.
Project details and progress
The following describes the results this project produced so far based on results from the related projects described in Section 2 below.
(1) Development of technologies for assessing the long-term integrity of fuel assemblies
Evaluation on the deposits on the surface of the fuel assemblies
Locknut
Materials of fuel assemblies (lock nuts) of Units 4 at theat Unit 4
Fukushima Daiichi NPS stored in the common pool weretransported to the post irradiation test facility, and component
analysis for white deposits and measurement of corroded crevice
re-passivation potential were carried out. Among the constituentsof the white deposit, the amount of Mg was the largest, and theamounts of Al and Si were about half of it. The amount of Cl wasFig. 2: Re-passivation potential
Fig. 1: Component analysis result of white
below the detection limit. Since Mg (OH)2 was separated and Cl
measurement result on corroded
deposition area of locknut at Unit 4
crevice of locknut
was not captured, it is considered that there is no possibility ofEnlarged view
Fuel
Damagedcorrosion (Fig.1). In the electrochemical test, there was no crevice
cladding
part
corrosion sensitivity in the area where the chloride ion
concentration was lower than 100 ppm, and it was confirmed that
there was almost no possibility of corrosion in the common pool ofless than or equal to 1 ppm (Fig.2).
HydrideEvaluation of integrity of fuel in dry storage conditions
Assuming dry storage of the fuel assemblies from the spent fuel Fig. 3: Hydride precipitation behavior
verification test result (Irradiated test
Fig. 4: Creep speed test result (Irradiatedpool of 1F, in order to confirm integrity of the fuel assemblies
piece: 300℃, cooling velocity: 0.04℃
test piece: 360℃, stress in a circumferential
affected by rubble falling and seawater components, which were /h, stress in a circumferential direction:
direction: 170MPa, attached scratches and
70MPa, attached scratches andseawater, fixed rubble)
to be stored in the dry system, a hydride precipitation behavior
seawater, fixed rubble)
verification test and a creep test were carried out, and the impact of the factors unique to 1F on material properties was evaluated. It was
confirmed that the impact on hydride precipitation behavior and creep behavior was small even in the condition affected by both rubble
damages and seawater injection (Fig. 3 and 4).
(2) Basic tests for long-term structural integrity
A test to evaluate seawater components transfer behavior in the crevice structure of fuel components was carried out, and it was found that theseawater components did not concentrate in the crevice structure and it changed according to the salt concentration outside the crevice
Implemented by International Research Institute for Nuclear Decommissioning (IRID) (FY 2013 -)2. Related projects
The following describes the results of previous related projects.
○ Evaluation of long-term structural integrity of fuel assemblies removed from the spent fuel pool (FY 2012 - 2014)(FY 2012, Fig. 0)- A corrosion test and a strength test were conducted using theunspent materials collected from irradiated materials. With this, the necessary data
was obtained for water quality as well as the impact of irradiation on theexamination of the long-term integrity of fuel assemblies.
- The experiment confirmed that the possibility of corrosion, etc. was low under the
Artificial seawaterElectrochemical
test of irradiated Crevice corrosion test under immersion test ofenvironment with the actual machine, such as seawater and gamma ray irradiation.
gamma
ray
irradiation
using
irradiated fuelfuel cladding
no-irradiated test piece(FY 2013 - 2014)- Examination of the test conditions for the assessment of longcladding
Structural integrity
term integrity :-Handle-Top tie plate
A test plan was developed, including a transportation plan, material matrix, and-Tie rod bolt joint
test manuals for fuel members that had been relocated to the common pool (Fig.
Cladding
airtightness
1).
-Fuel cladding- Development of technology to assess long-term integrity: A corrosion test and astrength test were conducted to evaluate the impact of corrosion, such as debris
integrityand stress, with a specimen that simulates the fuel members. It was confirmed that Structural
Test piece of simulated upper part
-Bottom tie plate
no corrosion or deterioration in strength that could impact the integrity was found Fig. 1: Evaluation items for long-term
Fig. 2:Appearance of locknut,
integrityof
fuelassembly
removedfrometc. after corrosion teston the screw part or in the cladding tubes in the area to be assessed. (Fig. 2).
spent fuel pool during wet storage- Survey of condition of spent fuel stored in the common pool: The appearance ofthe fuel assemblies that had been relocated from Unit 4 was observed, and theoxide film thickness of cladding tubes was measured. No abnormal corrosion wasobserved in the spent fuel investigated (Fig. 3).
- Evaluation of integrity of fuel in dry storage: Assessment was conducted for the
impact of hydrate separation behavior in cladding tube materials as well as thewater content of debris on the integrity of the fuel assemblies during dry storage.
Fig. 3:Appearance of top tie Fig. 4:Relationship between re- Evaluation of seawater component transfer behavior on fuel components:
plate of spent fuel of Unit 4 passivation potential for crevice
corrosion in diluted artificial seawater
Evaluation was conducted for the relocation of seawater component to the cladand chloride ion concentrationand the oxide film of cladding tube, etc.
- Evaluation of effect of corrosion derived from seawater or rubble components under radiation environment: An electrochemical corrosion test
under the gamma ray environment and a corrosion test were conducted using a specimen with a combination of zircaloy and stainless steel(Fig. 4).
FY2011FY2012FY2013
FY2014FY2015FY2016FY2017
Evaluation of long-term structural integrity of fuel assemblies removed from the spent fuel pool
Evaluation of long-term structural integrity of the fuel assemblies removed from spent fuel pool

Fig.A11-4-(1): Evaluation of long-term structural integrity of the fuel assemblies removed fromspent fuel pool

178

Project objectives
It can be thought that the salinity is due to the seawater adhered to the fuel of the spent fuel pool, and there is also the possibility that some of
the fuel has been damaged or leaked due to fallen concrete pieces, etc. For this reason, technical issues in the reprocessing of these fuels
were investigated and examined, and the development of indicators to decide the necessity of reprocessing was also examined.

Relative corrosion rate

Project details and progress(1) Evaluation of corrosive influence of impure substances on re-processing equipment
Corrosion tests (immersion, electro chemistry) was
conducted on materials in high-level concentrated
waste liquid storage tanksand high-level liquid waste
storage tanks, using simulated high-level liquid waste
solutions that took into account impurities and fission
products, etc. Under all conditions, results showed that
(1) RRP simulated (2) RRP simulated waste
while there was uniform corrosion in the form ofsolution + artificial seawater
waste solution
intergranular corrosion, pitting was not observed(chloride ion 0g/l) (chloride ion 20g/l)
Observation magnification: x50(258x257μm)(Fig.1). Furthermore, an increase in chloride ion
Fig. 1: Surface observation result of immersion test
Chloride ion concentration (g/l)
concentration was accompanied by a decrease inpiece of high-level liquid waste storage tank material Fig. 2: Corrosion rate of high-level liquid waste
corrosion (Fig.2).
storage tank material (relative corrosion rate:
corrosion rate of RRP simulated solution(2) Evaluation of in-process behavior of impurities
containing no impurities is valued as “1”)The extraction operation of impurity was conductedunder the condition of coexistence with FP to checkthe impact of FP on the extraction of impurity. As aresult, the distribution ratio of impurity was as low as
10-2 to 10-3 order (Fig. 3). In addition, the extraction
operation of U and Pu was conducted under theStandard
10x Standard 100x Standard
condition of coexistence with an anion, with theconcentration concentration concentration
possibility of obstructing the extraction of U and Pu,
Fig. 4: Homogeneity results of glass specimens
and the impact of the anion on the extraction of U and
Fig. 3: Distribution ratio of impurity components (visual observation)
under the condition of coexistence state with FP
Pu was checked. As a result, it was confirmed that the
distribution ratio of U and Pu would not be influenced by the anion.
(3) Evaluation of the influence of impurities on waste body
To determine the composition of waste solutions, a glass specimen was created from a formulation of powder materials with the major
components of seawater and mortar added as impurities, and an evaluation of homogeneity and so on was then performed. The result wasthat, under all conditions, there was no phase separation precipitation, and vitrification occurred (Fig.4).
(4) Identifying and outlining of other influences
The influences of the processing of damaged fuel in reprocessing facilities were comprehensively extracted, and the presence or absence of
required research elements and the findings obtained in this research were summarized.
Implemented by International Research Institute for Nuclear Decommissioning (IRID)
FY2011F2012

FY2013

FY2014FY2015FY2016FY2017FY2018Study of Methods to Process Damaged Fuel Removed from the Spent Fuel PoolFig. A11-4-(2): Study of Methods to Process Damaged Fuel Removed from the Spent Fuel Pool179

Project objectives
Concerning the contaminated water to be generated at the Fukushima Daiichi NPS, efforts are being made to remove the 62 kinds of nuclides,
but tritium remains because it cannot be isolated. Therefore, verification tests on the technologies of isolating tritium were conducted. Morespecifically, to verify the performance of separation in the water with the tritium generated within the power station (6.3 x 106 Bq/L to 4.2 x 106
Bq/L (the concentration depends on the timing of sampling)), an arbitrary-scale facility was built and verification tests were conducted toassess the performance of separation, the construction cost and the running cost.
Project details and progress
An arbitrary-scale facility was built, and assessment of the performance of separation and costs, etc. in the actual plant was
conducted as Category A (three projects), and assessment of the performance of separation and costs in the laboratorylevel tests was conducted as Category B (four projects). In both cases, a variety of issues were found, and it was notpossible to find technologies of separation that can be put into practical use immediately.
(1) Category A① Kurion Inc.: Water and hydrogen isotope exchange method (Combined Electrolysis Catalytic Exchange (CECE) method))- Based on the experimental data obtained from small tests and the building of a 1/10 scale (engineering scale) facility, theFig. 1: Kurion’s test
performance of separation in the actual plant was verified and the cost was estimated, etc.
equipment
3- In a design assuming a processing capacity of 400 m /day and an separation factor of 284 (H-3 concentration in the effluent: 4.4 Bq/cc), it
was estimated that the facility scale was 10,200 m2, the capital cost was $891,400,000, and the operating cost was$1,157,500,000 (per 800,000 m2 processing; the same applies hereinafter).
- The following issues were pointed out: there were variations, including instability and reproducibility in the test data;
further data must be acquired to assess the performance; the assumed performance level in the design of actual plant is
not obtained in the test plant; and the cost estimate related to the construction and demolition of the actual plant isthought to be underestimated.
② RosRAO: Combination of the water distillation method and the CECE method- Based on experimental data, etc. obtained from the construction of the actual-scale facility, the performance ofseparation in the actual plant was verified and the cost was estimated, etc.
Fig. 2: RosRAO’s test
- In a design assuming a processing capacity of 480 m3/day and an separation factor of 500, theequipment
construction cost was ¥38,500,000,000, and the operating cost was ¥40,500,000,000.
- The following issues were pointed out: there is a need to clarify evidence data for the performanceof separation, etc.; detailed investigation of mass balance including the amount of waste material on
the concentration side is required; testing is required for long-term operation and process stability;
and the cost estimation is considered to be underestimated.
③ Sasakura Engineering: Low-temperature vacuum distillation method with catalyst function- Based on experimental data obtained from the construction of an engineering-scale facility, theperformance of separation in the actual plant was verified and the cost was estimated etc.
- In a design assuming a processing capacity of 400 m3/day and an separation factor of 100, it wasFig. 3: Sasakura’s test equipment
estimated that the facility scale was 15,000 m2, the construction cost was ¥37,100,000,000, andthe operating cost was ¥21,200,000,000.
- While it is recognized that the graduation pre-processing test data was properly indicated, issues
including the following were pointed out: the test scale was small, an assessment in a one-stage
Crystallization
Solution ofCollection oflarger test plant is required to upgrade the scale to that of the actual plant, and cost estimationof HTO
crystalcrystalneeds to be carefully investigated in a larger-scale test.
Fig.
4:
Gas
hydridemethod(2) Category B① Sou Innovation: Two-stage gas hydride method- A laboratory-scale test was conducted for a tritium separation method in solid-liquid separation throughthe separation of gas hydride crystals that contain only the water with tritium in its structure.
- As a result of examination with laboratory-scale test equipment of a reaction vessel of 500 ml, with tritiumfrom the tritium-containing water in the first-stage separation processing reduction, a reduction in the tritium
concentration was achieved with a maximum separation factor of 341. On the other hand, of the tritium
separation processing in the two-stage gas hydride, there are still issues concerning the performance of
second-stage separation processing, such as that it would theoretically have been difficult to prove theperformance in this minimum-scale test equipment.
② Toshiba Corporation: Multiple-stage crystallization method
Fig. 5: Crystallization method- A laboratory-scale test was conducted for a method of incorporating tritium into ice based on the different
freezing points of water and the water with tritium and removing the ice with the increased tritium concentration.
- As a result of the test using the tritium liquid, there is a possibility of obtaining 1.02 of tritium separation ratio
per one-stage processing by setting the proper ice residence time. It was pointed out, however, that theperformance of separation was low, and it cannot be said that this is an advantageous approach compared toexisting methods, such as the water distillation and CECE methods.
③ Nextide: Multiple electrolytic cell-type electrolytic method- A laboratory-scale test was carried out for a method of separation based on the differences in the carriage ofhydrogen ions during electrolysis due to the different molecule sizes of the normal water that forms a cluster andthe water with tritium that is thought to exist alone.
Fig. 6: Electrolytic method- An electrolysis test was conducted by simple cell with the water with tritium, and the separation factor of 1.015
was obtained. It was pointed out, however, that the uncertainty of data was high and that it is not clear whether
tritium was selectively concentrated or depleted.
④ Hokkaido University: Electrolysis recombination method with fuel batteries- A laboratory-scale test was conducted for the separation method based on the differences arising in the
speed of the electrolysis reaction of ion-to-gas due to the mass number difference between light hydrogen andtritium.
- While it has been recognized as acquiring beneficial experimental data for the concentration of tritium in thefuel battery cell, issues including the following were pointed out: this experiment was carried out with heavy
water whose concentration is higher than the tritium concentration in the water to be processed, and theapplicability to isotope separation in a lower concentration region, such as the water with tritium in the
Fukushima Daiichi NPS, has not been confirmed.
Fig. 7: Electrolysis recombination
method with fuel batteries

Fig. A11-5-(1): Verification tests of tritium separation technologies

180Project objectives
In December 2013, as countermeasures for contaminated water at the Fukushima Daiichi NPS, the Japanese government compiled additional
measures for the problems of decommissioning and contaminated water at the Fukushima Daiichi NPS of Tokyo Electric Power Company
Holdings, Inc. Although it is stated in the document that the effects of additional measures are expected, in using technologies that requireconfirmation and verification, technologies with a high level of technical difficulty were verified, including seawater purification technology,
technology of capturing radioactive materials in soils, decontamination technology for contaminated water storage tanks, and unmanned boring
technology.
Project details and progress(1) Verification of seawater purification technology- Demonstration tests were conducted to verify the performance of purification technology for theradioactive cesium and radioactive strontium, etc. in the seawater of the port of the Fukushima
Daiichi NPS.
- Laboratory tests or tests with actual seawater were conducted for various adsorption methods
with inorganic and organic materials (Fig. 1), and a certain level of removal performance wasconfirmed. It can be thought that the outcome of this project may be used in the case of risingradioactive materials concentration in the seawater of the port.
Fig. 1: Adsorption performance test (case ofadsorption fiber)(2) Verification of the technology for capturing radioactive materials in soils- Demonstration tests were conducted to verify the performance of the technologies for
capturing radioactive materials (mainly radioactive strontium) in soils that is attributed to leakageof contaminated water at the Fukushima Daiichi NPS.
- Two types of permeable reactive walls were examined, and based on analysis, considering thesimulation tests, soil quality conditions, and groundwater flow conditions, there is a possibility
that the concentration of radioactive strontium in the groundwater can be reduced.
(3) Verification of the decontamination technology for contaminated water storage tanks- For the spent flange type tank to be generated after the replacement with welded type tanks,demonstration tests were conducted to verify the performance of decontamination methods
before demolition (Fig. 2).
- Tests were conducted in three approaches under the concept of neither generating nor
Fig. 2: Decontamination verification test
increasing liquid waste. This confirmed the remote operability and decontamination performance,
and a decontamination work plan was proposed based on the results.
(4) Verification of unmanned boring technology- For boring, which will be indispensable at the Fukushima Daiichi NPS in the future, demonstration tests were conducted to verify theperformance of unmanned boring under a high-dose environment from the viewpoint of reducing the exposure of workers.
- A remote boring system with a satellite communication network was developed and the feasibility of unmanned boring technology wasconfirmed through actual excavation work. Based on the test, issues during work within the power station premises were identified.
Implemented by
Mitsubishi Heavy Industries, IBC Advanced Technologies, Inc. Obayashi Corporation, the ATOX and AREVA NC and the AREVA ATOX D& D
Solution, JGC Corporation (verification of seawater purification technology)
The ATOX and AREVA NC and the AREVA ATOX D&D Solution, JGC Corporation (verification of the technology for capturing radioactive
materials in soils)
IH Corporation, Obayashi Corporation, and Kobe Steel (verification of the decontamination technology for contaminated water storage tanks)
Obayashi Corporation (verification of unmanned boring technology)

Fig. A11-5-(2): Verification of technologies for contaminated water treatment

181

Project objectives
To control the amount of groundwater flowing into the buildings at the Fukushima Daiichi NPS, a technology was established to surround thebuildings with large-scale frozen solid impermeable walls (frozen soil walls).
Project details and progress(1) Implementation period: from October 2013 to March 2021
Impermeable wall byfrozen soil method(2) Project operators: Tokyo Electric Power Company Holdings, Inc. and Kajima
Corporation(3) Objectives of frozen soil walls- To control increases of the contaminated water due to the inflow of groundwater into
the buildings by installing frozen soil walls around the buildings where contaminatedSea-sidewater has been stored.
impermeablewall- To control the water levels inside and outside the buildings with recharging (water
injection) to prevent from leaking the stagnant water outside the buildings after the installation of frozen soil walls.
- To limit the spread of leakage with the shielding effect of frozen soil walls during the period when the frozen soil walls are sound, even incases of leaking the stagnant water outside the building.
- For the research period of this project, even after the operation period, it should be possible to maintain the function of facilities throughmaintenance, replacement, and other responses.
(4) Overview of frozen soil walls
Land-side impermeable walls based on the frozen-soil method can shut out the flow of groundwater with the frozen construction methods,
maintaining a high impermeability. More specifically, coolant (brine, about -30°C) that was cooled in refrigerating machines and cooling towers
is pressurized and sent in the brine relocation tubes to be circulated through underground frozen tubes, which will freeze the ground in thesurrounding area. The total length is about 1,500 m, and the prepared amount of frozen soil is about 70,000 m 3, with 1,568 frozen tubes and359 temperature measurement tubes to be installed.
(5) Water injection system inside the frozen soil wall
As a supplementary system to prevent the contaminated water from the building from flowing underground, 33 holes of water injection wells
are arranged on the inner side of frozen soil walls, and water is injected into the ground as needed to keep the groundwater level around thebuildings higher (by 80 cm) than the level of water in the buildings. Observatory wells for groundwater level are installed at 82 locations.
(6) Long-term operational system
As to the maintenance of the function, it is said that stable operation will be possible for a long time with the replacement of brine (coolant),frozen tubes, and piping as needed. The frozen tubes have a triple-tube structure so that they can be replaced. Leakage detectors were
installed inside the protective tubes.
(7) Soil temperature measurement system
Soil temperatures at 359 locations are controlled to maintain the preparation of frozen soil. Instead of conventional platinum thermometers,
optical fiber thermometers are used to work out the complication of construction and systems.
First-Stage, Phase 1: March 31, 2016(8) Details
Start of freezing
June 2, 2014: Start of excavation for frozen tubes
April 30, 2015: Start of test freezing at 18 locations on mountain side
July 17, 2015: Completion of installing all 33 holes for recharge (water injection) wells
First-Stage, Phase 2: June 6, 2016
July 31, 2015: Approval for the execution of penetration work of buried materials for sea-side
Start of freezing
frozen tubes.
November 9, 2015: Completion of the positioning of piles of all 1927 frozen tubes andtemperature measurement tubes
Second-Stage: March 3, 2017
: Unfrozen areas
January 7, 2016: Completion of the installation of brine tubes
Start of freezingin First-Stage
February 1, 2016: Completion of the fulfillment of brine for all 1568 frozen tubes
March 31, 2016: First-Stage, Phase 1: Start of freezing at the sea side + part of the north side + the mountain side, where it will be frozen withpriority
June 6, 2016: First-Stage, Phase 2, Start of freezing on mountain side (except unfrozen areas at 7 locations)
December 3, 2016: Second-Stage, Start of freezing at 2 unfrozen areas on the mountain side
March 3, 2017: Second-Stage, Start of freezing at 4 unfrozen areas on the mountain side
May 22, 2017: Start of partial maintenance management operation
August 22, 2017: Third-Stage, Start of all freezing
November 13, 2017: Expansion of partial maintenance management operation(9) Status of preparation of frozen walls (Committee of Contaminated Water Measures, March 17, 2018)① The preparation of frozen walls was completed, except the areas in depth, judging from the status of temperature measurement tubes andthe water levels inside and outside.
② The impact of unfrozen areas in depth on the control of groundwater level within frozen soil walls seems minor. However, it is reasonable topromote freezing with a supplementary construction approach to take all possible measures for impermeability.
(10) Status of development of frozen soil walls (Committee of Contaminated Water Measures, March 17, 2018)① Although the groundwater levels around the buildings are controlled at higher levels than the water level from the viewpoint of preventing
water leakage in the building, they are being managed stably at low levels due to the water-impermeable effect of frozen soil walls. Combinedwith the effect of sub-drains, it is recognized that inflow into buildings is controlled.
② With the control of inflow into the building as well as a large reduction in the amount to be transferred to the building from the area of
T.P.2.5m, the amount of contaminated water to be generated was reduced to about one fourth.
③ In addition, the water-impermeable effect of frozen soil walls is found in the reduction of theBrine tubes
amount of sub-drain pumped up and the amount of groundwater pumped up in the area ofMountain side
T.P.2.5m.
Related project
Impermeable
Implementation period: From August 2013 to March 2015
wall line
Name of project: FY 2013 infrastructure development project of the decommissioning and safe
technology for nuclear reactor power (Feasibility study project for the water-impermeable
technology with frozen-soil method to control the inflow of groundwater)
Sea side Frozen tubes
Operator: Kajima Corporation
Objectives: - To verify the feasibility of impermeable walls with the frozen-soil method at the Fukushima Daiichi NPS- To obtain the necessary data for the construction execution plan and the construction execution management on the premise of long-term
use

Fig: A11-5-(3): Large-scale verification of impermeable walls (ice wall)

182

Project objectives
The conventional Advanced Liquid Processing System (ALPS, reducing the 62 types of nuclides from the contaminated water to less than orequal to the regulatory concentration), in which radioactive materials are removed from the contaminated water in the tank, generated a large
amount of slurry waste due to the pre-processing (iron coprecipitation processing and carbonate sedimentation processing) by the injection ofliquid reagent. However, development and demonstration was performed for high-performance equipment (high-performance ALPS) that
reduces more than 80% of the waste by adsorbing radioactive materials effectively to filters and adsorbents.
Project details and progress(1). Laboratory test to verify the performance of removal- With a test tube-scale laboratory test using the actual contaminated water, a filter technology and high-performance adsorbent with longer life
were carried out. The results of the laboratory test showed higher removal performance in most parts compared to other existing nuclide
removal equipment. In addition, the policy of adsorption tower configuration after the verification test was decided based on the test results.
- With the actual liquid test and a laboratory test, the characteristics of crevice corrosion resistance were assessed for the materials forpractical application and several candidate materials. In immersion tests with the actual liquid of about two months, no corrosion was found in
the two-phase stainless steel welded specimen, and in the assessment of the corrosion initiation life of the two-phase stainless steel weldedpart through an accelerated test, it was estimated that even with the water containing 6,000 ppm of chloride ion, if it is neutral, corrosion
resistance can be performed 76 years or longer.
(2) Verification of the removal performance with the small scale test equipment of the demonstration
plant.
- Using a 1/10 scale test equipment of the demonstration plant (Fig. 1), the removal performance of theadsorbent, the validity of the removal process, the amount of waste to be generated, and the properties ofwaste were verified. As a result, issues including the following were found: (1) the performance cannot besustained for long in the Cs and Sr adsorbent, and (2) the decontamination factor (DF*) of Cs and Sr
adsorbent (after the second tower) was low. It was decided that the factors causing these issues would beanalyzed using the demonstration plant because there was no impact of the scale.
*DF: Decontamination Factor)(3) Confirmation of removal performance with the improved ALPS developed, etc.
Fig. 1: Verification test equipment- A demonstration plant (500 m3/day: Fig.1) was designed, manufactured, and installed, and theperformance of the removal of nuclides was examined. It is a system consisting of a total of four filters intotal (of which three are chloride filters) and 20 adsorption towers, including seven to eight towers ofdeveloped Cs and Sr adsorbent materials.
- The following new issues were found in addition to the issues obtained from the verification test: (1) asabove, in the Cs and Sr adsorbent, the performance cannot be sustained for long; (2) the
B-line filters
Decontamination Factors of the second through fifth towers of the Cs and Sr adsorbent are low; and (3)the Decontamination Factors of the sixth through eighth towers of the Cs and Sr adsorbent are low. In
response to this, the analysis of factors through laboratory-scale basic testing, the planning of measures
A-line filters(4 filters)
and the verification of measures by a demonstration plant was conducted repeatedly.
- It was revealed that it was possible to sustain the performance in the adsorbent for longer by adjusting
the condition of feed liquid (pH) to Cs and SR adsorbent as an appropriate alkali condition. Therefore, forIssue (1), measures were taken by adding alkali after the carbonate ion removal with the addition of acidin the supply tank and before the liquid supplied to the adsorption towers of Cs and Sr adsorbent in theinitial stage of the system.
Adsorption tower- Actual contaminated water contains a chelate component, forming the complex with a part of Sr, which(20 towers)
was considered to be the cause of the low decontamination factor of Sr at the mid-stage adsorption towerin Issue (2). In response to this, countermeasures were taken with the dissociation of the complex by
Fig. 2: Demonstration plantmaking it acid at the mid-stage of the system.
- The processed water that had once been converted into acid by the countermeasure for Issue (2) is readjusted to alkali or neutral at the latterstage of the system. This was taken as the countermeasure for the declining decontamination factors in Issue (3).
- With the abovementioned measures, within the period of project, the Sr concentration at the exit of the system was less than or equal to theregulatory concentration, and a total of more than 34,000 tons of contaminated water was processed. In addition, it was confirmed based on
the life of adsorption tower that more than 80% (at that time) of the amount of waste to be generated could be reduced.
Implemented by
Tokyo Electric Power Company Holdings, Inc., Toshiba Corporation, Hitachi GE Nuclear Energy

Fig. A-11-5-(4): Development and verification of high-performance multi-nuclide removal system(High-performance ALPS)183

Attachment 12 Basic Direction of 6 Essential R&D Themes
December 12, 2017
Nuclear Damage Compensation and Decommissioning Facilitation Corporation (NDF)
The 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
center with 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 Scienceand Technology 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 researcheswith satisfactory understanding of the Needs.
Therefore, concerning the 6 Essential R&D Themes which are described in the interim report of the Task
Force on Research Collaboration (November 30, 2016), we compiled the Basic Direction of the 6 Essential R&D
Themes 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 fueldebris production. And since it is anticipated that the retrieval will require a long
period of time, the fuel debris will remain inside the reactors over 10 years. We
also need to remember that the retrieved debris must be stored safely. Choosing
the best possible methods of retrieve/transmission/storage of fuel debris requires
predictions of characteristic changes of fuel debris over time.
In relation to the accident of Chernobyl nuclear power plant, detection ofparticulates of the micron order which contain fuel components from around thefuel debris has been reported and the national report of the Ukrainian
government showed concerns about the increase risk of radioactive dust
emergence over time through self-decay. One of the possible reasons is that thefuel debris with high radioactivity exposed to rich humidity caused rapid progressof aging which quite slowly proceeds with normal uranium mineral in thegeological environment. It results in oxidation activated by radioactive dissolution
and generate hexavalent uranium compound. On the other hand, because thePrimary Containment Vessel (PCV) of the Fukushima Daiichi NPS (hereinafterreferred 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 the fueldebris. Since the radiation level is about one order higher in 1F than that of thefuel debris in the case of the Three Mile Island Unit 2 (TMI-2) accident (occurredin a short time after the operation started), it is under unexperienced condition. In
addition, it should be noted that it will take a longer period to complete theretrieval of the fuel debris from the time of accident than in the case of TMI-2.
Various factors are involved in the aging of such fuel debris in addition tothe 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 byalpha 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 alpha 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 confinement function, the criticality monitoring system, the PCV gas
control system, exposure evaluation, collecting, transferring and storing, andprocessing and disposal. In particular, as for the mid-and-long-term roadmap,

184

while the processing and disposal method of the fuel debris will be decided in
the third period (from 2022) after the retrieval of the fuel debris is started,
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 fromthe aging of the fuel 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 modelby clarifying 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 the matrix pattern of parameters (temperature, pH, etc.) to collectbasic data, and it needs to establish the prediction method of aging. In this case,
it is important to maintain the foundations for advancement of the actinoid
chemistry, which provides the basis for examination of the physical property offuel debris. In addition, heat analysis for 1F should be included in the basis ofinvestigation since the temperature distribution of the fuel debris has to beunderstood by calculating the heat distribution and the impact of regionaltemperature rise due to the decay heat should also be required to be examined.
Theme“Descriptions /Background issues”
on the interim report
Basic direction(2) To elucidate corrosion mechanisms under unusual/extreme
circumstances
It is required to collect data on corrosion under a variety of circumstances withconsideration of the circumstances specific to 1F decommissioning such as highradiation levels and unsteady routes of cooling water in order to prepare forpotential corrosion during decommissioning.
A boiling water reactor (BWR) consists of various metallic material. While
stainless steel, which is corrosion resistant, is used inside reactor where it ishigh temperature and high oxidizing environment, carbon steel, which is notcorrosion resistant, is used for the Primary Containment Vessel (PCV) that is theconfinement boundary and is assumed to be used in the normal atmosphere. On
the other hand, substantial knowledge has been collected so far about corrosionof 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 theoperation 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 have been generated through radiolysis ofwater. Currently, since nitrogen has been injected into PCV to prevent hydrogen
explosion and the oxygen density has been decreased, the densities of theoxygen and the hydrogen peroxide are considered to be decreased in the waterand 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 maintainsoundness of the structure and pipes that form the boundary for containment ofradioactive materials and it is also important to prepare countermeasures basedon the knowledge on corrosion in such environment.
Since the corrosion is essentially a kind of battery reaction, it is likely tohappen if the electric conductivity of water raises, pH falls, and the electric
potential raises under the condition of declining of surrounding water quality.
Although corrosion has been suppressed by nitrogen injection in general, it isstill under the condition of potential corrosion. Regional changes of theenvironmental condition may lead to an increase of corrosion speed at the part.
It is quite a special environment surrounded by various factors that promotes
corrosion such as a formation of liquid film of dew, a humid environment that
repeats wet and dry conditions near the water surface, an irregular flow ofcooling water, convection flow, or backwater due to irregular paths created bygaps between various shapes of fallen objects or deposited materials, a
progress of corrosion on the anode side between different kinds of metals
touching each other, a progress of acid-base reaction by microbes, or any otherpotential factors. In the future, further changes may occur in the internal
environment when the air that contains oxygen flows into the PCV when anegative pressure control is applied to retrieve the fuel debris. Since thecorrosion progresses over time during the decommissioning works under the185special environmental conditions, estimation of corrosion phenomenon andinvestigation of countermeasures are required based on the consideration on theenvironmental changes resulted from the progress of the decommissioning
process.
Therefore, it is necessary to collect basic data related to the progress ofcorrosion phenomena and systematically clarify and understand the phenomena
in order to provide satisfactory prediction and explanation of the risk changes
that follow corrosion of structures taking the permission and authorization of thesafety regulation into account with giving priority to the factors of higher needs
that may have critical impacts on the decommissioning works from the viewpointof the probability of occurrence such as the factors above and the impacts on thefunctionality (parts and severity), the scale and the timing. In this case, in order
to examine various approaches including not only the use of existing anticorrosive agents but also electric protection, it is important to accumulate andmaintain the knowledge related to the corrosion phenomena in addition to theelectronic state of materials in a special environment through principle analysis
and clarification of the corrosion progression mechanism.
Theme“Descriptions /Background issues”
on the interim report

Basic direction(3) Radiation measurement technologies adopting innovative
approaches
The radiation levels are still extremely high inside the 1F reactors/buildings dueto the accident and the existing measurement devices do not meet thecapability/functional requirements to provide accurate figures. It is vital todevelop 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 numberof radiation measurement products using various principles or materials
including ionization chambers, counting tubes, semiconductor detectors, andscintillation detectors are already offered. However, it is very important todevelop the human resources for measurement since it is necessary tounderstand the principles of the equipment in order to interpret the measurement
data and address possible troubles such as the case of the disorder (inversion ofdata value) between the all beta-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 performed by remote operation since the radiation level isextremely higher than the one in the work environment of existing nuclear
facilities. It is necessary to develop highly radiation resistant and small sizedmeasurement 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 isnecessary to develop measurement devices of neutron from the viewpoint ofcriticality prevention, real time measurement of alpha ray from the viewpoint ofidentification of the fuel debris, and gamma 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, high radiation resistance, energy
discrimination, space resolution (identification of radiation source position), easeof operation, and maintainability. As for the composition of the measurementtargets, development of the technology so-called "on-site analysis" is requiredsince 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
to promptly judge if the target is debris or not, and if the target is debris, thefunction to judge co-existence of reactor internals and neutron poison isrequired.
In addition, effective support tools for the decommissioning can be provided
by developing the technology to visualize the radiation field and thecontamination situation and clarify the profile of the fuel debris based on theinformation of the strength and the direction of the radioactive sources.
It is necessary to develop the generic technologies for innovative
measurement of radiation using new ideas and principles by considering on-site186

measurement requirements.
Theme“Descriptions /Background issues”
on the interim report

Basic direction

Theme“Descriptions /Background issues”
on the interim report
Basic direction(4) To clarify behavior of radioactive particulates generated duringdecommissioning (incl. alpha dust treatment)
As thermal cutting of the fuel debris via machine or Laser may produce a large
amount of alpha dust, it requires safety measures and dust containment
solutions. It is necessary to understand physical/chemical properties of alphadust, to predict the amount of dust to be produced for each method, and to
consider how to seal the dust according to the results in order to make sure theretrieval will be conducted in a safe and effective manner.
As the fuel debris retrieval work will start, cutting the fuel debris will create alot of radioactive dispersion particles (alpha dust) that contain alpha nuclide andthey will be dispersed within the boundary. When retrieving the fuel debris, since
the work will be performed in the confinement boundaries which are the brokenbuildings, it is important to understand the property of the alpha dust for studying
how to secure the confinement capability, designing the filtering system, andperforming the exposure evaluation of the surrounding environment and workers
including the time of accident.
With regard to the data about the scattering rate when alpha dusts are
generated, there is 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 not been collected just for the nuclear fuels
but collected for the objects polluted by the nuclear fuels, and the data has beencollected for the amount of the radioactive materials and their densities from theviewpoint of exposure control and it is not systematically organized.
On the other hand, the radioactive dispersion particles in 1F will be
generated directly from the fuel debris when retrieving the fuel debris and fromthe polluted objects in the decommissioning process. The types of radioactive
materials are alpha nuclides and beta gamma nuclides. While the alpha nuclideof which typical element is plutonium is important from the viewpoint of internal
exposure, the beta gamma nuclide such as cesium should be well considered aswell from the viewpoint of the total exposure evaluation.
In order to study on collection of radioactive dispersion particles, efficientfiltering and purification, criticality prevention, etc., it is necessary to grasp theamount of generated particles, distribution of particle diameters, radioactive
diameters, and the physical and chemical property of particles according to thedifferences of the target objects and the method of cutting. It is also important tounderstand the behavior of the generated particles in the gas phase, at the airliquid interface, and in the liquid phase during transportation or transition. For
example, it is important to understand the growth of particles throughcoagulation in the gas phase, evaluation of mist generation from the air-liquidinterface, 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 dispersion particles, it
is important to evaluate the impact of exposure to radioactive materials derived
from the fuel debris, especially the one of alpha nuclides. In this case, it isimportant to decide whether the conventional exposure evaluation methods canbe applied by judging if the chemical form and the particle diameter of theradioactive dispersion particles represented by plutonium is consistent with theones that have been used as the criteria of internal exposure evaluation forplutonium.
(5) To understand fundamentally mechanisms of radioactive contamination
To figure out the mechanism of radioactive contamination towards effective
decontamination; it is critical to implement effective approaches ofdecontamination based on the mechanism of the contamination to radiation
sources, and to decrease the volume of radioactive wastes as well.
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 so on, 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 and contaminated 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 othercontamination sources such as objects remained in pipes, and hidden sidebehind the pipes located at high inaccessible positions. In the future, when187

considering each step in the long decommissioning process, it is expected thatmany situations that require decontamination will arise, therefore, effective andefficient decontamination is considered highly necessary. With regard todecontamination, not only reduction of radiation but also reduction of wastes
should be taken into consideration.
For decontamination, while engineering approaches are required, including
physical methods 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 thecontamination mechanism of target objects in order to perform suchdecontamination 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 areused in pipes, tanks and so on to confine radioactive materials. However, there
is almost no research on 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 is important 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, the contamination mechanism must be clarified inprinciple on the concrete exposed to the accident and the subsequentenvironment 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 med- and long-term
viewpoint, it is necessary to establish the evaluation method based on theunderstanding of the contamination mechanism including the changes over timein the contamination 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 byradioactive materials, few number of researches are found on the contamination
mechanism of the metal of the piping and equipment in the environment of the
1F. While it is considered necessary to clarify the contamination mechanism
inside PCV or RPV exposed to the high 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 thedecontamination mechanism of resins and paints of cables, it is considered thatit 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 adsorption,
dispersion, moving along with groundwater flow in shallow underground in order
to conduct environmental fate studies to ensure they will not affect theenvironment.
In order to properly evaluate and reduce the risk of future environmental
impact caused by radioactive materials in the Fukushima Daiichi NPS site, it isnecessary to provide proper evaluation and estimation of the environmental fateof radioactive materials around the site via the shallow underground water andthe 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 waterleaked 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) the radioactive materials that are contained in thecontaminated water that will be generated as the result of retrieval of the fueldebris or decommissioning and dismantling of the buildings (including ion suchas actinide and suspended solids) that can be the future source 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 thetransport behavior of the radioactive materials as the required basic knowledge.
Specifically, the targets include the existence form of the radioactive materials in188the underground water, the distribution in the soil, the advection and diffusion
behavior in the underground water, the existence form and the advection anddiffusion in the surface layer, the existence form and the molten and diffusion
behavior of the radioactive materials in the seawater in the port and on thebottom 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 suchas the property 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 asimilar environment.
In addition, in order to provide the accurate future estimation of theenvironmental fate, it is also necessary to develop the monitoring technology toidentify the accurate contamination condition and the analysis technology tosimulate the transportation behavior of the radioactive materials. As for themonitoring technology, the technology for long term and continuous remotemeasurement and the mapping and behavior identification technology using thebig data are expected. As for the simulation technology, the creation model thatcan be used to analyze the behavior (influence of unsaturated layer, kinetic
evidence, etc.) specific to shallow underground and the estimation 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, adsorbent for purification of contaminated materials, and thepermeable reaction wall, it is necessary to examine the factors that have priority
since they may have critical influence on the decommissioning works.
In order to provide reasonable environmental fate studies for the radioactive
materials, it is important to proceed with considerations on the risk ofenvironmental influence so the development of the evaluation method related to
the risk of environmental influence has to be taken into account from thisviewpoint.

189

Attachment 13 Record of collaborative activities with foreign organizations
Activity
IAEA Project
IAEA side event

DAROD

OECD/NEA Project
BSAF

PreADESWGAMA-LTMNPP

TCOFF

EGFWMD

Contents

・Side events were held along with the IAEA General

Conference and explanation of the latest progresssituation in the Fukushima Daiichi decommissioning inorder to gain an understanding of the Fukushima
Daiichi decommissioning.
・Knowledge and experience obtained from the efforts
on challenges of decommissioning and recovery ofdamaged nuclear power facilities (regulations,
technologies, systems, and strategies) are shared
among the relevant countries.
・Research organizations and governmentalorganizations from eleven countries joined to conduct
benchmark study using severe accident analysis
codes developed by these organizations 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 tothe 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
are shared.
・Knowledge available to understand the characteristics
of fuel debris is shared and the methods for safety
evaluation on fuel debris sampling and retrieval areorganized.
・Regulations and standards in various countries andoperators’ efforts are shared and summarized on how
to ensure safe and stable status in the nuclear power
plants in which fuel remains after severe accidents.
・ From the viewpoint of proceeding with basic,
fundamental research for the Fukushima Daiichi NPS,a thermodynamic database, which is suited for thematerial science analysis to understand the migration
behavior of molten fuel and FPs and the characteristics
of fuel debris, is improved and enlarged.
・Expansion of knowledge for waste management anddecommissioning at the Fukushima Daiichi NPS
・ Advice to Japan’s R&Ds regarding waste in the
Fukushima Daiichi NPS

Participating
Organization
METI
NDF
TEPCO

NDF

JAEA
IRID
TEPCO

JAEA
IRID
TEPCO
NRA
NDF
TEPCO
JAEA

NDF

Inviting International Expert

International
Special Advisors(NDF)
International
Expert Group(TEPCO)International Advisors(IRID)

・To integrate and utilize a wide range of wisdom. To doso, NDF invites the experts in strategy development,
R&D, program/project management and safety
regulation as the International Special Advisors from
UK, US, France, and Spain.
・To obtain advice to contribute to the safer and more
efficient decommissioning of the Fukushima Daiichi
and its R&D, TEPCO invites experts with international
expertise and experience from UK, US, France, andUkraine.
・To gain the evaluation on the implementation status of
the R & D design review being undertaken,

190

NDF

TEPCO

IRID

enhancement of information dissemination andcommunication, and acquire knowledge including
failure experience, IRID invites the prominent experts
from the UK,US, and Spain.
Activities by each organizations
Cooperation with・Improvement of the accident analysis codes (US)overseas
・Development of technologies for prevention of leakageorganizations at（US）the stage of IRID’s ・ Developmentoftechnologies
forremotepractical
decontamination（UK, US）
application
・Development of internal investigation technologies（UK,research
US, France, Russia）
・Internal fuel debris detection and characterization（US,France）
・Fuel debris, structural component retrieval technology
development（US）
・ Fuel debris containment, transportation, storage

IRID

technology development (US)・Sampling Technology development（UK）
・ Solid waste processing and disposal technology

development (France)
1st International
scientific andpractical workshop
Insider Project

TOAL DECOMATOMEXPO

The 3rd Fukushima/
Chernobyl/ Three
Mile Island
International
Symposium・In associated with constructing new shelter of Unit 4 ofChernobyl Nuclear Power Plant, participating in theworkshop and gathering information on further policies.
(May 2017)
・In order to optimize waste management, this project aims
at establishing common method applicable to the specificcases by evaluating cost effective analysis andmeasurement methods for nuclear R&D facility anddamaged reactor, etc.
・Participating in the kick-off meeting of this project, andgathering the information.
・ Participating in meetings of nuclear and non-nuclear

companies related to decommissioning, held mainly by
British companies, and gathering information on effortsof each company. (April, 2018)
・ Participating in a forum where nuclear powercompanies related mainly to Russia and Europe, held
mainly by ROSATOM, and exchanging views with local
stakeholders. (May, 2018)
・Based on the lessons and cases from the accidents ofThree Mile Island and Chernobyl Nuclear Power Plant,
studies have been made for controlling the accident of the
Fukushima Daiichi NPS.

191

NDF

NDF

METI
NDF

METI
JAEA
TEPCO
NDF
JAEA
TEPCO