Subsidy Project of Decommissioning and Contaminated Water Management in the
FY 2018 Supplementary Budgets

Development of Technology for Further Increasing the Scale ofRetrieval of Fuel Debris and Internal Structures*
Research Report for FY2019
August, 2020
International Research Institute of Technology (IRID)*This project was named as Development of Technology for Retravel of Fuel Debris and Internal Structure, when the project started. According to
the development plan of decommissioning research in FY 2020 disclosed at the 75 th Secretariat Team Meeting for Countermeasures for
Decommissioning and Contaminated Water Treatment, the project name was changed to Development of Technology for Further Increasing the
Scale of Retravel of Fuel Debris and Internal Structures.

All rights reserved. International Research Institute for Nuclear Decommissioning (IRID)©International Research Institute for Nuclear Decommissioning

Contents

No.1

１． Purpose and goal of Development of Technology for Further Increasing the
Scale of Retrieval of Fuel Debris and Internal Structures
２． Achievements of projects implemented in previous fiscal year (FY2017-2018)
３． Project overviews
４．Implementation schedule
５． Project organization
６． Details of subsidy projects
７． Implementation items of this project1) Development of fuel debris retrieval methods
2) Development of technology for handling of fuel debris
3) Technological development for ensuring the safety during fuel debris retrieval
８． Summary
９． Specific goals for achieving the purposed of the project©International Research Institute for Nuclear Decommissioning1. Purposes and goals of Development of Technology for Further Increasing
No.2
the scale of Fuel Debris and Internal Structures【Purposes of Development of Technology for Further Increasing the scale of Retrieval of Fuel
Debris and Internal Structures】
In Tokyo Electric Power Company (TEPCO) Holdings Inc. Fukushima Daiichi Nuclear Power Stations (NPS), it isassumed that nuclear fuel has melted with reactor internals and existed as fuel debris at the reactor pressure vessel(RPV) and primary containment vessel (PCV).
Fuel debris in the RPV and PCV has been currently in a sub-critical state. However, as the reactor building (R/B),
RPV and PCV were damaged by the accident, the plant has been in unstable conditions which are different from theoriginal design. Therefore, fuel debris should be retrieved to maintain a sub-critical state so that radioactive materials
will not spread, aiming to ensure the stable conditions.
Against the above background, this project started, aiming at accomplishing a large scale of fuel debris retrieval incooperation with TEPCO Holdings for engineering and project management based on the Mid-and-Long Term
Roadmap toward the Decommissioning of TEPCO’s Fukushima Daiichi NPS (herein after Mid-and-Long Term
Roadmap). The project results will be utilized for engineering conducted by TEPCO.
This project aims at smooth responses to decommissioning and contaminated water management for the
Fukushima Daiichi NPS, as well as promotion of technology improvement in Japan, which contribute to technologicaldevelopment for the decommissioning of Fukushima Daiichi NPS and contaminated water management, based on the
Mid-and-Long Term Roadmap and the development plan of decommissioning study in FY2019 (the 63rd Secretariat
Team Meeting for Countermeasures for Decommissioning and Contaminated Water Treatment).
Specifically, the project development is planned for removal technology for interference objects to establish anaccess route in the R/B and PCV, fuel debris dust collecting technology and various element technologies to ensure
the safety system during retrieving fuel debris and the reactor internals.

【 Project goal】

The project sets a goal to accomplish a large scale of fuel debris retrieval and conducts study according to
the Mid-and-Long Term Roadmap.

【 Duration of project】 April 2019 – March 2021 (two years)©International Research Institute for Nuclear Decommissioning２． Project achievements in the previous fiscal year (FY 2017-2018)

No.3(1) Fundamental technology for retrieving fuel debris and internal structures① Development of fuel debris collecting systems
Assuming that fuel debris is collected from the bottom of primary containment vessel (PCV), technological information for powdery
fuel debris collecting system was organized. Additionally, the technological information on tools for processed fuel debris as well as
pebble-like fuel debris was organized, which is effective for collecting fuel debris.
② Development of fuel debris cutting and collecting systems
A confirmation test for process capability of the Molten Core Concrete Interaction (MCCI) by chisel process and ultrasonic core boring
was conducted to confirm processing characteristics and speed. In addition, components of the MCCI products and the test
production method for simulated fuel debris are studied, and data on the distribution of particle diameters was also obtained byanalysis of waste liquid that was processed in a test by using an actual prototyped test body.
③ Development of methods for preventing the spread of fuel debris
In order to prevent spreading fuel debris at the bottom of PCV into vent pipes and suppression chamber (S/C) during fuel debris
retrieval work, element tests were conducted and confirmed the feasibility of remote installation, assuming that a weir is installed at
PCV.
④ Development of element technology for work cells
By comparing confinement functions of the cell and technology for connecting PCV, these technologies were reviewed. In addition,
element tests for inflatable seals that seals when connecting the cell and PCV were conducted to confirm the feasibility of work steps
and to identify issues.
⑤ Development of removal technology for interference objects during fuel debris retrieval✓ Element tests for removing interference objects at the reactor bottom were conducted by the Partial Submersion Top Access
Method, and confirmed basic cutting methods such as CRD housing and collection methods for cut pieces.
✓ Element tests for removing interference objects of the biological shield wall (BSW) and inside/outside the pedestal wereconducted by the Partial Submersion Side Access Method, and offered the prospect of the feasibility for basic cutting andcollecting work in narrow places.
✓ Issues were identified by conducting combination tests of a robot arm and an access rail. The feasibility of basic actions related
to removal of fuel debris and interference objects was confirmed.
⑥ Development of remote maintenance technology for fuel debris retrieval device
A basic concept of remote maintenance, for example: equipment in the cell in case of the Partial Submersion Side Access Method,
was discussed to clarify maintenance classifications and maintenance equipment.
©International Research Institute for Nuclear Decommissioning２． Project achievements in the previous fiscal year (FY 2017-2018)(1) Fundamental technology for retrieving fuel debris and reactor internals

No.4

Technological development for removal ofinterference objects and the reactor bottom

Work cell sealing method (inflatable seal)
Confirmation of operation performance for

作業
ステップ

Ｇ．ペデスタル外干渉物撤去装置の準備

combination of robot arm and access rail
Ｈ．干渉物撤去用作業アーム
Ｉ．干渉物へのアクセス
の設置

撤去装置
プラットフォーム
装置
動作
状況
干渉物撤去用作業アーム

Opening holes of biological shielding wall
作業
時間
作業
ステップ

装置イ ンストール・設定：約2時間/ヶ 所※開口部への位置決め後に開始

干渉物Interference
object装置イ ンストール・設定：約6時間/ヶ 所
位置決め時間：約0.2時間/ヶ 所
Interference object※最大アクremovalセス長の干渉物把持method by using fuel
※開口部の位置決め後に開始

debris retrieval device as the side access method

Ｊ．干渉物の把持・切断・回収

Ｋ．干渉物を搬出ルート上へ回収
干渉物
回収

Ｌ．装置アンインストール
撤去装置
プラットフォーム

装置
動作
状況

Equipment for preventing the spreadof contamination into S/C

干渉物
作業
時間

・干渉物切断時間：約0.2時間/回
・切断片回収時間：約0.2時間/回

Effective processing methods from the aspect of

回収容器搬出時間：約0.3時間/回
装置引き上げ時間：約8時間/ヶ所
prevention of fuel debris
spreading©International Research Institute for Nuclear Decommissioning２． Project achievements in the previous fiscal year (FY 2017-2018)

No.5(2) Safety system technology during retrieving fuel debris and internal structures(i) Technology development for confinement functions

It was confirmed that Airflow analysis through the GOTHIC code could identify the shape of the primary containment vessel(PCV), geometry information such as locations of damaged openings, air volume of nitrogen injection and PCV exhaust, air flowcaused by air volume changes, distribution of hydrogen and oxygen concentrations, and qualitative characteristics of effects on
dust behaviors. As a result, an airflow analysis evaluation method was established to contribute the system design for nitrogen
injection and PCV exhaust system, aiming to secure confinement function during fuel debris retrieval work.

(ii) Technology development for collection and removal of dust generated from fuel debris (the air phase)

Existing technologies were surveyed to select technology that has advantages. The information on comparison study of equipment
type is necessary . Missing data in the necessary information, which is to be applied to the decommissioning of Fukushima Daiichi
Nuclear Power Station (NPS), was obtained from element testing.

(iii) Technology development for collection and removal of dust generated from fuel debris (the liquid phase:non-solubility)
Existing technologies were surveyed to select technology that has advantages. The information on comparison study of equipment
type is necessary. Missing data in the necessary information, which is to be applied to the decommissioning of Fukushima Daiichi
NPS, was obtained from element testing.

(iv) Technology development for collection and removal of dust generated from fuel debris (the liquid phase:solubility)
Existing technologies were surveyed to select technology that has advantages (adsorbent, etc.).Evaluation on removal capability
in the water quality conditions during fuel debris retrieval is difficult, therefore, element test for Am was conducted as Pu and Am
can be dominant in exposure evaluation. The data on removal capability which is used for the primary screening to apply
applicable technology for NPS was obtained.

(v) Study on the α nuclide monitoring system associated with fuel debris retrieval➢➢

The gas system can be monitored by using existing α monitoring technology.
A continuous monitoring for the liquid system might be difficult by using existing α monitoring technology. A target level ofconcentration in the liquid is not defined from the aspect of workers and public exposure up to Level 3, therefore, manual
analysis can be possible if measurement time requirement is long.

(vi) Optimization study on ensuring safety for fuel debris retrieval methods and systems

Exposure evaluation results were lower than the judgement criteria, and thus, the environment control system during fuel debris
retrieval can be applicable.
©International Research Institute for Nuclear Decommissioning2. Project achievements in the previous fiscal year (FY 2017-2018)

No.6(3) Criticality Control Technology for Retrieval of Fuel Debris and Internal Structures(i) Establishment of criticality control methods① Study of control methods and procedures reflecting the study on retrieval methods
The consistency between daily retrieval target, schedule and criticality control requirements was confirmed.

② Establishment of methods for controlling criticality while increasing the scale of fuel debris retrieval instagesA control method for full-scale retrieval was determined based on the internal investigation results.
(ii) Implementation of criticality control technology
Various tests were conducted to confirm how MCCI products can be processed using chisel and by means ofultrasonic core boring, and the processing characteristics and speed were verified. Moreover, the methods for creating
prototypes of the components of MCCI products or simulated fuel debris were studied, and in addition, data pertaining
to the particle size distribution, etc. was collected by processing actual prototypes during the processing test andanalyzing the wastewater generated during the processing.
① Sub-criticality measurement and criticality approach monitoring technology
・ Specifications for the monitoring systems and detector unit that can be operated with a robot arm, were formulated.
・ The feasibility of sub-criticality measurement in a large system was verified by simulating the actual equipment.
② Re-criticality detection technology
・ Detection performance of criticality was verified by using an operation method as part of the negative pressure
control system.
・ The calibration accuracy was quantified by establishing detector calibration technology.
③ Criticality prevention technology
・ The concept of the method of spraying non-soluble neutron absorber was developed and the measurement
concept for confirming the effect after spraying.
・ It was confirmed that long-term irradiation of non-soluble absorber did not affect the integrity of canisters.
・ The impact of using soluble neutron absorber (environmental impact at the time of leakage, compatibility withconcrete) was evaluated.
©International Research Institute for Nuclear Decommissioning3. Project overviews

No.7

3.1 Cooperation with other projects
This project is cooperated with other projects described below and holds joint meetings if necessary.

Sharing of analysis needs/results

Development of technology for further
increasing the scale of retrieval of fueldebris and reactor internals➢ Development of fuel debris retrieval method➢ Development of handling technology for fueldebrisCharacterization•••

Upgrading of the comprehensive understanding ofconditions inside the reactor
Development of technology for fuel debris
analysis/characterization
・ Investigation needs /sharing ofinvestigation results

Internal
investigation

Construction ofwater circulation
systemFuel debris collection and storage system
Treatment of fuel debris and deposits
Sorting of fuel debris and radioactive waste➢ Technological development related toensuring the safety during fuel debris
retrieval••・Sharing of safety scenarios (during operation tostop water and retrieve fuel debris)・Coordination of work steps for stopping waterleakage

Confinement functions
Prevention of criticality and monitoring
management method

Development of technology foraccessing and connecting inside PCV
that has high applicability to the water
circulation system.
Coordination of storage andtransfer methods

Collection, transfer
and storage
technologies

Development of Technology for
Investigation inside RPVDevelopment of Technology for
Detailed Investigation inside PCV

・ Sharing of safety scenarios
・Coordination of work steps

Increasing the scale offuel debris retrieval instagesDevelopment of technology for increasing
the scale of fuel debris retrieval in stages

・Needs of analysis
・Sharing of results

Dust collection
Development of technology for dust
collection system of fuel debris

Development of technology for collection,transfer and storage of fuel debris
Coordination of collection
and transfer methods

Solid waste
R&D for treatment and disposal of solid waste©International Research Institute for Nuclear Decommissioning3. Project overviews

No.8

3.2 Basic policy for developing technologies for fuel debris retrieval
The main response policies for implementing the plans under this project are described below.
[ Basic policy]
Technology for establishing the access route, technology for collecting fuel debris and technology for
maintaining the confinement function, which will be required for carrying out fuel debris retrieval work in aneven large scale, shall be studied.
Moreover, from amongst the matters that had undergone design study as part of the Project for Developmentof Fundamental Technology, Project for Advancement of the System, or the Criticality Control Project during
the previous fiscal year, studies on matters in which issues were found shall be materialized.

[Results based on the basic policy]⚫

Method for establishing the access route for fuel debris retrieval, work procedures and feasibility of the device⚫

Collection method depending on the condition of fuel debris, method of storing fuel debris in the canister, and system feasibility⚫

Results of the study of adsorbent related data pertaining to the treatment of fuel debris and deposits, and that of the conceptual system of
the wastewater treatment facility⚫

Results of technology investigation required for sorting the fuel debris and radioactive waste⚫

Technology related to the confinement function and the prevention and monitoring of criticality for ensuring safety during fuel debris
retrieval work

(Techniques for predicting the effect of removing aerosol from dust, method for connecting PCV and access tunnel, feasibility of subcriticality measurement, etc.)

Specific development items and implementation policies are indicated in the following slides.
©International Research Institute for Nuclear Decommissioning3. Project overviews

No.9

3.2 Basic policy for developing technologies for fuel debris retrieval

Based on the policies mentioned on the previous slides, implementation policies for each development item involving public
participation projects are specified below.
Development items involving public
participation projects

Implementation policy (proposed)

Reference

1) Development of fuel debris retrieval method

Technology shall be developed for establishing the access route based on site information, which will be needed toretrieve fuel debris existing in RPV and PCV. Moreover, a system for supporting the operation of the remote
operation device, technology for preventing the spread of contamination outside the PCV and technology believed tobe needed for establishing the access route, considering the construction of a shielding structure for exposure
reduction in the reactor building (R/B), shall be developed.

No.20
onwards2) Development of fuel debris handlingtechnology(i) Technological development of the fueldebris collection and storage system

The collection method depending on the condition of fuel debris, and the method and system for storing fuel debris
in the canister shall be developed for efficient fuel debris retrieval, and element tests shall be implemented as
required to check feasibility.

(ii) Technological development pertaining
to the treatment of fuel debris anddeposits

Element tests shall be implemented related to adsorption technology for removing soluble nuclides that are believed
to be dissolved in circulating cooling water from the fuel debris, and the most suitable adsorbent shall be selected.
Moreover, treatment technology for separating solid materials from liquid and storing them, shall be developed
considering the storage efficiency, remote operation, maintenance, etc. and the conceptual system design of thewastewater treatment facility shall be created.

(iii) Investigation of technologies related tosorting fuel debris and radioactive
wasteTechnologies required for sorting the matter retrieved from inside the PCV, into fuel debris and radioactive waste,
shall be investigated.

3) Technological development related toensuring safety during the fuel debris
retrieval work(i) Development of element technology for
confinement functions(ⅱ) Development of element technology
for preventing and monitoring criticality

No.143
onwards

No.167
onwards

No.204
onwards

Research and development shall be carried out concerning the combination of aerodynamic analysis of the inside of
the PCV, which is needed for estimating the behavior of dust containing alpha nuclides inside the PCV, and thetechnology for analyzing aerosol behavior, and in addition, the effect of removing aerosol from the dust shall be
estimated, appropriate monitoring technology shall be selected and the monitoring location shall be studied.
Moreover, the structure of the connection parts, method, inspection, maintenance of the sealing material, etc. shallbe studied, as technological development for ensuring the confinement function in the parts connected with existing
structures such as the PCV, and the necessary element tests shall be implemented to check technical feasibility.

No.229onwards

The accuracy of sub-criticality measurement with a large and complex fuel distribution system shall be verified tocheck feasibility of sub-criticality measurement. Moreover, the site applicability of the neutron detector shall be
confirmed.
The usage methods of non-soluble neutron absorbers, such as the method of using the non-soluble neutron
absorber differently depending on the condition of fuel debris during fuel debris retrieval, conceptual design of theabsorber injection device, etc. shall be studied.

No.271
onwards©International Research Institute for Nuclear Decommissioning4. Project Implementation Schedule

No.10[Legend]
【凡例】
Plan
：計画

Plan (After：計画（見直し後）
amendments)* *Consideration of：前倒し着手を検討

Implementation schedule for the development of technology for further increasing the scale of
燃料デブリ・炉内構造物の取り出し規模の更なる拡大に向けた技術の開発
retrieval of fuel debris and reactor internals (1/7) 実施スケジュール（1/7）
大分類
Category

小分類
Sub-category

１.Development
燃料デブリ取り出し工法開発
of method for fueldebris retrieval
a. Conceptual study①
of interference① Development
干渉物撤去技術の開発
a. 概念検討
removal technology
b. Element test plan
上アクセスによる大型構造物の取り
b. 要素試験計画
a. a.
Methodof retrieving and出し方法、搬送方法
transporting
large structures byc. Preliminary tests
accessing from the top.
c. 概念検討に必要な予備試験
needed for theconceptual study

２０１９年度FY2019an early start

Implementation as：必要に応じて実施
required

２０２０年度FY2020

備考

Mar
Apr 5月Aug Sep
Oct 11月
Dec 1月Jan 2月Feb 3月Jun 7月Jul 8月Nov 12月
Mar 4月Apr May
Jul Aug
Oct 11月
Nov 12月
Dec 1月Jan 2月May 6月Feb 3月
4月
9月 10月5月 Jun
6月 7月
8月 Sep9月 10月Remarks(最新状況)(lateststatus)Conceptual
概念検討 study
要素試験計画Element
test plan

Preliminary tests
予備試験
要素試験準備（試験装置製作）
Element
test preparation (Test device production)d. 要素試験準備

d. Element test
e.preparation
要素試験

Element
要素試験tests

e. Element tests

Summaryまとめf. まとめf. Summary
b. Method of removing
b. 横アクセスによるペデスタル内外のinterferences from inside and
干渉物撤去方法
outside the pedestal by accessing
from the side.

a.
study /
a. Conceptual
概念検討/課題抽出
Issue identificationb. Test planb. 試験計画c.
Test preparation / Test
c. 試験準備/試験装置製作
device production
d.
d. Element
要素試験tests

Conceptual
study / Issue identification
概念検討/課題抽出

Conceptual
study / Issue identification
概念検討/課題抽出
Test
plan
試験計画

Test plan
試験計画（ペデスタル外作業）(workoutside the pedestal)

Test preparation /Test device production
試験準備/試験装置製作

Test preparation/ Test device
試験準備/試験装置製作
production (work outside
the pedestal)（ペデスタル外作業）
Element test
要素試験（ペデスタル外作業）(workoutside the pedestal)e. Summarye. まとめSummary
まとめ（ペデスタル外作業）(work outside the pedestal)
干渉物撤去以外の技術の開発② Development②
of technology
other than that for interference
removala. 視界不良かつ狭隘環境での遠隔操作支援手法の開発

a. Development of remote
operation support method inenvironments with low visibility andnarrow spaces

a.a. Conceptual
概念検討 study
b.b. Preparationand verification
マニピュレータ制御ソフト
of the manipulator control
準備・検証
softwareモックアップ準備・検証
c.c. Mock-up
preparation andverification

Test plan(work inside the pedestal)
試験計画（ペデスタル内作業）

Element test
要素試験（ペデスタル内作業）(work
inside the pedestal)

Test preparation / Test device
試験準備/試験装置製作
production(work inside the pedestal)（ペデスタル内作業）
Element tests
要素試験

Summary
まとめ（ペデスタル内作業）

Summary
まとめ(work inside the pedestal)

概念検討 study
Conceptual

Preparationand verification of the
マニピュレータ制御ソフト準備・検証
manipulator control software

まとめd.d. Summary

The plan was revised to reflect the status of progress of the study and the results of element tests.
*：検討進捗状況および要素試験結果を反映し、計画の見直しを実施した。

Mock-upモックアップ準備・検証
preparation and verification
まとめ
Summary©International Research Institute for Nuclear Decommissioning4. Project Implementation Schedule

No.11

Implementation schedule for the development of technology for further increasing the scale ofretrieval of fuel debris and reactor internals (2/7)実施スケジュール（2/7）
燃料デブリ・炉内構造物の取り出し規模の更なる拡大に向けた技術の開発
大分類
Category

小分類
Sub-category

２０２０年度FY2020２０１９年度FY2019Aug Sep
Oct 11月
Nov 12月
Dec 1月Jan Feb
Apr 5月May JunJul 8月Aug 9月Sep 10月
Oct 11月
Nov 12月
Feb 3月Mar
Jun 7月Jul 8月Apr May
Dec Jan
4月
5月 6月
9月 10月2月 Mar
3月 4月
6月 7月
1月 2月

備考
Remarks(lateststatus)(最新状況)of method for fuel１.Development
燃料デブリ取り出し工法開発

debris retrieval② 干渉物撤去以外の技術の開発
Development of technology otherthanb. thatfor interference removal
サプレッションチェンバ（Ｓ／Ｃ）への汚染拡大防止方法
b. Methodof preventing the spread ofcontamination into the suppression
chamber (S/C)

Conceptual
概念検討studya. 概念検討

a. Conceptual studyb. Test
試験計画
b.
plan

Test plan
試験計画

c. 試験準備/試験装置製作
c. Test preparation / Test
device production
d. Element
要素試験testsd.

a.
of thea. Organization
検討条件案の整理
proposals for study
conditions
b. Creation
アクセスルート構築案
b.
of the proposals
foraccess routeの作成
establishment

Additional
test preparation
追加試験準備Element
要素試験tests

Additional tests
追加試験

Summaryまとめe.
まとめe. Summaryc. Establishmentof access route forc. セル設置に関わるアクセスルート構
cell installationand conceptual
study
築およびＲ／Ｂ
等への影響の低減
of the方法の概念検討
method for reducing impact on
R/B, etc.

Test preparation /Testdevice production
試験準備/試験装置製作

追加試験まとめ
Summary of additional tests

Organization of the proposals
検討条件案の整理
for study conditions
Creation of the proposals foraccess route establishment
アクセスルート構築案の作成

Narrowing down the proposals

for access route establishment
アクセスルート構築案の絞り込みc.
アクセスルート構築案
c. Narrowing
down theproposals
for access routeの絞り込み
establishment

Study on materialization of the proposals
for access route establishment
アクセスルート構築案の具体化検討

d. Study
アクセスルート構築案
d.
on materialization ofthe
proposals for accessの具体化検討
route establishment

アクセスルート構築案
Studyon feasibility of the proposalsの成立性検討for access route establishmente.
アクセスルート構築案
e. Study
on feasibility of the
proposals
for access routeの成立性検討
establishment

Summaryまとめ

まとめf.f. Summaryd. ユニット缶状態でのＲ／Ｂからの移送
d. Methodof transporting the unit cana.
of requirement
方法R/B
a. Formulation
要求仕様策定、概念設計
from thespecifications, conceptual
designb. Element
要素試験計画、要素評価b.
test plan, element
evaluation
装置仕様 device
specifications

c. Element
要素試験準備test preparationd.
d. Element
要素試験test

e. Summaryまとめ

Formulation of requirement
specifications,conceptual design
要求仕様策定、概念設計

Element test plan, element
evaluation
device specifications
要素試験計画、要素評価装置仕様
Test preparation / Test
deviceproduction
試験準備/試験装置製作
Element
test
要素試験
まとめ
Summary©International Research Institute for Nuclear Decommissioning4. Project Implementation Schedule

No.12

Implementation schedule for the development of technology for further increasing the scale ofretrieval of fuel debris and reactor internals (3/7)実施スケジュール（3/7）
燃料デブリ・炉内構造物の取り出し規模の更なる拡大に向けた技術の開発
大分類
Category

小分類
Sub-category

2.
of fuel debris
２.Development
燃料デブリ取り扱い技術の開発
handling technologya.
investigation(i)development of the(ⅰ)Technological
燃料デブリ回収・収納システムの技術
a. Technological
技術調査/概念検討/ conceptual study
fuel debris
開発 collection and storage
systemsa. 粒状燃料デブリの吸引回収システム b.b. Test
試験計画/試験装置製作
plan / test device
a. Development of suction collection
productionの開発
systemfor particulate fuel debrisc. Element
要素試験
c.
tests

２０１９年度FY2019２０２０年度FY2020

Apr May
Aug Sep
Oct 11月
Nov 12月
Dec 1月Jan Feb
Apr 5月May JunJul 8月Aug 9月Sep 10月
Oct 11月
Nov 12月
Dec JanFeb 3月Mar
Jun 7月Jul 8月4月
5月 6月
9月 10月2月 Mar
3月 4月
6月 7月
1月 2月

Technological investigation/ conceptual study
技術調査/概念検討
Test
plan / test device production
試験計画/試験装置製作
要素試験
Element
tests
まとめ
Summaryd.
d. Summaryまとめ

b.b.Developmentof technology for
燃料デブリ・廃棄物の搬出プロセス
transferring
fuel debris and waste
技術の開発

Conceptual
概念検討studya. Conceptual
概念検討 studya.
要素試験計画策定Formulationof element test planb. Formulation
要素試験計画策定b.
of element test
plan

Preliminary
tests
予備試験

c.
tests neededc. Preliminary
概念検討に必要なforthe conceptual study
予備試験

Element test preparation(Test device production)
要素試験準備（試験装置製作）

d. Element
要素試験準備d.
test preparation

Elementtests
要素試験

e. Element
要素試験
e.
tests

Summaryまとめf.f. Summaryまとめ(ii) Technologicaldevelopment of（ⅱ）
燃料デブリ・堆積物の処理に関わる技
treatment
術開発 for fuel debris and deposit①
technology for soluble① Removal
循環冷却水中の溶解性核種の除去技
nuclides in circulating cooling water術
a.a.Conceptual
system design of
溶解性核種除去設備の概念システ
solublenuclide removal facility
ム設計

備考
Remarks(lateststatus)(最新状況)

Conceptual
概念検討studya. Conceptual
概念検討 studya.
b. Element
要素試験計画b.
test plan

Element
test plan
要素試験計画Element
test preparation / Arranging test apparatus
要素試験準備/試験用品手配

c. Element
要素試験準備c.
test preparationd. Preliminary
要素試験計画に必要な予d.
tests neededforthe element test plan
備試験
e. Element
要素試験
e.
tests
f.f. Summary
まとめ

Am Am吸着試験
adsorbent test
U, Puなどの吸着試験
Adsorbent
tests of U, Pu, etc.

Summary
まとめ©International Research Institute for Nuclear Decommissioning4. Project Implementation Schedule

No.13

Implementation schedule for the development of technology for further increasing the scale ofretrieval of fuel debris and reactor internals (4/7)実施スケジュール（4/7）
燃料デブリ・炉内構造物の取り出し規模の更なる拡大に向けた技術の開発
大分類
Category
2.
of fuel debris
２.Development
燃料デブリ取り扱い技術の開発
handling technology（ⅱ）
燃料デブリ・堆積物の処理に関わる技(ii) Technologicaldevelopment oftreatment
術開発for fuel debris and deposit①
technology for soluble① Removal
循環冷却水中の溶解性核種の除去技
nuclides術 in circulating cooling water
b.b.Conceptual
design of the boric
ほう酸調整設備の概念設計

acid conditioning facility

小分類
Sub-categorya. Clarification
前提条件の整理a.
ofpreconditionsb.
of test items
b. Specification
試験項目の具体化(including
conceptual study(概念検討/調査含む)/ investigation)

FY2019２０１９年度

２０２０年度FY2020Dec 1月Aug Sep
Apr 5月May JunJul 8月Aug 9月Sep 10月
Oct 11月
Nov 12月
Dec JanFeb 3月Mar
Oct 11月
Jun 7月Jul 8月Nov 12月
Jan Feb
Apr May
4月
5月 6月
9月 10月2月 Mar
3月 4月
6月 7月
1月 2月

前提条件の整理
Clarificationof preconditions

Specification of test items (including

試験項目の具体化(概念検討/調査含む)
conceptual study / investigation)(Testplan if required)（要すれば試験計画）

Formulationof the element test plan
要素試験計画の立案

c. Formulation
要素試験計画の立案
c.
of the element
test plan

Element tests /organizationof results
要素試験/結果整理

d. Element
要素試験/結果整理
d.
tests /
organization of results(Element
tests if required)（要すれば要素試験）
Conceptual
概念設計 designe.
e. Conceptual
概念設計 design

Summaryまとめf.f. Summaryまとめ② Treatment
ＰＣＶ内から回収された堆積物等の処理②
technology for deposits
a. Element
test plan
技術 from inside PCV
要素試験計画
collected

a.a.Characterizationof collected
回収液や廃液等の性状把握
liquids, wastewater, etc.

b.
device production
b. Test
試験装置製作Element
test plan
要素試験計画
Test試験装置製作
device production

Parameter optimization

パラメータ最適化検討試験
study and testc. Parameter
パラメータ最適化検討試験
optimization
study and test

d.
water
d. Long-term
長期通水試験
conduction test

Long-term water
conduction
test
長期通水試験
Selection of 適用装置選定
device to be usedof technology toe. Selection
適用技術選定be usedConceptual
system design
概念システム設計

f.f. Conceptual
system design
概念システム設計

Summaryまとめg.
g. Summaryまとめb.
Development of technology forb. 回収液や廃液等からの固形分の分
separating
solid material from the
離・収納技術の開発
collected wastewater, etc. andstoring it

備考
Remarks(lateststatus)(最新状況)

Literature
survey文献調査
a. Literature
文献調査surveya.
b. Formulation
試験計画立案b.
of test plan

Formulation
試験計画立案of test plan
Element tests of sludge
separation technology
スラッジ分離技術要素試験

Element tests of the sludge storage
and handling technology
スラッジ収納・ハンドリング技術要素試験

c. Element
有意技術要素試験c.
tests of significant
technology
d. Conceptual概念システム設計d.
system design
・粒子回収システム
・Particle
collection system
・Solid materials separation
system・固形分分離システム
・Storage and handlingtechnology
・収納・ハンドリング技術
e.
e. Summary
まとめ

Conceptual systemdesign (Particle collection system)
概念システム設計(粒子回収システム)
Conceptual system design(Solid materials separation system)
概念システム設計(固形分分離システム)
Conceptual system design(Storage and handling technology)
概念システム設計(収納・ハンドリング技術)
Summary
まとめ

Summary
まとめ©International Research Institute for Nuclear Decommissioning4. Project Implementation Schedule

No.14

Implementation schedule for the development of technology for further increasing the scale ofretrieval of fuel debris and reactor internals (5/7)実施スケジュール（5/7）
燃料デブリ・炉内構造物の取り出し規模の更なる拡大に向けた技術の開発
大分類
Category

小分類
Sub-category

FY2019２０１９年度

２０２０年度FY2020

Apr May
Dec 1月Aug Sep
Apr 5月May JunJul 8月Aug 9月Sep 10月
Oct 11月
Nov 12月
Dec JanFeb 3月Mar
Jun 7月Jul 8月Oct 11月
Nov 12月
Jan Feb
4月
5月 6月
9月 10月2月 Mar
3月 4月
6月 7月
1月 2月

備考
Remarks(lateststatus)(最新状況)

２. Development
燃料デブリ取り扱い技術の開発
2.
of fuel debris

handling technology（ⅲ）
燃料デブリと放射性廃棄物の仕分けに(iii) Survey
on sorting technology for
関わる技術の調査
fuel debris
and radioactive wastes

a. Organizationof
前提条件と仕分シナリオ
preconditions
and scenario
案の整理
proposalsb.
of technologies
b. Survey
燃料デブリの仕分けに
required for sorting fuel
必要な技術の調査
debrisc. Evaluationof the feasibilityc.
燃料デブリの仕分
of fuel debris sorting
シナリオの実現性評価
scenarios

前提条件とシナリオ案の整理
Organizationof preconditions and scenario proposals

Technological
技術調査 survey
Feasibility
実現性評価
evaluation
Summaryまとめd.
d. Summary
まとめ

燃料デブリ取り出し作業時の安全確保
3. ３.
Developmentof technology forに関わる技術開発
ensuring
safety during fuel debris
retrieval work（ⅰ） 閉じ込め機能に関わる要素技術開発(i) Development of element
technology for confinement① functions
ＰＣＶ内でのダストの挙動予測技術① Technology for prediction of dust
behaviors in PCV

a. Improvement
現実的なダスト挙動評価手a.
in realistic
dust
behavior evaluation
法の高度化
methodsb. Aerodynamic
ＰＣＶ及びＲ／Ｂ全体におけb.
analysis oftheentire PCV and R/B
る気流解析
c. Evaluation
the behavior of
Ｒ／Ｂにおけるエアロゾルの
aerosolin R/B
挙動評価

d. Evaluation
入熱による気流への影響評
d.
of the effects ofheat
価 input on airflow
e. Summaryまとめe.

② Technology for ensuring② 接続部の閉じ込め機能確保技術
confinement
functions in connectionparts
a.
study /
a. Conceptual
概念検討/接続方式選定
Selection of connection
a. Methodof connection with the PCVa. アクセストンネルにおけるＰＣＶとの接methodfrom theaccess tunnel
b.
plan
続方法
b. Test試験計画
c. Testpreparation / Testc.
試験準備/試験装置製作
device production

d.
d. Element
要素試験tests

現実的なダスト挙動評価手法の高度化
Enhancementof realistic methods for dust behavior assessment

Aerodynamic
analysis of the entire PCV and R/B
ＰＣＶ及びＲ／Ｂ全体における気流解析
Evaluation
the behavior of aerosol in R/B
Ｒ／Ｂにおけるエアロゾルの挙動評価

Evaluation of
the effects of heat input on airflow
入熱による気流への影響評価
Summary
まとめConceptual study /概念検討/接続方式選定
Selection of connection method
Test
plan
試験計画
試験準備/試験装置製作
Test
preparation / Test device production

Element
tests
要素試験

Summaryまとめe. Summarye.
まとめ©International Research Institute for Nuclear Decommissioning4. Project Implementation Schedule

No.15

Implementation schedule for the development of technology for further increasing the scale ofretrieval of fuel debris and reactor internals (6/7)実施スケジュール（6/7）
燃料デブリ・炉内構造物の取り出し規模の更なる拡大に向けた技術の開発
大分類
Category
燃料デブリ取り出し作業時の安全確保
３. Development
3.
of technology forensuring
safety during fuel debrisに関わる技術開発
retrieval work(ii) Development of element technology

臨界防止・監視に関わる要素技術開発（ⅱ）
for preventing
and monitoring criticality

小分類
Sub-category

デブリを模擬した未臨界度
Sub-criticality
measurement
method
測定手法
with simulated debris

FY2019２０１９年度

Nov 12月
Aug 9月Dec JanMar
Nov 12月
Jul 8月Jan Feb
Apr 5月May JunSep 10月
Oct 11月
Apr MayOct 11月
Dec 1月Jun 7月Jul Aug
2月 3月
1月 Feb
6月 7月
3月 4月
2月 Mar
9月 10月8月 Sep
5月 6月
4月

Plan
計画

Preparation
準備① Technological development of① 臨界監視の管理方法の技術開発
criticality
monitoring control method
a. Verification of the feasibility of suba. 未臨界度測定の成立性確認
criticality
measurementb.
of applicability of the
中性子検出器の現場への適用性確
b. Verification
neutron detector to the field認

FY2020２０２０年度備考
Remarks(lateststatus)(最新状況)

Sub-criticality measurement
未臨界測定手法
methodverification test (KUCA)
確認試験(KUCA)
Evaluation
評価

Evaluation of alternative Service licensing procedures
役務許可手続代替中性子検出器の評価
neutrondetectors(RosRAO
CORONA)（RosRAO社CORONA)
Prototypespecifications
試作仕様
Deviceproduction
装置製作

Evaluationof alternative
代替中性子検出器の評価
neutrondetectors（IPL社半導体型)(IPL semi-conductors)

Preparation
準備

Devicedesign
装置設計

Transportation Detector
検出器
輸送
evaluation
評価試験 test（KUCA）
Evaluation
評価
Installationof test device
試験装置組み込み

Devicedesign and production
装置設計・製作
Study of testing system for
検出器評価試験体系検討
detectorevaluation

Detector
検出器
evaluation
評価試験
test
（KUCA）

デブリ上への設置装置
Deviceinstalled on thedebris

Evaluation
評価

Planningand design
計画・設計
Deviceproduction
装置製作
Environment
resistance test
耐環境試験

c.
of criticality approach
臨界近接管理手順の検討
c. Study
monitoring procedures

Criticality
approach
臨界近接管理手順
monitoring control
procedures

Summary
まとめ
測定手法
Establishmentofの成立 methodmeasurement

Selectionof options
検出器候補
forの選定the detector©International Research Institute for Nuclear Decommissioning4. Project Implementation Schedule

No.16

Implementation schedule for the development of technology for further increasing the scale of
燃料デブリ・炉内構造物の取り出し規模の更なる拡大に向けた技術の開発
retrieval of fuel debris and reactor internals (7/7)実施スケジュール（7/7）
大分類
Category
３. Development
燃料デブリ取り出し作業時の安全確保
3.
of technology forに関わる技術開発
ensuring
safety during fuel debris

retrieval work（ⅱ）
臨界防止・監視に関わる要素技術開発(ii) Developmentof element technology

for preventing and monitoring criticality② 臨界防止技術の開発② Development of technology forpreventing
criticality
a. 非溶解性中性子吸収材の投入装置、運用方法の検討
a. Study of the device and operating
method for injecting non-soluble
neutron absorber

小分類
Sub-category

２０２０年度FY2020Jun 7月Nov 12月
Nov 12月
Dec JanMar
Aug 9月Jul Aug
May 6月Jan Feb
Sep 10月
Oct 11月
Apr 5月Oct 11月
Dec 1月May JunJul 8月Apr 5月
4月
8月 Sep9月 10月2月 Mar
3月 4月
6月 7月
1月 Feb
2月 3月

備考
Remarks(lateststatus)(最新状況)

Scenario
clarification
シナリオ整理Clarificationof conditions
非溶解性中性子吸収材のStudyof the method of
条件整理
Criticality
臨界解析analysis
using
the non-soluble
使い分け方法の検討
まとめ
Summary
neutron absorberdifferently depending on
the case

Design,production and
投入装置の設計・製作・要
element
素試験 tests of theinjection device

Planning
計画

Preparation
準備
Underwater
injection
test
水中投入試験

デブリ加工による影響確Verificationof effects ofdebris認 processingb. Studyof corrosion effect consideringb. 放射線影響下での防錆剤の使用を
the use考慮した腐食影響の検討
of an antirust agent underradiation

FY2019２０１９年度Studyof corrosion effect
防錆剤を考慮した放射線
underradiation
影響下の腐食影響の検討
considering the use ofan antirust agent

Planning
計画

Evaluation
評価

Preparation
準備
Absorberprocessing test
吸収材加工試験
評価
Evaluation

Planning
計画

Preparation
準備

Absorber
irradiation test
吸収材照射試験
Evaluation
評価

主要なマイルストン
Major
milestones

Interimreport中間報告（10/15）

中間報告（4/22）
InterimreportInterimreport
中間報告

最終報告
Final
report©International Research Institute for Nuclear Decommissioning５．Project organization
Tokyo Electric Power Company
Holdings, Inc.
➢ Various coordination for site application(coordination to reflect the conditions ofapproval and authorization, anddecommissioning site)

No.17
International Research Institute for Nuclear
Decommissioning (IRID)➢ Coordination of overall planning and technology management➢ Coordination of technology administration including
technology development progress management

Project teams to cooperate fortechnological development

Hitachi-GE Nuclear Energy, Ltd.

Toshiba Energy Systems &Solutions Corporation

Mitsubishi Heavy Industries, Ltd.

Fuel Debris Characterization【Elemental test, technical development】(1) Development of fuel debris retrieval method【Elemental test, technical development】(1) Development of fuel debris retrieval method【Elemental test, technical development】(1) Development of fuel debris retrieval method

Development of Technology forCollection, Transfer and Storage ofFuel Debris① Development of technology for interference
object removal・Large-scale structure removal and transfer
methods by accessing from the top
・Interference object removal method forinside/outside pedestal by accessing from side
② Other technological development
・ Prevention method for spreading contamination
into S/C(2) Development of fuel debris handlingtechnology(ⅱ) Technological development for treatment of fueldebris and deposits① Removal technology for soluble nuclides incirculated water② Treatment technology for deposits collected from
PCV(ⅲ) Investigation of sorting technology for fuel debris
and radioactive waste(3) Technological development for ensuringsafety during fuel debris retrieval(ⅰ) Development of element technology for
confinement functions① Prediction technology for dust behaviors in PCV② Technology for ensuring confinement functions ofconnecting parts(ⅱ) Development of element technology for criticality
control and monitoring① Technological development of criticality
monitoring method② Development of criticality control technology① Development of technology for interference object
removal・Large-scale structure removal and transfer
methods by accessing from the top② Other technological development
・Transfer method for unit cans in R/B(2) Development of fuel debris handlingtechnology(ⅰ) Technological development for fuel debris
collection and storage systems
・Development of technology for processing fueldebris and waste transport(ⅱ) Development of treatment technology for fueldebris and deposits① Removal technology for soluble nuclides incirculated water② Treatment technology for deposits collected from
PCV(ⅲ) Investigation of sorting technology for fuel debris
and radioactive waste(3) Technological development for ensuringsafety during fuel debris retrieval(ⅰ) Development of element technology for
confinement functions① Prediction technology for dust behaviors in PCV(ⅱ) Development of element technology for criticality
control and monitoring① Technological development of criticality
monitoring method② Development of criticality control technology② Other technological development
・Development of remote operated support methodin poor visual and narrow environments
・Establishment of access route for cell installation
Conceptual study of impact reduction method for R/B(2) Development of fuel debris handlingtechnology(ⅰ) Technological development for fuel debris
collection and storage systems
・Development of particulate fuel debris suction andcollecting system(ⅱ) Development of treatment technology for fueldebris and deposits① Removal technology for soluble nuclides incirculated water② Treatment technology for deposits collectedfrom PCV(ⅲ) Investigation of sorting technology for fuel debris
and radioactive waste(3) Technological development for ensuring safety
during fuel debris retrieval（ⅱ）Development of element technology for criticality
control and monitoring① Technological development of criticality
monitoring method

R&D on Solid Waste Treatment
and Disposal

Development of Technology for
Investigation Inside the PCVDevelopment of Technology for
Investigation Inside RPVDevelopment of technology forconstruction of water circulation
systemDevelopment of technology fordust collection system of fueldebris©International Research Institute for Nuclear Decommissioning5. Project Organization
Tokyo Electric Power Company Holdings, Inc.
➢ Various coordination for site application(Coordination to reflect the conditions ofapproval and authorization, and the conditionsof the decommissioning site, etc.)

Hitachi-GE Nuclear Energy, Ltd.
TOKO Corporation (Chugai Technos Corporation)
・Development of technology for removing interferences fromwithin and outside the pedestal
・Development of technology for preventing the spread ofcontamination to the S/C

Shimizu Corporation
・Development of technology for preventing
the spread of contamination to the S/CMisuzu Seiko Co., Ltd.
・Characterization of collected wastewater

Hitachi Power Solutions Co., Ltd.
・Development of technologies forseparating solid material from collectedliquids and wastewater, and storing it・Study of corrosion effect under radiationconsidering the use of an anti-rust agent

Mitsubishi Hitachi Power Systems
・Method of connecting to the PCV fromthe access tunnel
Kyoto University
・Sub-criticality measurement test

No.18
International Research Institute for Nuclear
Decommissioning (IRID)➢ Coordination of overall planning and technology management➢ Coordination of technology administration including
technology development progress management

Toshiba Energy Systems &Solutions Corporation
ＩＨＩ
・Large-scale structure removal andtransfer methods by accessing fromthe top・Development of technology fortransferring fuel debris and waste
CLEO, CEGELEC, Irie Koken Co.,
Ltd.
・Method of transporting the unit canfrom the R/B
PaR Systems, Tokyo Denshi Kogyo
・Development of technology fortransferring fuel debris and waste

NFD
・Conceptual system design ofsoluble nuclide removal facility

Mitsubishi Heavy Industries, Ltd.
RosRAO・Evaluation of alternative neutron
detector
Shibuya Diving Industry
・Designing, production and element
tests of the injection device
Daiichi Cutter Kogyo K.K.
・Checking the impact of fuel debris
processing
CETD
・Device to be installed on top of fueldebris

Kyoto University
・Sub-criticality measurement test

Watari Co., Ltd.
・Method of transporting the unit
can from the R/B
Kurita Water Industries
・Characterization of collectedliquids and wastewater

MHI-NS Engineering
・Development of remote operation support method
in environments with low visibility and narrow
spaces.
・Development of suction collection system forparticulate fuel debris

ORANO
・Conceptual study on the establishment ofaccess route for cell installation and themethod of reducing the impact on R/B, etc.
MHI Solution Technologies Co., Ltd., NDC,
Sasakura Engineering Co., Ltd.
・Conceptual design of boric acid conditioning facility
Kobelco Research Institute Inc.
・Characterization of collected liquids andwastewater
・Development of technologies for separating solidmaterial from collected liquids and wastewater, andstoring it

Chugai Technos Corporation
・Development of technologies forseparating solid material from collectedliquids and wastewater, and storing it
Kyoto University
・Sub-criticality measurement test

IPL
・Evaluation of alternative neutron
detector©International Research Institute for Nuclear Decommissioning6. Details of Subsidy Project

No.19[Purpose]
To implement design study and element tests in coordination with engineering or project management
carried out by the Tokyo Electric Power Company Holdings, Inc., and make preparations so thatdevelopment of devices for the actual operation and systems for ensuring safety can be initiated for further
increasing the scale of fuel debris retrieval.

[Main scope of the project】
1) Development of fuel debris retrieval method
Technology will be developed for establishing the access route based on site information, which will beneeded to retrieve fuel debris existing in RPV and PCV. In particular, the technology for removing
interferences, which can shorten the process of fuel debris retrieval, will be studied assuming actual
conditions to confirm feasibility. Moreover, a system for supporting the operation of the remote-operated
device, the method for carrying out maintenance of the equipment inside PCV, technology for preventing
the spread of contamination outside PCV and technology believed to be needed for establishing the accessroute, considering the construction of a shielding structure for exposure reduction in the reactor building(R/B), will be developed besides the technology for removing interferences.
2) Development of fuel debris handling technology
The collection & storage method and system will be developed depending on the condition of fuel debris,the adsorbent most suitable for removing soluble nuclides will be selected, and technology for treating
wastewater generated as a result of filter backwash, etc. will be developed. Moreover, technology required
for sorting fuel debris and radioactive waste, will be surveyed.
3) Technological development related to ensuring safety during the fuel debris retrieval work
Element technologies for confinement of radioactive substances, prevention of criticality, etc., which areessential to ensure the safety of the public and workers during fuel debris retrieval work, will be developed.
©International Research Institute for Nuclear Decommissioning7. Implementation Items of this Project

No.20

1) Development of fuel debris retrieval method① Development of interference removal technology
Technology will be developed for establishing the access route based on site information, which will beneeded to retrieve fuel debris existing in RPV and PCV. Based on the results of investigating the R/B
equipment or the equipment inside PCV, in particular for the interference removal technology that canshorten the process of fuel debris retrieval, conditions such as misalignments, deformation, damage, etc., or
specific conditions such as heat induced deformation at locations that have not been investigated such as
the RPV, etc. will be assumed, the technology, basic specifications of equipment and procedures for removal
will be studied and feasibility will be confirmed by implementing element tests as required. Further,equipment, etc. assumed to become an interference for each of the access methods are mentioned below.
・Top access to RPV: Well shield cover, PCV head, RPV head, reactor internals, etc.
・Side access to PCV: Devices, equipment and piping, etc. inside R/B (specifically around the penetration),outside the pedestal, and inside the pedestal
The following will be included as the main development and study items. Element tests will be conducted
as required to identify problems and organize them. Further, maintenance will be considered as well during
the study.
a. Large-scale structure removal and transfer methods by accessing from the top

・Conceptual study pertaining to the access route and the method of transporting the structures will
be conducted.
b.Method of removing interferences from within and outside the pedestal by accessing from theside
・Element tests related to common utilities will be implemented to check the feasibility of the method.
©International Research Institute for Nuclear Decommissioning7. Implementation Items of this Project[1) ① Development of interference removal technology]

No.21

a. Large-scale structure removal and transfer methods by accessing from the top⚫ Purpose of development➢ To study the work procedures or the device used for establishing the access route to the reactor core,
which is necessary for fuel debris retrieval, in an effort to improve throughput.
⚫ Issues that must be resolved➢ Specification of work procedures and device used for establishing the access route➢ Confirmation of feasibility of the planned work procedures and equipment➢ Shortening of throughput⚫ Approaches to development➢ Procedures for removing interferences (cutting off, storing in canisters, transporting) up to the reactor
core will be specified. (desk study)➢ Procedures and equipment (hoisting attachments, shielding, fall protection gear) for removing
interferences even if the damage is severe, will be studied assuming the damage inside the reactor in
Units 1 - 3.
➢ Issues will be identified during the desk study, test plans will be formulated for the reactor internals forwhich element tests need to be conducted and for the stages of removal (cutting, holding, lifting, etc.),
and element tests will be conducted.
➢ The feasibility and throughput of the planned methods will be evaluated and a comparative evaluation
with past methods will be conducted.
⚫ Expected outcome➢ Feasibility of work procedures and device used for establishing the access route➢ Shortening of throughput©International Research Institute for Nuclear Decommissioning

７．Implementation items of this project【 1) ① Development of interference removal technology】

No.22

a. Top-access method for removal and transport methods of large-scale structures◼ Study flow

Plan in FY2018: Cutting of structures at the site and transport them in canister after storage• Long-term retrieval time schedule: 31.3 years (route A*) and 41.1 years (route B*) (original target
schedule : 10 years)• Many times of cutting: 2000 times, 2100 times for canister (ex. Steam-water separator)• Increasing size of the cell. No enough space due to severe loading conditions of container (4500t)

Achievements and issues inprevious subsidy projects

Approach policy of issues in FY2019
Preconditions for study

•
•
•

•

Removal and transport of the whole structure and fuel debris as muchas possible

Prediction of work area air dose rates in each stage of removing interference objects
Spread of contamination and confinement requirements in work area
Deformation of the structures during the accident

Study of specific plan 1:
Removal and transport of
the whole structure

Study of specific plan 2:
Combination of separating
structures and decontamination

Study of issues toimplement plan 1 (including
tests)

Study of issues to implement
plan 2 (including test)

Partly common issues (sharing oftechnology), Ex. Suspending technology
for structure

Applicable plan selection for practical use*：Overviews of the route A and B are described in No. 286.

：Explained at abriefing meeting

Executed in a subsidy project

Investigation results of the site condition (damaged
conditions of the reactor internals)©International Research Institute for Nuclear Decommissioning７． Implementation items of this project【 １）① Development of interference removal technology 】

No.23

a. Removal and transport methods for large-scale structures by accessing from the top◼ Development schedule
Study items
Major milestones

FY 2019
４

5

6

7

8

9

10

11

Interim report▼

FY 2020
12

1

2

3

4

Interim report
▼

5

6

7

8

9

10

11

Interim report
▼

12

1

2

Final report▼

１.Conceptual study
２.Planning ofelement test
３.Preliminary test
required forconceptual study

3(Testing will be conducted if necessary)

４.Preparation ofelement test (testdevice production)
５.Element test
６.SummaryRemark©International Research Institute for Nuclear Decommissioning

７．Implementation items of this project【 1) ① Development of interference removal technology】

No.24

a. Removal and transport methods for large-scale structures by accessing from the top◼ Major results of study in previous years•••

The image of the access route (to remove interference objects) that are from the
top floor (operation floor) to the reactor core and from the reactor core to thepedestal was established and the procedures was considered.
The duration of retrieval (throughput) was estimated, assuming unit cuts, theshape and thickness of the cut, and procedures of a cutting method,.
Additionally, element tests for the reactor bottom that is assumed to be the mostcomplicated and deformed were conducted by using simulated structures toverify the feasibility of the removal procedures.

Gas control
system inside
the building

Gas control
system inside
the building
Waste carryout area

Crane 1 inside
the cell

Crane 2 inside the cell
Canister storage area

Side for carrying in Side for carrying
the deviceout the waste
Canisterhandling crane 1
CanisterCanisterhandling
crane 2

Manipulator
airlock
Canisterinspection area

Crane 1 forcarrying out

Air Waster
Sealed Waste
Wastercanister carry- Shielding lock carry-out
carry-out
container door door 2
door 3
Canisterout door
installation stand1Sealed canister
Dolly cart for
Dolly cart forshieldingcontainer
shieldingdolly cart
inside the cell
container
outside the cell

Water control system
inside the building

Moved bymeans of lifter

100% operation at all times

Opening of PCV head

Cutting for opening shroud head
Access to steam-water separator

Water control system
inside the building
100% operation at all times
There are 3 or more doors in order tocontrol the release of radioactive
substances from inside the reactor
building, when the door is open.
Fuel debris
storage facility

CRD.H
Crump

Removal of large-scale structure at
the reactor bottom

AWJ nozzle

A transport method to transfer the structure
after stored in the unit canister (UC) would bechallenging in the aspect of workability andwork hours.
©International Research Institute for Nuclear Decommissioning

７．Implementation items of this project【 1) ① Development of interference removal technology】

No.25

a. Removal and transport methods for large-scale structures by accessing from the top
Preparation (set of device)◼ Challenges and issues based on the previous year’s results•
The duration of retrieval was estimated 31.3 years (route A) and 41.1 years(route B) despite the target duration of 10 years. A shortening of time schedule
is required.
Throughput【Ex.】 A number of cutting steam-water separator is 2000 times and2100 times for the canister. Most structures need to be cut many times.
•
A cell size is increased. The strict loading conditions of the building andcontainer (approx. 4500t)
Loading of the building•

Work hours of cutting (h)Improvement of loading in the building
A work implementation plan in the work cell was reviewed
and improvement of loading in the building was promoted
by optimizing requirement functions for the cell.

0.03

Total (h)

0.53
105
1060

Positioning (h)0.5

Cutting (water jet, h)

0.03

Total (h)0.53

Number of cutting

2000

Other work hours (h)5355

Fall protection device of cut pieces (h)Remote maintenance is required because contaminated devices are allocatedclosely, therefore a bad accessibility affects workers.
Maintenance performance◼ A policy and goal of this subsidy project✓

0.5

Cutting (h)Work hours of 11-devided height (10 cuttings, h)There is no enough space to allocate a cell on the operation floor and no bufferwhen trouble occur in specific equipment. Risk of suspending the whole work(a potential risk of delay in schedule).
Maintenance performance and throughput

Improvement of throughput
A duration of work was evaluated in previous years:
Approx. 4 years for establishment of the access route and27.2 – 37.2 years for fuel debris (estimation is different
depending on plans). A target goal was set to complete
within 10 years: 1.5 years for establishment of the accessroute and 8.5 years for fuel debris retrieval.

55.65

Positioning (h)Number of cutting

•

✓

5

1

Sampling transport (h)0.05Storage of cut pieces in collection canister (h)0.5Removal of collection canister (h)0.5transport of empty canister and preparation of
acceptance (h)0.5Total (h)2.55

Number of work

2100

Total work hours

6470.65

Days

269.6104

Clean-up

5

Total work days

279.6104

Separator removal procedures and duration ofremoval (route A)【results of FY2018】
No.

Method1

Route A

Shield plug
Including shroud head 3.3yrs.
Total: 1 yr.

2Route B

Shield plug
1.5yrs.
Including shroud head 2.4yrs.
Total: 3.9yrs.

Work hours to remove interferences

Work hours to retrieve fuel debris

Reactor ore and reactor
0.8yrs.
bottomConfirmation of the feasibility of a cutting method that a larger-sized structure
is removed and cut into pieces in another building (*1)
*1: After confirming the feasibility, the study of maintenance performance will start .

Inside pedestal
Total:

RemarkTotal

14.6yrs. 31.3yrs. Except for shroud
and jet pump
12.6yrs.
27.2yrs.

Reactor ore and reactor
bottomInside pedestal

24.1yrs. 41.1yrs. Plugs are transport
by using waste
13.1yrs.
canister.
16.8 yrs. For UC
Total: 37.2yrs.
400φ

Summary of the throughput (top-access method) 【results of FY 2018】©International Research Institute for Nuclear Decommissioning

７．Implementation items of this project【 1) ① Development of interference removal technology】

No.26

a. Removal and transport methods for large-scale structures by accessing from the top◼ Preconditions for study( surface dose of the structure that will be removed)⚫ Interference objects to be removed, contamination prediction, the surface dose and conditions of storage canister•

On a basis of MAAP analysis results and actual dose data, contamination levels and dose conditions were
calculated which will be preconditions for removal of the structure.
Structures
Contamination
Dimensions/mass
Surface dose calculated
from contamination
Shield cover

RPV head

Steam-water
separator•

1.0E+08Bq/cm2
The inner well isassumed to be evenly
contaminated.

11828×3200×t618
mm, 55ton(center, upper stage)Calculated from MAAP
analysis
Outer surface:
3.08E+13Bq
Inner surface:
2.86E+15Bq

Φ6200×h2800mm
Approx. 90ton

Calculated from MAAP
analysis
2.24E+16Bq(separator)
4.77E+15Bq (shroud
head)Φ4800×h5080mm
Approx. 51 ton

3.8E+03mSv/h (surface)
3.4E+01mSv/h（10m）
7.6E+02μSv/y（800m）
Firstly, it will be removed.

3E+04mSv/h (inner face)
1E+04mSv/h (outer face)

Dose rates increase after removal.

6E+04mSv/h (upper face)
7E+04mSv/h (side face)
Highest radiation dose

Vent

A storage canister for waste (under consideration by the waste project team)• Maximum surface contamination levels when storing waste: 10Sv/h• When transferring in the premises, wastes are stored in the transportcanister: contamination of the surface 30mSv/h©International Research Institute for Nuclear Decommissioning

７．Implementation items of this project【 1) ① Development of interference removal technology】

No.27

a. Removal and transport methods for large-scale structures by accessing from the top◼ Preconditions for study (air dose rates in work area)⚫ Dose calculation of work area in each step of interference object removal.

• Air dose rates of work area (mainly operation floor and RPV) in each construction step were calculated based on contaminationof the structure.
• A radiation dose was over 1 Sv/h at the time of removing upper part of the shield cover (the site boundary dose rate is approx.
800μSv/h（@1000m), approx.90μSv/h（@1000m)), therefore it is necessary to install the shield before removing the upper shied.

1.00E+05
Removal of three-step shield cover(exposing of PCV head)

1.00E+04

Exposing of dryer
Exposing of RPV head(removal of heat insulator)

Removal of upper shield cover( maintaining of installed middle
At the time of removing shroud head
(exposing of upper grid plate)

1000 mSv/h
1.00E+03
An evaluation point is positioned at the center of
RPV on operation floor surface level.

1.00E+02

1.00E+01

1.00E+00

Evaluation conditions (contamination conditions of the structure)
RPV external (PCV, well); The structure and wall surface of the well are evenly contaminated at
level of 1.0E+08Bq/cm2.
RPV internal devices: set at contamination levels obtained from MAAP analysis©International Research Institute for Nuclear Decommissioning

７．Implementation items of this project【 1) ① Development of interference removal technology】

No.28

a. Removal and transport methods for large-scale structures by accessing from the top◼ Preconditions for study (contamination levels in work area)⚫ Spread of contamination and confinement requirements of work area (when removing a shield cover)•••

The well reactor was exposed after removing the lower shield cover. When it was exposed, concentration of air contamination inwork area was evaluated and requirements of confinement were studied.
By setting contamination levels in the well internal (1.0E+8Bq/cm2), concentration of air contamination was predicted by dust
scattering rates and volumes of inside the well.
In case that contamination remains in the well, contamination levels are estimated to be several Bq/cm3. This estimate suggeststhat it is significantly a higher level of contamination than work standards in Fukushima Daiichi. In fact, contamination would
decrease affected by re-adhesion, however, confinement functions to prevent gases in the well from being released to outside
are required, which was reflected into work area plan.

Items

Setting values

RemarkAdhered contaminationof the well internal

4.14E+14Bq (well internal)
7.48E+13Bq (outer surface of PCV
head)

Assuming that same amount of Cs137 and Ba-137m exist.

Spread rate of adheredcontamination

1.0E-10/s*Cited by reference

Volume in the well

9.04E+08cm3=Radioactivity amount of wall
surface in the well * dust scattering
rates per 1 sec. * 24 hrs. / spatial
volume in the well

1.E+02Concentration of
air contamination（Bq/cm3）

Standards of equipment for work in
Fukushima Daiichi

1.E+01

1.E+00
0

100

200

300
Time (h)1.E-01* (Reference): G.A.Sehmel et.al, PARTICLERESUSPENSION RATES, BNWL-SA-5124, 1974

Equipment

Acceptable levels ofcontamination
concentration

Half face mask

～2E-03Bq/cm3

Hood mask～2E-03400
500
Full face mask

～2E-02

Electric fan with mask

～2E-01

Supplied-air respirator

2E-03～©International Research Institute for Nuclear Decommissioning

７．Implementation items of this project【 1) ① Development of interference removal technology】

No.29

a. Removal and transport methods for large-scale structures by accessing from the top◼ Preconditions for study (estimation of the structure conditions)⚫ Prediction of the structure deformation behaviors under high-temperature conditions after the structural accident

1F-3 構造物温度変化
1F-3: Temperaturechanges of structures

1F-2構造物温度変化
1F-2: Temperaturechanges of structures
1400

1400

1000

RPV上部ヘッドRPV upper head
セパレータ
Separator
シュラウドヘッドShroud head
Reactor core shroud
炉心シュラウド

800
600

PCVヘッド内面
PCV head inner
ドライヤ
Dryer
スタンドパイプStand pipe
上部格子板
Upper grid plate

1200

Temperature
温度 (℃) (℃)

PCVヘッド内面
PCV head inner
Dryer
ドライヤ
Stand pipe
スタンドパイプ
Upper grid plate
上部格子板

1200

温度 (℃)••

Analysis items (1) Estimation of static stress (estimation of short-time strength)(2) Prediction of creep deformation (estimation of long-time (max.3 weeks) deformation)
Analysis for: shroud head, shroud and dryer
Analysis temperature: a wide set of temperatures on 650℃, 700℃，800℃，900℃ (max.) at a basis of the MAAP analysis.

Temperature (℃)•

RPV upper head
RPV上部ヘッド
Separator
セパレータ
Shroud head
シュラウドヘッド
Reactor core shroud
炉心シュラウド

1000
800
600

400

400

200

200
0

0
0

24

48

72

96

120

144

168

192

216

0

24

48

72

96

120

144

168

192

216

(hr)
Time事故後の時間
after accident(hr)

事故後の時間(hr)Time after
accident (hr)Shroud headUpper unit

Baffle plate

Upper gridplate
ShroudReactor core
support plate©International Research Institute for Nuclear Decommissioning

７．Implementation items of this project【 1) ① Development of interference removal technology】

No.30

a. Removal and transport methods for large-scale structures by accessing from the top◼ Preconditions for study (estimation of the structure conditions)⚫ Prediction of the structure deformation behaviors under high-temperature conditions after the structural accident,
results and summary
•
•
•
•

•

•

The structure deformation was estimated based on temperature changes of the reactor internals that were obtained from the
MAAP analysis.
To widely estimate the structural deformation, temperature setting was increased by up to approximately 200℃, including
reduction of components that would restrict the deformation.
A short-time static stress estimation revealed that part of a steam dryer reached a yield stress at 900℃, but the whole part did
not cause the plastic deformation.
An estimation of three-week creep deformations revealed that the upper end of the shroud head was slightly deformed, but thedisplacement was less then 1mm and there was not a significant change of the shape such as inclining and decaying. On theother hand, a deformation evaluation revealed that the center of dryer dented and the skirt part was deformed, and eventually
the dent displacement increased at 600mm. Effects of contacting peripheral equipment will be discussed in future.
Additionally, these deformation might
be caused by a simplified model andtherefore, not-simplified model canbe considered.
The most significant change is the
position of the upper end of eachstructure. So, measuring of theposition changes for the upper endof the structure could be effective toestimate the changes of the whole
structure. Furthermore, the structure
information will be reflected for the
study of removing internal structures(study on removal of the deformed
state).

Figure of deformation after 3 weeks(magnification x 1)

Figure of deformation after 3 weeks(magnification x 1)

700℃

Displacement：51.2

900℃

Displacement：639mm©International Research Institute for Nuclear Decommissioning

７．Implementation items of this project【 1) ① Development of interference removal technology】

No.31

a. Removal and transport methods for large-scale structures by accessing from the top◼ Concepts of a new top-access method

Policy direction in FY 2019: Confirmation of the feasibility of the method that the large-scale structure istransported and cut into pieces in the other building.

【Plan ①】 Removal and transport of the whole unit✓✓✓

The whole structure in integrated state is removed andtransported.
The reactor core is divided into several units. The reactor
bottom is cut and removed the RPV end head from RPV.
Materials that will be transported are shielded and air-tighted
depending on a container and an access route, or their
combination.
Removed structures are cut into pieces and stored in acontainer, which will be performed in the building far from theR/B.
container
Reactorinternals Transport
炉内構造物
搬送容器

✓
✓
✓
✓

The structure that will be transported should be as large as
possible. The combination with decontamination and shielding
is performed depending on the transport condition.
Decontamination and dividing work are performed at the SFP
and DSP.
The reactor well is utilized to ensure shielding and airtightness.
The confinement area is formed by lines connecting the reactor
well and decontamination/dividing area (SFP or DSP). The
position of the confinement should be as low as possible toreduce the mass of the shield.
4.Decontamination/in④輸送コンテナのspection除染・検査
of transport
container

Reactorbuilding
原子炉建屋

搬送容器
Transport container

C4M

Expanded building
増設建屋

Transport to a
新設建屋へ
new building（構内輸送）(onsite transport)
Buffer
緩衝材

輸送車両
Transport
vehicleSecondary boundary

C4L

Primary boundary

約30m

✓

【Plan ②】 Combination with divided large structure andcontamination

3.Divided
structures are③細断した大型構造物をstoredin transport
輸送用コンテナに収納
container.
1.Remove large-sized①大型構造物を移動
structures.

C4H

2.②専用水槽で大型構造物を除Decontaminate or cut
large-sized structures in a
染または切断分割
water tank.

5.Transport
container is⑤輸送コンテナを車両に載loaded on transport container
せて搬出（有人作業）(manned work).

©International Research Institute for Nuclear Decommissioning

７．Implementation items of this project【 1) ① Development of interference removal technology】

No.32

a. Removal and transport methods for large-scale structures by accessing from the top◼ Study of specific plan 1 (Removal and transport of the whole structure in integrated state): Approach to study
To study the method for removing the structure as a unit, there are significant differences of issues to be discussed.
1. The structure is removed from the reactor and then transported. 2. The structure is separated from the reactor.

⇒ Two different methods are discussed (discussions on these methods are promoted whileproviding feedbacks each other).
Reactorinternals
炉内構造物

Transport

搬送容器
container

Reactorbuilding
原子炉建屋

搬送容器
Transport container

1. Study on the removal method of
the whole structure

Approx.
30m
約30m

Expanded
増設建屋 building

Transport
to a
新設建屋へ
new
building（構内輸送）(onsitetransport)

Buffer
緩衝材

輸送車両
Transport
vehicle2. Study on the removal method in the reactor
Conceptual diagram of the removal method of the whole structure©International Research Institute for Nuclear Decommissioning

７．Implementation items of this project【 1) ① Development of interference removal technology】

No.33

a. Top-access method for retrieving large-scale structures and transport method◼ Study of specific plan 1 (Removal and transport of the whole structure in integrated state): Study on transport
method and removal of the whole structure in integrated state
Comparative evaluation on three clarified plans was performed.

(Plan A)
Transport canister method
R/B
原子炉建屋

原子炉建屋
R/B

deviceLifting machine
揚重機 Removal
取り出し装置Conceptual diagram

Additional Lifting揚重機
machine増設建屋
building

搬送Transport
container
容器

Connecting
aisle
連絡通路

Ensuring
気密確保
airtightness
が課題
Airtightness
気密ゲート
gate

接続
Connecting
sleeveスリーブ(Plan B)
Connecting aisle method
原子炉建屋
R/B
取り出し装置RemovaldeviceAdditional
増設建屋
building

オペフロ上
Heavy
weight on
operation
重量大floor

Transport
搬送
台車
carriage

接続
Connecting
sleeve スリーブ

Transfer
to
新設建屋
new
building
へ移送

Issues• Feasibility of airtightness gate in canisters
and on the well.
• Feasibility of storing the lifting machine on
or in canisters.
• Maintenance is difficult, specifically, thelifting machine for canister installation.
• Emergency recovery of removal device isdifficult.

(Plan C)
Transport canister +Connecting
aisle method
Lifting揚重機
machineRemovaldevice
取り出し装置Gantry
門型架構
structure
Connecting
連絡通路aisle

Transport
搬送容器container
増設建屋

Additional building

Airtightness gate
気密ゲート

Transport
搬送
台車
carriage

Double二重蓋 lidTransport
輸送専用容器
container• Mass reduction of the connecting aisle• Maintenance of the lifting machine on
the connecting aisle/replacement
methodConnecting
接続
sleeve
スリーブ

搬送Transport
台車
carriage

二重蓋 lid
Double
輸送専用容器Transport
container• The height of the gate type structure
increases because the canister andrack are required, so the height of theconnecting aisle also increases.
• Maintenance of the lifting machine isdifficult.

Significant issues of each method* were studied: 1. Feasibility of ensuring airtightness, 2. Acceptability of the mass reduction of theconnecting aisle.
•

Maintenance performance is also significant issue, however above two issues are more important to evaluate the feasibility of each plan. Therefore, these issues started to be studied as priority issues.
©International Research Institute for Nuclear Decommissioning

７．Implementation items of this project【 1) ① Development of interference removal technology】

No.34

a. Removal and transport methods for large-scale structures by accessing from the top◼ Study of specific plan 1 (Removal and transport of the whole structure in integrated state): Feasibility of ensuring
airtightness【Issues】（A） Double door method✓ Possibly of radiation contamination•••

The shielding lid is installed for both side of thecontainer and the reactor well (for connectionof the container).
After contacting both lids of the container andthe well sides, they can adhere well by rotating
actions. Then, they performed open-close
movements.
When withdrawing from the container, theshielding functions did not lose and radiation
shielding can be secured in the reactor.

Container✓✓

in connecting surface of the gate
Gate driving method
Interference with the well
Double door base
ダブルドア
ベース

汚染拡大
Protection
防止容器
container

原子炉ウェルReactorwell

Interference

汚染拡大防止with
容器と干渉

ダブルドア閉
Doubledoor close（B）Double gate method•••

The shielding gate is installed for both sides of thecontainer and the reactor well.
Both lids of the container and the well side areconnected and adhered, and performed open-close
movements.
Therefore, shielding functions are not lose whenwithdrawing from the container and radiation
shielding can be secured in the reactor.

Interference
原子炉with reactor
ウェル
well
と干渉

protection
container
ダブルドア開Doubledoor open【Issues】✓
Container✓✓✓

Double gate

Spreading contamination in the lower
gate of sliding part (water contamination)
Possibly of radiation contamination inconnecting surface of the gate.
Gate driving method
Leakage of contaminated materials when
opening the gate.

Both methods were unsuccessful. The result revealed that there is a significant issue in ensuring airtightness.
©International Research Institute for Nuclear Decommissioning

７．Implementation items of this project【 1) ① Development of interference removal technology】

No.35

a. Removal and transport methods for large-scale structures by accessing from the top◼ Study of specific plan 1 (Removal and transport of the whole structure in integrated state): Feasibility of shieldthickness in connecting aisle
To assess the feasibility of shield thickness of the connecting aisle, the mass reduction of the connecting aisle was discussed.
【Conditions of preliminary calculation】【Study policy】①②③

Dose rates in the site boundary during opening the reactor (not moving thereactor internals toward the operation floor) are equivalent to effects of theoperation floor at the moment.
Dose rates of the connecting aisle surface during opening the reactor (notmoving the reactor internals toward the operation floor) are less than thesurface dose rate.
In case that the reactor internals are moved to the operation floor (inside theconnecting aisle), dose rates of the site boundary are less than 35μSv/yr.*1 .
Skyshine

Skyshine

Connecting
aisle
連絡通路

Connecting
aisle
連絡通路

＝
①炉内からの線量1.Radiation from thereactorLess
than 35γSv/y
35μSv/y以下•

Distances from each Unit to the site boundary: 800m•

Connecting aisle: [Inside dimensions] 14,400 x 37,500 x
H22,000mm(tentative values), iron

③構造物からの線量
3.Radiation from thestructure(direct)(直接線)

取り出し作業による追加分
Additionaldose caused by retrieval

Skyshine

Shield plug

【Results of preliminary calculation】
Dose rates ofoperationfloor
surface(average value)[mSv/h]Dose rates of
inside thereactor after
RPV headopen [Sv/h]

Unit 1

125

Unit 2
Unit 3Specifications of connecting
aisle
Shieldthickness
[mm]

Mass
[ton]

30

190

3,844

120

55

220

4,443

105

55

220

4,443

According to a preliminary calculation of shield thickness, dose rates in thesite boundary is more than 35μSv.

⇒In case that only surrounding of the structure is provided
additional shields at maximum of 130mm (approx. 120ton), dose
rate would be much less than 35μSv.

＝
Current
現状 state•[Unit 1]30Sv/h, [Unit 2&3]：55Sv/h,
Dose rates in the site boundary: Less than current dose (as ofFebruary 2020)

Connecting aisle
連絡通路

2.Radiation from the②構造物からの線量
structure(skyshine)(スカイシャイン)Less than average surface radiation
doseオペフロ表面線量率の平均値以下
rate of operation floor

Connecting
aisle
連絡通路

•

[Unit 1]125mSv/h, [Unit 2]120mSv/h, [Unit 3]105mSv/h
Dose rates in the reactor after opening the RPV head＋

現状維持current status
Maintaining
Skyshine

Dose rates of the operation floor surface (average value):

Skyshine＋

取り出し時の線量
Dose during removal

Averagesurface
オペフロ
radiation dose
表面線量率rate of operation
の平均値
floor

Connecting
aisle
連絡通路

•

Installation of connecting aisle
連絡通路設置状態

Shield thickness could be possibly reduced.

©International Research Institute for Nuclear Decommissioning*1： This value is per 1 plant and calculated by the Professional Standard, 1mSv/yr., and dose rate of the site boundary in the implementation plan, 0.89 mSv/yr.

７．Implementation items of this project【 1) ① Development of interference removal technology】

No.36

a. Removal and transport methods for large-scale structures by accessing from the top◼ Study of specific plan 1 (Removal and transport of the whole structure in integrated state): Comparativeevaluation on transport method of removal of the whole structure in integrated state

As a result of study on the feasibility of ensuring airtightness and shield thickness of the connecting aisle, candidate plans were re-compared.
Bold and underlined letters: Significant issues

Methods

Advantage

Disadvantage

Issues⚫

Plan A(Transportcanister method)⚫
⚫

Contamination ranges are
small. (In case that airtightnessof the gate is possible.)
The mass is assumed to besmall.

⚫
⚫

Complicated removal andtransport.
Large number of equipment⚫
⚫
⚫

Feasibility of gate airtightness on thecontainer and well.
Difficult maintenance. Specifically, liftingmachine for installation of the container.
Lifting machine cannot be stored in oron the container.
Difficult emergency recovery of
removal device

Assessment

Not
applicable

Plan B(Connecting aisle
method)⚫ Less equipment (lessmaintenance)⚫ Simple removal and transport⚫ Layout of equipment required formaintenance can be installed far
from the place on the well.

⚫ Contamination ranges are large.
⚫ The mass of equipment is large.

⚫ Heavy loads for the reactor building

Acceptable

Plan C(Connecting aisle＋Transportcanister method)⚫ The canister with shielding
function enables a smaller
thickness of the connecting aisle,
therefore, the mass reduction
may be possible.

⚫ As airtightness of the containeron the operation floor cannot
be ensured, contamination
would spread.
⚫ Complicated removal andtransport⚫ Large number of equipment⚫ The height in dimensions increases
because the container and rack arerequired. Therefore, the connecting aisle
would be larger (concerns about seismic
performance).

Acceptable

As a result of the above comparison, Plan B was adopted as the most priority and the detailed study will start innext fiscal year.
©International Research Institute for Nuclear Decommissioning

７．Implementation items of this project【 1) ① Development of interference removal technology】
◼

No.37

a. Removal and transport methods for large-scale structures by accessing from the top

Study of specific plan 1 (Removal and transport of the whole structure ): Study of Study of method to remove from the reactor

The study of method for lifting the reactor internals from the reactor to the operation floor level was performed (the transport method is basedon the study results in the previous pages).
【Concept of method for removing from the reactor 】
Removal and suspending functions are distributed (emergency recovery in the state of maintaining fall prevention for target objects, andimplementation of device maintenance)⇒Major component consists of two devices: Removal device and suspending device.
✓ Removal device: Cutting of the reactor internals, and wire implementation for fall prevention of removed objects✓ Suspending device (suspension rings and reactor wall rings): Suspending function to prevent falling target objects (reactor internals✓ Prevention of spreading contamination from the lower part of each device (RPV inner seal)✓

揚重機
Lifting machine

Divided分割吊りリング
suspension
ring fixing
jig
固定治具
吊りリング
Fixing jig(divided（分割構造）
structure)

Plate spring (upper)

板バネ付きのRemovaldevicewith取り出し装置がplate spring
内側を通過
passesthrough

RPV
RPV
壁面
wall

Water
curtainPlate spring (lower)

Suspension ring
吊りリング

Circular arc
Reactor炉壁リング
wall ring

Plate 板バネ
spring(complete（全周）
periphery)取り出し
Removal装置
deviceObject to
取り出し
be
対象
removedReactor
炉壁
wall
ring
リング

取り出し
Objectto be
対象removedExample of the RPV inner seal **Development Results by the Fundamental Technology Project in FY 2016-2017©International Research Institute for Nuclear Decommissioning

７．Implementation items of this project【 1) ① Development of interference removal technology】

No.38

a. Removal and transport methods for large-scale structures by accessing from the top◼ Study of specific plan 1 (Removal and transport of the whole structure in integrated state): Study of method to
remove from the reactor
On a basis of concepts in the previous pages, the removal procedures were studied.
Step
1
手順1

Step
2
手順2

Step
3
手順3

Step
4
手順4

Suspension
ring installation
吊りリング設置

Reactorwall ring installation
炉壁リング搬入

Connection
with removal object
取り出し対象接続

Removaldevice installation
取り出し装置搬入

据付
Removal装置
deviceObject to
取り出し
be
対象
removed••••A吊りリングを分割構造で搬入
suspension ring that is
dividedinto the structures is
し、バルクヘッド上で組み立て
installed
and assembled on theる（搬出容器内径に対して吊bulk head (external diameter of
りリングの外径大のため）。
the suspension ring is larger
バルクヘッドと吊りリングを接
thaninner diameter of the
続する。container).
transport
The bulk head is connectedwith the suspension ring.

据付
Removal装置
deviceObject tobe取り出しremoved
対象

The lower
搬入時は、据付装置
installation
deice is
下部と炉壁リングを
connected
with the
吊りワイヤで接続しreactorwall ring byた状態
suspension wire
during installation.

•••

After installing in

所定の高さまで搬入後、prescribed height,
作業アームを用いて吊りsuspension ring and
リングと炉壁リングを吊り
the reactor wall ring
ワイヤで接続する。
are connected with

据付
Removal装置
device取り出し
Removal装置
deviceObject to
取り出し
be
対象
removedObject to
取り出し
be
対象
removedsuspension wire byusing the work arm.

A炉壁リング外周の板バネを
plate spring of external
peripheryof the reactor wall
RPV内面に接触させる。
ringisconnected
with the RPV
炉壁リング上部と吊りリングを
inner surface.
吊りワイヤで接続する。
The upper part of the reactor
wall ring is connected by asuspension wire.

• • 取り出し対象に設定した吊りA suspension jig that will be
removed is connected with a
治具と、新炉壁リング下部を
new reactor wall ring by a
吊りワイヤで接続する。
suspension wire.

••

A plate spring of external
取り出し装置外周の板バネをperiphery of removal device is
炉壁リング内面と接触させる。
connected with the innersurface of the reactor wall ring.

©International Research Institute for Nuclear Decommissioning

７．Implementation items of this project【 1) ① Development of interference removal technology】

No.39

a. Removal and transport methods for large-scale structures by accessing from the top◼ Study of specific plan 1 (Removal and transport of the whole structure in integrated state): Study of method to
remove from the reactor
On a basis of concepts in the previous pages, the removal procedures were studied.
Step
手順55

Step
6
手順6

Step
7
手順7

Step
8
手順8

Cuttingof removal object
取り出し対象の切断

Pull-up of removal object
取り出し対象の引き上げ

Reactorwall ring suspension
炉壁リングの吊り下げ

Suspension wire extension
吊りリングからの吊りワイヤの延長

据付
Removal装置
deviceRemoval装置
device

吊上
Suspension
device装置
取り出し
Removal取り出し
Removal装置
device

取り出しtoObjectbe対象removedObject to be対象removedObject to be対象removed

据付

Several suspension
所定長さの吊りワイヤを
wires are prepared
複数本用意して、連結すto adjust theることで吊り位置（高さ）を調節する。
suspending
position(height).

••

••

The
removal object is cut by
取り出し装置に搭載した切断
using
a cutting tool that is
ツールを用いて、取り出し対
equippedwith the removal
象を切断する。
device.
取り出し装置を引き上げ搬出
The removal device isする。
suspendedand thentransported.

••

•

•

The suspension device is
吊上装置を搬入し、装置下部
installed and connected a jigと切断した取り出し対象の吊with suspending wire that is cut
り治具を吊りワイヤで接続すfrom the lower device.
る。
The suspending wire that is
吊上装置のアームで炉壁リン
connected the lower reactor
wall ring and removal objects is
グ下部と取り出し対象を接続
disconnected by using a
している吊りワイヤを解除し、
suspension arm. The whole
吊上装置ごと取り出し対象を
object including the suspension
引き上げる。
device is pull out.

••

•
•

The
installation device is
据付装置を搬入し、据付装置
transported, and connected the
下部と炉壁リング上部を吊りlower installation device and the
ワイヤで接続する。
upper
reactor wall ring with a
据付装置のアームで吊りリン
suspending
wire.
グと炉壁リングの接続を解除
The
connection with thesuspension
ring and the reactor
し、据付装置ごと炉壁リングを
wallring is disconnected by
吊り降ろす。
using the installation arm. The
installation device including thereactor wall ring is lowered.

•Object tobe対象removedA吊りリングから垂れている吊wire suspending from the ring
isりワイヤを延長し、吊りリング
extended and connected withthereactor wall ring.
と炉壁リングを接続する。

©International Research Institute for Nuclear Decommissioning

７．Implementation items of this project【 1) ① Development of interference removal technology】

No.40

a. Removal and transport methods for large-scale structures by accessing from the top◼ Study of specific plan 1 (Removal and transport of the whole structure in integrated state): Study on method to
remove from the reactor

In FY 2018, element tests for the reactor bottom were conducted (stored in the unit canister (UC).
⇒To comparably evaluate the removal method of the whole structure and the conventional cutting method, element tests that are similar to
the above test for the reactor bottom are being planned.
To conduct element tests, procedures of the reactor bottom removal method in an integrated state were studied.
【A concept of the method for removing the reactor bottom 】✓ Installation of the RPV opening to access the lower part
✓ After opening the RPV, fall prevention system is installed.
RPV bottomhead✓ Accessing from the opening, the structures are cut.
(grating, insulator, insulator support, HCU piping, signal cables and CRD
housing support beam)
Grating✓ After implementing fall prevention measures, the bottom head wasremoved from the RPV and then the reactor bottom is suspended in an
integrated state.
Route
Processing
range

establishment

Insulator
Insulator support

HCU piping
CRD housing
Support beam
Signal cable

Suspending
position of thereactor bottom(plan)
Cross sectional view of opening plan of
Horizontal view of planned opening of
the RPV bottom head
the RPV bottom head

Based on summary procedures, a method for removing each structure will be clarified and element tests will be conducted innext fiscal year.
©International Research Institute for Nuclear Decommissioning

７．Implementation items of this project【 1) ① Development of interference removal technology】

No.41

a. Removal and transport methods for large-scale structures by accessing from the top◼ Study of specific plan 2 (combination with removal of divided large-scale structures and decontamination: A basic policy⚫

The following work procedures and facility layout plan were set.

• Interference objects should be as large as possible to transport. In that case, the shielding of the objects should be reduced by decontamination or
divided into appropriate size to transport in combination with shielding (it will be conducted at DS pool or spent fuel pool (SFP)).
• The reactor well should be ensured shielding and confining before removing the shield cover. A concept of confinement area should be followed by
the results in the previous fiscal year; The primary/secondary boundary will be established, the primary boundary will be subdivided and dynamic
boundary will be applied by negative pressure.
• After removing large-scale structure, the area for cutting will be unnecessary. To minimize confinement area for contamination, area setting should
be modified.
• Fuel debris should be stored into the transport container at the location as near as the RPV and then transported. Necessary work for storage
including drying, sorting and storing in the canister should be performed in the another building away from the reactor building. Reduction of weightshould be promoted by reduction of cell functions.

Gas phase

TargetFunctions

Conditions for achieving goal

Necessary equipment(major dynamic
component)

Primary boundary(PCV, cell))Prevention of leaking fromprimary boundary to
secondary boundary

The primary boundary is
maintained at negative pressure
rather than in secondary boundary.

Exhaust fan (installation ofprimary boundary gas
control)

Secondary boundary(R/B, container andcell)Prevention of leaking fromsecondary boundary toenvironment (external air)The secondary boundary is
maintained at negative pressure
rather than external air.

<Overview>＜模式図＞

1次バウンダリ内
In primaryboundary※2-300Pa*-300Pa程度

In-leak
インリーク
損傷
DamageExhaust fan (installation of
secondary boundary gas
control)Pressure: -64Pa
Primary boundary ※圧力:－64Pa程度
１次バウンダリ(PCV + cell)（PCV＋セル等）

PCV
*Same
pressure as inside※圧力:PCV内と同圧
PCV*Pressure: Left +※圧力:左部＋数10Pa程度
several 10 Pa*Pressure:
Left + several※圧力:左部＋数10Pa程度
10 Pa

Container✓✓

万一の場合の隔離扉［検討中］Shielding door for emergency

燃料デブリ
Fuel debris（但し、PCV内の燃料デブリ の加工装置などに繋がるケーブ ル類

等も 撤去する必要があり 計画的な運転・保守の中では閉止不可）

扉
Door（同時に開しない運用）

2次バウンダリ内
In secondary boundary-64Pa*1 ※1-64Pa程度

External
air
外気
Atmosphericpressure
大気圧

インリーク
In-leak

インリーク
In-leakOut-leak
アウトリークアウトリークOut-leakConceptual diagram of boundary (formation image)✓

容器

Secondary boundary(R/B,
２次バウンダリ
container and cell)［R/B、コンテナ、セル］

排気ファン
排気ファン
Exhaustfan
Exhaustfan*1: Equivalent to target
negative pressure of secondary BWR※１：BWRの2次格納施設の目標設定負圧相当*2: -100Pa negative
pressure comparing to secondary boundary※2：２次バウンダリより-100Pa程度負圧(Tentatively, difference
pressures of cells in reprocessing facility are referred. However, conditions of the（暫定的に、再処理施設のセル等の差圧を参考。容器の状態はセル等と異なるため、
container are different有効な差圧については今後の検討結果を取り込む。）
from the cell. Effective difference pressures are required for future study result.

2次バウンダリ
Secondaryboundary（R/Bまたは新設建屋）(R/B
or new building)

Additional
generation of
放射性物質のradioactive
materials
追加発生

１次バウンダリ
Primaryboundary
(PCV,
cell)
（PCV、セル）The primary and secondary boundary are maintained at
negative pressure.
A policy of the primary boundary has several stages considering
prevention of the spread of contamination. (Pressure setting for
the PCV side are lower. Considering maintenance of the double
door, a door is designed with three-protection door).
Installation of isolation door at outer side of PCV is considered
for emergency.

Container
容器(遮蔽、密封)
扉
Door(shielding, sealing)（上流の扉メンテナンス時は当該扉を２重扉［外側］の位置づけで運用）©International Research Institute for Nuclear Decommissioning

７．Implementation items of this project

No.42【 1) ① Development of interference removal technology】

a. Removal and transport methods for large-scale structures by accessing from the top◼ Study of specific plan 2 (combination with removal of divided large-scale structures and decontamination: A basic policy⚫ Prevention measures for the spread of dust during removing the large-scale structure (a basic policy of confinement (ensuring gasphase boundary))•

According to a basic policy, an overall layout plan is developed for removal work of the structure (interferences) and fuel debris retrieval work.

•

Firstly, the structure is removed to the DSP or SFP and divided, and then divided structures are transported after decontamination. The load of
the building can be reduced by installation of shielding right above a pool (currently, the load to be reduced is estimated to be around 10001500 tons (fewer than half of the previous value).

•

After removing the structure, the overall layout is changed and the well is decontaminated by installation of shielding on the upper PCV.

•The DSP and SFP are under preparation for temporary placement of equipment and the transport container, and for maintenance. The overalllayout will be discussed in future.
Secondaryboundary

Secondaryboundary

Primary boundary

Primary boundary

C4M

Shield③Container
storage fortransport

C4L④Pretest fordecontaminationand transport②Dividing of structure anddecontamination (DSP or
SFP)①Move oflarge-scale
structure

C4H

Layout plan for structure removal

C4M

C4L

Shield②Storage ofunit can intransport
container①Storageof debris inunit can③ Pretest fordecontaminationand transport

C4H

Layout plan for fuel debris retrieval©International Research Institute for Nuclear Decommissioning

７．Implementation items of this project【 1) ① Development of interference removal technology】

No.43

a. Removal and transport methods for large-scale structures by accessing from the top◼ Study of specific plan 2 (combination with removal of divided large-scale structures and decontamination: Element
technology required for confirming feasibility planning

Transport crane, shielding suspension(Specifically, prevention measure for
spread of contamination)

C4M

Providing the prospect forelement technology

C4L

Well upper platform(openable, shielding andairtightness)
Primary boundary (inner well, etc.), gap
sealing and decontamination technology

Transport device that can transport
unit can storing fuel debris at the site
PCV upper platform(shielding, airtightness)
Fuel debris processing device
Fuel debris recovery device

Upper access device
for fuel debris
retrieval

Device for removing the whole structure
toward the DSP and removal method
Treatment before transporting the structure(cutting and decontamination)

Dust recovery
equipment (dustscattering prevention
technology)

C4H

Conducted under the previous subsidy
project
Employed the generally-used technology
Under development of subsidy projectin FY2019

Under development of subsidy
project in FY2019 (another project)©International Research Institute for Nuclear Decommissioning

７．Implementation items of this project【 1) ① Development of interference removal technology】
a. Removal and transport methods for large-scale structures by accessing from the top

No.44

Issues and development policy (development issues in element technology were clarified and development policy was established ）◼ Study of specific plan 2 (combination with removal of divided large-scale structures and decontamination: Element technology requiredfor confirming feasibility planning

ID

Element technologies

1

Transport device (canister)2

Well of upper platform

3

PCV upper platform

4

Gap sealing for inside the well,decontamination technology

5

Method for removing the whole
structure (including decontaminationand cutting)Issues of development

Development policy•••

Method for ensuring airtightness: Connection part and remote operation part (lid and lifting machine)*1
Hydrogen control during transportation and criticality control method
Development of basic requirement specifications (standards (site transport standards, etc.), fuel debris
specifications and work plan for workers)

Studied by the subsidy project in
FY2019 and 2020.

••

Devices required for installation before starting removal work of the reactor internals and fuel debris retrieval
work.
Verification of movable shielding operation and airtightness performance during suspending operation.

Element test were completed.
【It can be designed.】

•
•
•
•

Open-close method and its mechanism in the limited space such as the reactor well.
A method for fixing the PCV flange and fixing procedures
Connection with mobile cell and fuel debris retrieval device (airtightness structure)
Maintenance method and procedures

Studied by the subsidy project In
FY2019 and 2020 (to beplanned in 2020)•••

Necessary work before starting removal work of the reactor internals
Work for the well internal to build the primary boundary
Sealing technology for the gate plug gap and the construction method, decontamination of the well andtechnology for extending the PCV•

Selection of the transport method for large-scale structure, study of procedures, lifting machine andStudied by the subsidy project In
suspension tool
FY2019 and 2020
7.1)①a. Technology that the reactor well can be a boundary
Setting of decontamination ranges
Calculation of weight of shielding, available transport crane and estimation of minimum divided structures
Cutting technology for complicated structures such as dryer unit and separator.

•
•
•
•

6

Fuel debris processing and recovery
device••7

Dust recovery equipment••

8

Upper access device for fuel debris
retrieval••

Achieving of processing speed far beyond existing technology and investigation of processing technologythat exceeds the range of existing technology
Selection of processing and collecting methods according to type of debris (specifically, nuclear fuel that has
not been studied
7.1)①a. Technology for
Release of fission products (FP) and its responding method
removing shield cover
A collecting head that can collect dust generated from near the processing points, and study on the hood.
Considering maintenance performance, concepts of water dust collecting system, collecting effects, andmaintenance plan are developed.
Another subsidy project (development of dust
collectionbytechnology)
engages in.
A concept of access device that is improved maintenance performance
modulization.
Development of the maintenance method

7.1)②d. Method for transporting
the unit can from the R/B.

Studied by the subsidy project in
FY2019 and 2020 (existing
technology can be applied todecontamination).

The subsidy project hascompleted. Studied by thesubsidy project In FY2019 and2020, (continued, conceptual
design and element tests
continue)
Studied by the subsidy project In
FY2019 and 2020 (conceptual
study and element tests)
Element tests have done.
【Designing actual device】©International Research Institute for Nuclear Decommissioning

７．Implementation items of this project

No.45【 1) ① Development of interference removal technology】
a. Removal and transport methods for large-scale structures by accessing from the top

Study of specific plan 2 (combination with removal of divided large-scale structures and decontamination: Element technology requiredfor confirming feasibility planning⚫ Technology for removing the whole shield cover into the DSP (or SFP)

◼
•
•

•
•
•

•

It will be stored in a waste container and placed in the premises ofFukushima Daiichi.
Various plans are considered; Cutting in the R/B and stored in thecanister, cut and storage of the canister in the other building after
removing the whole unit, and dividing and transporting the largescale structure in the R/B and then cutting in the other building.
On the other hand, a critical process should be shorten andtherefore, operations in the R/B (cutting, etc.) and transportation
should be minimized.
Decontamination shall be considered, if effective, when transportingfrom the building.
Finally, several plans will be selected among plans for removal of
the whole unit, the divided large structure andcutting/decontamination, considering different states of each unit.
Element technologies will be studied to achieve a target according toeach scenario. In this fiscal year, a method for suspending the wholeunit and a dismantlement method on site will be considered.

Scope of technical study in this fiscal year
Availability of suspending thewhole unit

No

Yes

On site dismantlement

Moving to SFP or DSP fromthe well

Is it possible to transport the shielding
with decontamination? (Is the mass ofthe shielding acceptable?)

Storage cut pieces in waste container

No, the shielding is tooheavy to transport.

Decontamination/cutting (largedivided) can be effective for weight
reduction?

Decontamination (if necessary)

Top leftcover
Top center
cover

Cutting (if necessary)

11828
Top right
cover

Assuming surface
dose:3.8mSv/hCenter cover

Mass 55ton (max.)

Transport and storage of thecontainer
Not effective due to too many dividing
structures

On site dismantlement (cutting into
rubble)Top cover

3200

Storage of waste container intransport container

Storage of the cover in
returnable box with shielding
Transport of returnable box
into the other building

Storage of cutting pieces in waste
container
Storage of wastetransport container

container

inTransport and storage of container

Cutting in the other buildingandstorageofwaste
container

Thickness 618
Bottom cover©International Research Institute for Nuclear Decommissioning

７．Implementation items of this project

No.46【 1) ① Development of interference removal technology】◼
⚫

a. Removal and transport methods for large-scale structures by accessing from the top

Study of specific plan 2 (combination with removal of divided large-scale structures and decontamination: Element technology requiredfor confirming feasibility planning
(3)
(2)
(1)
Technology for removing the whole shield cover)Sequence plan for removal and transport•

Moving of the shield cover on the well for dividing
cutting, decontamination and packing•

Candidate destination is the DS pool or fuel pool.

•

A transport work should be completed inside the cover
to prevent from spreading contamination and to shield.

Therefore, a crane that can transport into the cover
should be implemented.

Bail coverFirstly, a deflection restraining frame isgrounded to control deflection around arobot by a wire tension.

Transfer a robot to the upper cover

(5)

(4)

Hook

Remove a bail cover by using a robot(work range of robot: 3m radius)

(6)

Hook

Contamination
prevention coverHandle a slinging hook by using a robot(7)

Handling of slinging hook by using a
robot

Pull a hook

(9)

(8)

Local exhaust
apparatus

Handle the shielding cover whenlowering the deflection frame.

Hook on four points, following the sameprocedure.

Handle the shielding cover whenlowering the deflection frame.

©International Research Institute for Nuclear Decommissioning

７．Implementation items of this project【 1) ① Development of interference removal technology】
a. Removal and transport methods for large-scale structures by accessing from the top

No.47◼ Study of specific plan 2 (combination with removal of divided large-scale structures and decontamination: Element
technology required for confirming feasibility planning
The most promising destination for⚫ Technology for removing the whole shield cover toward DSP (or SFP)
Applicability and issues for Unit 1 through 3, related to suspending move of the shield cover
Methods and target

Unit 1

Unit 2

Unit 3

Shielding cover andsurrounding
conditions

All the shield cover has fallen andthe lower part (the third layerfrom the top) has also fallen on
the well, which contacted thePCV top head. The well wall liner
was damaged (fallen off). Water
is accumulated in the well seal
bellows. Maximum dose of theupper surface: Approx. 2 sv/h.
The middle lower side: Approx. 1
Sv/h.

Damages are not identified fromobservation of the surface of theupper shield cover. It isestimated to be sound.

Damages were found in theupper of the shield cover (thefirst layer from the top) and theconcave surface was confirmedin the center. The condition of
the lower part is uncertain.

Reference of
information

TEPCO HD website
2019.8.29

TEPCO HD website
2019.1.31

Acceptability of
application△
It is hard to assess acceptabilityof application because of difficult
assessment of the soundness.

〇
Assessment reveals thesoundness.

Minimum conditions
for application of themethod forsuspending thewhole unit

・The shield cover has thestrength (it can enduresuspending transport).
・Amount of dust dispersion fromthe damaged part is withacceptable range.

Issues forassessment ofapplicability

Establishment of assessment
method for the soundness(confirmation of the soundness
for reinforcement arrangement
inside the shield cover)Confirmation of the conditions
including the lower partan entire shield cover to be removed
to is the DS pool, considering thelimit of dimension.

6248

DS pool

15417

Appearance investigation up to
the lower part

TEPCO HD website
2014.2.27△
It is hard to assess acceptabilityof application because of difficult
assessment of the soundness.
・The shield cover has thestrength (it can enduresuspending transport).
・Amount of dust dispersion fromthe damaged part is withacceptable range.
･The lower part should not fallwhen suspending the upper
cover.
Establishment of assessment
method for the soundness(confirmation of the soundness
for reinforcement arrangement
inside the shield cover)Confirmation of the conditions
including the lower part

9906

Fuel pool

12192©International Research Institute for Nuclear Decommissioning

７．Implementation items in this project

No.48【 1) ① Development of interference objects removal technology】◼
⚫

a. Removal and transport methods for large-scale structures by accessing from the top

Study of specific plan 2 (combination with removal of divided large-scale structures and decontamination: Element technology requiredfor confirming feasibility planning
Technology for dismantling the shield cover on the well (on site)

Applicability of the method of removing the whole shield cover for unit 1 and 3 has many issues, therefore comparative study of the dismantlement
method on site (cutting) was conducted.
MethodWater jet method (Method 1)

Wall saw method (disc blade)(Method 2)

Chipping method(Method 3)

Wet/ dryWet crushing method

Wet crushing method (low water amount)Dry crushing method

Assuming construction capability

0.2m3/a day

3.2m/a day (0.35 m depth)

0.3m3/a day

Characteristics

Image of implementation

1) The depth of cut can be controlled

1) General-purpose heavy machine can be used for2) Concrete and reinforcing steel can be cut at theconstruction.

same time.

2) Drainage water is not generated.

1) Reaction force is small.

1) Reaction force of a small type heavy machine

1) Cut pieces are grasped by jig or collected by

2) Abrasive is sprayed on reinforcing steels.

construction is larger than chipping method. 。

vacuum suction.

3）To avoid a sudden fall of the damaged part, it is2) To avoid falling cut parts, grasping method is2) To avoid a sudden fall of the damaged part, it isrepeatedly crushed from the top.

applied.

repeatedly crushed from the top.

1) Small reaction force
2) No effect on work even in buckling part.

1) Waste water treatment apparatus is required to
Major issues

treat large amount of water.
2) Particulate water injection would cause poor
visibility.

Risks of falling on the well

It is hard to collect drainage water, abrasive andfallen crushed materials. In addition, the damage

bellows)

prevention measure for the bellows is required.

Small

Dust dispersion

Abrasive volume: Large
Both of drained water and abrasive are collected.

Construction days
Assessment

internal corner.

3) It is necessary to move blade and install a jig to1) Micro cracks would generate due to vibrating andbeating.
2) Dust and flippered stones would generate.
3) Reinforcing steel cutting is required for long hours.

Fall prevention measures for cut pieces and debris

Fall prevention measures for processing pieces arefrom processing is required.

required.

Large

MediumWork under the contamination prevention cover
Drainage volume: Large

Secondary waste

2) Another construction method is required for

accept reaction force.

(specifically, on the well seal

Reaction force

1) Dust and flipped stones would generate.

Long
This method is not reasonable because large
amount of waste water would generate and theconstruction period is longer. However, if anothermethod causes trouble, (i,.e. Stacks due to a rockpocket), it might be effective during spot period.

Drainage volume: Small
Wasted blade volume: Small
Short
Dust and flipped stones can be prevented by theprotection cover. If the method 3 is applied for theinternal corner, this method can be applicable.

Wasted jig volume: Small
MediumDust and flipped stones can be prevented by theprotection cover. If the method 2 is applied for cuttingof a reinforcing steel in advance, the construction©International
Research
Institute for Nuclear Decommissioning
period would
be shorter.

７．Implementation items of this project

No.49【 1) ① Development of interference removal technology】
a. Removal and transport methods for large-scale structures by accessing from the top◼

Study of specific plan 2 (combination with removal of divided large-scale structures and decontamination: Element technology requiredfor confirming feasibility planning⚫

Technology for removing the whole shield cover

Summary⚫

⚫

⚫

A conceptual study of removing the shield cover that interferes with the top access method was performed.
•

In addition to a conventional plan that the shield cover is cut and stored in the waste container and then transported, various plans wereclarified; A plan for transporting the whole unit from R/B and other operations are performed outside the building including cutting the cover,
and another plan for transporting divided large-scale structures from the R/B (after reduction of weight) and the other operations are
performed outside the building including cutting.

•

Procedures and branch points for selecting scenarios of each plan (a flow chart) were defined. The applicability of the method differs
depending on units.

•

For the plan for suspending the whole unit from the reactor well, a conceptual design of device and the work procedures were performed.

•

For the plan for cutting in the R/B, a list of cutting method was created and comparative evaluation for each plan was performed.

•

To suspend the whole shield cover, it is necessary to study acceptability of feasibility for suspending the shield plug including the upper aswell as the lower part, and on-site investigation data is also required and significant.

Major issues that should be studied under a subsidy project in FY 2020.
•

Removal procedures for each unit are clarified. A plan for removing the whole unit was created and therefore a cutting method on the reactor
well will be developed, specifically prevention measures for collapsing the lowest part and falling plugs and debris will be developed.

•

Decontamination and paint treatment method to prevent spreading contamination adhered on the wall of the well after removing a plug•

Criteria for acceptability of the whole unit suspension method are required to determine. After that, transport device for the whole unit
suspension will be designed.

Response to construction and engineering in futureCollection of practical information to assess the whole unit suspension method (specifically, state of reinforcing steel, risks of fall, expectedradiation doses when suspending and contamination conditions for estimation of radiation emission, etc.)•

To consider cell configurations during removal, the time of installation for the shielding and scale of the shielding (thickness), contamination
state of inner surface of the well and the surface of the cover will be surveyed.
©International Research Institute for Nuclear Decommissioning

７．Implementation items of this project

No.50【 1) ① Development of interference removal technology】
a. Removal and transport methods for large-scale structures by accessing from the top◼ Study of specific plan 2 (combination with removal of divided large-scale structures and decontamination: Element
technology required for confirming feasibility planning⚫ Development of sealing technology for the opening part to establish a boundary of the reactor well
⚫ Basic requirement functions• Confinement of radioactive materials by building a boundary• Requirement conditions for the sealing part (target values)• Differential pressure: Leakage amount for differential pressure design will be set in future.
(Calculated from estimated flow rates in the system to maintain negative pressure in aboundary (-100Pa) and the ratio of the PCV surface area to the opening part.)• Differential pressure design: Differential pressure design will be set in future, which is required
the sealing part to be sound.
• Defect of the sealing part can be detected.
• Negative pressure of the primary boundary during maintenance (differential pressure during
normal operation) should be maintained.
• Reduction of radiation exposure for workers•
Measures to achieve requirement functions: Planning of a complete remote operation.
•
Requirement conditions for the sealing part: The sealing part can be installed/replaced byremote operation.
⚫ Conditions of environmental requirements• Radiation dose: (After installation of PCV flange platform): 5-10mSv/h(Before installation of PCV flange platform): Approx. 100 Sv/h• Temperature： -7～40℃• Humidity： Outside the boundary: Equivalent to external air
Inside the boundary: ≦99%
• Lifetime of design：50 years ( Consumables can be replaced during maintenance.)©International Research Institute for Nuclear Decommissioning

７．Implementation items of this project

No.51【 1) ① Development of interference removal technology】
a. Removal and transport methods for large-scale structures by accessing from the top◼

Study of specific plan 2 (combination with removal of divided large-scale structures and decontamination: Element technology requiredfor confirming feasibility planning

Development of sealing technology for the opening part to establish a boundary of the reactor well•Required sealing parts are identified.

⚫

Name of parts

#
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18

Studied

The reactor well air-conditioning duct opening part (23 parts)○

The reactor well side duct opening part of a skimmer surge tank (2 parts)○

The differential pressure detection tube opening part of the reactor well interruption part (2 parts)○

A gap between a slot and a slot plug (Officially, one side: 13mm)○(4) Erect sleeve(11 parts)(3) Skimmer surge tankdrain opening (2 parts)(6) Difference pressure
detecting pipe opening (2 parts)

G1/G2 gate tube drain of the slot part (1 part)A gap between D/S pit cannel and D/S plug (Officially, one side: 13mm)(5) Air conditioning
duct opening (22 parts)○(2) DS gate

A gap between the well cover and the upper stand of the reactor well (Officially, one side: 13mm)
A gap between the well cover and the other covers (Officially, 13mm)(7) Well drain(3 parts)

A gap between the well cover and the upper stand of the slot plug (Officially, 13mm)
A gap between the well cover and the upper part of the D/S plug (Officially, 13mm)(1) Fuel gate

The main part of the well sealing bellows (In case of damage)
The well sealing bellows and the welding part of the reactor well lower liner (in case of damage)
The welding part of the well sealing bellows and the PCV flange (in case of damage)The reactor well lining plate (in case of damage)
The welding part of the reactor well lining plate (in case of damage)The erecting sleeve of the upper part of the reactor well air-conditioning duct (11 parts)○

A screw hole for handrail on the reactor well cab liner ⇒ No penetration at t25
The well drain of the lower stand of the reactor well (3 parts)•○

A policy of response to Level 2 and 3 to evaluate radiation exposure.

Levelclassification

Evidence

Index of sealing performance (Ex.
Flow rates)

Level 1

Achieving a flow rate of **m3/h
at differential pressure **Pa.

**m3/hLevel 2** times of Level 1**m3/hLevel 3

When running out of sealing****m3/hIn this study, seven parts were selected as representative
structures from the left chart. The well cover related parts (7parts) are to be replaced with a new well cover that is
considered for leakage after removal, and another parts thathave potential damage in welding parts (5 parts) are to be
blocked depending on the damaged conditions or evenly sealedwith grout.
Detection
methodPrevention for
accident occurrence

Recovery for event
occurrence

Based on the sealing structure developed in
FY2019, the structure of each sealing part willbe created in FY2020.
©International Research Institute for Nuclear Decommissioning

７．Implementation items of this project

No.52【 1) ① Development of interference removal technology】
a. Removal and transport methods for large-scale structures by accessing from the top◼

⚫

Study of specific plan 2 (combination with removal of divided large-scale structures and decontamination: Element technology requiredfor confirming feasibility planning
Development of sealing technology for the opening part to establish a boundary of the reactor well(process chart for sealing)

Step 0: Precondition

Step 1: Cleaning

270°
313.5°

226.5°

46.5°

46.5°
Well side
Opening part of skimmer
surge tank drain

180°

Metal net
(existing)0°
Opening(2 parts)

Surface of 2nd layer ofshielding plug

・The 1st layer of the shield plug is removed and thesurface of the second layer completes decontaminationand cleaning.
・Metal net is installed in front of drainage.
・Damages and deformation are not seen
Step 2: Installation of sealing materials

Cleaning brushTechnical issues

Technical issue
・Setting of required cleaning levels

Step 3: Spraying of sealing materials

Step 4: Inspection

・Access route plan
・Plan of a monitoring method
・Clarification of requirement functions for self-driving
robot

Work camera

Fixing hook

Closing plate
A gap for fixing
metal net

Technical issue
・Acceptable value setting of clearance amountSpraying gun forsealing materials

Sealing materials

Technical issues
・Understanding of available construction hours
・Establishment of spraying procedures・Establishment of reasonable evidence todetermine to judge work completion

ExhaustP

Inspection box

Technical issues
・Structure of inspection box
・Method for identifying a leakage point©International Research Institute for Nuclear Decommissioning

７．Implementation items of this project【 1) ① Development of interference removal technology】

No.53

a. Removal and transport methods for large-scale structures by accessing from the top◼ Study of specific plan 2 (combination with removal of divided large-scale structures and decontamination: Element
technology required for confirming feasibility planning⚫

Development of sealing technology for the opening part to establish a boundary of the reactor well

Summary⚫ Sealing technology for the opening part to establish a boundary of the reactor well was studied, whichis required for the top-access fuel debris retrieval method.
⚫ The existing opening parts were surveyed and identified issues to be discussed.

⚫ General-use sealing technologies were surveyed and then sealing technology for each opening partwas selected.
⚫ A work flow chart for remote operated device was created as a implementation plan for selectedsealing technology.
⚫ The following items will be studied and implemented in fiscal year 2020.

•

Listing of issues based on the work step planning•

Maintenance plan for sealing, rescue plan in case of radiation leakage in the sealing part andmonitoring planning to avoid leakage in advance•A policy of solutions to identified challenges including maintenance and rescue, and test andanalysis plans•

A boundary establishment method except for sealing of the opening part; Ex. Comparison ofextension pipes installed from the PCV flange©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ① Development of interference removal technology]

No.54

b. Method of removing interferences from within and outside the pedestal byaccessing from the side
⚫ Purpose of Development➢ To conduct studies related to the method of establishing the access route from the equipment
hatch to the CRD opening, which is necessary for fuel debris retrieval.

⚫ Issues that must be resolved➢ Increase in scope of remote operation and enhancement of work efficiency.
➢ Method of handling cables of the various devices used within PCV.
⚫ Approaches to development➢ Drafting of work procedures from accessing the inside of PCV to removing interferences.
➢ Study of the methods of removing interferences that are on the access route from the equipment
hatch to the CRD opening, and element tests.
➢ Development of the method for installing utilities for the devices used inside PCV.
⚫ Expected outcome➢ Work procedures from accessing the inside of PCV to removing interferences.
➢ Method of establishing the access route from the equipment hatch to the CRD opening.
➢ Method for installing utilities for the devices used inside PCV.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ① Development of interference removal technology]

No.55

b. Method of removing interferences from within and outside the pedestal by accessing from the side

◼ Preconditions of the study✓ PCV and expanded building (*1) can be connected by means of a passage (access tunnel).
Further, the access tunnel is installed by launching it remotely from outside the R/B.
✓ The yard has been prepared for access tunnel installation, measures pertaining tounderground structures have been taken and structures on the ground have been removed.
✓ The work of removing interferences from inside PCV and the work of fuel debris retrieval is
carried out using multiple remote operation devices. Devices are assembled inside PCV as
required.
Carry-in
cell
搬入セル

Waste
廃棄物
carry-out
cell
搬出セル

燃料デブリ
Fuel
debrisretrievalcell
取り出しセル
保守セルcellMaintenance

PCV
Fuel
debris
燃料デブリ
carry
out cell
搬出セル
Passage(access tunnel)
通路(アクセストンネル)(*1): Since the detailed functions of the expanded building are still being studied, it is defined
as a facility with preliminary treatment functions for storage including fuel debris drying.

Image showing how the PCV andexpanded building are connected©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ① Development of interference removal technology]

No.56

b. Method of removing interferences from within and outside the pedestal by accessing from the side

◼ Study Flow

: Reported this time

Policy for initiatives for the current
fiscal year in response to issues
from the past fiscal year

• Outside the pedestal: Verification of feasibility of thework of installing utilities for implementing the work ofassembling remote operation devices• Inside the pedestal: Verification of feasibility ofinterference removal method / fuel debris retrieval
method using the CRD platform

Study of work outside the pedestal

Study of work inside the pedestal• Installation of common utilities for remote operation deviceson the grating on the first floor outside the pedestal

Issues faced in making
work outside the pedestal
possible and related
studies (including tests)

Issues faced in making
work inside the pedestal
possible and related
studies (including tests)• Verification of feasibility by means of element
tests related to installation of common utilities
Scope of implementation
under the subsidy project

Site situation investigation results

Actual construction plans©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ① Development of interference removal technology]

No.57

b. Method of removing interferences from within and outside the pedestal by accessing from the side

◼ Development Process
Discussion items
Major milestones：Planned: Planned (afterrevision)*: Actual

FY2019
４

5

6

7

8

9

10

11

FY2020212

1

2

Interim Report
▼

3

4

5

6

7

8

Interim Report
▼

9

10

11

12

1

2

3

Interim Report▼

1. Conceptual study /Issue identification
2. Test plan

Outside the pedestal

3 Test preparation /Test apparatus
prototype

4. Element tests

5. Summary

Outside the pedestal

Outside the pedestal

Outside the pedestal

Inside the pedestal

Inside the pedestal

Inside the pedestal

Inside the pedestal

Remarks* The plan was revised to reflect the study progress status and the results of element tests.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ① Development of interference removal technology]

No.58

b. Method of removing interferences from within and outside the pedestal by accessing from the side

◼ Overview of the method to be studied✓

The PCV and expanded building are connected by means of a passage (accesstunnel). The access tunnel is installed by launching it remotely from outside the
R/B.

The floor load inside R/B can be reduced by using the access tunnel.

➢

If the layout of the access tunnel interferes with other work, the interference
can be prevented by changing the route of the access tunnel (highly flexible
layout plan).
Amongst the existing penetrations, the equipment hatch is used as theaccess point for the access tunnel.

✓

The work of removing interferences from inside PCV and the work of fuel debris
retrieval is carried out using multiple remote operation devices. Devices areassembled inside PCV as required.

✓This method is applicable regardless of the unit and whether the interference isinside or outside the pedestal.

燃料デブリ
Fuel
debrisretrievalcell
取り出しセル
保守セル cell
Maintenance

➢

➢

搬入セル
Carry-in
cell

Waste
廃棄物
carry-out
cell
搬出セル

PCV
Fuel
debris
燃料デブリ
carry-outcell
搬出セル
Passage (access tunnel)
通路(アクセストンネル)

Image showing how the PCV andexpanded building are connected[Example of remote operation device]
Movementarm
移動アーム

Arm section

Transportation
deviceBase section

Flexible work arm (crawler, dual arm type)

CuttingdeviceTransportation device andcutting device

作業アーム
Work arm
Flexible work arm (multi-arm type)
Device for removing interferences from inside the pedestal©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ① Development of interference removal technology]

No.59

b. Method of removing interferences from within and outside the pedestal by accessing from the side

◼ Main work steps✓ Following is a series of the main work steps from preparatory work to fuel debris transportation.
Work
step

① Construction of provisional building
Opening up the R/B wall

② Launching the access tunnel

Provisional
仮設建屋
building
Shielding door

アクセストンネル
Accesstunnel

R/B壁開口部
R/Bwall opening③ Connection of the access tunnel

Expanded

増設建屋
building

Openingof R/B wall
R/B壁面開放Translation
並進

AccesstunnelTurning
旋回

仮設建屋内部
Insidethe provisional building

Remarks*

Conceptual study has been conducted.
Since it is presumed that existing technology
can be used, the following steps are being
studied on a priority basis.

••⑤ Removal of interferences outside thepedestal(Establishment of access route inside
D/W)

Work
step

④ Construction of expanded building

Balance weight

Inside
R/B
R/B内部

Step
diagram••: Studied in FY2019-20

取水開口部１
Hole 1 boredfor water intake

Accessアクセス
tunnel
トンネル

AT sleeve

Feasibility has been verified through element
tests. (FY2017-18)
A turning element test simulating actual weightis planned.

•⑥ Installation of weir for preventing
spread of contamination

Element tests related to the method of welding
the sleeve and PCV are planned to be
implemented as part of this project. (FY2020)•⑦ Laying of pipes in the D/WInstallation of submersible pump

Detailed study by the operator’s engineering
department is underway.

⑧ Removal of interferences from inside
the pedestal

HVH -12D

1470× 640
HVH -12E

SRV配
- 管
SRV
pipingAccess tunnel
アクセストンネル

SRV-piping
SRV配
管

9 0°

N

Equipment
hatch
機器ハッチ

干渉物撤去装置

Interference removal device

水中ポンプ

Submersible pump

Staircase階段
LP-12A

Step
diagram

エアロックステージ
Air lock stage

Pipe
supports
配管サポート類

1 8 0°
Air lock
エアロック

0°
HVH -12A

干渉物切断

Cutting the interference

SRV
SRV配piping
管

HVLP -D-1

Hole 2 bored for
取水穿孔部２water
intakeHole bored for detailed
investigationof the inside of PCV
内部詳細調査用穿孔部
PCV

Weir
堰

SRV
pipingSRV配管2 7 0°

HVH - 12C

ペデスタル内PCV内配管Piping inside PCV

1500×970

HPCI -21

Inside the pedestal

HVH -12B

PC V内部 干渉物撤去前 （ 平面図）•

Remarks*•

Interferences from inside the D/W (first floor) areremoved.
Element tests related to the tasks from accessing thebasement floors to the work of removing the piping
have been implemented. (FY2017-18)•••

The D/W basement floor is accessed.
A weir for preventing spread of contamination is
installed.
Preliminary tests related to installing the weir aroundthe jet deflector (J/D) have been implemented.
(FY2017-18)••

Construction of the water circulation system
Element tests related to laying of pipes and pumpinstallation have been implemented as part of the water
circulation PJ. (FY2019)* Items that have been underlined are being implemented as part of other projects or by the engineering department.

••

Element tests related to the method of interference
removal using a beam have been implemented.
(FY2017-18)
Interference removal element tests using the CRD
platform are planned to be implemented. (FY2020）©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ① Development of interference removal technology]

No.60

b. Method of removing interferences from within and outside the pedestal by accessing from the side

◼ Main work steps
Work
step

: Studied in FY2019-20⑨ Fuel debris retrieval⑩ Fuel debris transportation
Image of debris transportation (Overview)
Multi-axial robot

Dolly cart type multi-axial robot

Step
diagram

Unit canユニット缶

Muscle-robot

Transportation

搬出ルート
route

回収容器

Collecting canister••

Remarks*••

The figure illustrates an example of work carried out
outside the pedestal.
Work inside the pedestal is carried out by means of
the interference removal device.
Element tests related to the method of fuel debris
retrieval using a beam have been implemented.
(FY2017-18)
Element tests related to fuel debris removal using the
CRD platform are planned to be implemented.
(FY2020)••

Conceptual study has been conducted.
Although technological development is necessary fortransportation work in this step, if the preceding steps
are not feasible, this step won’t need to be carried out,
and hence the preceding steps are being studied on apriority basis.

Steps ③, ⑤, ⑥, ⑧, ⑨ are being studied as part of this subsidy project.
(The other steps are either being studied as part of other projects or by the engineering department, or element tests for those steps
have already been implemented.)✓ Detailed development content for step ③ is described under “3) (i) ② Technology for ensuring confinement functions in connection parts”.
(Slide 260 to 270)✓

Detailed development content for steps ⑤, ⑧, ⑨ is described under “1) Development of interference removal technology” in this section.
(next slide onwards)✓

Detailed development content for step ⑥ is described under “1) ② Development of various technologies other than interference removal
technology”. (Slides 88 to 107)
©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ① Development of interference removal technology]

No.61

b. Method of removing interferences from within and outside the pedestal by accessing from the side
◼ Overall strategy for the work of removing interferences from outside the pedestal✓

During the work of removing interferences from outside the pedestal, the work of providing utilities for removal work is carried out in parallelwith the interference removal work.
⇒ Verifying whether or not providing utilities is feasible, is as important as confirming whether or not the interference itself can be removed.
(Since the feasibility of provision of utilities has an impact on the interference removal method)⇒ As part of this project, efforts will be made to verify the feasibility of providing utilities prior to interference removal.
Interferences to
撤去対象干渉物
beremoved.

[Procedures for interference removal and for providing utilities](1) Work devices are installed through the equipment hatch.
(2) Interferences from around the equipment hatch (innerside of PCV) are removed.
(3) Utility (hydraulic pressure / electric) line is installedaround the equipment hatch.
(4) The cable connection of the work device is changed over(reconfigured) to the installed utility line.
(5) Surrounding interferences are removed.
(6) Utility line is extended.
(7) Steps (4) to (7) are repeated. (till CRD opening)(8) Installation of utility line is completed.

PLR TERMINAL BOX
DHC-1

PCV shell
PCVシェル

HPCI-210°MS-30CN

HVH-12C

Air lock
エアロック
Air lock stage
エアロックステージ

HVH-12B

PLR motor

Staircase
階段

PCVED-4270°
MS-30D

Utility(Piping, etc.)

Equipment
機器ハッチhatch

X-6
HVH-12D
MS-30AU stand

PLR motor

HVH-12A

90°
HVH-12E
MS-30BPLR TERMINAL BOX

Interferences to be removed from outside the pedestal (first floor)

Repeated in alternation

Utilities, such as piping, etc. areconnected on the utility stand (U stand)： Utility line
： U standInstallation of work devices
Interference removal
Utility line extension

Removal ofinterferences fromaround theequipment hatch,
installation of utility
line
Installation of work devices

Interference removal

Utility line extension

Utility line installation completed©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [1) ① Development of interference removal
No.62
technology]

b. Method of removing interferences from within and outside the pedestal by accessing from the side

◼ Study of the method of installing common utilities
Installation of common utilities around the area outside the pedestal was studied as a proposed solution for the issue identified lastyear, namely, enabling remote operation of the construction of removal devices and surrounding equipment of the remote operation
devices used inside the PCV.
: Utility line

[Concepts of common utilities]• Water (hydraulic pressure) and electricity,which is the drive source of the remote
operation device used inside PCV, is
supplied.

ユーティリティラインとPCV内でのThe utility line and the cable of thework device inside PCV are connected.
作業装置のケーブルを接続
Equipment
hatch
機器ハッチ

: Connector

ユーティリティライン：Utility line:
installed remotely
遠隔にて構築• The work of installing the utility line is carriedout by means of remote operation devices(flexible work arm, etc.) that operate
independently.
• The remote operation devices used for thework of removing interferences from within
and outside the pedestal, and for the fueldebris retrieval work are connected to theconnectors near the work place and utility(hydraulic pressure / electricity) is supplied.
• Since one end of the cable of the remote
operation device is connected to theconnector near the work place rather thanthat at the equipment hatch (expandedbuilding), there is no need to manage longcables.

Monitoring
監視装置
deviceConceptual diagram ofcommon utilities

Scope of installation of common
utilities (planned)©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ① Development of interference removal technology]

No.63

b. Method of removing interferences from within and outside the pedestal by accessing from the side

◼ Study of the method of installing common utilities
Preliminary tests pertaining to the connection for installing the common utilities, using the flexible work arm, are implemented.
[Preliminary test conditions pertaining to the connection with the connector under a simulated environment of the area outside the pedestal]Long shot
cameraConnector

Long shot
camera A

Cable
Flexible work arm(Type II)

Connector
Work arm handset

Connector
Stand

Connector installation status

Procedure

Criteria

Results

Is theconnection
possible?

Hold and convey theconnector with the flexible
work arm, and when thepositions of the terminal andconnector match, fix it with aclamp. Carry out work based
only on the image on thecamera.

・The connector should not
fall while holding andconveying.
・One should be able to fix
the clamp after the positionsof the terminal andconnector match.

Good

Confirmation
afterconnection

After the connection work iscompleted, check theconnection part between theterminal and the connector bymeans of the image on thecamera.

The connection part
between the terminal andthe connector should not be
deformed.

Good

Is there any
leakage?

After the connection is made,pass water through the pipe
inside the terminal.

There should be no leakage
from any of the pipes.

Good

Initial layout

Work arm handset

Terminal(Receiving end)

Items forVerification

Clamp part

Connector clamp status✓

It is confirmed that the connector can be connected through remote operation.

✓

There are still Issues such as the method of determining the position and the method of verifying completion of connection. Solutions forthe issues were studied. (Refer to the following slide)©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [1) ① Development of interference removal
technology]
b.

No.64

Method of removing interferences from within and outside the pedestal by accessing from the side

◼ Study of the method of installing common utilities
The configuration of common utilities studied based on the results of the preliminary tests, is as follows:
Positioning jig[Connection method]① Press the positioning jig on the guide of
the U stand and connect it to the position
determining pin.

U standPipe② Press a new U stand on the guide of thepositioning jig and connect it to the position
determining pin.
③ After repeating the above-mentioned
steps ① and ② and installing the U standinside D/W, connect the pipe on the U stand.

Positiondetermining pin(Positioning jig)
Pipe expansion
part

Guide

Procedure②

Top View

Procedure①

Lateral View

Positiondetermining pin(U stand)
Procedure③

Positiondetermining pin
connection holePipe
expansion part

Guide

The pipe is inserted in the V
groove of the U stand andV groove connected by moving the pipe
expansion part.

The applicability of remote operation in the above-mentioned configuration will be verified by means of element tests*.
*Refer to the table on Slide 71 for details on the contents of the element tests.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ① Development of interference removal technology]
b.

No.65

Method of removing interferences from within and outside the pedestal by accessing from the side

◼ Work steps for removal from outside the pedestal✓

The work steps for removal from outside the pedestal, which includes the installation of common utilities, were sub-divided, work contents
were specified and issues were identified.
[Step ④] Construction of the
expanded building(Refer to Slide 59)

1. Work device transportation, interference removal (around theequipment hatch), utility installation

Work step

Removalof
機器ハッチ周囲のinterferences from
干渉物撤去、ユー
around the equipment
hatch
and installation of
ティリティライン設置
utility line

Step
diagram: Items to be verified by meansof the element tests
3-1. Utility line extension*(Installation of positioning jig)

2. Removal of interferences from inside D/W (first floor)*

Transportation and installationof the positioning jig
Transportation of
作業装置搬入wok
deviceUtility (piping, etc.)

Interference
干渉物撤去
removalU standConceptual diagram of U standDetails

The work device is brought in from the equipment hatch and theinterferences around the equipment hatch (inner side of PCV) areremoved. Thereafter a utility line (hydraulic pressure / electricity) that is
extended from the access tunnel is installed around the equipment
hatch (inner side of PCV).
•

Issues

The cable connection of the work device is moved over to the utility
installed in the previous step. Thereafter, the surrounding interferences
are removed.

Method of pulling the utility extending from the access tunnel.
(Element tests concerning whether or not the connector can beconnected and the interferences can be cut off by means of remote
operation, have been implemented. [FY2015-16])

The positioning jig for installing the U stand is brought in up to thedesignated location and installed using the work device.

•
•
•
•

Whether or not the positioning jig can be held with the work device
Transportation of positioning jig using the work device
Accuracy of installing at the target position
Whether or not the positioning jig can be fixed*Since the utility line is installed (extended) while removing interferences from the first floor, steps 2 → 3-1 to 3-4 are repeated. (Until the area near the CRD opening is reached)©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ① Development of interference removal technology]
b.

No.66

Method of removing interferences from within and outside the pedestal by accessing from the side

◼ Work steps for removal from outside the pedestal
Work
step

Step
diagram

Details

Issues:: Target of element tests

3-2. Utility line extension*(Installation of U stand)3-3. Utility line extension*(Bringing in the segmented utility line)

Transportation andinstallation of U standBringing in the segmented
utility line (piping)

Utility (piping, etc.)

3-4. Utility line extension*(Connecting the segmented utility line)

Pipe (existing)Utility (piping, etc.)

U standU standConceptual diagram of U standConceptual diagram of U stand

Pipe (new)The U stand is transported to the designated locationand installed using the work device.

The segmented pipe is brought in to the designatedlocation using the work device.

The segmented pipe is fixed on the U stand andconnected to the existing pipe.

•

•
•

•
•

•
•
•

Whether or not the U stand can be held with thework device
Transporting U stand using the work device
Accuracy of installing at the target position
Whether or not U stand can be fixedTransportation of utility line using the work device
Accuracy of temporary positioning at the target
position

Fixing of the pipe on the U standConnection with the existing pipe using the work
device*Since the utility line is installed (extended) while removing interferences from the first floor, steps 2 → 3-1 to 3-4 are repeated. (Until the area near the CRD
opening is reached)©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ① Development of interference removal technology]
b.

No.67

Method of removing interferences from within and outside the pedestal by accessing from the side

◼ Work steps for removal from outside the pedestal
Work
step

4. Completion of interference removal frominside D/W (first floor) and laying of the utility
line
PCVシェル

5. Creating an opening in the grating, liftingand lowering the interference removal device

6. Removal of interferences from inside D/W(basement floor)

0°

エアロック

アクセストンネル

270°
X-6

45°

機器ハッチ

Step
diagram

90°

アイソメ図

Utility line
：ユーティリティライン：ユーティリティスタンド

Details

IssuesBy repeating steps 2 → 3-1 to 3-4, theinterferences on the first floor inside D/W are
removed and the utilities for the work device are
installed around the area outside the pedestal.

(Same as Steps 2 to 3-4)

Device formaking openings

Interference
removal device forthe basement floor

The grating is cut, and the interference removal
device is installed on the basement floor.

(Element tests concerning the tasks of cutting thegrating, making an opening and installing theinterference removal device on the basement
floor have been implemented. [FY2017-18])

Since the submersible pump is to be installed in
the pit, the interferences in the vicinity of the pit
such as the existing pump, etc. in the pit areremoved. If there are deposits, the deposits are
removed.
(Element tests concerning removal ofinterferences from the basement floor have been
implemented. [FY2017-18])©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ① Development of interference removal technology]
b.

No.68

Method of removing interferences from within and outside the pedestal by accessing from the side

◼ Work steps for removal from outside the pedestal
7-1 Installation of weir for preventing the spread ofcontamination*(Formwork installation)

Work
step

: Target of element tests**

7-2 Installation of weir for preventing the spread ofcontamination*(Dry mortar injection)

8. Deployment of piping inside D/W and installation ofsubmersible pump

Stagnant water
滞留水移送ポンプ
transfer pumpEquipment
hatch
機器ハッチ
Weir
堰

Stagnant water
滞留水移送ポンプ
transfer pumpEquipment
hatch
機器ハッチ
堰
Weir

Submersible

水中ポンプ
pump

Weir
堰

Step
diagram

Weir
堰
：堰

：堰

Piping inside

PCV内配管
PCV

Details

Issues(The figure indicates the image of installation of weirs in Unit 1.)(The figure indicates the image of installation of weirs in Unit 1.)

The folded weir (formwork) is brought in through the opening
in the grating using the interference removal device, etc. forthe basement floor. Thereafter, it is spread and installed on
the basement floor.

Dry mortar is injected in the folded weir that is now spread,with the help of the remote operation device (flexible workarm, etc.).

•

•

•

Whether or not the formwork (folded) can be carried inthrough the opening (1200 x 600mm)
Whether or not the formwork can be spread on thebasement floor

Pump
pit
ポンプピットJet
deflector

ジェットデフSump pit
サンプピット•Whether or not dry mortar can be injected into the spread
formwork
Whether or not the dry mortar can be transported to thebasement floor within the time period before it starts
hardening

Weir
堰

Pipes are deployed inside D/W and submersible pump isinstalled inside the pit.

(Element tests concerning laying out of pipes, installation ofpumps are planned to be implemented as part of the water
circulation PJ. [Implemented in FY2019])* Steps 7-1 to 7-2 are repeated until the desired weir is built for each unit.

Weir
堰
Worker access
作業員アクセス口
opening

Testtarget

試験対象

Image of installation in Unit 1

Worker access
作業員アクセス口
opening

Image of installation in Units 2 & 3©International Research Institute for Nuclear Decommissioning** Details of the test contents related to work steps 7-1/7-2 are described separately in section 7. (1) ② c “Method of preventing spread of contamination into S/C” (Slide 88 onwards).

7. Implementation items of this project[1) ① Development of interference removal technology]b.

Method of removing interferences from within and outside the pedestal by accessing from the side

◼ Work steps for removal from outside the pedestal
Work
step

No.69
: Target of element tests

9. Installation of drainage pump on the outer side of the weirs

D/W intake pump
D/W取水ポンプ
Equipment
機器ハッチhatch

Stepdiagram

Weir
堰

To theequipment
機器ハッチへhatch

Drainage

排水ポンプ
pump

[Step ⑧] Removal of interferences frominside the pedestal(Refer to Slide 59)
Drainage
排水ポンプ
pump

D/W intake pump
D/W取水ポンプ

Details

A submersible pump is installed on the outer side of the weirs. Further,
installation of submersible pump on the inner side of the weirs has been
completed in the previous step (Step 8). The submersible pumps areoperated after installation.

Issues(Same as Step 8)©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ① Development of interference removal technology]
b.

No.70

Method of removing interferences from within and outside the pedestal by accessing from the side

◼ Scope of simulation in element tests✓ Element tests concerning the work of installing common utilities are implemented for resolving theissues identified until the previous slide.
✓ Since the work of installing common utilities is a repetition of extension work, the final installation
location is simulated and it is assumed that the utilities up to the previous stage have been installed.

Floor above
the ground

Equipmenthatch
Utility line (installed)
U stand

Pipe

Weir installation
work area*

Items to be verified
through the tests

Scope of simulation in the common utility installation test
©International Research Institute for Nuclear Decommissioning*The details of the contamination spread prevention weir are described separately in Section 7. (1) ② c “Method of preventing spread of contamination into S/C” (Slide 88 onwards).

7. Implementation items of this project[1) ① Development of interference removal technology]
b.

No.71

Method of removing interferences from within and outside the pedestal by accessing from the side

◼ Element test items and criteria
ID

1

2

Test item

U standassembly

Connection of
the utility lines

Relevant
work
steps*1

Test outline and procedures
•
•

3-1
3-2

(1) Lift the positioning jig with the gantry-type lifting machine.
(2) Transport the positioning jig to the target position (existing U stand).
(3) Lower the positioning jig while pushing it against the existing U standguide and fix the jig to the existing U stand.
(4) Lift the U stand with the gantry-type lifting machine.
(5) Transport the U stand to the other end of the positioning jig installedin step (3).
(6) Lower the U stand while pushing it against the guide at the other endof the jig, and fix the U stand to the other end of the positioning jig
installed in step (3).

•

3-3
3-4

(1) Place the segmented pipes on the transport cart. [Manual work](2) Transport the pipes to the vicinity of the U stand by the transport cart.
(3) Install the pipes on the U stand with the gantry-type lifting machine.
(4) Fix the pipes to the U stand and connect them with the flexible workarm.

Criteria•••

Remarks

The positioning jig can be lifted.
The positioning jig and U stand can be transported tothe target position.
The positioning jig and U stand can be fixed together.

The pipes can be transported to the target position by
the transport cart.
The pipes can be installed from the transport cart on to
the U stand.
The pipes can be connected with the flexible work arm.

#1: Refer to the work steps on Slides 65 to 69[Devices used for the element tests]

Jig (beingtransported)
U stand
Lighting
CameraHolding the jig andassisting in the work

Transporting andinstalling the jigs

Transport
cart
Transporting thejigs

Flexible arm for remote operations*Gantry-type lifting machineTransport cart

Work monitoring

Camera cart

Positioningjig

U stand and positioning jig*There are several variations in flexible arms for remote operations. The above specification is an example of a crawler-type flexible arm.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ① Development of interference removal technology]
b.

No.72

Method of removing interferences from within and outside the pedestal by accessing from the side

◼ Test procedure (Assembling the stand) #Verified by a preliminary test
STEP

1. Carrying in thepositioning jig
Gantry-type
lifting
machine2. Connecting the jig to the U stand positioning pin
U stand

3. Verifying that theconnection is complete

Positioning pin (U stand)U stand
Positioning jig

Illustrationofoperation
Positioning jig
Positioning jig

STEP

4. Carrying in the U standGantry-type
lifting machine

Positioning pin (U stand)5. Connecting the U stand to the positioning pin of thepositioning jig
U stand

6. Verifying that theconnection is complete
U stand

Positioning jig

Illustrationofoperation

U stand

Positioning jig

Positioning pin (positioning jig)

U stand

Positioning jig

Positioning pin (positioning jig)

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ① Development of interference removal technology]
b.

No.73

Method of removing interferences from within and outside the pedestal by accessing from the side

◼ Element test results (Assembling the stand)
STEP

1. Carrying in thepositioning jig

2. Connecting the jig to the U stand positioning pin

3. Verifying that theconnection is complete

Positioning pin(U stand)Illustrationofoperation
Positioning jig

STEP

4. Carrying in the U standPositioning pin(U stand)5. Connecting the U stand to the positioning pin of thepositioning jig

Positioning pin(U stand)

6. Verifying that theconnection is complete

Positioning jig

Gantry-type
lifting machine

Illustrationof operation

U stand

Positioning pin(positioning jig)

Positioning pin(positioning jig)

It was confirmed that the positioning jig and U stand can be installed and connected remotely©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ① Development of interference removal technology]
b.

No.74

Method of removing interferences from within and outside the pedestal by accessing from the side

◼ Test procedure (Connection of the utility lines) #Verified by a preliminary test
STEP

1. Preparing the pipes [Manual work]
U stand

Cameracart

2. Transporting the pipes

Gantry-type
lifting machine

3. Installing the pipes

U standU standIllustrationofoperation
Pipeassembly

STEP

Gantry-type
lifting
machine4. Connecting the pipes

Pipe assembly(Existing:Connection
destination)

Gantry-type
lifting
machinePipe assembly(Connecting part)

5. Verifying that the connection is complete

Operate the lever with the handset
to connect the pipes

Illustrationofoperation

Pipeassembly

Flexiblework arm

Connecting
part

Pipe assembly(new)Pipe assembly(Existing)

Lever©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ① Development of interference removal technology]
b.

◼

No.75

Method of removing interferences from within and outside the pedestal by accessing from the side

Element test results (Connection of the utility lines)
STEP

1. Preparing the pipes [Manualwork]
U stand

Camera cart

2. Transporting the pipes

Pipe assembly (Existing:Connection destination)

Gantry-type
lifting machine

3. Installing the pipes

Pipe assembly(Connecting part)

Illustrationofoperation
Pipeassembly

STEP

Gantry-type
lifting machine

4. Connecting the pipes

U standGantry-type
lifting machine

U standLever

5. Verifying that the connection is complete

Operate the lever with the hand-set toconnect the pipes

Illustrationofoperation

Pipeassembly

Connecting
part

Flexible workarm

It was confirmed that the utility lines (pipe assembly) can be installed and connected©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ① Development of interference removal technology]
b.
◼

No.76

Method of removing interferences from within and outside the pedestal by accessing from the side

Summary of element test results
ID

1

2

Test items

Assembling
the standConnection of
the utility lines

Test outline and procedures(1) Lift the positioning jig with the gantry-type lifting machine.
(2) Transport the positioning jig to the target position (existing U stand).
(3) Lower the positioning jig while pushing it against the existing U
stand guide and fix the jig to the existing U stand.
(4) Lift the U stand with the gantry-type lifting machine.
(5) Transport the U stand to the other end of the positioning jig
installed in step (3).
(6) Lower the U stand while pushing it against the guide at the otherend of the jig, and fix the U stand to the other end of the positioning jig
installed in step (3).
(1) Place the segmented pipes on the transport cart. [Manual work](2) Transport the pipes to the vicinity of the U stand by the transport
cart.
(3) Install the pipes on the U stand with the gantry-type lifting machine.
(4) Fix the pipes to the U stand and connect them with the flexible
work arm.

Assembling the stand

Test results

•
•
•

•
•
•

Remarks

It was confirmed that the positioning jig can be lifted withthe gantry-type lifting machine.
It was confirmed that the positioning jig and U stand canbe transported to the target position with the gantry-type
lifting machine.
It was confirmed that the positioning jig and U stand canbe fixed together.

Good

It was confirmed that the pipes can be transported to thetarget position by the transport cart.
It was confirmed that the pipes can be raised from thetransport cart by the gantry-type lifting machine and canbe installed on the U stand.
It was confirmed that the pipes can be connected with theflexible work arm, using the pipe connecting jig.

Good

Connection of the utility lines

The feasibility of the work of establishing common utilities (assembling the stand, connectionof the utility lines), while satisfying all the criteria, was confirmed.
©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ① Development of interference removal technology]
b.

No.77

Method of removing interferences from within and outside the pedestal by accessing from the side

◼ Identified issues and response policies
No.

Issues

Details

Response policy

1

Method of handling theunevenness on the grating(floor surface)In this test, a U stand whose height was manually
adjusted was installed on a flat grating, but there is adifference in levels in the actual grating, whichcauses an error during assembly work such asalignment of pipes, etc.

Enable the U stand adjuster to expand and contractby means of remote operation.
Lay a scaffold on the grating to ensure a levelsurface.

Method of monitoring
during work (Camera andlighting arrangement)

It is difficult to check the alignment of low-lying
locations, such as the contact surface of U stand andpositioning jig.

Conduct a detailed study on the camera and lighting
arrangement so that each task is easily visible, forexample, mounting a camera on the tip of theoperation device and confirming the task at hand (inthe engineering work plan for actual operation).
Also conduct studies on structures that can be
indirectly verified. (The improved effect due to theremodeling of the camera cart has already beenverified)

Handling device cables

Manual assistance has to be provided for the extra
length of the cables outside the simulated test range(between equipment hatch to around 270°).

Study remote handling methods since devices andcables are crowded in a narrow environment.
(The effectiveness of handling the extra length with acable feeding cart has already been verified)2

3©International Research Institute for Nuclear Decommissioning7. Implementation items in this project

No.78

1) Development of fuel debris retrieval methods② Development of various technologies other than interference removal technology
A system for supporting the operation of the remote operation devices, technology for preventing the spread ofcontamination outside PCV and technology believed to be needed for establishing the access route, considering theconstruction of a shielding structure for exposure reduction in the reactor building (R/B), will be developed besides
the technology for removing interferences.
The following will be included as the main development and study items. Element tests will be conducted as
required to identify problems and organize them.
a. Development of remote operation support method in environments with low visibility and narrow spaces.
・Creation of an environment model (test version) and simulator are underway.
b. Methods to prevent the spread of contamination to the suppression chamber (S/C).
・Element tests will be performed concerning the method of carrying in segmented formwork as a way ofbuilding weirs remotely.
c. Conceptual study on the establishment of an access route for installation of cells and method of reducing
impact on R/B and other buildings.
・A structural study on determining the concrete details for reducing the load on R/B floor is underway.
d. Method of transporting the unit can from the R/B.

・ A conceptual study on how to carry out the fuel debris in unit cans is underway.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [1) ② Development of various
technologies other than the interference removal technology]
a.
⚫

No.79

Development of remote operation support method in environments with low visibility and narrow spaces
Need for development

Tip

Elbow

[General technologies]•
Operator carries out the operation looking at the “tip” of the manipulator•
The movements of other joints such as the manipulator’s “elbow” and “shoulder” areleft to take their own course ⇒ “Elbow”and “shoulder” are likely to interfere withmovements.

[Methods developed in the subsidy projects so far]•
The manipulator can be moved in any direction by pointing to the “elbow”“shoulder”
one at a time
[New issues]•
Which part of the manipulator is likely to interfere?
•
Which axis of the manipulator should be moved in which direction to avoid theinterference?
⇒ It is difficult to monitor all movements from a limited perspective/ field of vision⇒ One operator cannot operate many joints such as “tip", "elbow", and "shoulders"simultaneously (it is also difficult for several people to operate in cooperation witheach other)[Purpose of this development]•
To automate the movements to avoid "elbow" and "shoulder" interference so that theoperator can focus on the movement of the manipulator “tip”

Verification of workability of X-6 penetration(Decontamination project)

Picture of an operation performed by
avoiding interference of “elbow” and“shoulder”©International Research Institute for Nuclear Decommissioning7. Implementation items of this project
No.80[1) ② Development of various technologies other than the interference removal technology]a.

Development of remote operation support method in environments with low visibility and narrow spaces⚫

Purpose of development➢⚫

Issues to be solved➢➢⚫

In the PCV environment which has low visibility and narrow spaces, remote
operation of the access device while avoiding contact and with a limited field ofvision, is a heavy burden on the operator. Hence remote operation technology toassist the operations is essential.
The existing technology allows efficient robotic operations by teaching the robot, but
it is expected that it will take time for the operator to operate the robot inside thePCV where the situation is unclear and keeps changing.

Suggests an arm direction to
avoid the situation

Approaches to development
➢
➢
➢

⚫

To develop a control method that can reduce the workload of the operator who
remotely operates the robot and improve the efficiency of the operation inenvironments with low visibility and narrows spaces.

Warns that the arm is likely tocome in contact with theenvironment

Conceptual study of the interference avoidance methods
Development of an interference avoidance program (manipulator control software)
Verification of effectiveness through element tests

Expected outcome➢➢

Avoidance of interference with devices operating inside the PCVReduction in the operator’s load by enabling operations in a short time

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project
No.81[1) ② Development of various technologies other than the interference removal technology]a.

Development of remote operation support method in environments with low visibility and narrow spaces⚫

Study flow

Organization of issues pertaining to remote operations

Conceptual study of interference avoidance methods

Development of an interference avoidance program (manipulator control software)

Verification of effectiveness through element tests (simulator)

Generation of environmental map, application to theinterference avoidance program

Results of this fiscal year

Application to mockup test robots

Evaluation of effectiveness and feasibility through elemental tests (Mockup test)

Identification of future issues and response policies©International Research Institute for Nuclear Decommissioning7. Implementation items of this project
No.82[1) ② Development of various technologies other than the interference removal technology]a.

Development of remote operation support method in environments with low visibility and narrow spaces⚫

Development schedule

Study items
Major milestones[Legend] Planned:

FY2019
4

5

6

7

8

9

MRI interim report▼

10

Actual:FY2020
11

12

1

2

3

4

MRI interim report
▼

5

6

7

8

9

10

11

MRI interim report
▼

12

1

2

3

MRI final report▼

1. Conceptual study
2. Desk verification(Develop the
manipulator control
software, and verify iton the simulator)
3. Mockup verification(Apply the manipulator
control software toexisting devices,develop an
environmental map, andverify it on a mockup
equipment)

4. Summary

Remarks©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ② Development of various technologies other than the interference removal technology]No.83
a.

Development of remote operation support method in environments with low visibility and narrow spaces⚫

Preconditions for study

Set “Work of removing interferences from inside PCV” as the task.
← To perform operations remotely while avoiding contact in environments with low visibility andnarrow spaces.
← For difficult environments where the situation is unclear and keeps changing as the work
progresses➢ The target of the study is Unit 2 and 3. Use an environmental model that simulates the actual
situation as much as possible.
➢ For instance, the robot model uses two manipulators + access rails← Follow the current study on the “Large-scale fuel debris retrieval” method← Aim for a mechanism that can be deployed to other multi-joint robots in the future➢

Model of environment inside thepedestal (CAD model)Interference removal work (CAD model)
Model of environment inside PCV(CAD model)©International Research Institute for Nuclear Decommissioning7. Implementation items of this project
No.84[1) ② Development of various technologies other than the interference removal technology]a.

Development of remote operation support method in environments with low visibility and narrow spaces⚫ Conceptual study (Control flow)➢ Manipulator control flow for which a conceptual study was conducted• Incorporate movements as needed to “automatically avoid interferences” based on robotic
movements through operator-controlled remote operations (manipulator tip operations)

Information on robot/
environment
・Camera image
・Robot model・Environmental map

Presented tothe operator[Instruction]
Position andbearing of the tip

Trajectory planning•
Calculate the shortest pathto the target position•
Determining collision
between robot model andenvironmental map →Calculate a robot position
that does not interfere withthe environment[Instruction]
Angularvelocity ofeach joint

Robot operations
・Driving of each joint・Movement[Instruction] Position and bearing of the tip
Operator©International Research Institute for Nuclear Decommissioning7. Implementation items of this project
No.85[1) ② Development of various technologies other than the interference removal technology]a.

Development of remote operation support method in environments with low visibility and narrow spaces⚫ Conceptual study (Work procedure)
1. Scanning the environment

2. Setting the target position

3. Trajectory planning

Removal locations

Illustration of operation

Structures
SensorHand-set

CutterArea through which theaccess rail passes(= Scope ofinterference removal)

The operator sets the cutter /hand-set icon to the removal
location nearest on his side (=Target position of the cutter /hand-set)Move the sensor toscan the environment[To be implemented in
FY2020]Accessrail

Accessrail

Access
rail
アクセス
レール

The trajectory up to the target
position is planned, and is
replayed to present the route ina semi-transparent form.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ② Development of various technologies other than the interference removal technology]
a.

No.86

Development of remote operation support method in environments with low visibility and narrow spaces⚫ Verification by means of a simulator
Initial position of manipulator

Trajectory to avoid simulated interference and reach target value
Trajectory of
the right arm

Right arm
Simulated interference

Trajectory of
the left arm

Left arm

Trajectory planning
Targetvalue ofright arm

Targetvalue ofleft arm©International Research Institute for Nuclear Decommissioning7. Implementation items of this project
No.87[1) ② Development of various technologies other than the interference removal technology]a.

Development of remote operation support method in environments with low visibility and narrow spaces⚫ Implementation items
FY2019FY2020

System configuration

Information on robot/
environment
・Camera image
・Robot model・Environmental map[Status of progress in FY2019]・ Interference avoidance
program created・ Parameter tuning underway

Presented
to theoperator[Instruction]
Position andbearing of
the tip

Trajectory planning• Determining collision
between robot model andenvironmental map→ Calculate a robot
position that does notinterfere with theenvironment• Calculate the shortest
path to the target position[Instruction]
Angularvelocity ofeach joint

Robot operations• Driving of eachjoint
• Movement[Instruction] Position and bearing of the tip

Simulator screen being verified

Implementation items

3D mapOperation and verification of the robot model on the simulator•
Derivation of optimal values for the frequency of trajectory
planning•
Determination of optimal values of various parameters

Dual-arm manipulator

Operation and verification of the actual machine in the mockup•
Generation of a mockup environmental map and confirming forcollision•
Confirming the validity of simulator verification by using an actualmachine©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ② Development of various technologies other than the interference removal
technology]
b.

No.88

Methods for preventing spread of contamination to the suppression chamber (S/C)⚫ Purpose of development➢ Further improvement of workability considering remote installation of weirs➢ Performance evaluation of remotely installed weirs⚫ Issues to be solved➢ Determining the specific work steps considering leakage in the weirs➢ Expanding the scope of remote operations and improving work efficiency⚫ Approaches to development➢ Study of more efficient work steps for remote installation➢ Planning of element tests for confirming the feasibility based on work steps➢ Verification of workability through element tests, and evaluation of the amount of leakage fromthe weirs➢ Conceptual study on the long-term integrity of weirs considering the risk of leakage after the
installation of weirs⚫ Expected outcome➢ More efficient work steps for remote installation➢ Workability verification results through element tests, and evaluation results of the amount ofleakage from weirs➢ Results of the conceptual study on the long-term integrity of weirs considering the risk of leakage
after the installation of weirs©International Research Institute for Nuclear Decommissioning7. Implementation items of this project
No.89[1) ② Development of various technologies other than the interference removal technology]b.

Methods for preventing spread of contamination to the suppression chamber (S/C)◼ Preconditions

Expanded
増設建屋
building✓ The PCV and expanded building are connected by a passage(access tunnel), and interferences outside the pedestal havebeen removed.
(An access route has been established inside the D/W)

Accessアクセス
tunnel
トンネル✓ The deposits around the locations, where the weirs will be
installed, have been removed.

Expanded building and access tunnel✓ The D/W water level is below the current Unit 2 level, that is,
below the lower end of the vent nozzle.
◼ Development objectives

PCV
shell
PCVシェル

Air lock
エアロック

0°

アクセス

270°

PLR motor

X-6

45°

✓ To prevent leakage to the S/C, etc., in combination withdrainage from pumps installed inside and outside the weirs.
PLR motor

機器ハッ

90°

Weirs
堰

D/W
intake pump
D/W取水ポンプ

Connecting
連通管 pipe

Drainage
pump
排水ポンプ

Drainage
pump
排水ポンプ

アイソメ図

Status of removal of interferences
outside the pedestal

D/W intake pump
D/W取水ポンプ

Pump installation position (Example)©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ② Development of various technologies other than the interference removal
technology]b.

Methods for preventing spread of contamination to the suppression chamber (S/C)◼ Study flow

No.90
: Reported this time

Policy on initiatives of the subsidy project in the present
term in response to the outcomes and issues of thesubsidy project in the previous term•••

Comparison between proposed methods for preventing the spread of contamination to the jet deflector→ Selection of the weir installation method
Study on specific water stoppage methods→ Identification of methods for constructing weirs by pouring mortar.
Verification of feasibility of the methods for constructing weirs→Verification of feasibility of constructing weirs by pouring mortar (however, issues in execution
identified)Comparative study of streamlined weir construction
methods• Comparative study of the two methods, sandbag method and method of carrying in segmented
formwork.

Detailed study of the adopted weir construction method• Conduct a detailed study of the method of carrying in segmented formwork.

Elemental test for weirs
Scope of implementation in thesubsidy project

Survey results of site conditions
Engineering work plan for actual operations©International Research Institute for Nuclear Decommissioning7. Implementation items of this project
No.91[1) ② Development of various technologies other than the interference removal technology]b.

Methods for preventing spread of contamination to the suppression chamber (S/C)⚫ Development schedule
Study items
Major milestones

FY 2019
4

5

6

7

8

9

10

11

Interim report▼

FY 2020
12

1

2

3

4

5

6

7

8

Interim report
▼

9

10

11

12

1

2

Interim report
▼

3

Final
report
▼1. Conceptual study
2. Test plan
3. Preparation for test, test
device prototyping

4. Element tests

Additional element tests
Preparation for additional tests

Additional tests

5. Summary

Remarks

Additional element tests to be conducted to verify the response policy for issues identified in the element tests.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project
No.92[1) ② Development of various technologies other than the interference removal technology]b.

Methods for preventing spread of contamination to the suppression chamber (S/C)
Inside PCV

Outside PCV◼ Results of studies conducted until the previous fiscal year

Mixer

• Comparison between proposed methods for preventing the
spread of contamination to the Jet Deflector→ Selection of the weir installation method• Study on specific water stoppage methods→ Identification of methods for constructing weirs by pouring
mortar• Verification of feasibility of the methods for constructing weirs→ Verification of feasibility of constructing weirs by pouring
mortar(Following issues were identified)

Hose

Pump

MortarFormwork(inside)Arm for weir
installation

Formwork(outside)
Formwork

Image of weirs forpreventing spread ofcontamination

Image of pouring mortar forconstructing weirs

Arm for weir
installation

Issues identified through the tests and corresponding response policies
ID

IssuesResponse policy

1

It was difficult to arrange the hose that poursmortar, at the designated position above theformwork.
(In the current setup, the hose needs to belowered straight down and the opening of thegrating has to be larger.)

Other easier methods of creating weirs shall be
studied, including whether or not it is possible touse standalone weirs (e.g., sandbags) to build
weirs, and element tests shall be conducted.
(Tests to verify the ability to install weirs by
remote operation devices, tests to verify theability of weirs to stop water, etc.)

2

After the weirs are installed, pumps and piping
may be needed to drain the water on the outside of
the weirs.

Together with the tests mentioned above, tests toverify the ability to install pumps and piping fordraining water on the outside of the weirs shall be
conducted.

3

There is a risk of leakage after weirs are installed.

In order to ensure long-term integrity, remote
operation repair methods shall be studied.

Formwork

Weir formwork
being installed
Hose

Arm
NozzleFormwork

Mortar being filled in a weir formwork

Mortar©International Research Institute for Nuclear Decommissioning7. Implementation items of this project
No.93[1) ② Development of various technologies other than the interference removal technology]b.

Methods for preventing spread of contamination to the suppression chamber (S/C)

A comparative study was conducted on remote operation methods for constructing weirs to prevent the spread ofcontamination to the S/C.
Conventional method[Streamlining plan 1]
Sand bag method[Streamlining plan 2]
Method of carrying in segmented
formwork

Formwork is installed and animpermeable material (mortar) is
injected into the formwork.
[The water-added mortar is pumped
with a hose]

Impervious bags are stacked up[Bags filled with dry mortar are stackedup]

Several foldable formworks (mesh type)are used. After carrying in theformworks, they are unfolded andinstalled, and an impermeable material(dry mortar) is injected into the formwork.

〇▲
Depends on the stacking method. The
bags must be stacked without gaps.

△
Complete impermeability is difficult, butdifference in water levels can beensured▲Positioning of the hose, installation of
the formworks▲
Implementation of the method ofstacking the bags without gaps△
Formwork arrangement and installation

Ability to slide along an
uneven surface〇△
Sliding is limited even if the bags arepressed against each other△
Depends on leakage of dry mortar fromthe formwork

Long-term integrity(Frequency ofmaintenance)〇△
Depends on the material of the bag〇

Comprehensive evaluation

▲

▲

△

Features

Impermeability

Evaluation

Remote operability

✓
✓

Sand
bags

Sand bagmethodFoldable formworks

Mesh bags[Legend] O: Good, △: Some issues present, ▲: Issues present (Issues are bigger as compared to △)✓Conventional method

The sand bag method and the method of carrying in segmented formwork were studied as the methods for solving theissue of remote operability, which was a major issue with the conventional method.
The issues with the sand bag method are how to remotely stack the bags so that there are no gaps and the ability toslide the bags along the uneven surface of the floor.
The issues with the method of carrying in segmented formwork (method of injecting dry mortar in the foldable
formworks (mesh)) are the feasibility of work under the constraint of the mortar curing time, and selection of optimum
roughness and material for the mesh bag.

Dry mortar (After wetting)

Method of carrying insegmented formwork©International Research Institute for Nuclear Decommissioning7. Implementation items of this project
No.94[1) ② Development of various technologies other than the interference removal technology]b.

Methods for preventing spread of contamination to the suppression chamber (S/C)◼ Possibility of adopting Streamlining Plan 1 and 2✓✓

Ability to slide along the floor (uneven surface):•
With the sand bag method, the sliding ability is limited even if the bags are pressed by an
external force. The methods used for improvement, such as reducing the impervious
material in the bags, do not satisfy the required functions, and hence this method cannot beused.
•
In the method of carrying in segmented formwork, by using mesh bags, the ability to slide
along uneven surfaces can be ensured.
Remote operability:••✓

With the sand bag method, it is difficult to achieve remote stacking without gaps, which isalmost the same as the conventional method (the major issue with the conventional method
is remote hose positioning).
In the method of carrying in segmented formwork, remote-controlled carrying in andinstallation is possible by adopting an X-frame structure that is easy to fold and unfold.

Gaps
ManualRemoteComparison of sand bag stacking methods
Angle(L 90 x 90 mm)

Pressing

Ability to slide along unevensurfaces (sand bag)

Impermeability:••

With the sand bag method, since impermeability depends on the stacking method, ensuring
impermeability is difficult.
For the method of carrying in segmented formwork, the preliminary tests have confirmed thatthe water level difference can be ensured (On a separate note, it is necessary to check thefeasibility of working under the constraint of mesh mortar curing time).

⇒ The method of carrying in segmented formwork shall be adopted.

Formwork: Folded
Water
level

Formwork: Afterunfolding

Water
level

Status of drainage after mortar curing©International Research Institute for Nuclear Decommissioning7. Implementation items of this project
No.95[1) ② Development of various technologies other than the interference removal technology]
b. Methods for preventing spread of contamination to the suppression chamber (S/C)
The construction of weirs by the method of carrying in segmented formwork, which was studied in detail, isillustrated below.
Mortar filled in the formwork

Outer periphery(Shell side)

300 mmWater level

Pedestal

Robotic hand-set

Frame 650
mm

PCV
shell

Sides: Vinyl

Weir 600 mm

Unfolded after
installation

Approx. 900
mm

Lowered froma floor above
the ground

Inner periphery(Pedestal side)

Approx. 3 m
Mesh

Approx. 500 mm
When lowered from a floor above thegroundX frameAfter unfolding the frame

Preliminary test*1

Using a mesh for the surface that touches
the wall surface (pedestal, shell, floor)
creates a structure that enables sliding overthe gaps and uneven surfaces.

Mesh

Frame being unfolded

Completely unfolded frame*1: Prototype created with vinyl on the entire surface, as the preliminary test was conducted for the
purpose of confirming the unfolded structure.

Validity of remote operation in the above structure to be confirmed by element tests##For details on the content of the element tests, refer to the table on Slide 97.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [1) ② Development of various technologies other thanNo.96
the interference removal technology]b.

Methods for preventing spread of contamination to the suppression chamber (S/C)◼

Extent of simulation in element tests✓✓

Element tests pertaining to the construction of weirs are conducted for solving the issues identified in the work steps shown on Slides 65 to 69.
Weir construction work involves underwater installation work during the actual operation, and hence the tests are conducted under water-filledconditions. In order to confirm impermeability on to the floor and wall surface, the simulation is conducted by installing weirs touching thepedestal outer wall and PCV inner wall.
This Jet Deflector is not simulated
Jet Deflector
Extent of simulation of basement
floor

Equipment
hatch

Pit

Weir installation
work area(basement floor)

Waterfilling

Water-filled weir
(Test equipment)
Jet
Deflector

Opening of thegrating(first floor)
Target to be verified
by the tests

First Floor

Basement floor©International Research Institute for Nuclear Decommissioning7. Implementation items of this project
No.97[1) ② Development of various technologies other than the interference removal technology]
b. Methods for preventing spread of contamination to the suppression chamber (S/C)◼ Element test items and criteria
ID

1

2

Test items

Applicable worksteps#1

Test outline and procedures•

7-1(1) Lift the formwork with the gantry-type lifting machine.
(2) Move the gantry-type lifting machine to a position above the existing
opening.
(3) Use the winch of the gantry-type lifting machine to lower theformwork to the basement floor.
(4) Use the flexible work arm (basement floor) to hold the formwork andhoist the winch of the gantry-type lifting machine.
(5) Use the flexible work arm to unfold the formwork.

7-2(1) Use the flexible work arm to position the hose (to inject dry mortar)
directly above the formwork.
(2) Move the hose with the flexible work arm to inject dry mortar into theformwork#2.
(2) Drain the water between the weir and the simple weir (for testing).
[Manual work](3) Measure the amount of leakage. [Manual work]•••

Carrying inandunfolding theformworks

Injection ofdry mortar

Criteria•••

Remarks

The gantry-type lifting machine can be positioned above theopening.
The formwork can be carried in from the opening (1200 x 600
mm).
The formwork can be unfolded.

Dry mortar can be injected into the formwork.
The mortar solidifies.
A water level difference can be created inside and outside theweirs by draining water.
There is no leakage from the weirs. (Goal)#2:
Mortar is
injected into thehose manually#1: Refer to the work steps on slides 65 to 69.

[Devices used for the element tests]

ピット
Pit

Lighting
Camera水張り
Waterfilling

Holding the jig to assist in thework

Flexible arm for remote
operations*

Transporting and installingthe jigs

Work monitoring

Gantry-type liftingmachineCameracart

Water-filled weir
水張り堰（試験用設備）(Test equipment)
Jet
Deflector
ジェットデフグレーチング開口
Opening of thegrating（1階）(first floor)
Targetto be verified
試験での確認対象
by the test

Schematic image of test
Formwork*There are several variations in flexible arms for remote operations. The above specification is an example of a crawler type flexible arm.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project
No.98[1) ② Development of various technologies other than the interference removal technology]b.

Methods for preventing spread of contamination to the suppression chamber (S/C)

Test procedures (Carrying in and unfolding the formworks) #Verified by a preliminary test

◼
ID

1

Test items

Carrying inand unfolding
the formwork

STEP

Applicable
work steps

7-1

Test outline and procedures(1) Lift the formwork with the gantry-type lifting machine.
(2) Move the gantry-type lifting machine to a position above the existing opening.
(3) Use the winch of the gantry-type lifting machine to lower the formwork to thebasement floor.
(4) Use the flexible work arm (basement floor) to hold the formwork and hoist the
winch of the gantry-type lifting machine.
(5) Use the flexible work arm to unfold the formwork

1. Lifting the formwork with the gantry-type lifting machine → 2. Moving
the gantry-type lifting machine above the opening
Gantry-type
lifting machine

Criteria•••

Remarks

The gantry-type lifting machine can be positioned above the opening.
The formwork can be carried in from the opening (1200 x 600 mm).
The formwork can be unfolded.

3. Lowering the formwork

Form
work
OpeningIllustrationofoperation

4. Transferring the formwork

Flexible workarm

Holding thehandleOpening

STEP

5. Unfolding the formwork
Holding the top of the frame andunfolding it sideways

Illustrationofoperation

Pedestal

PCV shell

Temporarily
placed on
the floor©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [1) ② Development of various technologies other thanNo.99
the interference removal technology]b.

◼

Methods for preventing spread of contamination to the suppression chamber (S/C)

Element test results (Carrying in and unfolding the formwork)
STEP

1. Lifting the formwork with the gantry-type lifting machine → 2.
Moving the gantry-type lifting machine above the opening

3. Lowering the formwork

4. Transferring the formwork

Flexible workarm

Formwork

Gantry-type
lifting machine

Illustrationofoperation

HangerFormwork

Holding thehandleOpening

STEP

5. Unfolding the formwork
Holding the top of the frame andunfolding it sideways
Temporarily
placed on
the floor

Unfolding on the outer
periphery

Unfolding on the innerperiphery

Pedestal

PCV shell

Illustrationofoperation

It was confirmed that the formwork can be carried in and unfolded.
©International Research Institute for Nuclear Decommissioning7. Implementation items of this projectNo.100[1) ② Development of various technologies other than the interference removal technology]b.

Methods for preventing spread of contamination to the suppression chamber (S/C)

Test procedures (Injection of dry mortar) #Verified by a preliminary test

◼
ID

2

Test items

Injection ofdry mortar

STEP

Applicable
work steps

Test outline and procedures

7-2(1) Use the flexible work arm to position the hose (to inject dry mortar)
directly above the formwork.
(2) Move the hose with the flexible work arm to inject dry mortar into theformwork#2.
(2) Drain the water between the weir and the simple weir (for testing).
[Manual work](3) Measure the amount of leakage. [Manual work]

•
•
•
•

Criteria

Remarks

Dry mortar can be injected into the formwork.
The mortar solidifies.
A water level difference can be created inside and outside theweirs by draining water.
There is no leakage from the weirs. (Goal)#2:
Mortar is
injected into
the hosemanually

1. Positioning the hose overthe formwork
Flexible work arm
Manual mortar
injection

HopperHose

Illustrationofoperation

Hose
Formwork

Inner periphery (pedestal
side)

Outer periphery (shell side)

On the grating©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ② Development of various technologies other than the interference removal technology] No.101b.

Methods for preventing spread of contamination to the suppression chamber (S/C)◼ Element test results (Injection of dry mortar)
1. Positioning of hose (innerperiphery)

STEP

2. Injecting mortar

Hose

Illustrationofoperation

Water
level
300
mm

Cementoutflow

CementMesh

Formwork

It was confirmed that hose positioning and mortar injection is possible, but there was a large
outflow of cement from the gaps between the formwork and the wall surface.
⇒ The test was suspended, the mesh sieves were opened, and the test was resumed after measures were taken for the gaps
with the wall surface.
Inner periphery (pedestal side)[Countermeasures] Attach a non-woven fabric on the outside of
the mesh (sieve size 1 mm).
 Close the gaps with the wall with a sponge.
(Either on the outer or inner periphery)

Outer periphery (shell side) Sponge Sponge Nonwoven fabricState of the weir on the inner peripheryState of the weir on the outer periphery©International Research Institute for Nuclear Decommissioning7. Implementation items of this projectNo.102[1) ② Development of various technologies other than the interference removal technology]
b. Methods for preventing spread of contamination to the suppression chamber (S/C)◼ Element test results (Injection of dry mortar)✓ After taking measures for gaps as described on the previous slide, mortar injection was continued andcompleted. (See the next slide for the injection status)✓ For the mortar injection spots, the following procedures were carried out while observing the accumulated
mortar.
No.

Injection position

Details

Standard

Outer peripherybottomMortar is injected so that the gaps with the wall surface are filled
and the bottom surface is covered.

Mortar is injected until the top of the mortar moundemerges in the air.

Centre bottom

Mortar is injected so that the mesh at the bottom is covered.

Mortar is injected until the top of the mortar moundemerges in the air.



Inner peripherybottomMortar is injected so that the gaps with the wall surface are filled
and the bottom surface is covered.

Mortar is injected until the top of the mortar moundemerges in the air.

Entire area in the
airMortar is injected throughout so that the height is 600 mm.

Mortar is injected while confirming the mortar
accumulation emerging in the air.

: Hose installation
direction

Pedestal③②①

Weir 600
mm④

Mortar injection position and accumulation status (estimated)©International Research Institute for Nuclear Decommissioning7. Implementation items of this projectNo.103[1) ② Development of various technologies other than the interference removal technology]
b. Methods for preventing spread of contamination to the suppression chamber (S/C)◼ Element test results (Injection of dry mortar)✓ After taking the measures described on the previous slide, mortar injection was continued and completed.
STEP

1. Hose positioning, mortar injection( Outer periphery)

2. Hose positioning, mortar injection( Centre part)

3. Hose positioning, mortar injection( Inner periphery)

Hose

Flexiblework arm

Illustrationofoperation

STEP

Formwork

4. Mortar injection monitoring status ( Inner periphery)

5. After completion of mortar injection

Hose

Illustrationofoperation

Formwork

Mortar (in air)Mortar (in air)©International Research Institute for Nuclear Decommissioning7. Implementation items of this projectNo.104[1) ② Development of various technologies other than the interference removal technology]
b. Methods for preventing spread of contamination to the suppression chamber (S/C)◼ Element test results (Injection of dry mortar): Status of verification of leakage✓ It was confirmed that after the curing of mortar, different water levels could be achieved inside and outside theweirs by draining the water, but there was a major leakage from the outer periphery of the weirs and thedifference in water levels was lost soon (approximately 10 minutes).

→ Specifically on the outer periphery, the gaps were filled up with sponge, but the amount of leakage was large
because the mesh bags did not swell enough to adhere firmly to the wall surface.

Water level
Approx. 300
mm

Drainage
Water level
Approx.
300 mmWater level
Near thefloor

Submersible
pump

Before draining water outside the weirMinorleakageMajorleakage

Immediately after draining water outside the weir

©International Research Institute for Nuclear Decommissioning7. Implementation items of this projectNo.105[1) ② Development of various technologies other than the interference removal technology]

b. Methods for preventing spread of contamination to the suppression chamber (S/C)◼ Summary of element test results
ID

1

Test items

Test outline and procedures

Carrying inandunfoldingtheformwork(1) Lift the formwork with the gantry-type lifting machine.
(2) Move the gantry-type lifting machine to a position above theexisting opening.
(3) Use the winch of the gantry-type lifting machine to lower theformwork to the basement floor.
(4) Use the flexible work arm (basement floor) to hold theformwork and hoist the winch of the gantry-type lifting machine.
(5) Use the flexible work arm to unfold the formwork.

Injection ofdry mortar(1) Use the flexible work arm to position the hose (to inject dry
mortar) directly above the formwork.
(2) Move the hose with the flexible work arm to inject dry mortar
into the formwork#1.
(2) Drain the water between the weir and the simple weir (fortesting). [Manual work](3) Measure the amount of leakage. [Manual work]

Test results
•
•

•
•

2

Carrying in formwork

Unfolding formwork•••

Remarks

It was confirmed that the gantry-type lifting machine can bepositioned above the opening.
It was confirmed that the formwork can be carried in from theopening (1200 x 600 mm) into the basement floor with thegantry-type lifting machine, and that the formwork can be heldwith the flexible work arm (basement floor).
It was confirmed that the formwork can be unfolded using theflexible work arm (basement floor).

O

It was confirmed that the hose can be positioned with theflexible work arm (basement floor), and that dry mortar can beinjected into the formwork.
It was confirmed that the mortar injected under water solidifies.
The mortar could be injected into the formwork, but a large
amount of cement flowed out from the mesh.
Different water levels could be achieved inside and outside
the weirs by draining the water, but there was a major leakage
from the weirs and the difference in water levels was lost soon.

Injecting dry mortar▲#1:
Mortar is
injected into
the hosemanually

Outward appearance of weirs

It was confirmed that the criteria pertaining to remote operability were fulfilled, but there
are issues in the structure and impermeability of the weirs.
©International Research Institute for Nuclear Decommissioning7. Implementation items of this projectNo.106[1) ② Development of various technologies other than the interference removal technology]b.

Methods for preventing spread of contamination to the suppression chamber (S/C)◼ Identified issues and policy of response
No.

Issue

Details

Response policy

1

Method of closing the
gap between theformwork and the wall

If the gap between the formwork and the wall
surface is large, it cannot be filled completely
and the mortar does not adhere to the wall, so
the water cannot be stopped.

The roughness of the mesh sieve shall be
studied again and additional element tests
shall be conducted.

2

Method of preventing theoutflow of cement fromthe mesh

Since the mesh sieve is coarse, the cement
component of mortar flows out.

Additional element tests shall be conductedin addition to the above improvements.

3

Method of repair when aleak occurs

It is necessary to conduct repairs if the amountof leakage increases over a long period of time,
but repairs are difficult with the current method.

Desk study shall be performed and thenadditional element tests shall be conducted.
This will also be studied as a backup method forcases when water cannot be stopped as in thiselement test.

4

Method of confirming thestatus of mortar injection

It is difficult to monitor the mortar injection statusfrom outside the formwork. (This test was
conducted by manually installing a camera near
the hose.)

The issue shall be handled through detailed
arrangement of cameras and lighting (during
actual operation).

Additional element tests to be conducted for No. 1 to 3.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this projectNo.107[1) ② Development of various technologies other than the interference removal technology]b.

Methods for preventing spread of contamination to the suppression chamber (S/C)◼ Additional element test plan✓✓✓

An element test is being planned to check the items listed in the below table.
The structure will be reviewed focusing on "adhesion" between the wall surface and formwork and "measures against outflow"from the mesh, the preliminary tests will be performed, and then the following element test will be conducted.
The element test will be performed on a scale that takes into account the effect along the length of the weirs.

Interface: Cement
flowing out fromthe meshMortarFormwork

MortarInterface: BagsBags that do not leakcementSponge Workform
intersection

Water
tank

Weir

Illustration of a preliminary test
Items to be confirmed by means of the additional element test
No.

Check items

Criteria

1

Adhesion to the wall
surface

If there is a gap between the formwork and the wall
surface, can the water be stopped if the mortar adheresto the wall surface?

2

Adhesion of theformwork intersection

At the intersection of the two formworks, can the water
be stopped if the weirs adhere firmly to each other?

3

Ability to slide along
an uneven floor
surface

Can the water be stopped even if the floor surface isuneven?

4

Repair method in theevent of leakage

If there is a leakage in the above interface, can it berepaired? (Assuming work from the upstream side.) Wall surface

 Floorunevenness

Illustration of a specimen for the additional element test
©International Research Institute for Nuclear Decommissioning7. Implementation items of this projectNo.108[1) ② Development of various technologies other than the interference removal technology]c. Conceptual study for establishing an access route for installation of cells and method of
reducing impact on R/B and other buildings⚫ Purpose of development➢ To refine the planned implementation method for verifying feasibility, by reconsidering the
proposed configuration and layout of the fuel debris retrieval equipment, in addition to studying
the method of implementation that resulted from the studies conducted until last year, along withstudying the progress of other projects, the safety aspect, constraints of the installation location,etc., in preparation for the designing stage.
⚫ Issues to be solved➢ It is necessary to reduce the weight of the equipment to satisfy the allowable load value for thefloor surface inside and outside the R/B.
➢ It is necessary to optimize the equipment layout to avoid interference with existing equipment
outside the R/B, which are difficult to remove.
⚫ Approaches to development➢ Detailing and reorganization of study conditions for reducing the equipment weight andoptimizing the installation area
➢ Study of ideas on how to reduce the equipment weight and optimize the area
➢ Study of the methods of installation, carrying in, and fitting of retrieval facilities➢ Study for determining the specifics of the retrieval method and evaluation of its feasibility⚫ Expected outcome➢ Refinement of the feasible configuration, layout plan, and retrieval method for the fuel debris
retrieval facilities©International Research Institute for Nuclear Decommissioning7. Implementation items of this projectNo.109[1) ② Development of various technologies other than the interference removal technology]c. Conceptual study for establishing an access route for installation of cells and method of
reducing impact on R/B and other buildings⚫ Study flow
Organization of study conditions from issues and problems based on the
results of the subsidy project in the previous term

Study of proposed countermeasures to solve issues and problems
Results of this fiscal year

Determining the specifics of the access route establishment plan

Evaluation of the feasibility of the access route establishment plan

Identification of future issues and response policies©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [1) ② Development of various technologies otherthan the interference removal technology]

No.110c. Conceptual study for establishing an access route for installation of cells and method of reducing
impact on R/B and other buildings◼ Development schedule

Study items
Major milestones

FY 2019
4

5

6

7

8

9

10

FY 2020
11

Interim report
▼

12

1

2

3

4

Interim report
▼

5

6

7

8

9

10

11

Interim report
▼

12

1

2

3

Final report▼

1. Organization of proposed
study conditions
2. Creation of access routeestablishment plans
3. Narrowing down of theaccess route establishmentplans
4. Study to determine thespecifics of the access routeestablishment plan

5. Evaluation of thefeasibility of the accessroute establishment plan
6. Summary of results

Remarks©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [1) ② Development of various technologies other than the
No.111interference removal technology]c. Conceptual study for establishing an access route for installation of cells and method of reducing
impact on R/B and other buildings
Pressure vessel[Main outcomes up to the previous year]
Opening for CRD
With respect to the method of collecting fuel debris by securing the boundary exchange workwith cells and accessing the debris from the side of the PCV with the robotic
X-6 penetration
arm, the configuration and layout plan of the fuel debris retrieval facility wasdevised with the below development concept.
Concepts of development (Aims of development)○ Consideration of access with the shortest distanceAirtight cell→Study of methods to quickly access fuel debris based on the idea ofcarrying in large equipment into the PCV with minimal interference removal○ Adoption of the cell structure cultivated at the Rokkasho Reprocessing
Manipulator
Plant (RRP)
Rail→Based on the technology that has a proven track record in reducing worker
exposure, a highly reliable system configuration was created adding the latest
findings○ Utilization of remotely operated manipulators for which knowledge isavailable in the field of maintenance and reprocessing of light water reactors→Study of methods taking into account the special environment based onrealistic work / maintenance plan considering actual results

Equipment hatch cell
ceiling crane
Carrying-in/out cell

Crane maintenance module

Primary
containment vessel
Robotic arm, etc.
(debris outside the pedestal)
Airtight cell

Equipment hatch

Pedestal

Robotic arm (debris inside the pedestal)

Fuel debris

Conceptual diagram of access route

Reactor Building (RB)

Reactor Pressure Vessel

Expanded building
PCV opening(Position of X-6 penetration)

Equipment temporary
placement module
Maintenance Fuel debris
cell
retrieval cell

Shielding door
Cell adaptor

RB wall opening
Cell installation deck
Mini cask cell
Canister cell

Access device
carrying-in/out port
Radioactive waste carrying-in/out port
Cask carrying-in/out port

Canister carrying in station

Cask cell(Cask carrying-out inspection area)
Radioactive waste carrying-out cell

Facility configuration plan up to the previous fiscal year

Facility layout plan up to the previous fiscal year
©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [1) ② Development of various technologies other than
No.112the interference removal technology]c. Conceptual study for establishing an access route for installation of cells and method of reducing
impact on R/B and other buildings: Scope of study in FY2019-2020

Overview of work steps(1/2)

Installation of expandedbuilding

Work step

Reactor Pressure Vessel

Reactor Building (RB)ExpandedbuildingGreen house

BSW opening
Temporary cell for BSW opening work

Scope of interference removal

Wall reinforcement

Expanded building

PCV opening(Position of X-6
penetration)

Removal of interferences
inside R/B

Work to open R/B wall

Illustrationof steps

Core boring device

Size of opening: W
2.5 m x H 3.5 m

Approx. 2 m

RB wall opening

PCV 38 mm

Vinyl covering
Covering
sheet

Shield (Tungsten jacket + Partition)
ShieldRemarks•

Work step

A brief study of scale wasconducted in FY2017-2018•

A conceptual study was
conducted in FY2017-2018

Welding of cell adapter

Installation of shielding door

• The opening method developedby Toshiba in FY2017-2018 isplanned to be applied.

Installation of fuel debris
retrieval cell

Installation of external R/B
cellRadioactive waste
carrying-out cell

Fuel debris retrieval cell

Shielding door

Fillet welding

Gap between biological shielding
and PCV: Approx. 40 mm

Illustrationof steps
Covering
sheet
ShieldRemarks•

Element technology wasdeveloped in FY2016-2017•

A conceptual study was
conducted in FY2017-2018•

A conceptual study was
conducted in FY2017-2018•

A conceptual study was
conducted in FY2017-2018©International Research Institute for Nuclear Decommissioning7. Implementation items of this projectNo.113[1) ② Development of various technologies other than the interference removal technology]c. Conceptual study for establishing an access route for installation of cells and method of reducing
impact on R/B and other buildings: Scope of study in FY2019-2020

Overview of work steps(2/2)
Work
step

Removal of interferences
inside PCV

PCV wall opening

Fuel debris retrieval

Carrying out fuel debris

Illustrationof steps

Remarks•

A conceptual study was
conducted in FY2017-2018•

A conceptual study was
conducted in FY2017-2018• A conceptual study was
conducted in FY2017-2018• A conceptual study was
conducted in FY2017-2018

To solve the issue that the R/B floor load will exceed the allowable value due to the weight of the fuel debris retrieval cell to beinstalled in the R/B, in this subsidy project, studies will be conducted to review the access method, the equipment layout inside thefuel debris retrieval cell, and the cell fixing method with the intention of reducing the size and weight of the fuel debris retrieval cell.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this projectNo.114[1) ② Development of various technologies other than the interference removal technology]c. Conceptual study for establishing an access route for installation of cells and method of reducing
impact on R/B and other buildings
Remote controlled MSM

Vacuum cleaner for decontamination

Jib crane

Platform to temporarily
place LLW basket (Platform
to temporarily place unitcans)

Travelling crane[Issues / problems based on the results of the previous fiscal year]① Increase in the size of the cell due to the relatively large access device

Hydraulic jack forrestoration of tilted accessrails (Mobile)

Remote controlled MSM
Cell monitoring camera

Double doors(for canisters)② Increase in the weight of the cell due to the increase in the number of shields
to be installed

Remote controlled MSM
Rack to temporarily place end tools

Remote controlled MSM

Remote controlled MSM
Cell monitoring camera

Shielding door

③ By conducting a study on a suspension bridge-shaped cell installation
platform and bringing the cell load inside the R/B out of the R/B, the R/B floorload can be reduced (zero by offsetting)Sliding table tosupply end tools④ However, a relatively heavy load is applied to the columns and anchor fixing
parts of the suspension bridge-shaped cell installation platform,
corresponding to the cell load inside the R/B.

Unit can measurement chamber[Study policy and goals]• Aim to reduce the R/B floor load by reducing the cell's weightand distributing the load by the cell's weight.
• Prevent local concentration of load outside the R/B byeliminating the suspension bridge-shaped cell installation
platform

Radioactive wastetransport passage

Access rail
Access rail emergency
recovery winch⑤ There are existing structures such as underground tanks near the R/B, and astudy found that it would be difficult to construct the strong foundationnecessary to support the load applied to the columns and anchor fixing parts

Load reduction measures other than suspension bridgeshaped cell installation platform were required for R/B firstfloor.

Approx. 5,400

Double doors(for radioactive waste
storage containers)

Access device
power supply
rail
Robotic arm
Access device
traveling rail

Structural drawing of the fuel debris retrieval cell up
to the previous fiscal year

PCV frame
PCV frame

RB wall opening sealing plate

RB outer
periphery wall

Shielding door
Fuel debris
retrieval cell

Deck plate
mounting platform

Equipment temporary
placement module

Cell transport rail

Hydraulic unit

Anchor cableStand jack

Crane maintenance module
Anchor cable

Deck plate

RB wall opening
sealing plate
Mini cask cell

Canister cell
Cask cellRadioactive waste
carrying-out cell

Supporting column
Anchor plate

Supporting column
Deck plate

Hydraulic unit
Stand jackSupporting column

Supporting column

Schematic diagram of suspension bridge-shaped cell
installation platform up to the previous fiscal year
©International Research Institute for Nuclear Decommissioning7. Implementation items of this projectNo.115[1) ② Development of various technologies other than the interference removal technology]c. Conceptual study for establishing an access route for installation of cells and method of reducing
impact on R/B and other buildings◼ Pre-conditions
•

•
•
•

The ground has been levelled in the area for the cells andexpanded building outside R/B and the cells and the
expanded building can be set up.
An opening is provided on the side wall of the R/B, andcells can be carried into the R/B and installed.

BSW opening
RB outer periphery wall
PCV frame

Shielding door
Fuel debris
retrieval cell

Interferences in the specified area of the R/B have been
removed, and the cell installation area has been secured.

Equipment temporary
placement module

An opening is provided in BSW, and the area forconnecting cells to PCV is secured.

Crane maintenance module

Mini cask cell

Canister cell
Cask cellRadioactive waste
container cell

◼ Development objectives• Prevent local concentration of load outside the R/B byeliminating the suspension bridge-shaped cell installation
platform.

RB wall opening
sealing plate

R/B side wall opening

Anchor cable• Aim to reduce the R/B floor load by reducing the weightof the cells and distributing the load resulting from theweight of the cells.

Cell installation area inside R/B

Supporting column

Deck plate

Hydraulic unit

Stand jackSupporting column

Area outside RB for
constructing the expandedbuilding and installing cells©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ② Development of various technologies other than the interference removal
technology]

No.116c. Conceptual study for establishing an access route for installation of cells and method of reducing
impact on R/B and other buildings

Study items for reducing the size and weight of the cell in R/B (fuel debris retrieval cell)

PurposeTo reduce the weight ofthe cellTo distribute load
resulting from theweight of the cell

Study items

Overview① Changing the accessmethodEliminate the access rail method and use a fixed rail
method to reduce the cell height.

② Reducing the height of thecellReduce the cell height by reviewing the configurationand layout of the equipment in the cell.

③ Studying the shield
mounting range of the cell

Reduce the weight of the shield by limiting the fueldebris transportation system area.

④ Reducing the shieldthickness

Reduce the weight of the shield by reducing the shieldthickness within a reasonable range.

⑤ Distributing the floor load
by using an installation
platform

Distribute the floor load resulting from the cell using
the platform provided at the bottom of the cell.

• The results of studies conducted on the cells outside the R/B in the subsidy projects up to the previous fiscal year indicate
no major problem. Therefore, in order to reduce the load due to the installation of cells inside and outside the R/B, a study
will be conducted focusing on reducing the size and weight of the fuel debris retrieval cells installed inside the R/B.
©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [1) ② Development of various technologies other than
No.117the interference removal technology]c. Conceptual study for establishing an access route for installation of cells and method of reducing
impact on R/B and other buildings

Flow of future studies

 Study on downsizing of the access device
 Study on downsizing of cells

Equipment temporary
placement module

Equipment carrying-in/out
container yard
Equipment transport
container

Crane maintenance module

Decontamination module

RHR piping support

Debris retrieval cell

Debris retrieval reserve cellRadioactive waste container cell
Mini cask cell

 Study on the mounting range of the shield Study on the shield thickness

Canister carrying-in station
Canister cell
LLW transport container

Cask container yard

Cask cell

Cask container

LLW transport container yard

#Comfort: The cells outside the R/B andthe entire yard will be covered by tents.

Schematic image of facility configuration review

 Study on the distribution of floor load by using an
installation platform

Mini cask cell

Evaluation of cell size, weight and estimated floor load
Canister carrying-in station

RHR piping support
Canister cellRadioactive waste container cell

Debris retrieval cell

LLW transport container

Debris retrieval reserve cell
Maintenance cell

Evaluation of the effectiveness and feasibility of the review
plan from the viewpoint of fuel debris retrieval work, facility
operations, facility layout, installation work, maintenance etc.

Cask cell

Crane maintenance module

Cask container yard

Decontamination module

Cask container

Equipment temporary placement module

LLW transport container yard

Access device carrying-in/out port
Equipment transport container#Comfort: The cells outside the R/B and theentire yard will be covered by tents.

Equipment carrying-in/out container yard

Schematic image of facility layout review©International Research Institute for Nuclear Decommissioning7. Implementation items of this projectNo.118[1) ② Development of various technologies other than the interference removal technology]c. Conceptual study for establishing an access route for installation of cells and method of reducing
impact on R/B and other buildings① Changing the access method

Access rail

Robotic arm

Access device-related issues up to the previous fiscal year•
The method involves tilting and extending the access rail to secure
an access route, and a space is required vertically in the cell for
tilting.
•

A relatively large space is required in the cell to move a relatively
large access device in and out of the PCV.

Study policy•
Reducing the height-wise space of the cell by eliminating theaccess rail and adopting the fixed rail method•

Reducing the space required for carrying the access device in/out
by stopping the use of the largest access rail (on contraction: 8.7 m
x 1.9 m x 2.5 m; 24 ton)•

Determining the specifics of the structure, installation method,
robotic arm, and unit can transportation methods for a fixed rail as
an alternative to the access rail, and verifying their feasibility.

In order to tilt, the end part of
the arm is raised, so space isrequired in the verticaldirection
Schematic diagram of the access device up to theprevious fiscal year

Future response•

The specifics of the study item will be verified in the next
fiscal year.

Conceptual diagram of the fixed rail method©International Research Institute for Nuclear Decommissioning7. Implementation items of this projectNo.119[1) ② Development of various technologies other than the interference removal technology]c. Conceptual study for establishing an access route for installation of cells and method of reducing
impact on R/B and other buildings② Reducing the height of the cell

Cell-related issues up to the previous fiscal year•••

LiningLining

The weight of shields including structural materials (360 tons)
accounts for most of the weight of the fuel debris retrieval cell (382
tons).
Shields are installed on the sides of the cell to protect workers, and Shielding
material
reducing the cell height is effective for reducing weight.
In order to collect the unit cans from the access rail, it wasnecessary to lift the unit cans from the top of the inclined accessrail, and it was necessary to provide a travelling space for the craneon the top.

Shielding material(Iron plate)

Shielding material (PE)Shielding material filling up
the pillar part

Outer wall plate

Outer wall of cell

Column (H steel)

LiningShielding material (PE)Shielding material fillingup the pillar part

Outer wall plate

Supporting beam
(H steel)

Structural drawing of the shielding material in a fuel debris
retrieval cell up to the previous fiscal year

Study policy•

Due to the elimination of the access rail, it is necessary to review themethod of collecting and transporting the unit cans from the access
rail.

•

In order to reduce the cell height, review the method of collecting theunit cans using a ceiling-mounted traveling crane.

Future response•
The specifics of the study item will be verified in thenext fiscal year.

Access rail

Unit can(Delivery position)

Unit can transport cart

Structural drawing of the unit can transport cart up to
the previous fiscal year
©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ② Development of various technologies other than the interference removal technology]

No.120c. Conceptual study for establishing an access route for installation of cells and method of reducing impact on R/B
and other buildings③ Studying the shield mounting range of the cell

Sliding table tosupply end tools

Platform to temporarily
place unit cans

Jib crane

Travelling craneShield-related issues up to the previous fiscal year•
It is necessary to install a thick shield suitable to the source
strength for fuel debris.
•
Inside the cell, fuel debris was being handled in unit cans without
the shielding effect.
Shielding door•
In the fuel debris retrieval cell, it was necessary to secure a
relatively large space for handling a large access device.
•
It was also necessary to install a thick shield suitable to the fueldebris over the entire side surface of the cell.
•
As a result, the shield mounting area became larger and the cellweight increased.
Receipt of unit cans

Review the unit cantransportation method to
reduce the required space

Unit can measurement station
Double doors (for canisters)

Rack to temporarily place end tools
Mini cask cell

Double doors(for radioactive waste
storage containers)

Reduce the thickness of theshields mounted in parts otherthan the transportation area ofunit cans containing fueldebrisfrom the access rail

Measurement of unit cans

Store the unit can ina canister

Study policy•
Reduce the handling area of fuel debris and the mounting area of the
thick shield suitable to the fuel debris by reviewing the unit canhandling method.

Reduce or separate thetransportation area of unitcans containing fuel debris to
reduce the mounting range ofthe shields suitable to the fueldebrisAccess device
traveling rail

Conceptual diagram on the study of the shield mounting
range

Assumed unit can specifications
•
•
•
•
•

Future response•
The specifics of the study item will be verified in the next
fiscal year.

External dimensions: Ф210 mm x 200
mm
Dry weight: 10 kg
Maximum debris storage capacity: 50 kg
Average debris storage capacity: 15 kg
Maximum gross weight: 60 kg

Unit can©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ② Development of various technologies other than the interference removal technology]

No.121c. Conceptual study for establishing an access route for installation of cells and method of reducing impact on R/B
and other buildings④ Reducing the shield thickness

Source conditions inside the PCVShield thickness-related issues up to the previous fiscal year•
The shield thickness was being specified conservatively.
•
The shield thickness can be reduced to some extent byreviewing the conditions within a reasonable range
Response•
Review the source conditions within a reasonable range•
Review the criteria for dose rate around the cell within areasonable range•
Review the considered source study conditions taking into
account the operating conditions

Source conditions of unit cans

Primary study results•
Required shield thickness (2 cm iron + 5 cm PE + 21 cm iron ⇒iron 2 cm + PE 5 cm + iron 18.5 cm)•
Shield weight reduced by approximately 12% in the same
mounting area
Future response•
Reduce the cell weight further in combination with other items.

Analysis result in normal times©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ② Development of various technologies other than the interference removal technology]

No.122c. Conceptual study for establishing an access route for installation of cells and method of
reducing impact on R/B and other buildings⑤ Distributing the floor load by using an installation platform Fuel debris retrieval cell
Hydraulic drive cell
Issues related to the cell installation platform up to theprevious fiscal year•
Allowable load value on R/B first floor•
The design for R/B does not take into account the installationof heavy objects such as cells that were considered up to theprevious fiscal year.
•
The methods like the cell mounting platform, used up to theprevious fiscal year, is advantageous from the viewpoint of
reducing the floor load.
•
However, dealing with earthquake load and displacement will
become complicated and difficult.
Study policy•
Study the method of installing and fixing the cell to the R/Bfloor together with distributing the cell load properly byproviding a platform at the bottom of the cell.

•

Although an orthodox method that does not draw attention, it isa realistic method that makes it relatively easy to deal with loadand displacement during an earthquake

transport cart

Anchor cable

Deck plate
mounting platform
Deck plate
Cell transport rail

Schematic diagram of the cell mounting platform up to
the previous fiscal year

Cell

Cell platform

R/B floor

Future response•
Reduce the floor load in combination with other items.
Schematic diagram of a cell platform©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [1) ② Development of various technologies otherthan the interference removal technology]

Exhaust pump
Filterd. Transfer method for unit cans from R/B
⚫
⚫

⚫

⚫

Basket liftingmechanism

Purpose of development➢ Develop a method for transporting unit cans to a building away from the R/B.

BasketIssues to be solved➢ If the treatment for carrying-out fuel debris storage containers, such as sealing, isperformed in a building away from the R/B, a confinement device for transportingthe containers to the building is required.
➢ Conceptual study on the confinement device and verification of its feasibility isnecessary.
➢ The concept of a building that performs confinement treatment on fuel debris
storage containers has not been studied.
Approaches to development➢ Conduct a study on transportation of unit cans to a building away from the R/B.
- Primary evaluation of the transportation time, formulation of confinement device
specifications- Identification of technology development items, assessment of feasibility
through element tests and analysis (Hydrogen generation, cooling, shielding,etc.)➢ Study the concept and requirement specifications for a building that will perform
treatment required for fuel debris storage presuming the application of aconfinement device.
➢ Study the required number of treatment systems taking into account themaintenance, normal inspection, failure, etc. of confinement device and building
equipment.
Expected outcome➢ Establishment of a device to transport fuel debris in unit cans is expected.

No.123

ShieldShield shutter
Basket lift door

Basket lift hatch

Concept of a transportation device

Transfer tool withshield lining

Dispensing cell
Receiving/ (temporary storage)measurement cell
Temporary storage container

DryingdeviceWeight measurement device

Surface nuclide composition analyzer
Dose rate measurement device

Treatment building to store fuel debris (Schematic image)©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ② Development of various technologies other than the interference removal technology]

No.124d. Transfer method for unit cans from R/B■ Study flow

Problems and issues in existing design,
aiming at improvements

•
•
•
•

Work procedure in existing study
Study of building layout
Solutions to issues
Ideas for fuel debris transportation

Conceptual study of the design to be
improved•

Conceptual study of facilities combined with fuel debris retrieval method
and layout
Combination of transportation device and building boundary at each step
Basic functional requirements
Safety functions required for the transportation of unit cans storing fueldebrisBasic design conditions based on safety requirements
Basic design conditions based on handling requirements
Defining design specifications and issues

•
•
•
•
•
•

Response to technical issues•••

Confinement mechanism for the door part
Prevention of hydrogen explosion
Evaluation of handling work steps, facility layout, and work
schedule

Summary (Outcomes of FY2019 / Plan for FY2020)©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ② Development of various technologies other than the interference removal technology]
d. Transfer method for unit cans from R/B

No.125Development schedule
Study items

Major milestones

FY 2019
4

5

6

7

8

9

10

11

Interim report▼

FY 2020
12

1

2

3

Interimreport
▼

4

5

6

7

8

9

10

11

Interimreport
▼

12

1

2

3

Final
report▼1. Formulation ofrequirement specifications,conceptual design
2. Element test plan
Element evaluation device
specifications
4. Preparation for elementtests

5. Element tests
6. Summary

Remarks©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ② Development of various technologies other than the interference removal technology]
d. Transfer method for unit cans from R/B

No.126■ Problems and issues in existing design, aim of improvements (Work procedure in existing study)•••

Dry the unit can, then store it in a canister, and store the canister in the storage building
When transporting from the R/B, store in an airtight/ transportation/ shielding container
Carry out the work before transportation in the expanded building to be constructed adjacent to the R/B or in a container in R/B
R/B

STEP 1Retrieve fueldebrisExpanded building (or container)

STEP 2
Store fueldebris in unit
can

STEP 3

STEP 4

Dry the unit
canStore unit canin canister

STEP 5Store thecanister in atransportation/ shielding
container

Storage building

Transportation

STEP 6

STEP 7
Receive thetransportation
container/
shieldingcontainer

Transport thetransportation/ shielding
container

STEP 8Remove thecanister

Fuel debris
Unit can⚫

Canister
Problems in the existing procedure•
The expanded building is large in size and affects the site layout•
Top access container also requires a 3000 to 5000 ton shield•
Expanded building (container) is required for each top/ side access in each
Unit

Container(Top access)

R/B

Transportation container/ Shielding container
Carrying-in
cellRadioactive waste
carrying-out cellRadioactive waste
carrying-out port
Fuel debris
carrying-out port

Expandedbuilding(Side access)1F on-sitetransportation

Storagebuilding

Fuel debris
retrieval cell
Maintenancecell

Access tunnel

Fuel debris
carrying-out cell

Example of expanded building©International Research Institute for Nuclear Decommissioning7. Implementation items of this projectNo.127[1) ② Development of various technologies other than the interference removal technology]
d. Transfer method for unit cans from R/B■ Problems and issues in existing design, aim of improvements (Study of building layout)• The effect of the establishment of an expanded building in 1F-1 to 3 was studied.
• If an expanded building is placed in the mountain side area of each Unit, a large area will be occupied by the expandedbuilding, and problems such as inability to carry out other work (for example, installation of a crane for carrying-in/out large
objects) are assumed.

Sea side

Unit 1
T/B

Approx. 44 m

Expanded building

Unit 2
R/B

Approx. 37 m

Road adjacent to thebuilding

Unit 1
R/B
Maintenance
cell

Unit 3
T/B

Unit 2
T/B

Approx. 29 m

Unit 3
R/B

Approx. 21m
Approx. 54 m

MaintenancecellExpanded building

Road adjacent to the building

MaintenancecellExpanded building
Approx. 30 m
Road adjacent to the building

Elevation
Elevation©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[1) ② Development of various technologies other than the interference removal technology]

No.128d. Transfer method for unit cans from R/B■ Problems and issues in existing design, aim of improvements (Solutions to issues)•••

Establish the expanded building away from the R/B and transport on-site.
The established expanded building shall be a common facility for Units, top/ side access.
In this process, transportation will take place without draining the fuel debris and storing it in canisters.

R/B

STEP 1Retrieve fueldebrisExpanded building (or container)

Transportation

STEP 2

STEP 3

STEP 5Store fueldebris in unit
can

STEP4

Store unit can in
a transportation
deviceTransport onsite

Receive thetransportation
deviceSTEP 7
STEP 6
Remove theunit canPre-storage
treatmentsuch as drying
and sorting

Storage building

STEP 8STEP 9Store the unit
can in acanister

To storage steps(same
operations
hereafter)

Fuel debris
Unit canUnit 1 R/B

Transportation device Transportation
deviceSide accessTop access

Unit 2 R/B

Has the function of expanded building as a common facility between UnitsSide accessTop access

Unit 3 R/B

Top access1F on-site transportation

Expandedbuilding

Storagebuilding

Side access©International Research Institute for Nuclear Decommissioning7. Implementation items of this projectNo.129[1) ② Development of various technologies other than the interference removal technology]
d. Transfer method for unit cans from R/B■ Problems and issues in existing design, aim of improvements (Ideas for fuel debris transportation)

・For transportation to the remotely established expanded building, Idea 2 which links the site with a controlled area (trench, etc.), and Idea 3 which transports the cans on site shall beconsidered.
・ Idea 3 has great merits, but it involves development elements, so a conceptual study will be conducted in this project.

Schematic image

Idea

Fuel debris
transfer flow line

R/B

1

Overview

Merit/ Demerit

Development elements[Existing study idea]
Dry the fuel debris and perform
various container storage operationsin the expanded building adjacent to
the R/B.

[Merits]
The drained fuel debris can be dried
without being transported on site.
[Demerits]
Since an expanded building will be
provided for each Unit, there are major
restrictions on the site layout andcontainers.

There are no development
elements in transportation
because the controlled areas
are connected.

[Idea of a tunnel]
Provide a controlled area (tunnel)
between the R/B and the expandedbuilding to transport the unit can as
is without drying.

[Merits]
Fuel debris can be transported without
providing an expanded building aroundthe R/B.
[Demerits]
A study on the possibility of installing atransportation tunnel (radiation controlledarea) for each unit must be conducted.

There are no development
elements in transportation
because the controlled areas
are connected.

Expanded building

Idea
2

R/B
Expanded

Access tunnelbuilding(radiation controlled area)

Idea
3

R/B

On-sitetransfer

Expandedbuilding[Idea of on-site transportation]★ Fuel debris transportation system
Transport the unit cans on site fromthe R/B before the drying process.

[Merits]
Minimizes the amount of facilities added
to the R/B.
Great freedom for selecting the
installation location of the expandedbuilding.

Development of a device for onsite transportation of unit cansthat have not been dried.
Conduct a study on a transportation
device that can transport unit cans on the
1F site and a transportation system tooperate the same.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this projectNo.130[1) ② Development of various technologies other than the interference removal technology]
d. Transfer method for unit cans from R/B■ Organization of preconditions for the design to be improved (Conceptual study of facilities combined with fuel debris retrieval method and layout)⚫
⚫

To ease the load conditions on the building and container,
minimize the shielding on the refueling floor installation cell
Carry out the unit cans outside the building without using canisters
and transport them on site

Transportation device
Body of transportation
device (with shield)

Secondary boundary
Primary boundary
C4M

C4L

Platform Test before
decontamination andcarrying out

Platform

Unit can /fuel debris

Storing unit cans on a transportation device⚫ Install a work platform with shield at the top of the PCV⚫ Connect the transportation device to the platform and insert
the unit can Store the unit canin the transportation
deviceWork platform with theshielding function

On-site transportation

 Store the fueldebris in the unitcan

C4H
 Load onto
a vehicle tocarry out

⚫ Double the confinement by storing
the transportation device in acontainer⚫ Monitor the hydrogen concentration
inside the transportation device⚫ If it exceeds the standard value,
inject air or nitrogen to dilute it⚫ Diluted gas is stored in the exhaustgas tank

Inject air or
inert gas

[Hydrogen treatment
mechanism]
Gas tank for dilution
or compressor

Hydrogentreatment
container

Collectionof exhaustgas[Hydrogen treatment
mechanism]
Pump[Hydrogen treatment
mechanism]
Exhaust gas tankOn-site transportation vehicle©International Research Institute for Nuclear DecommissioningNo.131
7. Implementation items of this project[1) ② Development of various technologies other than the interference removal technology]
d. Transfer method for unit cans from R/B■ Organization of preconditions for the design to be improved (Alignment of the transportation device and the building
boundaries during each step)
Building (Secondary boundary)

Cell (Primary boundary)

Building (Secondary boundary)Hydrogen treatment container(Secondary boundary)

Boundary of thetransportation
device[Hydrogen Treatment
Mechanism] Gas tankfor dilution orCompressor[Hydrogen
Treatment
Mechanism] Pump

Hydrogentreatment
container

Exhaust gas
collection
Hydrogen treatment container[Hydrogen
Treatment
Mechanism]
Exhaust gas tankBoundary of thetransportation
deviceBoundary of thetransportation
deviceOn-site transport vehicle[At the time of transportation within
the building cell (Primary boundary)][Outside the building cell
(Secondary boundary)][At the time of transportation
within the site]

Confinement to be executed asfollows:• Main body of the transportation
device• Cell (Primary boundary of thebuilding)• Building (Secondary boundary)Confinement to be executed
as follows:• Main body of thetransportation device• Building (Secondary
boundary)

Confinement to be executed asfollows:• Main body of thetransportation device• Container for transportation©International Research Institute for Nuclear DecommissioningNo.132
7. Implementation items of this project[1) ② Development of various technologies other than the interference removal technology]
d. Transfer method for unit cans from R/B■ Organization of preconditions for the design to be improved (Basic functional requirements)

Table. Specifications of the unit can to
be transported (Set based on
information from the canister project)⚫ Work requirements
・ Unit cans containing fuel debris will be loaded and then transported on-site from the R/B to
Type 1
Type 2the distantly located expanded building where the unit cans will be lifted out.
⚫ Transportation targets
External
dimensions
Φ206×h403
Φ386×h403
・Capacity (Number of unit cans stored): 1 canmm
mm
・Unit can: Type 1 mentioned on the right is the basic type.
UO2 weight
41.5 kg
150 kg(Maximum), #1(The feasibility of Type 2 will also be studied from the view point of enhancement.)
Moisture
30%
30%⚫ Environmental conditions (Set values)
content %・Temperature: 0 to 40℃
Amount of
0.5 L/hr
1.7 L/hrhydrogen・Humidity: 100% or less (with condensation)
generation, #1
・Environmental dose: 3000 Gy/h or less (Same as the environment inside the reactor)）
Amount of
0.25 L/hr
0.85 L/hr⚫ Unit can storage function
oxygen・It should be possible to internally store a unit can by lifting it from the reactor (reactor bottom). generationof fuelParticle diameter of 0.1 mm or・Lifting height: About 14m (The lifting is assumed to be from the height of the core support plate Sizedebrismore
up to the PCV flange.)#1: In this project, the⚫ Transport specifications
specifications have been
provided for 1 unit can・Number of cans transported in a day: 10 cans (24 hour work)
although in the canister project,
・Scope of handling: Buildings
the specifications have been
Limited to inside the R/B and the Expanded building and within the site of the power station.
formulated for collectivelytransporting 2 unit cans
(Does not include any handling outside the site.)・Access to workers (Presence during work)
Workers will be present during the inspection prior to carrying out the cans from the primary boundary in
C4L.
The design should be such that the handling work, verification of lid closure, and decontamination work in C4M
can be carried out with remote operated devices.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this projectNo.133[1) ② Development of various technologies other than the interference removal technology]
d. Transfer method for unit cans from R/B■ Conceptual study of the design to be improved (Safety functions required for the transportation of unit cans storing fuel debris)
Safety functions

Contents

Leakage prevention

Prevention of exposure of workers or public to radioactive materials due to leakage from the transportation
deviceShielding

Prevention of exposure of workers or public to radiation

Subcriticality maintenance

Maintenance of subcriticalities

Cooling or heat removal
Functions necessary to maintain thesafety functions

Prevention of safety inhibitors due to a rise in the temperature of the fuel debris
Appropriate structural strength so as to maintain the necessary safety functions such as leakage prevention
and subcriticality maintenance, etc. considering the handling of the transportation device
Prevention of explosion of the generated hydrogen
Maintenance of confinement function etc. to deal with aging degradation such as corrosion or degradation ofmaterial during the period of usage
Fire prevention with powder metal such as zircaloy, etc.

Basic requirements for functional
maintenance (General common
matters)

Maintaining functions under usage environment conditions
Essential reliability and multiplexing and diversification as and when necessary
Equipment with which necessary inspections and maintenance can be executed
Equipment for the necessary monitoring functions
Equipment with which the necessary operations, especially, the necessary site operations can be executed

Equipment with which appropriate counteraction is possible against any hazard (internal or external)©International Research Institute for Nuclear DecommissioningNo.134
7. Implementation items of this project[1) ② Development of various technologies other than the interference removal technology]
d. Transfer method for unit cans from R/B■ Conceptual study of the design to be improved (Basic design conditions based on the safety requirements)
Safety principles
Prevention ofexcessive
release ofradioactive
materials

Confinement ofradioactive
materials

Functional requirements

Design policies

Prevention of leakage ofcontaminants during storage
and handling of the unit canscontaining the fuel debris

・In addition to the confinement area (primary boundary) in the form of the main body ofthe transportation device (including the hydrogen treatment mechanism), the main bodyof the transportation device and the hydrogen treatment mechanism will be stored in acontainer (secondary boundary), and the container will be handled at the time of carryingout of the controlled area.
・The exhaust gas stored in the tank of the hydrogen treatment mechanism will not bereleased in the air outside the controlled area, but will be passed through the vapor
treatment system of the R/B or the Expanded building.
・There is a possibility that a positive pressure is formed in the transportation device main
body during abnormal circumstances following hydrogen generation, hence by managing
negative pressure in the container, the release of contaminants outside the container will
be prevented even if the contaminants are released from the transportation device mainbody.
・By installing ventilating and air conditioning equipment inside the container, the gas
inside the container will be purified even if there is any release of contaminants from thetransportation device main body. Moreover, the designing will be such that it will bepossible to carry out decontamination inside the transportation device main body and thecontainer by remote devices.
There will be two doors at the opening of the container and even if one door is open, theother should be closed. Thus, there will always be a double partition together with theconfinement of the main body of the transportation device.
Enable the replacement of the seal parts to ensure their integrity considering the aging
degradation such as corrosion or degradation of material during the period of usage.
Appropriate seismic class settings or fall prevention measures will be carried out for thehandling equipment so that the transportation device does not suffer an impact due tofalling or tumbling etc. during handling. Acceleration etc. will be studied in future.

©International Research Institute for Nuclear DecommissioningNo.135
7. Implementation items of this project[1) ② Development of various technologies other than the interference removal technology]
d. Transfer method for unit cans from R/B■ Conceptual study of the design to be improved (Basic design conditions based on the safety requirements)
Safety principles
Prevention ofexcessive
release ofradioactive
materials

Prevention ofabnormal
additional
generation ofradioactive
materials

Functional requirements

Design policies

Prevention of additional nuclear
fission reaction

The sub-criticalities of the fuel debris are maintained due to the geometrical shape of the main body ofthe transportation device.
Appropriate seismic class settings or fall prevention measures will be carried out for the handlingequipment so that the transportation device does not suffer an impact due to falling or tumbling etc.
during handling so as to not cause serious damage during handling by which the criticality preventionfunction might be lost.

Prevention of abnormal heatdissipation

The structure should be such that the integrity can be maintained (the permissible temperature rangeof the seal is maintained) by heat dissipation of the main body of the transportation device.

Prevention of abnormal diffusion of radioactive materials by cutting (Not applicable)
Prevention of excessive exposuredue to radiation

Prevention of excessive exposure orinternal exposure of workers

Prevention of a fire or explosion due
to retained hydrogen

Implementation of radiation
shielding in such a way that theworkers can directly access theexternal part of the transportation
device main body.

Install a shield structure so that the radiation dose will be 2 mSv/h or less on the outer surface of thetransportation device main body.

Design for reduction of exposure ofworkers

Specify appropriate settings for shielding, contamination, and dose classification, carry out remote
maintenance to reduce exposure, and use on-site fuel debris transportation flow lines.

Operation management for
reduction in exposure of workers

Use operation methods, maintenance plans, and work management in order to reduce exposure.

Management of hydrogen density
or oxygen density to maintain thelower limit of hydrogen explosion

The structure should not retain hydrogen.
Use the hydrogen treatment mechanism to dilute the hydrogen such that its density can be
maintained at less than the design value. The exhaust gas due to dilution will be released in the tankconnected to the main body of the transportation device, which will be its primary storage duringtransportation.

Confine the contamination such that work outside the main body of the transportation device becomes
possible.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this projectNo.136[1) ② Development of various technologies other than the interference removal technology]
d. Transfer method for unit cans from R/B■ Conceptual study of the design to be improved (Basic design conditions based on the handling requirements)
Safetyprinciples
Remoteoperability

Functional requirements

Design policies

It should be possible to set up the main
body of the transportation device on thework platform by remote operation

Positioning, installation and fixation of the transportation device on theplatform should be possible by remote operation.
The opening lid of the platform and the transportation device can be
opened and closed by remote operation.

The transportation device can be decontaminated by remote
operation.
Confinement ofradioactive
materials

MaintenanceIn case of an emergency, the worker
should be able to access the platform(outside the main body of thetransportation device)

Prevent leakage of radioactive materials from the interface between
the platform and the main body of the transportation device. Moreover,set up a seismic class to enable prevention of leakage of radioactive
materials even during an earthquake.

In order to be able to carry out the main
body of the transportation device, it
should be possible to decontaminate theouter surface

In order to carry out the main body of the transportation device fromthe primary boundary, the structure should be such that the outersurface of the main device body can be decontaminated.

Remote maintenance

Remote maintenance and replacement of consumables should bepossible for the devices on the PCV platform part.
Equipment that is always accessible to workers should be able to be
manually maintained and the consumables replaced.
©International Research Institute for Nuclear Decommissioning7. Implementation items of this projectNo.137[1) ② Development of various technologies other than the interference removal technology]
d. Transfer method for unit cans from R/B■ Conceptual study of the design to be improved (Defining design specifications and issues)
Identification of design specifications and issues
based on design policies and regulation compliancepolicies

Issues that need to be verified by studies and tests inparticular

Design specifications for transportation device

Nothing specific (the structure should be such that thedecontamination of the locations assumed to becontaminated at the design stage, is easily possible)•

The material of the transportation device
should maintain the temperature such thatno function is lost.
Surface of the transportation device: 2
mSv/hDistance of 1 m from the surface: 0.1
mSv/h

•

The shape of the unit can should bemaintainable without damaging thetransportation device in view of theassumed events during handling.

••

Hydrogen explosionprevention

Outer surface of container: 0.4 Bq/cm2 orless #1••

・It is necessary to study a structure wherein transportation
within the site is possible while both "Confinement" and"Prevention of contamination" is carried out for the door partwhich will open and close during the storage or lifting out of
the unit cans.
・A structure in which operations including the opening andclosing of the door can be done by remote operations, isnecessary.
•

•
Heat
Criticality
removal prevention•

The leakage from the transportation device,and the leakage from the part connecting
the platform and the transportation device
is below the limit value.
The confinement is maintained without
damaging the transportation device in viewof the assumed events during handling.

ShieldingConfinement•

As regards the detailing and refinement of design specifications, the plans will be
formulated after reflecting on the details of actions to deal with future technological issues.

A state where the levels are kept
sufficiently below the lower hydrogen
explosive limit (4vol%) should bemaintained. In order to maintain this, highpressure gas should not be used whiletransporting within the 1F site premises.

Studies under this PJ••

Planning and risk evaluation of work flow. As
regards the events which carry a risk, verify thepresence of structural strength required to maintainconfinement by means of simulation or testing, etc.
Feasibility study by means of formulating a device
concept and conducting seal testing, etc.

•

The structure should allow easy decontamination.

Nothing specific (Note that it is necessary to study theinterface conditions after the basic design of the unit can
is completed)•

Planning and risk evaluation of work flow. As
regards the events that carry a risk, ensure thestructural strength to maintain the shape by meansof simulation or testing, etc.

•

Confirm the feasibility of the structural design•

Confirm by means of simulation, etc.

•

Confirm the feasibility of the structural design•

Confirm by means of simulation, etc.

•

Measures to prevent rise in pressure inside the main
body of the transportation device following hydrogengeneration → If high pressure-resistant specifications
are used, the transportation device becomes larger insize (+ weight increases). Study systems which can
reduce the density of hydrogen while avoiding rise inpressure.
It is necessary to have a structure wherein the densityof hydrogen inside the transportation device can bemonitored remotely and hydrogen treatment is possible
through remote operations during routine work.

•

Verify the feasibility of the hydrogen treatment
system through desk studies or testing, etc. by
injecting air or inert gases such as nitrogen.

•©International Research Institute for Nuclear Decommissioning#1：Cited from the Atomic Energy Society of Japan (AESJ) " Safety Design and Inspection Criteria for Transfer Container for Spent Fuels, New Mixed Oxide Fuels, and High Level Radioactive Wastes: 2013"No.138
7. Implementation items of this project[1) ② Development of various technologies other than the interference removal technology]
d. Transfer method for unit cans from R/B■ Response to technical issues (Confinement mechanism for the door)••

Separation or integration of shielding and confinement functions for the structure of the door. Moreover, three kinds of door structures are planned as per thedoor opening operation mechanism.
All have their own merits and demerits, and evaluation will be carried out by continuing the studies, such as assigning priority levels to the evaluation items
or testing the seals, etc.
"Separation of Shielding and Confinement" Concept
Transportati
on device
main body

Sealinglid

Evaluation items[Legend]◎ ： Especially excellent as comparedwith other concepts〇：Superior than △, but inferior to ◎△：Too many issues compared to otherconcepts

Ability tohandleShield shutter

Shield shutter

Platform

Platform

Platform

Platform
Red line: PCV boundary

Tra
nsp
orta
tion
devi
ce
mai
n
bod
y

Unit can storage and body
Sealing/Shielding lid(plug)

Platform
Platform

Platform

Platform

Red line: PCV boundary

Sealing lid[Linear method]
The sealing lid is attached and detached by the
direct movement of the rail.

[Swing method]
The sealing lid is attached and detached by thearc-like movement caused by the arm installedinside the PCV.

[Plug method]
The sealing lid (plug), which has the thickness of the shield, is attached anddetached by the arm installed inside the PCV.

Occupied area on
platform (footprint)〇〇◎
It can be expected to become compact as compared to the other concepts since
the plug will include both the shielding and the confinement functions.

Weight of the door〇
Since the clamping mechanism of the sealing lid
is needed, the weight increases as compared tothe swing method.

◎△
It becomes heavier due to the thickness of the shield provided to the sealing lid.

△△〇
As compared to other concepts, comparatively lightweight with few equipment
components

Evaluation scheduled

Evaluation scheduled

Evaluation scheduled

Weight of the entire
equipment installed on theplatform
Speed of the door
opening/closing operation

Nuclearsafety"Integration of Shielding and Confinement" Concept

Unit can storage△
The interface of the platform with the shieldmust be estimated.

△
As the arm rises gradually over the PCV, it becomes a source of radiation andhence it is necessary to add dose reduction measures.

Shielding function〇

Confinement function

Evaluation scheduled

Evaluation scheduled

Evaluation scheduled

Risk of door falling〇〇△
It is necessary to study the fixing mechanism to be used when the arm ishandling the plug and study the fall risk reduction measures.

©International Research Institute for Nuclear DecommissioningNo.139
7. Implementation items of this project[1) ② Development of various technologies other than the interference removal technology]
d. Transfer method for unit cans from R/B■ Response to technical issues (Confinement mechanism for the door)"Integration of Shielding andConfinement" Concept"Separation of Shielding andConfinement" Concept• A structural image for each method is shown below.

[Linear method]# Each structure is likely to change in future with the progress in design[Swing method]
Unit can storage

Sealing lid(Transportation
device main body)

Sealing lid(Platform)

PCV Platform(Cell side)

Unit canstorage

PCV Platform(PCV side)

Platform

Arm

Guide-rail(Vertical
movement)

Guide-rail(Horizontal
movement)

Guide-rail[Plug Type]Transportation
device main
body
Unit canPlug(Transportation
device main body)

Plug(Platform)

Plug
Handling arm©International Research Institute for Nuclear DecommissioningNo.140
7. Implementation items of this project[1) ② Development of various technologies other than the interference removal technology]
d. Transfer method for unit cans from R/B■ Response to technical issues (Prevention of hydrogen explosion)• Maintain the hydrogen density below the explosion lower limit (4Vol%) and manage it below the set value (2.5Vol%)• Always monitor the density of hydrogen in the main body of the transportation device. If a standard value is exceeded, gas will be injected for
dilution. After purification, the exhaust gas will be stored temporarily in the waste gas tank by the compressor. It is assumed that the exhaustgas tank can store at the most for seven consecutive days. Also conduct a study on reducing the size of the tank (add as needed when diluting).
• Exhaust gas which is thus stored will be passed through the heating, ventilating and air-conditioning system of the transportation destination or
R/B.

① During routine monitoring② When hydrogen is generated (During dilution)

Nitrogen or compressed air
for dilution

Nitrogen or compressed air
for dilution (19.3 L/h)Gas
circulation
Filter(#1)

DilutionCompressor

H2 Sensor

Exhaust gas
tankFilter(#1)

H2 Sensor

Exhaust gas
generation: 480 L/day

Compressor

Exhaust gas
tankApprox.Φ600×1600 mm(400 L, Max.0.8 Mpa)

#1: Also conduct a study on installinga filter inside the main body of thetransportation device considering arise in the radiation

Piston-type vacuum /pressurized airtightcompressor©International Research Institute for Nuclear DecommissioningNo.141
7. Implementation items of this project[1) ② Development of various technologies other than the interference removal technology]
d. Transfer method for unit cans from R/B■ Response to technical issues (Evaluation of work steps, equipment layout ,and work schedule during handling)➢ Goal: To transport 10 unit cans per day➢ Select a series of handling work sequences such as transportation between
cells using cranes and vehicles, opening/ closing operations between thecell partitions, inspection prior to carrying out, etc.
Evaluate the overall handling time by allotting working hours for each task.

New refuelingfloor➢ Plan the overall handling plan for a day taking into account the prohibitionof opening of the partition doors simultaneously or the settings for thewaiting time when multiple transportation equipment are in use.

Cell 1
Cell 2 (Front
room)➢ The cans can be carried out in about 2.5 hour intervals. Assuming 10
vehicles a day with a target time of 2.4 hours or less, two carrying out
routes will be established. Note that if the movement of one vehicle isfocused on, the handling work inside the R/B will require 11.5 hours, so
altogether ten transportation devices will be needed.

R/B carrying-in/out area 1

Cell 2

New refuelingfloor

Table 1. Carrying out timing chart for fuel debris transportation device focusing on the line
区分

機器

扉

PCV内

1h

取合い位置

2h

3h

4h

5h

6h

7h

8h

9h

10h

11h

12h

13h

14h

15h

16h

17h

18h

19h

PCV上
プラット

－
－

ホーム

（PCV上プラット
ホームセル内）セル

PCV上プラット

TD取合い位置

ホームセル/セル1
昇降装置

扉

－

Ｐ

Ｐ

開

TD T
設置 D

閉
UC投/デブリ充填
/UC収

下
扉降

T
D

接続

H2起

20h

21h

22h

PCV Platform
Cell－
プラットホーム
移送装置

扉閉

確認

開
扉
移動 閉
TD

－
セル1
－
レール

（セル1内）
台車取合い位置

H2
起

H2
解

Continue the same steps

扉開 TD

セル1/セル2前室扉

空台

移動

移動 扉
閉

セル2前室

ターンテーブル－（セル2前室内）台車
回転

取合い位置

扉

開

セル2前室/セル3扉

セル2

ターンテーブ
ル

－

空台 扉

移動

移動 閉
空調

（セル2）
台車取合い位置

セル3/
RB搬出入エリア1扉
ターンテーブ（R/B搬出入エリア1

ル内）台車取合い位置－TD

Continue thesame steps

回転

検査

扉
開 TD
移

空台
移動 扉

動

閉
回転

ク
レ

R/B搬出入
エリア1

クレーン

－

レール

－

(R/B搬出入エリア1

ー

接続

H2

H2

起

解

TD
移動

ン

内）水素処理コンテ

ターンテーブル－

昇降装置↓
水素処理コンテナ

ナ取合い位置

C
開

H2

H2

起

解

（昇降装置）

C

閉

検査
空C

TD
移動

移動

Carry out about every 2.5 hours

回転

オペフロ位置TD
移動（昇降装置）
RB搬出入

扉

扉 空C 扉

扉

開 TD 閉移動

地上位置

エリア1/2扉

開 移動 閉

車両
固定

検査

車両 車両
入替 解除

（R/B搬出入エリア2

R/B搬出入エリア2 水素処理コン－

内）水素処理コンテ
TD
移動

ナ取合い位置テナ

Platform

PCV

空調

R/B carrying-in/out area 1

R/B carrying-in/out area 2

Ｒ/Ｂ外搬出入扉
Ｒ/Ｂ外

Ｒ/Ｂ外

11.5 h/vehicle

扉

扉

開

閉

©International Research Institute for Nuclear DecommissioningNo.142
7. Implementation items of this project[1) ② Development of various technologies other than the interference removal technology]
d. Transfer method for unit cans from R/B■ Summary⚫

Results in FY2019• The building cell layout was determined based on the plan for fuel debris on-site transportation by transportation devices.
• The basic functional requirements for transportation were determined based on the assumed parameters of the unit cans/
fuel debris.
• The basic functional requirements for safety were determined based on the nuclear safety and related regulations.
• The design policies were formulated and the technical issues were identified based on the above-mentioned basic
functional requirements for transportation and safety.
• The conceptual design for the important technical issues of doors (double doors) and hydrogen treatment mechanism was
prepared.
• The evaluation and feasibility of fuel debris transportation throughput using the transportation devices was confirmed.

⚫ Details of the implementation schedule for FY2020• Narrowing down of the concept of door (double doors) and its feasibility evaluation by means of analysis and seal unit
testing, etc.
• Feasibility evaluation of the concept of hydrogen treatment mechanism by means of element testing, etc. (prototyping ofelement composition)• Preliminary analysis of the strength reinforcement of the transportation device main body in anticipation of safety analysis.
• Analysis and testing to evaluate the structural feasibility of the winch used for unit can collection.
• Structural designing and feasibility evaluation of the unit can grasping mechanism, study of the procedure for the handling
operations, and formulation of a plan for issues and countermeasures. Study of problem events in particular.
• Study of facility layout inside the R/B with a view to further improving the fuel debris transportation throughput.

©International Research Institute for Nuclear DecommissioningNo.143

7. Implementation items of this project2) Development of technology for handling of fuel debris(i) Technological development for fuel debris collection and storage systems

In addition to the molten-damaged fuel from the RPV, there is a pile up of various states of fuel debris (loose fueldebris, sludge, particulate (powdered) fuel debris, fuel debris due to processing such as crushing/cutting etc.) atvarious places on the structures such as the pedestal bottom etc. inside the PCV. For efficient fuel debris retrieval,a collection method and a method and system for storage in containers will be developed according to the state of
the fuel debris, and as and when necessary, the feasibility will be verified by means of element tests.
During the element tests, data to evaluate the collection efficiency will be acquired and this data will be used forevaluating the fuel debris retrieval efficiency considering the location and state of the fuel debris. Moreover, aconceptual study will be undertaken considering assurance of safety and on-site applicability for the storage
methods and systems, which can improve the retrieval efficiency.
The following are the main development study items, and issues will be identified and consolidated by carryingout element tests as and when necessary:a. Development of particulate fuel debris suction and collecting system
・The optimal shape of the pump strainer is being studied.
・A method of detecting the fullness of the separator without using a pressure gauge/ flowmeter is being
studied.
b. Development of technology for transferring fuel debris and wastes
・The retrieval method is being planned by presuming the state and properties of the fuel debris.

©International Research Institute for Nuclear DecommissioningNo.144

7. Implementation items of this project[2) (i) Technological development for fuel debris collection and storage systems]
a. Development of particulate fuel debris suction and collecting system⚫ Purpose of developing a suction and collecting system for particulate fuel debris➢ The suction and collecting system for particulate fuel debris is being
developed with an aim to collect the particulate fuel debris that is difficult tocatch with the robot arm or in liquid phase so as toprevent missing out on any fuel debris.
⚫ Issues to be solved➢ Determining the details of the specifications, equipment, maintenance andoperability etc. considering the 1F environment.
⚫ Approaches to development (from the main issues identified during theprevious term of the subsidy project)a) Study of a pump strainer shape suitable for this systemb) Study of a method to detect the fullness of the unit can (separator)c) Study of a method for replacing the unit can (separator)d) Study of a suitable pump model for suction from the air(A pump model for suction under water has been studied.)
Based on the above results, feasibility of the system will be studied by
performing element tests.

⚫ Expected outcome➢ Specifications with details determined for the suction and collecting systemfor particulate fuel debris suitable for the 1F environment.

Particle size
Large

Block-like fuel debris
Particle size: 100 mm or
more
Needs to be processed
before storing in a unit
can

Small

Pebble-like fueldebrisParticle size: 10 mm to
100 mmCan be stored in theunit can without any
processing

Suction scheduled with the robot arm

Particulate fueldebrisParticle size: 0.1 mm to10 mm
Needs to be collectedby suction as it cannot
be handled by a tool

Powdered fuel debris
Particle size: 0.1 mm or
lessWill be collected in theliquid phase of thesystemCollection scheduled in liquid phase

Target scope at this time

<Separator>
・ Study of method of replacement by
remote operation
・ Study of separator sieve opening
・ Study of a method to detect theclogging and fullness of the separator<Pump>
・ Selection of pump model<Pump strainer>・ Study of a shape suitable for thesuction of the particulate fuel debris

Suction and collecting system for particulate fuel debris Study issues©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[2) (i) Technological development for fuel debris collection and storage systems]

No.145a. Development of particulate fuel debris suction and collecting system⚫ Study flow
Results and issues in thesubsidy project in theprevious term

The following issues have been selected presuming suction and collection from water:・Since the strainer (suction port) of the submersible pump has been designed empirically as per the
manufacturer's know-how, it is difficult to say whether it is a suitable shape for the suction and collectionof the particulate fuel debris.
・ The method of detecting the fullness of the particulate fuel debris unit can (separator) is undecided.
・ The method of replacing a filled unit can (separator) is undecided.

Approach towards the issues to be handled during thesubsidy project of the current term

(Suction from water)

Study of the shape of thestrainer・Solutions for the issues identified during the subsidy project of the previous term will be
studied.
・A suction and collecting system for particulate fuel debris, applicable even in anenvironment (in air) where some of the fuel debris has been exposed to air, will be
studied.

(Suction from air)Study of the suction device

Results of this fiscal
year(Common equipment to be used under water/ in air)Study of methods to detect the
fullness of the separator
Study of methods to replace theseparator

Verification of the feasibility of the study results bymeans of element tests

Identification of future issues and response policies©International Research Institute for Nuclear DecommissioningNo.146

7. Implementation items of this project[2) (i) Technological development for fuel debris collection and storage systems]
a. Development of particulate fuel debris suction and collecting system◼ Development Schedule
FY2019Study items
４

Major milestones

5

6

7

8

9

10

FY2020
11

Interim Report

▼

12

1

2

3

4

Annual Report

▼

5

6

7

8

9

10

11

Interim Report

▼

12

1

2

3

Final Report▼

1. Technologicalinvestigation/ Conceptual
study
2.Test plan/ Test device
prototyping
3. Element tests
4. Summary

Remarks©International Research Institute for Nuclear DecommissioningNo.147

7. Implementation items of this project[2) (i) Technological development for fuel debris collection and storage systems]
a. Development of particulate fuel debris suction and collecting system⚫ Preconditions for study➢ The robot arm should be able to access the particulate fuel debris.
➢ At the time of suction from water, the water level necessary (roughly
about 20 cm) for operating the submersible pump should beavailable.
➢ The collection equipment (separator) that is filled with theparticulate fuel debris will be removed by the manipulator inside thecell after the robot arm is stored in the cell.
➢ The maintenance of the particulate fuel debris suction andcollecting system will be also carried out by the manipulator inside
the cell.
⚫ Development objectives➢ Development of a suction and collecting system which will
efficiently collect the particulate fuel debris (0.1 mm to 10 mm)
believed to exist at the bottom of the reactor.

Separation device(separator)

Separator
replacement
mechanism

Transfer part(pipe)

Suctionequipment©International Research Institute for Nuclear DecommissioningNo.148

7. Implementation items of this project[2) (i) Technological development for fuel debris collection and storage systems]
a. Development of particulate fuel debris suction and collecting systema) Study of the shape of the pump strainer suitable for the suction and collecting system for particulate fuel debris➢ In order to perform suction and collection of the particulate fuel debris in a stable manner,
the strainer shape satisfying the following was studied:• In order to prevent clogging of the pump, solid objects that are larger than thepredetermined size should not be sucked.
• Suction occurs at a flow velocity of 2.7 m/s or more, which is terminal settling velocity x
1.5 times• The sucked particles do not remain in the strainer.
➢ As a result of the study, it was possible to develop a strainer shape that satisfies the abovementioned conditions (Refer to the lower right side figure).
➢ For information, the results of the flow analysis done on the manufacturer standardized
strainer are given in the lower left side figure.
The flow velocity at the strainer inlet port is slow
The flow velocity stagnates in thebottom part of the strainer

Strainer(suction port)

Image of the submersible pump strainer

Improvement in the flow velocity at the strainer inlet port
Reduction in stagnation inside the strainer

Review of shape notcontributing to thesuction performance

Flow analysis results of the manufacturer
standardized strainer (suction port)

Flow analysis results of themodified strainer (suction port)©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[2) (i) Technological development for fuel debris collection and storage systems]

No.149a. Development of particulate fuel debris suction and collecting systemb) Study of the method to detect the fullness of the unit can (separator)➢ The method that measures the changes in the discharge pressure or the flow rate of the pump to detect the fullness is a method that
detects fullness indirectly and not directly, hence, the issues of erroneous detection or accuracy remain.
➢ Here, a method that directly detects the fullness of the separator using a robot-arm mounted camera for monitoring, was studied.
➢ The method studied involved making holes on the top part of the separator and when the separator was filled with fuel debris, the fueldebris seemed to spout out from the holes at the top of the separator and this would be monitored by the camera (lower left side
figure).

c) Study of the method to replace the unit can (separator)➢ One-touch installation/ removal mechanism with toggle mechanism + spring was examined as the structure which can enable easy
replacement by remote operations. (Lower right side figure)Fullness monitor

Suctionequipment(pump)
NozzleSuction part monitor

Separation
device(separator)

Transfer part (pipe)

Separator
HandleSpringParticulate
debrisImage of the separator fullness detection method

Separator replacement mechanism concept©International Research Institute for Nuclear DecommissioningNo.150

7. Implementation items of this project[2) (i) Technological development for fuel debris collection and storage systems]
a. Development of particulate fuel debris suction and collecting systemd) Study of a pump model suitable for suction from air➢ Screening was conducted for the selection of a pump model suitable for suction from air.
➢ Ejector pump which uses the compressed air as the driving source was selected from the screening results.
Ejector pump
(Driving source: Compressed air)Model

Principles and Features

Ejector pump
(Driving source: High-pressure water)This type of pump is categorized as a special pump.
It has three ports, namely, a driving fluid inlet port, transportation fluid inlet port, and discharge port, and the driving
fluid injected from its driving fluid inlet port gives rise to negative pressure which causes suction. Reliability is highbecause it has no moving element.

Vacuum suction type
Negative pressure is generated inside the container bymeans of the piston type or the root type of pump and thefluid is sucked using this negative pressure. There is alarge number of maintenance parts as compared to theejector pump.

Schematic drawing

FilterCompressor air

Compressed air (Nitrogen (100 m3/h) used for in-leakprevention is assumed to be the driving source)

High-pressure water (Circulating water (10 m3/h) from theliquid phase is assumed to be the driving source)Electricity (Supplied by a cable via the robot arm)

From airSpecific gravity of 11 and particle size of approximately
10 mm is presumed

Specific gravity of 11 and particle size of approximately 0.1
mm is presumed

Specific gravity of 11 and particle size of approximately 10
mm is presumed

Under
water

Not clear

Specific gravity of 11 and particle size of approximately 5
mm is presumed

Not clear

Power

Ability to remove
solid objects (Deskstudy)

Resistance to radiation

Resistant to radiation (metallic)

Consists of components which have poor resistance toradiation

Draining

Even if water is sucked or transferred, draining is possible by the separator in the latter stage

It is necessary to separately study the method of draining
since there is no discharge mechanism for water.

•

Evaluation

It is necessary to consider the impact on thegaseous phase inside the PCV, but it can beapplied in the range which will not cause any
impact.

•As per the information from the catalog, since thesuction performance (flow velocity) in air is clearly low,
and it cannot suck the target particles, so this pump
was eliminated from consideration.

•

It is necessary to study the maintenance of negative
pressure, the method of draining, water-proofing, andmaintenance methods and there are a lot ofdevelopmental issues, so this pump was eliminatedfrom consideration.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[2) (i) Technological development for fuel debris collection and storage systems]

No.151a. Development of particulate fuel debris suction and collecting system● Results of the conceptual study for this fiscal year
➢ The status of study for suction under water or in air are as follows.
➢ In the next fiscal year, element tests are scheduled after studying the pump suction port for "Suction from air”.
Items

Suction from water

Suction from air

Selection of suction
equipment

Submersible pump selected

Ejector pump selected（Driving source: Compressed air）

Study of pumpsuction port

An appropriate strainer shapewas evaluated through CFD
analysis

Not yet(The study is scheduled for the
next fiscal year with the method
applied for suction from water)

Notes

Separator fullness
detection method

The method of detecting thefullness with a surveillancecamera was designed

Same as left

・Equipment used in common forwater and air・Study scheduled for a shapewhich can collect a size of 0.1 mm
or moreSeparator
replacement
methodOne-touch installation/ removal
mechanism with toggle
mechanism + spring was studied

Same as left

・Equipment used in common forwater and air

Evaluation

To be carried out in the next fiscal
year

To be carried out in the next fiscal
year©International Research Institute for Nuclear DecommissioningNo.152

7. Implementation items of this project[2) (i) Technological development for fuel debris collection and storage systems]b. Development of technology for transferring fuel debris and wastes⚫ Purpose of development➢ For efficient fuel debris retrieval, a collection method, and a method and system for storage in containers will be developed according to thestate of the fuel debris, and as and when necessary, the feasibility will be verified by means of element tests.
⚫ Issues to be solved➢ Determining the specific collection procedures and devices to be used for the assumed variety of fuel debris➢ Feasibility verification of planned work procedures and devices➢ Refinement and shortening of work time evaluation accuracy⚫ Approaches to development➢ Study of collection methods (procedures and necessary devices) by consolidating the various forms of fuel debris, such as slag, particulatematter, pellets, damaged fuels, etc., as a precondition (study of collection methods depending on the state such particle size or fuel rod
stubs, etc.)➢ Consideration of shortening of work time, maintenance during operations, response during breakdowns, etc.
➢ Study of investigation and adoption of cutting methods (processing ability and dust evaluation etc.), and element tests as and whennecessary➢ Study of feasibility of the methods of processing-grasping-unit can storage and carrying out➢ Evaluation of falling of debris during work and study of necessary functions➢ Coordination with other project "Development of Technology for the Retrieval of Fuel Debris and Internal Structures (Development ofTechnology for Dust Collection System of Fuel Debris)”
Presumptions andclassification of the fueldebris properties
Creation of a presumptive
diagram of the status inside
the reactor based on theexisting study results, and use
this as the precondition for this
study

⚫

Planning the retrieval method dependingon the properties

Planning the composition of thedevices to be used

Planning the retrieval methods for each of
the presumed states of the fuel debris

Creation of conceptual diagrams forwork systems and devices bystudying not just the usage state ofthe devices, but also theprocedures and compositions forthe replacement of consumables ormaintenance, etc.

Selecting the processing method andproviding information for the development ofcollection technology

Element tests
Measurement of retrieval methods,
work time, and collection efficiency by
the retrieval element tests ofsimulated fuel debris.
Reflection in the conceptual diagramand procedure chart for the device.
Estimation of the fuel debris retrieval
period (review of existing results)

Expected outcome (TRL：3)➢ Conceptual planning of the collection methods (procedure and necessary devices) depending on the state of the fuel debris (slag,
particulate matter, pellets, damaged fuel etc.)➢ Feasibility evaluation of the planned methods by means of element tests➢ Estimation about the work time and the assumed debris retrieval period©International Research Institute for Nuclear Decommissioning7. Implementation details of this projectNo.153[2) (i) Technological development of fuel debris collection and storage systems]b. Development of technology for transferring fuel debris and wastes◼ Development schedule
Study items

Major milestones

FY2019
４

5

6

7

8

9

10

11

FY2020
12

1

2

3

4

5

6

7

8

9

10

11

Interim Report

Interim Report

Interim Report

▼

▼

▼

12

1

2

3

Final Report▼

1. Conceptual Study
2. Formulation of theelement test plan

3. Preliminary tests
necessary forconceptual study
4. Preparations forelement tests
5. Element tests

6. Summary

Remarks©International Research Institute for Nuclear Decommissioning7. Implementation details of this projectNo.154[2) (i) Technological development of fuel debris collection and storage systems]b.

Development of technology for transferring fuel debris and wastes■Preconditions of the study (Estimation of the state of fuel debris inside RPV: Around theupper grid plate and reactor core)
Creation of a presumptive diagram for the retrieval study by putting together the existing
estimated results

Undamaged fuel assemblies
Undamaged control rods
Openings due to melting around thecenterCreep deformation around the centerNo adherence of molten fuelNo damageNo adherence of molten fuelPowdered/particulate fueldebris (including abrasives)or fuel pellet deposits on thefuel rods or the reactor core
support plates

Burnt pellets
Particulate slag fuelAbrasives
No damage

Mostly standing upright
There are fuel assemblies
which can fall from the outer Inclined and touching thearea into the central area and surroundings
collide with the processing/
Toppledcollection devices

Damaged. There are no fueldebris adhesions
Fuel support metal fittings
remain on the reactor core
support plates

Fuel debris adherence on theinsideAdherence to the reactor
core support plate

Surrounding structures melted and solidified on the reactor
core support plates
Melted fuel debris which has entered between the reactor
core support plates and the shroud bodyPresumptive Chart for the State of Debris Distribution, RPV, and PCV "Update of the overall situation inside the reactor" (FY2017 Result Report; June 2018)
Created on the basis of the presumptions about the fuel debris distribution at Units 1 to 3 "Technical Strategic Plan FY2018 for the Decommissioning ofFukushima Daiichi Nuclear Power Station"©International Research Institute for Nuclear Decommissioning7. Implementation details of this projectNo.155[2) (i) Technological development of fuel debris collection and storage systems]b.

Development of technology for transferring fuel debris and wastes■Preconditions of the study (Estimation of the condition of fuel debris inside the RPV: Lower plenum and reactor
bottom)Upright position: No damage

Collapsing on top of one
another

Adhesion of a little molten
fuel debris

Adhesion of a lot of molten
fuel debris

Scatteredpellets/claddings andstructures (handles/fuelsupport metal fittings)

Non-fused

Granulated cooled andhardened molten
objects (metal rich)Non-fused

Fused

Fused
Exists independently

Crusted fuel debris

Molten metal: PoolshapedBall-shaped molten
metal

Fused with other structures(CRD/ bottom head)
Metal rich
Ceramic rich
Metal rich
Ceramic richPresumptive Chart for the State of Debris Distribution, RPV and PCV "Update of the overall situation inside the reactor" (FY2017
Result Report; June 2018)
Created on the basis of the presumptions about the fuel debris distribution at Units 1 to 3 "Technical Strategic Plan FY2018 forthe Decommissioning of Fukushima Daiichi Nuclear Power Station"©International Research Institute for Nuclear Decommissioning7. Implementation details of this projectNo.156[2) (i) Technological development of fuel debris collection and storage systems]b. Development of technology for transferring fuel debris and wastes■ Preconditions of the study (Classification by type of the fuel debris conditions inside the RPV)Classification of Fuel Debris

Mass (t)Composition

Porosity

Hardness

Melting point

Fuel
6 GPa
Up to10%
Cladding
1 to 3 GPa

Fuel
2850℃
Cladding
1850℃

Undamaged fuel assemblies
Vicinity of the upper gridplate

Undamaged fuel assemblies and internal
structures

Undamaged upper grid plate
Damaged upper grid plate

Fuel debris on the upper
part of the reactor core
support plates (Fuel rod
stubs)

Undamaged control rods

Damaged or deformed upper grid plate
Upright fuel rod stubs with little damage

There are fuel rods with shapes maintained Greatly damaged fuel rod stubs(identifiable as fuel rods)
Fuel rod stubs that are damaged but are supported by surrounding structures

17

Toppled fuel assemblies
Undamaged fuel support metal fittings

51

Fuel support metal fittings remaining on the Undamaged reactor core support plate
reactor core support plate
Damaged fuel support metal fittings that have melted and solidified in the surroundings

In the vicinity of thereactor core support
plates and the fuelsupport metal fittings

17

Damaged reactor core support plates that have melted and solidified in the surroundings
Block-like fuel debris piled up without adhesion
Deposited fuel debris

Pebble-like fuel debris piled up without adhesion
Powdery/ particulate fuel debris piled up without adhesion

Melted and solidified structures or fuelIn the vicinity of CRGT
(From the reactor core
support plates up to thereactor bottom)

UO2
ZrO2
(U,Zr)O2Zr(O)Fuel debris adhering to the reactor core support plates

Block of fuel debris that has solidified at the
Fuel debris that has melted and solidified on the internal structures
bottom of the reactor
Reactor bottom

Up to
88%

6 to
10 GP

2500 to
2700℃

35(U,Zr)O2(Zr,U)O2
ZrO
Fe

Up to
88%

6 to 17 GPa1075 to
2700℃

35(U,Zr)O2(Zr,U)O2

Up to
30%

6 to 10 GPa

2500 to2700℃

Block-like fuel debris that has melted and solidified between the shroud and the reactor
core support plates

CRGT that is undamaged and standing upright
CRGT and the control rod drive mechanism
which are either standing or collapsed due CRGT that has collapsed but has little damage
to damage
CRGT that has collapsed and has fuel adhering to it

CRD housing(U,Zr)O2(Zr,U)O2

17

CRD housing present without damage

15

85

Block-like fuel debris piled up without adhesion

Deposits at the bottom of the reactor

Pebble-like fuel debris piled up without adhesion
Powdery/ particulate fuel debris piled up without adhesion©International Research Institute for Nuclear Decommissioning7. Implementation details of this project[2) (i) Technological development of fuel debris collection and storage systems]b.

No.157Development of technology for transferring fuel debris and wastes■Planning the collection methods according to the types of fuel debris inside the RPV⚫ Policies for planning the collection methods
⚫

⚫

⚫

⚫

Fuel debris with a size of 200 mm or less to be collected without any processing•
Fuel debris with a size of 10 mm or more to be picked up (buckets, jamming adsorption pads etc.)•
Fuel debris with a size of 10 mm or less to be suctioned, etc.
Melted and solidified blocks•
Priority to be given to breaking down the debris with a chisel (spiky hammer) that has a simple operation and a fast
processing speed, or to breaking the brittle parts with bucket-like tools, but prepare the following alternate plans
assuming a scenario where the debris cannot be broken down due to accessibility or its physical properties.
⇒Disc Saw (Less powder dust and simple operation)⇒Core boring⇒Reactor core may not be stable (might collapse), so consider using it in combination with ultrasonicwaves ⇒Widening of opening with a splitting implement.
⇒Laser (and gouging)
Fuel assemblies and fuel rods•
Draw the fuel assembly or fuel rod, which is not damaged and can be removed as is, into long mobile cells•
Priority to be given to cutting out (while tearing out) the damaged fuel using a hydraulic cutter or bucket-like tools, but
prepare the following alternate plans assuming a scenario where the material is hard and cannot be cut.
⇒ Cutting out small parts with a disc saw
Structures•
The melted and solidified internal structures to be chipped off and cut with an ultrasonic chisel, etc.
⇒ Disc saw to be used for the block-like fuel support metal fittings or the reactor core support plates⇒ Laser or AWJ to be used for the thick shroud or the CRGT etc.
⇒ Hydraulic cutter to be used for the thin upper grid plate©International Research Institute for Nuclear Decommissioning7. Implementation details of this projectNo.158[2) (i) Technological development of fuel debris collection and storage systems]b.

Development of technology for transferring fuel debris and wastes■Planning the collection methods according to the types of fuel debris inside the RPV
Targets

Core boring
Ultrasonic core drill

Disc Saw
Circular saw*1

Cutting wheel
grindstone*2

Undamaged fuel assemblies

Hydraulic cutter

Bucket

Chisel hammer

AWJ

Laser cutting
Laser gouging

△
△
△

◎
◎
◎

△
△

◎
〇

△

〇

△
△
△

〇
◎
◎
◎
〇

△

◎

Collected without processing〇〇〇

Undamaged control rods
Undamaged upper grid plate
Damaged or deformed upper grid plate

〇
△
◎
△

Upright fuel rod stubs with little damage
Greatly damaged fuel rod stubs
Fuel rod stubs that are damaged but supported by surrounding
structures
Toppled fuel assemblies
Undamaged fuel support metal fittings

◎
〇
〇
〇

◎
◎

Collected without processing〇

Undamaged reactor core support plate
Damaged fuel support metal fittings that have melted and solidified
in the surroundings
Damaged reactor core support plates that have melted andsolidified in the surroundings◎◎◎

Block-like fuel debris piled up without adhesion
Pebble-like fuel debris piled up without adhesion〇
Collected without processing

Powdery/ particulate fuel debris piled up without adhesion

Collected without processing◎

Fuel debris adhering to the reactor core support plates△〇

Block-like fuel debris that has melted and solidified between the
shroud and the reactor core support plates〇〇

CRGT that is undamaged and standing upright
CRGT that has collapsed but has little damage◎〇

CRGT that has collapsed and has fuel adhering to it
Fuel debris that has melted and solidified on the internal structures△

CRD housing maintaining its shape
Block-like fuel debris piled up without adhesion

△

◎

〇

Pebble-like fuel debris piled up without adhesion

Collected without processing

Powdery/ particulate fuel debris piled up without adhesion

Collected without processing#1：The targets are processed by cutting#2：The targets are processed by grinding with an abrasive coating◎〇◎： Processing method with highest applicability〇： Processing method with high applicability△： Processing method which may be applicable, but has issues©International Research Institute for Nuclear Decommissioning7. Implementation details of this project[2) (i) Technological development of fuel debris collection and storage systems]b.

No.159Development of technology for transferring fuel debris and wastes■Planning the collection methods according to the types of fuel debris inside the RPV⚫ Basic requirement functions• The tip tools (processing tools and collection tools) should be able to access the entire internalarea of the pressure vessel below the height of the upper grid plate (specific limitations will be
formulated in FY2020).
• The tip tools can be attached/ detached and replaced inside the PCV.
• For maintenance such as replacement of consumables, etc., the tip tools shall be carried out
from the platform at the top of the PCV. Therefore, the tools should be able to pass through the
opening for carrying out the unit cans.
• The power at which a manipulator can operate is given below (the payload of a general power
manipulator is 2000 N or less)• Processing tool mass should be 500 N and the reactive force during processing should be1000 N or less.
• The following utilities can supply to the reactor core and the reactor bottom• Power supply: 3φ, 200 V, 10 A
• Water: Approx. 1 m3/h
• High-pressure air: (The pressure and flow rate will be formulated in FY2020 by referencing
the existing study results of the gaseous phase system)⚫ Environmental requirement conditions• Dose rate: Inside the reactor, reactor bottom: 3000 mSv/h• (After setting up the PCV flange platform)： 5 to 10 mSv/h• Temperature： -7 to 40℃• Within the boundary: ≤ 99%• Design life: 50 years (However, the consumables can be replaced during maintenance.)©International Research Institute for Nuclear Decommissioning7. Implementation details of this projectNo.160[2) (i) Technological development of fuel debris collection and storage systems]b.

Development of technology for transferring fuel debris and wastes■ Planning the collection methods according to the types of fuel debris inside the RPV

The opened hole is enlarged with asplitting implement and fuel debris
is crushed.

Opening of holes with theultrasonic core drill

Chisel hammer

Overall view of the reactor bottom work status●Study of the collection method
Suitable processing collection tools were selected by using
the assessment model diagram. A step diagram was createdby using these tools and a comprehensive study was
conducted with respect to the effectiveness, workability, andefficiency of the collection method.
● Content of future studies: Selection of issues
・ Feasibility of the interface between the tools
・ Confirmation of visibility by means of camera
arrangements, etc.
・ Identification and rescue method of assumed problems
・ Feasibility of comprehensive maintenanceThe crushed small-diameter fueldebris is collected with the suction
deviceCutting wheel
grindstone
Bucket

200 mm or more
fuel debris piled
up withoutadhesion

The large fuel debris amongst theones crushed with the splitting
implement are picked up

The picked up fuel debris is broken
down with the cutting wheel
grindstone

The debris is crushed to a size thatwill fit into a unit can and thenstored in the unit can

Example of reactor bottom work step diagram©International Research Institute for Nuclear Decommissioning7. Implementation details of this projectNo.161[2) (i) Technological development of fuel debris collection and storage systems]b. Development of technology for transferring fuel debris and wastes■ Planning the collection methods according to the types of fuel debris inside the RPVInvestigation of the processing capability of the processing technologies being considered
The cutting process is carried out by using the tools normally suitable for metals or ceramic respectively.
Based on the estimation of the conditions inside the reactor, element tests were carried out for the cutting process using the same
tools for composite materials containing metals and ceramics, and the issues were identified.
Tools

Targets

Issues➢ Possibility of cutting the fuel rods or stubs containing a

HydraulicRemaining fuel rods orcutterfuel rod stubs

mixture of both metals (claddings) and ceramics (fuelpellets), and understanding the behavior during cutting➢ The effect on the behavior during cutting due to thevariations in the shape of the cutting blade➢ Can the fuel-rich fuel debris containing both metals(internal structures) and ceramics (molten fuels) be
Cutting wheel
grindstone

Fuel-rich fuel debris

processed with the same tools, under the same processing
conditions and at the same time?
➢ Tool life when metal and ceramics are processed at thesame time➢ Can the metal-rich fuel debris containing both metals(internal structures) and ceramics (molten fuels) be
Chip Saw

Metal-rich fuel debris

processed with the same tools, under the same processing
conditions and at the same time?
➢ Tool life when metal and ceramics are processed at thesame time

Test details➢ Possibility of processing withhydraulic cutter➢ Identification of issues during the
cutting of fuel rods
➢ Possibility of processing with a
cutting wheel grindstone➢ Processing speed and tool life
➢ Issues when processing with cutting
wheel grindstone➢ Possibility of processing with chipsaw
➢ Processing speed and tool life
➢ Issues when processing with a chipsaw

©International Research Institute for Nuclear Decommissioning7. Implementation details of this projectNo.162[2) (i) Technological development of fuel debris collection and storage systems]b. Development of technology for transferring fuel debris and wastes■ Planning the collection methods according to the types of fuel debris inside the RPVInvestigation of the processing capability of the processing technologies being considered (Hydraulic cutter)⚫

Used tools and test apparatus• Commercially available handy hydraulic cutter was used.
Test specimen

Test apparatus

One orthogonally cut test
specimen

3x3 orthogonally cut test specimen

Could be cut (10 secs required)
The blade tip was damaged anddeformed.
Zirconia was crushed into pebblelike pieces.

Of 9 rods, only 1 rod could be cut.
4 rods out of the remaining 8 could not be
cut although zirconia was crushed.
The blade tip was damaged anddeformed.

Could be cut (8 secs required)
There was no significant damage tothe tip of the blade.
Mullite was crushed into powder.

Could be cut (10 secs required).

Max. 285 mm

Zirconia

Soundness andhardness are
assumedHydraulic cutter
50 MPaTest pieces⚫

Blade tip of the hydraulic cutter(LUKAS LSI240)

Test specimen••

Application to the residual fuel and the fuel rod stubs was
assumedThe fuel rods were simulated. One in nine rods were compressed
bydifferent strength
Zirconia/ Mullite(φ10 mm)

MulliteAssumed to
become brittle
under harshconditions

SUS304(φ13 mm t1 mm)
Reference: Compressive strength of ceramics

3 x 3 test specimen

Zirconia

Approx.4000 MPa

MulliteApprox.160 MPa(Reference) UO2

Approx.1000 MPa⚫
⚫
⚫

Minor damage caused near the base of
the blade.

It was confirmed that processing is possible with the hydraulic cutter, since the cladding
was cut by shearing while the fuel was crushed due to the compressive strength.
It is estimated that the hydraulic cutter is effective when the fuel is brittle.
In case the fuel is not brittle, then it is estimated that initially crushing the fuel pellets withbucket-like crushing tools etc. and then carrying out the processing with a combination oftools, is effective.
©International Research Institute for Nuclear Decommissioning7. Implementation details of this projectNo.163[2) (i) Technological development of fuel debris collection and storage systems]b.

Development of technology for transferring fuel debris and wastes■ Planning the collection methods according to the types of fuel debris inside the RPVInvestigation of the processing capability of the processing technologies being considered (Cutting wheel grindstone)
Used tools

Two types of diamond abrasive coating tools for ceramic processing
A vertical slicing device was used for precision processing

Diamond abrasive
coating

Diamond abrasive
coating
Electroplated
bladeElectroplated blade
Metal blade⚫

Metal blade
Base metal

Base metal

Abrasive coating
strength

Abrasive
coatingFeatures

StrongSingle layer

Abrasive coating is strongly retained

Somewhat strong

Multi-layer

If there is wear and tear, the lower layer of theabrasive coating is exposed.

Tool diameter

200 mmFeed speed

4 mm/minRotating speed of tool

2000 rpm

Cutting fluidNot used and only a rust inhibitor added

Target processing depth

17 mm

VerticalElectroplated blade••

Horizontal

State of abrasive coating
after cutting(microphotography)

If processing is continued, the abrasive
coating is lost and this leads to clogging.

An abnormal noise was generated during processing
and the processing stopped. Abrasive coating is lost
and this leads to clogging.

Processing ability is excellent.
Due to the wearing out of the grindstone, thediameter shortens.

Processing ability is excellent.
Even if the grindstone wears out, processing
ability continues.

Test specimen••
Feeddirection

Four laminated layers of SUS304 (t5 mm) plates and zirconia plates (t5mm)
Changes in the direction of lamination and processing (notch)

Metal blade⚫

Rotational
direction
Vertical⚫

Cutting
position

Horizontal⚫
⚫

It was confirmed that the electroplated blade and the metal blade could be used to cut
the test specimen laminated with metal and ceramic even when the cutting fluid was
not used.
It is estimated that the tool life of the metal blade is longer that is approx. 19,000 mm 2.
Processing speed (cross-sectional area) is 80 mm2/min.
©International Research Institute for Nuclear Decommissioning7. Implementation details of this projectNo.164[2) (i) Technological development of fuel debris collection and storage systems]b.

Development of technology for transferring fuel debris and wastes■ Planning the collection methods according to the types of fuel debris inside the RPVInvestigation of the processing capability of the processing technologies being considered (Chip Saw)⚫

Used tools•

Test
specimen

Commercially available dry cermet cutter was used.

Feed 12 mm/min

Feed 60 mm/min

油圧カッターFeeddirection

Tool rotation

Metal-rich

試験装置
Test
specimen⚫

Cermet cutter

Possible to process metal, zirconia wasfragmented.

There were notches in both zirconia and metal,
but the cutter was damaged as well.

Zirconia cracked, metal had notches and thechip corners were lost.

Zirconia cracked, chipcorners were lost and thebase metal of the cutter
was also damaged.

Test specimen•

Simulation of the melted solidified fuel (zirconia) andsimulation of the laminated structure of metal(SUS304)

Zirconia
2 mm thick
SUS3044 mm thick
Metal-richCeramicrich

SUS3041 mm thick
Zirconia
5 mm thick
Ceramic-rich

Idling speed

3500 rpm

External diameter ofsaw blade

185 mmThickness of saw
blade

1.6 mmNumber of blades

48

⚫

⚫
⚫

It was confirmed that cracks developed at the locations hit by the blade and with further
progress, the zirconia plate cracked.
The tip of the blade was damaged by the collision of zirconia and the blade.
The processing ability with respect to stainless steel decreased due to the damage tothe tip and burrs developed.

©International Research Institute for Nuclear Decommissioning7. Implementation details of this projectNo.165[2) (i) Technological development of fuel debris collection and storage systems]b. Development of technology for transferring fuel debris and wastes■Study of the fuel debris processing and collection procedures
Investigation of the processing capability of the processing technologies being considered - Summary

Tools

Test details

Hydrauliccutter➢ Possibility of processing theassumed fuel debris (especially,
the fuel rods that remained
unmelted)➢ Identification of issues at thetime of cutting the fuel rods

Cuttingwheel
grindstone➢ Possibility of processing theassumed fuel debris (especially,
the melted and solidified fueldebris)➢ Processing speed and tool life
➢ Issues when processing with a
cutting wheel grindstone

Chip Saw➢ Possibility of processing of theassumed fuel debris (especially,
the melted and solidified fueldebris)➢ Processing speed and tool life
➢ Issues when processing withchip saw

Test results
⚫

⚫

⚫

⚫
⚫
⚫
⚫

⚫

Identified issues

Claddings were crushed by shearing (plastic
deformation) and fuel was crushed bycompressive strength. The required time was a
few seconds up to about ten seconds
When hardness level of the fuel is high, it isdifficult to cut multiple fuel rods together. Study of
the work procedures etc. is necessary.

⚫ Cutting process for objects thatare not brittle➢ Procedures to be combined withcrushing➢ Improvement of cutting power

It was possible to cut the test specimen
consisting of laminated metal and ceramic. The
melted and solidified debris consisting of metals
and fuels mixed together could be processed.
The processing speed is slow at 80 mm2/min fora cross-sectional area.
Tool life of the metal blade is longer.
It is estimated that approximately 19,000 mm2 ofarea can be processed with the metal blade.

⚫ Procedure to cut out block-likedebris⚫ Processing support technology
such as position controls thatemulates the unevenness of thefuel debris⚫ Size or quantity of abrasive
coating, types of bonds,
introduction of slits, etc.

In case of ceramics, with the development of
cracks, the processing moved forward with thecrushing of the material, but it was found that the
tip of the blade also gets damaged similarly.
Metallic bodies could be processed without
trouble, but the processing ability with respect tostainless steel decreased due to the damage tothe tip and burrs developed.

⚫ The ability to process ceramics islow, and the grinding stone ismore applicable to fuel debris
mixed with melted and solidified
UO2.

©International Research Institute for Nuclear Decommissioning7. Implementation details of this project[2) (i) Technological development of fuel debris collection and storage systems]b.

No.166Development of technology for transferring fuel debris and wastes■ Planning the collection methods according to the types of fuel debris inside the RPV⚫

Summary and future policies⚫ An imaginary diagram was created estimating the conditions inside the RPV putting together the information on Units 1 to 3.
•
Information about the existing understanding of conditions inside the reactor and the fuel debris characterization project.
•
The summarized information was used in the study of methods for retrieving fuel debris inside the RPV.
⚫ Selection of collection methods for each type of fuel debris classified on the basis of shape, presence/ absence of melting, etc.
•
Study of the presumed operating procedures by planning the collection devices to be used inside the reactor based on
the collection methods⚫ Execution of preliminary tests to verify the feasibility of the hydraulic cutter and the disc saw•
Assessment of the general scope of applicability by verifying the applicability to the processing of metals or ceramics by
the tools used
⚫ Issues for the future (FY2020)•
Summarization of issues on the basis of the concept of fuel debris retrieval devices and operating procedures planned
currently.
•
Study of the assumed problem events such as falling of equipment or falling of fuel debris during transport, etc.,
and study of action policies for the same•
Transportation to the container in which the fuel debris will be stored and transportation of the container•
Impact on the safety systems (outflow of chips, utility (water and gas) supply and criticality monitoring procedures)•
The tools on the tip coming in contact with fuel debris were studied in FY2019 and will be summarized in future in acomprehensive configuration diagram.
•
Incorporation of the maintenance policies that are the outcome of the previous year and their reflection in thedevice maintenance plans•
Incorporation of the device maintenance plans and the response policies for the problem events, and theirreflection in the layout plans•
Study of cell installation and construction procedures to establish the overall configuration•
Calculation of the processing abilities necessary for the device plan and for the achievement of goals etc. when
the plan is applied, and evaluation of conformity to the desired values©International Research Institute for Nuclear Decommissioning7. Implementation details of this projectNo.1672) Development of Technology for Processing Fuel Debris(ii) Technological development for the treatment of fuel debris and deposit① Removal technology for soluble nuclides in circulating cooling water
As regards the removal technology for soluble nuclides believed to flow into the circulating
cooling water from the fuel debris, optimal adsorbents were selected after carrying out element
tests related to the adsorption technology for the removal of soluble nuclides. The element tests
were conducted for alpha nuclides. The conceptual system design of the adsorption removal
equipment was carried out based on these results.
The main development study items included the following, and as and when necessary, issues
were identified and organized by conducting element tests.
a.Conceptual system design of soluble nuclide removal facility
・The preliminary test necessary for drafting the plan for the adsorption test using alphanuclides (U and Pu, etc.) is underway.
b.Conceptual design of boric acid conditioning facility
・ The deposition behavior assessment test of boric acid ions is underway.

©International Research Institute for Nuclear Decommissioning7. Implementation details of this projectNo.1682) Development of Technology for Processing Fuel Debris(ii) Technological development for the treatment of fuel debris and deposit① Removal technology for soluble nuclides in circulating cooling water
The conceptual diagram of the contaminated water treatment system for the retrieval of fuel debris is shown below.
: Facilities of these items were tested in this project.
Boric acid (Silo)

Item: ②a. Characterization of collected wastewater
CoolerSoluble nuclide
removalParticulateremoval

Pump

Boric acid solution
tank＋pump

Pump

Tank

Boric acidconditioning
facility

Adjusted boric acidTo water treatment
facility/ tank

Supernatant liquid

Drain water/
Backwash water

Sludge collection

SludgeItem: ①b. Conceptual design of boric acidconditioning facility
Waste treatment facility

Item: ①a. Conceptual system design of soluble nuclide removal facility
Ground water

Particulateremoval

Pump

Soluble nuclide
removalDrain water/
Backwash water

Item: ②b. Development of technologies forseparating solid materials from collected liquids,
wastewater and storing them.

Sludge collection

To water treatment facility/ tank

Supernatant liquid
SludgeWaste treatment facility

Note: This conceptual diagram is an
example of equipment configuration (Whenthere is no leakage from PCV)

Figure. Liquid system during fuel debris retrieval (conceptual diagram) and implementation targets for element tests©International Research Institute for Nuclear Decommissioning7. Implementation details of this projectNo.169[2) (ii) ① Removal technology for soluble nuclides in circulating cooling water]a.

Conceptual system design of soluble nuclide removal facility⚫

Purpose of development
To acquire data in order to select the adsorbent types and amount used for theadsorption removal of alpha nuclides (Pu, U, Np, Am, Cm).

⚫

Issues to be solved
The data regarding the nuclide adsorption capability of the alpha nuclide
adsorbent used at the soluble nuclide removal facility, is insufficient.

⚫

Approaches to development(1) Preparation for element tests① Organization of water quality assumed for actual operations② Study of the form of existence of the alpha nuclides for the assumed water
quality③ Literature research on the adsorbents for the adsorption removal of alphanuclides other than Am (Pu, U, Np, etc.), and selection of the candidate adsorbent④ Study of methods to test adsorption (conditions and procedures)(2) Preliminary tests required for element test planning② Study of the form of existence of the alpha nuclides for the assumed water
quality(3) Adsorption tests using alpha nuclides (U, Pu, etc.) (Immersion test andcirculation test)<<Example of the test to verify the nuclide adsorption ability>>
1) Immersion test
• Establishment of high accuracy
test procedures•
Execution of test for Pu, U etc.

2) Circulation test
• Establishment of a test
procedure which can reduce thequantity of nuclides used
• Execution of test for Pu, U etc.

Measurement of the adsorbed
amount by measuring theconcentration of adsorbed
substances in a canister
before/after starting the test

Understanding the initial DF bymeasuring the changes occurringin the concentration of adsorbed
substances with the lapse of time

Adsorbent A

Adsorbent B

Metering pump

Sample foranalysis⚫

Simulated raw
water

Expected outcome
Data for selecting the alpha nuclide adsorbents to be used at the soluble nuclide removal facility©International Research Institute for Nuclear Decommissioning7. Implementation details of this project[2) (ii) ① Removal technology for soluble nuclides in circulating cooling water]a.

Conceptual system design of soluble nuclide removal facility◆ Study flow

The initiatives until FY2018, the initiatives in this project, and the initiatives after this project
are shown in the following flow.

[Up to FY2018]

・Applicable candidate adsorbent (Am)・Necessity of test procedure review
・Necessity of evaluation of nonevaluated nuclides

Conceptual study of theliquid phase system

In FY2019, the preconditions for the adsorption tests (for theselection of the candidate adsorbent) were mainly organized withpreliminary tests, etc.

[FY2019 to FY2020 (This project)]
A: Adsorption test plan

No.170[FY2021 onwards][Results in FY2019]
Test plan drafted. Approach toadsorption test priority organized.

・ Test procedures (After review)
・Adsorbent to be evaluatedAdsorption performanceverification test using
the actual liquid #3

Exposure evaluation

Large share of Pu andAm
Necessity of removal ofsoluble alpha nuclides

B: Preliminary test
Selection of the applicablecandidate adsorbent by lab
tests (Am)・ALARA・ (Discharge to water treatment
facility)
Temporary settings ofthe assumed water
quality during actual
operations
Refer to the water quality ofstagnant water inside thePCV of Units 2 and 3
Narrowing down of theapplicable candidate
technology[Results in FY2019]
The preliminary test for Am,
Cm, and U are executed. The
feasibility estimations of theAm, Cm adsorption test andthe rising tendency ofsolubility under boric acidsolution conditions, areconfirmed.

・Test procedures・Adsorbents tobe evaluatedAdsorption test plan(Am)

Selection of RO membrane removal
and adsorption removal. Selecting
an appropriate adsorbent isnecessary.

・Form of existence of thealpha nuclides in the liquid
phase・Container adhesion effect

C:Selection of the applicablecandidate adsorbent by lab
tests (Pu etc.)

・Applicable adsorbent
・Differential pressure
data during long-termwater flow
・Removal performance
during long-term waterflow・Adsorbent replacement
frequency (amount ofwaste generated)

Applicable candidate
adsorbent #1

System conceptual
designLab test result

D: System conceptual study #2#1: It is necessary to conduct a performance verification test
which finally uses the actual liquid for the selection of theadsorbent. Under this project, the policy is not to narrow down toa single candidate adsorbent, but to make multiple selections.
#2:The study of a conceptual system and the identification ofissues shall be performed assuming the application of theadsorption system.

#3: It is considered necessary to
conduct tests using the retrieval of fueldebris from the first Unit, which isplanned to start in 2021.

©International Research Institute for Nuclear DecommissioningNo.171

7. Implementation items of this project[2) (ii) ① Removal technology for soluble nuclides in circulating cooling water]a. Conceptual system design of soluble nuclide removal facility◆ Development Schedule

Study Items
Major milestones

FY2019
４

5

6

7

8

9

10

11

Interim Report▼

FY2020
12

1

2

3

4

Interim Report
▼

5

6

7

8

9

10

11

Interim Report
▼

12

1

2

3

Final Report▼

1.Conceptual study
2.Planning for elementtests
2- (1).Preparation forelement tests (Includes
making arrangements fortest supplies)
2- (2).Preliminary tests
required for element testplanning

3.Element tests

Preliminary tests using Am / U, etc.

Adsorption tests for U / Pu, etc.

4.SummaryRemarks©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[2) (ii) ① Removal technology for soluble nuclides in circulating cooling water]

No.172a. Conceptual system design of soluble nuclide removal facility◆ Selection of a candidate adsorbent: The candidate material was selected based on a literature research and the form ofexistence of alpha nuclides in the solution.
① A literature research was conducted and the candidate adsorbent for each nuclide element to be removed was selected.
AlphanuclidesCandidate adsorbent

PuAm

Approach to selection (literature research)

Clinoptilolite

In the test results using actual underground water, Kd was present to the order of 10 3

Chabazite

In the test results using actual underground water, Kd was present to the order of 10 2

Clinoptilolite

In the test results using actual underground water, Kd was present to the order of 10 3

Crystalline silicotitanates

Cm

Not applicableThough there is literature on adsorbents evaluated in the acidic region, there are no research results evaluated near neutrality

MordeniteU

In the test results with NaCl (10-3 mol/l ~ 10-1 mol/ｌ) pH 6~7, Kd was present to the order of 103

Activated carbon attached toadsorption site

Np

In the test results with 30% diluted seawater (3000 ppm of boron), Kd was present to the order of 10 3

X-type zeolite

In the test results with KNO3 (I = 0.1) pH 7~8, Kd was present to the order of 104
In the test results with 0.1 M NaCl / 0.001 mol/l HCl pH 6, Kd of approximately 800 ml/g was present② Candidate adsorbents were selected considering the existence form of alpha nuclide to be removed
Existence form of alphanuclidesIon

Candidate adsorbent

Selection method

Divalent cation
Molecular ions

Zeolite / Titanic acid /Crystalline silicotitanates

Selected as an adsorbent that can adsorb cations. In addition, a high adsorption rate was confirmed in the
Am test during the last year’s subsidized project.

Trivalent cation

Zirconium phosphate

Colloid

Activated carbon

Selected as an adsorbent that can adsorb trivalent ions.
As it has excellent adsorption of neutral substances, it is expected to remove molecules that reacted withcolloid and organic substances.
©International Research Institute for Nuclear DecommissioningNo.173

7. Implementation items of this project[2) (ii) ① Removal technology for soluble nuclides in circulating cooling water]a. Conceptual system design of soluble nuclide removal facility◆ Preliminary test: Confirms the difference in the degree of dissolution of Am, Cm, and U depending on the assumed
water quality. Pu and Np will be evaluated in FY2020.
Conditions
SS
concentration
[ppm]

<1

Near neutral<1

Stagnant water inside PCV
Stagnant water inside the building

No

No

No

10 ~ 40

About 20

2

Yes

Yes

No

10 ~ 40

About 350 About 40<1

Weak alkali

<1

3

No

No

Yes

10 ~ 40

About 20

7,000

Weak alkali

<1

4

Yes

Yes

Yes

10 ~ 40

About 350 About 40

7,000

Weak alkali

<1

5

No

Yes

No

10 ~ 40

About 20

<1

High

<1

1.0

Preliminary
test results
using Cm

1.00

0.96

<1

pH

1

<1

Saturated

0.97

0.8

0.6
0.4

0.41

The concentration of dissolved
Cm is high in solution containing
sodium pentaborate and in
solution assuming the water
stagnant inside the building

0.2

Percentage of U remaining insolution (after ultrafiltration)

Conditions for
conducting
preliminary
tests

Addition
Test
Seawater Concrete of boric
Cl ionCa ionBoric acidsolutioncomponent component acid Temperature concentration concentration concentration
conditions[℃]
solution
[ppm]
[ppm]
[ppm]

Percentage of Cm remaining insolution (after ultrafiltration)

List ofactual andassumedwater
qualityconditions

Preliminary
test results
using U

1.0

Assumptions

Stagnant water inside PCV (boric acidsolution added)
Stagnant water inside the building (boric
acid solution added)
Elution of concrete components into
stagnant water inside PCV

0.92

0.98

0.8

The concentration ofdissolved U is high insolution containing sodium
pentaborate

0.6
0.4

0.34

0.2
0.06

0.0

0.0
条件１ 1
Condition

条件２ 2Condition

条件３ 3Condition

条件４ 4Condition

条件１
Condition
1

条件２
Condition2

条件３
Condition
3

条件４
Condition
4©International Research Institute for Nuclear DecommissioningNo.174

7. Implementation items of this project[2) (ii) ① Removal technology for soluble nuclides in circulating cooling water]a. Conceptual system design of soluble nuclide removal facility◆ Adsorption test: Adsorption test is being conducted after prioritizing the tests based on the desk evaluation andpreliminary test results.

• Pu, Am, and Cm, that have a high effect of exposure and a low regulatory concentration limit, have a high priority for adsorption
removal.
• Seawater components and boric acid-added conditions, that were suggested to increase alpha nuclide concentration in solution inpreliminary tests, have relatively high priorities.
• The coexistence of Cs and Sr may reduce the alpha nuclide adsorption performance, so it has the second highest priority. However,
U and Np, which have low solubility, have a relatively low priority.
• Although the elution effect of Ca has a high priority for data collection, it is difficult to evaluate Cm and Am as their carriers get
sedimented with the increase in pH. The possibility of evaluation for U, Np, and Pu will be confirmed in the preliminary tests in the
future.
• The initially planned water flow test is expected to be excluded by conducting high-priority tests.
Immersion test (confirmation of adsorption capacity at equilibrium)
Alpha
Coexisting ion effectnuclide Circulating
Ca
Carbonic
Seawaterelements coolingCs
Srwater components (concrete)acid

Boric acid

OthersWater flow test
DF
DuringDuring Adsorbent
life
initial water long-termflowwater flow

U

△/x

△/x

(▲/x)

x

(▲)

(▲)

○

▲

▲

x

x

Np

(△/x)

(△/x)

(▲/x)

x

(▲)

(▲)

(○)

▲

▲

x

x

Pu

(◎)

( ○)

(○/x)

x

(△)

(△)

(○)

▲

▲

x

x

Am

◎

○

x

x

△

△

○

▲

▲

x

x

Cm

◎

○

x

x

△

△

○

▲

▲

x

x

・Legend ◎: High priority, ○: Relatively high priority, △: Relatively low priority, ▲: Low priority, x: Test evaluation is difficult in this PJ
・ The parentheses indicate assumptions since these tests were conducted before the preliminary test.
©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[2) (ii) ① Removal technology for soluble nuclides in circulating cooling water]

No.175b. Conceptual design of boric acid conditioning facility⚫ Purpose of development
As a safety system, it was necessary to organize the design requirements (equipment needs) essential for
criticality control using soluble neutron absorbers and to determine the specific details of the process.
The data on physical properties when using sodium pentaborate (physical properties of sodium pentaborate,
deposition behavior, migration data during evaporation, etc.) will be obtained and the design requirements will
be organized.
⚫ Issues to be solved
To obtain the data on physical properties such as deposition behavior of sodium pentaborate.
To organize the design requirements for criticality control systems considering the data on physical properties of
sodium pentaborate.
To study the requirements for operation and maintenance of the criticality control systems under the actual
environment (dose, etc.).
⚫ Approaches to development
Regarding the behavior of sodium pentaborate, literature research and element tests (test assuming actual
operations if necessary) will be performed, and the data necessary for design will be acquired.
The requirements for criticality control systems will be organized using the acquired data.
The amount of boric acid migrating to the destination (equivalent to the existing water treatment facility) will besuggested and the impact (element test etc. shall be conducted if necessary) will be confirmed.
⚫ Expected outcome
Physical properties data required for actual equipment design
Concept of criticality control systems using sodium pentaborate
Amount of boric acid migrating to the destination and its impact©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[2) (ii) ① Removal technology for soluble nuclides in circulating cooling water]

No.176b. Conceptual design of boric acid conditioning facility[Background: Overview]
・ In the subsidized project (Criticality Control Project), initiatives are being taken for development of technology forcases using both soluble and insoluble neutron absorbers as the criticality control methods for retrieving fuel debris.
・ As technology developments of Criticality Control Project (FY2018), a system configuration plan was created, issues
were identified and studied after tentatively determining the preconditions including the liquid phase system, as thesystem using soluble neutron absorbers is closely related to the liquid phase system (circulation loop system).
・ In continuation to last year, this project will continue to carry out studies aimed at solving the identified issues.
・ Identified issues (see next slide for details)
Issue ① : Deposition of sodium pentaborate
Deposition of boric acid inside the system (PCV, circulation loop, boric acid conditioning facility) and its
effect on the systems
Issue ② : Amount of sodium pentaborate and pH regulator to be injected
Loss of boric acid inside the system (PCV, circulation loop, boric acid conditioning facility) (increased
burden on the boric acid injection system)
Issue ③: Amount of sodium pentaborate migrated
Amount of boric acid migrating to the destination and its impact
Issue ④: Operation and maintenance policy of boric acid conditioning facility
Organizing the operations and maintenance policies under actual operating environment based on issues① to ③©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[2) (ii) ① Removal technology for soluble nuclides in circulating cooling water]

No.177b. Conceptual design of boric acid conditioning facility[Challenges of a pentaborate operating system]
Issue ② Amount of sodium pentaborate and pH regulator to be injected
Loss of boric acid inside the systems (PCV, circulation loop, boric acidconditioning facility) (increased burden on the boric acid injection system)•Confirmation of deposition conditions, confirmation of boric acid adsorption
amount in nuclide removal (confirmed in the previous section a.)•Evaluation on the amount of loss of boric acid

Issue ① Deposition of sodium pentaborate
Deposition of boric acid compounds inside the system (PCV,
circulation loop, boric acid conditioning facility) and its effect on thesystems• Organizing the deposition conditions for sodium pentaborate withrespect to the assumed water quality at each location (literature
research, element test)• Study on the effects of boric acid deposition and countermeasures
for the same
(Study of operating condition range where there is no deposition)

Emergency boric acidsolution injection system
Sodium pentaborate(powder)

Silo

Treated water
Pentaborate solution
supply tank

Emergency cooling water
circulation system
Wastewater
receiving tank

CoolerSilo
Cooling water
circulation system

Boric acid solution
injection system

ParticulateremovalpH adjusted water, etc.
SolublenuclideremovalTorus room drainage
systemParticulateremovalSodium pentaborate(powder)
Treated water

CoolerPump

SolublenuclideremovalWastewater
receiving tankPentaborate
solution supplytank

Conditioning
tank

Boric acidconditioning facility
Torus room
wastewater
receiving tankConcentratecanister unit

Concentratereceiving tank

Pump

Condensate
receiving tank
Pump

Issue ③ Amount of sodium pentaborate
migrated
Amount of boric acid migrating to thedestination and its impact• Confirmation of boric acid recoverability(element test)• Checking the amount migrating to thedestination and its impact

Line of Emergency
cooling water circulation
system: not shown

Treated water(discharge)

Groundwater

Issue ④ Operation and maintenance policy of boric acidconditioning facility
Operation and maintenance policy under actual operating
environment• Confirmation of DF performance in boric acid environment(confirmed in the previous section a.)©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [2) (ii) ① Removal technology for soluble nuclides No.178
in circulating cooling water]b. Conceptual design of boric acid conditioning facility○ Workflow
To prevent insufficient data at the system design stage, the basic physical properties of sodium pentaborate will be organized basedon the results of studies conducted in the subsidy projects so far and additional literature research, and the missing data will becollected through element tests.
1.Organizing the preconditions

FY2019 toFY2020Study and investigationof the issues(first half)① Organizing the preconditions for using sodium
pentaborate solution(a) Organizing the criticality control plan(b) Organizing the assumed conditions(c) Creation of system function diagram

2.Identifying the test items② Creation of list of properties of sodium pentaborate
solution(a) Literature research (includes organizing the results ofpast subsidy projects)(b) Creation of physical property list
[Preliminary test]
Confirmation of missing
physical properties③ Material balance evaluation under the conditions assumed in 1F(Identifying the concerns)(a) Conditions for material balance evaluation(b) Results of material balance evaluation④ Study on measures for addressing theconcerns
Study and investigationof the issues

FY2020Organizing therequirement
specifications for
operations[Preliminary test]
Confirmation of evaluation
conditions / evaluation results
related to the concerns⑤ Organizing the insufficient data / technical issues (test items) from the viewpoint of 1F design
3. Drafting the element test plan / 4. Element tests⑥ Element tests(a)Setting of test conditions ….Setting the conditions based on the results of material balance evaluation.
(b)Drafting of testing method ….Drafting the testing method using the literature research results in ①(c)Element tests⑦Creating the list on physical properties of sodium pentaborate solution for the 1F design

Goal

Implementing the important processes required in criticality control using soluble neutron absorbers: Organizing the design requirements (equipment needs),
determining the specific details for the configuration of facilities©International Research Institute for Nuclear DecommissioningNo.179

7. Implementation items of this project[2) (ii) ① Removal technology for soluble nuclides in circulating cooling water]b. Conceptual design of boric acid conditioning facility○ Future Plans<Plan for FY2020>• Evaluation of deposition element test results using Ca and B deposits and Na/B ratio change.
• Additional tests shall be performed if required.
• At the same time, countermeasures for the deposits shall be studied.

○ Schedule
FY2019Study Items
４

Major milestones

5

6

7

8

9

10

FY2020
11

Interim Report
▼

12

1

2

3

Annual report
▼

4

5

6

7

8

9

10

11

Interim Report
▼

12

1

2

3

Final Report▼

1.Organization of
preconditions
2. Identification of test items(Includes conceptual study/
investigation of issues)
3.Drafting the element testplan
4. Organizing the element
test/ results
5.Conceptual design

6.Summary©International Research Institute for Nuclear DecommissioningNo.180

7. Implementation items of this project [2) (ii) ① Removal technology for soluble nuclides incirculating cooling water]b. Conceptual design of boric acid conditioning facility① Organizing of preconditions for using sodium pentaborate solution● Organizing the Criticality Control Plan
・ Safety requirements
Prevention of generation of abnormal radioactive materials due tonuclear reaction
・Functional requirements
Level 1: Maintaining subcriticality within the control criteria values
Level 2: Preventing criticality by detecting the approach to criticality of
the assumed abnormality and stopping the retrieval operation
Level 3: Detecting criticality and stopping the operation immediately.

・ Means of achievement
Maintaining subcriticality using neutron absorber (sodium pentaborate
solution) and stopping abnormalities● Organizing the assumptions● Creation of system functions diagram

To organize the material balance evaluation
conditions assumed in 1F based on thepreconditions➢ Example of assumed conditions (around PCV)
Processed water
When sodium pentaborate is used as the neutron absorber, it is assumed that
sodium pentaborate is also used for the processed water.
The maximum flow rate is set to 2.2 m3/h.

Unit 3Presenceof spentfuelLeakage from PCV (D/W and S/C) to torus room
Assume two patterns, one with leakage and one without leakage. The maximum leakage
amount is set to 3 m3/h.
(Assuming the maximum leakage amount based on the study conditions for the liquid phasesystem last year)

Processedwater
Groundwater

In addition to studying the development issues in using sodium
pentaborate solution, the following was studied for organizing thepreconditions.

Water level
inside PCV :About 6 m

Groundwater
Assume that a certain amount will flow in (This will have an impact if there is
PCV leakage)
The maximum flowrate is set to 2.1 m3/h.
(Follow the preconditions of last year's liquid phase system)

Access routefor operators

Fuel debris(3) Water level
The water level is assumed to be equal to or above a certain level.
(Assumed to be at least 30 cm more than the bottom of the D/W in Unit 2)

Leakage points (visual confirmation)➢ System functions diagram[Example]

Prerequisite: Presence oflarge pH fluctuation⇒Required function: Add
a pH regulator to maintainthe form of sodium
pentaborate

Boron injection(Concentration control)
Boron injection (endof critical time)

Nuclideremoval(Ensure DF)

Analysis andadjustment(storage tank)

Nuclide removal(Ensure DF)

pH regulator injection(pH control)

Removalof factors
inhibiting
operation

Boron collection(evaporative
separation)

Analysis(Storagetank)

Existing
treatment
facility

Prerequisite:
Presence of PCV
leakage=> Required
function: Remove
the Cl originatingfrom groundwater
Prerequisite: Presenceof PCV leakage⇒Required functions:
To concentrate theboric acid diluted withgroundwater©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[2) (ii) ① Removal technology for soluble nuclides in circulating cooling water]

No.181b. Conceptual design of boric acid conditioning facility② Organizing the physical properties of sodium pentaborate solution
Studies on the physical properties of sodium pentaborate solution and literature research on
deposition were carried out.
● Study results(i) Ca ion: Possibility of formation of deposits in coexistence with approximately 200 mg/L of Ca
ions(ii) Mg ion: Possibility of formation of deposits in coexistence with approximately 350 mg/L of
Mg ions(iii) Na/B ratio: Possibility of formation of deposits when the NaB ratio (molar ratio) deviates
Cooling waterfrom the range
of of0.12
toions0.43
sodium pentaborate it is usually
circulation0.2)
systemElutioncalciumfrom(forconcrete

Change in Na/B ratio(Changes in Na/B ratio may occur in nuclide removal equipment,etc.)
Na ionBorate ion
Ca ionCa adsorption
increases Na ionin the solution

CoolerBorate ion

Ca ion(Eluted from the new concrete surface)

Crushed concrete pieces(Generated during fuel debris retrieval)Removal ofsolubleneutrons

ParticulateremovalTorus room
drainage system

Ion Exchanger

Mixing of groundwater components

Removal ofsolubleneutrons

ParticulateremovalFormation of deposits due tochanges in Na/B ratio

Ca ionBorate ion

Mg ionNa ionConcrete

K ion
Groundwater

Formation of deposits
with Borate ion

Cl ionSO4 ionGroundwater components

Formation of deposits
with Borate ion

Pump

Figure. System during fuel debris retrieval, and contaminants from concrete and groundwater③ Material balance evaluation under the conditions assumed in 1F
Based on the preconditions, the conditions for material balance evaluation assumed in 1F are organized,
and material balance evaluation is carried out for items of concern such as usage amount of B.
Determine the need for data acquisition by evaluating the level of impact, and deploy it in the test plan.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[2) (ii) ① Removal technology for soluble nuclides in circulating cooling water]

No.182b. Conceptual design of boric acid conditioning facility④ Study on measures for addressing the concerns⑤ Organizing insufficient data / technical issues (test items) from the viewpoint of 1F design
Based on literature research, etc., the initiating events, event progression and issues, countermeasures, and test items
related to the concerns that may be expected when using sodium pentaborate solution, were organized.
● Organizing the results related to deposition

Initiating
eventsPossible event progression

Issues, countermeasures and test items

Cutting of
fueldebrisElution of Ca ions(Generated from concrete)

Deposition reaction of Ca ion and Borate ion

Contamination ofgroundwater

Increase in Na ions
➡ Change in Na/B
ratio

Deposition due to change in form of borate ion insolution (formation of low solubility compound)

Form

Chemical formula

Solubility (g/L as
B)

Sodium pentaborate

Na2B10O16・10H2O

49 (30℃)

H3BO3

12 (30℃)

Na2B4O7・10H2O

7.8 (30℃)

Boric acid
Sodium tetraborate

Increase in Ca / Mg ions

Deposition reaction of Ca/ Mg ions and borate ions

[Issues]
・Reduction of boric acid concentration
・Blockage due to deposition
・Impact on system equipment(increased burden on filter)(Degraded performance of coolers, etc.)[Countermeasure]・Prevention of deposition by operation
management→ Control of Ca ions/ Mg ions
→ Adjustment of Na/B ratio
・Removal of deposits[Test items]
・Checking the criteria for operations management→ Allowable limits for Ca ion and Mg ion→ Allowable limit of Na/B ratio・Checking the removal method of the deposits→ Composition of the deposits, etc.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[2) (ii) ① Removal technology for soluble nuclides in circulating cooling water]

No.183b. Conceptual design of boric acid conditioning facility⑥ Element test
[Purpose]• To confirm the deposition behavior of sodium pentaborate solution, and study thecountermeasures (Control criteria for preventing deposition, method for maintaining boric acidconcentration, etc.)[Issues]• The Ca ions, Mg ions, etc., contained in concrete may react with borate ions to form deposits.
The impact in this case (concentration range, temperature range, and impact of coexisting ions)is unclear.
• The borate ions may form deposits due to an increase in low-solubility boric acid compounds
due to changes in Na/B ratio. The impact in this case (dissolution / deposition range) is unclear.
[Implementation Details]• Confirming of dissolution / deposition behavior of sodium pentaborate in the event ofcoexistence with Ca ions, Mg ions, and other impurities• Confirming of dissolution / deposition behavior during change in Na/B ratio• Study on the methods for removing deposits• Study on the control methods for impure ions and Na/B ratio[Goal]• Determination of control criteria for concentration of impure ions (specifically Ca, Mg), Na/B ratio, etc.
• Drafting of countermeasures for deposits©International Research Institute for Nuclear DecommissioningNo.184

7. Implementation items of this project[2) (ii) ① Removal technology for soluble nuclides in circulating cooling water]b. Conceptual design of boric acid conditioning facility⑥ Element test
[Test condition range and expected outcome]
Table

Test Conditions

Table

Items

Testing range

B concentration (mg/L as
B)

510 ~ 10000

Na/B ratio

0.06 ~ 1.0

Set based on the literary factof possibility of deposition at
25℃

0 ~ 800

Set based on the saturated
concentration of Ca (OH) 2

0 ~ 800

Set by assuming equivalence
to Ca

Ca ion concentration(mg/L)
Mg ion concentration(mg/L)

Remarks

Diagram on control criteria values

Items

Control values

B concentration (mg/L as
B)

510 ~ 10000

Na/B ratio

Remarks

XX ~ XX

Range where there is no
deposition at 20℃

Ca ion concentration(mg/L)< XX

Range where there is no
deposition at 20℃

Mg ion concentration(mg/L)< XX

Range where there is no
deposition at 20℃
Heater
ヒーター

Floating-type
浮き型
insulation
保温材
material[Testing Method]• Prepare the specified solution and stir it for the specified amount of time and temperature• After the specified time, filter the solution and separate the filtrate, filter and beaker
Test solution
cleaning solution.
供試液• Perform component analysis of filtrate and filter (B, Na, Ca, Mg, etc.)[Evaluation method]• Analyze the target components in the solution and in the collected solids, and determine
the amount of deposition of the added elements and boron components• Find the concentration range of each component where there is no deposition of boron
components.

Water
bath
ウォーター
バス
Stirrer撹拌子
Magnetic
マグネチック
stirrer
スターラー

Figure. Test equipment©International Research Institute for Nuclear DecommissioningNo.185

7. Implementation items of this project[2) (ii) ① Removal technology for soluble nuclides in circulating cooling water]b. Conceptual design of boric acid conditioning facility⑥ Element test

[Effect of Ca ions]• While adding Ca (OH)2 , deposition was confirmed at 380 mg/L or higher Ca.
• The Ca/B molar ratio of deposits was approximately 0.3. Based on the literature information, the deposits are assumed to be
Ca2B6O11･13H2O.
[Effect of Na/B ratio]• Deposition was seen when Na/B ratio was < 0.1.
• Most of the deposits are B, and based on the literature information, the deposits are assumed to be H3BO3.
• The tests are progressing as planned, and the results are also being obtained as per the plan.
<Test conditions>
Solution: sodium pentaborate 6000 mg/L as B
(Testing underway for cases where sodium
pentaborate concentration is changed)
Temperature: 20 ℃

<Test conditions>
Solution: With sodium pentaborate 6000 mg/L as B, adjust Na/B ratio by adding boric
acid or sodium hydroxide(Testing underway for cases where sodium pentaborate concentration is changed)
Temperature: 20 ℃

Appearance of deposits
when 590 mg/L as Ca is
added

40000

Deposits
formed600
400 380 mg/L200
00 200 400 600 800 1000
Concentration of added Ca(mg/L)

Figure.1 Overview of deposition test results byadding Ca

Appearance of deposits
when Na/B ratio is 0.08

Concentration ofdeposits (mg/L)

Ca concentration in
filtrate (mg/L)

800

Deposits
formedNa / B ratio 0.2
Equivalent to sodium
pentaborate

30000
20000
10000
0

0

0.2

0.4 0.6
Na/B

0.8

1

① Na/B ratio decreases ② Na/B ratio increases
Figure.2 Impact confirmation test for change in Na/B ratio©International Research Institute for Nuclear DecommissioningNo.186

7. Implementation items in this project2) Development of technology for processing fuel debris(ii) Technological development of treatment for fuel debris and deposit② Treatment technology for deposits collected from inside PCVWhen collecting deposits containing fuel debris that is present at the bottom of PCV, wastewater containing
deposits is generated. Also, during the fuel debris retrieval operation, wastewater containing solid matter isgenerated when the filter of the circulating water system is backwashed. When handling and storing these, itis necessary to separate the solid components from the liquid from the viewpoint of safety, volume reduction,and the like. Therefore, regarding the treatment technology for separating solid material from the wastewaterand storing it, technology considering storage efficiency, remote operations, maintenance, etc., wasdeveloped and a conceptual system of the wastewater treatment equipment was designed.
The main development study items included the following, and the issues were identified and organized by
performing element tests wherever required.
a. Characterization of collected wastewater・ A differential pressure countermeasure test was partially conducted on an intermediate removal
filter, which is a candidate device for removing insoluble nuclides.
b. Development of technologies for separating solid material from wastewater and storing it
・ As a technology for separating solids from collected wastewater, the applicability of separation
using sedimentation separation is being verified.

©International Research Institute for Nuclear DecommissioningNo.187

7. Implementation items in this project2) Development of technology for processing fuel debris(ii) Technological development of treatment for fuel debris and deposit② Treatment technology for deposits collected from inside PCV
The conceptual diagram of the contaminated water treatment system during fuel debris retrieval is shown below.
: Equipment for testing this item in the project
Boric acid (silo)

Item: ②a. Characterization of collected
wastewater

Pump

CoolerPump

ParticulateremovalSolublenuclideremovalDrain water/
backwash water
Sludgecollection
Groundwater

Pump

Boric acidsolutiontank＋pump

Tank

Supernatant liquid

ParticulateremovalDrain water/
backwash water
Item: ②b. Development of
Sludgetechnologies for separating solid
collection
material from wastewater and storing it

Boric acidconditioning
facility

To water treatment facility /tank

Item: ①b. Conceptual design of boric acidconditioning facility

Sludge Waste treatment
facility

Adjusted boric acid

Item: ①a. Conceptual system design of soluble nuclide
removal facility

Solublenuclideremoval

To water treatment facility /tankSupernatant liquid
Sludge Waste treatment
facility

Note: This conceptual diagram is an example ofequipment configuration (When there is no leakage
from PCV).

Figure. Liquid system during fuel debris retrieval (conceptual diagram) and implementation targets for element tests©International Research Institute for Nuclear DecommissioningNo.188

7. Implementation items of this project[ 2) (ii) ② Treatment technology for deposits collected from inside PCV]
a. Characterization of collected liquids, wastewater⚫ Purpose of development
In order to select the particulate collection equipment to be used in the insoluble nuclide removal facility,
data such as particulate collection rate, differential pressure behavior, and properties of wastewater was
acquired by performing a filter water flow test.
⚫ Issues to be solved➢ Particulate collection technology to be used has not been selected.
➢ Optimization of particulate collection equipment design is required.
➢ There is insufficient data on the backwash performance, etc. of particulate collection equipment for
long-term water flow.
⚫ Approaches to development① Filter load optimization test
② Backwash conditions optimization test
③ Impact assessment test of insoluble criticality inhibitor (conducted as needed)④ Long-term water flow test
⚫ Expected outcome
Design optimization and acquisition of operational data on particulate collection equipment candidates to be used inparticulate removal equipment.
Insoluble alpha nuclide removal equipment (particulate removal equipment)
Stagnant
water
insidePCV

Water containing
particles of size 100
μm or lessRoughremoval
(Largeparticles)< several tens of μm

Drain waterIntermediate
removal
(Large tosmall particles)< several μm

Backwash water /drain water

Final treatment(Fine particles)< Several tens toseveral hundreds of nm

To soluble alpha nuclide
removal equipment

Backwash water

Drain water/ backwash water treatment equipment (Sludge collection)

Figure. Conceptual diagram on the configuration ofparticulate removal equipment system# Described in Section (2) (ii) ② b. Development of technologies forseparating solid material from wastewater and storing it©International Research Institute for Nuclear DecommissioningNo.189

7. Implementation items of this project[2) (ii) ② Treatment technology for deposits collected from inside PCV]
a. Characterization of collected liquids, wastewater➢ Study Flow: In this project, design optimization of the candidate equipment identified in the previous subsidy
project and acquisition of long-term operation data was performed and was reflected in the conceptual systemdesign.
FY2019Previous subsidy project

FY2021This PJ

After FY2021

Study on the liquid-phase particulate removal
concept flow

Optimization of particulate collection
candidate equipment design

Confirming the properties ofactual liquid

Study on technology benchmarking

Acquisition of operational data for long-termwater flow

Conceptual design based on the properties ofactual liquid

Candidate device applicability confirmation
test

Conceptual system design

Basic design and detailed design< several tens toseveral hundreds of nm

Large differential pressure

Insoluble alpha非溶解性α核種除去設備(粒子除去設備)
nuclide removal facility (Particulate removal facility)
StagnantPCV内
water
滞留水
inside PCV
粒径100μm
Water以下の粒子を
with
particlesof size
含む水100 μm or lesser< several tens
to several
hundreds<数十 ofμm(Roughremoval) < (Intermediate
treatment)
粗取り
中取りremoval) (Final
最終処理<

To soluble～数百nm 溶解性
<数μmseveral Particulatealpha nuclide
Particulate removal
粒子除去
粒子除去
μm
removalα核種(Largeto small particles)(大粒子)(大～小粒子)(微小粒子)
removal facility(Fine
particles)
除去設備へ
Treatedwater isBackwash
water
逆洗水/
逆洗水
ドレン水
Backwashwater sent to
Drain
water
Supernatant/ドレン水
drain water
water
上澄液
circulating
処理水はスラッジ回収設備
Sludge collection facility
system
循環系統へ
Debris particles
デブリ粒子
Matterstored in container is
容器収納物はデブリ収納エリア
transferred
to debris storage area
または廃棄物保管エリアへ移送
or waste storage area
Figure. Conceptual flow of liquid-phase particulateremoval systemseveral<数十μm
Particulatetens of
粒子除去removal(Large particles) μm

Table: Test results of FY2018

RoughremovalIntermediate
removalEquipment

Auto strainer
Liquid cyclone

Auto strainer
Inorganic filters#1

Inorganic filters #1

Collection
rate

High

High

High

MinimalLarge(High frequency
of backwash)

Increase indifferential
pressure

Final treatment<Several tens
of μm

Rough
removalequipment

Finaltreatment

Drainbackwash water

Drain
backwash water

Backwash waterSpecific approaches
after FY2021 will
require coordination
with operator’s
engineering department.

① Check whether the increase in differential pressure can be suppressed by changing
the filter pore size to an appropriate value.
② If the increase in differential pressure cannot be suppressed with step ①, try for overall
optimization by distributing the load during final treatment.

INPUT theoptimized
conditions

Figure. Illustration of the study on intermediate removal
optimization

Rough removal (up to several
tens of μm)

Intermediate removal (up to
several μm)
Ｐ

Ｆ

Less

Final treatment (up to
hundreds of nm)
Ｐ

Ｆ

Ｐ

Ｐ

# Ceramic filter
or
Sintered metal
filters#Auto straineror
Sintered metal
filtersor
Bug filter# AutostrainerorLiquidcycloneＦ

Ｐ

Ｐ

Tank for treatedwater
Tank fortreated water

Tank fortreated water
Original water tank

#1: Candidates for Inorganic filters include sintered metal filters and ceramic filters.

< Several μm

Intermediat
e removalequipment

Tank for drained
water

Tank for backwash water
Tank for drained wateror backwash water

Figure. Illustration of long-term water flow test

©International Research Institute for Nuclear DecommissioningNo.190

7. Implementation items of this project[ 2) (ii) ② Treatment technology for deposits collected from inside PCV]
a. Characterization of collected iquids, wastewater◼ Development Schedule

Study Items
Major milestones

FY2019
４

5

6

7

8

9

10

FY2020
11

Interim Report
▼

12

1

2

3

4

5

Interim Report
▼

6

7

8

9

10

11

12

Interim Report▼

1. Planning for elementtest
2. Manufacturing of test
equipment
3. Element test
(Parameter optimization
test)
4. Long-term water flow
test

1

2

3

Final
Report▼① Filter load optimization test
② Backwash condition optimization test

④ Long-term water flow test

5. Selection of applicabletechnology

6. Conceptual systemdesign7. Summary

Remarks③ Impact evaluation test of insoluble criticality inhibitors: Study on its implementation period based on necessity©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[2) (ii) ② Treatment technology for deposits collected from inside PCV]

No.191a. Characterization of collected liquids, wastewater➢ Preconditions✓ Applicable particulate collection equipment candidates

The selections were made as shown in the table below based on the test results up to the previous year, and literature research.
Table: Filter water flow test
No.
1

2

3

Particulate collection
equipment

SystemRoughremovalList of particulate collection equipment

Mesh size of filter

Remarks

Auto strainer

50 μm

Select the same mesh size as used in the FY02018 subsidy project

Auto strainer

10, 20 μm

As the differential pressure increased significantly when using auto
strainer with a mesh size of 5 μm, the test was conducted using filterswith a larger mesh size.

Sintered metal filters2, 5, 10, 20 μm

As the differential pressure increased significantly with the sinteredmetal filters with a mesh size of 2 μm, the test was conducted using
filters with a larger mesh size.

10, 50, 100 nm

As the increase in differential pressure was gradual for UF membranes
with mesh sizes of 50 and 100 nm, tests were also conducted for a 10
nm UF membrane.

Intermediate
removalFinal
UF membrane (Ceramic✓
Simulated particle
composition
treatment
filter)Current
report✓ Simulated particle composition
The components of fuel debris are roughly divided into fuel components, structural material components, and concrete components.
The composition of particles used for simulating each component were set as shown in the table below.
Table: Filter water flow test
Simulated particle composition
No.

Simulation target

Simulated particles

Particle composition

Relative density

Particle composition

Relative density

1

Uranium dioxide

Approx.11

Tungsten carbide

Approx.15

Tungsten dioxide

Approx.11

2Core internals

Approx.8

SUS316L

Approx.8

3Concrete

Approx.2 to 3

Silica sandApprox.3©International Research Institute for Nuclear DecommissioningNo.192

7. Implementation items of this project[ 2) (ii) ② Treatment technology for deposits collected from inside PCV]
a. Characterization of collected liquids, wastewater➢ Preconditions✓ Filter water flow test - Testing Conditions
The common conditions for the filter water flow test were set as shown in the table below based on theliquid conditions of the water to be treated.
Table: Filter water flow test - Testing Conditions
No.
1

Items
Temperature

Conditions

Remarks

Room temperature (approximately
25℃)

The temperature was set to room temperature as it was considered that the removal performance
had less dependency on temperature.

2

pH

Neutral (unadjusted)

Though the particles may aggregate, deposit, and dissolve due to the impact of pH, the factorsthat change in such cases are particle size and SS concentration. Since the particle size and SS
concentration are evaluated as test parameters in this development, it is considered that data
contributing to the performance evaluation can be obtained even if the actual pH and the test pH
are different.

3

Sodium pentaborate

0 ppm

Not added, as the effect of sodium pentaborate was considered to be small based on the resultsof the subsidy project in FY2018.

4

Salt content

0 ppm

The operation during fuel debris retrieval varies depending on the presence or absence of PCV
leakage. However, even under the condition of PCV leakage with high salt concentration where
the salt concentration is about several hundred ppm of chloride ions, its contribution to the densityand viscosity of the fluid is small, and hence not considered.

5

SS concentration

30 ppm ~ 300 ppm #The basic condition was set to 100 ppm, the accelerated test condition was 300 ppm, and thecondition considered for rough removal was 30 ppm.

6

Particle size

0.1 μm ~ 100 μm

Estimated particle size collected with non-soluble nuclide removal equipment

7

Flow rate3 ~ 10 m3/h #The basic condition was set to 10 m3/h considering the current rate of coolant injection as 3 m3/h,
and anticipating the amount of water expected to be used in the retrieval operation.

# The test conditions were adjusted so that the load on the filters is the same (Example: Flow rate 10 m3/h x SS
concentration 100 ppm ≒ Flow rate 3 m3/h x SS concentration 300 ppm )©International Research Institute for Nuclear DecommissioningNo.193

7. Implementation items of this project[ 2) (ii) ② Treatment technology for deposits collected from inside PCV]
a. Characterization of collected liquids, wastewater➢ Test results✓ The test was conducted under the condition of changing the filter pore size (5 → 20 μm) and eliminating themaintainability of SS concentration (100 ppm → 30 ppm), but there was no change in the differential pressure
increase time (liquid flow time).
✓ When the particle size was set to 50 to 0.1 μm, which is the assumed size in the actual water system, the increasein differential pressure was mitigated, but the particle collection rate decreased.
✓ The impact of the decrease in particle collection rate needs to be evaluated because it will be a load during the final
treatment in the subsequent stage.
Table: Summary of FY2018-2019 test results
FY2018FY2019Equipment

Auto strainer

Auto strainer

Filter pore size

5 μm

20 μm (close to the pore size of rough removal where the differential pressure was relatively small)Particle size

5 μm each

Flow rate

10 m3/hSS concentration

100 ppm(Not considering rough removal)

Mix of 50 ~ 0.1μm(Assumption in actual water system)

50 μm each
10 m3/h100 ppm, 30 ppm(Considering the presence or absence ofrough removal)

30 ppm(Considering rough removal)

Particle collection
rate

Approximately 90%

About 30%
(The remaining about 60% will become the load
for final treatment)

Increase indifferential pressure

Large(Liquid flow time approx. 10 sec)

Small
(Liquid flow time approx.10 min#1)#1: Value estimated from differential pressure trend (calculated by creating an approximation curve from the test data)©International Research Institute for Nuclear DecommissioningNo.194

7. Implementation items of this project[ 2) (ii) ② Treatment technology for deposits collected from inside PCV]
a. Characterization of collected liquids, wastewater➢

Summary✓ In the current test, a prospect of approximate liquid flow time of 10 min was obtained. However, as thefrequency of backwashing is high even under these conditions, it is presumed that measures such asadding multiple intermediate removal systems is necessary while implementing intermediate removal.

➢

Future plans✓ Load optimization test for final treatment.
✓ Study on adding multiple intermediate removal systems.
✓ Long-term water flow test by combining rough removal, intermediate removal (if necessary) and final
treatment
Table: Relationship between duration of liquid flow andamount of drain / backwash water (reference)

Durationof liquid
flow

Amount of drain /backwash water #1(m3/y)

Frequency of drain /backwash(times/y)Concentration rate #2(times)10 sec22075203153600

0.04

10 min

36792

52560

2.4

1h

6132

8760

14

24 h

256

365

343

#1: The assumed total amount of drained and backwash water per 1 time is
700 L.
#2: Amount of water flow per year (10 m3/h x 24 h/d x 365 d/y = 87600 m3/y)
÷ amount of drained or backwash water② If the intermediate removal load cannot be absorbed in the
final treatment, measures such as adding multiple intermediate
removal systems and increase in filtration area shall be required.
Roughremovalequipment

Drainbackwash water< Several tens of μm

Intermediate
removalequipment 1< Several μm

Drain
backwash waterIntermediate
removalequipment 2
Drain
backwash water

Final
Treatment< Several tens toseveral hundreds of nm

Backwash water① Evaluate whether theload of intermediate removal
can be handled in final
treatment and check theneed for intermediate
removal.

Figure. Future Study Policy©International Research Institute for Nuclear DecommissioningNo.195

7. Implementation items of this project[ 2) (ii) ② Treatment technology for deposits collected from inside PCV]
b. Development of technologies for separating solid material from wastewater and storing it⚫ Purpose of development➢ The technology for the separation of solid materials present in wastewater generated from particulate removal
equipment, was selected.
➢ The practical applicability of the selected technology for solid-liquid separation was evaluated through element tests.
⚫ Issues to be solved➢ It takes some hours to several tens of hours for the fine particles contained in the wastewater to settle naturally, and itis necessary to develop a technology that promotes solid-liquid separation.
➢ Insufficiency in the data required to identify the equipment considering the storage, remote operations, maintenance,
etc., of separated solid materials.
⚫ Approaches to development➢ Selection of technology for solid-liquid separation through literature research.
➢ Selection of aggregating agents through literature research.
➢ Preliminary tests for the selection of aggregating agents.
① Confirming the sedimentation separation effect of aggregating agents using simulated wastewater (beaker scale
test)② Evaluating the effects of aggregating agents on adsorbents (batch test)③ Designing and manufacturing of testing facility (sedimentation separation tank)➢ Element tests to confirm the actual applicability of sedimentation separation tank.
④ Aggregation sedimentation test using the sedimentation separation tank
⑤ Evaluating the effects of aggregating agents on adsorbents (batch test, column water flow test)⚫ Expected outcome➢ Evaluation of the results on actual applicability of aggregation sedimentation technology by theaddition of aggregating agents.
➢ Conceptual design of sedimentation separation tank and acquisition of operational data.

©International Research Institute for Nuclear DecommissioningNo.196

7. Implementation items of this project[ 2) (ii) ② Treatment technology for deposits collected from inside PCV]
b. Development of technologies for separating solid material from wastewater and storing it➢ Study Flow: In this project, the applicability of aggregation sedimentation was confirmed using a beaker test, andthe operation and operability was tested using device testing.
FY2019 (Start of this PJ)

Selection oftechnology forseparation(See table below)

FY2021 (End of this PJ)

FY2020Aggregation
sedimentation
beaker test

(Results)
・Aggregation sedimentation
・ Applicability of inorganic aggregating
agentsAfter FY2021

Device testing
and conceptual design

Solid-liquid
separation
methodBasic / detailed
design(Results)
・ Equipment / process concept
・Operation and operability(Results)
・Applicability of aggregation
sedimentation
・ Candidates for aggregating
agentsTable Features of solid-liquid separation method and applicability to sludge
collection (Summary of literature research)

Confirmation of
actual liquid
properties andconceptual design

StirrerP

Aggregating
agentsM

P

pH regulator

pH regulating
reagent

pH調整試薬Features

Applicability to sludge
collection

Aggregating
agent
凝集剤

M

・ Separability mainly depends on surface properties of theparticles (charge, etc.).
⇒ There is redundancy even when the particle size fluctuates
Aggregation slightly.
sedimentation ・Aggregating agents are basically inorganic in consideration of
their contact with fuel debris.
・The effect of aggregating agents on the adsorption systems inlater stages is unknown.

pH pH計meter

Outlet for supernatant water

○
Valve

P
Container for collecting
supernatant water
Extraction pump

Centrifugal
separation

・The structure is simple, and maintainability is excellent.
・Separability strongly depends on particle size and relative density.

△

Filtration

・Separability strongly depends on particle size.
・Separation of fine particles might be accompanied by a significant
increase in differential pressure.
・Excellent for dehydration (filter press, etc.).

△

Floatation

・Strongly dependent on the relative density of the particles (it does
not float if the relative density is high).
・It is necessary to study the recovery method of floating particles.

x

Illustration of beaker test Illustration of device test
Items to be confirmed in the tests
・Time for sedimentation, sludge volume and impact on theadsorbent
・ Impact of parameters such as pH, water temperature,presence or absence of boric acid, etc.

Specific approaches
after FY2021 will
require coordination
with operator’s
engineering department.

©International Research Institute for Nuclear DecommissioningNo.197

7. Implementation items of this project[ 2) (ii) ② Treatment technology for deposits collected from inside PCV]
b. Development of technologies for separating solid material from wastewater and storing it➢ Development Schedule
FY2019Study Items
４

Major milestones

5

6

7

8

9

10

11

Interim Report▼

FY2020
12

1

2

3

4

5

Interim Report
▼

6

7

8

9

10

11

12

1

Interim Report
▼

2

3

Final
Report▼

1. Literature research
2. Drafting of test plan
3.Element test
(Includes preliminary test /equipment manufacturing)① Beaker scale test
② Evaluating the effects on adsorbents (batch test)③ Design and manufacture of test facility (sedimentation
separation tank)④ Testing of sedimentation separation facility⑤ Evaluating the effects on adsorbents (column water flow test)

4.Conceptual system design
・ Particulate collection system
・ Solid materials separation
system・ Storage and handlingtechnology

5.SummaryRemarks©International Research Institute for Nuclear DecommissioningNo.198

7. Implementation items of this project[ 2) (ii) ② Treatment technology for deposits collected from inside PCV]
b. Development of technologies for separating solid material from wastewater and storing it
Test details and preconditions➢

Evaluating the effects of aggregating agentson adsorbent (batch test)
Tracer: Any one of Cs, Sr, Am
Added concentration: Am approximately 10 Bq/ml, Cs, Sr
100 ppm
Supernatant
Sampling & analysis
water

Inorganic aggregating agents: PAC, high-base PAC, aluminium sulphate, ferric sulphate, ferric chloride
Simulated wastewater(SS particles + purewater)

pH regulator (NaOH or HCl)

Stirrer

Adsorbent
Sludge component

BeakerCreation ofsimulated
wastewater

Addition of
aggregating
agentsStirring(150 rpm)

Stirring(50 rpm)

Shaking (0th, 7th,
14th day) )Leave tostand

Concentration analysis

Figure. Illustration of beaker scale sedimentation separation test and batch test for evaluating impact on adsorbent
Table: Preconditions for beaker scale sedimentation separation test
No.

Items

Test conditions

Basis for setting the conditions

1

Concentration of aggregating
agents100 ~ 1500 ppm

Set based on the value recommended by the manufacturer.

2

SS type① SiO2 only (standard condition)② Three component mixture (WC, SUS, SiO2)① Only SiO2 was set as a conservative condition (which is difficult to settle).
② For comparison, the test was also conducted with the condition of three-component mixture
simulating the actual wastewater.

3

SS particle size

0.1 ~ 100 μm (Standard condition: 0.1 μm)

Set based on the particle size range of the particulate collection facility used in the previous stage.

4

SS concentration

2000 ~ 20000 ppm(Standard condition: 10000 ppm)

With 10000 ppm as the base, which is 100 times more concentrated than the inlet condition (100 ppm)of the particle recovery equipment in the previous stage, the fluctuating conditions are evaluated asparameters.

5pH

7 ± approx.2 (Standard condition: 6 ~ 7)

The fluctuating conditions are evaluated based on the value 7 recommended by the manufacturer.

6

Temperature of water

10 ~ 40 ℃ (Standard condition: 25 ℃)

Standard condition is 25℃, taking seasonal fluctuations into consideration.

7

Impact on adsorbent

Adsorbent for Cs and Sr: Titanium silicate
Adsorbent for alpha nuclide (Am) : ActivatedcarbonA typical adsorbent is selected from the results of previous studies.
Tested with supernatant water with an aggregating agent added concentration of 1000 ppm (set fromthe previous beaker tests).

8

Sodium pentaborate

0 ~ 7000 ppm as B

Evaluate the effects of mixing boric acid in treated water.
©International Research Institute for Nuclear DecommissioningNo.199

7. Implementation items of this project[ 2) (ii) ② Treatment technology for deposits collected from inside PCV]

b. Development of technologies for separating solid material from wastewater and storing it➢ Test results (example)✓ Evaluation of the concentration of added aggregating agents and sedimentation time in boric acid-free system✓ All aggregating agents underwent aggregation sedimentation with an added concentration of 1000 ppm or more(sedimentation time was within 40 minutes).
✓ PAC and high-base PAC sedimented well even at an added concentration of 500 ppm.
✓ Sedimentation time increased with PAC of 1500 ppm (If the amount added is too large, the sedimentation time becomes longer).
<Test conditions>Concentration of aggregating agents: 100 ~ 1500 ppm, concentration of wastewater, silica sand (SiO2) 0.1 μm 10000 ppm, Boric acid (B) 0 ppm, pH ofwastewater (when adjusting pH after adding aggregating agents): 7±0.5
Temperature of water: 25℃, stirring speed / time: strong stirring 150 rpm x 5 min → weak stirring 50 rpm x 15 min, liquid volume 440 mL
Relation between concentration of
Sedimentation time indicates the time
aggregating agents and sedimentation timewith turbidity 20o or lesser

Ferric Chloride

Sedimentation time (min)High-base PAC

Volume after
aggregation
About 15 ~20%

High-base
PAC

Aluminumsulphate

FerricSulphate

FerricChloride
0 minutes after left to stand (1000 ppm)

10 minutes after left to stand (1000ppm)©International Research Institute for Nuclear DecommissioningNo.200

7. Implementation items of this project[ 2) (ii) ② Treatment technology for deposits collected from inside PCV]
b. Development of technologies for separating solid material from wastewater and storing it➢ Test Results (Summary)✓ The effectiveness of aggregation sedimentation in a boric acid-free system was confirmed.
✓ In the future, appropriate aggregating agents shall be selected by evaluating the effect of adding
boric acid.
Table Test result summary of beaker scale sedimentation separation test
No.

Items

Test conditions

Test results

Concentration of
aggregating agents100 ~ 1500 ppm

2

SS type① SiO2 only (standard condition)② Three-component mixture (WC, SUS, SiO2)・Aggregation sedimentation was seen under both conditions ① and ②. Sedimentation time was about the same.
・The volume of the aggregated material is less in the case of the three-component mixture (presumed to beconsolidated as it contains particles with a large relative density).

3

SS particle size

0.1 ~ 100 μm (Standard condition: 0.1μm)

・Sedimentation time is about 10 to 15 minutes at 0.1 μm and about 5 minutes at 1 μm or more.
・Larger particle size is more favorable for aggregation sedimentation.

4

SS concentration

2000 ~ 50000 ppm(Standard condition: 10000 ppm)

・ Good sedimentation can be seen up to 20000 ppm (sedimentation cannot be seen for 50000 ppm)
・ High-base PAC has a shorter sedimentation time with increasing SS concentration.

5pH

7 ± approx.2 (Standard condition: 7±0.5)

・The suitable pH range for high-base PAC is pH 6 ~ 8.5, and ferric chloride is pH 4.5 ~ 7.1.
・Assuming that the actual pH is neutral to alkaline, it may be possible to reduce the amount of pH regulator used
for high-base PAC.

6

Temperature of water

10 ~ 40 ℃ (Standard condition: 25 ℃)

・No impact was seen in this temperature range (the result was the same as the standard 25 ℃).

7Impact on adsorbents

Adsorbent for Cs and Sr : Titanium silicate
Adsorbent for alpha nuclide (Am) : Activated carbon

・No major change in adsorption performance even when supernatant water was added (removal rate wasalmost 100%)

8

Sodium pentaborate

0 ~ 7000 ppm as B

・Testing in progress.

©International Research Institute for Nuclear Decommissioning

Boric acid-free system

1・All of the aggregating agents showed aggregation sedimentation with added concentration of 1000 ppm or
more.
・ The relative merits and demerits of the aggregating agents is as shown below when determined based on
sedimentation time and added concentration of the aggregating agents
From the top: high base PAC, PAC ≒ ferrous chloride, aluminum sulphate, ferrous sulphate
・The standard added concentration is set to 1000 ppm for the parameter tests other than No.2

No.201

2.Progress of technology development (Element test results)2) (ii) ②b. Development of technologies for separating solid material from wastewater and storing it➢ Summary of FY2019✓ The applicability of separation using aggregation sedimentation was confirmed in a boric
acid-free system.
➢ Future plans✓ To confirm the applicability of aggregating agents in boric acid-added system andconceptual process design based on those results.
✓ To confirm the operability using aggregation sedimentation device testing, identify theissues, and study the countermeasures.
Impact evaluation test of boric acid

Case 1: There is no impact ofboric acid on aggregation
sedimentation (impact whenusing up to 7000 ppm as B)
・As it has the same trend as
the boric acid-free system, it
does not require any
additional study.

Case 2: A permissible boric acidconcentration (for example, up to 1000ppm) is required for aggregation
sedimentation to come into effect.
・ Study the aggregating agents /aggregation conditions that can beapplied even in the presence of boric
acid.
・Study processes where boric acidconcentration is lower than thepermissible limit (image shown in next
slide)
・Perform aggregation sedimentation
after removing boric acid up to thepermissible concentration.

Case 3: Aggregation
sedimentation does not comeinto effect in the presence ofeven a small amount of boric
acid・ Study the aggregating agents /aggregation conditions that canbe applied even in the presenceof boric acid.
・ Perform aggregation
sedimentation after removing
boric acid completely.

Figure. Trifurcation of study cases based on the impact of boric acid
©International Research Institute for Nuclear DecommissioningNo.2022.Progress of technology development (Element test results)2) (ii) ②b. Development of technologies for separating solid material from wastewater andstoring it➢

Proposal for a conceptual system (reference) for the case with permissible boric acid concentration✓ Aggregation sedimentation may be applicable by using the external water that does not contain boric acid,for backwash water.

Circulating cooling water (with boric acid) is also acceptable

Stagnant water inside PCV(with boric acid)

Rough removal
equipment(Cyclone,
Auto strainer)< Severaltens ofμm

External water(without boric acid)

Intermediate removal
equipment(Sintered metal filters,
bug filter)

Drain or backwash water< Severalμm

Backwash water

Extraction

SupernatantNatural sedimentation separation tank

Regulate the amountand frequency of
backwash so thatboric acidconcentration does
not hinder
aggregation
sedimentation

To circulating
cooling water
systemTo waste treatment facility or fueldebris treatment facility

RetrievalExternal water(without boric acid)

Final treatment
equipment(UF membrane, MF
membrane, etc.)< Several tens toseveral hundreds of nm
To soluble alpha nuclide removal facility

Backwash water

Extraction

To circulating coolingwater system

Aggregating agentsSupernatant

To waste treatment facility or
fuel debris treatment facility

Retrieval
Aggregation sedimentation separation tank

Figure. Conceptual plan for the aggregation sedimentation system (for the case with permissible boric acid concentration)©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[ 2) (ii) ② Treatment technology for deposits collected from inside PCV]

No.203b. Development of technologies for separating solid material from wastewater and storing it➢ Design and manufacture of testing device (sedimentation separation tank)✓ The conceptual diagram of a testing device was created and the required equipment
was identified.
Table: Proposal for testing device specifications
No.

P

Sludge collection pump(For emergency use
during valve blockage)

Scraper
Stirrer
M

P

Aggregating
agentsP

NaOH tankP

H2SO4 or HCltank

Sedimentation
separation tank

Outlet for supernatant water
ShakerValve
P

Container forsludge collection

Pump for extracting
supernatant water incollection container

Equipment
nameSedimentation
separation
tank

No of
units
1

1

Made of acrylic. The shape, dimensions, and capacity are
set with reference to the beaker test results and canisters.

3

Container forsludgecollection
Gate valve

1

Electrically-operated gate valve

4
5

Tank
Pump

3
5

Aggregating agents reagent x 1, pH reagent x 2
For supplying aggregating agents x 1, for supplying pH
reagent x 2, for extracting supernatant x 1, for extracting
sludge x 1

6Stirrer

1

Set based on the stirring speed in the beaker test

7

Scraper1

Manual type. To remove particles adhering to the tank.

8

Shaker

1

10

Pipe

1

For removal of particles adhering to the tank and to prevent
valve blockage
For transferring water before and after treatment

11

pH meter

1

12Turbidity
Meter

1

13

Rack

1

1

2

Remarks
The tank height is set based on the sedimentation time inbeaker test.

Assumed to have the same specifications as in the beaker
test.
Assumed to have the same specifications as in the beakertest.
For installing the facility

Figure. Conceptual diagram of testing device©International Research Institute for Nuclear DecommissioningNo.204

7. Implementation items in this project2) Development of technology for processing fuel debris(ⅲ) Survey on sorting technology for fuel debris and radioactive wastes
The technologies required for sorting the fuel debris and radioactive wastes retrieved from inside
the PCV were surveyed.
Study on the sorting methods and their feasibility were evaluated in coordination with related
projects such as “Development of Technology for Containing, Transfer and Storage of Fuel Debris”
and “R&D for Treatment and Disposal of Solid Radioactive Wastes”.

① Study on the formulation of scenarios related to sorting
・The preconditions (sorting requirement) were organized, and the sorting requirement flow
was studied.
② Study on the required sorting technology
・ The locations for applying each decided sorting method were organized and the necessity
for survey on technologies was studied.
・ Surveys on technologies were conducted and the applicability to sorting was studied.
③ Evaluating the technical feasibility of the fuel debris sorting scenarios
・ Based on the technology survey results, the technical feasibility of each sorting point wasevaluated and sorting scenarios considering the feasibility were formulated.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[ 2) (iii) Survey on sorting technology for fuel debris and radioactive wastes]

No.205⚫ Purpose of development➢ Sorting (separating) fuel debris and radioactive wastes to reduce the amount stored asfuel debris➢ Evaluating the feasibility of technology that can sort fuel debris and radioactive wastes⚫ Issues to be solved➢ Identifying methods for sorting fuel debris and radioactive wastes➢ The procedures (scenario) for sorting fuel debris and radioactive wastes, and locations
where sorting must be applied have not been decided yet.
⚫ Approaches to development➢ Organization of the sorting requirements from the perspective of retrieval, transportation,and storage➢ Creation of a plan for sorting method of fuel debris based on the requirements
organization results➢ Identification of the treatment and measurement techniques required for sorting fueldebris➢ Survey on treatment and measurement techniques required for sorting fuel debris➢ Creation of a plan for the scenarios and locations for application of fuel debris sorting➢ Organization of the development issues and difficulty levels required for sorting fueldebris and radioactive wastes⚫ Expected outcome➢ Identification of the sorting method for fuel debris and radioactive wastes➢ Identification of the development issues required for sorting fuel debris and radioactive
wastes©International Research Institute for Nuclear DecommissioningNo.206

7. Implementation items of this project[ 2) (iii) Survey on sorting technology for fuel debris and radioactive wastes]◼ Study flow
Survey on sorting requirements from each projectClassification and organization of sorting
requirements

Questionnaire survey on the purpose, target, sorting details, implementation method,
location of application, required technology, issues assumed in the requirements,
related projects, etc.
The study targets were set after the following classification and organization:① Requirements to be studied in each project② Requirements to be studied in design as interface conditions③Sorting requirements subject to study

Excluded fromstudy targets

Study of sorting requirements flow

A flow chart that logically and mechanically arranged all sorting requirement points
considered from the viewpoint of sorting requirements and the location of application,
was created

Organizing the plans for sorting scenarios

Multiple sorting scenario plans were created from the sorting requirements flow chart
based on the evaluation of advantages and disadvantages of each sorting point andthe assumption on technical feasibility.

Identification of measurement technology required for sorting

Regarding the measurement technology, which is required as the deciding method, thelocation of application was classified based on the size of the background and thetargets for technology survey were identified.

Survey on technologies

Evaluating the technical feasibility of the fuel debris
sorting scenarios

Surveys were conducted on visual (image) determination technology, PCV nuclear
material distribution estimation technology, and nuclear material measurement
technology, and their applicability was evaluated.

Based on the results of technology surveys, the technical feasibility of each sorting point
was evaluated and sorting scenarios considering the feasibility were formulated.

©International Research Institute for Nuclear DecommissioningNo.207

7. Implementation items of this project[ 2) (iii) Survey on sorting technology for fuel debris and radioactive wastes]◼ Development Schedule
FY2019Study Items
4

Major milestones

5

6

7

8

9

10

Interim Report
▼

11

12

1

2

3

Year-endreport
▼1. Organizing preconditions andsorting scenario plans
2. Survey on technologies required
for sorting fuel debris
3. Feasibility evaluation of fueldebris sorting scenarios
4. Result Summary
Remarks©International Research Institute for Nuclear DecommissioningNo.208

7. Implementation items of this project[ 2) (iii) Survey on sorting technology for fuel debris and radioactive wastes]◼ Organising the preconditions (sorting requirements)

From the identified sorting requirements, the sorting requirements to be studied (purpose, targets, sorting location, determination
method, criteria, expected issues, advantages and disadvantages of sorting, etc.) were organized.

● Sorting requirements to be studied
No.

1

2

PurposeEnsuring criticality
safetyIncrease in retrieval
throughput

3

Sorting location

Required technology

Items containing significant nuclear fuel material are stored in shapecontrolled containers (canisters) to ensure criticality safety.

A: Inside PCV

Nuclear material measurement for visual (image)determination, nuclear material distribution
estimation and criticality safety control

Collected materials with a small amount of nuclear material are stored
in mitigation canisters that are larger than the storage canisters.

A: Inside PCV

Measurement of nuclear material for measurementof fissile material concentration

Collected materials with large amount of target materials are stored inlarge-size subcritical containers.

A: Inside PCV

Visual understanding of shape and dimensions

Collected materials with a small amount of nuclear material are stored
in mitigation canisters that are larger than the storage canisters.

4
5

Goal and sorting details

Reduction in storage
scale of fuel debris

Collected materials with a small amount of nuclear material are storedin waste containers.

Measurement of nuclear material for measurementof fissile material concentration
B: Outside the PCV of eachunit
C: Storage facilities

Measurement of nuclear material for criticality
safety control

6Collected materials that are not covered by safeguards are stored inwaste containers.

7

Sorting of waste into waste covered by safeguards and waste notcovered by safeguards

B: Outside the PCV of eachunit
C: Storage facilities

Measurement of nuclear material for deciding thetermination of safeguards, PP

Confirmation of the fact that the amount of nuclear material containedin the waste does not exceed the subcritical limit.

A: Inside PCV
B: Outside the PCV of eachunit
C: Storage facilities

Measurement of nuclear material for criticality
safety control

8Streamlining of waste
storage andmanagement

Measurement of nuclear material for deciding thetermination of safeguards, PP©International Research Institute for Nuclear DecommissioningNo.209
7. Implementation items of this project[ 2) (iii) Survey on sorting technology for fuel debris and radioactive wastes]◼ Summary of study results on the formulation of scenarios for sorting (Sorting Requirements Flow Chart)
A logical and mechanical arrangement according to the logistic flow of all possible sorting requirement points, considered fromthe viewpoint of sorting requirements and location of application
R/B, Expanded building (Each Unit)

Storagefacility (shared)
保管施設（共用）

R/B、増設建屋（各号機）

B：PCV外B:
Outside PCVA:
Inside
A：PCV内
PCV

形状・
Shape
濃度
/ 寸法
Size Concentration
Aa3

Aa2

Concentration
Ba

PP

Criticality
臨界

濃度

PP

Criticality

Ba6

Ba5

Ba4

Ca6

Ca5

Unit
cans
ユニット缶

デブリ
DebrisC:
Storage facility (During
C：保管施設（保管前準備時）
preparation before storage)

保管施設（デブリ保管）
Storagefacility(Debris storage)

Concentra-

Ca

臨界 濃度
tion

Inner diameter
220mmx
内径 220mm×Height 高さ
1m 1m～1.5ｍ
to 1.5m

Ca4

Canister
収納缶

Canister
収納缶

Bf

Bd

Cf

Cd

Bb6

Bb5

Cb6

Cb5

a: a：収納缶
Canister

Inner diameter
400mmx
内径 400mm×Height 高さ
1m1m～1.5ｍ
to 1.5m

Ad

Mitigating
緩和型
unit can
ユニット缶

PP
Bf

Criticality
臨界

Mitigating
緩和型収納缶canister

Bd

Ba

PP
Cf

Ba

Mitigating
緩和型収納缶canister

臨界

Criticality

b: Mitigatingb：緩和型収納缶
canisterCd

Ca・A further elongated storage
canister・Waste storage containers
・収納缶を更に長尺化したもの
made with flat plates with a
・廃棄物保管容器を現状の臨界制限
thickness lesser than the
値以下の厚さの平板状にしたもの
current criticality limit value

Ca等

Targets
対象物(Before(解体前)
dismantling)

Aa1

目視/
Visual /Criticality
臨界

Container inside a
大型未臨界容器
large subcritical
内容器
container

PP

Bc5

Cc6

Criticality
臨界

PP

Bf

Ad

Waste
廃棄物

Bc6

Large
subcritical
大型未臨界容器
canister

Cf

Ba

Bd

Bd8

Criticality
臨界

Bd7

臨界
Criticality

Ca

PP

Cd8

Bf

Waste storage
廃棄物保管容器
container（汚染レベル）(Contaminatedlevel)

臨界
Criticality

c: Large subcriticalc：大型未臨界容器
container

Cc5

Storagefacility
保管施設（廃棄物保管）(Waste storage)
PP、保障措置対象
Subject
to PP, safeguards

Cd

Waste storage
廃棄物保管容器
container（管理レベル）(Controlledlevel)

Waste collection
container(Amount of
廃棄物回収容器
nuclear material less
（核物質量 臨界未満）
than criticality)

Preconditions① Inside PCV, the fuel debris andwastes are put in the inner container② Outside PCV, the inner container isplaced in corresponding outer
container (canisters, waste storage
canister, etc.)③ As large-size subcritical container is a
new concept that has not beenstudied so far, it is indicated by adotted line.

Cd7

PP

Waste storage
廃棄物保管容器
container（管理レベル）(Controlledlevel)
Cf

Waste storage
廃棄物保管容器
container（汚染レベル）level)(ContaminatedSorting points are indicated by●.The sorting points indicate thesorting location and order and do
not necessarily mean that sortingis performed at those points.
If sorting is completed on the
upstream side, it need not be
implemented on the downstreamside.

Width 1.42m x Depth 1.42m
幅 1.42ｍ × 奥行 1.42ｍ
x Height 1m
× 高さ 1m
d: Waste storaged：廃棄物保管容器
container（管理レベル）(Controlledlevel)

Not
subject to PP,
PP、保障措置対象外
safeguards
f: Wastestorage container
f：廃棄物保管容器(Contaminated（汚染レベル）level)

Notations used for sorting points
1st digit: Area symbol
2nd digit: Line symbol
3rd digit: Sorting requirement
number©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[ 2) (iii) Survey on sorting technology for fuel debris and radioactive wastes]

No.210◼ Summary of study results on the formulation of scenarios for sorting● Sorting classification results and criteria (Tentative plan)
Criteria (tentative)*1

Amount of nuclear
Amount of *2
material is less than
nuclearor equal to thematerial isamount to terminate
3.7 kg /safeguards andcontainer or
nuclear material
lessprotection target(Neither ofthe items on
the right issatisfied)

U-235 concentration infuel debris: less thanor equal to thereactivity equivalent to
1.5 wt%Debris that isdifficult tostore in acanister

1

O

-

-

-

-

1, 2, 4

-

O

-

-

-

c: Large-size subcriticalcontainers

1, 3

-

-

O

-

-

d: Waste storage container(controlled level)

1, 5, 7, 8

-

-

-

O

-

f: Waste storage container(contaminated level)

1, 5, 6, 7, 8

-

-

-

O

O

Sortingrequirement
numberSorting classification

a: Canister(Inner diameter 220 mm,
height approximately 1 m)
B: Mitigating canister(Inner diameter 400 mm)*3*1: As the results of past evaluations in the Canister Project show that canisters with an inner diameter of 400 mm can be used when the weightratio of U-235 to the fuel debris is approximately 1.7 wt% or less on the assumption that the fuel debris is composed entirely of U-235 and U-238,
this value is set to 1.5 wt% providing a margin to the 1.7 wt%.
*2: Set by equally dividing the minimum critical mass (approximately 30 kg) into 8 waste storage containers, considering the arrangement andstacking during storage*3: The large-size subcritical containers are at the idea stage, and no concrete specifications have been defined.

©International Research Institute for Nuclear DecommissioningNo.211

7. Implementation details of this project[2) (iii) Survey on sorting technology for fuel debris and radioactive wastes]◼ Study results on technologies required for sorting (Identification of targets for technology survey)
Regarding the determining methods identified while organizing sorting requirements, the location of application was classified basedon the size of the background affecting the difficulty level of the decision and the target of each technology survey was set. (O shownin the table below refers to survey targets)Standard for applicable location and background
A: Inside PCV(Before storing thecontainer)
Several 10 ~Several 100 Sv/h

B: Outside PCV ofeach unit(RB, expandedbuilding)
Several mSv/h level

C: Storage facility(During
preparation before
storage)500 μSv/h or less(Amber zone)

1

Visual (image) determination
・Comparison by means ofidentification sample*1
・Determination through image
processing (Includes AI)

O
(1-A)

O
(1-B)

O
(1-C)

2

Estimation of nuclear material
distribution inside PCV

O
(2-A)

3

Measurement of nuclear material(Nuclide, concentration)

No.

Determining (measurement)
technology

Remarks*1) Shape data for structures etc., which
provide information for visual determination

Create distribution diagram based on the
results of disaster behavior analysis andsampling analysis⇒ Directly investigate the technology formeasuring distribution① For measuring theconcentration of nuclear fissile
materialO(3①-A)

O(3①-B)

O(3①-C)

・For determining the feasibility of storage
in a mitigating canister② For criticality safety control

O(3②-A)

O(3②-B)

O(3②-C)

・For determining the feasibility of storage
in a waste storage containerO(3③-B)

O(3③-C)

・For the determining the targets for PP
and safeguards③ For decision on termination of
PP and safeguards

Note) Since the symbols in the lower row represent the applicable locations, the symbols are a combination of determination(measurement) technology number and symbol for the applicable locations (A, B, C)©International Research Institute for Nuclear DecommissioningNo.212

7. Implementation details of this project[2) (iii) Survey on sorting technology for fuel debris and radioactive wastes]◼ Study results on technologies required for sorting (Technologies subject to technology survey)
The existing determination (measurement) technologies were investigated and the principle, performance, usage
environment, device configuration/size, measurement time (measuring speed), usage results, applicable locations andissues were organized in a survey form.
●: Estimations
Purpose of determination (measurement) assuming application
Visual (Image)determination

Technologies surveyed

Estimation of
distribution

Concentration
measurement

Criticality safetycontrolPP,
safeguards

1-A

1-B

1-C

2-A

3①-A

3①-B

3①-C

3②-A

3②-B

3②-C

3③-B

3③-C

Reflectance measurement

●

●

●

-

-

-

-

-

-

-

-

-

Image determination

●

●

●

-

-

-

-

-

-

-

-

-

Passive alpha ray measurement

-

-

-

●

●

●

●

●

●

●

-

-

Passive gamma ray measurement

-

-

-

●

●

●

●

●

●

●

-

-

Passive neutron measurement

-

-

-

●

●

●

●

●

●

●

-

-

Active X ray measurement

-

-

-

-

●

●

●

●

●

●

-

-

X ray transmission

-

-

-

-

-

●

●

-

●

●

-

-

Cosmic rays scattering measurement

-

-

-

-

-

●

●

-

●

●

-

-

Active neutron measurement

-

-

-

-

-

-

-

-

●

●

●

●

Passive neutron measurement +gamma ray measurement

-

-

-

-

-

-

-

-

●

●

●

●

Passive/Active neutron measurement +gamma ray measurement

-

-

-

-

-

-

-

-

●

●

●

●

©International Research Institute for Nuclear DecommissioningNo.213
7. Implementation details of this project[2) (iii) Survey on sorting technology for fuel debris and radioactive wastes]◼ Study results on technologies required for sorting (Main measurement technology survey form (1/4))
Table-1 Measurement
technology survey form
表-1 測定技術調査票

Table-2 Measurement
technology survey form
表-2 測定技術調査票

Technology
name Reflectancemeasurement technology (Hyperspectral camera)
反射率計測技術（ハイパースペクトルカメラ）
技術名称

determination technology (Determination of structure using camera image, 3D shape)
Technology
画像判定技術（カメラ画像、3D形状を活用した構造物判定）
技術名称name Image・Identify the type of the structure by comparing the image capturing the appearance of the object with a・可視カメラで対象の外観を撮影した画像や、レーザー光線によるTOF（タイム・オブ・フライ
visibility camera,
and the external shape of the object measured by the TOF (Time Of Flight) laser
ト、TimeOf Flight）方式のレーザスキャナーなどで測定した対象の外観形状を構造物の外scanner method using laser light etc., with the external photographs of the structure and drawing data.
観写真や図面データなどと比較し構造物の種類を識別する。
・Application is limited to the less damaged structure in order to compare with drawing data and・構造物の外観写真や図面データなどと比較するため損傷の少ない構造物に適用が限定
external photographs of the structure, but the type of the structure might be identifiable using
されるが、市販センサ（認識処理は別）で構造物の種類が識別できる可能性がある。
commercial sensors (recognition process is separate).

・ハイパーカメラで対象表面の反射率分布を取得し、事前測定した各種材料とのスペクトル・Acquire
reflectance distribution of the object surface with hyper camera and compare the spectrum
withvarious materials measured earlier in order to identify the material.
と比較することで、材質を識別する。
・Sinceinformation obtained is limited to the surface, evaluate the material of the structural surface
・得られる情報は表面に限定されるため、比較的損傷の少ない構造物等に適用すること(such as adhesion/non-adhesion of fuel component) broadly by application to relatively less damaged
で、構造物表面の材質（燃料成分の付着の有無等）を大まかに評価する。
structures.
原理
Principle

Principle
原理

RGB image

Pseudo-color image

Figure 1:図1：鉱石の材料評価の例
Example of material evaluation of ore

Source:https://www.klv.co.jp/hyperspectral/case_study_case_study06.htmlり引用
https://www.klv.co.jp/hyperspectral/case_study/case_study06.htmlよPerformance
性能

分解能
Resolution

Figure 2: Spectrum of Uranium ore
図2：ウラン鉱石のスペクトル

Figure図1：車載カメラによる画像認識の例
1: Example showing image recognition using in-vehicle camera
https://jp.mathworks.com/discovery/image-recognition.html
よ り引用
Source:https:/jp.mathworks.com/discovery/image-recognition.html

Source:PRELIMINARY STUDY ON THE DETECTION OF RADIOACTIVITY WITH AIRBORNE REMOTESENSINGstudy
SYSTEMS
よ り引用
Preliminary
on thedetection of radioactivity with airborne remote sensing system

・To the extent to which the rough composition distribution of concrete, metal etc. on
・測定対象表面のコンクリート、金属などの大まか組成分布が分かる程度
thesurface to be measured can be understood・Sincethe data is only for the surface of the target to be measured, it is not possible
・測定対象表面のみのデータになるため、測定対象物の燃料成分の重量やUO2等の割
to measure the weight of the fuel composition and percentage of UO2 in the material
合の測定は不可tobe measured.
・測定対象物の損傷が比較的少ない場合、その画像と構造物の設計図面を照合するこ・Incase the damage of the target to be measured is relatively low, it may be possible
toestimate the target to be measured by comparing this image with the designとで、測定対象物が何かを推定できる可能性あり
diagram of the structure.

Performance
性能

Lower limit of

Depends
on the surrounding conditions and target to be measured.
測定下限値
周囲の状況や測定対象に依存。
measurement

Lower limit of

測定下限値
絶対値での評価は困難
Evaluation
with absolute value is difficult
measurement

Usage
使用環境
詳細不明（遮蔽体が必要）
environment
Details
not known (Shield is required)（バックグラウンド）(background)[Device configuration]【装置構成】
・Camera and mechanism for its operation are mandatory
・カメラ及びそれを操作する機構が必要
・Since camera is not radiation resistant, shield is required
・カメラは対放射線性がないため、遮蔽体が必要
・Data is transferred with a cable
・データはケーブルで伝送する[Size]【サイズ】
・Since it depends on the size of the shield, evaluation by radiation resistance test is required
・遮蔽体のサイズ依存するため、対放射線性試験などで評価必要
・カメラ単体(Headwall社Nano-Hyperspec)：約76
× 76Approx.
× 12076mm
mm（移動用機構、遮蔽体
・Camera unit (Nano-Hyperspec manufactured by Headwall):x 76mm x 120mmを除く）(Mobile mechanism and shield are excluded)
Device
装置構成・サイズ
configuration / size

Shielding
遮蔽体

図2：形状フィッテイングの例
Figure2: Example of shape fitting

https://jp.mathworks.com/discovery/image-recognition.html
よ り引用
Source:https:/jp.mathworks.com/discovery/image-recognition.html

・Identification of structure type with characteristic differences in
・外観（形状・寸法）に特徴的な違いがある構造物の種別を識別可能
appearance(shape/size) is possible.
Resolution ・外観（形状・寸法）と図面データなどを比較するため、損傷のある対象には
分解能
・Objects with damages are not applicable since appearance
適用不可。
(shape/size) and drawing data are compared.

Usage
environment Shieldis required in high-dose environment since semiconductor sensor is used.
使用環境
半導体センサを使用するため、高線量環境では遮蔽体が必要。
(background)
Shieldis not required in case of radiation-resistant camera.
（バックグラウンド）
耐放射線カメラの場合は、耐放性があるため遮蔽体は不要。
[Device configuration]【装置構成】
・Camera, laser scanner, and mechanism for operating them are required.
・カメラ／レーザスキャナー及びそれを操作する機構が必要。
・Requires the process of identifying the structure from the appearance image and shape.
・外観画像や外観形状から構造物を識別する処理が必要。
・Since semiconductor sensor (camera, laser scanner) is not radiation-resistant, shield is required
・半導体センサ（カメラ／レーザスキャナー）は耐放射線性がないため、遮蔽体が必要（耐(excluding when radiation-resistant camera is used).
放射線カメラ使用時は除く）。
・Data is transferred with a cable.
・データはケーブルで伝送。
[Size]【サイズ】
・Since it depends on the size of the shield, evaluation by radiation resistance test is required
・遮蔽体のサイズ依存するため、耐放射線性試験などで評価が必要
・耐放射線性カメラ単体：約Φ41×250[mm]・Radiation-resistant camera unit: Approx. Φ41mm x 250mm
・レーザスキャナー単体(FARO社
FOCUS150)：約230x183x103mm
Device・Laser scanner unit (FARO FOCUS 150):
Approx. 230mm x 183mm x 103mm
装置構成・サイズ
configuration / size ※移動用機構、遮蔽体を除く※Mobile mechanism and shield are excluded

Cameraカメラ

ミラー
Mirror

図3：カメラの外観例（Headwall社Nano-Hyperspec）

Figure 3: Example of the appearance of the camera(Nano-Hyperspec manufactured by Headwall)

Figure図4：構造のイメージ
4: Image of the structure

Measuring
測定時間size 1画像あたり、数秒
Several seconds per image(Measuring
speed) ※画角はレンズによって異なる。
（測定速度）※The angle of view depends on the lensUsage
results
使用実績

We have a record of general ore sorting, but no record of UO2 and debris
一般的な鉱石の仕訳実績あり。UO2やデブリの測定実績なし
management.

1A, 1B, and 1C. However, tasks such as opening the lid after container storage, etc. are required.
Applicable
locations １A,１B.1C。ただし、容器収納後は蓋を開けるなどの作業が必要
適用場所

Figure 図3：耐放射線性カメラの外観例
3: Example of radiation-resistantcamera(DiakrontSTS-40M）
STS-40M)（Diakont社

Usage
results
使用実績※Ifused only for general confirmation of surface condition (composition distribution), then △※表面状態（組成分布）の大まかな確認にのみ使用する場合は△Issues
課題

・カメラの操作機構の開発が必要(△)
・Developmentof the operating mechanism for camera is required (△)
・照明の種類、配置の最適化検討が必要(△)
・Studyon the types of lighting, layout optimization is required (△)
・水分などによるスペクトルの変動影響評価が必要(△)
・Evaluating
the influence of spectral changes due to moisture, etc. is required (△)
・事前にウランなどの測定対象材料のスペクトルデータの作成が必要(△)
・Advance
creation of spectral data for the material to be measured such as Uranium, etc. is required (△)

・Image recognition using in-vehicle camera
・車載カメラによる画像認識
・PCV内部調査（3号機）で取得した画像評価の一部
・A part of image evaluation acquired by PCV internal investigation (Unit 3)

Applicable
locations １A
適用場所

・Information
related to interior composition of the structure is not acquired (×)
・対象内部の組成情報が得られない（×）

・遮蔽体の設計および耐放射線性の評価が必要(△)
・Shield
design and evaluation of radiation resistance are required (△)

Figure 図4：レーザスキャナーの外観例
4: Example of the appearance of（FARO社
FOCUS150）
a laser scanner(FARO
FOCUS 150)

Visibility camera: 1 second or less per image (Angle of view depends on the lens)Measuring
測定時間size 可視カメラ：1画像あたり1秒以下(画角はレンズによって異なる)
Laser scanner: Several minutes to several tens of minutes (Varies depending on measurement range
and measuring pitch)（測定速度）(Measuring
speed) レーザスキャナー：数分～数十分（測定範囲、測定ピッチによって異なる）

Issues
課題

・Not applicable to damaged objects (×)
・損傷のある対象には適用不可（×）
・It is possible that identification with similar structures is not allowed (×)
・類似の構造物と識別が不可能な可能性あり（×）
・Shield design and evaluation of radiation resistance are required (△)
・遮蔽体の設計および耐放射線性の評価が必要(△)
・Developing the operating mechanism for camera/laser scanner is required (△)
・カメラ／レーザスキャナーの操作機構の開発が必要(△)
・Study on identification process is required (△)
・識別処理の検討が必要（△）
・Training data (learningdata) is required (△)
・トレーニングデータ(学習データ)が必要（△）
・照明の種類、配置の最適化検討が必要(△)
・Study on the types of lighting, layout optimization is required (△)©International Research Institute for Nuclear DecommissioningNo.214
7. Implementation details of this project[2) (iii) Survey on sorting technology for fuel debris and radioactive wastes]◼ Study results on technologies required for sorting (Main measurement technology survey form (2/4))
表-16 測定技術調査票
Table-16 Measurement
technology survey form

Table-7 Measurement
technology survey form
表-7 測定技術調査票
Technology
name パッシブγ線計測技術（CdTeセンサ）
技術名称
Passive gamma ray measurement technology (CdTe sensor)

Principle
原理・Measurement of the charge generated from the interaction between CdTe
semiconductor and gamma rays
・Conversion of the charge into the energy given to CdTe semiconductor by
・CdTe半導体とγ線との相互作用で生じた電荷量を測定
using the conversion coefficient prepared in advance
・事前に準備した変換係数を用いて、電荷量をCdTe半導体への付与エネルギーに変換
・Evaluation of gamma ray emitting radionuclides incident from energy
・エネルギー分布から入射したγ線の放出核種を評価
distribution
・Eu-154由来γ線束の有無やその量から核燃料の有無やその量を推定
・Estimation of the presence or absence of nuclear fuel and its amount fromthe presence or absence of gamma ray flux and its amount originating from
Eu-154
Resolution
分解能

Performance
性能

Lower limit of

測定下限値
measurement

12.9keV @662keV

neutron measurement technology (Advanced type compact B-10 sensor)
Technology
パッシブ中性子計測技術（改良型小型B-10センサ）
技術名称name Passive

Principle
原理

性能
Performance

Depends
on measuring time
測定時間に依存
例：8.6×10 -3 /cm2/s @1274keV
Ex:

Usage

使用環境
environment
使用済み燃料集合体測定時の実績値：19.4Gy/hActualvalue at the time of measurement of spent fuel assembly: 19.4Gy/h（バックグラウンド）

外寸：φ7mm×100mm
Outer
dimensions: Φ7mm x 100mm
ケーブル：同軸ケーブル1本
Cable:
1 coaxial cable

Measuring the response when using spent fuel with the fuel debris detection device

Measuring the response when using spent fuel with the fuel debris detection device

Equipment
CdTe
semiconduc
tor detector

Equipment
CdTe
semiconduc
tor detector

Advancedcompact B-10
detector

Tungsten-shielded watertight container

DeviceShieldCdTe semiconductor detector

configuration /装置構成・サイズ

Watertightcontainer

size

Advanced type compact B-10
detector

Tungsten shield

Item

Outer
dimensions

Cable

Specification
2 coaxial cables (signal, high pressure),
1 preamplifier drive cable: 60m

Advancedcompact B-10
detectorOuter
dimensions

Cable

Outer
dimensions

1 coaxial cable: 60m

Material Tungsten
Weight In air: 8.5kg, in water: 7.0kg
ShieldLower part: 40mm,
thicknessside and back faces: 24mm
Outer

dimensions

Material Aluminum

Device
装置構成・サイズ
configuration /size

Composite Outer diameter
cable
No. of cores Coaxial: 3 cores, simple line: 4 cores(between
Coaxial: 26, simple line: 40
detector-ROV)
Powersource/signal
processing
deviceSignalprocessingPowersupplyTungsten-shielded watertight container
CdTe semiconductor detector

Advanced type compact B-10
detector

Tungsten shield

High wave value processing function, PHA
display function
High voltage power supply: Maximum +1,000V
Preamplifier power supply: NIM standard

ShieldWatertightcontainer

Item

Outer
dimensions

Cable

Outer
dimensions

Cable

Outer
dimensions

Specification
2 coaxial cables (signal, high pressure),
1 preamplifier drive cable: 60m

1 coaxial cable: 60m

Material Tungsten
Weight In air: 8.5kg, in water: 7.0kg
Lower part: 40mm,
Shieldthicknessside and back faces: 24mm

Outer
dimensions

Material Aluminum

Composite
Outer diameter
cable
No. of cores Coaxial: 3 cores, simple line: 4 cores(between
Coaxial: 26, simple line: 40
detector-ROV)
Powersource/signal
processing
deviceSignalHigh wave value processing function, PHA
processing display function
High voltage power supply: Maximum +1,000V
Preamplifier power supply: NIM standardPowersupplyData collection Notebook PC

Data collection Notebook PC

測定環境に依存測定時間
Measuringtime Depends
on measuring environment
例：使用済み燃料集合体測定時の実績：3分以上（測定速度）(Measuring
speed) Ex:
Actual value at the time of spent fuel assembly measurement: 3 minutes or above
・Operation
was confirmed in a radioactive environment using spent fuel assembly
・使用済燃料集合体を利用した放射線環境下で動作確認済み
・To
be used mounted onto ROV in 1F-1PCV internal detailed investigation
・1F-1PCV内部詳細調査でROVに搭載して使用予定

・2-A (Sensor scanning required)・2-A（センサ走査要）
Applicable
locations ・3(2)-A, 3(2)-B, 3(2)-C (Must be combined with other technologies and the results of
適用場所
・3②-A、3②-B、3②-C（他技術や事前分析・解析結果と組合せ要）
pre-analysis/analysis)

課題
IssuesResolution－
分解能
Lower limit of
Depends on measuring time (Thermal neutron sensitivity:cps/nv)
0.2cps/nv)
測定時間に依存（熱中性子感度：0.2
測定下限値
measurement

Usage
線量率耐性：650Gy/h
Doserate tolerance: 650Gy/h
使用環境
environment
累積線量：1MGy以上（付帯ケーブルが制約）（バックグラウンド） Cumulative
dose rate: 1MGy or above (Constrained by accessory cables)(background)(background)Usage
results
使用実績

・Measurement of charge (wave height value) which is correlated with the energyof charged particles generated by nuclear reaction between B-10 and thermal
・B-10と熱中性子との核反応で生じた荷電粒子のエネルギーに相関を持つ電荷量（波高neutrons
値）を測定
・Measurementof wave height spectrum, determining any area as a neutron
・波高値スペクトルを測定し、任意の領域を中性子計測領域と定め、計数値を取得
measuringarea and acquiring values
・事前に取得した熱中性子感度を利用して計数値から熱中性子束を評価
・Evaluationof thermal neutron flux from the values by using thermal neutron
・熱中性子束の有無やその量から核燃料の有無やその量を推定
sensitivity
acquired in advance・Estimation of the presence or absence of nuclear fuel and its amount from thepresence or absence of thermal neutron flux and its amount・Shielding
and collimator design in accordance with applicable environment (△)
・適用環境に応じた遮蔽及びコリメータ設計(△)
・Methodof conversion into Uranium weight (△)
・ウラン重量への換算方法(△)
・Concentration
measurement for isotopic elements (×: 3 (1)-A, B, C)
・同位体元素の濃度測定(×：3①-A,B,C)※他技術や事前分析・解析結果と組合せで測定対象の有無や濃度の高低の推定に使※In
case of combined usage with other technologies, results of pre-analysis/analysis
用する場合には△。
inestimating the presence of target to be measured and concentration levels (△)

Depends on measuring environment
測定環境に依存Measuringtime
測定時間
Ex: Actual results when measuring spent fuel assembly (implemented in internal detailed
例：使用済み燃料集合体測定時の実績（PCV内部詳細調査PJで実施）：1分程度（測定速度）(Measuring
speed)
investigation of PCV PJ): Approx. 1 minute

使用実績
Usage
results

・Operation was confirmed in a radioactive environment using spent fuel assembly
・使用済燃料集合体を利用した放射線環境下で動作確認済み
・To be used mounted onto ROV in 1F-1PCV internal detailed investigation
・1F-1PCV内部詳細調査でROVに搭載して使用予定

・2-A (Sensor scanning required)・2-A（センサ走査要）
Applicable
locations ・3(2)-A, 3(2)-B, 3(2)-C (Must be combined with other technologies and the results of
適用場所
・3②-A、3②-B、3②-C（他技術や事前分析・解析結果と組合せ要）
pre-analysis/analysis)

Issues
課題・Shielding and collimator design in accordance with applicable environment (△)
・指向性（中性子減速材やコリメータの搭載性）（△）
・Method of conversion into Uranium weight (△)
・ウラン重量への換算方法(△)
・Concentration measurement for isotopic elements (×: 3 (1)-A, B, C)
・同位体元素の濃度測定(×：3①-A,B,C)※In
case of combined usage with other technologies, results of pre-analysis/analysis※他技術や事前分析・解析結果と組合せで測定対象の有無や濃度の高低の推定に使
in estimating the presence of target to be measured and concentration levels (△)
用する場合には△。

©International Research Institute for Nuclear DecommissioningNo.215
7. Implementation details of this project[2) (iii) Survey on sorting technology for fuel debris and radioactive wastes]◼ Study results on technologies required for sorting (Main measurement technology survey form (3/4))
Table-20 Measurement
technology survey form
表-20 測定技術調査票

Table-19 Measurement
technology survey form
表-19 測定技術調査票
Technology
name X
X線透過(高エネルギX線CT法)
ray transmission (High energy X ray CT method)
技術名称

原理
Principle

技術名称
Technology
name 宇宙線散乱計測技術（ミュオン散乱法）
Cosmic ray scattering measurement technology (Muon scattering method)
・宇宙線ミュオンが物質を通過するに、・撮像原理は医療用X線CTと同様。
・Imaging
principle is the same as medical X ray CT.
・測定対象物を回転テーブルにより一回転させて撮像。
・Imaging
is carried out by rotating the target to be measured using a rotating table.
・X線源に加速器(LINAC)を使用することにより，大型構造物，高密度物質の・Seismic
tomography for high density material and large structures is possible
断層撮像が可能。
using
accelerator (LINAC) as the X ray source.
※1 http://www.hitachi.co.jp/products/healthcare/products-support/catalog/industrial_ct/pdf/HiXCT-BR-070_j.pdf※2 上坂，他：日本原子力学会2019秋の大会予稿集※2: Uesaka, et al.: Collection of Theses, Atomic Energy Society of Japan 2019
Resolution
分解能

原理
Principle※1
Image resolution: 0.4mm※1
画像分解能：0.4mm

Performance
性能
Lower limit of
測定下限値
measurement

Muon Radiography Data

Usage
使用環境
In a normal environment (Assuming the use outside the PCV of each unit and at
通常環境程度(各号機PCV外，または保管施設での使用を想定)
environment（バックグラウンド）storage facilities)(background)<以下は，※1関連情報><Given below is the information concerning ※1>
構成：X線源（加速器)，回転テーブル，X線検出器(ラインセンサ)
Configuration:
X ray source (accelerator), rotating table, X ray detector (line sensor)
サイズ：概略4m×4m(フットプリント，メンテナンスエリア含む)
Size:
Approx. 4m x 4m (includes footprint, maintenance area)
測定対象物重量：最大100kg程度
Weightof the target to be measured: Maximum 100kg
測定対象物高さ：最大1000㎜程度
Heightof the target to be measured: Maximum 1,000mm
X X線透過能力：鉄320mm，コンクリート1080mm
ray transmission capacity: Iron 320mm, concrete 1,080mm(When 9MV X ray source is used)(9MVのX線源を使用した場合)
Others:Install collimator on the front of the X ray detector to reduce the impact of scattered
その他：X線検出器前面にコリメータを設置し，測定誤差(画質劣化)の原因となる
rays which散乱線の影響を低減
cause measurement error (image degradation).

装置構成・サイズ
Deviceconfiguration /
size

ウラン
通過する物体の原子番号に応じた角度
・Whena cosmic ray muon passes through a
Uranium
substance, the composition in the substance is
で散乱する特性を利用し、散乱角の大
analyzed from the magnitude and spread of the
きさ、広がりから物質中の組成を解析す
scattering
angle by utilizing the property of
scattering
at an angle according to the atomic
る。
number of the object the ray passes through.
・原子番号が大きい物質ほど散乱角が・If
the atomic number of the material is high,scattering
angle is also high; Uranium can be
大きく、鉄などに比べウランを選択的に
selectively evaluated when compared to iron etc.,評価できるが、同位体分析は不可
but isotopic analysis is not possible
・宇宙線を利用するため線源が不要
・Sourceis not required since cosmic rays areused.
・大まかな3次元分布も取得可能
図1
鉄14tonの中の20kgのウラン球を検知
1 Detectionof 20kg Uranium sphere in 14 tons of iron・3 dimensional distribution can also be acquired, Figurealbeit roughly[1]LA-UR-04-3985, Information Extraction from

Object under study Detection device
X ray beam
X ray source(accelerator)

分解能
Resolution
性能
Performance

使用環境
Usage
environment

Lower limit of
測定下限値
measurement

Uranium:
Diameter 8cm, weight 20kg [1]ウラン:直径8cm、重量20kg[1]※Depends on the size of the container, air dose rate in the environment, and measuring※容器の大きさ、環境の空間線量率および測定時間に依存
timeAbout 1 to 10kg when measuring Uranium alone
ウラン単体の測定では1～10kg程度※Depends on the size of the container, air dose rate in the environment, and※容器の大きさ、環境の空間線量率および測定時間に依存
measuring time

Assuming
use in the lowest possible background
可能な限り低バックグラウンドでの使用を前提※The
upper limit of air dose rate changes depending on the size of the detector.
（バックグラウンド）※空間線量率の上限は、検出器の大きさで変わる。
(background)【装置構成】[Device configuration]
２つのミュオン軌跡検出器から構成され、上段のミュオン軌跡検出器で所得した軌跡と、下Made of 2 muon trajectory detectors, and scattering spot and scattering angle are measured
from the trajectories acquired by the muon trajectory detectors at the upper and lower levels.
段のミュオン軌跡検出器での軌跡から、散乱地点および散乱角を測定する。
[Size]【サイズ】
Size is approx. 2.5m x 1.5m x 3m (power supply panel and control panel are not included).
サイズは2.5m×1.5m×3m程度(電源盤、制御盤含まず。)※Optimization required since this depends on environmental dose rate and measuring time.
※環境の線量率や測定時間に依存するため最適化が必要
Muon
Muon trajectory detector (1)Deviceconfiguration /装置構成・サイズ
size

Target to bemeasured

Measuringarea

Drift tube
Muon
trajectory
detector (2)Rotating table

10 to 15 seconds per(第3世代撮像(一回転撮像方式)，シングルエネルギ)
section (※1) (when 3G imaging (single rotation imaging
Measuring
測定時間time 10～15秒/断面(※1)
single energy) → When a unit can was measured (inner height 360mm),(Measuring
speed) system),→ユニット缶(内寸高さ360㎜)を測定対象とした場合，3600～5400秒(約1～1.5時間)（測定速度）
Usage
results
使用実績

3,600 to 5,400 seconds (approx. 1 to 1.5 hours)
・Observation of internal conditions of waste drum (※1)
・廃棄物ドラム缶内部状態観察(※1)
・Fuel assembly inspection (rotation and translational imaging method)
・燃料集合体検査(回転＋並進撮像方式)
石見，他：JAEA-Data/Code 2014-012
Iwami, et al.: JAEA-Data/Code 2014-012
・Various non-destructive inspections (Machine components such as automobile engines etc. (※1), bridge (※2) etc.)
・各種非破壊検査(自動車エンジン等機械部品(※1)，橋梁(※2)等)

Applicable
locations ３①－Ｂ、3①－Ｃ、３②－Ｂ、３②－Ｃ
適用場所

課題
Issues・Only density distribution information is obtained from cross-sectional imaging by
・単一X線エネルギでの断層像から得られるのは，密度分布情報のみだが，single energy X ray, but it is possible to acquire elemental information by a method
複数のX線エネルギを用いる方法(※2)により元素情報を取得し，ウラン等の(※2) using multi-energy X ray and identify the nuclear material such as Uranium
核燃料と，鉄，コンクリートを識別できる可能性あり(△)
etc., iron and concrete (△)
・同位体識別不可。同位体比測定等，他の計測・評価方法等との・Isotopic identification is not possible. Must be combined with other measurements,
組み合わせ要(△)
evaluation methods etc. such as isotopic ratio measurement etc. (△)
・デブリ由来のγ線による測定精度への影響低減(△)
・Reduction of impact of debris-derived gamma rays on measuring accuracy. (△)
・BG低減(△)
・BG reduction (△)

Figure 2:
Exampleof device configuration
図２：
装置構成の例・Ionization during muontransmission
・Multi-layered trajectorymeasurement (6 layers, 2
axes)

FigureConfiguration of detector
図３3:検出器の構成

Measuringtime Approx.
測定時間
1ユニットの測定あたり約数分～1時間[1]several minutes to an hour per measurement of 1 unit
(Measuring
speed) ※Varies（測定速度）※空間線量率および検出器の有効測定面積で変わる
due to effective measurement area of detector and air dose rate

使用実績
Usage
results

・セキュリティスキャナ
・Security
scanner・Results
of measuring spent fuel containing casks (MC-10 cask for PWR) under low dose rate
・低線量下で使用済み燃料入りのキャスク（PWR用MC-10Cask）の測定実績あり
Phys. Rev. Applied 9, 044013,"Verification of Spent Nuclear Fuel in Sealed Dry Storage Casks via
Measurements of Cosmic-Ray Muon Scattering”

Applicable
locations ３①－Ｂ、３①－Ｃ、３②－Ｂ、３②－Ｃ
適用場所
・Muon measuring efficiency is reduced by the impact of gamma rays from debris and BG. If a shield isIssues課題

added, the effect of gamma rays is like to be reduced, but since the precision of the fuel detection
・BGおよびデブリからのγ線の影響でミュオンの測定効率が低下する。遮蔽体を追加するとγ線の影響
system would then deteriorate, it is necessary to conduct a study on optimization (△)
は低減可能だが、燃料の検知精度が低下するため、最適化の検討が必要(△）
・Identification of isotope and measurement of concentration levels cannot be performed, so it is
・同位体の識別、濃縮度が測定できないため、平均濃縮度等から導出する方法の検討が必要(△）
necessary to conduct a study on the derivation method from average concentration level (△)※同位体比や、燃焼度が測定できる手法(燃焼度モニタなど)と組み合わせることで、測定精度、確度の※It is possible to increase measuring accuracy and degree of certainty by combining the techniques
向上の可能性あり。
for measuring isotopic ratio and degree of burn-up (Burn-up monitor etc.)
・1F環境下（放射線、粉塵、振動など）での長期的な運用、メンテナンス方法の検討が必要(△）
・It is necessary to conduct a study on long-term operation and maintenance methods under 1F
environment (radioactivity, dust, vibrations etc.) (△)©International Research Institute for Nuclear DecommissioningNo.216
7. Implementation details of this project[2) (iii) Survey on sorting technology for fuel debris and radioactive wastes]◼ Study results on technologies required for sorting (Main measurement technology survey form (4/4))
Table-21 Measurement
survey form
表-21 technology
測定技術調査票
Active neutron measurement technology (Fast Neutron Direct Interrogation
技術名称name アクティブ中性子計測技術（高速中性子直接問いかけ(FNDI)法）
Technology(FNDI) method)

原理
Principle(1) Sample is irradiated with fast neutrons from neutron generator.
(2) Thermal neutrons inside the system of measuring device and sample are①中性子発生装置から高速中性子を試料に照射
sloweddown.
②試料及び計測装置の体系内で熱中性子に減速される。
(3)
Nuclear fission material contained within the sample induces nuclear③試料内に含まれる核分裂性物質が核分裂を誘発し即発中性子が放出される。
fissionand neutrons are emitted.
④計測体系に設置された中性子検出器により計測される。
(4)
Measurement is carried out with neutron detector installed in measuring⑤核分裂中性子の消滅時間と計測により核物質量を計算評価する。
system.
(5) Nuclear material weight is computationally estimated by measuring theextinction time for fission neutrons.

Resolution
分解能

Performance
性能

核物質(U-235,Pu-239,Pu241)の総量が計測可能
Total amount of nuclear material (U-235, Pu-239, Pu-241) can be measured.
比較的短時間で計測可能。
Measurable in a comparatively shorter time.
Can be applied to large and high density containers because fast neutrons are
高速中性子を使用するために、大型で高密度の容器に適用できる。
used.

Resolution is not known.
分可能に関しては不明。
Lower limit ofUnknownmeasurement
測定下限値
不明

Usage

使用環境
可能な限り低バックグランドでの使用を前提とする。
environment
Assuming use in the lowest possible background（バックグラウンド）(background)

Table-23 Measurement
survey form
表-23 technology
測定技術調査票
パッシブ・アクティブ中性子計測+γ線計測技術（圧縮減容ドラムを対象としたγスペクトロメトリー
Passive/active neutron and gamma ray measurement technology (Gamma
Technology
技術名称name
及び中性子計測装置）
spectrometry for volume compressed drum and neutron measuring device)
1.γスペクトル計測装置
Based on the measuring results obtained with:
2.Passive中性子計測装置
1. Gamma spectrum measuring device
3.Active中性子計測装置

2. Passive neutron measuring device

3. Active neutron measuring deviceで得られる計測結果の基づく原理
Principle

d. Decay heatを評価する。

Resolution
分解能

性能
Performance

Deviceconfiguration /装置構成・サイズ
size

Metal scrap
FNDI method
DT neutron
Detector
generator tube
bank
Drum

Device developed for
Ningyo-toge

核分裂性物質量Amount of nuclear fissile material
・計測結果から得られる統計誤差：約10%
Statistical error obtained from measuring result: approx. 10%
Calibration error: approx. 30%
・校正誤差：約30％

検出限界：
Detection limit
Lower limit of・U235+Pu239+Pu241の合計で100mg
100mg of U-235 + Pu-239 + Pu-241 in total
測定下限値
measurement
・U235+U238+Pu239+Pu241の合計で100g
100g of U-235 + U-238 + Pu-239 + Pu-241 in total
The
gamma
spectrummeasuring device must be limited to sufficiently low background.
γスペクトル計測装置は十分に低いバックグランドに抑える必要がある。
The level of gamma rays from the surroundings and target to be measured must be reduced
中性子計測装置は検出器の特性により検出器周辺で2rad/h以下となるように周辺及び測定対
Usage使用環境
environment
so that the neutron measuring device is maintained at a max. 2rad/h due to the properties of
象からのγ線レベルを低減する必要あり。
（バックグラウンド）(background)the detector.
従って、周辺バックグランドレベルの高い場所は不向き。
Therefore, locations with high surrounding background levels are not suitable.
Drum to be measured: Diameter
approx.
1m; Height approx.
計測対象ドラム：直径
約1m 高さ
約1.5m
1.5m

構成設備
Device configuration①中性子検出器バンク（中性子検出器、減速材等含む)(1) Neutron detector bank (includes neutron detector, moderator etc.)②中性子発生装置(DT管）(2) Neutron generator (DT tube)③信号処理部(3) Signal processing unit
④計算評価用計算機(4) Calculator for computational estimation

It is a structure in which a detector bank is installed around the debris canister. In
デブリ収納缶の周辺に検出器バンクを設置する構造であり、また高速中性子を減速させる
addition, it is necessary to provide a moderator for decelerating the fast neutrons and
ための減速材及び中性子発生装置動作時の周辺の放射線レベルを低減するための遮へ
a shield for reducing the radiation level in the surroundings when the neutron
いを周辺に設ける必要があり、かなり大規模となる。
generator is operating, which makes the size considerably large.
装置サイズ：2m(W)x x2m(D)x 2.2m(H)Device size: 2m (W)
2m (D)
x 2.2m (H)a.全U及び全Pu量評価
Will evaluate:
b.AlphaActivitya. Gross
U and(Pu,Cm,全α)
Pu amounts
c.BetaActivity(Cs137,Sr90-Y90,Pu241)b. Alpha
activity(Pu, Cm, gross alpha)d.崩壊熱
c. Beta activity (Cs-137, Sr-90, Y-90, Pu-241)

Deviceconfiguration /size

装置構成・サイズ

γスペクトル計測装置：
Gamma spectrum measuring device:
高純度Ge検出器
High purity Ge detector
ドラム回転及び上下駆動装置
Drum
rotating and vertical driving device
設置スペース：不明
Installation
space: Not known
中性子計測装置：
Neutronmeasuring device:
計測セル3面方向に各83本の中性子検出器（有感長：1000mm)
合計249本83 neutron
directors each in 3 発生中性子：2E+9
directions of measuringcell
中性子発生装置2セット
n/sec
中性子発生管（2セット）
Neutron generator
tubes (2 sets)(finite length: 1,000mm); 249 total
中性子減速材で構成する計測セル
Neutron generators: 2 sets; generated neutrons: 2E+9n/sec
設置スペース：計測セル10mx10mx10m
Measuringcell made of neutron moderator(中性子発生装置付属機器及び計測装置は含まず）
Installation space: Measuring cell 10m x 10m x 10m(Auxiliary equipment for neutron generator and measuring
device are not included)
Target to be measured

測定対象物（ドラム内）(inside the drum)

Sediment CaF2

計測セル（3面）
Measuring cell
(3 sides)
中性子計測装置
概念図 device
Conceptual diagramof neutron measuring

Concrete shield
Irradiated neutrons

Fission neutrons

Measuringtime
測定時間（測定速度）(Measuring
speed)

Measuring time varies in accordance with required detection limit.
要求される検出限界に応じて測定時間が変化する。
一般的にIn general, it is about 1 hour in total for background measurement of
計測セルのバックグランド計測、収納缶の搬入、計測、搬出の合計で１時間程度。
measuring cell, carrying-in of canister, measurement, carrying-out.

Usageresults
使用実績

JAEA Ningyo-toge
Environmental Engineering Center Measuring
device for measurement and management of Uranium waste (JAWAS-N)
JAEA
人形峠環境技術センター
ウラン廃棄物の計量管理用測定装置(JAWAS-N)

Applicable
適用場所
locationsGamma spectrum measuring
device: 15 to 45min.
～ 45min
Measuringtime γスペクトル計測装置：15
測定時間
Neutron measuring～device:
60min 45 to 60min.
（測定速度）(Measuring
speed) 中性子計測装置：45
Usageresults
使用実績

3 (2)-B, 3 (2)-C, 3 (3)-B, 3 (3)-C (However, the device being large-sized, it is considered difficult to install in R/B and

3②-B、3②-C、3③-B、3③-C

Issues
課題
Issues
課題

収納缶内のデブリの状況把握が最も大きな課題。（△）
Major issue is identifying the status of debris inside the canister.

La Hague圧縮減容施設
La
Hague volume compression facility

3②-B、3②-C、3③-B、3③-C ただし、装置が大型でありRBへの設置は難しいと思われ、装置
Applicable
locations it is assumed that it will be installed in storage preparation facility and new expanded building which may be
適用場所に応じた設計が可能となる新増設建屋や保管準備施設への設置が想定される。
designed in accordance with the device.)
Measurement target is the hull-end piece evacuated from the retreatment plant and the background of
計測対象が再処置工場から排出されるハル・エンドピースであり内容物の素性が明らかである。
thecontents is known.
Onthe other hand, the fuel debris contents cannot be identified, so a major problem is how to
一方、燃料デブリに関しては、内容物を特定できないことから密度、材質等の評価に重要なパラ
evaluate
critical parameters while evaluating density, material etc.
メータをどの様に評価するのかが大きな問題となる。
Accordingly,
in order to understand these uncertain and unknown elements, it is highly likely that there
従って、これらの不確定要素を把握するために、tomographyの様な内容物の密度、分布の状態
is a need for additional installation of instrumentation devices to be able to estimate and evaluate the等を推定評価できる計装装置の追加設置が必要となる可能性が高い。（△）
status of distribution and density of the contents as in case of tomography. (△)
その結果、装置の規模が大きくなることが予想される。（△）
Asa result, the size of the device is expected to increase. (△)©International Research Institute for Nuclear DecommissioningNo.217

7. Implementation details of this project[2) (iii) Survey on sorting technology for fuel debris and radioactive wastes]◼ Study results on technologies required for sorting●Evaluation of each surveyed technology (1/3)

O: Prospects for application△: Feasible
x: Not feasible at this stage-: Not applicable

A: Inside PCV (Before storage in the container)
B: Outside PCV of each Unit (RB, expandedbuilding)
C: Storage facility (During preparations before
storage)[2]

Main specifications

Performance
Technologies
surveyed

Reflectancemeasurement

Image
determination① Resolution② Lower limit ofmeasurement① Radiation tolerance or
performance in theenvironment used
② Equipment size
③ Measuring time

①To the extent that roughsurface composition distribution
can be understood② Evaluation by absolute value
is difficult① Details not known (Shield is
required)② 76 x 76 x 120 mm③ Several seconds per image

・Relevant internal
information is not available
・Creation of spectrum data
for material to bemeasured

A: x
B: x
C: x

① Shield is necessary② Approx.230 x 183 x 103 mm③ Camera: 1 second or lessper image; laser scanner:~several tens of minutes

・Not applicable todamaged objects
・Learning data is required

A: △
B: △
C: △

① Position resolution: About 2cm② Alpha radioactivity: 4 Bq/cm 2① Details not known (Shield is
required)*3② 400 x 300 x 250 mm③ From several seconds toseveral minutes, maximum 1000
seconds per image

・Effects of high dose
gamma ray and beta ray
emission・Relevant internal
information is not available①② Depends on measuring time

① Dose rate tolerance: 350
Gy/h② Sensor size 3 x 3 mm
③ Depends on measuring
environment

・Alpha ray sensitivity
・Method for conversion
into uranium weight・Measurement of isotope
concentration① Types of structures withcharacteristic differences are
identifiable②-

Passive alpharaymeasurement(Alphacamera)Passive alpharaymeasurement[1] Visual(Image)determination*3

Main issues[3①] Formeasuring theconcentrationof nuclear
fissile material[3②] For
criticality
safetycontrol[3③] Fordecision on
terminationof PP andsafeguards

Concentrationof U-235 in fueldebris:Reactivity
equivalent to
1.5 wt% or less

Mass ofnuclearmaterial 3.7
kg / containeror less

Not yetdetermined

-

-

-

-

-

-

-

-

-

-

-

A: x
B: x
C: x

A: △ *4
B: x
C: x

Estimationof nuclear
material
distribution
inside PCV

*1

*2

*1

-

-

A: x

A: △

*1
*4

*1

-

*1: There is a possibility to acquire relevant surface information*2: Identification of nuclear material is not possible. However, might support sorting by shape measurement etc.
*3: Currently under development in Solid Radioactive Waste PJ. Provisional values.
*4: Combination with other sensors and techniques required©International Research Institute for Nuclear DecommissioningNo.218
7. Implementation details of this project[2) (iii) Survey on sorting technology for fuel debris and radioactive wastes]◼ Study results on technologies required for sorting● Evaluation of each surveyed technology (2/3)

O: Prospects for application△: Feasible
x: Not feasible at this stage-: Not applicable

A: Inside PCV (Before storage in the container)
B: Outside PCV of each Unit (RB, expandedbuilding)
C: Storage facility (During preparations before
storage)[2]

Performance[1] Visual(Image)determination

Main specifications① Resolution② Lower limit ofmeasurement① Radiation tolerance or
performance in theenvironment used
② Equipment size
③ Measuring time

Passive
gamma raymeasurement①12.9keV @662keV②Depends on measuring time

①19.4Gy/h (Actual values at
the time of spent fuel assembly
measurement)②Detector size Φ7x51 mm③Depends on measuring
environment

・Shield/collimator
designing
・Method for conversion
into uranium weight・Measurement of isotope
concentration

Passive
neutronmeasurement①②Depends on measuring time

①Dose rate tolerance: 650
Gy/h②Detector size Φ7x100mm③Depends on measuring
environment

・Directivity
・Method for conversion
into uranium weight・Measurement of isotope
concentration-

Active X ray
measurement①②Uranium in the order of μg
is detectable①Laboratory analysis ispresumed②Unknown③Unknown

・Measurement under highBG environment
・Measurement of isotope
concentration
・Sample removal method-①Image resolution: 0.4 mm②-①Normal environment level② (Footprint) 4 x 4 m③10 ~ 15 sec/Section (Unitcan: 1 ~ 1.5 Hr)・Distinguishingiron/concrete fromuranium etc.
・Reduction of impact of
gamma rays originatingfrom BG and fuel debris

Technologies
surveyed

X raytransmission

Main issues

Estimationof nuclear
material
distribution
inside PCV[3①] Formeasuring theconcentrationof nuclear
fissile material[3②] For
criticality
safetycontrol[3③] Fordecision on
terminationof PP andsafeguards

Concentrationof U-235 in fueldebris:Reactivity
equivalent to
1.5 wt% or less

Mass ofnuclearmaterial 3.7
kg /container or
less

Not yetdetermined

*4

-

*4

A: x
B: x
C: x

A: △
B: △
C: △

A: △

A: x
B: x
C: x

A: △
B: △
C: △

-

A: x

A: x
B: x
C: x

A: x
B: x
C: x

-

-

A: B: △
C: △

A: △

*4

*4

*5

-

-

*6

A: B: △
C: △-*4: Combination with other sensors and techniques required*5: Estimate the weight of nuclear material. The conversion of U-235 concentration needs to be reevaluated by assuming the concentration and composition ratio.
*6: Must be re-evaluated by correcting the total weight of the measurement target.
©International Research Institute for Nuclear DecommissioningNo.219

7. Implementation details of this project[2) (iii) Survey on sorting technology for fuel debris and radioactive wastes]◼ Study results on technologies required for sorting● Evaluation of each surveyed technology (3/3)

O: Prospects for application△: Feasible
x: Not feasible at this stage-: Not applicable

A: Inside PCV (Before storage in the container)
B: Outside PCV of each Unit (RB, expanded building)
C: Storage facility (During preparations before
storage)[2]

Technologies
surveyed

Performance[1] Visual(Image)determination

Main specifications① Resolution② Lower limit of measurement① Radiation tolerance or
performance in the environment used
② Equipment size
③ Measuring time

Cosmic rays
scattering
measurement①Uranium: Diameter 8 cm,
weight 20kg
②About 1 ~10 kg for uranium
alone①Upper limit of air dose rate in therange*7of approximately 0.3 ~ 0.5 mSv/h②2.5m x 1.5 m x 3 m③Several minutes to one Hr per unit

・Effects of gamma
rays from BG and fueldebris・Isotope identification,
derivation ofconcentration level

Active neutron
measurement①Unknown (Total quantity of
nuclear material is measurable)②Unknown①Usage at low BG is presumed②2 m x 2 m x 2.2 m③ About 1Hr (BG measurement,
carrying in, measurement, carrying out)・Identifying the statusof material stored in
the container

Passive
neutronmeasurement+ gamma ray
measurement①30% of measured values bycomprehensively evaluating all
error factors②Unknown①To the extent that the waste drum canbe operated by manipulator②Unknown③Unknown

・Identifying the statusof material stored in
the container

Passive/Active
neutronmeasurement+ gamma ray
measurement①Statistical error approx.10%,
Calibration error approx.30%②U235+Pu239+
Pu241: 100 mg
Above-mentioned + U238: 100 g①Unsuitable for high BG areas② (Installation space: Measuring cell) 10
m x 10 m x 10 m
③Gamma spectrum measurement: 15 ~45 minutes, neutron measurement: 45 ~60 minutes

・Identifying the statusof material stored in
the container
・Increased size ofequipment

Main issues

Estimation of
nuclearmaterial
distribution
inside PCV[3①] Formeasuring theconcentrationof nuclear
fissile material[3②] For
criticality
safetycontrol[3③] Fordecision on
termination of
PP andsafeguards

Concentration of
U-235 in fueldebris:Reactivity
equivalent to 1.5
wt% or less

Mass ofnuclearmaterial 3.7
kg / containeror less

Not yetdetermined

*5

-

-

*6

A: B: △C: △A: B: △C: △-A: B: △C: △-A: B: △
C: △

-

*9

*8

-

-

B: △
C: △

*8

-

-

B: x
C: x
*8
*9

*8

-

-

-

A: B: △
C: △

B: △
C: △

*5: Estimate the weight of nuclear material. The conversion of U-235 concentration needs to be re-evaluated by assuming the concentration and composition ratio.
*6: Must be re-evaluated by correcting the total weight of the measurement target.
*7: Performance when the target is loaded into the canister with shielding function*8: Evaluation was conducted on the premise that the composition ratio of fissile material nuclides is known.⇒ Combined usage of other measurement techniques such as estimating thedegree of burn-up in fuel debris.
*9: It is necessary to re-evaluate the detection limit and measurement error when fuel debris is the measurement target.

©International Research Institute for Nuclear DecommissioningNo.220

7. Implementation details of this project[2) (iii) Survey on sorting technology for fuel debris and radioactive wastes]◼ Study results on technologies required for sorting

O: Technology that can be applied△: Feasible technology
x: No feasible technology at this stage●Summary
Application location
No.

Determination (measurement)
technology

1

A: Inside PCV(Before storage in the container)

B: Outside PCV of each Unit
(RB, expanded building)

C: Storage facility(During preparations before storage)

Visual (Image) determination(1-A)△Image determination(1-B)△Image determination(1-C)△Image determination

2

Estimation of nuclear material
distribution inside PCV(2-A)△Passive alpha ray measurement△Passive gamma ray measurement△Passive neutron measurement

3

Measurement of nuclear
material (Nuclides,
concentration)(3①-A)
x(3①-B)△X ray transmission (High energy X ray CT method)△Cosmic rays scattering measurement (Muon scattering
method)(3①-C)△X ray transmission (High energy X ray CT method)△Cosmic rays scattering measurement (Muon scattering
method)(3②-A)△Passive alpha ray measurement△Passive gamma ray measurement△Passive neutron measurement(3②-B)△Passive gamma ray measurement△Passive neutron measurement△X ray transmission (High energy X ray CT method)△Cosmic rays scattering measurement (Muon scattering
method)△Active neutron measurement (FNDI method)△Passive neutron measurement + gamma ray
measurement△ Passive /active neutron measurement + gamma ray
measurement(3②-C)△Passive gamma ray measurement△Passive neutron measurement△X ray transmission (High energy X ray CT method)△Cosmic rays scattering measurement (Muon scattering
method)△Active neutron measurement (FNDI method)△Passive neutron measurement + gamma ray
measurement△Passive /active neutron measurement + gamma ray
measurement-(3③-B)△Active neutron measurement (FNDI method)△Passive /active neutron measurement + gamma ray
measurement(3③-C)△Active neutron measurement (FNDI method)△Passive /active neutron measurement + gamma ray
measurement①For measuring theconcentration of nuclear fissile
material② For criticality safety control③For decision on terminationof PP and safeguards©International Research Institute for Nuclear Decommissioning7. Implementation details of this project[2) (iii) Survey on sorting technology for fuel debris and radioactive wastes]

No.221◼ Evaluation of technical feasibility of fuel debris sorting scenarios (Evaluation of sorting scenarios)①②

Preconditions➢

Do not consider sorting with the same content in multiple locations. (If sorting is performed on the
upstream side, do not perform sorting on the downstream side.)➢

The storage facility (during preparations before storage) has a lower background than outside thePCV of each unit (RB, expanded building), it is easy to secure the installation space for the sorting
equipment, and installability of sorting equipment is higher on the storage facility side.

➢

It is likely that measurement of nuclear material inside PCV is difficult when compared to otherlocations and it is also assumed that handling sorting (Aa1) of fuel debris and wastes inside PCV bymere visual (image) determination might be difficult.

Formulation of sorting scenarios from the perspective of feasibility➢

Scenario A: Scenarios in which sorting is performed at the earliest possible stage➢

Scenario B: Scenarios focusing on the installability of sorting equipment➢

Scenario C: When sorting (Aa1) of fuel debris and waste inside PCV is difficult ① (Sorting at theearliest possible stage)➢

Scenario D: When sorting (Aa1) of fuel debris and waste inside PCV is difficult ② (Focus on
installability of sorting equipment)©International Research Institute for Nuclear Decommissioning7. Implementation details of this projectNo.222[2) (iii) Survey on sorting technology for fuel debris and radioactive wastes]◼

Evaluation of technical feasibility of fuel debris sorting
scenarios (Evaluation of sorting scenarios)

Scenario A: Scenarios in which sorting is performed at the earliest possible stage
Storagefacility (shared)
保管施設（共用）

R/B,
Expanded building (Each Unit)
R/B、増設建屋（各号機）
B:B：PCV外Outside PCV

A: Inside
A：PCV内PCVC:Storage facility (During
C：保管施設（保管前準備時）
preparation before storage)

Shape
形状・/
Size

寸法

PP

Aa3

Ba6

Debrisデブリ

Concentration

Ba

濃度

Criticality

臨界

Inner内径diameter
220mm
220mm ×x Height
1m to 1.5m
高さ 1m～1.5ｍ

Ba4Canister
収納缶

Unit
cans
ユニット缶

Bf

Bd

・Active
neutron・アクティブ中性子計測
measurement
・パッシブ・アクティブ中性子
・Passive/active neutron and
計測＋γ線計測
gamma ray measurement

・Image decision
・Passive alpha ray
・画像判定
measurement
・パッシブα線計測
・Passive gamma ray
・パッシブγ線計測
measurement
・パッシブ中性子計測
・Passive neutron
measurement

Storagefacility
保管施設（デブリ保管）(Debris storage)

Mitigating

a: a：収納缶
Canister・X ray transmission
・X線透過ray・Cosmic
scattering
・宇宙線散乱計測
measurement

Inner 内径diameter
400mm400mm
×x Height
to 1.5m
高さ 1m
1m～1.5ｍb: Mitigatingb：緩和型収納缶
canister

緩和型収納缶canister・A further elongated storage
canister・Waste storage containers
・収納缶を更に長尺化したもの
made with flat plates with a
・廃棄物保管容器を現状の臨界制限
thicknesslesser than the
値以下の厚さの平板状にしたもの
current criticality limit value
等

Ba

Targets
対象物(Before(解体前)
dismantling)

Aa1

目視/
Visual /Criticality
臨界

Criticality
臨界

Container
inside
大型未臨界容器a large subcritical
内容器
container

Large subcritical
大型未臨界容器
canister

Bc6

PP
Bf

Ba

Waste
廃棄物

Waste collection
container(Amount of
廃棄物回収容器
nuclearmaterial
less
（核物質量
臨界未満）
than criticality)

c: Large subcriticalc：大型未臨界容器
container

Storagefacility
保管施設（廃棄物保管）(Waste storage)
Subject
to PP,
PP、保障措置対象
safeguards

Bd
Width
1.42m
x Depth
幅 1.42ｍ
× 奥行
1.42ｍ
1.42m x×Height
高さ 1m 1m
Waste storage
廃棄物保管容器
container（管理レベル）(Controlledlevel)

Bd8

Criticality
臨界
Needs confirmation in order toensure criticality safety

・Passive gamma ray measurement
・パッシブγ線計測
・Passiveneutron measurement
・パッシブ中性子計測
・X
ray transmission
・Cosmic
・X線透過ray scattering measurement
・Active
neutron measurement
・宇宙線散乱計測
・Passiveneutron and gamma ray
・アクティブ中性子計測
measurement
・パッシブ中性子計測＋γ線計測
・Passive/active
neutron and gamma ray
・パッシブ・アクティブ中性子計測＋γ線計測
measurement

・Active neutron
・アクティブ中性子計測
measurement
・パッシブ・アクティブ中性子
・Passive/active
neutron and
計測＋γ線計測gamma
ray measurement

d: Waste storaged：廃棄物保管容器
container（管理レベル）(Controlledlevel)

Bd7

PP

Bf
Waste storage
container
廃棄物保管容器(Contaminated（汚染レベル）level)

PP、保障措置対象外
Notsubject to PP,
safeguards
f: Waste storage
f：廃棄物保管容器
container(Contaminated（汚染レベル）level)©International Research Institute for Nuclear Decommissioning7. Implementation details of this projectNo.223[2) (iii) Survey on sorting technology for fuel debris and radioactive wastes]◼

Evaluation of technical feasibility of fuel debris sorting
scenarios (Evaluation of sorting scenarios)

Scenario B: Scenarios focusing on the installability of sorting
equipment

R/B,
Expanded building (Each Unit)
R/B、増設建屋（各号機）
B:B：PCV外Outside PCV

A: Inside
A：PCV内PCV

Storagefacility (shared)
保管施設（共用）
C:
Storage facility (During
C：保管施設（保管前準備時）
preparation before storage)

Shape
形状・
/ Size
寸法

Ba

Aa3

デブリ
DebrisConcentration
濃度

PP
臨界

Criticality

Ca6

ユニット缶
Unit
cans

収納缶
Canister・Active neutron
・アクティブ中性子計測
measurement
・パッシブ・アクティブ中性子
・Passive/active neutron and
計測＋γ線計測
gamma ray measurement

Inner diameter
内径 220mm220mm
×x Height
to 1.5m
高さ1m
1m～1.5ｍ

Ca4収納缶
CanisterCf・Image decision
・画像判定alpha ray
・Passivemeasurement
・パッシブα線計測
・Passivegamma ray・パッシブγ線計測
measurement
・パッシブ中性子計測
・Passiveneutronmeasurement

保管施設（デブリ保管）
Storagefacility(Debris storage)
Ca

Cd

Inner 内径diameter
400mm400mm
×x Height
to 1.5m
高さ1m
1m～1.5ｍ

・X ray transmission
・X線透過ray・Cosmic
scattering
・宇宙線散乱計測
measurement

Mitigating

緩和型収納缶canisterAa1

・A further elongatedstorage canister

Container inside
大型未臨界容器a large subcritical
内容器
container

Needs confirmation in order toensure criticality safety

c:c：大型未臨界容器
Large subcriticalcontainer

Cc6

Ba

廃棄物
Waste

Criticality
臨界

Large subcritical
大型未臨界容器
canister

目視/
Visual /Criticality
臨界

Waste collection
container(Amount of
廃棄物回収容器
nuclearmaterial
less
（核物質量
臨界未満）
than criticality)b: Mitigatingb：緩和型収納缶
canister・Waste storage containers
・収納缶を更に長尺化したもの
made with flat plates with a
・廃棄物保管容器を現状の臨界制限
thicknesslesser than the
値以下の厚さの平板状にしたもの
current criticality limit value
等

Ca

Targets
対象物(Before(解体前)
dismantling)a: Canister
a：収納缶

・Passivegamma ray measurement
・パッシブγ線計測
・Passiveneutron measurement
・パッシブ中性子計測
・X ray transmission
・X線透過・Cosmic ray scattering measurement
・宇宙線散乱計測
・Active
neutron measurement
・Passiveneutron and gamma ray
・アクティブ中性子計測
measurement
・パッシブ中性子計測＋γ線計測
・Passive/active neutron and gamma ray
・パッシブ・アクティブ中性子計測＋γ線計測
measurement

PP

Storagefacility
保管施設（廃棄物保管）(Waste storage)
Subject
to PP,
PP、保障措置対象
safeguards

Cf

Cd

Waste storage
container(Amount of
廃棄物保管容器
nuclear material less
（核物質量
臨界未満）
than criticality)

Bd8

臨界Criticality

Waste storage
廃棄物保管容器
container（管理レベル）(Controlledlevel)

Width
1.42m
x Depth
幅 1.42ｍ
× 奥行
1.42ｍ
1.42m x×Height
高さ 1m 1m
d: Waste storaged：廃棄物保管容器
container（管理レベル）(Controlledlevel)

Cd7

・Active neutron
・アクティブ中性子計測
measurement
・パッシブ・アクティブ中性子
・Passive/active
neutron and
計測＋γ線計測gamma
ray measurement

PP

Cf

Waste storage
container
廃棄物保管容器(Contaminated（汚染レベル）level)

Not
subject to PP,
PP、保障措置対象外
safeguards

f: Waste storage
f：廃棄物保管容器
container（汚染レベル）level)(Contaminated©International Research Institute for Nuclear DecommissioningNo.224

7. Implementation details of this project[2) (iii) Survey on sorting technology for fuel debris and radioactive wastes]◼

Evaluation of technical feasibility of fuel debris sorting
scenarios (Evaluation of sorting scenarios)

Scenario C: When sorting (Aa1) of fuel debris and waste inside PCV is difficult① (Sorting at the earliest possible stage)

R/B,
Expanded building (Each Unit)
R/B、増設建屋（各号機）
B:B：PCV外Outside PCV

A: Inside
A：PCV内PCV

形状・Shape /
Size
寸法

PP

Aa3

Ba6

Concentration

Bf

Ba

濃度

Criticality

臨界

Unit
cans
ユニット缶

デブリ
DebrisStoragefacility (shared)
保管施設（共用）
C：保管施設（保管前準備時）
C:
Storage facility

保管施設（デブリ保管）
Storagefacility(Debris storage)
内径 220mm
×Inner diameter220mm
高さ
1m～1.5ｍx Height
1m
to 1.5mBa4a: a：収納缶
CanisterCanister
収納缶

Bd・Active neutron
・アクティブ中性子計測
measurement
・Passive/active
neutron・パッシブ・アクティブ中性子
andgamma ray
計測＋γ線計測
measurement(During preparation
before storage)・X
ray transmission
・X線透過・Cosmic ray scattering
・宇宙線散乱計測
measurement

Mitigating

緩和型収納缶canister

Inner diameter
400mm
内径 400mm×x Height高さ1m
to 1.5m
1m～1.5ｍb: Mitigatingb：緩和型収納缶
canister・A further elongated storage
canister・Waste storage containers
・収納缶を更に長尺化したもの
made with flat plates with a
・廃棄物保管容器を現状の臨界制限
thicknesslesser than the
値以下の厚さの平板状にしたもの
current criticality limit value
等

Ba

Targets
対象物(Before(解体前)
dismantling)

Aa1

Criticality
臨界

Container inside
大型未臨界容器a large subcritical
内容器
container

Bc6

Large
subcritical
大型未臨界容器
canisterPP

c: Large

subcriticalc：大型未臨界容器
container

Storagefacility
保管施設（廃棄物保管）(Waste storage)
PP、保障措置対象
Subject
to PP,
safeguards

Bf

Bd

Waste
廃棄物

Waste storage
廃棄物保管容器
container（管理レベル）(Controlledlevel)
Bf

Waste storage
廃棄物保管容器
container（汚染レベル）level)(Contaminated

Width
1.42m
x Depth
幅 1.42ｍ
× 奥行
1.42ｍ
1.42m ×x Height
高さ 1m 1m
d: Waste storaged：廃棄物保管容器
container（管理レベル）(Controlledlevel)

Not
subject to PP,
PP、保障措置対象外
safeguards
f: Waste storage
f：廃棄物保管容器
container(Contaminated（汚染レベル）level)©International Research Institute for Nuclear DecommissioningNo.225

7. Implementation details of this project[2) (iii) Survey on sorting technology for fuel debris and radioactive wastes]◼

Evaluation of technical feasibility of fuel debris sorting
scenarios (Evaluation of sorting scenarios)

Scenario D: When sorting (Aa1) of fuel debris and waste inside PCV is difficult② (Focus on installability of sorting equipment)
Storagefacility (shared)
保管施設（共用）

R/B, R/B、増設建屋（各号機）
Expanded building (Each Unit)
B：PCV外B:
Outside PCVC:Storage facility (During
C：保管施設（保管前準備時）
preparation before storage)

形状・/
Shape
Size

A: Inside
A：PCV内PCV
Debris
デブリ

寸法

PP

Aa3

Ca6

ユニット缶
Unit
cans

Ca

Concentration
濃度
Criticality
臨界

Canister
収納缶

Cf・Active neutron
・アクティブ中性子計測
measurement
・Passive/active
neutron・パッシブ・アクティブ中性子
andgamma ray
計測＋γ線計測
measurement

Storagefacility
保管施設（デブリ保管）(Debris storage)
内径 220mm
×Inner diameter220mm
高さ
1m～1.5ｍx Height
1m
to 1.5m

Ca4

Cd

・X
ray transmission
・X線透過・Cosmic ray scattering
・宇宙線散乱計測
measurement

内径 400mm
×Inner diameter
400mm
x Height
1m
to 1.5m
高さ
1m～1.5ｍ

Mitigating
緩和型収納缶Aa1

Container insidea 大型未臨界容器
large subcritical
内容器
container等

c: Large

c：大型未臨界容器
subcriticalcontainer

Cc6

PP

保管施設（廃棄物保管）
Storagefacility(Wastestorage)
PP、保障措置対象
Subject to PP, safeguards

Cf

Cd

Waste
廃棄物

canister・A further elongatedstorage canister
・Waste storage containers
・収納缶を更に長尺化したもの
made with flat plates with a
・廃棄物保管容器を現状の臨界制限
thickness lesser than the
値以下の厚さの平板状にしたもの
current
criticality limit value

Criticality
臨界

Large
subcritical
大型未臨界容器
canisterb: Mitigatingb：緩和型収納缶

canisterCa

Targets
対象物(Before(解体前)
dismantling)a: a：収納缶
Canister

収納缶
Canister

Waste storage
廃棄物保管容器
container（管理レベル）(Controlledlevel)
Cf

Waste storage
container
廃棄物保管容器(Contaminated（汚染レベル）level)

Width
1.42m
x Depth
幅 1.42ｍ
× 奥行
1.42ｍ
1.42m x×Height
高さ 1m1m
d: Waste storaged：廃棄物保管容器
container（管理レベル）(Controlledlevel)

Not
subject to PP,
PP、保障措置対象外
safeguards
f: Waste storage
f：廃棄物保管容器
container（汚染レベル） level)(Contaminated©International Research Institute for Nuclear DecommissioningNo.226

7. Implementation details of this project[2) (iii) Survey on sorting technology for fuel debris and radioactive wastes]◼ Evaluation of technical feasibility of fuel debris sorting scenarios (Comparison of sorting scenarios)
Sorting scenario (Mainapplication location)

MeritsDrawbacks➢

Scenario A
A: Inside PCV
B: Outside PCV

Sorting at theearliest possible
stage

➢
➢

➢

Scenario B
A: Inside PCV
C: Storage
facility

Focus oninstallability ofsorting equipment➢

Scenario C(B: Outside
PCV)

When sorting (Aa1)of fuel debris andwaste inside PCV
is difficult ①(Sorting at theearliest possible
stage)➢

Scenario D(C: Storage
facility)

When sorting (Aa1)of fuel debris andwaste inside PCV
is difficult②(Focus on
installability ofsorting equipment)

➢

➢
➢

➢
➢

Hot cell can be shared
Amount of transferredmaterial can be optimized
since sorting is completed
before transportation

Sorting equipment can besharedIt is easier to secure theinstallation site for sorting
equipment when comparedwith Scenario A➢➢➢

Work inside PCV is lesswhen compared to
Scenarios A and B
Sorting points are less whencompared to Scenario A(Merits are same as
Scenario A)➢

Work inside PCV is lesswhen compared to
Scenarios A and B
Sorting points are less whencompared to Scenario B(Merits are same as
Scenario B)

➢

➢
➢

➢
➢
➢

Requires sorting equipment forevery Unit
Since fuel debris retrieval
equipment and sorting
equipment coexist, there may beproblems in securing a place for
installation

Storage facility requires hot cells
Less effective in reducing theamount of material transferred incanisters when compared to
Scenario A
Delivered quantity of unit cans ismore when compared toscenarios A and B
Ba6 treated quantity increases
when compared to Scenario A(Drawbacks are same as
Scenario A)
Delivered quantity in case of unitcans is more when compared toscenarios A and B
Ca6 treated quantity increases
when compared to Scenario B
Storage facility requires hot cells
Amount transferred in canisters
is the highest of all the scenarios

Evaluation
➢

➢

➢

➢

Sorting has major benefits suchas optimization of transferred
amount, but issues associatedwith installability of equipment are
expected
Difficulty level of sorting inside
PCV is highHas merits such as sharing ofequipment and installability, butless effective in reducing theamount of material transferred incanisters
Difficulty level of sorting inside
PCV is high.

➢Sorting work inside PCV is less,but issues associated withinstallability of equipment are
expected➢

Sorting work inside and outside
PCV is the least and the level ofdifficulty is low, but it is lesseffective in reducing the volumeof unit cans delivered and amountof material transferred incanisters©International Research Institute for Nuclear Decommissioning7. Implementation details of this projectNo.227[2) (iii) Survey on sorting technology for fuel debris and radioactive wastes]◼ Summary (Issues and future action)

Issues① The technology survey revealed that at present there is no technology that can be applied tosorting technology as is, and in case of technology that may be feasible, it is necessary totake countermeasures for each issue, for example, understanding the properties of themeasurement target and correcting the measurement errors. <Measurement technology>② Since it is highly difficult to measure nuclear material inside PCV, combination of multiple
techniques is required to implement sorting. <Measurement technology>③ When loading into a waste container, some kind of confirmation method for safety isdefinitely required in order to “Secure” criticality safety and this method of measurement
needs to be developed. <Measurement technology>④ The amount of fuel debris stored can be reduced in any scenario. However, sorting inside
PCV is effective for reducing retrieval workload and sorting at an earlier stage (inside oroutside PCV) will be effective in achieving an improved effect of reduction in the amount oftransported material. On the other hand, sorting at an early stage has a lot of installability
issues pertaining to setup of background level, space for installation etc. and leads toconflicting effects. <Scenario>⑤ Sorting itself leads to increase in costs and it is necessary to evaluate the relativeadvantages and disadvantages of cost reduction effects of fuel debris storage and waste
storage by sorting. <Scenario>©International Research Institute for Nuclear Decommissioning7. Implementation details of this projectNo.228[2) (iii) Survey on sorting technology for fuel debris and radioactive wastes]◼ Summary (Issues and future action)

Future course of action
Matters to be taken up in future for implementation of sorting① For finalizing the scenarios, it is necessary to decide on implementation of
scenario and decide on the stage at which sorting is to be carried out. It isnecessary to select the scenarios based on the overall plan for fuel debris
retrieval and a study by operators is required. <Scenario>② It is essential to develop the technologies for image determination and nuclear
material measurement required for each scenario. It is difficult to judge theapplicability by conducting technological survey focusing on published informationand literature. Hence, for future development, it is important to study matters suchas measuring instruments, shield structure, data analysis technologies etc.
together with the vendors. <Measurement technology>③ In this study, scenarios were examined assuming the implementation of sorting.
In future, it will be necessary to verify the effectiveness of sorting by determining
the concrete details of the specifications, layout study, economic evaluation etc.
of sorting equipment. <Scenario>©International Research Institute for Nuclear DecommissioningNo.229

7. Implementation items in this project3) Development of technology for ensuring safety while retrieving fueldebris(i) Development of element technology for confinement functions① Technology for prediction of dust behaviors in PCV
Confinement function is essential to ensure the safety of the public and workers.
The project aims at conducting research and development towards combined
technology for the analysis of aerosol behavior and airflow analysis inside PCV,
which is essential for predicting the behavior of the dust containing alpha nuclides
produced during fuel debris processing, and also aims at enhancing the analysis
model for predicting the behavior inside R/B. In this way, this project aims at
conducting a study on predicting the effects of removing aerosol contained in thedust, selecting appropriate monitoring technology and monitoring positions.
Main study items for development include the following:a. Airflow analysis
・ Realistic dust behavior evaluation techniques were enhanced.
・ Airflow analysis was performed for containment vessel and entire R/B.
・ Evaluation was conducted to find out the impact of heat input on airflow.

©International Research Institute for Nuclear Decommissioning7. Implementation details of this project[3) (i) ① Technology for prediction of dust behaviors in PCV]
No.238~

No.233~No.241~

Airflow analysis for containment
vessel and entire R/B

Assessment of aerosol
behavior in R/B
No.247~

Enhancement of realistic
methods for dust behavior
assessment

Assessment of the impact ofheat input on airflow

Enhancement of FY2018GOTHIC model

Dust occurrence conditions (Dustcollection PJ, Fuel debris
characterization PJ)

No.230Target of thissubsidized project

Reflection of the damaged condition ofbuildings and design study in the model

Study on secondary
containment vessel

Implemented from theoutcomes of engineeringand other projects©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[3) (i) ① Technology for prediction of dust behaviors in PCV]

No.231⚫

Purpose of Development➢ From the study conducted in the previous subsidy projects, a method for predicting dust behaviors inside PCV
using the GOTHIC code was created and the possibility of applying it to system design was obtained. Our goal
this year is to study the issues that were not studied last year and to achieve the following two items:•
Formulation of requirement specifications (ventilation rate, negative pressure) for R/B air tightness and gas
management system, to secure the confinement function of radioactive dust in the event of an abnormality•
Identification of effective monitoring positions from the viewpoint of reducing the exposure to the public andworkers while monitoring within the R/B, in the event of an abnormal dust leak from PCV⚫

Issues to be solved➢ Study on evaluation methods considering resuspension in the aerosol removal mechanism.
➢ Study of impact on airflow analysis considering the R/B.
➢ Study of the effects on airflow due to the heat input during decay heat / fuel debris processing.

⚫

Approaches to development➢ Advancement of realistic dust behavior evaluation method➢ Airflow analysis in containment vessel and entire R/B➢ Evaluating the behavior of aerosol in R/B➢ Evaluating the impact of input heat on airflow⚫

Expected outcome➢ Requirement specifications (ventilation rate, negative pressure) for R/B
air tightness and gas management system to secure the confinement
function in the event of an abnormality➢ Identification of effective monitoring positions from the viewpoint of reducing
the exposure to the public and workers while monitoring within R/B, in theevent of an abnormal leak from PCVOpening (1)↓開口①

複数開口・入熱等の対流への影響把握

Understanding the impact ofmultiple openings, input heat etc.
on convection

エアロゾル挙動
模擬の高度化

Advancement of aerosol
behavior simulation

Opening(2)↓開口②Input heat
崩壊熱・デブリaccompanying
切削に伴う入熱
decay heat anddebris cutting

二次バウン
ダリ機能等に対する要件
把握

Understanding therequirements related
to secondary
boundary function©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[3) (i) ① Technology for prediction of dust behaviors in PCV]

No.232◼ Development Schedule
Study Items

Major milestones

FY 2019
4

5

6

7

8

9

10

11

Interim Report
▼

12

1

2

3

Year-endreport▼1. Advancement of realistic dust behavior
evaluation method
2. Airflow analysis in containment vessel
and entire R/B
3. Evaluating the behavior of aerosol in
R/B

4. Evaluating the impact of input heat on
airflow
5. Summary
Remarks©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[3) (i) ① Technology for prediction of dust behaviors in PCV]
① Advancement of realistic dust behavior evaluation method (1/5)

No.233✓ Implementation items for this year [Ｉ] - Creation of evaluation methods considering dust removal mechanisms
other than gravitational sedimentation ・ Applying the knowledge related to dust removal mechanisms other than gravitational sedimentation→ Investigation was conducted on each of the GOTHIC removal mechanisms and phenomenon analysis
models. As a result of the investigation, it was found that the aerosol behavior in the environment inside the 1F
containment vessel can be evaluated by appropriately considering gravitational sedimentation.
・ Conducting study on appropriate spatial dust concentration distribution evaluation methods after consideringvarious removal mechanisms→ While evaluating spatial concentration distribution, the results show that the mesh sensitivity (especially in
the gravitational direction) is high when gravitational sedimentation is dominant. By changing to a modelwith fine mesh, it has been possible to properly evaluate the dust floating in the air (and amount transferredthrough exhaust).
✓ Implementation items for this year [ＩＩ] -Confirming the effect of dust removal using a moist environment・ Confirming the effect of accelerated removal using particle growth→ The growth of particles in a moist environment happens only in the case of hygroscopic aerosols. As it isconsidered that the dust generated by fuel debris retrieval does not include hygroscopic particles, it became
clear that it is not necessary to pay special attention to the moist environment.
・ Confirming the effect of accelerated removal by dropping condensed water→ The removal effect using the water flowing like raindrops that were observed inside the PCV internal survey
was studied. However, this does not work like spray removal and was less effective as well. From this, it wasunderstood that it need not be considered.
©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[3) (i) ① Technology for prediction of dust behaviors in PCV]
① Advancement of realistic dust behavior evaluation method (2/5)

No.234✓ Implementation items for this year [ＩＩＩ] - Creation of dust behavior evaluation methods consideringresuspension ・ Studying the methods for appropriately simulating the behavior of resuspended dust
→ The GOTHIC V8.3 resuspension model has been created for droplet entrainment at the twophase flow boundary and it was concluded that it cannot be applied for aerosol analysis at 1F.
Though it became clear that a mechanistic model can be incorporated, from the flow rate it
became clear that resuspension will not occur. For this reason, it was clear that dust suspension
need not be considered.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[3) (i) ① Technology for prediction of dust behaviors in PCV]
① Advancement of realistic dust behavior evaluation method (3/5)
Mechanism

Controlling
factors fortheenvironment

No.235

Study on applicability

PCV
environment

Airflow analysis handling policy

Natural removal

Gravitational
sedimentation

None

Its applicability was confirmed through the validity confirmation conducted in
FY2018－

As it is dominant in natural removal and its applicability has
been confirmed, the GOTHIC model shall always beenabled and evaluated.

Inertial
impaction

Stokesnumber(Flow rate
Nozzlediameter)・Its applicability in the Stk number region (~ 0.3) was confirmed, where thecapturing rate of the GOTHIC model is approximately 50%
・ Though inertial impaction is less likely to occur when the Stk number issmall, it tends to be non-conservative as it is possible for the GOTHIC model
to calculate the removal even in such areas.
・However, as the capturing rate evaluated in the low Stk number region isextremely small (< 1E-5) and is negligible, it has no effect.

Low Stk region(~ 1E-5)

・As the flow rates are such that do not cause inertial
deposition inside the PCV environment, the GOTHIC model
shall be disabled and evaluated. (As it tends to be nonconservative, there is virtually no problem even whenenabled)・In situations where a high Stk number is expected, it isenabled, as the impact will be large and the applicability canalso be confirmed.

Diffusion
deposition

Flow rate

・Diffusion deposition is a phenomenon that occurs near the wall surface (~1E-4m). However, even if the aerosol is transported near the wall surface due
to turbulence, the effect is limited as the area targeted for diffusion depositionis extremely small with respect to the entire space.
・The diffusion model of GOTHIC is a model that fits the test results assuming
in-tube gas-liquid two-phase flow (turbulent flow) and does not simulate a dry
state in a large space.

Extremely
large spacecompared tothe inside of
the pipe, etc.

・Turbulent diffusion acts in limited regions and does nothave significant impact in large spaces when compared togravitational sedimentation. Therefore, it is disabled andevaluated.
・When focusing on turbulence in piping, it shall be enabled
and evaluated.

Thermophoresis

Temperature
gradient

・In GOTHIC, it occurs on the surface of the heat sink・It was confirmed that the GOTHIC model is a model based on the thermal
kinetics of the medium gas. Therefore, it is deemed to be applicable in a wide
temperature range.
・However, in situations where the temperature difference is extremely small, it
has no effect as gravitational sedimentation becomes dominant.
・It may contribute to situation with particles with size to the order of submicron
・It has limited effect in large systems

Expected
temperature
difference isapproximately
5℃

・ Though its usability was confirmed, it has no effect incases where there is almost no temperature difference. So,
it can be disabled and evaluated. (Not a problem even if
enabled)・In small systems such as pipes where there is a large
temperature difference as well as handling of small particle
sizes, it shall be enabled and evaluated.

Diffusiophoresis

Concentration
gradient

・In GOTHIC, this happens when water vapor condensation occurs on a cold
wall surface.
・When calculating the amount of water vapor transferred in an environment
with less concentration difference (when diluted water vapor is contained innitrogen gas), it was confirmed that the value calculated with GOTHIC isseveral tens of times larger than when evaluated using the diffusion
coefficient.
・The reason for this is thought to be that GOTHIC calculates the amount ofwater vapor transfer from the water vapor condensation rate, but thecondensation rate and gas diffusion transfer are different.
・The deposition rates of thermophoresis and diffusiophoresis are generally
the same, however, there are concerns on the GOTHIC evaluated values

The expectedconcentrationdifference isapproximately(0.3 mol / m3)

・As it is not effective in situations where there is almost no
difference in concentration (almost no water vapor
condensation occurs), it is not necessary to consider it. So, it
can be disabled and evaluated.
(Even if there is a concentration difference, it should not beenabled as it may overestimate the deposition.)(Or water
vapor
condensation
rate)

Note)
Difference insaturatedwater vapor
concentration
at 5℃©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[3) (i) ① Technology for prediction of dust behaviors in PCV]
① Advancement of realistic dust behavior evaluation method (4/5)

No.236

Moisture EffectMechanism

Controlling
factors fortheenvironment

Study on applicability

1F PCVenvironment

Airflow analysis handling policy

Growth of
hygroscopic
particles

Relative
humidity・Though there is an accelerated removal effect due to theabsorption of moisture by the aerosols resulting in increased
particle size, it has not been modelled in GOTHIC.
・ Some of the particles with remarkable particle growth areaerosol with deliquescent properties such as sea salt particles,
zeolites, silica gel, etc., that absorb moisture. (These absorb
moisture even if the relative humidity is less than 100%)
・The particle growth of aerosols other than these (hydrophobic)
due to moisture absorption is limited.
・U3O8, Fe oxide, etc., have been reported to have the effect ofpromoting gravitational sedimentation by changing the shapefactor during the aggregation process in a moist environment.

Maximum
relative
humidity is
100%

・It is unlikely for aerosols with hygroscopicity
to be generated due to mechanical cutting.
・Non-hygroscopic aerosols may have someparticle growth, but their effects are limited.
・Therefore, as it is not necessary to consider
particle growth due to moisture absorption
even in a moist environment, no newmodeling shall be performed. (However,when considering polydisperse, there arecases where it shall be taken into
consideration depending on the aerosol
properties)

Removalusing water
falling likeraindrops

Dropletdensity
Temperature
difference

・A spray effect can be considered as a removal mechanism
using raindrops
・As a method for aerosol removal using spray, GOTHIC deals
with agglomeration of spray droplets and aerosols
・Turbulent diffusion and gravitational sedimentation are
considered for agglomeration.
・However, the removal mechanism using spray is generally due
to inertial impaction, interception, brown diffusion, and phoresisthat are related to airflows around the droplet.
・When the number density of droplets becomes large as in theengineered spray, the target range for removal becomes large
and removal effect can be expected.
・In cases where the droplet temperature is low and vapor
condensation occurs on the droplet surface, phoresis systemremoval mechanism works, and a significant capturing rate canbe expected.

Low droplet
densitySmalltemperature
difference(approx.
5℃)

・As the GOTHIC removal model is notbased on airflows around the droplet, it isnecessary to incorporate the model
separately.
・However, with the amount of raindrops
caused by cooling water and the amount of
evaporation and condensation of drain water,
density of droplets is low and spray removal
is not possible.
・In addition, when there is no temperature
difference between the droplet and the space,
the phoresis system removal does not work,and the removal efficiency is extremely small
・Therefore, it is not necessary to consider
removal using raindrop-like water that iscurrently seen, and new modeling shall not
be performed.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[3) (i) ① Technology for prediction of dust behaviors in PCV]
① Advancement of realistic dust behavior evaluation method (5/5)

No.237

Mechanism

Controlling
factors fortheenvironment

Study on applicability

1F PCVenvironment

Airflow analysis handling policy

Resuspension(Dryenvironment)

Flow rate

・The GOTHIC V8.3 resuspension model has beencreated for droplet entrainment at the cyclic gas-liquid
two-phase flow boundary and is not applicable to dry
conditions.
・Though re-scattering of primary particles depositedunder dry conditions is unlikely to occur, the fineagglomerated particles re-scatter and form a powder
layer when the amount of deposition increases.
・It is considered possible to create a model that
evaluates the amount of resuspension based on themechanical balance between shear stress andadhesive force.
・However, continuous re-scattering does not occur at
a flow rate of less than 10 m/s.

・Expected
flow rate isless than 0.6
m/s (inside
PCV)
Less than 0.3
m/s(In R/B)

・GOTHIC resuspension model cannot
be applied・Though it is possible to incorporate amodel to evaluate the re-scattering of
agglomerated particles, there is notenough airflow to cause continuous
resuspension even in situations where
dry aerosols are deposited on dry
surfaces.
・ Therefore, no new modeling shall be
performed.

#99%
cumulative
value©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[3) (i) ① Technology for prediction of dust behaviors in PCV]
② Airflow analysis in containment vessel and entire R/B (1/3)

No.238✓ Creation of GOTHIC analysis model considering R/B→ Surveys were conducted on the characteristics of each simulation method of GOTHIC codeand construction methods for the building model were studied. Though it was not possible tocalculate the transport of aerosols through vertical flow path using the previous surveys, an R/Bmodel using vertical and horizontal flow paths was constructed as transportation accompanied
by airflows can be evaluated.

✓ Confirmation on the atmospheric pressure followability of PCV internal pressure taking R/B into
consideration, and as a result, the evaluation on the maintainability of negative pressure→ After the construction of R/B model, the pressure behavior was evaluated under the conditionof fluctuating atmospheric pressure and it was confirmed that, like the evaluations in FY2018,
the PCV pressure does not follow atmospheric pressure.
✓ Understanding the airtightness of the building using the damage status of R/B→ The pressure behavior was evaluated with the building damage conditions. If the building issound, then airtightness can be ensured using negative pressure. Also, even in cases where theoperating floor is open to the atmosphere, it was confirmed that it is possible to ensure
airtightness using negative pressure by ventilating air from the lower floor and narrowing the flow
path to the operating floor.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[3) (i) ① Technology for prediction of dust behaviors in PCV]
② Airflow analysis in containment vessel and entire R/B (2/3) ~Analysis Conditions~[Unit 2: Absence of damage in building]Operati✓ Exhaust at 10,000 [m3/h] from the operating floor
ng floor(Uniform flow rate boundary)✓ Pressure boundaries due to damage in operating floor are notconsidered[Unit 3: Presence of damage in building]✓ Assuming damage in operating floor ceiling (pressure
4thboundary)floor✓ Exhaust at 10,000 [m3/h] from the 4th floor of the building(Uniform flow rate boundary)✓ Negative pressure countermeasures for the building areexpected to be implemented in the area of the stairs leadingfrom the 4th floor to the operating floor(The flow path area is set to 1/100)[R/B shared items]✓ Secondary containment on the floors above the ground(1st to 5th floor) are considered as target for modelling✓ The equipment hatch between the building floors is kept
open. However, the equipment hatch on the operating
floor is closed.
✓ The staircase door is open
✓ The flow path between floors are staircases andequipment hatches, and each of them are modeled tohave vertical flow paths✓ Though the shutter used for carrying equipment in/out
remains shut, it is considered as the atmosphericpressure isobaric boundary due to lack of airtightness✓ Areas on each floor (PCV and MS tunnel rooms) thatare not included in the free space are closed using
blockages

No.239

Damage on operating floor (Unit 3)

Operatingfloor

R/B exhaust(Unit 2)

R/B exhaust(Unit 3)

PCV

3rd
floor

Nitrogensealing

PCV
2nd
floor

Pressure boundaries(simulated shutter for
carrying-in/out oflarge objects)

PCV exhaust

Dust injection

1st
Floor

PCV

PCV

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[3) (i) ① Technology for prediction of dust behaviors in PCV]
② Airflow analysis in containment vessel and entire R/B (3/3)✓✓

In cases where the air tightness of the building is ensured (absence of damage in the building), it is possible to create negative pressure in the building byperforming exhaust at 10,000 [m3/h] from the current operating floor. In addition, it is expected that a negative pressure of approximately 0.1 kPa can beachieved inside the PCV as well against the building.
When the air tightness of the building was lost (operating floor ceiling and building damaged), exhaust from the operating floor became meaningless and theevaluation was conducted by relocating the building exhaust to the 4th floor. However, the result was that negative pressure could not be created due to the
inflow of outside air from the operating floor side.
As countermeasures for the building, cross-sectional area of the stairs between the operating floor and 4F was reduced to 1/100 (0.09m2 x 3) which resulted increation of approximately 0.1 kPa negative pressure in the building. In addition, it is expected that a negative pressure of approximately 0.1 kPa can beachieved inside the PCV also against the building.

[Presence of damage in building][Absence of damage in building]
Stairs(9m2)

Equipment
carrying-in hatch(25 m2)

Flow rate boundary 4F
Exhaust at10,000[m3/h]

Operat
ing
floor

3F

4F
3F

Exhaustgas
Nitrogensealing

2F
1F

Air flow

Pressureboundary

Nitrogensealing

Stairs (9 m2) x 3
15.6 Nm3/h
It is assumed that all nitrogen sealed → 0.09 m2 x 3from the top of the RPV flows out fromthe bottom of the RPV to the PCV.

Damagedopening

Assuming D/W spherical shell
Based on the previous year’s study, thearea was set to 1. 0 E-4m2

Exhaustair flowfrom PCV

16.8 Nm3/h

Based on the previous year’s study, it is
the air flow required to achieve the PCV
negative pressure of 0.1 kPa when theoutside of the PCV is assumed to be theatmosphere

Damagedopening

Equipment
carrying-in
hatch (25 m2)

Flow rate boundary
Exhaust at 10,000
3F[m3/h]

Oper
ating
floor

2F

4F

1FExhaust
gas

3F
2F

Pressureboundary

1F
Damagedopening

AnalysisConditions

Air flow

Nitrogensealing

1700 Nm3/h
Based on the previous year's study, the airflow
required for the negative pressure and mitigation ofoxygen concentration is sealed in from the bottomof the RPV

Damag
ed
opening

Assuming D/W spherical shell
Based on the previous year’s study, the area was
set to 1. 0 E-2m2

Exhaustair flowfrom
PCV

2000 Nm3/h
Based on the previous year’s study, it is the airflow
required for achieving the PCV negative pressureof 0.1 kPa and for the mitigation of oxygen
concentration assuming that the outside of the
PCV is the atmosphere

Nitrogen sealing

PCV pressure (PR1s2982)
R/B 1F pressure (PR4s488)

Pressure (kPa)

Pressureboundary

Analysis
Conditions

Environment-Building Negative pressure approximately 1
kPa
Building-PCV Negative pressure approximately 0.1
kPa

Analysis duration (h)Pressure (kPa)✓No.240

PCV pressure (PR1s2982)
R/B 1F pressure (PR4s488)

Environment-Building Negative pressure approximately 0.1 kPa

Building-PCV Negative pressure approximately
0.1 kPa

Analysis duration (h)©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[3) (i) ① Technology for prediction of dust behaviors in PCV]
③ Evaluating the behavior of aerosol in R/B (1/6)

No.241

The analysis conditions including the R/B model were set for evaluating the aerosol behavior inside R/B. The behavior
of aerosols inside R/B was evaluated by assuming cases such as most severe PCV leakage, continuation of nitrogensealing and fuel debris cutting after the stoppage of PCV exhaust, etc., in an environment equivalent to Unit 3 where theairtightness of the building has been lost.
✓ Evaluating the behavior at the time of abnormal dust leakage in R/B
The following findings were obtained by conducting the R/B dust behavior evaluation:• It was confirmed that in areas where the gas flow rate was fast, the amount of dust deposition was small, andin areas where the flow rate was slow (stagnation), the amount of deposition tends to be large. It became
evident that the air flow rate is an important parameter.
• From the perspective of leak detection monitoring, it became clear that it may not be possible to measure theair dose rate correctly in areas with large amount of deposition after a certain period due to the backgroundeffect of the deposited FP.
✓ Sensitivity analysis for R/B dust behavior
As sensitivity analysis, the following findings were obtained by evaluating the impact of damaged opening position in
PCV, impact of response at the time of stoppage of PCV exhaust system and impact of the state of air-conditioning
in the building:• It is possible to prevent the spread of contamination by stopping dust cutting immediately after detecting anabnormal leakage.
• Even in cases where the R/B cannot be maintained in a negative pressure state, exhaust from the building
results in the prevention of dust leakage into the environment. As a result, the confinement function may besecured even when negative pressure inside the building is not attempted.
• However, if negative pressure is not attemted inside the building, the contamination on the leaked floor
spreads. In addition, as the deposition in the building increases, the dust transfer to the building exhaust
decreases.
• If the building exhaust is not performed, the gas leaking from the PCV will cause dust to leak into theenvironment. Ventilation of the building is indispensable to confine the dust.
©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[3) (i) ① Technology for prediction of dust behaviors in PCV]
③ Evaluating the behavior of aerosol in R/B (2/6)

No.242

If the exhauststops for somereasonNitrogeninjection
stoppedNONO
Cutting
stoppedYES• Cutting started: 1 hour later• Exhaust stopped: 11 hours later• Dust particle size: 1 μm

Exhaust stopped
after 0h, 1h, 2h

YESAmount of accumulated dust leakage [kg]✓ As a result of sensitivity analysis conducted in FY2018 to study the amount of dust leak from PCV, it was foundthat continuous nitrogen injection and continuous dust cutting while PCV exhaust was stopped was the most
severe condition . Therefore, the same conditions are used to evaluate the dust leakage inside the building.

2 hour difference
Stopped at the same time

Analysis duration [Hr]Aerosol injection
Damagedopening

Nitrogeninjection

Exhaustopening* From the results of technology development research
conducted in FY2018 to secure confinement function©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[3) (i) ① Technology for prediction of dust behaviors in PCV]
③ Evaluating the behavior of aerosol in R/B (3/6) ~ Mass distribution of aerosol ~[Base case analysis result (1st floor to 2nd floor)]
2nd floor of R/B

Amount of deposited dust

# central heightNortheast Stairs

Hatch forcarrying-in
large objects

Amount of deposited dust (kg)1.
0Amount of airborne dust

Northwest Stairs

m/s

No.2430.0Southeast Stairs

1st Floor of R/B (Leakage floor)m/s# 24 hours after dust leakage

Amount of airborne dust

Amount of depositeddust# central height

Hatch for carryingin large objects
Damaged opening

0.0Southwest area
Shutter for carrying large
objects in/out
Southeast Stairs

Amount of deposited dust (kg)Northeast
Stairs1.
0# 24 hours after dust leakage©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[3) (i) ① Technology for prediction of dust behaviors in PCV]
③ Evaluating the behavior of aerosol in R/B (4/6) ~ Mass distribution of aerosol ~[Base case analysis result (3rd floor to 4th floor)]m/s
1.
0 Building exhaust

Northwest Stairs (withAmount of deposited dust

Amount of airborne dust # central height

countermeasures)Northeast
Stairs(withcountermeasures)

0.0

Amount of deposited dust (kg)

4th floor of R/B

No.244Southeast Stairs(withcountermeasures)# 24 hours after dust leakage
3rd floor of R/B

Northwest Stairs

m/s
1. 0

Amount of deposited dust

# central heightNortheast
StairsHatch forcarrying-in
large
objects

Southeast Stairs

Amount of deposited dust (kg)0.0Amount of airborne dust

# 24 hours after dust leakage©International Research Institute for Nuclear Decommissioning7. Implementation items of this projectNo.245[3) (i) ① Technology for prediction of dust behaviors in PCV]
③ Evaluating the behavior of aerosol in R/B (5/6) ~Monitoring dose rate contribution~
Amount of airborne dust

Dose rate (Mev/s/m2)✓ From a monitoring point of view, inareas with a high amount of dust
deposition, the air dose rate may not
be measured properly after a certain
period of time.

Amount of deposited dust (kg)✓ When doing a simple evaluation on thedose rate contribution when FP issuspended in a cube with a side of 3mand when all of it is deposited on thefloor, the contribution of suspended FP
was about 3 times than that of deposited
FP.
✓ In areas where the deposited amount is
the most severe, the dose rate
contribution of floating dust is dominant
up to 5.5 hours, and after that, thecontribution by the deposited dust is high.

Amount of deposited dust

Floating
dust

Analysis duration (h)©International Research Institute for Nuclear Decommissioning7. Implementation items of this projectNo.246[3) (i) ① Technology for prediction of dust behaviors in PCV]
③ Evaluating the behavior of aerosol in R/B (6/6) ~Sensitivity analysis using building countermeasures~✓

Amount of airborne dust on 1st floor

Negative pressurecountermeasures
implemented

Negative pressurecountermeasures
not implemented

0.09 [m2]9.0 [m2]

Negative pressure

0.1 [kPa-diff]

0.0 [kPa-diff]

Intake of outside
air through theoperating floor

2.0 [kg/s](Gas)

2.9 [kg/s](Gas)

5.2 [kg]

2.1 [kg]

4F

0.3 [kg]

0.2 [kg]

3F

0.4 [kg]

0.5 [kg]

2F

0.9 [kg]

2. 4 [kg]1F

2. 1 [kg]3. 8 [kg]

0.9 [kg/s](Gas)

0.04 [kg/s](Gas)

Staircase area

Amount of dust
exhaust from 4F

Dust mass#

Intake of outside
air through the
shutter for large
objects

Amount of deposited dust on 1st floor# Sum of the floating and deposited amounts

Amount of deposited dust (kg)✓Dust may not be released to the environment even if negative pressure countermeasures are not implemented.
When negative pressure countermeasures are not implemented, dust inside the building increases and contamination on1st floor and 2nd floor of
the building spreads.
By implementing negative pressure countermeasures, it may be possible to reduce the amount of dust inside the building and prevent contamination.
However, the exhaust dust from the building increases.
Furthermore, if the building exhaust is not performed, dust is released into the environment through the operating floor and the shutter for
carrying large objects in/out.

Amount of deposited dust (kg)✓✓# 24 hours after dust leakage
Negative pressurecountermeasures not implemented(0 [kPa - diff])

Negative pressurecountermeasures implemented(0.1 [kPa - diff])©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[3) (i) ① Technology for prediction of dust behaviors in PCV]
④ Effect of heat input on airflows (1/9)

No.247・ In cases where the dust and the area near the floor become hot due to the decay heat from fuel debris or due to the
cutting of fuel debris etc., the ascending air flow will cause more dust to fly up, possibly increasing the amount ofleakage and the amount of dust migrating to the exhaust.
・ When a temperature gradient occurs inside PCV and multiple damaged openings are present in the PCV, thechimney effect will cause the airflow to enter and exit through the damaged opening. This is believed to have impact
on the dust behavior.
✓ Implementation items for this year [I]- Study including the method for evaluating the effect of heat input on airflows due to decay heat of fuel debris or
cutting of fuel debris → As typical cutting methods, plasma cutting that has a large thermal effect, and laser cutting that has a large
amount of assist gas, were selected, and the amount of dust generated and the amount of heat input (heatdissipation at the time of cutting and decay heat of the cutting dust) were evaluated based on the literature.
Also, a model adding heat input to airflow was built.
→ Through analysis, the increase in airflow temperature due to heat input was confirmed by taking into accountthe increase in exhaust airflow due to the blowout of assist gas and heat input.
✓ Implementation items for this year [II]- Understanding the temperature gradient due to heat input and understanding the relationship between the positionof damaged openings and the chimney effect based on it → A sub-volume was added to the GOTHIC model to simulate the gaps between the PCV and building concrete.
Two damaged opening locations with different heights and flow conditions were set.
→ When the negative pressure inside the PCV was set from -200 Pa to -300 Pa, there was no leak due to thechimney effect in the gaps between the inside of the PCV and the outside of the PCV. Heat dissipation at thetime of cutting is the dominant factor for increasing the pressure inside the PCV.
©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[3) (i) ① Technology for prediction of dust behaviors in PCV]
④ Effect of heat input on airflows (2/9)✓✓

No.248

As cutting methods, plasma cutting that has a large thermal effect, and laser cutting that has a large amount of assist gas, were selected, and theamount of dust generated and the amount of heat input (heat dissipation at the time of cutting and decay heat of the cutting dust) were evaluatedbased on the literature.
A model with addition of heat input was built. The heat loss due to exhaust was added to the flow of plasma gas and assist gas at the fuel debris
cutting position.

Study on the proposed parameters for evaluating the effect of heat input
Analysis Conditions

Plasma cutting

Laser cutting

Cutting current / laser output

200 kW (600 A)

6 kW

50%, 25%

50%, 25%

Ar, N2N2

19.5 L/min

175 L/min

Loss of exhaust heat *3

7%

7%

Amount of dust generated(underwater cutting) *2

3. 7 x 10-6 kg/s

1. 8 x 10-6 kg/s

Size of dust particles *40.05 x 10-6 m

1. 0 x 10-6 m

Operating rate *1
Plasma gas / assist gas
Gas flow rate *2

Dust

Assist gas

Heating part

Dust, gas and amount of heat added*1：It is the percentage of the net cutting time (50%, quoting the estimated value) excluding the non-heating time such as pulling up and moving the torch, from theworking hours.
*2：Value when the operating rate is 50%.
*3： The evaluated value of the exhaust heat loss ratio of plasma gas in the casting field has been used. In underwater cutting, it is assumed
that except for the amount of heat that accompanies the exhaust, the rest of the heat gets transitioned to the water that has large heat capacity and subcooling.
*4：The dust particle size distribution during plasma cutting in JPDR dismantling field test is quoted. The dust particle size distribution during SUS304 laser cutting isquoted.
©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (i) ① Technology for prediction of dust behaviors in PCV]
④ Effect of heat input on airflows (3/9)

No.249✓ To analyze the chimney effect, a sub-volume simulating the gaps between the PCV and the building concrete isadded outside the PCV in the GOTHIC model. For the openings simulating the damaged openings in the upper
part, the positions where the height and flow conditions are different (A1: Top head flange, or A2: Top of tophead) were studied.
✓ The open part near the PCV top head is connected with the building.*1

Simulated opening A2

Building opening

Simulated opening A1

Simulated opening A2

Exhaust nozzle *2(CCS Sparger)

30 m

Simulated opening A1N2 injection

Pedestal

30 m
Simulated opening B

Supply ofSolid particles(dust)

Supply ofplasma gas
or assist gas

Control Volume Settings

Simulated opening B

Sub-volume settings for gaps between PCV and building*1： The plan was to set up a model that considers gas replacement withthe building through pipe opening holes in the case of a leak andGaps between
PCV and building increase in temperature in the gaps. However, the results of this analysis
confirmed that there was no leakage, so it was decided that it need notconcretebe considered.

* 2: In this analysis, the current exhaust system is simulated.
©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[3) (i) ① Technology for prediction of dust behaviors in PCV]
④ Effect of heat input on airflows (4/9)✓✓

No.250The chimney effect and dust diffusion behavior were evaluated by performing an analysis simulating the fuel debris cutting state.
Based on the evaluation [5] of this project, the exhaust flow rate in the analysis was set so that the negative pressure in the PCV became -200 Pa to 300 Pa.

Exhaust flow rate for each cutting method

Position of the opening
Cutting operating rate

Exhaust flow rate

Top head flange50%

3.96 kg/min(0.066 kg/s)

Maximum pressure during
cutting[Negative pressure]101.11 kPa[-215 Pa]

Top head flange25%

3.38 kg/min(0.056 kg/s)

101.05 kPa[-275 Pa]3

Top of top head50%

3.96 kg/min(0.066 kg/s)

101.11 kPa[-215 Pa]4

Top head flange50%

2.88 kg/min(0.048 kg/s)

101.11 kPa[-215 Pa]

Top head flange25%

2.88 kg/min(0.048 kg/s)

101.05 kPa[-275 Pa]Top of top head50%

2.88 kg/min(0.048 kg/s)

101.11 kPa[-215 Pa]

Top head flange50%

11.8 kg/min(0.196 kg/s)

101.04 kPa[-285 Pa]

Case No.

Cutting method

1
2

5

Plasma(Underwater)

Laser(Underwater)

6
7

✓
✓
✓

Plasma(in air)

Though the plasma gas flow rate (19.5 L/min, at an operating rate of 50%) used for underwater plasma cutting, where a large
amount of heat is dissipated, was one order less than the assist gas for laser cutting (175 L/min), the exhaust flow rate formaintaining negative pressure increased by approximately 38%.
For plasma cutting performed in air (13. 5 L/min) where the amount of gas generated was relatively low, the exhaust flow rateincreased compared to the underwater cutting.
It is considered that heat dissipated at the time of cutting is the dominant factor for the increase in pressure inside PCV.
©International Research Institute for Nuclear Decommissioning7. Implementation items of this project[3) (i) ① Technology for prediction of dust behaviors in PCV]
④ Effect of heat input on airflows (5/9)
Behavior of PCV internal pressure and temperature while cutting(Comparison of cutting methods)

101

Set negative pressure range-300 Pa

100
Cutting duration(Operating rate 50%)

99
60

40
30

Cutting duration(Operating rate 50%)

0

5

10

15

20

Time ( x 103 s)
Underwater plasma cutting (Flange opening)✓✓✓ The analysis of Case 1 took a total of 5 hours, including
2 hours for stabilization of negative pressure for computation, 1hour for actual cutting and 2 hours for cooling down.

102

Set negative pressure range-200 Pa

101-300 Pa

100
Cutting duration(Operating rate 50%)

99

Pedestal region
Spherical shell
Cylindrical partTop head part50

20

Pressure (kPa)-200 Pa

Average Temperature (℃)

Average Temperature (℃)

Pressure (kPa)102

No.25160

Pedestal region
Spherical shell
Cylindrical partTop head part50
40

Cutting duration(Operating rate 50%)

30
20

0

5

10

15

20

Time ( x 103 s)
Underwater laser cutting (Flange opening)

As there is less change in pressure and increase in temperature in underwater laser cutting, the chimney effect was conservatively evaluated using
underwater plasma cutting.
The pressure distribution graph did not show any local pressure deviations due to nitrogen sealing, exhaust, in leak from the opening, etc.
©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (i) ① Technology for prediction of dust behaviors in PCV] No.252④ Effect of heat input on airflows (6/9)

Pedestal region
Spherical shell
Cylindrical partTop head part

50

40
30
Cutting duration(Operating rate 50%)

20
0.050

Total flow rate (A1 + B)

Simulated opening at the top A1

0.025

Spherical shell B

Out leak0-0.025In leak-0.050

Average Temperature (℃)

60

Passing flow rate (kg/s)

Passing flow rate (kg/s)

Average Temperature (℃)

Behavior of temperature and flow rate through openings in various parts of PCV during plasma cutting(Comparison between underwater cutting and cutting in air)

Cutting duration(Operating rate 50%)

0

5

10

15

20

Time ( x 103 s)
Underwater plasma cutting (Flange opening)
✓
✓
✓

200

Pedestal region
Spherical shell
Cylindrical partTop head part

150

100
50
0
0.05

Cutting duration(Operating rate 50%)0Total flow rate (A1 + B)
Out leak
Simulated opening at the top A1
Spherical shell B-0.05

In leak

-0.10
-0.15
0

Cutting duration(Operating rate 50%)

5

10

15

20

Time ( x 103 s)
Plasma cutting in air (Flange opening)

In underwater plasma cutting, there were temperature differences in various parts of PCV depending on the heat input, but it did not lead to out leak.
In addition, there was almost no difference due to the position of the opening.
During plasma cutting in air, the temperature of various parts of PCV increased significantly compared to underwater cutting, but there was no out
leak.
©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (i) ① Technology for prediction of dust behaviors in PCV]
④ Effect of heat input on airflows (7/9)

No.253Temperature distribution inside PCV during the completion of cutting
150

100

50

0
Temperature (℃)
Case 1 (10800 s)
・ Underwater plasma cutting・ Operating rate 50%
・Top head flange opening✓✓

Case 4 (10800 s)
・ Underwater laser cutting・Operating rate 50%
・Top head flange opening

Case 7 (10800 s)
・ Plasma cutting in air・Operating rate 50%
・Top head flange opening

A height-wise change in temperature was observed due to the difference in heat flow in each cutting method. The pressure
distribution was also evaluated, and height-wise pressure changes were observed due to the differences in density.
The airflows were compared for plasma cutting underwater and in air where the temperature difference was relatively large.
©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (i) ① Technology for prediction of dust behaviors in PCV]No.254④ Effect of heat input on airflows (8/9)
Velocity distribution inside PCV at the time of cutting (position of openings and cutting conditions)
Top opening↓
Top headpart

Cylindricalpart

Spherical
shell

Flange opening

Flange opening

Pedestal region

Case 3 (9600 s)
Case 1 (9600 s)
・Underwater plasma cutting Gap between・ Underwater plasma cutting
PCV and ・ Operating rate 50%
・Operating rate 50%
building ・ Top of top head opening
・Top head flange opening

Underwater plasma cutting

Case 7 (9600 s)
・ Plasma cutting in air
・ Operating rate 50%
・ Top head flange opening

Plasma cutting in air✓ In-leak flow toward the inside of the PCV was confirmed at the flange opening. For the top opening, the in-leakflow was confirmed from the velocity vector at the gap.
✓ During plasma cutting in air, convection in the PCV cylindrical part and spherical shell became significant, butthere was no occurrence of out leak due to the chimney effect.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (i) ① Technology for prediction of dust behaviors in PCV]No.255④ Effect of heat input on airflows (9/9)
Dust concentration distribution inside PCV (Underwater plasma cutting, Operating rate 50%, Flange opening)
10-9

Exhaustnozzle10-10

Nitrogen sealing

Cutting position

10-11

10-12
t = 7200 s (Cutting started)✓✓✓t = 10800 s (Cutting completed)t = 18000 s (2hrs after cutting)

Dust density(kg/m3)The dust concentration in the pedestal region was high when cutting was completed. Two hours after the completion of cutting,though the concentration decreased to some extent, dust was distributed throughout the PCV.
In addition, there was no significant difference in the top opening conditions compared to flange opening.
In underwater plasma cutting, the reduction of dust concentration took some time. When managing the dust concentration in theexhaust, it may be necessary to increase the exhaust flow rate.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (i) ① Technology for prediction of dust behaviors in PCV]No.256⑤ Summary (1/4)◆ The following four items of technology development were carried out to evaluate the behavior of dust
generated due to fuel debris retrieval. The results of each implementation item are summarized below.
① Advancement of realistic dust behavior evaluation method
・ The validity of the aerosol model in the GOTHIC code and its applicability to 1F were investigated, and it
was concluded that the aerosol behavior could be evaluated by considering only gravitational sedimentation.
② Airflow analysis in containment vessel and entire R/B
・ The pressure behavior analysis was conducted by creating a new R/B model and by using the presence orabsence of damage in the building as a parameter. Even under conditions assuming building damage, it wasestimated that negative pressure (air tightness ensured) can be achieved by taking measures such as
narrowing the exhaust position and flow path.
③ Evaluating the behavior of aerosol in R/B・ By evaluating the aerosol behavior using the R/B model, the findings for identifying appropriate monitoring
positions for detecting abnormal leaks and deposition tendency of aerosols, were understood.
・ By performing sensitivity analysis changing the state of air-conditioning in the building, it was estimated thatnegative pressure inside the building may not always be necessary to secure the confinement function, but
exhaust inside the building is mandatory.

④ Effect of heat input on airflows
・ A model adding heat input to the airflow was built by studying the amount of dust generated and heat input
such as heat dissipation during decay heat and fuel debris cutting.
・ The effects of temperature gradient and damaged opening position were evaluated in addition tounderstanding the required exhaust flow rate for each cutting method through analysis. With this, it was
confirmed that the chimney effect does not occur at the assumed negative pressure.
©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (i) ① Technology for prediction of dust behaviors in PCV]
No.257⑤ Summary (2/4)
The results of the following two items that were specified as development items are as follows:① Requirement specifications for R/B air tightness and gas management system to secure theconfinement function in the event of an abnormality⇒ It can be stated that it is necessary to reduce the flow path area between the operating floor
and 4F to 1/100 of the current level (approximately 0.09 [m2]), with the prerequisite ofventilation rate due to exhaust set to 10,000 [m3/h] from the 4th floor of the building. In otherwords, a negative pressure countermeasure plan that narrows down the flow path area
between the operating floor and 4F can be one of the requirement specifications for the air
tightness and the gas management system of the R/B.

② Identification of effective monitoring positions from the viewpoint of reducing the exposure to the
public and workers while monitoring within R/B, in the event of an abnormal leak from the PCV⇒ Appropriate monitoring positions will be the positions where the flow rate of the gas leaking
from PCV is high. The position where the flow rate of the gas leaking from PCV is fast is near
the damaged opening in PCV. However, in situations where the location of damage is
unknown, the appropriate positions may be the staircase and areas near the hatch, which is
the main flow path of leaked gas with a continuous fast flow rate between the floors.
⇒ In order to select an appropriate monitoring technology, measures that do not pick up thebackground dose from the deposited FP (for example, installing a collimator) are believed tobe effective as it is considered necessary to measure the amount of airborne dust from theviewpoint of leak detection.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (i) ① Technology for prediction of dust behaviors in PCV]No.258⑤ Summary (3/4)◆ Dust behavior evaluation method was created based on the physical behavior
of aerosols considering environment of Fukushima Daiichi◆ By using the results obtained here, various studies have become possible withthe following engineering performed for fuel debris retrieval:

•
•
•
•
•
•

Cutting methods and PCV environment during cutting
Image training for cutting work sequence
Optimization of PCV gas management system
R/B countermeasures and optimization of ventilation system for securing
the confinement function
System operating conditions that can prevent out leaks and safety margin
evaluation during transient
Study on leakage monitoring during abnormality and mitigation measures◆ For individual engineering, conditions that appropriately reflect the results offield surveys and actual tests were set using the analysis methods developedin this PJ. Also, it is necessary to ensure the reliability of analysis byexpanding and modifying the model as necessary and comparing it with thetest results.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (i) ① Technology for prediction of dust behaviors in PCV]No.259⑤ Summary (4/4)◆ The remaining issues include the following:➢ Issues related to improving the analysis accuracy• Confirming the polydisperse behavior considering particle size distribution• Evaluating the effect of local heat flow behavior below the cell size (CFD,etc.)⇒ Development of an evaluation method that can appropriately reflect
aerosol conditions (particle size distribution, composition, shape factor)
and their generation conditions (generated amount, generation rate, etc.)
in accordance with the processing method carried out through subsidy
projects, based on the results of Dust Collection Project (PJ) and Fuel
Debris Characterization PJ.
➢ Challenges for evaluating the actual equipment• Understanding the state of dust adhesion before the start of cutting bycollaborating with the Internal Investigation PJ• Creating the model of actual PCV, R/B in accordance with the retrieval
method• Evaluating the physical properties and generated amount of dust bycollaborating with the Fuel Debris Characterization PJ• Specifying actual aerosol conditions in accordance with the cutting method,etc.
⇒ Implemented based on the progress of other PJs and engineering.
©International Research Institute for Nuclear Decommissioning7. Implementation items in this projectNo.2603) Development of Technology for Ensuring Safety while Retrieving Fuel Debris(i) Development of element technology for confinement functions② Technology for ensuring confinement functions in connection parts

Regarding the new large-sized equipment installed in R/B, such as work cells forfuel debris retrieval, as technology development for securing the confinement
function of connection parts with existing structures such as PCV, the structure,methods, inspection, maintenance of sealing materials, etc., of the connection parts
were studied, and necessary element tests to confirm the feasibility of thetechnology were conducted.
The important development study items include the following, and issues shall beidentified and organized by conducting element tests, as necessary:a. Method for connecting PCV and access tunnel.
・The organization of basic requirements for access tunnel connection parts isunder progress.
・ Work steps to complete access tunnel connection considering the procedure
for removing the shielding plug in front of the equipment hatch, were studied.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [ 3) (i) ② Technology for ensuringconfinement functions in connection parts ]No.261⚫ Purpose of Development➢ To study the connection procedures with PCV and the sealing methods of the access tunnels forwhich the feasibility of the cantilever delivery method has been confirmed.
⚫ Issues to be solved➢ Connection structure with PCV (sealing method) and creation of installation procedures.
⚫ Approaches to development➢ Plan the work procedures starting from the removal of shielding structure in front of equipment
hatch to the connection of access tunnel to PCV, considering remote operations.
➢ Study on the connecting structures of PCV and sleeve, sleeve and access tunnel, and check thefeasibility using element tests.
➢ Review the shielding function and conduct conceptual study on weight reduction.
⚫ Expected outcome➢ Work procedures starting from the removal of shielding structure in front of equipment hatch tothe connection of access tunnel to PCV, considering remote operation.
➢ Confirmation results on the feasibility for connecting structures of PCV and sleeve, sleeve andaccess tunnel.
➢ Results on the shielding function review and conceptual study on weight reduction.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (i) ② Technology for ensuringconfinement functions in connection parts]

No.262◼ Preconditions for access tunnel installation✓ Completion of yard maintenance, underground structure measures andaboveground structure removal for installing the access tunnel.
✓ The possibility to build an expanded building that connects to PCV through accesstunnel.
✓ The possibility of creating an opening on the outer wall of R/B to install the accesstunnel. Furthermore, the access tunnel will be laid remotely by delivery fromoutside the R/B.
Expanded

増設建屋
building

PCV
Access tunnel
アクセストンネル

Image showing the connection of PCV and expanded building©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (i) ② Technology for ensuringconfinement functions in connection parts]◼ Study Flow

No.263：This report

Outcomes and issues of the subsidy
project in the previous term
• Conduct element tests for delivering the access tunnel. Confirm the feasibility of a remotedelivering operation.

Study on the connection methods between
access tunnel and PCV• Organize the basic requirement specifications for connection parts between access tunnel and PCV.
• For connection with PCV, conduct a detailed study on the work steps to complete access tunnel
connection considering the procedure for removing the shielding structure in front of the equipment
hatch.
• Detailed study on the access tunnel sleeve structure.
• Study the proposed connection structures between access tunnel sleeve and equipment hatch.

Element test plan for the connection between
access tunnel sleeve and equipment hatch

Connection element tests

Scope of implementation in
the subsidy projectPossibility of delivering the access tunnel with the actual weight

Survey results on site conditions

Actual construction plan
©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (i) ② Technology for ensuringconfinement functions in connection parts]

No.264◼ Development Schedule
Study Items
Major Milestones

FY 2019
４

5

6

7

8

9

10

11

Interim Report▼

FY 2020
12

1

2

3

4

Interim Report
▼

5

6

7

8

9

10

11

Interim Report
▼

12

1

2

3

Final
Report
▼1. Conceptual study/
selection ofconnection method
2. Test plan

3. Testing
preparations /Manufacturing oftest equipment
4. Element tests
5. Summary

Remarks©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [ 3) (i) ② Technology for ensuringconfinement functions in connection parts ]◼ Basic requirement specifications for access tunnel connection parts

No.265

Expanded
増設建屋
building

Regarding the access tunnel installation method, element tests on the remote delivery method
were conducted and the prospect on feasibility was obtained in previous fiscal years.
✓ In this subsidy project (FY2019 - 2020), the connection method with the PCV will be studied.
✓ Along with the study on connection parts, organization on the basic requirements for accesstunnel connection parts is in progress.
✓ After studying the response policy based on the basic requirements in the future,
implementation of the required element tests will be planned.
( Element tests are scheduled to be conducted in FY2020)✓

PCV
Access tunnel
アクセストンネル

Concept of access tunnel

Table. Basic requirements for the connection parts between PCV and access tunnel [DRAFT]
Items
Required functionality

Specifications
Confinement of radioactive materials
Amount of leakage is 〇〇 or less when the differential pressure is -100 Pa

Requirement specification
conditions for connectionparts

The connection parts must be sound when the designed differential pressure is 〇〇Pa.

Basic requirements
Environmental
conditions

Reduction in exposure ofworkers

Connection work and maintenance work must be completely remotely operated.

Appearance of confirmation tests
on the feasibility of remotedelivery (results of previous
years)

Design Life

50 years

PCV side
PCV側

Maximum horizontal
displacement

Can withstand seismic motion (Maximum horizontal displacement: ± 〇〇 mm)

Floor load capacity

Shall satisfy the condition of 750 ton/m 2 (load capacity value of the shielding pluginstallation part).

Effective diameter of the
opening□ 1500 mm

Dose rate(R/B 1st Floor) 5 ~ 10 mSv/h(Near the outer wall of the PCV shell) 10 ~ 100 Sv/h

Temperature-7 ~ 40 ℃

Humidity≤ 100%The negative pressure (differential pressure during normal operation) at the primaryboundary shall be maintained during maintenance.

アクセストンネル
Access tunnel

AT側

AT sideSleeve
スリーブ

Image of access tunnel sleeve©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (i) ② Technology for ensuringconfinement functions in connection parts]

No.266◼ Work steps for access tunnel construction✓ Up to the previous fiscal year, the steps related to delivering the access tunnels from outside the R/B → inside
the R/B were mainly studied.
✓ This time, for the connection with the PCV, the work steps up to the completion of access tunnel connection
considering the procedure for removing the shielding structure in front of equipment hatch will be studied.

[Diagram on access tunnel installation steps (1/2)]
Step.1

Step.2

Step.3

Step.4

Simulated AT
ATｽﾘｰﾌﾞ模擬体

sleeveTemporary
仮設遮へい
shieldAT
ATｽﾘｰﾌﾞsleeveEquipment hatch
機器ﾊｯﾁ遮へい体

Water
shielding
水遮へい注水coolant

シール溶接
Welded seal

shielding structure

ホース
Hose

•
•
•
•

Simulated AT sleeve: Equipped withrust remover

Carry-in, temporary storage of accesstunnel sleeve (hereinafter referred to as
AT sleeve)
Carry-in, temporary storage ofsimulated AT sleeve
Installation of shield when pulling out
the equipment hatch shielding structure
Pulling out and removal of equipment
hatch shielding structure

•
•
•
•
•

Performing 3D scanInsertion and installation of simulated
AT sleeve
Pretreatment before AT seal welding(rust removal)
Removal of simulated AT sleeve
Installation of AT sleeve

AT sleeve: Equipped with welding equipment, replaced with a water
shielding tank

•
•
•
•

Performing 3D scanInsertion and installation of AT sleeve
Welding of AT sleeve seal
Welding inspection(Inspection items will be studied
separately)•

Injecting water shielding coolant inside
AT sleeve• Dismantling and removal of temporary
shield• Dismantling and removal of equipment
hatch shielding structure(Carrying out from the north opening)©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (i) ② Technology for ensuringconfinement functions in connection parts]

No.267◼ Work steps for access tunnel construction[Diagram on access tunnel installation steps (2/2)]
Step.5

Step.6

Step.7

Step.8

Balancing
ﾊﾞﾗﾝｽｳｴｲﾄ
weightAccesstunnelAccess

ｱｸｾｽﾄﾝﾈﾙ

ｱｸｾｽﾄﾝﾈﾙ
tunnelShieldingwater
水遮へい水槽tank•••

Installation of temporary building
AT installation
AT delivery

Expanded
増設建屋
building

シール溶接
Weldedseal

Shielding water
水遮へい水槽

tank

•
•
•

Water shielding drainage inside AT
sleeveRemoval and carry-out of shielding
water tank inside AT sleeve
Connecting AT

ATｽﾘｰﾌﾞAT
sleeve•Dismantling the temporary building(partial)

•

•
•
•

Construction of expanded building(completion of installation of heating,ventilating and air conditioning systemand other equipment)
AT edge cap removal
Welding of sleeve with AT
Opening equipment hatch©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (i) ② Technology for ensuring confinement
functions in connection parts]

Horizontal displacement *◼ Study on AT sleeve connection methods[Setting the amount of displacement at time of earthquake]
It is necessary to secure a displacement of ± 12.5 mm in the horizontal direction.

Evaluation
point [mm]Horizontal
displaceme
nt [mm]Top end of
hatch

12905

13490

12. 5

Centre of
hatch

11260

11180

9

Bottom endof hatch

9675

9760

7

Vertical displacement *Position

The inter-plane distance at the application part is approximately 350 mm.
As it is difficult to use existing technologies such as bellows structure, study new
structures.
12. 5mm

Level fromOP [mm]Position➢ AT sleeve structure✓ Conduct a detailed study on the displacement absorbing mechanism, which
is a component of the AT sleeve.

Flexiblerod

No.268Top end of
hatch

Centerof AT
ATスリーブ軸中心
sleeve axis

Level fromOP [mm]Evaluation
point [mm]Verticaldisplaceme
nt [mm]

12905

13490

0.12

Displacement
absorbing mechanism
Access tunnel bodyBSW

フローティング
Floating
フランジ

Equipment
hatchflange12. 5mm

アクセストンネル側
Accesstunnel side

Proposed structure for displacement
absorbing mechanism which isunder study

PCV側
PCV side

Details of displacement absorbing
structure

AT Sleeve

Example of the structure of connection parts
between the access tunnel and PCV

In the next fiscal year, the study will continue on the detailed structure in combination with the AT sleeve.
©International Research Institute for Nuclear Decommissioning* Set based on the report details of Subsidy Project of Decommissioning and Contaminated Water Management in FY 2015 Supplementary Budgets “Development of Seismic Resistance and Impact
Assessment Method on PCV and RPV”

7. Implementation items of this project [3) (i) ② Technology for ensuringconfinement functions in connection parts]

No.269◼ Study on the connection method of AT sleeve➢ Structures connecting equipment hatch and AT sleeve✓ Multiple welded connecting structures were studied and compared
Case 1Case 2•

Confinement ofcontamination inside theAT sleeve

Partial removal of scaffolding inside the equipment hatch
AT sleeve is welded to the end surface of the equipment
hatch shell.
The lower side of the AT sleeve is welded to the inner surfaceof the equipment hatch shell via the scaffolding seal cover.

AT sleeve welding position

AT sleeve welding position

AT sleeve welding position••

Case 3•••(The scaffolding inside the equipment hatch need not be
removed)AT sleeve is welded to the end surface of the equipment
hatch shell.
Place the plate on the scaffold and weld the lower side of the
AT sleeve, the inner surface of the equipment hatch and theequipment hatch door.

Extended sleeve

•
•
•
•

Partial removal of scaffolding inside the equipment hatch
A plate that aligns with the AT sleeve is welded on the lower
side inside the equipment hatch.
Extended sleeve is added to AT sleeve.
The inner surface of the equipment hatch and the plate thataligns with the AT sleeve are welded.

Workability◎
All connection parts are welded〇
No seal on the door opening / closing part under the plate◎
All connection parts are welded

Number of
installation steps△
Scaffolding removal◎
No scaffolding removal△
Scaffolding removal

Number of parts
other than AT
sleeve○Presence of scaffolding seal cover drive mechanism◎
Plate only◎
Plate only

Groove alignment(Inside theequipment hatch)○
It is necessary to study whether the scaffolding seal cover works
well with the equipment hatch.

◎
Assume that the scaffolding plate is just placed△
Installation accuracy of the alignment plate is required(Level / distance from AT sleeve)

Welding
workability○
Presence of discontinuous welded structures○
Relatively easy◎
Peripheral welding only

Overall evaluation

〇

〇

△

Cases 1 and 2 will be the main proposals, and studies will be carried out on them in future.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (i) ② Technology for ensuringconfinement functions in connection parts]

No.270◼ Study on the connection methods of AT sleeve➢ Structures connecting the equipment hatch and AT sleeve

Welding position of AT sleeve to equipment hatch (Case 1)

Welding position of AT sleeve to equipment hatch (Case 2)

After deciding the structure of the AT sleeve early next fical year, an element test plan
for welded connections will be formulated and conducted in the next fiscal year.
©International Research Institute for Nuclear Decommissioning7. Implementation items in this projectNo.271

3) Development of Technology for Ensuring Safety while Retrieving Fuel Debris(ii) Development of element technology for prevention and monitoring of criticality⚫
Current
scope

Overview of technology development for criticality safety control② Criticality approachmonitoring① Criticality
prevention

・ Monitoring the approach to
criticality(Near the retrieval work area)
・Response before criticality

a. Controlling andmanaging work procedures
so that they do not lead to
criticality

Neutrondetector③ Detection ofcriticality

・ Prompt detection ofcriticality signs(entire containment
vessel)Gas radiation
monitor

Emergency
boric acidinjectionb. Addition of neutron absorber① Soluble neutron absorber (boric acid solution)② Insoluble neutron absorber④ Mitigation of
impact・ Stopping immediately if
criticality is reachedB4C Sintered Glass material Gd2O3
particles
metal material with B/Gd

Water glass /Gd2O3 granulated
powder©International Research Institute for Nuclear Decommissioning7. Implementation items in this projectNo.272

3) Development of Technology for Ensuring Safety while Retrieving Fuel Debris(ii) Development of element technology for prevention and monitoring of criticality① Technological development of criticality monitoring method
Development results till date and implementation details for this fiscal year
Final Goal

Creation of
subcriticality
measurement
technology

Goal to be achieved before
actual application① Creation of measurement
methodSelection of a method that combines the reactor noise
method and the neutron source multiplication method(completed)
Study on detector sensitivity and placement to reduce
measurement error② Formulation of system
specifications

Formulation of system specifications for detectors,
measurement circuits, etc. (completed)③ Specifications for neutron
detector

Formulation of specifications for neutron detector andselection of detectors for prototyping④ System design / prototyping

Prototyping of test system consisting of neutron detector,
measurement circuit and PC for analysis⑤ Confirmation on the feasibilityof subcriticality measurement

Incorporation into
fuel debris
retrieval system

Achievement status

Confirmation under homogeneous fuel debris simulated
conditions (KUCA test # 1 / # 2)
Confirmation under large-scale fuel debris simulated
conditions (KUCA test # 3)

Details for this fiscal year

Study on the specifications of alternative
neutron detector candidates (RosRAO /IPL) for size and weight reduction

Tests for confirming the conditions using
non-homogeneous fuel debris / insoluble
neutron absorber (KUCA test # 4)

① Confirmation of theapplicability of neutron
detectors to the field

Organization of specification items of the neutron detector
unit for transfer with robot arm
Formulation of cable handling concept

Study on the structure of neutron detectorunit② Measures againstelectromagnetic noise

Evaluation of simulated noise effect, and organization ofcountermeasures

Study on measures against noise③ Study on the criticality
approach management
procedures

Evaluation on the time required for measurement

Procedures for monitoring the approach to
criticality at each step of the debris retrieval
method©International Research Institute for Nuclear Decommissioning7. Implementation items in this projectNo.273

3) Development of Technology for Ensuring Safety while Retrieving Fuel Debris(ii) Development of element technology for prevention and monitoring of criticality① Technological development of criticality monitoring method

To prevent the occurrence of criticality due to fuel debris retrieval operation, careful retrieval operation along with
monitoring is required. Therefore, the feasibility of subcriticality measurement shall be confirmed by confirming theaccuracy of subcriticality measurement in a large and complicated fuel distribution system simulating the state of
the Fukushima Daiichi Nuclear Power Station. Also, based on the results of the related “Project of Development ofFundamental Technology for Retrieving Fuel Debris/ Internal Structures (development of small-size neutron
detectors)”, the applicability of neutron detectors to the site shall be confirmed. At the same time, the procedures formonitoring the approach to criticality shall be formulated, management procedures including the methods for setting
the management parameters, etc., shall be created and the formulation of a demonstration plan for on-site
application shall be studied in cooperation with the engineering department of TEPCO .
The following are included as important development study items, and issues shall be identified and organized by
performing element tests, as necessary:a. Confirmation of the feasibility of subcriticality measurement
・ A subcriticality measurement test was conducted at the Kyoto University Criticality Assembly (KUCA).

b. Confirmation of the applicability of neutron detectors to the field・ Neutron detectors found to have good prospects were selected from the perspective of subcriticality
measurement.
c. Study on the criticality approach management procedures・ The procedures for monitoring the approach to criticality at each step of the debris retrieval method wereorganized.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (ii) ① Technological development of criticality
No.274
monitoring method]⚫

Development Schedule

Study Items

FY 2019
４

5

6

7

8

10

11

12

1

2

Interim Report▼

Major Milestones
a. Confirmation of thefeasibility of
subcriticality
measurement

9

FY 2020

Plan

3

4

5

6

7

8

9

Interim Report
▼

10

11

12

1

2

3

Final
Report▼Interim Report▼

Subcriticality measurement
method confirmation test (KUCA)

Preparations

Evaluation

Service permit procedures
Prototype specifications

b. Applicability ofneutron absorbers to
the field

Preparations / Prototype
specifications

Equipment manufacturing (B10 Corona counter)

Equipment design (SiC typesemiconductor detector)

Transport
Incorporation into
testing device
Equipment design and manufacturing
Study of testing system for detector
evaluation

Evaluation
Detector evaluation
test (KUCA)

Planning and designing of structure to
be installed on fuel debris

Evaluation

Manufacturing of equipment
Environmental resistance testc. Procedures formonitoring theapproach to
criticality

Creation of measurement methods

Selection of detector candidates

Summary

Remarks©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (ii) ① Technological development of criticality
No.275monitoring method]a. Confirmation of the feasibility of subcriticality measurement⚫ Purpose of Development➢ To develop methods for measuring the subcriticality of the widely-spread out fuel debris, like that in Unit 2
pedestal.
⚫ Issues to be solved➢ In large fuel debris (which is larger than the minimum critical mass), it is believed that there are parts thatare likely to reach criticality locally and parts that are not likely to reach criticality, and there is noknowledge of subcriticality measurement for such a non-homogeneous system.
➢ There is a concern about the impact on subcriticality measurement as there is a possibility of injecting
insoluble neutron absorbers during fuel debris processing and retrieval.
⚫ Approaches to development➢ A subcriticality system simulating large non-homogeneous fuel debris will be
constructed at the Kyoto University Criticality Assembly (KUCA) and a
subcriticality measurement test will be conducted using the criticality approachmonitoring system. (Figure 1)➢ In addition, a subcriticality measurement test will be conducted in a system inwhich a neutron absorber is placed between the core and the detector. (Figure
2)⚫ Expected outcome➢ Errors which should be expected in thesubcriticality measurement of fuel debris➢ Methods for measuring fuel debris(detector sensitivity, placement,
measurement time)

Region中性 子where
増倍neutron
しやすい領域

multiplication is
easy

Region where
中性子増倍しneutronにくい領域

multiplication isdifficult

Neutron detector
中性子検出器
Neutronabsorber
中性子吸収材Reactor
炉心部

中性子
Neutron検出器
detector

Figure 1 Image showing thetesting of non-homogeneous system

core

Figure 2 Image showing the test
using neutron absorbers©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (ii) ① Technological development of criticality
No.276
monitoring method]a. Confirmation of the feasibility of subcriticality measurement⚫

Overview of testing facility➢ Kyoto University Criticality Assembly (KUCA) (Figure 1)➢ Fuel: Enriched uranium➢ Moderator: Polyethylene (solid moderator)➢ A desired fuel debris simulated core is prepared bychanging the number of enriched uranium cells, thenumber of polyethylene cells, the number of absorber cells,etc.

⚫ Test period
February 18 to 27, 2020

5 cm x 5 cm

Approx.
50cm
Polyethylene

Unit cell
36 unit cells⚫

Criticality approach monitoring system➢ Neutron detector: 3 units of B-10 proportionalcounter type (Figure 2 and 3)➢ The time history of the observed neutron signal
is identified and recorded at intervals of 100
nsec (Figure 4).
➢ Analysis performed using Feynman-alphamethodApprox.
150cm

1/16” enriched Uranium
1/8” Polyethylene
1/4” Polyethylene

Approx.
50cm
Polyethylene

Source: http://www.rri.kyotou.ac.jp/shiroya-lb/kuca.htm

Aluminumsleeve pipe

Figure 3 Measuring Equipment[During approach to criticality]
The neutrons created by uranium
chain reaction are predominant

U235

Multiple signals in a short time
/ Time correlation

Time T

Figure 4 Time series data of neutron signals

Sensitive length 150 mm Φ 25.4 mm

Figure 2 B-10 proportional countertype neutron detector

Figure 1 KUCA Criticality Testing Facility©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (ii) ① Technological development of criticality
No.277
monitoring method]a. Confirmation of the feasibility of subcriticality measurement⚫

Purpose of the test
➢ To confirm the applicability of subcriticality measurement by Feynman-alpha method to non-homogeneous debris distribution➢ To confirm the applicability of subcriticality measurement to distribution of fuel debris mixed with neutron absorbers.

⚫

Testing method➢ Non-homogeneous distribution containing three types of fuels with different reactivity was simulated by changing the composition of U,
polyethylene and graphite (Figure 1)➢ Neutron absorber (LiF) was mixed to one part, and a state where an insoluble absorber is mixed was simulated (Figure 2)➢ Subcriticality was measured by changing the arrangement of the three neutron detectors (Figure 1, 2)⚫

Test results➢ Neutron signals (time series data) required for the Feynman-alpha method (for this fiscal year, only till data collection)
were collected.
➢ Analysis and evaluation of the obtained data will be carried out in the next fiscal year

1
ぬ
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2

3

4

5

6

7

8

9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29●試験炉心の特徴
2Features
2 2 of2 the2 test
2 core2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2Simulates
2. ・燃焼度が低い（反応度大）デブリと燃焼度が高い
2 2 2the2 distribution
2 2 2 of2debris
2 2in which
2 2 debris
2 2with
2 low
2 degree2 2 of2 2 2 2 2 2 2 2
burn-up (high reactivity) and debris with high degree of burn-up (low2 （反応度小）デブリが隣接しているデブリ分布を模擬
2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 3 3 2 2 2reactivity) are adjacent.
2. ・デブリ取り出し箇所に水が浸入した場合を模擬
2 2 2the2 case
2 where2 2 water
2 2 had2 entered
2 2 the2 debris
2 2 retrieval
2 2 2location.
2 2 2 3 3 2 2 2Simulates
2. ・中性子源は燃焼度が高い側に存在
2 neutron
2 2 source
2 2 is2 on2 the2 side2 with2 2high2 degree
2 2 of2burn-up.
2 2 2 2 2 2 2 2 2 2 2
The(2) Polyethylene
2・k・ｋeff＝０．８９（Ｍ＝９．４）
2 2 2(M=9.4)
2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 ②ポリ
2 2 2 2
eff=0.89
2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2
2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2
・反応度小
・Low
reactivity
・反応度大
・反応度小
2 2 2 2 2 2 ・High
2 2 reactivity
2 2 2 2 2 2 2 2 2 2 2 2 2 2 ・Low
2 2reactivity
2 2 2・Waterpresent
・水あり
・Waterpresent
・水あり
2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 ・Water
・水なし
2 2 absent
2 2 2
2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2
2 3 3 2 2 2 2 2 2 2 21 21 21 21 21 21 23 23 23 26 26 26 2 2 2 2 2
2 3 3 2 2 2 2 2 2 2 21 21 21 21 21 21 23 24 23 26 27 26 2 2 2 2 2
2 2 2 2 2 2 2 2 2 2 21 21 21 21 21 21 23 23 23 26 26 26 2 2 2 2 2⑪：中性子検出器
2 (11)
2 Neutron
2 2 2 detector
2 2 2 2 2 11 2 11 2 2 11 2 2 2 2 2 2 2 2 2 2 2
中性子源
2 2 2 2 2 2 2 2 2 2 3 3 3 3 3 3 3 3 3 3 3 3 2Neutron
2 2 source
2 2
2 2 2 2 2 ③ボイド
2
2
2
2
2
3
3
3
3
3
3
3
3
3
3
3
3
2
2 2 2 2
(3) Void
2 2 2 2 2 2 2 2 2 3 3 3 3 3 3 3 3 3 3 3 3 2 2 2 2
3

Figure 1 Test core simulating non-homogeneous debris distribution

1
ぬ
る
を
わ
か
よ
た
れ
そ
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な
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の
お
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や 3

2

3

4

5

6

7

8

9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29●試験炉心の特徴
Featuresof the test core
2. ・燃焼度が低い（反応度大）デブリと燃焼度が高い
2 2 2the2distribution
2 2 2 of 2debris
2 in
2 which
2 2 debris
2 2with
2 low
2 degree2 2 of2 2
Simulates
2burn-up
2 2(high
2 reactivity)
2 2 2 and2 debris
2 2 with
2 2
2 2 of 2burn-up
2 2 (low
2 2 2
high2degree（反応度小）デブリが隣接しているデブリ分布を模擬
2reactivity)
2 2 are2 adjacent.
2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2・デブリ取り出し箇所に水が浸入した場合を模擬
location.
2. Simulates
2 2 2the2case
2 where2 2 water
2 2had2 entered
2 2 the2 debris
2 2 retrieval
2 2 2
2 2
・中性子源は燃焼度が高い側に存在
2. The2 neutron
2 2 source
2 2 is2 on2 the2 side
2 with2 high2 2degree
2 2 of 2burn-up.
2 2 2 2 2・中性子吸収材の存在を模擬of neutron
2・Simulates
2 2 2the2existence
2 2 2
2 2 absorber.
2 2 2 2 2 2 2 2 2 2 2・k eff=0.78 (M=4.6)
2 ・ｋeff＝０．７８（Ｍ＝４．６）
2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2
2 2 2 2 2 2 2 2 2・反応度大
2 2 2 2・High
2 2reactivity
2 2 2
・水あり
2 2 2 2・Water
2 2 present
2 2 2
2 2 2 2 2 2 2 2 2
2 3 3 2 2 2 2 2 2
2 3 3 2 2 2 2 2 2
2 2 2 2 2 2 2 2 2
Neutron⑪：中性子検出器
2 (11)2 2
2 2 detector
2 2 2 22 2 2 2 2 2 2 2 2
2 2 2 2 2 ③ボイド
2 Void2 2 2(3)
2 2 2 2 2 2 2 2

2 2 2 2 2 2
2 2 2 2 2 2
2 3 3 2 2 2
2 3 3 2 2 2
2 2 2 2 2 2(2) Polyethylene
2 2 ②ポリ2 2 2 22 2 2 2 2 2
2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2
・中性子吸
・反応度小
・Low reactivity
2・Neutron
2 2 absorber
2 2 2 2 2 2 2 2 2 2 ・反応度小
2 2reactivity
2 2 2
・Low
・Water present
included
収材含入・水あり
・Water2 2 2 2 2 2 2 2 2 2 2 2 2 ・水なし
2 2 absent
2 2 2
2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 22 21 21 21 21 21 21 23 23 23 26 26 26 2 2 2 2 22 21 4 4 4 21 21 23 24 23 26 27 26 2 2 2 2 22 21 4 4 4 21 21 23 23 23 26 26 26 2 2 2 2 2
2 2 2 11 2 2 2 2 11 2 2 2 2 2 2 2 2 2
中性子源
2 3 3 3 3 3 11 3 3 3 3 3 3 2 Neutron
2 2 source
2 2
2 3 3 3 3 3 3 3 3 3 3 3 3 2 2 2 2 2
2 3 3 3 3 3 3 3 3 3 3 3 3 2 2 2 2
3

Figure 2 Test core simulating the mixing of neutron absorbers©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (ii) ① Technological development of criticality
No.278monitoring method]b. Confirmation of the applicability of neutron detectors to the field⚫ Purpose of Development➢ To formulate the basic specifications of a neutron detector unit that can be transported by a robot arm andplaced on top of fuel debris for measuring the subcriticality of fuel debris. (Figure 1)(Note) Unit: Refers to an integrated structure including multiple sensors and shield, and in some cases a preamplifier⚫ Issues to be solved➢ Neutron detectors for measuring the subcriticality of fuel debris are required to have high sensitivity, hightime resolution, gamma ray resistance, and electromagnetic noise resistance.
➢ The detector needs to be compact and lightweight, so that it can be transported with a robot arm.

⚫ Approaches to development(b-1) Study on the specifications of alternative neutron
detector candidates for size and weight reduction(1: Corona type manufactured by RosRAO)(b-2) Study on the specifications of alternative neutron
detector candidates for size and weight reduction(2: SiC type manufactured by IPL)(b-3) Study on the structure of detector unit (B-10 / He3 proportional counter)(b-4) Study on measures against noise⚫ Expected outcome➢ Basic design of neutron detector unit
➢ Alternative candidates for neutron detector

コネクタでケーブル接続
Cable connection using
connector (through the arm)
(アーム内を通す)
Tool changer
ツールチェンジャー

Figure 1 Image of neutron detector unit in side-access retrieval method©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (ii) ① Technological development of criticality
No.279monitoring method]b. Confirmation of the applicability of neutron detectors to the fieldb-1. Study on the specifications of alternative neutron detector candidate for size and weight reduction (1: Corona typemanufactured by RosRAO)⚫ Implementation details
Based on the results of the related project “Project of Development of Fundamental Technology for Retrieving Fuel
Debris/ Internal Structures (development of small-size neutron detectors)”, the specifications of the prototype used for
confirming the applicability of the small-size neutron detector (Corona detector) to the field were summarized.
⚫ Future implementation plan
The designing and prototyping of a small-size neutron detector will be carried out, and its applicability to the field will be
confirmed by a performance confirmation test with the Kyoto University Criticality Assembly (KUCA).

Requirement specification of prototype (detector) for performance
testing with KUCA
Requirement specification

Environmental
conditions

Temperature: 0 ~ 40 ℃
Humidity: Should not condense
Gamma-ray background dose rate: ~100 mGy/hr

Dimension

The size should be accommodated inside the aluminum squarepipes making up KUCA, and the sensitive length is set to a coreheight of 350 mm.

Material

Enclosed gas: Ar, He
Neutron converter: 10B and 3He

Thermalneutronsensitivity

2 cps/nv and above (10B type)
15 cps/nv and above (3He type)

Pulse amplification factor(signal magnitude)

Items

Corona detector

Corona discharge area
Uses He as the main componentof the enclosed gas. Electronicavalanche is suppressed byusing a high load resistor(~100MΩ) etc.
Figure 1 Overview of corona detector©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (ii) ① Technological development of criticality
No.280monitoring method]b. Confirmation of the applicability of neutron detectors to the fieldb-2. Study on the specifications of alternative neutron detector candidate for size and weight reduction (2: SiC typemanufactured by IPL)

⚫ Implementation details
Based on the results of the related project “Project of Development of Fundamental Technology for Retrieving Fuel
Debris/ Internal Structures (development of small-size neutron detectors)”, the specifications of the prototype used for
confirming the applicability of the small-size neutron detector (semiconductor neutron detector) to the field weresummarized.
⚫ Future implementation plan
The designing and prototyping of a small-size neutron detector will be carried out and its applicability to the field will be
confirmed by performing confirmation test with the Kyoto University Criticality Assembly (KUCA).
Requirement specification of prototype (detector) for performance
testing with KUCA
Items

Requirement specification

Environmental
conditions

Temperature: 0 ~ 40 ℃
Humidity: Should not condense
Gamma-ray background dose rate: ~100 mGy/hr

Dimension

The size should be accommodated inside the aluminum squarepipes making up KUCA, and the sensitive length is set to a coreheight of 350 mm.

Material

Base material: Thin SiC semiconductor
Neutron converter: 10B and 157Gd

Thermalneutronsensitivity

As a requirement specification, thermal neutron sensitivity is set
to 10 cps/nv, but the specific feasible sensitivity shall be
evaluated from the sensitive length, outer diameter size, etc., at
the detailed design / manufacturing stage.

Polyethylene
moderator

SiC detector

Preamplifier (tungsten
shield)(Reference) Proposed detector (detection unit) under study
in the Development of Technology for Investigation inside
Primary Containment Vessel©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (ii) ① Technological development of criticality
No.281monitoring method]b. Confirmation of the applicability of neutron detectors to the fieldb-3. Study on the structure of detector unit (He-3 proportional counter)⚫ Implementation details➢ The design of the detector unit (and neutron detector) was reviewed to address the design /manufacturing issues of the detector unit (#).
# The detector unit is intended to be used during the preparation period for fuel debris
retrieval and for subcriticality measurement during fuel debris retrieval (see next slide).
Table 1 Issues in the design / manufacture of detector unit, and the response policies
IssuesResponse policy for the issues

Overview of
implementation in
FY2019Weight reduction(Target: less than 50kg)

Improving the radiation resistance by reviewing the design
parameters (of the detector) of the He-3 proportionalcounter.

Study on the designof He-3 proportionalcounter

To confirm the feasibility of
He-3 proportional counter
through prototype tests

Response toenvironmental
conditions

Studying structures that can withstand the environmental
conditions of 1F. Confirming its ability through prototyping /tests.

Reducing the impactof electromagnetic
noise

Improving the S/N ratio (signal-to-noise ratio) byincorporating a preamplifier in the detector unit.

Study on the designof detector unit

To confirm the feasibility ofdetector unit throughprototype tests

Handling cables
through remote
operations

By adopting the tool changer method (the cable is laid
inside the robot arm), the handling of the cable itself inside
the containment vessel is eliminated.

-

To confirm the feasibility oftool changer throughprototype tests

Scheduled to be
implemented in FY2020©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (ii) ① Technological development of criticality
No.282
monitoring method]b. Confirmation of the applicability of neutron detectors to the fieldb-3. Study on the structure of detector unit (He-3 proportional counter) (Continued)⚫

(Supplement) Operation of detector unit
➢ The detector unit is used during (A) the preparation period for large-scale retrieval and for (B) dailysubcriticality measurement.
Table 1 Subcriticality measurement / criticality approach monitoring during fuel debris retrieval (large-scale)Work phase

Measurement details

Measurement
frequency (goal)

First time

Measurement
time (goal)Once a week

A

Preparation period forlarge-scale retrieval① The local subcriticality distribution before starting largescale retrieval measurement using the detector unit.
Subcriticality is measured for a long duration only the first time
to confirm that the subcriticality is properly obtained.

B

Before the start of daily
retrieval operation① The detector unit is used to confirm adequately deep
subcriticality before starting the daily fuel debris retrieval (thesubcriticality and counting rate established as criteria for themeasurements in C are to be obtained)

Twice a day

Once per hourC

During the daily retrieval
operation② A unit detector is used to ensure that the situation is not
approaching criticality during daily fuel debris retrieval.

Once perprocessing orcollection task

Once in 10 min

DWhen an effective
change in neutron
source intensity isobserved

When a change in the shape of the fuel debris / movement of
the detector is observed, the subcriticality and counting rateestablished as criteria for the measurements in C are to beobtained using ① the detector unit and ② the unit detector.

When required

Once per hourE

When it is deemednecessary

Subcriticality is measured when it is deemed necessary to
identify the extent of subcriticality for some reason (in case ofproblems).

When required

Once a day

After that, about
several tens oftimes

Once per hour©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (ii) ① Technological development of criticality
No.283
monitoring method]b. Confirmation of the applicability of neutron detectors to the fieldb-3. Study on the structure of detector unit (He-3 proportional counter) (Continued)⚫

Results obtained so far➢ By reviewing the design parameters of the detector (adopting a multi-cell He-3 proportional counter), the prospect
for significant weight reduction was obtained.
•
Last year’s design plan 120 kg ⇒ Approx. 40 kg (without preamplifier), Approx. 60 kg (with preamplifier)⚫

Future implementation plan
➢ To prototype the detector unit (including the detector and tool changer) and confirm the feasibility of the design bytesting.

Weight is reduced by thinning the lead shield as theradiation resistance of the detector is improved.
Approx. 42 cm

Neutron moderator(polyethylene)[Principle]
By arranging multiple anodes and cathodes in one He-3 gas
container, it is possible to perform measurements equivalent tomultiple proportional counters
CathodeAnode

Diameter
About 15 cm

Preamplifier

Gamma shield (lead)
Multi-cell typeHe-3 proportional counter# The plan is to install a neutron shield (polyethylene) on the top and sides
to shield neutrons from fuel debris that is not subject to measurement.

Figure 1 Structure of detector unit (draft)External appearanceof detector

Internalstructure

Cross-sectionof detector

Suppressing the impact of noise (pile up) due
to gamma rays by increasing the diameter
Figure 2 Principle and structure of multi-cell He-3
proportional counter (draft)©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (ii) ① Technological development of No.284
criticality monitoring method]b. Confirmation of the applicability of neutron detectors to the fieldb-4．Countermeasures for noise generated from retrieval equipment⚫ BackgroundIn the side-access retrieval method, the cables of the fuel debris retrieval equipment and cables of the neutron
detector are likely to pass through the same pedestal opening or PCV penetration (Figure 1), so the neutron
measurement system gets impacted by the induced noise due to the proximity of the cables [1].
⚫ Implementation details
・ By changing the cable between the B-10 proportional counter and preamplifier with a noise-resistant cablecovered with a copper mesh shield (Figure 2), it was confirmed that the peak value of the induced noise was
reduced by a maximum of one digit (Figure 3).
・ By a test in which the retrieval equipment cables and the detector cables were brought close to each other(Figure 4), it was confirmed that their coexistence is possible without the impact of induced noise (Table 1).
⚫ Remaining issues
・ Noise countermeasures using wraparound from the power system (Independent electrical system, noise
removal / suppression)

Contracting
tube
収縮チューブ

Ground wire
接地線

・Absorbs electromagnetic waves as
induced current and eliminates it fromthe system through the grounding part

Noise level corresponding
to maximum wave heightvalues of neutron

Figure 2 Noise resistant cable

1/8

200

SHV
connector
SHVコネクタ
Pipe saw

Copper銅網シールド
mesh shield100

PCV penetration

Preamplifier
Cable length 20 - 30 m

Fuel debris
Detector

Preamplifier

PCV penetration
Pedestal opening

Figure 1 Image of installation in 1F

30 m

Conventional
従来ケーブルcable
SHVコネクタSHV
connector
Contracting
tube
収縮チューブ

0Inside thecontainment vessel

Coexistence test

Simulated noise test

4
3
2
1
0

Outside thecontainment vessel

1/16

Pipe saw drive cable
Neutron detector cableHydraulic arm
mechanism

Vicinity areas

Vicinity areas

Crawler mechanism
Hydraulic arm cable Electric crawler cable

Neutron detector cable

Figure 4. Coexistence test with retrieval equipment

Conventional cable
従来ケーブルTable 1. Sensitivity reduction rate due to induced noise from the retrieval device

Noise resistant cable
耐ノイズケーブル

Cable type

Conventional
cable

Noise resistant cable

Pipe saw

32 %0%Hydraulic arm

0%0%Electric crawler

50 %

0%

1/14

300
400
500
600
Frequency F [kHz]Figure 3 Induced noise reduction effect (when the induced
noise current value is 200 mA)

Maximum
wave heightvalue ofneutrons

700[1] IRID, Research report on Subsidy Project of Decommissioning and Contaminated Water Management in©International Research Institute for Nuclear Decommissioning
FY 2016 Supplementary Budget "Upgrading of Methods and Systems for Retrieval of Fuel Debris and Internal Structures (development of technology for the creation of criticality safety control methods)"

7. Implementation items of this project [3) (ii) ① Technological development of criticality
No.285monitoring method]c. Study on the criticality approach management procedures ~ Coordination with methods ~⚫ Implementation details
To study the criticality approach management procedures during fuel debris retrieval, the criticality safetycontrol method at each retrieval step of each fuel debris retrieval method currently under study (three sideaccess methods, two top-access methods, figures on next slide) was studied, and the criticality approachmonitoring methods to be applied were studied.
⚫ Future implementation plan・To create the concept of criticality approach monitoring system based on the constraints related to the methods
and retrieval facilities organized this time.
・To create on-site management procedures by reflecting the requirements and concerns of criticality safetycontrol in the study on methods.
Table. Study on the criticality safety control method at each retrieval step (From the viewpoint of coordinating the methods)
Items

Points to be clarified

Viewpoint

Criticality approachmonitoring system

・Steps to be monitored, monitoring methodand detector to be used
・Loading and allowable weight of retrieval
system・Getting a clarity on the criticality approachmonitoring system study conditions
・Confirmation of consistency with the retrieval
procedures

Criticality safetycontrol method(Processing
restrictions, etc.)

・Steps that require criticality safety control,and criticality safety control requirements

・Confirmation of consistency with the retrieval
procedures

Insoluble neutron
absorbers

・Steps that may require the addition ofabsorbers
・Loading and allowable weight of retrieval
system・Getting a clarity on the study conditions
related to loading an absorber injection device
on the arm

Notes on criticality
safety control

・Items to be noted from criticality safetycontrol point of view

・Getting a clarity on the criticality concerns
related to analysis and response to equipment
failure modes.
©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (ii) ① Technological development of criticality
No.286
monitoring method]c. Study on the criticality approach management procedures ~ Coordination with methods ~FigureMethods under study (Side-access: PLAN-A~ C, Top-access: routes A, B
PLAN-A

PLAN-B

PLAN-C
機器搬送台車Equipment
transport cart

Crane for carrying-in equipment

機器搬入クレーン

Equipment hatch cell

機器ハッチセル

Maintenance
保守セル cell

Carrying-in
搬入セルcellWaste
carry廃棄物
搬出セル
out cellReactor building
wall opening
原子炉建屋壁開口

Reactor building wall

原子炉建屋壁横

Maintenance
cell
保守セル

燃料デブリ
Fuel
debrisretrievalcell
取り出しセルSide access

Equipment
機器ｾﾙcellFueldebris
燃料デブリ
carry-out
cell
取り出しセル

廃棄物搬送台車
Waste
transport cart

BSW

Cell transportation
system
セルの搬送システム

PCV
Pedestal

Fuel debris
燃料デブリ
搬出セル
retrieval
cell

ペデスタル

Mini
cask system
ミニキャスクシステム

Wastetransport cell
廃棄物搬送ｾﾙ

Route (access tunnel)
通路(アクセストンネル)

・Access via X-6 penetration
・X-6ペネからアクセス
・Large size arm used
・大型アーム使用

・Access
via equipment hatch
・機器ハッチからアクセス
・Multiple
operating devices used
・複数の作業装置を使用

ルートA
Route AGas management
system inside thebuilding

Waste carrying-out area

Upper
container
Container crane

Crane 2 inside the cell

Crane 1 inside the cellPressure
vesselCarrying-out
crane 1

Canister handling crane 1

・Separateroute for carrying-in devices and carrying-out waste
・装置搬入・廃棄物搬出の動線を別系統

Route B
ルートB

Gas management
system inside thebuilding

Transfer container storage area

Device carrying-in side・X-6および新開口からアクセス
・Access
via X-6 and new opening

Waste carrying-in side

Canister handlingcrane 2

Manipulator
airlock

Canister

Waste
Sealedcontainer carrying-out
door 1
Canister
installation platform

Shieldedcontainer

Airlock door

Sealed container cart

Fuel debris
retrieval cell

Transfer
container
inspection
area

Waste
carrying-out
door 2

Shielded container
cart inside the cell

Waste
carrying-out
door 3

Shared devices

共通装置

Shielded container
cart outside the cell

Liftingmechanism(hanging tool)

Fuel debris retrieval
cell crane

Carrying-in
equipment etc.

Well

Top access上Shroud support
System for controlling water
inside the building

Dedicated task

Canisters
Inside thereactorTransferred using lifter

Shieldcontainer

Individual
devices

個別装置

Always at 100% operation

System for controlling water
inside the building

Fuel debris

Fuel debris

Always at 100%
operation

3 or more doors to control the releaseof radioactive materials from within
the R/B when the doors are opened

CRD housing

Fuel debris storage
facility

・Simple route・シンプルな動線
・Access to thebottom of the core without removing the shroud, etc.
・シュラウド等は撤去せず、炉底部へアクセス

・オペフロレベルで遮へい、DSP経由でデブリや干渉物を搬出・共通装置および対象ごとの個別装置で炉内構造物を全て撤去

・Shielding at operating floor level, debris and interferences carried out through the DSP・Complete removal of all internal structures via shared devices and individual devices for each target©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (ii) ① Technological development of No.287
criticality monitoring method]c. Study on the criticality approach management procedures ~Concept of criticality approach monitoring~

Criticality approach monitoring is
implemented using 3 types of detectors

Table. Organization of functionality of the detectors used for criticality
approach monitoring
Name

Neutron detector unit

Unit detector

Detector for re-criticality detection

Purpose of use

To understand the pre-job state

To decide on start of processing

To detect unexpected changes

Subcriticality measurement

Criticality approach monitoring before
and after processing

Continuous monitoring of neutron flux
during processing

Functionality

Neutron detector unit
・ Includes multiple detectors for high sensitivity
・ To locally measure the fuel debris processing
location

Measurement of absolute valueof neutron multiplication factor

Measurement of relative change of neutron multiplication factorForm

A type that can be installed on fueldebris

Arm mounted type

A type that can be installed on fueldebris

Weight120 ~ 150kg
Currently, size reduction is understudy

30 ~ 50㎏

Less than or equal to 30 ｋｇ

Calculation time

A few days to a week

About 10 minutes

Continuous

Measurement points

One point near the retrieval startpoint

Moved as appropriate according to theprocessing position.

Same as on the left

Unit detector
・ Comprises of one detector・ To locally measure the fuel debris
processing location

Detector for re-criticality detection
・ Comprises of one detector・ Monitors the entire area away from theprocessing location from a bird's-eye view
©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (ii) ① Technological development of criticality monitoring
No.288method]c. Study on the criticality approach management procedures ~ Deployment of criticality safetycontrol to retrieval steps ~・ The concept of criticality safety control was deployed to the retrieval steps, and the constraints on monitoring system and points to be
noted on criticality safety control were organized.
Example of Criticality Safety Control methods and requirements in the retrieval steps
Step

Sub-steps

Maximum arm
mounted weight

Processing restriction /Criticality safety control
methodPossibility ofpre-fillingabsorberMonitoring the approach to criticality
Subcriticality
measurement

Monitoring for each
fixed amount ofprocessing

Continuous monitoring
during processing

Other criticality prevention
measures

Remarks (*2)

2. Removal of interferences inside PCV and preparation for fuel debris retrieval
Absence of deposits
presumed to be fuel debris:no constraints③ Removal of 4) Processing
interferences and cutting ofin PCV
interferences

Presence of deposits (small
amount of deformation *1):
Limitation on the amount of(Carrying into cutting powder that falls on
PCV) ~ 100kg lower deposits(Arm mounting) Presence of deposits (large~ 30kgamount of deformation):・ Processing along withcriticality approachmonitoring・ Limitation of the amount of
cutting powder that falls on
lower deposits

Classification of
retrieval method steps(Sub-steps inserted forthe steps with criticality)

Criticality safetycontrol method,
processing
restrictions, etc.
Constraints on detectors
and insoluble absorber
injection device- (*1)-〇
Duringapproach to
criticality

-

-

-

-

-

Neutron flux
measurement andcriticality approachmonitoring before andafter processing andcutting

Method of performing
criticality approachmonitoring
Steps in which
insoluble
absorbers may be
applied・Preventing the cut pieces*1 The criterion forfrom falling towards thedetermining the amount ofdeposits in the pedestal
・Preventing heavy objects deformation is the minimum
critical weight (30-60kg)
from falling on top of
deposits in the pedestal(Measures to prevent
As a backup, continuous contact with CRD, measures
to prevent fall of cutting
monitoring may betools)
required during
processing* 1 "-" means not applicable.
* 2 If there are no special notes, it is left

Precautions from criticality safetycontrol point of view (mainly
measures to prevent theoccurrence of incidental events)©International Research Institute for Nuclear Decommissioning7. Implementation items in this projectNo.2893) Development of Technology for Ensuring Safety while Retrieving Fuel Debris(ii) Development of element technology for prevention and monitoring of criticality② Development of technology for criticality prevention
Development results till date and implementation details for this year
Final Goal

Creation ofcriticality
prevention
technology

Incorporation into
fuel debris
retrieval system

Goal to be achieved before actual
application

Achievement status

Details for this fiscal year

① Selection of candidates

Selection of candidate materials for insoluble neutron absorbers byconfirming basic characteristics, irradiation characteristics, andnuclear characteristics (4 types of candidate materials selected)② Confirmation of workability
and specifications ofsubcriticality maintaining
conditions

Estimation of required injection amount based on the assumed
usage
Confirmation of workability with fuel debris crushed using chisel,and confirmation of absorber distribution and subcriticality
maintenance evaluation (solid type absorber)③ Evaluation of corrosion
effects

From the long-term irradiation test, it was found that the hydrogengeneration G value is less than the design value, and the diluted
seawater in which the absorber component is eluted has a pH of 6
or more.

-To evaluate the impact of rust
inhibitor effect under irradiation③ Evaluation of side effects

The number of canisters and the amount of waste increased by up
to approximately 10% for solid absorbers and up to approximately
40% for solidified absorbers.

① Study on the insoluble
neutron absorber injection
device and operation methods

The method for injecting absorbers was formulated
A procedure for injecting absorber for fuel debris processing using
chisel was formulated (assuming MCCI fuel debris present underpedestal)② Design of absorber injection
device

The concept of absorber injecting device was formulated based on
the weight / dimensional restrictions of the absorber injecting device
and the absorber transfer route restrictions.

③ Verification of combination
with the retrieval equipment

The absorbers for the fuel debris processing methods for each fueldebris position were organized.

To confirm that it can be injectedunder water
To confirm the impact due toprocessing©International Research Institute for Nuclear Decommissioning7. Implementation items in this projectNo.2903) Development of Technology for Ensuring Safety while Retrieving Fuel Debris(ii) Development of element technology for prevention and monitoring of criticality② Development of technology for criticality prevention

The proper use of insoluble neutron absorbers based on the state of fuel debris at the time offuel debris retrieval, conceptual design of injection device and operational methods of insoluble
neutron absorbers shall be studied. In addition, the impact of fuel debris processing on thefunctionality of insoluble neutron absorbers shall be confirmed using element tests, and thoseshall be reflected in the study on operational methods. Furthermore, the corrosion effect of
insoluble neutron absorbers on structural materials in an environment considering the use ofrust inhibitors under irradiation effects shall be studied.
The following are included as important development study items, and issues shall beidentified and organized by conducting element tests, as necessary:a. Study on insoluble neutron absorber injection device and operation methods• The possibility of injecting insoluble neutron absorbers in water was confirmed
through tests.
• The impact of processing absorbers was confirmed through tests.
b. Study on the corrosion effects considering the use of rust inhibitors under irradiation
effects•The possibility of using a rust inhibitor and a glass material containing B/Gd
together was confirmed by an irradiation test.

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (ii) ② Development of technology for criticality prevention] No.291
a.
⚫

⚫

Study on insoluble neutron absorber injection device and operation methods

Purpose of development➢ To develop a device to inject an insoluble neutron absorber so as to prevent fuel debris fromapproaching criticality.
Issues to be solved➢ So far neutron absorbers of the solid type or the solidified type (changed from liquid to solid) havebeen developed (Figure 1). The concept of devices to inject these were studied (Figure 2), but it isunclear whether they can be used under water.
➢ By injecting the neutron absorber prior to the processing of the fuel debris, it is assumed that it will
be processed together with the fuel debris during the processing operations. Although the impact ofprocessing on the solid type absorbers was confirmed by means of testing, the impact on solidified
type absorbers is unclear.

5 mm

5 mm

B4C sinteredmetal material(Solid)Glass material
with B/Gd(Solid)

5 mm

10 mm

Water glass/ Gd2O3
Gd2O3 particles
granulated powder(Solid)(Viscous)⚫

⚫

Approaches to development➢ Study of the design specifications and the operation method of the absorber
injection device.
➢ Study of the proper usage of the neutron absorber assuming the state of the fueldebris.
➢ Verification of the impact of fuel debris processing on the functionality of theneutron absorber by carrying out element tests wherein the simulated debris
injected with a neutron absorber is processed.
Expected outcome➢ Basic specifications of the neutron absorber injection device➢ Operation method of the neutron absorber

Figure 1 Insoluble neutron absorber

Example of injection device
for solids (powder)

Figure 2 Image showing the absorber
injection device mounted on the robot arm©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (ii) ② Development of technology for criticality prevention] No.292
a.

Study on insoluble neutron absorber injection device and operation methods⚫ Development Schedule

Study items

FY2019
４

5

6

7

8

10

11

12

1

2

▼

▼
Organization of
scenarios

Organizationof conditions

1. Proper usage method

3Year-end Report

Interim Report

Major Milestones

Criticality analysis

Summary

Planning

Preparation

2. Design, manufacture andelement test of the injection device

Underwater injection test

Planning

3. Verification of impact due to fueldebris processing

9

Evaluation

Preparation

Processing test of absorber
Evaluation

Remarks©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (ii) ② Development of technology for criticality prevention] No.293a.

Study on insoluble neutron absorber injection device and operation methods
Proper usage method of the absorber•

The fuel debris is presumed to be in various shapes, such as rod shape or granular shape etc. The type ofneutron absorber which is best for the various shapes of fuel debris, should be used.
The insoluble neutron absorbers are divided into the solid type and the type that has solidified (water glasstype) from the liquid state due to lapse of time. The water glass type has fluidity or viscosity immediately after
injection, so it is effective when the residual fuel (fuel rod stubs) is standing close together vertically (Figure 5),
when the gaps between the fuel debris are small (Figure 3, 4), and when there is a lot of unevenness on thefuel debris surface (Figure 6).

•

DebrisAbsorber (solid)

Figure 1 Solid absorber is used
for pebble-like fuel debris

Figure 3 Liquid to solidified absorber isused for pebble-like fuel debris

Figure 2 Solid absorber is used
for the bedrock-like fuel debris

Figure 4 Liquid to solidified
absorber is used for the
bedrock-like fuel debris

Figure 5 Liquid to solidified
absorber is used for rod-shaped
fuel debris

Figure 6 Liquid to solidified absorber is used for fueldebris with large uneven surfaces©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (ii) ② Development of technology for criticality prevention] No.294
a.

Study on insoluble neutron absorber injection device and operation methods
Underwater injection test for neutron absorbers⚫ Purpose of the test
• The technology which forms the basis for the injection device is presumed to be used in air and there
are no results so far regarding its use under water.
• It is anticipated that as the water depth becomes deeper, it will be difficult to discharge the absorber
against the water pressure.
• To verify whether the solid and viscous insoluble neutron absorbers can be injected under water.
⚫ Testing methods• By presuming the water level at Unit 3 of 1F, the test was carried out inside the water tank set up in the
Recompression chamber in order to simulate the water pressure (0.5 atmospheric pressure) at 5 munder water. (Figure 1)• The injection speed of the absorber was evaluated.

Tray for evaluation

Neutron absorber
injection device

Tool changer to connect
with the robot arm

Recompression chamber

470

Acrylic water tank

Figure 1 Overview of the test device©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (ii) ② Development of technology for criticality prevention] No.295
a.

⚫

Issues at the time of actual
application• When the water glass typeabsorber was injected, fineparticles that did not settle
Masterplate
マスタープレート
whirled up and the water
turned turbid.
Air pressureport
空気圧ポートRc1/4
Rc1/4,
２カ所2 places
It is necessary to confirm thepressureport
空気圧ポートRc1/8
impact on the water treatment Air
Rc1/8,
6カ所 6 places
system or the impact on theElectric signal
connector
電気信号コネクタ
visibility of the fuel debris(masterside)（マスタ側）
processing work.
Electric
signal
電気信号コネクタ（ツール側）
connector(tool side)

ツールプレート
Tool
plate

Tray weight / arbitrary unit

Test results• Both solidified type (Gd particles type) and viscous type (water glass type)absorbers could be injected at a stable speed in water with depth equivalent to 5
m. (Figure 1)• The intended injection speed could be achieved. (Figure 2)
Gd particles type: 3.3g/sec
Water glass type: 2.8L/min(Calculated on the assumption that it will take 30 minutes to inject the quantity
needed for the 300 kg of fuel debris retrieved in a day)• Water resistance performance of the tool changer prototyped for underwater
testing was confirmed at a water depth of 5 m. (Figure 3)(Static immersion test of 30 minutes based on the JIS waterproofing standard
IP7)

Stable injection
at a constant
speedTime / sFigure 1 Temporal changes in the amount of absorber injected(Water glass type)

Discharge speed / g/s⚫

Study on insoluble neutron absorber injection device and operation methods
Underwater injection test for neutron absorbers (Continued)

Design targets(2.8L/min)

Air pressure port
空気圧ポートRc1/4
２カ所
Rc1/4, 2 places

Air pressure port
空気圧ポートRc1/8
6カ所
Rc1/8, 6 places

O ring (forOリング（防水用）

waterproofing)

Condition 2

Condition 4

Condition 1

Condition 3

Condition 5

Condition 6

Figure 2 Injection speed of absorber (Water glass type)Figure 3 Tool changer prototyped for underwater testing©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (ii) ② Development of technology for criticality prevention] No.296
a.
⚫

⚫

•
•
•
•

Study on insoluble neutron absorber injection device and operation methods
Verification of impact due to fuel debris processing

Purpose of the test
To assess the impact of processing when the fuel debris is processed with the absorber scattered on the debris (Figure 1).
To confirm that the absorber is functioning effectively (the fuel debris and the absorber should mix together after processing).
To confirm that the fuel debris processing is performed without trouble even when the fuel debris is covered with the absorber.
To examine the size (particle size) of the fine powder of the processed absorber.

Testing methods• The MCCI debris was assumed as the simulated fuel debris and a mortar with regulatedcompressive strength was created.
• Absorber (water glass type) was dropped on the surface of the simulated fuel debris (Figure 2).
• The surface of the simulated fuel debris (approximately 5 cm in depth) was crushed (9 places at 5cm intervals) using a chisel processing device (electric hammer) (Figure 3).
• Measurements were done under water (particle size measurement) and in air (measurement of
mixing).

X-Y Positioning jig

Electric chisel
Pressing by the weightof the chisel (28kg)

Absorber (Water glass type)

Simulated fueldebris (Mortar)

Figure 2 Water glass type absorber (viscous solidified)Figure 1 Image of chisel processing

Figure 3 Outline image of test device©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (ii) ② Development of technology for criticality prevention] No.297

The test sample after processing was
solidified by impregnation with epoxy
resin and on collecting a Φ30 mm
sample for analysis by core boring, it
was found that the neutron absorber(Gd) had infiltrated (got mixed) deeply(5 to 7 cms). (Figure 2; White part)(Figure 3)

Direction of depth

Case 4: Normal mortar
Compressive strength: 20Mpa

Case 5: Normal mortar
Compressive strength: 10Mpa

Mass (%)

Absorberpresent

Amount of fine particles ofsize less than 250μm or lessincreased when there wasan absorber

Although thevariation is large,
there is no
significant
difference

Particle size
Figure 1 Particle size distribution of pieces crushed due to chisel
processing

Case 6: Lightweight mortar
Compressive strength: 20Mpa

Case 4(Position A)

Position A•

Test results
The processing at 9 locations required 90 to 150 seconds and no
significant difference was noted due to the presence or absence of theabsorber.
• The broken pieces were collected after chisel processing, sorted
through a strainer, and their weight was measured to obtain the particle
size distribution data (Figure 1). When the absorber was present, there
was a tendency that the number of fine particles of size less than 250μm
or less increased. This is believed to be because of the minute scatteredparticles resulting from the crushing or decomposition of the neutron
absorber. This result will be presented as an input for the water
treatment system design.
•

Position B⚫

Study on insoluble neutron absorber injection device and operation methods
Verification of impact due to fuel debris processing (Continued)
None

Case 5
Case 4(Position B) (Position A)

Case 5(Position B)

Case 6
Case 6(Position A) (Position B)

Direction of depth
Gd amount / ga.

Estimate of Gd amount

7cm
Surface

Figure 2 Sample for analysis obtained by core boring

Interior 1

Interior 2Interior 3

Figure 3 Weight of absorber (Gd) included in the direction of
depth after processing©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (ii) ② Development of technology for criticality prevention] No.298b.

Study on the corrosion effects considering the use of rust inhibitors under irradiation effects⚫

Purpose of development➢ To develop an insoluble neutron absorber to prevent criticality.

⚫

Issues to be solved➢ So far the corrosion effect due to neutron absorbers has been evaluated from theperspective of changes in water quality associated with long-term gamma ray irradiation.
The remaining issues arise when the use of a rust inhibitor (Table 2) is considered forcontrolling PCV corrosion in the reactor water environment.

⚫

⚫

Table 2 Rust inhibitor and water
quality to be studied

Approaches to development➢ Experimental evaluation of the impact on the rust inhibitor effect of neutron absorbers
containing rare earths whose effect on the corrosive behavior in a gamma-ray irradiatedenvironment is unknown. (Table 1)

Expected outcome➢ A combination of rust inhibitor and insoluble neutron absorber that can be applied

Water quality to be studied

Classification

Rust inhibitor

B
Concentration
(ppm)

pH

I.

Na2B10O16

≧4000

(8-9)

II.

Na2WO4+Na2B10O16

＜4000

(8-9)

III.

ZSMMP

＜(40)

6-8

IV.

ZSCMP

＜(40)

≧10

Values in the parentheses
are the result values

Table 1 Dissolution characteristics and secondary impact evaluation
status of the selected candidate material
Dissolution characteristics
Classification

SolidsCandidate material

Rust prevention measures

Canister(Chisel)(Desk-top study #)(Long-term irradiation)Fe, B

No effect

No effect

In design

Evaluated

6 to 9B, Gd, Rare earth typeNo effect

Experimental evaluation required

In design

Evaluated

7 to 9Gd

Not evaluatedNo effect

In design

Evaluated

12

Na, Gd, Ca

No effect

Water quality regulation required

In design

Evaluated

Dissolution components

B4C sintered metal material

7 to 9Glass material with B/Gd
Gd2O3 particles
Water glass/Gd2O3

Solid

granulated powder

Subsequent processes

Fuel debris processing

pH

Liquid toDuring retrieval

ZSMMP: Zinc/Sodium carbonate mixed phosphate, ZSCMP: Zinc/Sodium molybdate mixed phosphate

Waste# Desk study based on the pH and dissolution components©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (ii) ② Development of technology for criticality prevention] No.299b.

Study on the corrosion effects considering the use of rust inhibitors under irradiation effects⚫ Development Schedule

Study items

FY2019
４

5

6

7

8

9

10

11

Interim Report

Major Milestones

12

1

2

3

Year-end Report▼▼

Detailed test plan
Selection of QST Takasaki Advanced Radiation ResearchInstitute experimental issues

1. Plans
2. Test preparations

Setting up of the Test Ancillary Facility
Preparation of test specimen and testing
solution
Absorber irradiation test

3. Absorber irradiation test
Evaluation

4. Evaluation of test results

Remarks©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (ii) ② Development of technology for criticality prevention] No.300b.
⚫
⚫

Study on the corrosion effects considering the use of rust inhibitors under irradiation effects
Combination with rust prevention measures➢ Evaluation was carried out for the classifications I., II., and III., in which the pH range matches the glass material eluate with B/Gd (Table 3).
Experimental methods➢ The corrosion test was executed for the PCV carbon steel SGV480 by using the testing solution containing dissolved absorber components (Table 4)
by conforming in principle to the testing methods stipulated in the rust prevention measures.
➢ The test was conducted at the dose rate of 4 kGy/h which was the same as that stipulated while deciding the rust prevention measures at the QST
Takasaki Advanced Radiation Research Institute's gamma ray irradiation facilities.
➢ An immersion corrosion test was carried out for uniform corrosion, and an electrochemical test was carried out for oxidation membrane type rust
inhibitor for localized corrosion (Classifications I. and II.).

Table 3 Composition and experimental evaluation of the testing solution
Clas
sific
ation

Rust inhibitor
Elements

Mechanism

AbsorberDensity ofartificial sea
water

Water quality

Experimental evaluation

B
Concentration
(ppm)

pH

Uniform
corrosion

Localized
corrosion

I.

Na2B10O16 4000 ppm(as B)

Oxidation
membrane

Glass material
with B/Gd

Diluted 1000
times

4000

8

Immersion
corrosion

Electrochemical

II.

Na2WO4 5000 ppm+Na2B10O16 1000 ppm

Oxidation
membrane

Glass material
with B/Gd

Diluted 1000
times

1000

8-9

Immersion
corrosion

Electrochemical

III.

ZSMMP 4000 ppm

Precipitation
membrane

Glass material
with B/Gd

Diluted 10000
times

0.6 - 0.8

7-8

Immersion
corrosion―ZSMMP: Zinc/Sodium carbonate mixed phosphate

Table 4 Dissolution components [ppm] from the absorber (Glass material with B/Gd)
B

Ba

Zn

Bi

La

Gd

0.6-0.8

0.4-0.6

0.1-0.2

0.07

0.01

0.01

©International Research Institute for Nuclear Decommissioning7. Implementation items of this project [3) (ii) ② Development of technology for criticality prevention] No.301b.

Study on the corrosion effects considering the use of rust inhibitors under irradiation effects[Test results]

⚫

1.2
20

15

10

5

0
0

ガンマ線
Gamma
ray
dose rate
線量率[kGy/h][kGy/h]Rustinhibitor
防錆剤

Fixed potential crevice corrosion
testing

1.0

Potential (V vs. SSE)
(V vs. SSE)
電位⚫

Uniform corrosion (Immersion corrosion test) (Figure 1)➢ The amount of corrosion decreased significantly with any of the rust prevention measures, and there was no effect of the
dissolved absorber components.
Localized corrosion (Electrochemical measurement and testing) (Figure 2)➢ When the dissolved absorber components were present in a gamma ray irradiated environment of dose rate 4 kGy/h, thenatural immersion potential increased and changed in the direction in which localized corrosion was likely to occur.
➢ When the progress of localized corrosion was evaluated by setting up conditions based on the measured potential value, it was
found that even in the presence of dissolved absorber components, the measured value of the natural immersion potential had
a margin of 0.45V or more and no progress in corrosion was seen.
Coexistence with rust prevention measures➢ It was confirmed that the dissolved components of glass material with B/Gd could coexist with the Classification I., II., and III.
rust prevention measures.

2)
Corrosionamount (mg/cm2)
腐食量(mg/cm

⚫

0.8

0

4

No
rust inhibitor
防錆剤なし

0.0

0.0

0.0

0

4

0

4

Classification
分類I. IClassification
分類II.II

There is no
progress〇

Potential measurement results

0.4
0.2

₋0.2

0.0

×

0.6

0.0

データ
No
data
なし

There is progress

Repassivation
potential for
crevice corrosion▲

Natural immersion
potential◆-0.2

ガンマ線線量率Gamma
ray dose rate[kGy/h][kGy/h]04Classification
分類III.III

Figure 1 Amount of corrosion based on immersion corrosion test (500 h)(presence of dissolved components of glass material with B/Gd)

Absorber
吸収材成分
Rust inhibitor
防錆剤

0

4
―

0

4

B・Gd入
Glass
material
ガラス材
with
B/Gd

Classification
I
分類I.

0

4
―

0

4

Glass
material
B・Gd入
with
B/Gd
ガラス材

Classification
分類II. II

Figure 2 Evaluation of progress of localized corrosion based on electrochemical test
#The data not containing absorber components is cited from the Corrosion Project results©International Research Institute for Nuclear Decommissioning8. Summary

No.302(1) Development of Fuel debris Retrieval method① Development of technology for removing interferences➢ A conceptual study was undertaken for the access routes and methods of carrying out structures with the
top-access method. Next fiscal year, the feasibility of the methods will be verified by drafting plans forelement tests and by conducting the tests.
➢ For the side-access method, the feasibility of the method was verified by conducting element tests for thecommon utilities.
② Development of various technologies other than removal technology for interferences➢ As regards the remote operation support method, an environmental model and simulator were created andthe robot model was used on the simulator to confirm that interferences can be avoided. Next fiscal year,
the effectiveness of the remote operation support methods will be verified by means of a partial mock-up.
➢ As regards the method of preventing the spread of contamination to the S/C, element tests were conducted
for building a weir remotely. The ability of building a weir by remote operation could be confirmed, butissues regarding the ability of stopping water were identified, hence next fiscal year, re-verification will be
carried out by means of additional testing.

➢ In order to reduce the load on the R/B floor, ideas on reducing the weight of the cells inside the R/B werestudied by optimizing the settings for the shield thickness required around the cell and reducing the cellheight by reviewing the access rail structure. Next fiscal year, the effectiveness and the feasibility of themethods studied will be evaluated.
➢ As regards the method of transporting the unit cans from the R/B, a conceptual study was conducted on
the confinement device (mobile cell) for transporting the fuel debris in unit cans. Next fiscal year, thefeasibility of the studied method will be verified by means of element tests.

©International Research Institute for Nuclear Decommissioning8. Summary

No.303(2) Development of Technology for Processing Fuel Debris(i) Technological development of fuel debris collection and storage systems➢ As regards the suction collecting system for particulate fuel debris, the shape of the pump strainer was
optimized and a pump model was selected according to the application environment. The verification will
be conducted by means of element tests next fiscal year.
➢ For the development of technology for transferring fuel debris and wastes, retrieval methods are being
planned presuming the state and characteristics of the fuel debris. Moreover, by carrying out processing
element tests for mechanical cutting, issues regarding cutting tools or cutting targets were identified. The
targets of application will be organized next fiscal year.
(ii) Technological development of treatment for fuel debris and deposit① Removal technology for soluble nuclides in circulating cooling water➢ The preliminary tests necessary for drafting the plan for adsorption tests using alpha nuclides (U, Pu,etc.) were partially executed, and the container adhesion effect and differences in the dissolution levels
were verified by adding Am, Cm, U to the solution having the actually presumed water quality. In future,adsorption testing will be conducted after carrying out similar preliminary tests for Np and Pu too.
➢

As a result of studying the deposition behavior of boric acid ions, the deposition risk due to the Ca ionsor the Na/B ratio changes was confirmed. Details are under study by means of the ongoing element tests.

② Treatment technology for deposits collected from inside PCV➢ A differential pressure countermeasures test of the intermediate removal filter which is a candidate
device for removal of insoluble nuclides was conducted, and issues and future policies were organized.
In the current scope of study (only changes in test conditions for intermediate removal), it was difficult
to control the rise in differential pressure during intermediate removal, so distribution of load to the final
treatment stage, or creating multiple intermediate removal systems, will be studied in future.
➢

As a technology for separating solid materials in collected liquids and wastewater, the applicability ofseparation using aggregation sedimentation was confirmed in a boric acid-free system. In future,operations or applicability verification testing (device testing) will be carried out after confirming theeffects of boric acid addition.

©International Research Institute for Nuclear Decommissioning8. Summary

No.304(2) Development of Technology for Processing Fuel Debris (Continued)(iii) Survey on sorting technology for fuel debris and radioactive wastes➢ As regards the sorting technology for fuel debris and radioactive wastes, the technical feasibility of eachsorting point was evaluated and sorting scenarios were formulated taking feasibility into account.
(3) Development of Technology for Ensuring Safety while Retrieving Fuel Debris(i) Development of element technology for confinement functions① Technology for prediction of dust behaviors in PCV➢ Advancement of realistic dust behavior evaluation method was implemented to conduct airflow analysis
for the containment vessel and entire R/B. The dust behavior evaluation method was established basedon the physical behavior of aerosol.
➢

The airflow analysis was conducted considering the decay heat of fuel debris and heat input during
processing. It was confirmed that there was no out leak due to the chimney effect under the assumed
negative pressure conditions.

② Technology for ensuring confinement functions in connection parts➢ As regards the method of connecting to PCV using the access tunnel, the basic requirements of theconnection parts are being organized by studying the work steps. In addition, a comparative study of theconnection methods was conducted. Next fiscal year, the feasibility of the method of connecting to PCV
by means of welding will be verified by element tests.

©International Research Institute for Nuclear Decommissioning8. Summary

No.305(3) Development of Technology for Ensuring Safety while Retrieving Fuel Debris (Continued)(ii) Development of element technology for prevention and monitoring of criticality① Technological development of criticality monitoring method➢ In order to verify the feasibility of the subcriticality measurements, a neutron measurement test was
carried out with the Kyoto University Criticality Assembly, and the gathered data is being analyzed.
➢

As regards the applicability of the neutron detector to the field, neutron detectors found to have goodprospects were selected from the perspective of subcriticality measurement. The structure of theneutron detector unit was designed and it was expected to be much lighter in weight than the earlier
ones. Moreover, the effect of a noise-resistant cable was verified after studying countermeasures fornoise generated by the processing devices.

➢

The monitoring methods and limiting conditions of criticality approach and the main points of attention
from the viewpoint of criticality safety control were consolidated by organizing the concepts of criticality
safety control involved at each step of each of the various fuel debris retrieval methods.

② Development of technology for criticality prevention➢ An insoluble absorber injection device was prototyped and tests were conducted and it was confirmedthat under water injection at the target speed was possible. Moreover, a test of processing with a chisel
was carried out after injection of absorber, and it was confirmed that the fuel debris and absorber get
mixed after processing.
➢

As regards the study on the corrosion effects considering the use of rust inhibitors under irradiation
effects, an irradiation test was conducted and it was confirmed that the rust inhibitor and glass material
with B/Gd can be used together.

©International Research Institute for Nuclear Decommissioning9. Specific Goals for Achieving the Purpose of the ProjectNo.306(1) Development of Fuel Debris Retrieval Method① Development of technology for removing interferences

As regards the methods to secure an access route by the top-access method (technology for・Top-access and transportation methods for removing

removing interferences, basic specifications of devices and procedures etc.), it should be possible

large-scale structures

to confirm the feasibility of the methods by means of element tests after conducting studies whichtake into account remote operations, maintenance or reduction of exposure, etc.

・ Side-access method for removing interferences
inside/outside of the pedestal(Target TRL at the end of the project: Level 3)
Element test should be completed and the criteria should be fulfilled at the plant with respect tothe methods for establishing utilities for the remote equipment, which influence the methods ofremoving interferences on the route, for the route inside the PCV from the equipment hatch to the

CRD opening. Or the issues to be solved and the action policies should be clear.
② Development of various technologies other thanremoval technology for interferences(Target TRL at the end of the project: Level 4)
By developing control methods to avoid interferences by automatic operation of remotecontrolled robots with an aim to decrease the work load on the operators and improve efficiency,

・Development of remote operation support method

the effectiveness and feasibility of applying these control methods must be evaluated by meansin environments with low visibility and narrowof simulator verification and element tests, etc., and the identification of issues and action

spacespolicies for the future should be presented.

・Method for preventing the spread of contamination

into the suppression chamber (S/C)(Target TRL at the end of the project: Level 3)
As regards the method of preventing the spread of contamination into the suppression chamber(S/C) using weirs, element tests should be completed and the criteria for the element testingshould be fulfilled at the plant, or the issues to be solved and the action policies should be clear.

・Establishment of the access route for cell installation(Target TRL at the end of the project: Level 4)
In order to reduce the load both inside and outside the R/B resulting from the installation of a cell,and conceptual study of reducing the impact on thethe method of downsizing the cell and the method of cell installation should be evaluated for theirR/B

effectiveness and feasibility, and the identification of issues and action policies for the future
should be presented.
(Target TRL at the end of the project: Level 3)©International Research Institute for Nuclear Decommissioning9. Specific Goals for Achieving the Purpose of the ProjectNo.307(1) Development of Fuel Debris Retrieval Method (Continued)② Development of various technologies other than

Conducting a conceptual study regarding the formulation of the safety design or equipment

removal technology for interferences

specifications, etc. concerning confinement, criticality prevention, or hydrogen generation

・A method of transporting the unit can from the R/B.

countermeasures etc. for the various transport devices used for transporting the unit cans storing
fuel debris from the R/B to another building within the premises, and confirming the feasibility ofthe methods or equipment by means of element tests of the confinement mechanism and thehydrogen treatment mechanism.
(Target TRL at the end of the project: Level 3)©International Research Institute for Nuclear Decommissioning9. Specific Goals for Achieving the Purpose of the ProjectNo.308(2) Development of Technology for Processing Fuel Debris(i) Technological development of fuel debris collection andstorage systems
・Development of suction collecting system for particulate
fuel debris

By building the proposed suction collecting system for particulate fuel debris, theeffectiveness of the suction speed of fuel debris and the ability of mounting the device on the
robot arm or the feasibility of the operation methods, etc. shall be evaluated by means ofelement tests, and the identification of issues and action policies for the future should bepresented.

・Process of transferring fuel debris and wastes(Target TRL at the end of the project: Level 3)
As regards the variety of fuel debris (slags, particulate, pellets etc.) presumed till date in the"Upgrading of the Comprehensive Identification of Conditions inside Reactor (FY2017)", byconducting a conceptual study on the necessary devices and series of collection procedures
such as cutting, container storage, transport etc. of fuel debris by accessing various locations
such as inside the reactor core, bottom of the reactor and the pedestal, the feasibility of themethods or devices should be verified by means of element tests, etc. of the processing or
handling devices.

(ii) Technological development of treatment for fuel debris
and deposit① Removal technology for soluble nuclides fromcirculating cooling water(Target TRL at the end of the project: Level 3)
As regards the soluble alpha nuclides removal facility and the boric acid conditioning facility,
the effectiveness and feasibility of the facilities must be verified by means of element tests, etc.

based on the results of the conceptual design.
(Target TRL at the end of the project: Level 3 to 4)②Treatment technology for deposits collected from inside
PCV

The effectiveness and feasibility of the results of the conceptual design of the wastewater
treatment equipment that is related to the separation of solids and liquids and the collection
and storage of solids must be verified by means of element tests, etc.

(iii) Survey on sorting technology for fuel debris andradioactive wastes(Target TRL at the end of the project: Level 4)
By formulating scenarios related to fuel debris sorting and conducting surveys of technology
necessary for sorting based on the proposed scenarios, the feasibility of the sorting scenarios

must be evaluated based on the results of technology survey considering the site environment,
and the issues must be identified.
(Target TRL at the end of the project: Level 2)©International Research Institute for Nuclear Decommissioning9. Specific Goals for Achieving the Purpose of the ProjectNo.309(3) Development of Technology for Ensuring Safety while Retrieving Fuel Debris(i)

Developmentofelementtechnology

for

By enhancing the analytical model for predicting the behavior of airflow and aerosol within theconfinement functions

boundaries related to the confinement function, more apt methods of evaluation must be studied.

① Technology for prediction of dust behaviors in

Moreover, the requirements for the boundary function (static) and the gas management function

PCV(dynamic) of PCV and R/B for ensuring the confinement function must be studied.
By predicting the aerosol behavior inside the R/B during times of dust leakage from the PCV, study
results regarding the monitoring positions etc. must be presented.

(Target TRL at the end of the project: Level 3)②Technology for ensuring confinement functions
in connection parts

As regards the structure and the connection methods of the connection parts between the accesstunnel and the PCV (equipment hatch), the criteria of the element test concerning the connection
between the access tunnel sleeve and the equipment hatch should be fulfilled, or the issues to besolved and the action policies should be clear.

(ii)

Developmentofelementtechnology

forprevention and monitoring of criticality① Technologicaldevelopmentof(Target TRL at the end of the project: Level 4)
The applicability of the candidate neutron detector for the measurement of subcriticality must be

confirmed.
criticality

monitoring method

The basic design must be carried out for the structure to be used for installing a neutron detector on
top of the fuel debris, and the issues and feasibility must be presented.
(Target TRL at the end of the project: Level 4)② Development of technology for criticality
prevention

The corrosion effect which is the secondary effect of an insoluble neutron absorber or the effectassociated with fuel debris processing must be confirmed.
The basic design must be developed for the device to be used for injecting an insoluble neutron
absorber, and the issues and feasibility must be presented.
(Target TRL at the end of the project: Level 4)©International Research Institute for Nuclear DecommissioningNo.310

Terminology (1/3)
No.

Terms

Definition

1

1F

Fukushima Daiichi Nuclear Power Station

2

R/B

Reactor building

3

PCV

Primary containment vessel

4

RPV

Reactor pressure vessel

5

CRD

Control rod drive

6

Operation floor

Operation floor

7

DSP

Device storage pool

8

SFP

Spent fuel pool9X-6 penetration

One of PCV piping openings

10

S/C

Suppression chamber

11

Jet deflector

Jet deflector

12

Inflate seal

Inflatable sealing material

13

Cell adapter

A member that connects the cell with the PCV

14

M/U

Mockup15

BSW

Biological shielding wall

16

MCCI

Molten core concrete interaction

17

UC

Unit can (fuel debris storage container)

18

AWJ

Abrasive waterjet

19

Heating Ventilating Handling Unit
(HVH)

Air conditioning unit

20

HCU

Hydraulic control unit

©International Research Institute for Nuclear DecommissioningNo.311

Terminology (2/3)
No.

Terms

Definition

21

CRGT

Control rod guide tube

22

JPDR

Japan Power Demonstration Reactor

23

MAAP

The modular accident analysis program owned by the Electric Research Institute (EPRI), U.S.

24

GOTHIC code

One of the general purpose thermal-hydraulic analysis code developed by EPRI/ZACHRY

25

Chimney effect

The movement of air leakage resulting from air circulation flow through multiple damaged openings
due to differences of air density in inside/outside PCV.

©International Research Institute for Nuclear DecommissioningNo.312

Terminology (3/3)
No.

Terms

Definition

1

TRL7

Practical application is complete.

2

TRL6

Onsite demonstration phase

3

TRL5

Actual-scale prototypes are produced, and validation tests are performed in a plantunder conditions that simulate the actual environment.

4

TRL4

Functional tests are performed using testing mock-ups as part of the development andengineering processes.

5

TRL3

Development and engineering work is performed within the range of conventional
experiences or their combination, or development and engineering work in new areas
virtually without past experience.

6

TRL2

Development and engineering work is performed, and the required specifications aredeveloped in areas where there is almost no applicable past experience.

7

TRL1

Basic requirements are identified for the methods and systems to be developed andengineered.

©International Research Institute for Nuclear Decommissioning