Subsidy Project of Decommissioning and Contaminated Water Managementin the FY2015 Supplementary Budgets

Development of Technology for Collection,
Transfer and Storage of Fuel Debris
FY2017 Research Report

June 2018
International Research Institute for Nuclear Decommissioning (IRID)
All Rights Reserved. International Research Institute for Nuclear Decommissioning©International Research Institute for Nuclear Decommissioning

No.1

CONTENTS
1. Research Background and Purposes
2. Project Goals
3. Implementation Items, Their Correlations, and Relations with Other Research Development
4. Schedule
5. Project Organization Chart
6. Implementation Details
6.1 Investigation and Establishment of Research Plans for Transfer and Storage
6.2 Study of Safety Requirements and Specifications of Storage Systems for theTransfer/Storage of Fuel Debris Canisters
6.3 Development of Safety Evaluation Methods and Validation
6.4 Study on Fuel Debris Collection Method
7. Overall Summary[Supplement-1] Position of Output©International Research Institute for Nuclear Decommissioning1. Research Background and Purposes

No.2

1.1 Background
Technology for collection, transfer and storage of fuel debris is required to retrieve fuel debris safety andefficiency for the decommissioning of the Fukushima Daiichi Nuclear Power Station (1F).
Fuel debris contains nuclear fuel, therefore it is necessary to consider, in particular, confinement ofradioactive materials (prevention of spreading contamination) and sub-criticality.
When the Three Mile Island Nuclear Power Generation Station Unit 2 (TMI-2), in the United States, wasdecommissioned, fuel debris was collected in a dedicated container (canister) and handled by the canister. Thisrationally fulfilled requirements, such as confinement of radioactive materials, by using existing technologies
for transfer and storage of spent fuel and management of radioactive waste. This example led us to believe that
it is reasonable to use existing technologies effectively by developing canisters to meet individual
circumstances. Based on the concepts, it has been decided to focus our development on canisters fordecommissioning the 1F.
The plant type of 1F is different from that of TMI-2. In addition, seawater was injected during the initial phaseof the accident, and the molten core reached the pedestal at the bottom of the reactor pressure vessel.
Therefore, the requirements for 1F canisters are more complex and advanced compared with TMI-2. To collect,transfer, and store fuel debris safely and rationally, a canister especially for 1F needs to be developed.
In previous studies, elemental technologies were evaluated and a basic canister design created. However,since there is limited information on 1F after the accident, it is important to lay down design conditions for thecanister and reflect updated knowledge as occasion arises to optimize 1F decommissioning (e.g., the amount offuel debris including high radiation waste).
1.2 Purpose
In this project, fuel debris canisters conditions and technology for handling the canister which are applicablefor the 1F have been developed. The information and requirements provided from the IRID projects to thisproject (input conditions) and provided from this project to the IRID projects (output conditions) are organized
and clarified by close cooperation with the related projects.
©International Research Institute for Nuclear Decommissioning1. Research Background and Purposes

No.3(Appendix) Compare to the precedent
The precedent of TMI-2 had completed to collect, transfer, and store fuel debris generated in the core meltdown accident.
This achievement is an excellent reference; however, it is different from 1F in the following points.
•

•
•

•

•
•

Fuel debris in 1F is distributed from the reactor pressure vessel to the pedestal inside the containment.
⇒ Products of concrete interactions (MCCI) and adherence to concrete during collection need to be considered.
✓ The alkaline component in concrete must also be taken into account.
Seawater was injected into the 1F reactor.
✓ Residual salt (chlorine) in fuel debris needs to be considered.
The working environment may be severe due to damage on the building. RPV and PCV may be damaged.
⇒ Collection of fuel debris by methods other than the submersion-top entry method, such as the partial submersionside entry method, are being considered.
✓ Studies appropriate for methods other than the submersion-top entry method, such as the partial submersion-side
entry method, (e.g., fuel debris canister design, know-how on handling of canister) are necessary.
Concentration is high.
⇒ The canister’s inner diameter will be smaller to maintain sub-criticality. It is difficult to place canisters side by side.
✓ Considerations on workability suitable for retrieval and collection of fuel debris and reduction in storage area areimportant.
⇒ Risk of re-criticality when retrieving fuel debris needs to be considered.
✓ The possibility of adding neutron-absorbing material to the fuel debris when retrieved also needs to beconsidered.
Burnup (source strength) is high.
✓ Measures against increasing hydrogen generated by water radiolysis are necessary.
Amount of fuel debris is large.
✓ Considerations on workability suitable for retrieval and collection of fuel debris and reduction in storage area areimportant.

©International Research Institute for Nuclear Decommissioning2. Project Goals

No.4

2.1 The overall goal of the project
Assuming the retrieval of fuel debris in 2021, our goal is to establish a method to collect, transfer, and store fueldebris safely and efficiently.
2.2 Goal of FY2018E
The project aims to develop a plan for how fuel debris will be collected, transferred, and stored safely and rationally.
The canister specifications, developed mainly from the perspective of safety design during FY2015 and 2016, will be
evaluated and improved with an eye to the fuel debris retrieval methods and transferring tasks. The canister
specification (prototype) will be established by performing verifications with tests.
(Remaining tasks)
As a result of the previous studies, the following tasks remain and must be taken on.
(1) Optimization in terms of safe and efficient collection, transfer, and storage
In the previous technical development, the issues were identified in terms of safety assessment by referring toexamples such as TMI-2. Study conditions were postulated based on expert opinions, and a provisional draft ofcanister specifications and multiple storage methods were proposed. However, in order to store the retrieved fueldebris smoothly, optimization in terms of treatment capacity and practical equipment is necessary.
In addition, rational measures against the possibility of conditions exceeding projections (e.g., fuel debris
characterizations) must be proposed from the aspects of management and equipment. (-> This will be studied in theimplementation items (1) and (2) on the next page.)(2) Safety evaluation with consideration given to the system
In the previous technical development, evaluations focusing on element tests were performed to round up the basic
canister design and the transfer/storage system. However, evaluation of the entire system is necessary. In addition,some of the tests require evaluations that will be added to previously acquired knowledge. (-> This will be studied in
the implementation item (3) on the next page.)(3) Management of fuel debris collected during water and gas treatment
In the previous technical development, block, granular, and powder fuel debris collected by the equipment weretargeted. However, as fuel debris is collected from water and gas treatment, conducted by the incidental equipment offuel debris retrieval, measures are required to store them. (-> This will be studied in the implementation item (4) on
the next page.)©International Research Institute for Nuclear Decommissioning3. Implementation Items, Their Correlations, and Relations with Other
Research Development

No.5

3.1 Implementation Item
The following verifications of safety and handling will be conducted from FY2019 with the canister prototype
test manufacturing for mockup tests and performing the mockup tests.

(1) Investigation and establishment of research plans for transfer and storage
The latest information on the situation, along with knowledge of related projects and sites, will be collected. Further
analysis of technical requirements related to overseas safety will be performed based on already obtained
information. These will be reflected in the research plan along with expert opinions in IRID. In addition, if further
information is required, additional investigation will be conducted, such as organizing a workshop with foreign
engineers.

(2) Study of safety requirements, specifications, and storage systems for the transfer/storage offuel debris canisters
Prerequisite technical requirements for the safety of transport casks and storage facility will be established totransfer and store fuel debris safely and rationally in canisters. In addition, each task’s processing capacity and otherrelated conditions will be evaluated and reflected in the requirement specifications for handling flow and thetransfer/storage system. It will also be reflected in the specifications of devices for handling canisters.

(3) Development of safety evaluation methods and safety validation
Based on the studies of (2) and (4), safety will be evaluated using a partial model of canisters and analysis, andfeasibility of the canister and transfer/storage system will be confirmed.

(4) Study on fuel debris collection method
Based on the study of (2), a storage method that is adapted to the fuel debris properties and retrieval method will be
formulated with a related technical development to transfer and store fuel debris safely and rationally in canisters.
The method will be reflected in the specifications of each part of the canister.
©International Research Institute for Nuclear Decommissioning3. Implementation Items, Their Correlations, and Relations with Other
Research Development

No.6

R&D Project(i) Establish collection, transfer, and storage flow (provisional)• Examples of work flow for damaged fuels in overseas (e.g.,
TMI-2)• Formulate flow proposals based on conditions of 1F

Revise as necessary

Results of internal PCV Survey
Results of internal RPV Survey
Results of fuel debris analysis
Set the basic conditions of fueldebris properties etc.
• Information on 1F, technicalliterature on TMI-2• Results of Fuel Debris
Characterization Project Team,etc.

(ii) Formulate the concept of the collection, transfer, and storage system(provisional)• Investigate foreign systems• Formulate concept of the storage system (wet and dry) based on conditions of 1F(iii) Establish the evaluation method and share of safety functions required incanisters and the system (provisional)• Survey systems developed in overseas for evaluation methods and share the safety functions• Develop methods for examining the applicability to 1F (identifying issues) and evaluating hydrogenetc., and propose measures(iv) Draw up the specifications based on thesafety assessment (provisional)• Specify properties based on foreign examples• Specify requirements based on parametric analysis

Set a range of basic
specifications for canisters(inner diameter, total height)
Organize the correlation with each
design factors
Organize the correlation of the basic
specifications for canisters with designsof other facility and equipment, work
efficiency, and approximate cost

Start of thissubsidized project*1: Related projects
Method Project Team, Fundamental
Technology Project Team, Fuel Debris
Characterization Project Team, Criticality
Control Project Team, Waste Project Team• Reflect study results and knowledge ofrelated projects*1 to canister specifications• Optimize the system specifications withconsiderations to safety, treatment
performance, etc.

Previous technicaldevelopments will
be reviewed andupdated with thelatest knowledge
and optimization.

Inspection of canister using partial
model and analyzation (evaluation on
function feasibility)

Finalize the canister’s basic
specifications
FY2019 onwards

Verification of the canister design’s validity(overall verification)©International Research Institute for Nuclear Decommissioning3. Implementation Items, Their Correlations, and Relations with
Other Research Development

No.7

3.2. Relation of Implementation Items (1/2)
Assuming the retrieval of fuel debris from the initial unit in FY2021, development will be pursued by the following schedule.
Phase 1

Phase 2
2014

Item/FY

2015

2016

2017

2018

2019

2020

2021

2013
(Mid)

(Early)Characterization using simulated fuel debris

Related projects(Late)Development of Technology for Fuel Debris Analysis/ Fuel Debris Characterization

Upgrading of Approach and Systems/ Upgrading of Fundamental Technology for Retrieval of Fuel Debris and Internal
Structures

Development of Technology for Criticality Control Methods

1. Investigate and establish researchplans for transfer and storage ofdamaged fuel2. Study on the fuel debris storage
system
3. Development of safety assessment
technology
4. Development of fuel debris storage
technology
5. Develop canister transfer and storage
technology(FY2017 to FY2018)
1. Investigate and establish researchplans for transfer and storage
2. Study safety requirements,
specifications, and storage systems
for the transfer/storage of fuel debris
canisters
3. Develop safety evaluation methods
and safety validation

4. Study on fuel debris collection method

Coordination with related projects
Input: Composition, chemical properties, thermal
properties, and handling restrictions of fuel debris; subcriticality management method, etc.
Output: Concepts of canister (including design concept)
etc.

▼
Retrieval of fuel debris(initial unit)

FY2015 to FY2016:
Laid down basic specifications of thecanister, the mockup test, and handlingdevices.

FY2017 to FY2018:
Dealt with basic specifications of canisters andhandling devices, optimizing them for
operation. Clarified safety requirements of thetransfer/storage system. Evaluated safety.

Study on the canister storage system
Safety analysis, material selection
Development anddesign of canister

Mockup test*1

Study on canister
for mockup

Development and design ofhandling device

Reflection of the latest
knowledge of related projects
Clarification of safety
requirements, optimization of
Study oncanister specifications
handling device
Proposal of action policy for
issues in previous
Safety evaluation
developmentof canistersCompletion of canister designfor mockup test
Test
manufacturing

Optimize canister specifications Evaluate thebased on knowledge of related canister handling
projects

Final verification test

Test
Mockup test (conducted in *1)
properties
manufacturing©International Research Institute for Nuclear Decommissioning

The project’s scope of implementation

3. Implementation Items, Their Correlations, and Relations with Other
Research Development

No.8

3.2. Relation of Implementation Items (2/2)
International Research Institute for
Nuclear Decommissioning (IRID)
Fuel Debris Characterization Project[Information provided to the Canister ProjectTeam]
Knowledge of fuel debris shape
Knowledge of fuel debris properties
Knowledge of basic physical properties of fueldebrisKnowledge of thermal properties of fuel debris
Knowledge of chemical properties of fueldebrisWaste Management Project[Information provided to the Canister
Project Team]
Knowledge on waste sorting, etc.

Canister Project[Information provided to related project
teams](1) Studies of safety requirements,
specifications, and storage systems forthe transfer/storage of fuel debris
canisters• Details of inspections for safetyassessment(4) Study of fuel debris collection method• Canister specifications• Conditions of structure, weight,
dimensions, etc.
• Basic canister handling method• Canister handling device specifications*• Handling method of canister handlingdevices**Canister lid closing device, canister hoisting device, etc.

Note)
Method Project Team: Project on fuel debris retrieval method and system
Fundamental Technology Project Team : Project on fundamental technologies for fueldebris retrieval
Debris Characterization Project Team : Project on fuel debris characterization
Criticality Control Project Team: Project on fuel debris criticality control
Waste Project Team: Project on treatment and disposal of solid waste
Canister Project Team: Project on collection, transfer, and storage of fueldebrisFuel Debris Retrieval Method Project,
Fundamental Technology Project[Information provided to the Canister Project
Team]
Limiting conditions for handling fuel debris• Dimension conditions and weight
limitations
Storage location and procedure for fueldebris• Canister installation location• Canister hoisting information
Fuel debris collection method• Knowledge of fuel debris cutting, particle
size, etc.
• Collection method (e.g., pump suction)
Handling procedure of special tools

Criticality Control Project[Information provided to the Canister
Project Team]
Limiting conditions to maintain subcriticality
Neutron-absorbing material used tomaintain sub-criticality

Consistent results are obtained by sharing information provided from related projects of IRID and informationdelivered from this project, working in cooperation with those projects, and making adjustments.
©International Research Institute for Nuclear Decommissioning4. Schedule
Development of Technology for Collection, Transfer and Storage of Fuel Debris

No.9

FY2017
April

May

June

July

AugustSeptember

October

Analyzation of other R&D results and evaluation of foreign safety requirements

1

Investigation and establishment of researchplans for transfer and storage

2

Changes in safety requirements andspecifications related to transfer and storageof fuel debris canisters(1) Establishment of safety requirements
and specifications related to transfer andstorageNovember

December

January

FebruaryMarchRefinement of plan

Study on safety requirements
Reflection of study results (method, etc.)

System optimization (flow, etc.)

Organization of technical requirements for wet and dry storage methods (continues in FY2018)(2) Study of storage system(3) Study of storage method

Study of conditions

Preparation of model(4) Study of drying system(5) Study and review of specifications forhandling device
3

Development of safety evaluation methods
and safety validation(1) Safety evaluation of sub-criticality

Study of drying procedure

Reflection of study results(method, etc.)

Inspection of water volume measurement technology

Study of applicability of water volume restriction

Survey of sub-criticality conditions
Formulation of requirements for lids etc.

Preparation of materials for lid structure specimen (used in FY2018 test)
Formulation of test plan using real size canisters(2) Safety evaluation of structural strength
Study of stress corrosion cracking (SCC) conditions

Study of effect upon SCC(3) Safety evaluation regarding aging
degradation

Optimization of corrosion measures including operation (continues in FY2018)
Inspection of gas phase reaction(4) Safety evaluation of measures against
hydrogen gas
Verification of the amount generated

4

Gamma-ray irradiation test
Study of alpha-ray effectHydrogen evaluation test using spent fuel (continues in FY2018)

Study of countermeasures from an operational perspective(5) Verification of measures against
hydrogen

Study of flow characteristics inside canister

Study of fuel debris collection method(1) Study of the canister’s specification thatsuits fuel debris properties

Study of canisters that suit fuel debris properties

Revision of canister specifications (continues in FY2018)(2) Review of canister design

Presentation at academic conferences▼

Briefing▼

Briefing▼

Briefing, presentation, etc.
©International Research Institute for Nuclear Decommissioning5. Project Organization Chart (as of the end of March 2018)
External stakeholder

No.10

International Research Institute for Nuclear
Decommissioning (Head Office)Tokyo Electric Power Company Holdings, Inc.

 Formulation of overall plan and technical management
 Technical management, including technical development progress

Mitsubishi Heavy
Industries, Ltd.

Toshiba Energy Systems &Solutions Corporation

Relevant projects

Hitachi-GE Nuclear
Energy, Ltd.

Development of Technology for Fuel Debris Analysis/
Fuel Debris Characterization

Upgrading of Approach and Systems for Retrieval of
Fuel Debris and Internal Structures(1) Investigation and establishment of research plans for transfer and storage(2) Study of safety requirements, specifications, and storage systems for the transfer/storage of fuel debris canisters

Upgrading of Fundamental Technology for
Retrieval of Fuel Debris and Internal Structures(3) Development of safety evaluation methods and safety validation

Development of Technology for Criticality Control
Methods(4) Study of fuel debris collection method

R&D for Treatment and Disposal of Solid
Radioactive Waste

NDC• Development of MCCI products and physicalproperties of fuel debris onto design parameter• Study on measures for fuel debris against
hydrogen generation

MHI-NS• Study on basic specifications of canisters(including analysis), preparation of designand drawing of canisters and handlingdeviceSoltec• Tests related to heat transfer and heatflow inside canister

Fuji Electric Co., Ltd.
• Study on specifications for transfer andstorage of fuel debris canisters

MPR Associates, Inc.
• Study on specifications for transfer andstorage of fuel debris canisters

EIWA Corporation, Nippon Swagelok FST, Inc.
• Consumables used in tests on evaluation of
the amount of hydrogen generated

KOKUYO Co.,Ltd.
• Organization of documents and data

NDC: Nuclear Development Corporation

GNF• Study on concepts related to the storage
systemMHI-NS: MHI-NS Engineering Co., Ltd.

Hitachi Power Solutions Co., Ltd.
• Verification of feasibility of canister lid
structure

Soltec: MHI Solution Technologies Co., Ltd.

GNF: Global Nuclear Fuel Japan Co., Ltd.
TOKO Corporation• Study on method of handling canister lids
Hitachi Power Solutions Co., Ltd.
• Analyzation of specimen for tests to evaluate
the amount of hydrogen generated

GNF• Study on evaluation on the amount ofhydrogen generated using spent fuelPacific Northwest National Laboratory• Study on the effect of alpha-ray on theevaluation on the amount of hydrogengenerated©International Research Institute for Nuclear DecommissioningNo.11

6. Implementation Details
6.1 Investigation and Establishment of Research Plans for Transfer and Storage(1) Gathering of the latest knowledge on other technical development(2) Analyzation on foreign knowledge(3) Formulation of research plan
6.2Study of Safety Requirements and Specifications of Storage Systems for theTransfer/Storage of Fuel Debris Canisters(1) Establishment of Safety Requirements and Specifications Regarding Transfer andStorage of Fuel Debris Canister(2) Study of storage system(3) Study of storage method(4) Study of drying system(5) Study and review of specifications for handling device
6.3 Development of Safety Evaluation ,Methods and Validation(1) Safety evaluation of sub-criticality(2) Safety evaluation of structural strength: Study of lid structure(3) Safety evaluation regarding aging degradation: Study of canister materials(4) Safety evaluation of measures against hydrogen gas
a. Verification of the amount generated
b. Verification of measures against hydrogen
6.4 Study on Fuel Debris Collection Method(1) Study on the canister’s specifications that suit fuel debris properties(2) Review of canister design©International Research Institute for Nuclear DecommissioningNo.12

6. Implementation Details
6.1 Investigation and Establishment of Research Plans for Transfer and Storage(1) Acquiring the latest knowledge on related technical development
On formulating the flow from retrieval to storage of fuel debris in 6.2 “Study of safety requirements,
specifications, and storage systems for the transfer/storage of fuel debris canisters,” information was
exchanged in a joint meeting among the Method Project Team, Fundamental Technology Project Team, Waste
Project Team, and experts of material accountancy from outside IRID. The latest knowledge (limiting
conditions of the fuel debris retrieval method, amount of fuel debris, material accountancy and control policy
for fuel debris, etc.) was gathered and reflected in the workflow.
(2) Analysis on knowledge obtained from overseas
The TMI-2 case was referenced in establishing safety requirements of 6.2 “Study on safety requirements andspecifications and storage systems for the transfer/storage of fuel debris canisters.”
In addition, on studying the shape of canister lids and other specifications in 6.4 “Study on fuel debris
collection method,” the approval application form of a domestic example (waste management facility of Japan
Nuclear Fuel Limited reprocessing business) was obtained and referred.

(3) Formulation of research plan
The 4 “Schedule” was formulated as a research plan. Additionally, the 6.2 (2) “Study of the storage system”
and individual topics were discussed to revise based on experts.

©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.13
6.2 Study of Safety Requirements and Specifications of Storage Systems for the Transfer/Storage ofFuel Debris Canisters(1) Establishment of safety requirements and specifications regarding transfer and storage of fuel debris canister(i) Purpose and overview
In terms of clarifying the requirements for designing the fuel debris transfer/storage system, safety requirement proposals were formulated in
the following terms based on the safety requirement proposals of fuel debris retrieval that share the same basic safety principles. In formulating the
proposal, the Act on the Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactors was referred to, as well as requirements in the related
laws concerning handling and storage of spent fuels.
• Basic requirements
Have function necessary to ensure safety of the facility (e.g.,preventive measure against leakage of radioactive materials)

Act on the Regulation of Nuclear Source Material, Nuclear Fuel Material andReactors• External event and human event
Prevent human caused events (e.g., trespass) and damage caused byexternal event (e.g., earthquake, tsunami)

Regulations for Transport of Nuclear Fuel Material Outside Plants• General design requirements for design basis facility
Give considerations to design in safety facilities (e.g., environmental
conditions, reliability), prevent operational error, prepare countermeasureequipment for station blackout, ensure operation during abnormality, etc.

The NRA Ordinance on Technical Standards for the Capabilities ofSpent Fuel Interim Storage Facilities• Individual design requirements for target facility
Prevent leakage of radioactive materials, retain integrity of the canister andcanister handling facility, criticality control, cool fuel debris, prevent fireand explosion, etc.

• Worker exposure

Reduce exposure dose as much as
possible/keep as low as possible• Public exposure• Transfer of canister and storage task
Manage so that works and operation that deviate from design limitations
are not conducted• Prevention of escalation in design basis accident (DBA)
Identify accident event that triggers leakage of radioactive materials andre-criticality, prevent escalation when the event occurs, and takeappropriate measures that satisfy the criteria• Laws and regulations(ii) Future plans
The task is to create specific design requirements from the safety
requirement proposals formulated here. The evaluation of the fuel debris
retrieval safety requirements will be coordinated with and the requirementsof spent fuel storage facilities considered.

The NRA Ordinance on Standards for the Location, Structure andEquipment of Spent Fuel Interim Storage Facilities
The NRA Ordinance on Standards for the Location, Structure, andEquipment of Commercial Power Reactors

Ordinance on the Security of Nuclear Reactor Facilities andProtection of Specific Nuclear Fuel Material at the Fukushima
Daiichi Nuclear Power Station
The NRA Ordinance on Technical Standards for the Design andConstruction Methods of Spent Fuel Interim Storage Facilities
Figure: Laws and regulations particularly referred to in the study©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.14

6.2 Study of Safety Requirements and Specifications of Storage Systems for the Transfer/Storage ofFuel Debris Canisters(2) Study on the storage system (1/13)(i) Purpose
The focus of our study was the safety assessment methods that are necessary to design the transfer/storage
system. However, to materialize these facilities, it is necessary to consider the demands and requirementsfrom another point of view, including the operation of throughput. To fulfill these demands and requests, aprovisional process flow up to the storage of fuel debris was laid down, based on expert opinions about
related projects in IRID. In addition, the quantity which will be the premise of the throughput study is in theprocess of being studied.
(ii) Future plans
The validity of the system will be evaluated by related projects based on the provisional flow and quantity.
Requirements for the canisters and transfer/storage system will be made concrete.

©International Research Institute for Nuclear Decommissioning6. Implementation Details
6.2 Study of Safety Requirements and Specifications of Storage Systems for the Transfer/Storage ofFuel Debris Canisters(2) Study on the storage system (2/13)

No.15

Process flow establishment
On establishing a process flow, the basic concepts of boundaries (confinement boundary) and share of safety functions werereconfirmed.
(i) Sub-criticality(ii) Heat removal(iii) Shielding (double
layer)(iv) Confinement(double layer)

Reactor building(shielding,confinement)

Reactorbuilding
Inside the canister (contains UC*1 with water
adhesion)(sub-criticality, heat removal,
confinement (*in terms of fuel debris discharge))

New building(exterior of storage canister
handling cell and carry out cell)(shielding, confinement)

Load into transport cask
(sub-criticality, heat removal,
shielding, confinement (double
lid))
Carry out cell

Maintenancecell

Fuel debris retrieval
storage cell

Hatch foraccess(additional)

Fuel debris carry out
building

Send out in a transport cask
(sub-criticality, heat removal,
shielding, confinement (double
lid))

Storage canister
handling cell

Fuel debris retrieval
storage cell(shielding, confinement)

Storage facility

Place fuel debris in
UC*1 (sub-criticality)

Fuel debris
retrievalStorage in canister

Storage canister
handling cell(shielding, confinement)

Cleaning, etc.
of canister

Carry out cell(shielding, confinement)

Storage of canisterin transport cask

*1: unit canister (UC) is short for unit can. It is acontainer to put fuel debris, which can be brought
into confined spaces in the primary containment
vessel. The can has a cylindrical shape and amesh structure on the bottom and side. Several
unit cans can be stored in a canister.

Send out transport
cask

Figure: Canister management in 1F (example of partial submersion©International Research Institute for Nuclear Decommissioningside entry method)

6. Implementation Details
6.2 Study of Safety Requirements and Specifications of Storage Systems for the Transfer/Storage ofFuel Debris Canisters(2) Study on the storage system (3/13)

No.16

Process flow setting
Based on the concepts of the basic system safety design as previously mentioned, a provisional flow was estimated byexchanging opinions among the Fuel Debris Retrieval Method Project Team, the Waste Project Team, and IRID experts to ensure
safety. Several flows that would be necessary in processes such as sorting and wet (pool) storage were created in parallel.
[Fuel debris retrieval process][1] Accept empty
canisterEmpty canister[2] Open lid

Scenario branching A(use unit can)Canister lid[3] Take out the empty
unit canEmpty the canister(lid open)

Empty the unit can[4] Carry into thepedestal
Scenario branching B(sort discharge)[5] Cut andprocess

Discharge[6] Sort

Fuel debris[7] Collect and fill

Waste

[9] Bring out of thepedestal[8] Drain

Canister (lid open)

Unit canCanister (lid open)

Go to the waste treatment
line
No.

[1]

[2]

Handled item

Empty canister

Canister lid

Details andpurpose of task

Carry an empty
canister in which
fuel debris will be
stored in thespecified positionin the fuel debris
retrieval storage
cell.

Open the lid of
the empty
canister before
collecting fueldebris.

[3]
Empty the unit
can

Empty
canister[4]
Empty the unit
can

[5]

[6]

[7]

[8]

Discharge

Discharge

Fuel debris

Unit canCut and process
To place and store discharges,
From the empty
the discharge into Sort the cut andbring the empty canister orcanister, take out
a shape andprocessed
unit can from the fuel debris
an empty unit candimension thatdischarge into
retrieval storage cell into thefor storing fuelwill fit in thefuel debris andpedestal and have it stand bydebris.
canister or unit
waste.
in a predetermined position.
can.

Collect the fueldebris and place
it in the empty
canister or unit
can.

[9]

Canister

Unit canReduce entrainedwater whencollecting andplacing fuel debris
Transport the canister or unit
for purposes
can filled with fuel debris toincluding reduction the fuel debris retrievalof the amount ofstorage cell that is outside ofhydrogen generated the pedestal.
and relaxation of theinitial drying
treatment conditions.

Scenario branching A[Details]“Bring the canister
directly into thepedestal” or “bring theunit can into thepedestal”

Scenario branching B[Details]“Sort the discharge into
fuel debris and waste at
the site, depending onwhether criticality may
occur or not” or “sort thedischarge in advance into
fuel debris and waste in
the area where thedischarge is located andnot at the site of retrieval”

Figure: Example of process flow: Retrieving fuel debris ~ Storing in the unit can
Research Institute for Nuclear Decommissioning(Details are studied by the Fuel Debris Retrieval Method Project©International
team.)

6. Implementation Details

No.17

6.2 . Study of Safety Requirements and Specifications of Storage Systems for the Transfer/Storageof Fuel Debris Canisters(2) Study on the storage system (4/13)[Collecting process]

Scenario branching C(perform drying treatment)

Unit canCanister (lid open)[10] Dry[14] Close lid[11] Load

Canister (lidopen)[12] Close lid[13] Confirm contamination on
surface, decontaminate[15] Wash surface

Canister lid

No.
Handled item

[10]
Unit can[11]
Unit can[12]
CanisterDry the entrainedwater and reduce theamount of water at an
early stage before theunit can is stored in
Details andLoad the unit can into Close the lid of thethe canister to reduce
purpose of task
the canister.
canister.
the risk of criticality(during arrangement),risk of corrosion, theamount of hydrogengenerated, etc.

[13]
Canister[14]
Canister

Confirm forcontamination on thesurface of thecleaned canister.
Decontaminate if
Close the lid of thecontamination iscanister.
confirmed. Cut andprocess into an
appropriate
dimension.

[15]
Canister

Scenario branching C[Details]“Perform drying treatment on theunit can” or “no drying treatment”

Wash the surface ofthe canister toprevent the spread ofcontamination.

Figure: Example of process flow: Storing in the unit can ~ Closing lid of canister can©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.18

6.2. Study of Safety Requirements and Specifications of Storage Systems for the Transfer/Storage ofFuel Debris Canisters(2) Study on the storage system (5/13)[Process before sending out the canister]
Scenario branching D(perform drying
treatment)

Canister[16] Dry[17] Confirm
carry out[18] Measure amountof hydrogen generated

Scenario branching E(use transport cask)[22] Accept empty hermetic
container

Empty the hermetic
container[23] Open lid

Empty the hermetic
container (lid open)[19] Load[20] Close the lid

[16]

[17]

[18]

[19]

Handled item

CanisterCanisterCanisterCanister

Details andpurpose of task

Measure theOn carrying out the amount of
Dry the entrained canister to thehydrogenwater and reduce storage facility,
generated tothe amount ofconfirm the canisters confirm that thewater to reduce
to collect and linkhydrogen
Load thethe risk ofdata and conditions concentration
canister into
criticality (during of each canisters
inside the canister
the hermetic
arrangement), risk (mass measurement, will not reach thecontainer.
of corrosion, the surface dose rate
lower explosion
amount ofmeasurement, visual limit within ahydrogeninspection (including specified time
generated, etc.
verification ofwhen transferringcanister ID)).
to the storage
facility.

[20]
Hermetic
container lid

Note: Suppose that only one canister isstored in the hermetic container

Hermetic container(lid open)

Hermetic container lid

No.

Canister[21]
Hermetic
container[22]
Empty thehermetic container[23]
Hermetic
container lid

To ensure thecontainer’s fueldebrisAfter loading the confinement
canister, confirm property (including Open the lid offor contamination gas) at an early
the empty
Close the lid ofon the hermetic
stage, carry in the hermetic
the hermetic
container surface. canister to acontainer
container.
Decontaminate if specified position before sealing
contamination is inside the empty in the canister.
confirmed.
hermetic canister
which the canister
will be sealed
inside.

[21] Confirm for contamination
on surface, decontaminate

Scenario branching D[Details]“Perform drying treatment
before carrying out thecanister” or “no drying
treatment”

Scenario branching E[Details]“Store the canister in a dual container(hermetic canister + shielded container)that share the functions between the two”
or “store the canister in the transport cask
only (double lid)”

Figure: Example of process flow: Confirming of the closed canister ~ Preparation of transferring the canister©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.19

6.2. Study of Safety Requirements and Specifications of Storage Systems for the Transfer/Storage ofFuel Debris Canisters(2) Study on the storage system (6/13)[Send out process][24] Accept the empty transport
cask/empty the shielded container
Empty the transport cask/
empty the shielded container[25] Open lid

Lid

Empty the transport cask (lid open)/empty theshielded container (lid open)[26] Load

CanisterTransport the cask (lid open)/shieldedcontainer (lid open)

Hermetic
containerNo.
Handled item

Details andpurpose of task

[24]
Empty thetransport cask

[25]
Empty theshieldedcontainer

Transport cask
lid

[26]
Shieldedcontainer lid

Open the lid of the container
On transferring thebefore storing thecanister/hermetic container tocanister/hermetic canisters inthe storage facility, carry in the the transport cask/shieldedtransfer cask/shielded container container.
for external transportation to a (Open both the primary andspecified position.
secondary lids of the transportcask.)

Canister

Shieldedcontainer[29] Confirm sendoutTransport cask

[28] Measure the amountof hydrogen generated[27] Close lid[27]
Hermetic
container

Transport cask
lid

[28]
Shieldedcontainer lid

Once the specified number ofcanisters/hermetic containers
Load a specified number ofare loaded, close the lid of thecanisters/hermetic containers in transport cask/shielded
the transport cask/shieldedcontainer.
container.
(Close both the primary andsecondary lids of the transportcask.)[30] Load

Transport
vehicle[29]

Transport cask Transport cask

[30]
Shieldedcontainer

On sending out
the container,
measure theamount of
On transporting the containerhydrogeninside the premises of 1F,
generated per confirm the containers for any
transport cask problem (confinement
and confirm
confirmation, inspection forthat thecontamination on surface,hydrogendecontamination, surface doseconcentration
rate measurement, visual
will not reachinspection (including
the lower
confirmation of canister ID)).
explosion limit
within thespecified time.

Transport cask

Shieldedcontainer

Load and secure the transport
cask/shielded container onto
the transport vehicle.

Figure: Example of process flow: Loading the canister ~ Preparation to sending out the transport cask
©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.20

6.2. Study of Safety Requirements and Specifications of Storage Systems for the Transfer/Storage ofFuel Debris Canisters(2) Study on the storage system (7/13)[Dry storage and acceptance process][Transfer process inside building]

Scenario branchingF(perform wet storage)

Transport
vehicle[31] Accept[32] Unload

Shieldedcontainer[33] Confirm
acceptanceTransport cask

No.
Handled item

Details andpurpose of task

[32]

[31]
Transport cask

Shieldedcontainer

Transport cask

[34]

[33]
Shieldedcontainer

Transport cask

Shieldedcontainer

Confirm the accepted transfer
casks/shielded containers for
Accept the transport vehicle with Unload the transport
any problem (inspection for
the transport cask/shielded
cask/shielded container from the contamination on the surface,container into the dry storagetransport vehicle to a specified decontamination, surface dose
facility.
location.
rate measurement, visualinspection (including
confirmation of canister ID)).

Transport cask[35] Open lid[36] Take out
canister/hermetic
container

Canister[34] Measure hydrogenconcentration

Hermetic
container[35]
Transport cask
Shieldedlid
container lid[36]
Canister

Hermetic
container

On opening thetransport cask lid,measure the amountof hydrogen
Open the lid of the transport
Take out the canister/hermetic
generated inside the cask/shielded container. (Opencontainer from the transport
cask and confirm the both the primary and secondary
cask/shielded container.
hydrogenlids of the transport cask. )concentration has notreach the lower
explosion limit.

Scenario branching F[Details]“Perform temporary wet
storage” or “no wet
storage”

Figure: Example of process flow: Transfer on site~ Receiving canister into dry storage facility ~ taking out canister©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.21

6.2. Study of Safety Requirements and Specifications of Storage Systems for the Transfer/Storage ofFuel Debris Canisters(2) Study on the storage system (8/13)[Dry storage process and process before storage][37] Confirm
acceptance

Canister

Scenario branching G(go through wet
storage)[39] Dry[40] Confirm before
storage

Canister[38] Drain[42] Measure hydrogenconcentration
Hermetic
containerNo.
Handled item

Details andpurpose of task

[43] Open lid[44] Take out
canisterCanister[41] Confirm contamination on
surface, decontaminate

[37]

[38]

[39]

[40]

[41]

CanisterCanisterCanisterCanister

Hermetic container

[42]
[43]
[44]
Hermetic
Hermetic
Canistercontainer
container lid
On opening the lid
On storing the canister,of the hermetic
Confirm the accepted
confirm the canister for
After the canister has
container,
containers for any
any problem (massgone through wetmeasure theproblem (inspection formeasurement,
Confirm forstorage, drain residual
amount ofsurface contamination,
Dry out the water inside inspection forcontamination on thewater from the canisterhydrogenOpen the lid of the Take the canister
decontamination, mass
the canister to reduce contamination on
surface of the accepted
through the drain pipe
generated inside hermetic container out from themeasurement, surface
the amount of hydrogen surface,hermetic container.
as much as possible tothe container and to take out thehermeticdose rate
generated, risk of
decontamination,
Decontaminate if
relax the initialconfirm thecanister.
container.
measurement, visualcorrosion, etc.
surface dose ratecontamination isconditions of dryinghydrogeninspection (including
measurement, visualconfirmed.
treatment as much asconcentration has
confirmation of canister
inspection (including
possible.
not reach theID)).
confirmation of canisterlower explosion
ID)).
limit.

Scenario branching G[Details]“Go through wet
storage” or “no wet
storage”

Figure: Example of process flow: Confirmation for acceptance of canister ~ Confirmation before storage©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.22

6.2. Study of Safety Requirements and Specifications of Storage Systems for the Transfer/Storageof Fuel Debris Canisters(2) Study on the storage system (9/13)[Dry storage process and process before storage][45] Accept empty metallic cask

Empty metallic cask[46] Open lid

Lid

Empty metallic cask (lid open)

Canister[47] Load

Metallic cask (lid open)[48] Close lid

Metallic cask

[49] Confirm confinement[50] Connect vent piping[51] Confirm confinement of ventpiping[52] Long-term storage

No.
Handled item

[45]
Empty metallic cask

[46]
Metallic cask lid[47]
Canister[48]
Metallic cask lid

Details andpurpose of task

Carry in an empty
metallic cask inwhich the canister
will be stored for along period to thespecified position.

Open the lid of theempty metallic cask
before storing
canisters.

Load a specified
number of canisters
inside the empty
metallic cask.

Once the specified
number of canisters
is loaded, close thelid of the metallic
cask.

[49]
Metallic cask

[50]
Metallic cask

Connect the ventConfirm that thepiping to the metallic
metallic cask
cask. The piping
satisfies thereleases gas insideprescribed
the metallic cask
confinement function. outside in a
managed manner.

[51]
Metallic cask

[52]
Metallic cask

Confirm that the ventpipe connected to
Store the canister inside
the metallic cask
the metallic cask in asatisfies thestable state for a longprescribedperiod of time.
confinement function.

Figure: Example of process flow: Preparation for dry storage ~ Dry storage©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.23

6.2. Study of Safety Requirements and Specifications of Storage Systems for the Transfer/Storageof Fuel Debris Canisters(2) Study on the storage system (10/13)[Transfer processinside building][Wet storage and acceptanceprocess]

Transport
vehicleNo.

[53] Accept[54] Unload

[54]

[53]
Shieldedcontainer

Transport cask[55] Confirm
acceptance

Shieldedcontainer

Transport cask[56] Measure hydrogenconcentration

Transport cask

Shieldedcontainer[58] Take out

Canister

Hermetic
container

[56]

[55]
Shieldedcontainer[57] Open lid

Transport cask

[57]
Transport cask
Shieldedlid
container lid

Handled item

Transport cask

Details andpurpose of task

On opening the lid of
the transport cask,
Confirm the accepted transfer
measure the amountcasks/shielded containers for any
Accept the transport vehicle with Unload the transportof hydrogen generated Open the lid of the transport
problem (inspection for surface
the transport cask/shielded
cask/shielded container from theinside the transport
cask/shielded container. (Opencontamination, decontamination,
container into the wet storagetransport vehicle to a specified
cask and confirm that both the primary and secondary
surface dose rate measurement,
facility.
location.
hydrogenlids of the transport cask.)
visual inspection (including
concentration has notconfirmation of canister ID)).
reach the lower
explosion limit.

[58]
Canister

Hermetic
container

Take out the canister/hermetic
container from the transport
cask/shielded container.

Figure: Example of process flow: Transfer inside the premises ~ Receiving the canister in wet (pool) storage facility ~ Taking out
the canister©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.24

6.2. Study of Safety Requirements and Specifications of Storage Systems for the Transfer/Storage ofFuel Debris Canisters(2) Study on the storage system (11/13)[Wet storage and temporary storage process][Wet storage process and process before temporary storage][59] Confirm
acceptance

Canister[63] Measure hydrogenconcentration
Hermetic
containerNo.
Handled item

Details andpurpose of task[64] Open lid[65] Take out
canister[60] Inject water into
canister[61] Store in pool

Canister[62] Confirm contamination
on surface, decontaminate

[59]
[60]
CanisterCanister
Confirm the accepted
containers for any
problem (inspection forsurface contamination,
On putting thedecontamination, mass containers in wetmeasurement, surface
storage, inject water indose rate measurement, the canister.
visual inspection(including confirmation ofcanister ID)).

[61]
Canister[62]
Hermetic container

Confirm forTransfer and settle thecontamination on thecanister to a specified
surface of the accepted
position in the pool.
hermetic container.
(Depending on the form of
Decontaminate if
canister storage, connect acontamination isvent pipe to the canister.)confirmed.

[63]
Hermetic container[64]
Hermetic container lid

On opening the lid of thehermetic container,
measure the amount ofOpen the lid of thehydrogen generated inside
hermetic container tothe container and confirm
take out the canister.
the hydrogen concentration
has not reach the lower
explosion limit.

[65]
CanisterTake the canister out
from the hermetic
container.

Figure: Example of process flow: Preparation of wet storage ~ Wet storage©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.25

6.2. Study of Safety Requirements and Specifications of Storage Systems for the Transfer/Storageof Fuel Debris Canisters(2) Study on the storage system (12/13)[Wet storage process and process before sending out]

Canister[62] Take out[63] Drain[64] Wash surface

Scenario branching H(perform drying treatment on fueldebris)[65] Dry surface[67] Confirm sendout[68] Measure amount ofhydrogen generated[66] Dry
The following process is omitted
as it is the same as the processfrom SH-3 “Scenario branchingE” to SH-4 “Send out.”No.
Handled item

[62]
Canister[63]
Canister[64]
Canister

Discharge water inside thecanister through the drainpipe for the purposes
including reduction of theTake the canister out of the
Wash the canister surface
Details and purpose
amount of hydrogenpool to transition to dry
to prevent the spread ofof task
generated and risk ofstorage.
contamination.
leakage, as well as
relaxation of the initialconditions of drying
treatment.

[65]
Canister[66]
Canister[67][68]
CanisterCanisterOn sending out the canister tothe dry storage facility, confirm
Measure the amount of
Dry the surface and inside the canisters to collect and linkhydrogen generated toof the canister to reduce
data and conditions of eachconfirm that the hydrogen
Dry water on the canister
the risk of criticality (during canisters again (massconcentration inside thesurface to prevent thearrangement), risk ofmeasurement, inspection forcanister will not reach the
spread of contamination bycorrosion, the amount ofcontamination on surface,lower explosion limit withinwater drops.
hydrogen generated, etc.
decontamination, surface dosea specified time when
Confirm
rate measurement, visualtransferring to the dry
inspection (including confirmation storage facility.
of canister ID)).

Scenario branching H[Details]“Perform drying treatment on thefuel debris before sending out thecanister” or “no drying treatment”

Figure: Example of process flow: Taking out canister in a storage facility ~ Preparation for transport©International Research Institute for Nuclear DecommissioningNo.26

6. Implementation Details
6.2 Study of Safety Requirements and Specifications of Storage Systems for the Transfer/Storage ofFuel Debris Canisters(2) Study on the storage system (13/13)

Issues that need to be solved to make the system work were identified based on the provisionally established process flow. (The figure
below shows typical discussion topics.)• Canister, items to be placed in the waste container,
definition of fuel debris, method of confirming fueldebris, quantity, location of sorting (inside furnace,
Reactor building
before temporarystorage, predisposal)• Function of canisters, functionof facilities, share of functions,
restrictions

Maintenancecell

Fuel debris retrieval
storage cell

Hatch foraccess(additional)• Concepts of transfer, necessary functions, items forconfirmation and treatment before sending out (for safety
and material accountancy), frequency of send out(capacity of transport cask)

Fuel debris carryingout building

Carry out cell

Storage canister
handling cell

Storagefacility

Fuel debris
retrievalStorage incanister

Cleaning ofcanister

Storage of canister intransport cask

Send out transport
cask

Figure: Bottom of the primary containment vessel; example of partial submersion-side entry method©International Research Institute for Nuclear Decommissioning6. Implementation Details
6.2 Study of Safety Requirements and Specifications of Storage Systems for the Transfer/Storage ofFuel Debris Canisters(3) Study of storage method

No.27

As an example, a case study was conducted and the decision flow of temporary storage (wet storage) organized in a joint meeting withexternal experts, experts of IRID, the Method Project Team, and the Waste Project Team. As a result, since wet storage may be required, it
has been decided to continue negotiations in FY2018 and work on making the method concrete.
Storage of fueldebrisConfirmation
methodIs it possible to complete constructionof a dry storage facility before fueldebris retrieval begins?

Can fuel debris be completely
dried*2?

Sealed dry storage*2: Level of the amount of hydrogengenerated for which long-term sealed
storage is possible

Is confirmation by hot run for dry
storage necessary?
Vent method, treatment method ofgaseous radioactive waste

Vent dry storage

Can existing wet storage
facilities be used?*3

Is it possible to complete construction of a wet
storage facility before fuel debris retrieval begins?

Vent wet storage(modify existing pool)*3: If the facility can be put to use with minor modifications

Assume that a small wet storage
facility for temporary storage will be
newly constructed

What are the requirements for wet
storage of fuel debris? (Can it be mixed
with spent fuel? What is the level ofmodification?)

Upon completion of construction of
the dry storage facility, shift to dry
storage

Can the commencement of fueldebris retrieval be rescheduled?

Is fuel debris retrieval
not being able to start?

Vent wet storage(storage facility newly
constructed)

Will fuel debris not be retrieved until constructionof the storage facility completes?

The need for wet storage will be determined by the timing of fuel debris retrieval
and completion of the dry storage facility, rather than the technical aspect of fueldebris drying.

Figure: Decision flow of primary storage (wet storage)

Temporary storage
inside the reactor
building©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.28

6.2 Study of Safety Requirements and Specifications of Storage Systems for the Transfer/Storage ofFuel Debris Canisters(4) Study on drying system (1/6)(i) Purpose
Dry storage is a reasonable method for long-term storage of fuel debris. On the other hand, a method must be devised to
dry the fuel debris for dry storage. To create the drying system’s concept plan, this study aims to ensure throughput andprovisionally determine a safe drying method, considering the basic requirements of the system (especially discharge offission products).
(ii) Target drying conditions
Free water elimination was chosen as the drying goal.
The features of dry storage are that by drying fuel debris and storing it in an inert atmosphere, it suppresses aged deterioration(e.g., corrosion) and hydrogen generation by water radiolysis. This can be mostly achieved by eliminating free water contained infuel debris. Crystal water contained in concrete (e.g., MCCI products) does not contribute to cell formation which causes corrosion.
In addition, data from previous research shows that it does not contribute to hydrogen generation. For this reason, it was excluded
from the subject.
(iii) Summary of FY2017 studies
In addition to heated vacuum drying conducted in TMI-2, high-temperature drying which heats the inside of the canister
to a higher temperature and hot air drying that blows in hot air were studied.
• Regarding 1F, the method of storing collected fuel debris into a unit can and then storing that into a canister for reasons ofworkability is being studied. It is assumed that heat will not easily transfer to the fuel debris when the canister is heated.
• When drying the unit can, it is expected the fuel debris will dry faster due to direct heating. As for technical development, study
results of drying method for canisters can be diverted so it will not be a subject for study as of now.

Internal pressure of canister:
vacuum to atmosphericpressure
Flow of air

Heatingsurface(e.g., heater)

Vacuumdrawing

Flow of heat

Canister wall (approx. 120C)
Moisture

Fuel debris

Radiation

VacuumHeatingsurface(e.g., heater)

Canister wall (max. 300 C)
Fuel debris[Heated vacuum drying]

Canister wall (max. 300 C)
Fuel debris

Moisture

Moisture

Radiation
Thermalconduction

Radiation
Convection

Thermalconduction of air(still air)Figure: Overview of the drying
methods
Inside canister: vacuum (decompressed)

Heatingsurface(e.g., heater)

Inside canister: air[High-temperature drying]

Forcedconvection(hot air)Inside canister: hot air (forced convection)©International Research Institute for Nuclear Decommissioning[High-temperature heating & hot air drying]6. Implementation Details

No.29

6.2 Study of Safety Requirements and Specifications of Storage Systems for the Transfer/Storage ofFuel Debris Canisters(III) Falling rate drying(II) Constant rate(4) Study on drying system (2/6)
perioddrying period(I) Preheating
period(iv) Concepts

b. Constant rate drying period (II): Since water starts to boil and drying progresses tobalance the external heat input and evaporation latent heat, the moisture content will
decrease at a fixed rate with respect to time.
The evaporation time can be calculated by dividing the residual water volume by theevaporation rate.
-> External heat input is calculated by simulating a heat transfer model. (Assumed
that the water volume is equivalent to 30 vol.% of fuel debris volume, storage
efficiency is 30 vol.%, and heat generation of fuel debris is zero according to thedraining test)c. Falling rate drying period (III): Drying is dominated by the amount of mass transfer,
such as water evaporation, expansion, and seepage in narrow parts. The shape of thefuel debris (state of water in narrow parts) is unknown and elaboration is difficult
without examining it. III will also be qualitatively short if II is short.
-> This phase is ignored since the period is short in the element test conducted in the
Fuel Debris Characterization Project Team. However, the period may have beenshort since it was a small-scale test, so it will be our future task.

-dW/dt: Drying rate

a. Preheating period (decompression period in vacuum drying) (I): Period in which thewater inside starts to boil due to heating (in vacuum drying, the water inside will start
to boil due to decompression inside the container)
The time it takes is determined by the heat input and the heat capacity of the fueldebris. (In vacuum drying, it will be determined by the discharge capacity of thevacuum pump and the capacity of the container.)-> This phase is ignored in this study because the period is short.

Surface dries

C
SurfaceWater
evaporates

InsideWater
evaporates

Suspended
water

Equilibrium moisture content
Critical moisture content

Moisturecontent

Drying rate [mg-H2O/cm2*min]

Calculate the evaporation rate based on the constant rate drying period (area II
in the right figure)

Al2O3
SiO2
ZrO2
UO2

20

300C

200C
10

100C
0
-200

0

Ceramics

200

400

Water content [mg]Figure: Drying characteristics (with fuel debris substitute material)
Reference: FY2016 study results of Fuel Debris Characterization
Project Team
Note: The graph shows that the drying rate changes as water content©International
Research
Institutefor Nuclear Decommissioning
decreases (dries)(decreases
from rightto left).

6. Implementation Details
6.2 Study of Safety Requirements and Specifications of Storage Systems for the Transfer/Storage ofFuel Debris Canisters(4) Study on drying system (3/6)No.30(v) Evaluation method
In the FY2016 fuel debris preliminary draining test, residual water was about 10 to 50 vol.% of the fuel debris volume even whendrained. Parameters such as decompression and convection inside canisters under the premise of heating were surveyed, and thedrying rates were compared. The drying rate was calculated by dividing the heat input to the fuel debris by the evaporation latent
heat of water.

Canister walls Th: max. 300C (determined as range of study based on information on TMI-2)
Internal pressure ofcanister:
Decompressed toatmospheric pressure

Fuel debris
Study conditions: Assume that theconstant rate drying period (II) is the
only drying process(no falling rate drying period)

Air flow
Td: Fuel debris
temperature

Heat flow

Radiation(determined as boilingpoint of water based on
the pressure inside thecanister when drying)

Moisture inside the fuel debris is assumed
to be supplied to the surface and heat
input will all be used to evaporate themoisture (evaporation heat). Therefore,
the fuel debris temperature will be theboiling point of water.

Forced convection: Nu number is set
Convection: When not considered, flowrate of approx. 0.2 m/sec

Figure: Overview of the heattransfer model©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.31

6.2 Study on Safety Requirements and Specifications of Storage Systems for the Transfer/Storage ofFuel Debris Canisters(4) Study on drying system (4/6)(vi) Studied cases
The following cases were specified for study.
Heat drying (basic case)
Heated vacuum drying
High-temperature drying
High-temperature heating & hot air drying

Table: List of conditions of studied cases
Heated vacuum
dryingHigh-temperaturedryingHigh-temperature
heating & hot air
dryingList of conditions

Unit

Heat drying(basic case)

Canister temperature

C

120

120

300*1

300

Warm air temperature
at opening
Fuel debris
temperature(assumed to be theboiling point of water)
Inner diameter ofcanister

C

-

-

-

300*1

C

100

11*2

100

100

mm

220

220

220

220

m/s

-

-

-

0.2

Flow rate of warm air*1: A value provisionally specified as the upper limit based on the information on drying fuel debris in TMI-2*2: A value specified on the assumption that the pressure during vacuum drying is 10 mmHg (boiling point ofwater at pressure of 10 mmHg)©International Research Institute for Nuclear Decommissioning6. Implementation Details
6.2 Study of Safety Requirements and Specifications of Storage Systems for the Transfer/Storage ofFuel Debris Canisters(4) Study on drying system (5/6)

No.32(vii) Evaluation results
The drying time was compared between the drying systems.
1.00

0.30
0.05
Heat drying(basic case)

Heated vacuum drying(effect of decompression)(fuel debris temperature 11 C: 10 mmHg)

0.02

High-temperature drying
Effect of heating thecanister to 300C

High-temperature heating &hot air drying
Effect of forced convection in addition
to heating up the canister to 300C

Figure: Comparison of drying time when heat drying (basic case) is expressed as 1
(Basic case: Fuel debris temperature 100C (atmospheric pressure), canister temperature 120C, time required without hot air)In the basic case, when ⌀220 mm canister is filled up to 30% with fuel debris as well as 30 vol. % of water in respect to the fuel debris
volume, the time required to dry is about 3 days (area II only). More time may be required if area III, which has a large uncertainty, istaken into consideration.
-> According to the results of trial calculation for the processing capacity, based on the quantity studied in 6.2.(2) (assuming
10.8 cans/day), rationalization of facilities and shortening of period may be important and, in this case, high-temperature
heating and high-temperature heating & hot air drying are effective drying methods.
©International Research Institute for Nuclear DecommissioningNo.33

6. Implementation Details
6.2 Study of Safety Requirements and Specifications of Storage Systems for the Transfer/Storage ofFuel Debris Canisters(4) Study on drying system (6/6)(viii) Conclusion• A heating method that is also effective to reduce pressure and warm air injection into the canisterare effective in drying.
(ix) Future plans
Along with heated vacuum drying which gave results in TMI-2, high-temperature drying and hightemperature heating & hot air drying will be studied since effect on ensuring throughput can be
expected.

a. Uptake results of fission product behavior (chemical form, volatility) of nuclide inside the fueldebris studied by the Fuel Debris Characterization Project Team and specify the drying
temperature condition.
b. Study the drying method in terms of arranging apparatus for the drying system and examine thespecification proposal for the drying equipment.

©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.34

6.2 Study of Safety Requirements and Specifications of Storage Systems for the Transfer/Storage ofFuel Debris Canisters(5) Study and review of specifications for handling device(i) Purpose
When handling canisters, there are processes such as closing lids and hoisting canisters. Devices used for such processes are dedicated tothe canisters and, as part of the system study, the basic specifications of these devices will be determined along with the canister
specifications.
(ii) Summary of FY2017 studies
Examples of handling devices (lid closing device, canister hoisting attachment) based on the study on lid structure of 6.3 (2) as of now are
shown below.
(iii) Future plans
The processing capacity study will be continued as the lid structure updates while taking requirements into consideration.

Main shaft

Liftingequipment
Arm open/close
motor
Bolt tightening
deviceCamApprox.
2500 mmApprox.
2000 mmNut runnerArm

Figure: Example of lid closing device (for bolt structure 2)

Figure: Example of canister hoisting attachment(when grabbing onto the canister itself)©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.35

6.3 Development of Safety Evaluation Methods and Validation(1) Safety evaluation on sub-criticality (1/5)(i) Study on sub-criticality maintenance measures

Purpose: To select a sub-criticality maintenance measures necessary to expand the canister’s inner diameter, feasibility will be
evaluated in terms of inner diameter expansion effect and explainability of safety assessment.
Fuel debris conditions

Inner diameter of canister

Strictest conditions(Approx. 5 wt% UO2 + optimumBasedeceleration state of water)
No.

(i)

(ii)

(iii)

(iv)

Sub-criticality maintenancemeasures(design condition restriction items)• Approx. 220 mm(10 mm thick, made of SUS)

Explainability of safety assessment



Inner diameter expansion effect (target inner diameter: 400 mm)• No problem in explainability

Explainability of safety assessment• 400 mm cannot be achieved within the realistic range of settings fordegree of enrichment.
(Effect of inner diameter expansion is small between approx. 2 to 5
wt% and concentration must be reduced to about 1.7 wt% to achieve
inner diameter of 400 mm.)

• Although it can be explained based on combustion results and fuel design
specifications, they need to be guaranteed as a design condition. A low degreeof enrichment with predictable effects cannot be guaranteed as the design
condition.
• The maximum degree of enrichment inside the reactor vessel with consideration
to combustion depends on the outcome of the Criticality Control Project Team,but it is unlikely that significant reduction can be expected.

Consider possibility of structure material
mixing into fuel debris• Effect is in a certain level. Expansion is dependent on mixture ratio offuel debris and structure material inside canister.

• It needs to guarantee the mixture ratio of fuel debris and structure material andstate of distribution as the design condition.

Consider the possibility of neutronabsorbing material inside the reactor
vessel (Gd/B4C) mixing into fuel debris• The effect of inner diameter expansion is large. The diameter can
expand to 400 mm depending on the condition but depends on the
mixture ratio of fuel debris inside the canister and neutron-absorbing
material inside the reactor vessel.

• It needs to guarantee the mixture ratio and state of distribution as the design
condition.

• The effect of inner diameter expansion is large. The diameter can
expand to 400 mm depending on the condition.

• It needs to guarantee the concentration and state of distribution as the design
condition.
(Inspect possession of knowledge and conditions.)

Decrease enrichment degree

Inject boric acid solution

(v)

Restrict water volume

• The effect is in a certain level. The expansion depends on the
distribution condition of fuel debris and water inside the canister.

• It needs to guarantee the moisture content and water volume (amount ofresidual water) as the design condition by water volume measurement
technology if applying to the 1F fuel debris retrieval site.
• Less residual water compared to draining can be expected if put through drytreatment, but it needs to guarantee water volume after drying.

(vi)

Lay neutron-absorbing material inside
the canister• Inner diameter of the canister can be expanded to 400 mm by laying B
SUS plate and B4C pipe.

• The effect of accidents (e.g., deformation) needs to be confirmed, but it canbasically be set as a design condition and has explainability.

(vii)

Change the canister material• It is possible to expand the inner diameter by about 20 to 30 mm whenthe material is changed to B-SUS.

• There is explainability and sufficient usage achievement as materials for racks
and baskets.

[Effect] C: Low possibility of reaching 400 mm, B: Medium possibility of reaching 400 mm, A: High possibility of reaching 400 mm, [Explainability] D: No explainability, C: Low explainability, B: Medium explainability, A: Highexplainability

In terms of explainability and effect, application of neutron-absorbing material is effective for expanding the inner diameter of the canister and preventing fueldebris re-criticality.
©International Research Institute for Nuclear Decommissioning
However, although restriction of water volume is simple, there is an issue in explainability. -> Consider restricting the water volume by draining the fuel debris.

6. Implementation Details

No.36

6.3 Development of Safety Evaluation Methods and Validation(1) Safety evaluation on sub-criticality (2/5)(ii) Study on operation of fuel debris water volume restriction (inspection on water volume measurement technology)

Purpose: To apply the restriction on water volume to enlarge the inner diameter of the canister, it was judged that quantitative guarantee bymeasurements is necessary since there is little margin in terms of criticality control with just draining the fuel debris. The water volume
measurement methods were investigated and their applicability to 1F fuel debris retrieval at the site evaluated.
Electric resistanceElectric capacity

Microwave

Near-infrared (light)

Calculated from thePrinciple of resistance value given when Calculated from the change in Calculated by irradiating the Calculated by irradiating
subject of measurement withmoistureelectricity is applied to theelectric capacity when AC
infrared light on the subject
microwaves and measuring
voltage is applied to theof measurement andmeasurement subject of measurement
the amount of energy before(insulator or semiconductor) subject of measurement
through electrodes

Major
applications Lumber, construction(example of materials, sand, gravelapplication)

Construction materials,
paper, gas, liquidand after the wave passes

measuring its reflectivity

Lumber, construction
Food, minerals, chemicals,
materials, sand, gravel, food,flour, sludge
sludge, powder

NeutronDry weight

Chemical measurement

Calculated by irradiating fastneutrons on the subject of
Measured the weight of themeasurement and measuring
subject of measurement
the amount of neutrons that
before and after drying
change their state into
thermal neutrons

Standard method of moisturemeasurement
Measured by titrating with the
moisture of the subject ofmeasurement and Karl
Fischer reagent, andmeasuring the amount of
titration which thepolarization voltage changes
drastically due to free iodine

Minerals, sintering raw
materials, coke, sand,concrete

Food, oil, fat, drugs

Food, lumber, coal, coke,minerals

Determined as not applicable
Determined as not applicablefor the following reasons
Determined as not applicablefor the following reasons➢ Cannot be used as afor the following reasons
Determined as not applicable
Determined as not applicable➢ Cannot be measured sincemeasurement method to ➢ The fuel debris needs tofor the following reasonsfor the following reasons
there is a high possibility
ensure sub-criticality
be dried to measure its➢ Only the surface of the➢ Cannot be conductedthat metal from thesince the measurement
water volume, but it
Determined as not applicable Determined as not applicablesubject of measurement
inside the pedestal or
Applicability to
structural materials
value fluctuates greatly
cannot be dried inside thefor the same reasons as
for the same reasons as
can be measured, so the
RPV because reagents(conductive
materials)
isdependingonthepedestalor
RPV
fuel debris
electric resistanceelectric resistancemoisture inside the fuelare very sensitive tomixed into the fuel debris
composition of the fuel
➢ It only measures thedebris or fuel debris that
atmospheric moisture➢ Measurements are taken
debris and accuracy
weight before and afteris stacked up cannot be
and must be isolated
via unit can (steel
cannot be guaranteed
the fuel debris is dried;measuredfrom moist environment
container), so high➢ Performance cannot be
however, it may still
accuracy is hard to obtain
guaranteed under highcontain water after drying
radiation environments⚫

At the 1F fuel debris retrieval site (remotely, under radiation environment), measurement method applicable to fuel debris, withindefinite properties, composition, and shape via canisters, was not confirmed. Since water volume cannot be quantitatively
guaranteed, water volume restriction cannot be applied to the design conditions of the criticality prevention function.
The results reveals that it is currently difficult to be designed for the condition of water volume restriction. However, waterdrain can be effective to reduce drying time, so this process will be applied.
©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.37

6.3 Development of Safety Evaluation Methods and Validation(1) Safety evaluation on sub-criticality (3/5)(iii) Criticality evaluation assuming dry storage (infinite array condition)

Purpose: To streamline the dry storage facility on the premise of drying, the close-packed storage conditions for canisters wereclarified from the aspect of maintaining sub-criticality.
<Evaluation conditions>◆ Assumed that fuel (initial maximum concentration of pellet 4.9 wt%) and water (residual water after drying treatment) are an evenand homogeneous mixture◆ Assumed that volumetric water content is reduced to 0.2 or less by sufficient drying◆ As a condition for safety, the model is created assuming the canister is arranged indefinitely (see the computational model
below)

Canister
UO2 + H2O

Modeled as an infinite array*1

Height direction: Infinite

Figure: Computational model

Effective multiplication factor (keff + 3σ)<Evaluation results>
Even with a canister with a 220 mm inner diameter that can maintain sub-criticality on its own, the effective multiplication factor
will increase when multiple canisters are arranged. In this evaluation condition, even if the inner diameter of the canister is 220 mm,
water needs to be removed by drying treatment down to a volumetric water content of 0.1 or lower (margin considered) to maintain
sub-criticality.-> Examine as a precondition of drying evaluation
Exceeds the evaluation*1: Since it is an infinite array model, the effective
multiplication factor does not depend on the center-tocenter distance of canisters.

standard (0.95)

For infinite array, volumetricwater content of 0.1 or less isrequired in canister with innerdiameter of 220 mm
Inner diameter
220 mmInner diameter
400 mmVolumetric water content©International
Research
Institute forresults
Nuclear Decommissioning
Figure:Evaluation

6. Implementation Details

No.38

6.3 Development of Safety Evaluation Methods and Validation(1) Safety evaluation on sub-criticality (4/5)(iv) Evaluation of criticality that assumes temporary storage (e.g., inside hot cell) (finite array conditions)Purpose: Temporary storage of canisters, such as inside the R/B cell, is likely to be difficult, especially in terms of space. Closepacked storage conditions are clarified from the aspect of maintaining sub-criticality.
<Evaluation conditions>◆ Assumed that fuel (initial maximum concentration of pellet 4.9 wt%) and water (residual water after drying treatment) are aneven and homogeneous mixture◆ Assumed that volumetric water content is reduced to 0.2 or less by sufficient drying◆ Assuming temporary storage, modeled based on finite array for canister arrangement (10  10)

When volumetric water content is 0.2 for canisters with
inner diameters of 220 mm, if center-to-center distance of 33
cm or greater is ensured between the canisters, sub-criticality
can be maintained in a 10  10 arrangement.

On the other hand, if volumetric water content can be
reduced to 0.1 or less, canisters can be placed side by side incontact with each other (maximum area efficiency). Here, thestorage efficiency*1 does not depend on the inner diameter andis about 0.7*2. The increase in the diameter of the canister
opening does not contribute to the improvement of storage
efficiency.
*1: Storage efficiency is the area occupied in the canister per unit area
*2: Assume that the thickness of the canister is 10 mm

Effective multiplication factor (keff + 3σ)<Evaluation results>

Evaluationstandard
Evaluation(0.95)
standard(0.95)For volumetric water content of 0.2, the center-to-center distance mustbe 33 cm or more between the canisters to maintain sub-criticality

Center-to-center distance between canisters [cm](Example of canister with inner diameter of 220 mm, 10  10 array system)©International
Research Institutefor Nuclear Decommissioning
Figure:Evaluation
results

6. Implementation Details

No.39

6.3 Development of Safety Evaluation Methods and Validation(1) Safety evaluation on sub-criticality (5/5)(v) Conclusion

a. Study on sub-criticality maintenance measures⚫ Measures for maintaining sub-criticality when expanding the inner diameter of the canister were studied in terms of
expansion effect and explainability of the safety assessment.
⚫ As expansion of the inner diameter and the work volume of fuel debris retrieval are proportional, measures to be
applied will be established in the future, taking into consideration the viewpoints other than maintenance of subcriticality, such as requirements on handling.

b. Study on operation of fuel debris water volume restriction⚫ It was confirmed that restricting water volume is difficult as a means of maintaining sub-criticality when expanding
the canister’s inner diameter. However, as is assumed that draining fuel debris has merits in terms of operation,
incorporating draining into the fuel debris handling process as a premise in each study at a degree which will noteffect throughput or incidental equipment will be considered.

c. Criticality evaluation assuming dry storage⚫ If drying treatment performance can be obtained, the canisters can be arranged side by side. In addition toincorporating it as a requirement for examining the drying method, a sub-criticality maintenance scenario will becreated, reflecting the study results of the Method Project Team, Fundamental Technology Project Team, Criticality
Control Project Team, etc. concerning the handling of residual water.

(vi) Future plans⚫ Reflect the study results of the Method Project Team, Fundamental Technology Project Team, Criticality Control
Project Team, etc. and decide on a sub-criticality maintenance scenario.

©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.40

6.3 Development of Safety Evaluation Methods and Safety Validation(2) Safety evaluation on structural strength: Study on lid structure (1/8)(i) Purpose
The structure of a canister lid required to store fuel debris will be studied as part of the canister design.
By FY2018E, a proposal of the lid’s structural design will be provided, that achieves safety (necessary
structural integrity), workability (necessary throughput), and other factors.
(ii) Required lid specifications
Based on the function required for canisters, proposals for required specifications of the lid structure
were developed in FY2016 with consideration of remote control (see the table below).
 Confinement that will not release fuel debris pieces outside the canister (includes structural
integrity that guarantees confinement) and workability by remote control are necessary.
In FY2016, simple installation structures, bolt structures, and welded structures were selected as lid
structures that satisfy these required specification proposals and their outlines were briefly studied.
Table: Proposal on required specifications of the canister lid structure
Items

Requirements

Confinement
property

Fuel debris pieces shall not be released outside*1.

Structural
strength

The lid shall not open or break even when the canister receives an effect (e.g., toppling).
The canister shall maintain its integrity against internal pressure*2.

Workability

The lid shall be able to open and close above and under the water by remote control.
The lid shall be able to close with simple movements (e.g., turning lid) from the aspect of
workability.

*1: Since vents are basic measures of canisters against hydrogen, airtightness is not required of the lid area.
*2: Although airtightness is not required of the canister, as one of the conditions of the lid’s structural design, it is required to withstand a certain level of internal pressure.

©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.41

6.3 Development of Safety Evaluation Methods and Safety Validation(2) Safety evaluation on structural strength: Study on lid structure (2/8)(iii) Organization of assumed events
Based on the latest study by experts inside and outside of IRID, the Fuel Debris Retrieval Method Project
Team, and the Fundamental Technology Project Team, the canister handling flow (from retrieval to storage)
was reviewed, and events that should be evaluated to ensure safety were re-identified.
After loading fuel debris into the unit can, transfer it inside building (method ofaccessing from the side of the nuclear reactor)
Reactorbuilding

Hoist

Additionalbuilding

Hoist

Manipulator

Hoist

Canister lid
closing system

Fuel debris
retrieval
deviceCanisterPort

Primary lid closing Secondary lid
device, gasclosing device, gasinsertion
insertionequipment, leakage equipment, leakage
confirm equipment confirm equipment

Seal cell

Debris treatment cellGas insertion system

Inside nuclear reactor

Dryingequipment• Load unit can into drying
equipment• Dry fuel debris*1: Not just unit cans, but fuel debris blocks
and structures, also go through theaccess tunnel to move from the nuclear
reactor to the additional building. The
figure shows the transportation of unitcans as a typical example.

Carry in of transport cask into storage
building to storage
Hydrogen gas
Hoist treatment system

Weight measuringsystemCanister transfer
systemDebris treatment cell

Access tunnel• Transport UC into an
additional building*1

CanisterTransport
vehicleUnit can transfer
system• Collect fuel debris into UC using
the fuel debris retrieval device• Transport UC to access tunnel

Handling jig

Drying equipment

Access tunnel

Fuel debris

Hoist

Preparation
cell

Transport
caskTransport cask
transfer system

Seal cell• Load UC into the canister• Temporarily close the canister lid• Move the canister to the lid closingarea• Close the canister lid• Confirm the canister lid is closed(• Connect canister to drying
equipment*2)(• Dry canister*2)• Connect the canister to gas
insertion equipment• Insert inert gas into canister

Insidebuilding

Preparation cell
• Place the canister onto the weightmeasuring system• Measure the weight of canister• Move the canister onto the port

*2: Conducted when the canister cannot be driedin the debris treatment cell

Ventilation system

Transfer frame• Load canister into the transport cask
• Open the canister’s vent pipe
• Close lid of the transport cask (primary lid)
• Insert inert gas (inside cavity)• Confirm the lid of the transport cask is closed (primary lid)
• Close the lid of the transport cask (secondary lid)
• Insert inert gas (between the primary and secondary lids)• Confirm the lid of the transport cask is closed (secondarylid)• Carefully load the transport cask onto the transport
vehicle• Confirm the transport cask before transfer• Transfer inside
the buildinga. Verticaldrop

c. Vertical
drop onto
canisterb.
Toppling

Storagebuilding

Area monitor

Handling jig

Gas treatment system

Canister

Gate
Transport cask

Canister transfer
systemCamera Shielding coverStored item
confirming
systemDryingequipment

Storagecontainer

Primary lid ofstoragecontainer

Primary lid ofstoragecontainer

Vent pipe

Scaffold for
operations

Unit
can

Port
Approx.
8.6 m

Approx.
6.1 m

Shielding wall
(approx. 2 m(provisional value))

Transport cask*2(approx. 5.6 m)

Measuring
deviceTransport cask
frame

Storage facility truck bay
door

Acceptance area (i)Stored item confirming
area

Drying area

• Carry in transport cask through truck
bay door• Confirm acceptance of transport cask

• Remove lid of transport cask (primary lid)(Measure amount of hydrogen)• Remove lid of transport cask (secondary lid)
• Close canister’s vent pipe
• Take out canister from the transport cask*3• Load canister into
stored item
confirming system• Confirm stored item

• Load canister into
drying equipment• Dry fuel debris*4*4: Conduct if drying isnecessary before storage

Acceptance area(ii)

• Load canister into storage
container• Open canister’s vent pipe
• Close primary lid of storage
container• Insert inert gas (inside cavity)• Confirm primary lid of storage
container is closed

Storage preparationarea• Close secondary lid ofstorage container• Insert inert gas (between
primary and secondary lids)• Confirm secondary lid isclosed• Confirm before storage

Storage area
• Prepare forstorage• Store*3: Put the shielding cover on thecanister and transfer

Figure: Proposal of canister handling flow

Canister(approx. 2 m)
Transport cask transfer system (approx.
0.5 m (provisional value))

Figure: Example of event identification*1*1: Confirmed that events in the storage facility is the same or includes these events*2: Assumed double-stacked onto the existing spent fuel transport cask

Confirmed that a. vertical drop height is included in the 9-m height specified by FY2016 and identified that b.
toppling and c. vertical drop onto canister (assume 7 m considering all events) as new events.
©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.42

6.3 Development of Safety Evaluation Methods and Safety Validation(2) Safety evaluation on structural strength: Study on lid structure (3/8)(iv) Example of lid structure design (1/2)
Along with the brief study conducted in FY2016, the lid structure proposals were studied with considerationof the radioactive waste containers. The study focused especially on the confinement property of fuel debris
and workability in 1F.

Lid(integralstructure*1)
Groove forhoisting lid
Anti-rotation
fitting

Fitting
part

Locknut

Hole forhoisting lid

Bolt

Grip for hoisting lid
Lid

Bolt

Body of canister(flange)
Vent pipe
attachingposition (a) Simple installation
structure

Lid

Vent pipe attachingposition

O-ring(attached toside of lid)

*1: A structure with an
integrated barrel and clamp(binding) that fastens the lid
plate.

Vent pipe
attachingposition

Body of canister(flange)O-ring(attached to sideof lid)(b) Bolt structure 1
Figure: Example of the lid structure design (whenthe inner diameter of the canister is 220 mm)(c) Bolt structure 2

O-ring(attached toflat surfaceof lid)Body ofcanister(flange)©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.43

6.3 Development of Safety Evaluation Methods and Safety Validation(2) Safety evaluation on structural strength: Study on lid structure (4/8)(iv) Example of lid structure design (2/2)
Comparisons were made on the designed lid structures and future evaluation policy was determined.
Table: Example of comparison between lid structure plans
Items

Simple installation structure

Bolt structure

Welded structure

ExamplesketchNot studied
here since itwill be studied
in detail for
weld groove

Bolt structure 1
Features<Workability>• Lid can be closed with simple steps (pushing,rotating).

<Confinement property>• Confinement is ensured by squeeze of O-ring
resulting from push-in of lid.

Bolt structure 2<Workability>• Remote bolt tightening is probably possible, but
know-how, including torque management, isrequired.
• Measures against galling of bolt are necessary.
<Confinement property>• Confinement is ensured by squeeze of O-ring
resulting from push-in of lid (bolt structure 1) ortightening of bolt (bolt structure 2).

Examplerecord ofuse• Integral structure itself has been used in a superhigh pressure vessel, but changes need to bemade to 1F canister (e.g., sealing method).
• There are design examples for waste containers.

• Examples of remote control bolt structure
containers include Rokkasho low-level waste
container.

Evaluation

This will continue to be taken into consideration
as one of the options for its ease of installation.
Feasibility of the proposed lid structures (e.g.,
structural integrity) will be evaluated in FY2018.

This will continue to be taken into consideration
as one of the options for its robust lid structure.
Throughput evaluation will be conducted on thelid closing task, and applicability will be evaluatedin FY2018.

<Workability>• There are actual results in remote welding, but
the feasibility in 1F environment must be
studied.

<Confinement property>• Confinement is ensured by welded structure.

• Rokkasho glass solidified container• Canister for spent fuel (overseas)

Considering the welding work environment, this
proposal is difficult to adopt at the moment.
However, there may be welding treatment at thestorage facility for long-term storage, so
evaluation will continue while keeping an eye on©International Research Institute for Nuclear Decommissioning
the specific progress of storage.

6. Implementation Details

No.44

6.3 Development of Safety Evaluation Methods and Safety Validation(2) Safety evaluation on structural strength: Study on lid structure (5/8)(v) Study on a lid structure verification test (1/3)
Matters to be confirmed to assess the feasibility of designed lid structure proposals were evaluated.
Lid damage modes were studied from the aspectof the lid’s confinement property (refer to thefigure). The following damage modes wereidentified for events which the lid take effect load(e.g., drop).

Leakage offuel debris

Rupture of lid

Corrosion, pitting
corrosion

Environment (hightemperature and humidity)

Deformationof lidDrop of heavy
weight object

Material
selection

Drop of canister

Rupture of seal

Biting of foreign
matterAdherence of fuel debris etc.

Heat deterioration of Oring

Environment

Life

Material
selection

Irradiation deterioration of
O-ringEnvironment

Life

Material
selection

Rupture of O-ring

Burr
Failure of lid closing
task

a. Deformation of lid
b. Rupture of O-ringc. Deformation of canisterd. Rupture of fitting part and bolt

Biting of foreign matter

Failure of O-ring attachment
Drop of heavy
weight object

Drop of canister

Among these, regarding a., c., and d., evaluation
by dynamic analysis was performed on the events
identified in the study on events to be evaluated.
For b., it is currently difficult to precisely simulate
damage behavior of O-ring only with analysis andrequires confirmation by tests.

Toppling of canister

Deformation of canister(expansion in seal area)

Drop of canister

Toppling of canister

Thermaldeformation
Rise in internalpressure

Rotation oflid

Damage of antirotation pin andbolt

Drop of heavy
weight object

Drop of canister

Falling oflid

Rupture of fitting part andbolt

Drop of heavy
weight object
Drop of canister

Figure: Example of identification of lid damage modes©International Research Institute for Nuclear Decommissioning6. Implementation Details
No.45

6.3 Development of Safety Evaluation Methods and Safety Validation(2) Safety evaluation on structural strength: Study on lid structure (6/8)(v) Study on a lid structure verification test (2/3)
Dynamic analysis was conducted using the designed lid structure proposals on the events selected in (iii).
Integrity of the confinement function against damage modes identified from the analysis results were
11.7 m/s
confirmed.
CanisterbarrelBuffering structure
on bottom

Canister lid
Storeditem(equivalent to 7-m drop)

Dropped canister(simulated
canister: Approx.
1.5 t)

Plastic strain
near the seal area

420 mm

110 mm
10 mm

57 mm

1500 mm

Canisterlid

545 mm
400 mm

400 mmCanisterbarrel2000 mm

Give angular velocity at
topping, the origin being
the center(i) Analysis model

Plastic strain
near the seal area(ii) Equivalent strain around the seal area(i) Analysis model

1.0

Rigid
floor

(ii) Equivalent strain around the seal area

1.0
Exceeds theprovisional acceptancevalue and leakage may
occur

0.8
0.6
0.4

Provisional acceptancevalue: Half the O-ringsqueeze prescribed by JIS
is provisionally set

0.2

0.0
-0.2

in the seal area [mm]Gap[mm]
シール部の隙間量in the seal area [mm]Gap[mm]
シール部の隙間量

Original
point

10 mm

0.8

Provisional acceptancevalue: Half the O-ringsqueeze prescribed by
JIS is provisionally set

0.6
0.4
0.2
0.0

Doesn’t exceed the

-0.2
0

30

60

90
120
Angle
角度 [degree]
[度]

150

180

(iii) Gap in the seal area*1 (180 degrees on the top)(a) When toppled

0

30

60

90
120
Angle
角度 [degree][度]

150 provisional
180 acceptancevalue, and leakage is not
likely to occur(iii) Gap in the seal area*1(b) Vertical drop onto canister (without buffer structure)

Figure: Example of dynamic analysis results (simple installation structure, inner diameter of canister 400 mm, temperature 300C)*1: Amount of gap is evaluated from the amount of deformation in the canister barrel and the lid near the sealing surface.

©International Research Institute for Nuclear Decommissioning6. Implementation Details
No.46

6.3 Development of Safety Evaluation Methods and Safety Validation(2) Safety evaluation on structural strength: Study on lid structure (7/8)(v) Study on a lid structure verification test (3/3)
From the analysis results, the integrity of the confinement function of the lid structure proposals wasevaluated and a method of examining the validity of the evaluation results and dynamic analysis wasstudied.

Vertical drop

Table: Example results of dynamic analysis of parts relevant to the confinement function of the lid (for simple installation structure) and proposalon the method of validation (for simple installation structure (inner diameter of canister: 400 mm))
Event

Deformations of lid/canister*1

No buffering structure on bottom

No plastic deformation, approx. 0-mm gap ()Buffering structure on bottom

No plastic deformation, approx. 0-mm gap ()Cannot simulate failure
behavior of the- O-ring byanalysis alone

Plastic deformation, approx. 0-mm gap ()-Collision
between stored
item and lid*3

No buffering structure on
bottomBuffering structure on
bottomToppling
Drop onto
canister*4

No buffering structure on bottom
Buffering structure on bottom

Validity evaluation method proposal

*1:
*2:
*3:
*4:

Rupture of O-ring

Rupture of tab-

No rupture*2
No rupture*2
No rupture*2

No collision between stored item and lid
Plastic deformation, approx. 0.9-mm gap ()-No rupture*2

Plastic deformation, approx. 0.02-mm gap ()Cannot simulate failure
behavior of the O-ring byanalysis alone

No rupture*2-

No rupture*2

Confirm the condition of the Oring after the test by element
test; confirm for leakages byleakage inspection

Confirm the condition of thefitting part after the test byelement test; confirm forleakages by leakage inspection

Plastic deformation, approx. 0-mm gap ()Measure the dimensions near the sealing
surface after the test by element test; confirm
for leakages by leakage inspection

Amount of gap is evaluated from the amount of deformation in the canister barrel and the lid near the sealing surface. Half the O-ring squeeze prescribed by JIS is set as theprovisional acceptance value of the gap (0.325 mm). When smaller than the provisional value,  is given (no leakage) and when greater than the value,  is given (leakage).
“No rupture” in the table means that the strain was 0.3 (30%) or less (provisional value) and was judged that rupture will not occur.
An event in which the stored item jumps up and collides with inner part of the lid when the canister is dropped vertically.
The collided canister is the subject of integrity evaluation of the confinement function.
©International Research Institute for Nuclear Decommissioning6. Implementation Details
6.3 Development of Safety Evaluation Methods and Safety Validation(2) Safety evaluation on structural strength: Study on lid structure (8/8)No.47(vi) Conclusion• Based on the latest study state of the containing, transfer, and storage WG, the Method Project Team, and theFundamental Technology Project Team, the canister handling flow (from retrieval to storage) was reviewed andevents that should be evaluated to ensure safety were re-identified.
• The lid structure proposals were studied from the aspect of confinement of fuel debris and workability in 1F.
• Regarding the identified events, lid damage modes were studied from the aspect of the lid’s confinement
property (release of fuel debris).
• Dynamic analysis was conducted for the damage modes and items to be evaluated were determined from the
aspect of lid integrity.
• Methods to evaluate the validity of the evaluation items and integrity of the lid were studied for the specified
evaluation items.

(vii) Future plans
Confirmation of feasibility including the handling of designed lid structures obtained from the study results of
FY2017, and the study on integrity (especially the simple structure), validity of the evaluation items concerning
the confinement property of the lid will be confirmed by element tests and analysis in FY2018. Specificallyspeaking, the following items will be implemented.
• Design and manufacturing of lid structure specimen• Handling property verification test
• Structural integrity verification test (load test)
Proposal of the lid structure design will be provided based on these results.
Regarding the remote lid closing technology, since there is no need for development in the technology itself if
the structure is remotely controllable, the handling confirmation test will confirm its structure.
In addition, separate from the lid structure study, a full-scale drop test scheduled for FY2020 (study of specimen
design and test plan in FY2018) will be prepared.
©International Research Institute for Nuclear Decommissioning6. Implementation Details
6.3 Development of Safety Evaluation Methods and Validation(3) Safety evaluation regarding aging degradation: Study on canister material (1/6)

No.48(i) Purpose
Purpose: Select canister material.
Fuel debris contains chloride ions attendant upon introduction of seawater. Taking this into account, material
candidates will be selected from the aspect of crevice corrosion and chloride SCC occurrence.

(2) Details of study up to FY2016⚫ Nickel-based and titanium-based alloys were selected as candidates, assuming the fuel debris properties andenvironments with a margin. Not only is it still impossible to completely eliminate the possibility of corrosion aslong as there is uncertainty in fuel debris properties, but it is also difficult to process and it was judged that it is notnecessarily a rational choice.
⚫ Taking this to account, the applicability of carbon steel and SUS316L that has been used before was evaluated. The
retrieval and storage under water (pool), retrieval in air, and environment of dry storage from the aspect of crevice
corrosion and SCC which are aging degradation modes of SUS316L were evaluated. The evaluation confirmed that
there is almost no degradation.
In addition, it was confirmed that sodium pentaborate that could possibly be used in terms of sub-criticality is also
effective regarding corrosion resistance.

(iii) Details of study in FY2017⚫ The entire surface of carbon steel has corroded. Although a method to maintain structural integrity is established bydefining corrosion margins from the estimated amount of corrosion, if stored in the pool for a long time, there is apossibility that corrosion occurred on the outer surface may cause difficulty in workability when re-opening the lid
after storage, etc.  Carbon steel is not suitable for long-term pool storage.
⚫ Additional evaluation was conducted mainly on SUS316L, which is effective for the still needed pool storage, since
there were unstudied transfer and drying environments and the possibility of storing drained fuel debris in air was
assumed. In these environments, condensation of chloride ions and high-temperature environment can be assumed.
As the period is short, it may be a difficult condition for evaluation.
Reference: Chloride ion concentration record (Fukaya et al., Current Status and Challenges Related to Corrosion Control of Containment and Piping at Fukushima
Daiichi, The 63rd Japan Conference on Materials and Environments, 2016)

Unit 1: 19 ppm (10/12/2012), Unit 2: 2.9 ppm (8/7/2013), Unit 3: < 1 ppm (10/22/2015)©International Research Institute for Nuclear Decommissioning6. Implementation Details
6.3 Development of Safety Evaluation Methods and Validation(3) Safety evaluation regarding aging degradation: Study on canister material (2/6)

No.49◆ Evaluation results of corrosive resistance of SUS316L in the canister’s usage environment
There is no issue in temperatures up to about 50C. However, crevice corrosion and cracks due to SCC become more apparent as temperaturerises and their occurrence cannot be denied.
Case

Flooded(fuel debris isflooded)Process

PeriodRetrievalMax. 10 days

TransferMax. 10 days

Wet storage
Max. 50 years(pool)

Dry(fuel debris is
dry)DryingMax. 10 days

TransferMax. 10 days

Dry storage Max. 50 years
RetrievalMax. 10 days

TransferMax. 10 days
Aerial(fuel debris is Wet storagedrained and wet) (pool)
AerialMax. 50 years
storage(Inside hotcell etc.)

: Low risk of corrosion

Temperature

Evaluation

Ambient temperature() There is a low risk of crevice corrosion and SCC.
() Ambient temperature or up to about 50C is an environment equivalent to that of retrieval, andthere is a low risk of crevice corrosion and SCC. Crevice corrosion may occur in higher
Ambient temperaturetemperatures; however, it is likely to be minor as the time period is short. Given the results of theor up to 150C
chloride ion concentration, cracks caused by SCC are likely to appear when the temperature
exceeds 70C but there is no concrete knowledge that allows us to determine a threshold. In
addition, SCC tends to become more apparent as temperature increases.
Max. 40C (temperature () Since the temperature inside the canister is 40C or lower like the retrieval environment, there is acontrol by pool water)
low risk of crevice corrosion and SCC.
() There is no problem when dried at ambient temperature or up to about 50C. Given the results ofthe chloride ion concentration, cracks caused by SCC is likely to appear when the temperature
exceeds 70C. However, since there is no concrete knowledge and cracks are dependent on the
behavior of the ion concentration, a specific threshold cannot be determined. In the water poolMax. 300C
inside the canister, crevice corrosion is also expected due to the rise of corrosion potential causedby hydrogen peroxide solution. However, since hydrogen peroxide is easy to escape into the gas
phase at high temperatures and since the period is short (about 10 days), it is believed that it isunlikely that it will be an issue. However, there is no specific knowledge to base our judgment.
Ambient temperature() Corrosion and SCC will not occur after fuel debris is dried.
or up to 150C
Ambient temperature() Corrosion and SCC will not occur after fuel debris is dried.
or up to 150C() There is a low risk of crevice corrosion and SCC. In addition, since the drying and concentration
Ambient temperatureenvironments are not active, the risk of SCC is low.
Ambient temperature() Same as the drying process.
or up to 150C
Up to 40C() The temperature is low so there is a low risk of crevice corrosion and SCC.

Ambient temperatureor up to 150C() Moisture remains in the fuel debris. When it contains moisture and is stored in a high-temperature
for a long period of time, hydrogen peroxide may generate. Therefore, compared with the drying
process, there is a higher risk of crevice corrosion and formation of cracks caused by SCC.

: Cannot deny the possibility of corrosion at present

Common matters) Atmosphere: Nitrogen or argon gas atmosphere (Wet storage pool is assumed to be installed in an atmosphericenvironment)
Water quality: Chloride ion concentration around 1 ppm (real results), hydrogen peroxide around 3 ppm (set as ambient
temperature based on documents)©International Research Institute for Nuclear Decommissioning
Concentration of hydrogen peroxide is affected by the environment (e.g., temperature).

6. Implementation Details
6.3 Development of Safety Evaluation Methods and Validation(3) Safety evaluation regarding aging degradation: Study on canister material (3/6)

No.50◆ Evaluation results of SUS316L crevice corrosion susceptibility⚫

In temperatures between about 30 and 50C, the risk of crevice corrosion is low even when the potential rises due to hydrogen
peroxide.
⚫ At about 80C, crevice corrosion may occur even with a chloride ion concentration of 1 ppm when the potential rises due tohydrogen peroxide.
-> Possibility of crevice corrosion remains. However, the rise in potential will be minor in a short period of time, and the option of
dealing with it by creating a design that will not increase the effect is available.
400

No susceptibility tocrevice corrosion

Repassivation potential for crevice
corrosion/mV vs. SSE

Estimated water quality inside
canister < 1 ppm as Cl

300

200

Susceptibility tocrevice corrosion

100

H2O2: Results of trial calculation on rise of
potential at 0.4 to 3.4 ppm
Approx. 250 to 280 mV vs. SSE (100C or lower)

Tolerance to crevice corrosion changes
depending on the concentration ofhydrogen peroxide but is almost constant
between 0.4 to 3.4 ppm*5*5: Kikuchi et al. (1987). Effects of Hydrogen Peroxide and Dissolved
Oxygen on Corrosion Potential of Type 304 Stainless Steel in HighTemperature Water. Corrosion Engineering. (Vol. 36, pp. 721–724)0
: 30C*1
: 40C *2
: 50C*2, 3◆: 80C*4

-100

-200
0.1

Source of data
*1: Imai et al. (1992).
Repassivation Potential for Crevice Corrosion of SUS 316 StainlessSteel/Teflon Ball Crevice. Surface Finishing. (43, 357–358)*2: IRID, 2016 Annual Research Report*3: Matsuhashi et al. (2009). Estimation of Crevice Corrosion Life Time for
Stainless Steels in Sea Water Environments. NIPPON STEEL &SUMITOMO METAL CORPORATION. (Vol. 389, 62–72)*4: Tsujikawa et al. (1980). Development of A New Test Method for Chloride
StressCorrosion Cracking of Stainless Steels and Its Application
to That of Type 316 in Dilute NaCl Solutions. (Volume 66, Issue 14, 2067–
2075, 263)

1
10
100
1000 10000 100000
Chloride ion concentration/ppm as Cl

Figure: Evaluation of crevice corrosion susceptibility (SUS316L) in each chlorine ion concentration using
crevice corrosion re-passivation potential©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.51

6.3 Development of Safety Evaluation Methods and Validation(3) Safety evaluation regarding aging degradation: Study on canister material (4/6)◆ Atmospheric stress corrosion cracking evaluation of SUS304 and SUS316L⚫ Stress corrosion cracking in the atmosphere is sensitive to relative humidity and depends on the temperature, relative
humidity, and concentration of chloride ion etc. in a solution. Crack sensitivity is lost below criticality conditions. (Refer tothe figure below)⚫ The chloride ions in the canister is thought to derive from sodium chloride. No cracks developed in SUS316L in the twoweek test (70C, 12,700 mg/m2 of NaCl deposit). From this it can be expected that by further reducing the concentration ofchloride ions, cracks will not appear in SUS316L in a short period of time, even in a 70C environment.
⚫ However, the threshold cannot be determined at this point because there is no knowledge in the existing documents thatquantitatively shows the relationship between chloride ion concentration and occurrence of SCC in SUS316L in thetemperature range (70C or higher) that is assumed for 1F canisters.
2 as Cl
70C,
RH
100 60%mg/m(i) Artificial
seawater(ii) NaCl

Amount of adhered NaCl:
13.9 µg/mm2 as Cl

Amount of adhered NaCl:
12.7 µg/mm2 as Cl= 12700 mg/m2

Rcrack (%): Crack rate, Nmean: Mean number of cracks, Lmean (mm): Mean length of largest crack

Figure: U-bend SCC test results of SUS304 and SUS316 that were dripped of chloride
solution and exposed to radiation for two weeks under constant temperature and humidity*1

Figure: U-bend SCC test results of SUS304 that was
dripped of NaCl solution and exposed to radiation for aweek under constant temperature and humidity*2*1: Shoji et al. (1986). Effects of Relative Humidity on Atmospheric Stress Corrosion Cracking of Stainless Steels. Corrosion Engineering. (35, 559-565)*2: Shoji et al. (1988). Effects of Deposit Amount of Chlorides on Atmospheric Stress Corrosion Cracking of Stainless Steels. Proceeding of the 35th Japan Corrosion Conference (Japan Society of Corrosion Engineering). (263)©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.52

6.3 Development of Safety Evaluation Methods and Validation(3) Safety evaluation regarding aging degradation: Study on canister material (5/6)◆ Atmospheric stress corrosion cracking evaluation of SUS304 and SUS316L⚫ The chloride ion concentration of stagnant water inside the primary containment vessel of Units 1 to 3 of 1F is 0.1 to19 ppm*1. From the diagram below, the concentration is assumed to be 10,000 times lower than that of a solution
with the amount of chlorine which would cause cracks and similar to that of fresh water. However, the drying
environment (adherence environment) differs from the environment inside the canister, so it is unknown how muchchloride ions will actually adhere inside the canister.
⚫ According to the report in the previous section, cracks did not form on SUS316L during the two-week test under theadherence condition of 12,700 mg/m2 of sodium chloride (70C, 60% RH) (condition in which cracks formed on
SUS304). Considering that the water is diluted by 10,000 times with respect to the chloride ion concentration of thesolution in which cracks form, it can be expected that the risk of cracks is extremely low.
⚫ On the other hand, as mentioned in the previous section, since cracks caused by SCC tend to appear more astemperature rises, the possibility of cracks forming on SUS316L at 70C or higher cannot be denied.
Logarithm of amount of adhered “sea salt,”
log W [g/m2]

Shimizu city, Shizuoka Prefecture

Instrument Rooftop
screenNishihara, Okinawa Prefecture

50-cm above sea level,facing downwards

Above
the roof90 south(below the roof)

Nogi, Tochigi Prefecture
OutsideInsideSeawater spraying
Range of water quality of 1F Units 1
to 3

10-3 seawater
Tap water, clear
water

10-2 seawater

Range in which cracks have not formed

Seawater on SUS316L in the two-week test
according to the past report

5% NaCl solution

Figure: Amount of adhered sea salt in various environments*2*1: Yuichi Fukaya, Katsuhiko Kumagai. (2016). Current Status and Challenges Related to Corrosion Control of Containment and Piping at Fukushima Daiichi. The 63rd Japan Conference on Materials and Environments. (C-101)*2: Shinohara. (1998). Evaluation of Corrosivity of Atmosphere (Monitoring of Corrosive Environment and Concepts on Corrosion Classification). Retrieved from JWTC (Proceedingof the1998
WeatheringInstituteTechnology©International
Research
forResearch
Nuclear Decommissioning
Presentation. (No. 3, 15-24)

6. Implementation Details
6.3 Development of Safety Evaluation Methods and Validation(3) Safety evaluation regarding aging degradation: Study on canister material (6/6)

No.53(iv) Future plans
In a high-temperature environment, risk of crevice corrosion and SCC in SUS316L material tend to increase.
However, the risk can be dealt with by restricting the canister shape and handling method to prevent the effect ofcorrosion from surfacing, and a margin can be set so that minor damages on the canister will not effect safety. In
addition, nickel-based alloys etc. with better corrosion resistance have issues such as difficulty in processing. For
these reasons, SUS316L will be a candidate material and study of countermeasures (limiting conditions) will beprioritized.
Even if element tests are conducted as follows, the corrosive environment itself depends on the canister’s design
condition. Therefore, to confirm the possibility and degree of corrosion, the necessity of the test again based on theabove study will be judged.
• Water resides for a long period of time in a high-temperature in aerial storage*1 and there is a possibility thathydrogen peroxide solution that greatly effects the occurrence of crevice corrosion will form. On the other hand,since the concentration of the hydrogen peroxide solution depends on the state of water, it is predicted that it
will be affected by the canister shape. In addition, the assumed period of aerial storage*1 is not clear.
• Although it is believed there is margin in the concentration of chloride ions in the drying process, it is assumed
that it will be affected by the drying method and canister shape.
*1: The fuel debris is drained and wet.

©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.54

6.3 Development of Safety Evaluation Methods and Safety Validation(4) Safety validation on countermeasures against hydrogen gas; a. Verification of the hydrogen generation
amount: Confirmation on effect of gas phase (gamma-ray irradiation test) (1/10)(i) Purpose
Up to FY2016, the hydrogen generation that considers the 1F water quality conditions (seawatercomponent, iodine, concrete) has been studied by gamma-ray irradiation test. As a result, the hydrogengeneration under gamma-ray irradiation can be evaluated using the primary G value of hydrogen generation(0.45 molecules/100eV), even when taking 1F water quality conditions into account.
However, when the gas-liquid ratio is high, there is a possibility that the decline in pH of the liquid phasecaused by the nitric acid generated by the radiolysis and oxidization of nitrogen in the gas phase effects thehydrogen generation.
In addition, in a condition where the hydrogen generation amount is small and the error will be large,
there were cases where the apparent G value exceeded the primary G value of hydrogen generation (0.45molecules/100eV).
Therefore, an additional gamma-ray irradiation test was conducted to confirm the effect of nitric acidproduction as well as evaluate the reason the apparent G value exceeded the primary G value of hydrogen
genetation (0.45 molecules/100eV). It was aimed to reflect it as necessary to evaluate hydrogen generation
amount.

(ii)Implementation details
Tests with different gasses in the gas phase (air, nitrogen, argon) were conducted to confirm its effect. In
addition, considering the current chloride ion concentration, tests with reduced chloride ion concentration
were conducted to confirm the effect of chloride ion.
Regarding the FY2016 test, it was confirmed the validity of the apparent G value obtained under thecondition which the apparent G value exceeded the primary G value of hydrogen generation (0.45molecules/100eV).
©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.55
6.3 Development of Safety Evaluation Methods and Safety Validation(4) Safety validation on countermeasures against hydrogen gas; a. Verification of the hydrogen generation
amount: Confirmation on effect of gas phase (gamma-ray irradiation test) (2/10)(iii) Test method
Similarly to FY2016, gamma-ray was irradiated from the outside to the container which the test water wassealed in and the apparent G value was calculated from the pressure variation.
Temperature:
ambient temperature

Co60 gamma-ray
source(center)Pressure
indicator

Valve
Valve

Co60 gamma-ray source

Pressure indicator

Gas
phase

Valve

Valve
Sample connector

Gammaray

Test
water

Safety valve

Test container (300 mL)

Amount of hydrogengenerated

Figure: Overview of the test system

Jig

Calculate G value
from gradient b

Test container

Figure: Test conditions

Absorbed dose
Figure: Calculation method of apparent G value©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.56
6.3 Development of Safety Evaluation Methods and Safety Validation(4) Safety validation on countermeasures against hydrogen gas; a. Verification of the hydrogen generation
amount: Confirmation on effect of gas phase (gamma-ray irradiation test) (3/10)(iv) effect assessment test on gas phase (1/2)
By FY2016, it was revealed that the decline in pH of the liquid phase caused by the nitric acid generated by radiolysis
and oxidization of nitrogen in the gas phase under gamma-ray irradiation promotes the hydrogen generation in a theseawater component coexistence system.
-> The tests that charged the gas type in the gas phase conduct to confirm the effect on suppression of nitric acid andhydrogen generation.
Table: Test conditions for effect assessment of gas phase

Case

(i)

(ii)

(iii)

(iv)

(v)

Gas in gas phase

Seawater component
concentration*1(chloride ion concentration) [mol/L]

Iodide ion concentration[mol/L]

Gas-liquid ratio*2[%]

Temperature

Absorbed dose [kGy]

Number of
tests

Air

2.8  10−3 *3

1.0  10−4 *4900

Ambient
temperature*5

Max. 1,000*52

Nitrogen

2.8  10−3 *3

1.0  10−4 *4900

Ambient
temperature*5

Max. 1,000*52

Nitrogen

2.8  10−3 *30900

Ambient
temperature*5

Max. 1,000*5

2

Argon

2.8  10−3 *3

1.0  10−4 *4900

Ambient
temperature*5

Max. 1,000*5

2

Argon

2.8  10−3 *30900

Ambient
temperature*5

Max. 1,000*5

2

Notes

Same conditions as FY2016

Conditions that exclude initial oxygen whichcauses nitric acid generation

Conditions that exclude nitrogen which causes
nitric acid generation*1: Uses diluted artificial seawater*2: Volume of gas phase / volume of liquid phase*3: Equivalent to 100 ppm as chloride ion concentration, determined by the water quality criteria*6 of 1F stagnant water*4: Determined on the assumption that 10% of the iodine inventory*7 in the core fuel of 1F leaches*5: Same conditions as FY2016*6: Council for the Decommissioning of TEPCO’s Fukushima Daiichi. (Mar. 28, 2013). Evaluation on the Early Realization of Circulation Loop in the Building and Reduction of Circulation Line. 1st Secretariat Meeting. (Material 31, Circulating Injection Cooling)*7: Kenji Nishihara, Hiroki Iwamoto, Kenya Suyama. (Sept. 2012). Estimation of Fuel Compositions in Fukushima Daiichi Nuclear Power Plant. Japan Atomic Energy Agency. (JAEA-Data/Code2012-18)

Case (i)-1 (with correction)
Case (i)-2 (with correction)

Absorbed dose [kGy](a) Case (i)Case (ii)-1

Pressure variation [Mpa]

Pressure variation [Mpa]

Case (i)-2Case (ii)-2

Pressure variation [Mpa]

To reduce the effect of variances of thetwo tests, the measured value ofpressures were corrected using themeasurement results of hydrogenconcentration

Case (i)-1Case (ii)-1 (with correction)
Case (ii)-2 (with correction)

Absorbed dose [kGy](b) Case (ii)Case (iv)-1
Case (iv)-2
Case (iv)-1 (with correction)
Case (iv)-2 (with correction)

Absorbed dose [kGy](c) Case (iv)

©International
Research Institute for Nuclear Decommissioning
Figure: Example of pressure measurement (includes correction by hydrogen generation
amount)6. Implementation Details

No.57
6.3 Development of Safety Evaluation Methods and Safety Validation(4) Safety validation on countermeasures against hydrogen gas; a. Verification of the hydrogen generation
amount: Confirmation on effect of gas phase (gamma-ray irradiation test) (4/10)(iv) effect assessment test on gas phase (2/2)
a. When iodine concentration is high: Regardless of the pH (the degree of nitric acid production), the apparent G
values are about 0.20 and same when errors are considered.
-> Under iodine-dominant conditions, it is assumed that replacement of gas in the gas phase isineffective.
b. When iodine concentration is low: There is a possibility that, in conditions which the pH declines (nitric acidgenerates), hydrogen generation is promoted by seawater component andin conditions which the pH does not change (nitric acid does not
generate), hydrogen generation is suppressed.
-> It is assumed that replacement of gas in the gas phase is effective.
However, iodine is assumed to leach out from the fuel debris and cannot be controlled.
-> It is necessary to consider the presence of a certain amount of iodine.
Table: Test results of effect assessment of gas phase (evaluated based on the value corrected by the measured value of hydrogen concentration in gas phase*1)

Test conditions

Test results

Seawater component
concentration*2(chloride ion
concentration) [mol/L]

Iodide ion
concentration[mol/L]

Gas-liquidratio*3[%]

Hydrogenconcentrationin gas phase*4[vol/%]

a. Comparison between
conditions with highiodine concentration

Nitric acid ion
concentration inliquid phase*4[mg/L]

pH of test
water*4, *5[-]

Apparent G
value*6[molecules/100eV]

Case2

Gas in gas
phase

(i)

Air

2.8  10−3 *7

1.0  10−4 *8

900

3.78/4.05

130/130

3.4/3.3

0.09  0.23

(ii)

Nitrogen

2.8  10−3 *7

1.0  10−4 *8

900

3.92/4.82

92/1104.4/3.4

0.11  0.29

(iii)

Nitrogen

2.8  10−3 *70900

2.44/4.13

120/140

3.1/3.1

0.09  0.04

(iv)

Argon

2.8  10−3 *7

1.0  10−4 *8900

6.53/7.20

13/13

6.1/6.4

0.20  0.05

(v)

Argon

2.8  10−3 *70900

1.94/2.93

14/16

5.6/6.4

0.05  0.02

*1: Corrected the measured value of pressure using the measurement results of hydrogen concentration in the gas phase*2: Uses diluted artificial seawater*3: Volume of gas phase / volume of liquid phase*4: Two tests are conducted per case and both test results are shown*5: Value at water temperature of 25C, pH value measured before the test was 6.5 (almost neutral)*6: Results of the two tests are used as a single data to calculate the apparent G value*7: Equivalent to 100 ppm in chloride ion concentration, determined by the water quality criteria *9 of 1F stagnant water*8: Determined on the assumption that 10% of the iodine inventory*10 in the core fuel of 1F leaches*9: Council for the Decommissioning of TEPCO’s Fukushima Daiichi. (Mar. 28, 2013). Evaluation on the Early Realization of Circulation Loop in the Building and Reduction of Circulation Line. 1st Secretariat Meeting. (Material 3-1,
Circulating Injection Cooling)*10: Kenji Nishihara, Hiroki Iwamoto, Kenya Suyama. (Sept. 2012). Estimation of Fuel Compositions in Fukushima Daiichi Nuclear Power Plant. Japan Atomic Energy Agency. (JAEA-Data/Code2012-18)

b. Comparison between
conditions with low
iodine concentration©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.58
6.3 Development of Safety Evaluation Methods and Safety Validation(4) Safety validation on countermeasures against hydrogen gas; a. Verification of the hydrogen generation
amount : Confirmation on effect of gas phase (gamma-ray irradiation test) (5/10)(v) effect assessment of seawater (1/2)
In the test conducted up to FY2016, the concentration of chloride ions was determined based on the water
quality criteria of 1F stagnant water in the reactor building. Here considering the results of the investigation
inside the PCV, the effect on hydrogen generation with reduced concentration of seawater component will
be confirmed.
-> In the tests up to FY2016, the chloride ion concentration was 2.8  10−3 mol/L (equivalent to 100 ppm*1).
Tests were conducted with concentration of 0 mol/L (0 ppm) and 5.6  10−4 mol/L (equivalent to 20 ppm*2).
Table: Test conditions for effect assessment of seawaterGas in gas phase

Seawater component
concentration*3(chloride ion concentration)[mol/L]

Iodide ion concentration[mol/L]

Gas-liquidratio*4[%]

Temperature

Absorbed dose
[kGy]

Number of
tests

(vi)

Air

0 (0 ppm)

0*6

900

Ambient
temperature*7

Max. 1,000*7

2

(vii)

Air

5.6  10−4 *5

0*6

900

Ambient
temperature*7

Max. 1,000*7

2

Case

Notes

Case (vi)-1
Case (vi)-2Case (vi)-1 (with correction)
Case (vi)-2 (with correction)

Absorbed dose [kGy](a) Case (vi)Pressure variation [Mpa]

Pressure variation [Mpa]*1: Determined by the water quality criteria*8 of 1F stagnant water*2: Determined by the measurement results of Unit 1 (approx. 19 ppm)*9*3: Uses diluted artificial seawater*4: Volume of gas phase / volume of liquid phase*5: Equivalent to 20 ppm in chloride ion concentration*6: Iodine ion is not added so that effect of seawater can be observed*7: Same conditions as FY2016*8: Council for the Decommissioning of TEPCO’s Fukushima Daiichi. (Mar. 28, 2013). Evaluation on the Early Realization of Circulation Loop in the Building and Reduction of Circulation Line. 1st Secretariat Meeting. (Material 3-1, Circulating
Injection Cooling)*9: Kenji Nishihara, Hiroki Iwamoto, Kenya Suyama. (Sept. 2012). Estimation of Fuel Compositions in Fukushima Daiichi Nuclear Power Plant. Japan Atomic Energy Agency. (JAEA-Data/Code2012-18)
Case (vii)-1
Case (vii)-2Case (vii)-1 (with correction)
Case (vii)-2 (with correction)

Absorbed dose [kGy](b) Case (vii)©International
Figure: Example of pressure measurement (includes correction by hydrogen generation
amountResearch Institute for Nuclear Decommissioning6. Implementation Details

No.59
6.3 Development of Safety Evaluation Methods and Safety Validation(4) Safety validation on countermeasures against hydrogen gas; a. Verification of the hydrogen generation
amount: Confirmation on effect of gas phase (gamma-ray irradiation test) (6/10)(v) effect assessment of seawater (2/2)
Comparison between this test result and the test result of FY2016 shows that, even under a condition with
reduced pH that would promote hydrogen generation, the apparent G value will be about 0 if theconcentration of the seawater component is low (chloride ion concentration of 5.6 × 10-4 mol/L (equivalent to
20 ppm) or lower).
-> It was revealed that when the concentrations of seawater component and iodine are sufficiently low,
the apparent G value will reduce due to progress in reaction of hydrogen recombination.
Table: Test results for effect assessment of seawater (evaluated based on the value corrected by the measured value ofhydrogen concentration in gas phase*1 (results of FY2017 only))
Test conditions

Test resultsCase

Gas ingas
phase

Seawater component
concentration*2(chloride ion
concentration) [mol/L]

Iodide ion
concentration[mol/L]

Gas-liquidratio*3[%]

Hydrogenconcentration ingas phase*4[vol/%]

Nitric acid ion
concentration inliquid phase*4 [mg/L]

pH of test
water*4, *5[-]

Apparent G
value*6[molecules/100eV]

(vi)

Air

0

0*7

900

0.80/0.68

160/160

2.9/2.9

0

(vii)

Air

5.6  10−4 *8

0*7

900

0.93/1.68

150/130

3.0/2.9

0.01 ± 0.02

Reference(FY2016 test results)Air

2.8  10−3 *9

0

900

10.1

270

2.8

0.21 ± 0.01

*1: Corrected the measured value of pressure using the measurement results of hydrogen concentration in the gas phase*2: Uses diluted artificial seawater*3: Volume of gas phase / volume of liquid phase*4: Two tests are conducted per case and both test results are shown; however, only one test is conducted in FY2016*5: Value at water temperature of 25C, pH value measured before the test was 6.5 (almost neutral)*6: Results of the two tests are used as a single data to calculate the apparent G value; however, evaluation in FY2016 uses the result from a single test
*7: Iodine ion is not added so that effect of seawater can be observed*8: Determined on the assumption that 10% of the iodine inventory*10 in the core fuel of 1F leaches*9: Equivalent to 100 ppm in chloride ion concentration, determined by the water quality criteria*11 of 1F stagnant water*10: Kenji Nishihara, Hiroki Iwamoto, Kenya Suyama. (Sept. 2012). Estimation of Fuel Compositions in Fukushima Daiichi Nuclear Power Plant. Japan Atomic Energy Agency. (JAEA-Data/Code2012-18)*11: Council for the Decommissioning of TEPCO’s Fukushima Daiichi. (Mar. 28, 2013). Evaluation on the Early Realization of Circulation Loop in the Building and Reduction of Circulation Line. 1st Secretariat Meeting. (Material 31, Circulating Injection Cooling)©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.60
6.3 Development of Safety Evaluation Methods and Safety Validation(4) Safety validation on countermeasures against hydrogen gas; a. Verification of the hydrogen generation
amount: Confirmation on effect of gas phase (gamma-ray irradiation test) (7/10)(vi) Verification of validity of G value (1/2)
In the FY2016 test, there were several cases where the apparent G values exceeded the primary G value ofhydrogen generation (0.45 molecules/100eV).
It was revealed that, in conditions where rise in pressure is small, the apparent G value is more susceptible
to changes in atmospheric pressure and temperature, and there are issues in measurement and evaluation
accuracy.
-> Regarding the behavior in the initial phase of irradiation which has great effect on the evaluation of
apparent G value, it was aimed to improve accuracy by measuring the hydrogen concentration directly.
*: In previous evaluations on the apparent G value by pressure measurement, changes over time can be observed without disturbing the system. It
benefits us to see the changes in reaction when evaluating a closed system such as the canister during transfer. However, when the amount ofhydrogen generation is small, changes in atmospheric pressure and temperature will be similar to the amount of hydrogen generation. This effects themeasurement and causes greater errors. The method of measuring the hydrogen concentration at regular time intervals can evaluate the amount ofhydrogen generation at that point in time very accurately, but sampling requirements disturb the system. In closed systems, the test would have to beterminated.

Table: Test conditions for verification of validity of G value
Case

Gas in gas
phase

Seawater component
concentration*1(chloride ion concentration)[mol/L]

Iodide ion
concentration[mol/L]

Gas-liquidratio*2[%]

Temperature

Absorbed dose
[kGy]

Numberof tests(viii)

Air

2.8  10−3 *3

1.0  10−4 *4500

Ambient
temperature*5

10

3

(ix)

Air

2.8  10−3 *3

1.0  10−4 *4500

Ambient
temperature*5

20

3

(x)

Air

2.8  10−3 *3

1.0  10−4 *4500

Ambient
temperature*5

30

3

(xi)

Air

2.8  10−3 *3

1.0  10−4 *4500

Ambient
temperature*5

300

3

Notes

Pressure was measured for reference in
one of the three tests*1: Uses diluted artificial seawater*2: Volume of gas phase / volume of liquid phase*3: Equivalent to 100 ppm in chloride ion concentration, determined by the water quality criteria *6 of 1F stagnant water*4: Determined on the assumption that 10% of the iodine inventory*7 in the core fuel of 1F leaches*5: Same conditions as FY2016*6: Council for the Decommissioning of TEPCO’s Fukushima Daiichi. (Mar. 28, 2013). Evaluation on the Early Realization of Circulation Loop in the Building and Reduction of Circulation Line. 1st Secretariat Meeting. (Material 3-1, Circulating Injection Cooling)*7: Kenji Nishihara, Hiroki Iwamoto, Kenya Suyama. (Sept. 2012). Estimation of Fuel Compositions in Fukushima Daiichi Nuclear Power Plant. Japan Atomic Energy Agency. (JAEA-Data/Code2012-18)©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.61
6.3 Development of Safety Evaluation Methods and Safety Validation(4) Safety validation on countermeasures against hydrogen gas; a. Verification of the hydrogen generation
amount: Confirmation on effect of gas phase (gamma-ray irradiation test) (8/10)(vi) Verification of validity of G value (2/2)
In the test conducted in FY2017, the apparent G value evaluated based on the measurement results ofhydrogen concentration was 0.32 molecules/100eV. This is lower than the primary G value of hydrogengeneration (0.45 molecules/100eV) and is about half the test result of FY2016 (0.57 molecules/100eV).
In addition, the apparent G value evaluated from the pressure measurement results as a reference value*1 was
0.23 molecules/100eV. This was almost the same as the apparent value evaluated based on the measurement
results of hydrogen concentration.
-> It is assumed that the apparent G value in the FY2016 test results may have been evaluated larger.

Amount of hydrogengeneration increased
relatively linear

Absorbed dose (kGy)

Figure: Dependency of hydrogen generation amount to absorbed dose
(Cases (viii) to (xi))

Apparent G value [molecules/100eV]

Hydrogen generation amount (10−5 mol)

*1: The data is of the initial phase of irradiation up to 300 kGy. There are only three points of pressure measurement and the apparent G values are reference values.
FY2016 test results
may have been
evaluated larger

FY2016 test results
Hishida M, 2016 (pressure measurement)
FY2017 test results (gas analysis)
FY2017 test results (pressure measurement (withcorrection))

G value at initial generation (0.45)

Apparent G values evaluated from thehydrogen concentration measurement
result and pressure measurement results
were both about the same

Gas-liquid ratio [%]Figure: Dependency of apparent G value to gas-liquid ratio (error: 2σ)(iodine ion concentration: 1 10−4 mol/L)©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.62
6.3 Development of Safety Evaluation Methods and Safety Validation(4) Safety validation on countermeasures against hydrogen gas; a. Verification of the hydrogen generation
amount: Confirmation on effect of gas phase (gamma-ray irradiation test) (9/10)(vii) Comparison with analysis
The test result were compared to the analytical results using the evaluation method that combines theradiolysis model and the liquid-gas distribution with the equation of state and Henry’s law (same as method
used until FY2016). The primary G value of hydrogen generation (0.45 molecules/100eV) was used in theanalysis.

∆ Hydrogen partial pressure∆ Gage pressure∆ Gage pressure

Hydrogen concentration

Absorbed dose [MGy](a) Case (i)(Chloride ion concentration: 2.8  10−3
mol/L, iodide ion concentration: 1  10−4
mol/L, gas phase gas: air, gas-liquid ratio:
900%)

Gage pressure (experimental value): Case (v)-2Hydrogen concentration

Absorbed dose [MGy](a) Case (v)(Chloride ion concentration: 5.6  10−4 mol/L,
no added iodine (0 mol/L), gas phase gas:argon, gas-liquid ratio: 900%)

Gage pressure (experimental value): case (vii)-1
Gage pressure (experimental value): case (vii)-2Hydrogen concentration

Pressure variation [Mpa]Gage pressure (experimental value): case (i)-2Gage pressure (experimental value): Case (v)-1

Pressure variation [Mpa]

Pressure variation [Mpa]Gage pressure (experimental value): case (i)-1

Hydrogen concentration in liquid phase [ppb]∆ Hydrogen partial pressure∆ Gage pressure

Hydrogen concentration in liquid phase [ppb]∆ Hydrogen partial pressure

Absorbed dose [MGy](c) Case (vii)(Chloride ion concentration: 5.6  10−4
mol/L, no added iodine (0 mol/L), gas
phase gas: air, gas-liquid ratio: 900%)

Figure: Comparison between analysis results and measured hydrogen generation amount©International Research Institute for Nuclear Decommissioning

Hydrogen concentration in liquid phase [ppb]-> By using the primary G value of hydrogen generation (0.45 molecules/100eV) in each of the conditions,the analytical values enveloped the test results. This shows that the amount can be evaluated using the
primary G value of hydrogen generation (0.45 molecules/100 eV) under gamma-ray irradiation, eventhough considering conditions of water quality in the 1F.

6. Implementation Details

No.63

6.3 Development of Safety Evaluation Methods and Safety Validation(4) Safety validation on countermeasures against hydrogen gas; a. Verification of the hydrogen generation
amount: Confirmation on effect of gas phase (gamma-ray irradiation test) (10/10)(viii) Conclusion• Chloride and effect of gas phase showed that;(a) When iodine concentration is high, hydrogen generation is dominantly affected by iodine and not by the typeof gas in the gas phase.
(b) When iodine concentration is low, hydrogen generation is accelerated by the decline in pH caused by nitricacid production, so replacing gas with argon which does not produce nitric acid is effective in suppressinghydrogen generation.
• In a system which iodine does not coexist, the apparent G value of hydrogen generation will decrease when theseawater component has a chloride ion concentration of about 20 ppm, even if the gas phase is atmospheric.
• A high G value was observed in FY2016. When the amount of hydrogen generation is small, the error becomeslarger in pressure measurement due to changes in atmospheric pressure and temperature, which may cause G
value to be evaluated larger. It was confirmed that the apparent G value, evaluated based on the measurement
results of hydrogen concentration conducted in FY2017, was lower than the primary G value of hydrogengeneration (0.45 molecules/100eV).
• It was predicted that the evaluation method that combines the radiolysis model of water with the primary G valueof hydrogen generation (0.45 molecules/100eV) and the liquid-gas distribution with equation of state and Henry’s
law can evaluate the amount of hydrogen generation that envelope the test results.
Based on the above results, as a result of evaluating the hydrogen generation amount under irradiation of gamma-ray
with consideration to 1F water quality conditions (seawater, iodine, concrete) that have been considered up to FY2017,it was revealed that the hydrogen generation amount under gamma-ray irradiation can be evaluated using the primary
G value of hydrogen generation (0.45 molecules/100eV), even when taking 1F water quality conditions into
consideration.
Evaluation related to gamma-rays had finished in FY2017. Along with the results of the separately implemented
evaluation on effect of alpha-rays, it was aimed to appropriately evaluate hydrogen generation inside 1F fuel debris
canisters.
©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.64

6.3 Development of Safety Evaluation Methods and Safety Validation(4) Safety validation on countermeasures against hydrogen gas; a. Verification of the amount generatedhydrogen generation amount: Confirmation of effect of alpha-ray (1/6)(i) Purpose
Thus far, the data have been improved by the gamma-ray irradiation tests to evaluate the method to predict
the amount of hydrogen generation inside the canister. However, effect of alpha-rays on fuel debris needs tobe considered as claddings cannot be expected to shield the radiation*1. Therefore, the effect of alpha-ray
under 1F conditions will be confirmed by conducting tests that use spent fuel pellet pieces.
In FY2018, based on these test results, the amount of hydrogen generation inside the canister will be
predicted by hydrogen generation amount evaluation suitable for 1F condition (water quality and fuel debris
conditions). Specifically, the hydrogen generation amount during transfer will be predicted and will be used
to determine the time limit for transfer.
*1: Although beta-ray may have an effect when shielding by cladding cannot be expected, it is known that the effect of LET (linear energy
transfer) of beta-rays is similar to that of gamma-rays (Spinks, J.W.T. and R.J. Woods, An introduction to radiation chemistry, 1990, Wiley)and knowledge on gamma-rays may be applicable in terms of hydrogen generation by radiolysis of water.

(ii)Implementation details
The hydrogen generation amount by gamma-ray was estimated by the gamma-ray irradiation test andanalysis up to FY2016 with consideration to the 1F conditions (water quality and mixed concrete pieces).
Since alpha-rays is expected to contribute in hydrogen generation, tests using spent fuel pellet pieces will
be conducted from FY2017 (continued in FY2018). The main effective factors identified are shown below.
• Presence or absence of alpha-ray (conducted in FY2017)• Effect of water volume (conducted in FY2017 and FY2018) (FY2017 is preliminary study)• Effect of particle size (conducted in FY2018)• Effect of moisture in concrete (associated item when collecting MCCI products) (conducted in FY2018)©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.65
6.3 Development of Safety Evaluation Methods and Safety Validation(4) Safety validation on countermeasures against hydrogen gas; a. Verification of hydrogen generation
amount: Confirmation of effect of alpha-ray (2/6)(iii) Test details
Using spent fuel pellet pieces, the hydrogen generation amount is measured for both conditions when alpha-ray
is considered and when not. From the difference between the two amounts, contribution of the alpha-ray isverified. In addition, a preliminary test is conducted with a different water volume to roughly understand its
effect which is used to set out the test conditions for FY2018.
Table: Test conditions
Items
Test system

Cases 1 and 2

Case 3

Immerse pellet pieces insufficient amount of testwater

No contact between pellet pieces andtest water (shielding of alpha-ray)Specimen weight/particle size

Test water

Amount of water

Case 4

Seawater component
concentration

Immerse pellet pieces in small amountof test water

Sized after being washed and dried (approx. 100C, total of 6.5
hours)

About 80 g/20 to 300 μm
100 mL

Notes
Case 2 has same conditions since it is for confirming the
reproducibility of case 1

8 mL (50 vol.% of pellet pieces)

5.6  10−4 mol/L (Equivalent to 20 ppm) as chloride ion concentration

50 vol.% is based on the draining test conducted in FY2016
Determined by the measurement results of IF of Unit 1 (approx. 19ppm)*1

Iodide ion concentration1  10−4 mol/L

Assumed that 10% of the iodine inventory*2 will leach into the
water

pH

No adjustment

Measured before and after test

Type of gas in gas phase/initialinternal pressure

Period of immersion
Test temperature

Atmosphere/approximately the atmospheric pressure (no added pressure)

The pressure is atmospheric assuming a case where gas cannot
be replaced

Max. 20 days

Specified with margin for the assumed transfer period of 10 days

Ambient temperature

Temperature inside container (measurement during test): 18.3C
to 24.2C*1: Tokyo Electric Power Company, Inc. (Oct. 22, 2012). Unit 1 Primary Containment Vessel (PCV) Internal Investigation Results.
*2: Kenji Nishihara, Hiroki Iwamoto, Kenya Suyama. (Sept. 2012). Estimation of Fuel Compositions in Fukushima Daiichi Nuclear Power Plant. Japan Atomic Energy Agency. (JAEA-Data/Code2012-18)

Inner container (SUS316L)

Inner container(SUS316L)
Test water(considers seawater andiodine)(low water volume in case 4)(a) Cases 1, 2, and 4

Outercontainer

Outer
container
SUS316L leaf
(thickness: 20 μm)
PelletpiecesPelletpiecesFigure: Overview of the test

SUS316L mesh and frame(minimize shielding part asmuch as possible)
Test water(considers seawater andiodine)(b)Case 3 Research Institute for Nuclear Decommissioning©International6. Implementation Details
6.3 Development of Safety Evaluation Methods and Safety Validation(4) Safety validation on countermeasures against hydrogen gas; a. Verification of the hydrogengeneration amount: Confirmation of effect of alpha-ray (3/6)(iv) Test conditions

No.66

Change of pressure in the container during the test was measured. After the test, the gas and test water were analyzed.
Compoundgage

Micromanometer

Compoundgage

Micromanometer

Safetyvalve

Piping for
gas/test
water supply

Piping for gas
collection

Thermocouple

Safety valve
Piping for gas
collection

Piping for gas/testwater supply

Test
container

Test water

Test container
Thermocouple

Spent fuelPellet pieces

Innercontainer

Figure: Appearance oftest equipment

Figure: Test equipment diagram

Table: Specifications of spent fuelSpent fuel used for testing*1 (result of combustion calculation)(Reference) 1F fuel (e.g., Unit 1)*2

Approx. 57 GWd/t (fuel element average)

Approx. 26 GWd/t (reactor vessel average)

Cooling period

About 15 years

10 years

radioactivity

2.12  107 GBq/t

1.08  107 GBq/t

Decay heat

2.18  103 W/t

7.97  102 W/t

Gamma-ray intensity

5.56  1015 photon/s/t

3.04  1015 photon/s/t

Neutron intensity

1.86  109 neutron/s/t

1.07  108 neutron/s/t

Burnup*1: Used fuel pellet pieces separated from BWR 9 × 9 fuel (type A) LUA (Lead Use Assembly).
*2: Kenji Nishihara, Hiroki Iwamoto, Kenya Suyama. (Sept. 2012). Estimation of Fuel Compositions in Fukushima Daiichi Nuclear Power Plant. Japan Atomic Energy Agency. (JAEAData/Code2012-18)©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.67
6.3 Development of Safety Evaluation Methods and Safety Validation(4) Safety validation on countermeasures against hydrogen gas; a. Verification of the amount generatedhydrogen generation amount: Confirmation of effect of alpha-ray (4/6)(v) Test results
The hydrogen generation rate was evaluated and confirmed from the measurement results of hydrogenconcentration inside the gas phase after the test.
• In the condition that considers alpha-rays (cases 1 and 2) and condition that shields alpha-rays (case 3),
the former had the hydrogen generation rate 10 times or greater than that of the latter.
• The hydrogen generation rate under the condition in which fuel pellets were immersed in small amount ofwater (case4), was greater than that under condition in which fuel pellets were immersed in sufficientamount of water (case1 and 2).
In this test condition, the pressure variation during the test is small and the hydrogen generation rate couldn’t
be evaluated from the pressure measurement in all the cases.

Indicated value of micro-manometer> 0 kPa
Indicated value of micro-manometer= 0 kPa (reference value)

Time (day)(a) Case 1

Error bar:  0.56 kPa
Dashed line: 0  0.56 kPa

Hydrogen generation rate*1(Reference) Hydrogen concentration in gas
phase (measured value)

Case 1

Approx. 7.5  10−8 L/h/gUO2Approx. 0.73%

Case 2

Approx. 6.0  10−8 L/h/gUO2Approx. 0.51%

Case 3

Approx. 4.6  10−9 L/h/gUO2Approx. 0.04%

Case 4

Approx. 1.1  10−7 L/h/gUO2Approx. 0.78%

Indicated value of micro-manometer> 0 kPa
Indicated value of micro-manometer= 0 kPa (reference value)

Error bar:  0.56 kPa
Dashed line: 0  0.56 kPa

Pressure variation inside
test container (kPa,converted value at 25C)

Pressure variation inside
test container (kPa,converted value at 25C)*2: Inner diameter of canister: ⌀220 mm  800 mm, fueldebris density: 11 t/m3 (equivalent to UO2), filling rate:
50 vol.%, water volume: 50 vol.% relative to fuel debris
volume.

Table: Evaluation results on the hydrogen generation rate

Pressure variation inside
test container (kPa,converted value at 25 C)<Reference evaluation>
For example, in case 4, it takes about 2.5 days
for the hydrogen concentration inside thecanister to reach 4% (lower explosion limit)
when evaluated under conservative
conditions*2.
-> There is a prospect that safe transportationis possible by reviewing conditions etc.

Indicated value of micro-manometer>0
Indicated value of micro-manometer= 0 (reference value)

Time (day)Time (day)(b) Case 3(c) Case 4

Figure: Pressure variation inside the test container during the test (measured value)*1: Calculated fromthe hydrogenconcentration in
the gas phase(measured value),
volume of test
container, weightof pellet piece,
test period, etc.
(converted value
at 25C)

Error bar:  0.56 kPa
Dashed line: 0  0.56 kPa©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.68
6.3 Development of Safety Evaluation Methods and Safety Validation(4) Safety validation on countermeasures against hydrogen gas; a. Verification of the hydrogengeneration amount: Confirmation of effect of alpha-ray (5/6)(vi) Comparison with analysis
The absorbed dose rate was specified based on the hydrogen generation rate (average value) obtained
from the measurement results. An analysis using the radiolysis model was conducted under the test
conditions and the hydrogen generation amount was evaluated. The results show that hydrogen partialpressure in the gas phase is usually approximately the same in both the test and the analysis.
∆ Gage pressure
Measured hydrogen partialpressure (case 3)

Hydrogen concentration (ppb)

Hydrogen concentration∆ Hydrogenpartial pressure∆ Hydrogen partial pressure

Hydrogen concentration

Partial pressure of hydrogen inside
the test container is evaluatedconservatively in cases 1 to 3

∆ Gage pressure

Change in pressure inside container(MPa)∆ Gage pressure
Measured hydrogen partialpressure (case 1)
Measured hydrogen partialpressure (case 2)

Measured hydrogen partialpressure (case 4)
Hydrogen concentration

Time (day)Time (day)(a) Cases 1 and 2

Absorbed dose rate may
have been evaluated less incase 4
Hydrogen concentration (ppb)

Hydrogen concentration (ppb)
Change in pressure inside container(MPa)

Change in pressure inside container (MPa)∆ Hydrogen partial pressure

Time (day)(b) Case 3(c) Case 4

Figure: Example of comparison between test results and analysis results

Table: Analysis conditions
Items
Absorbed
dose rate

Cases 1 and 2

Case 3

Case 4

Notes

Alpha-ray

21.5 Gy/h-263 Gy/h

Beta-ray,gamma-ray

Determined from the hydrogen generation rate (average value) obtained from the test
results

4.7 Gy/h

4.0 Gy/h

4.7 Gy/h

Water volume
Seawater component
concentration
Iodide ion concentration

100 mL

8 mL

Used to calculate hydrogen concentration in gas phase (not considered in analysis byradiolysis model)

5.6 × 10−4 mol/L (Equivalent to 20 ppm) as chloride ion concentration

Test conditions

1.0 × 10−4 mol/L

Test conditions

Temperature

25C

Ambient temperatureTime

20 days

Test conditions©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.69

6.3 Development of Safety Evaluation Methods and Safety Validation(4) Safety validation on countermeasures against hydrogen gas; a. Verification of the amount generatedhydrogen generation amount: Confirmation of effect of alpha-ray (6/6)(vii) Conclusion• In the condition that considers alpha-rays (cases 1 and 2) and the condition that shields alpha-rays (case 3),
the former had the hydrogen generation rate 10 times or greater than that of the latter, which confirms theeffect of alpha-rays.
• The hydrogen generation rate under the condition in which fuel pellets were immersed in small amount ofwater (case4), was greater than that under condition in which fuel pellets were immersed in sufficient amountof water (case1 and 2). This result suggests that, in this test condition (fuel pellet particle size 20 to 300 μm),
the alpha-ray is the dominant condition with small amount of water (small amount of water is the amount ofextent to which drain water).
• Analysis by the radiolysis model that uses the absorbed dose rate which was specified based on the hydrogengeneration rate (average value) obtained from the measurement showed that the results are usually about
approximately the same in both the test and the analysis.
(viii) Future plans
The effect of alpha-ray on hydrogen generation for the identified effect factor shown below will continue to be
confirmed by conducting tests using spent fuel.
• Study on effect of water volume• Effect of particle size
• Effect of moisture in concrete (associated item when collecting MCCI products)
When doing so, test methods and test systems will be reviewed to solve the issues obtained in the evaluation
conducted in FY2017 (establishment of absorbed dose rate, evaluation on evaluation method, and measurementof hydrogen generation rate) and conduct tests and evaluations.
Based on these results, the hydrogen generation evaluation method under 1F conditions will be summarized andtest calculation on the hydrogen generation amount inside the canister will be performed.
©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.70

6.3 Development of Safety Evaluation Methods and Validation(5) Safety evaluation on measures against hydrogen gas; b. Verification of measures against hydrogen (1/6)(i) Purpose
It is assumed that the canister will be transported from the nuclear reactor building to the storage facility in a transport cask.
Therefore, as the transport cask needs to be sealed during transfer, measures against explosion of residual hydrogen isnecessary.
In addition, confining the canister as much as possible is effective in suppressing spread of contamination outside the canister.
Therefore, as a method which treats hydrogen generated in the canister while it is in the can (proposal for the measurement
against hydrogen), the possibility and effectiveness of the measure that recombines oxygen with hydrogen generated inside thecanister will be clarified.
(ii) Basic concepts
Catalysts and hydrogen absorption alloy are generally used for hydrogen treatment. [Supply of gas]
However, since hydrogen absorption alloy has issues in capacity, our study on the Gas supplied to the catalyst
by gas flow inside canister
catalyst will be started.

Recombination catalyst[Performance of recombination catalyst]• Hydrogen oxidation rate per surface
area of catalyst• Coordination withoperating
temperature• Recombination property by water
repellency• Radiation resistance• Poison resistance (chloride, iodine)

To effectively recombine the two with a catalyst, the following needs to be satisfied.
a. The performance related to recombination of the catalyst itself has the ability that meets theamount generated.
• Have recombination performance
The hydrogen recombination rate per catalyst surface area is high and it is in a size that canbe installed inside the canister.
Fuel debris• Functions effectively in the used environment
The catalyst can maintain its performance that is necessary in terms of temperature, water
Canisterrepellency, radiation resistance, and resistance to poison (chlorine component derived fromseawater, anti-poisoning property against iodine produced by fission).
b. The target gas (hydrogen/oxygen) is supplied to the catalyst effectively.
Gas is supplied to the catalyst by the flow inside the canister and the hydrogenconcentration inside the canister is maintained to be or lower than the explosion limit. Figure: Flow inside canister andexample of catalyst placement-> In FY2017, evaluation was conducted focusing on b.

©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.71

6.3 Development of Safety Evaluation Methods and Validation(5) Safety evaluation on measures against hydrogen gas; b. Verification of measures against hydrogen (2/6)(iii) Process of evaluation
Taking into account the previous section, the evaluation proceeds in the following steps.
a. Conduct feasibility study assuming a realistic amount of hydrogen generated (conducted in FY2016)
Assuming an average fuel debris of 1F (average source strength of furnace vessel fuel, size distribution using core samplingof TMI-2), the required amount of catalyst was calculated, with respect to the amount of hydrogen generated and theconcentration distribution inside the canister, when assuming that the catalyst on the ends of the canister will conservatively
recombine by diffusion only. The hydrogen concentration at the center of the canister which had the highest concentration
was lower than 4 vol. %.
b. Examine a feasibility study assuming hydrogen is generated under stricter conditions (details of study in FY2017)
When examining the hydrogen countermeasure scenario, the possibility that the conditions will be stricter cannot be denied(e.g., distribution of fuel concentration in fuel debris), so a scenario for such case will be constructed.
• Study on effect of flow inside canister
The effect of the flow inside the canister will be analyzed and the placement of catalysts and gaps necessary inside thecanister will be studied.

c. Verification assuming canister shape (details of study in FY2018)• Confirm the effectiveness of the catalyst by verification test on flow
inside the canister.
• Obtain basic data of resistance against environmental conditions, suchas resistance to radiation and poison, from the aspect of placing
recombined hydrogen catalysts inside the canister.
In addition, consider the flow inside the canister and verify its
effectiveness regarding the catalyst selected for its resistance toradiation and poison.

[Arrangement of recombination
catalyst]
Study the placement of therecombined catalysts inside thecanister using a flow evaluation
model that reflects the resultsof flow test
[Flow analysis]
Understand flow of hydrogenby diffusion and convection(cylinder part, upper space)

Recombination catalyst[Information on recombination
catalyst]
Obtain information on
recombination catalyst• Operational temperature
conditions• Coefficient of hydrogenoxidation rate per surface
area of catalyst• Water repellency• Radiation resistance• Poison resistance, etc.

Fuel debris

CanisterFigure: ©International
Study items
regarding measures against
Research Institute for Nuclear Decommissioninghydrogen gas inside canisterNo.72

6. Implementation Details

6.3 Development of Safety Evaluation Methods and Validation(5) Safety evaluation on measures against hydrogen gas; b. Verification of measures against hydrogen (3/6)(iv) Hydrogen concentration inside the canister when hydrogen does not flow but spreads by diffusion alone
On conducting the study, the following is the hydrogen concentration in the center of the canister that is premised
only on the amount of hydrogen assumed to generate inside the canister (trial calculation) and diffusion.
a. Concentration distribution of hydrogen
Catalysts will be placed on both ends inside the canister. Assuming that hydrogen molecules move by diffusion only
when it generates from fuel debris, hydrogen concentration at an arbitrary position in the canister is given by afunction of diffusion coefficient and the amount of hydrogen generated. Hydrogen is distributed in a parabolic shapeand its concentration is the highest at the center of the canister.
b. Hydrogen concentration
The amount of hydrogen generated is calculated from the amount of heatgeneration, G value, and absorption rate, a method which was also used
for TMI-2. Conservatively, G value was specified as (0.45) and absorption
rate F as (0.6)*1. The amount of heat generated was also conservatively
defined assuming that pellets of aggregates with maximum burnup (55GWd/t) (10 years of cooldown) were the only fuel debris. In addition, it
was assumed that the fuel debris filling rate is 30 vol.%.
The hydrogen concentration on the ends was assumed as 1 vol. %based on the general performance of the catalyst. As a result, theconcentration in the center was evaluated to be 6.4 vol. %.
-> Convections actually generated inside the can and by making
allowance for such feasibility under strict conditions can be
expected.
-> Flex will be studied.

Place hydrogen catalyst on both ends
Hydrogen gas
generates
uniformly

CanisterFigure: Concentration distribution
inside the canister*1: F = 0.6 is for the assumption that fuel debris is a fine powder(from the study results of FY2016)©International Research Institute for Nuclear Decommissioning6. Implementation Details(v) Study on effect of convection (1/2)
a. Target convection flow rate
As shown in the previous page, if only hydrogen diffusion is assumed, themaximum concentration at the center of the canister was 6.4 vol.%. Given
this situation, the target convection flow rate was evaluated in which
effect of convection can be expected.
• In order for the maximum concentration to go below 4 vol.% (the rightfigure), a flow that can diffuse hydrogen two times (or greater) more
efficiently than the diffusion coefficient is necessary.
• Considering the relation between diffusion term and convection term in
the diffusion equation, only a natural convection flow of the level of 0.01mm/s (= 10-5 m/s) same as the diffusion coefficient is necessary.
However, since it is likely that there will be great uncertainty in suchflow rate, in the analysis the possibility that a flow of about 1 mm/s may
be created was tentatively studied.
𝐷
5 × 10−5=≈ 10−5𝐻2
1.52
𝑢 > 10−5 [m/s]Convection term of diffusion equation

𝑢
𝐻

𝑢
2
𝐻 2 ≈ 10−5
𝐻 > 𝐷 ×(Double or more)

Hydrogen concentration(diffusion only) 6.4 vol%Limit of hydrogen explosion (lower
limit) 4 vol%Doubling for diffusion coefficient of hydrogen [-]Figure: Relation between hydrogen diffusion
coefficient and hydrogen concentration

One-dimensional diffusion equation

𝑑𝜔
𝑑𝜔
𝑑2 𝜔
+ 𝑢
= 𝐷 2
𝑑𝑡
𝑑𝑧
𝑑𝑧

Diffusion term of diffusion equation

𝐷
𝐻2

No.73
Hydrogen concentration at center ofcanister [vol%]6.3 Development of Safety Evaluation Methods and Validation(5) Safety evaluation on measures against hydrogen gas; b. Verificationof measures against hydrogen (4/6)

Convection term Diffusion term

D: Hydrogen diffusion coefficient [m2/s]
H: Height of canister [m] (= 1.5 m)
t: Time [sec]u: Flow rate [m/s]z: Distance in direction of flow [m]ω: Hydrogen concentration (massfraction) [-]©International Research Institute for Nuclear Decommissioning6. Implementation Details

No.74
6.3 Development of Safety Evaluation Methods and Validation(5) Safety evaluation on measures against hydrogen gas; b. Verification of measures against hydrogen (5/6)(v) Study on effect of convection (2/2)
b. Prediction of flow effect by analysis
From the results of thermal flow analysis, it was revealed that the average rising flow rate was 10 to 100 mm/s for gap of 20 mm (cases
1 to 4) and 1 to 4 mm/s for gap of 5 mm (case 5) and that flow can be expected.
Case1 220 W/m3, 20 mm
Case2 700 W/m3, 20 mm
Case3 2630.7 W/m3, 20 mm
Case4 7000 W/m3, 20 mm
Case5 220 W/m3, 5 mm
Document formula⌀220 mm10⌀180 mm or⌀210 mm

Nul

1400 mm H*Vertical sealed parallel flat plates (Source:
S.W.Churchill (1983). Heat exchanger design
handbook)

1

Nu l = 1+

1500 mm

Gap width (20 mm, 5mm)b (= 1.5m): Height of flat plate [m]l (= 0.02 m): Width between parallel
flat plates [m]Average temperature of the fuel0.1
1.00E-02
debris side surface
Average temperature of
the canister side surface

Gap

Figure: General formof the canister

Nu: Nusselt number [-]1.00E-01 1.00E+00 1.00E+01 1.00E+02 1.00E+03 Ra: Rayleigh number [-]Index: Representative length used
Ral (l/b)

Figure: Document formula of vertical sealed parallel flat
plates and analysis results of gap

120
0.12

cross-section average
Horizontalcross-section
flow
averageflow
Horizontal[mm/s]
velocityrate [m/s]

Used to find the values ofNusselt number andRayleigh number

Ral l
720 b

Case1 220 W/m3, 20 mm

100
0.1

Case2 700 W/m3, 20 mm

80
0.08

Case3 2630.7 W/m3, 20 mm

60
0.06

Case4 7000 W/m3, 20 mm

40
0.04

Case5 220 W/m3, 5 mm

At height of 1.4 m
Average rising flow rate: 103[mm/s]

Figure: Example on flow
distribution in the direction ofheight(Case 4)20
0.02

Hightemperaturewall

Lowtemperaturewall

Figure: Vertical sealed parallel flat plates

0
0

0.5

1

1.5

Height from bottom [m]Figure: Rising flow rate of each cross-section in the gap©International Research Institute for Nuclear Decommissioning

No.75

6. Implementation Details

6.3 Development of Safety Evaluation Methods and Validation(5) Safety evaluation on measures against hydrogen gas; b. Verification of measures against
hydrogen (6/6)(vi) Conclusion
Flow inside the canister is effective for promoting hydrogen diffusion. It was revealed that, by
anticipating this effect, the hydrogen concentration in the canister can be expected to go below thelower explosion limit of 4 vol.% or less even with the amount of hydrogen generated based onconservative assumption.
(vii) Future plans
Measures will be organized using catalysts by the following studies and clarify the benefits if catalysts
are adopted.
• Study on analysis of flow inside the canister
From the analysis, flow is expected to be generated between the unit can and the canister wall.
However, the gap between the two is narrow and as disclosed knowledge on natural flow has yet tobe found, data on the effect of convection by tests will be collected and confirmed.
• Organization of candidate catalyst (including consideration on poisoning, heat resistance, radiation
resistance, etc.)
Catalysts are required of resistance to poison (e.g., chloride ion), heat, and radiation. To collectinformation on these properties from documents and tests, and to base our judgments on whether toadopt the catalysts, the environment during transfer must be considered and the validity of thecatalyst candidate must be organized.

©International Research Institute for Nuclear Decommissioning

No.76

6. Implementation Details
6.4 Study on Fuel Debris Collection Method(1) Study on the canister’s specifications that suit fuel debris properties(i) Purpose
A canister for block-like fuel debris has been developed to store an unit can after collecting fuel debris
by griping or scooping in a basic way. On the other hand, there are powder type of fuel debris. In
example of TMI-2, canisters with filters inside were used. Therefore, canisters that can be used tocollect fuel debris of various forms will be developed.
(ii) Result of study
Currently, collection method of powder fuel debris is being studied in the Method Project Team andFundamental Technology Project Team. As of now, it has been decided to proceed with our study on
using canisters for fuel debris blocks, regardless of the fuel debris shape.
In this project, designing canisters for fuel debris blocks will be started. From FY2018 onwards, issues
that arise when using canisters for block-like fuel debris to collect powder fuel debris will be shared andadjustments made.

©International Research Institute for Nuclear Decommissioning

No.77

6. Implementation Details
6.4 Study on Fuel Debris Collection Method(2) Review of canister design
LidLid(i) Purpose
Develop design of canister that reflects theresults of 6.2 “Study on safety requirements andspecifications and storage systems for thetransfer/storage of fuel debris canisters” and 6.3“Development of safety evaluation methods andsafety validation.”(ii) Result of study
The study of 6.3.(2) “Safety evaluation on
structural strength: Study on lid structure” will bereflected on the lid structure design as for now.
The example of the reflected design is shown in
the right figure.

Body ofcanister

Drain pipe

In the future, requirements based on handling
methods and safety assessment will be reflectedto the design and summarized as the basic
specification (study will continue in FY2018).
Assembled canister

Body of canister

Figure: Canister design (example)©International Research Institute for Nuclear Decommissioning7. Overall Summary

No.78• To efficiently retrieve, collect, transfer, and store fuel debris, steps of each task were
defined and a reasonable process flow was provisionally determined. In addition, theamount of items to be treated was studied from the aspect of overall rationalizationand evaluated from the aspect of future throughput. They will be responded to byreflecting it on the canister design and requirements of the transfer/storage system.
• Related projects will be consulted with on the issues newly identified in the abovetask and will reflect on the plans of FY2018 onwards.
• Regarding the evaluation method required for the safety assessment oftransfer/storage system that is conducted continuously from the previous research,
the evaluation will be continued for achieving results in FY2018. The results will bereflected to the canister specifications.

©International Research Institute for Nuclear Decommissioning[Supplement-1] Positioning the Output

No.79

The conditions such as safety requirements, knowledge of fuel debris properties, and request for optimization
will change in the future.
The canister to be developed should be a prototype. The protocol of the canister design will be reviewed toflexibly change the conditions.
Requirements for optimization (feasibility): All times
Safety requirements (restrictions)ProtocolImprovement of work
efficiency

Ensure subcriticality
Size, guarantee on amount ofcontent, absorption material…

Requirements

Measures against
hydrogen and fire

Expansion of sizes of opening andstorage
Manipulation (e.g., restriction inheight direction)
Canisters for each target objects

Change in basic
conditions
Improvement ofknowledge on fuel debris
Technological innovation

Manufacturing process

Structure of lid and vent, restriction oftransfer time, hydrogen confinement…

Manufacturability (quality),
manufacturing and inspection
cost, production volume

Control of temperature
increase

Rationalization of storage
management

Heat removal design, guarantee onamount of heat generation,
continuous immersion…

Dry storage, external event,number of stored items,
storage area, discharge
control

Discharge
suppression
Structural strength, wet storage,
high corrosion resistance material,building ventilation system (drystorage)

Actual conditions of on-site work
Unknown factors

Change in design

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