FY2021 Subsidy Project of Decommissioning and Contaminated Water
Management

Development of Safety System (Liquid/Gas Phase Systems,
Criticality Control Technology)

Final Report(1) Liquid/Gas Phase Systems(2) Criticality Control Technology
June 2023

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

0FY2021 Subsidy Project of Decommissioning and Contaminated
Water Management
Development of Safety System (Liquid/Gas Phase Systems,
Criticality Control Technology)

Final Report(1) Liquid/Gas Phase Systems
June 2023
International Research Institute for Nuclear
Decommissioning (IRID)
All rights reserved.
©International Research Institute for Nuclear Decommissioning (IRID)

0

No.1

Table of Contents
1. Purpose and goals of the subsidy project
2. Overview of subsidy project
3. Development results up to the previous PJ and remaining issues
4. Input and output information
5. Implementation schedule
6. Implementation framework
7. Implementation items① Development of soluble α nuclide removal technology② Development of RO concentrated water treatment technology③ Development of secondary waste treatment technology③-1 Development of non-soluble nuclide removal technology③-2 Development of sludge dehydration treatment technology

8. Specific goals for achieving the purpose of the project©International Research Institute for Nuclear Decommissioning1. Purpose and goals of “Development of Safety System (Liquid/Gas Phase Systems, Criticality
Control Technology)”

No.2[Purpose of Development of Safety System (Liquid/Gas Phase Systems, Criticality Control Technology)]
At Tokyo Electric Power Company Holdings Inc. Fukushima Daiichi Nuclear Power Station (1F), it is considered that nuclear fuelmelted with reactor internals and exists as fuel debris in the reactor pressure vessel (RPV) and primary containment vessel(PCV).
Fuel debris in RPV and PCV is currently considered to be in subcriticality, but since the plant itself is in an unstable state which
is different from the initial design, such as the reactor building (R/B), RPV, PCV, etc. being damaged due to the accident, it isnecessary to prevent the diffusion of radioactive materials and achieve a stable state by removing fuel debris and maintaining
subcriticality of fuel debris.
Against the above background, this project will conduct studies to realize large-scale fuel debris retrieval work underengineering and project management performed by Tokyo Electric Power Company Holdings Inc. (TEPCO) in accordance withthe “Mid-and-Long-Term Roadmap towards the Decommissioning of TEPCO's Fukushima Daiichi Nuclear Power Station Units 14” (hereinafter Mid-and-Long-Term Roadmap). Development results of this project will be utilized in engineering by TEPCO.
The purpose of this project is to facilitate decommissioning and contaminated water measures of 1F and raise the level ofscience and technology of Japan, by implementing the project that supports the development of technologies that contribute todecommissioning and contaminated water measures of 1F in accordance with the Mid-and-Long-Term Roadmap and “FY2021
Decommissioning Research and Development Plan” (Decommissioning and Contaminated Water Team Meeting / Secretariat
Liaison Meeting (86th)).

In specific, technology related to on-site operation methods shall be developed regarding soluble α nuclide removal technology
for α nuclides considered to be eluted in circulation cooling water from fuel debris, RO concentrated water (*) treatment
technology, secondary waste treatment technology, as well as criticality approach monitoring technology and neutron
absorbent technology.
*Liquid waste with high impurity concentration generated by removing impurities from cooling watercontaining impurities with reverse osmosis membrane (RO membrane)[Overall development goals]
Perform necessary element technology development and tests regarding system technology andtechnology for ensuring safety, required to further expand the scale of retrieving fuel debris and reactor
internals, in accordance with R&D results obtained up to date.

©International Research Institute for Nuclear Decommissioning

No.3

2. Overview of subsidy project

Here is the system configuration diagram for the liquid/gas phase systems studied in the subsidy project.
This term, soluble nuclide removal equipment and sludge dehydration and stabilization treatment system weredeveloped in the subsidy project.

Pretreatment
systemGas phasefilter: Development itemof each equipment

To exhaust port
Exhaust fan

CoolerParticle
removalfilterDrain water /backwashwater

Soluble nuclide
removal adsorption
tower

Tank
(including boric acidadjustment)

B

Supernatant water /permeable waterA

Ambient air
NitrogenP
Ground waterP

Soluble nuclide
removaladsorption tower

Particle
removal filter

P

Drain water /Supernatant water
backwashwater
Sludge collection
Dehydration and

Sludgesedimentation
separator

Permeable waterA*

Sludge collection
all filtering
equipment

B
Sludge (filter)

Supernatant waterC

A

Sludge collection
sedimentation
separator

Boric acid (silo)*Used only during regular use of
boric acidCollected boric acid solution

Collected boric
acid tank: Equipment subject
to this project

Fuel debris retrieval
work water③ Development of secondary waste
treatment technology
a. Inspection of pretreatment technology
b. Study of applicability to actual
equipment② Development of RO concentrated water
treatment technology
a. Selection of adsorbent and condensing agent
b. Study of applicability to actual equipment① Development of soluble α nuclide removal technology
a. Study of soluble α nuclide removal test assuming use of actual liquid
b. Element test assuming fuel debris retrieval work

stabilization
treatment system

Boric acidcollection
equipment*RO membrane: Reverse
osmosis membrane

Soluble nuclide
To existing water
removal RO
treatment facility
membrane
Concentrated water
Supernatant water
To existingwater treatment
Sludge collection
facility
sedimentation
separator

SludgeDehydration andstabilization
treatment
systemC

B
Sludge Dehydration andstabilization
treatment system

CC
To wastestorageC
To
canisterstorage*: Used to collect drain water from rough particle removal treatment

Figure. Liquid/Gas Phase Systems during fuel debris retrieval (subsidy project basis)
Note: This conceptual diagram is an example of equipment configuration (case with no leakage from PCV)
©International Research Institute for Nuclear Decommissioning3. Development results up to the previous PJ* and remaining issues

No.4① Development of soluble α nuclide removal technology: To be studied in this PJ
No.

Item

Initiatives and results in the previous PJ

Remaining issues

Particle generation impact of 15 nuclides of 5 elements,U, Np, Pu, Am andCm, was confirmed through tests. Regarding U, it was confirmed that the
Dissolution behavior and adsorption behavior of α nuclides ofneed for adsorption removal was low, since solubility near neutrality of pH U, Np, PU, Am and Cm, respectively, assuming nitrogen
5-9 in air atmosphere and concentration in accumulated water in PCV are environment in PCV
about two digits lower than the announcement density.

1

Nuclides to be
removed

2

Water quality of Evaluation was conducted assuming leaching of seawater components,
water to beleaching of concrete components, and injection of sodium pentaborate
treatedwhich is a criticality prevention agent.
Treatment flow rate conditions when installing adsorption removal system
inside equipment intended for circulation cooling of accumulated water in
PCV, and treatment flow rate conditions when installing adsorption
removal system outside equipment intended for circulation cooling ofaccumulated water in PCV were tentatively set.

3

Treatment flowrate

4

Water quality was evaluated assuming leaching of seawater components
and injection of sodium pentaborate which is a criticality prevention agent,
α nuclideand dissolution behavior of each α was grasped. Regarding impact ofconcentration inleaching of concrete, it was confirmed that evaluation in nitrogenassumed water
atmosphere close to PCV environment is necessary, since α nuclides coquality
precipitate due to generation of calcium carbonate in atmosphericenvironment.

5

Concentration
reduction
requestImpact of oil, paint components, rust inhibitor components
and non-soluble criticality prevention agent

Setting of treatment flow rate considering equipment
operation schedule

Dissolution behavior and adsorption behavior of α nuclides inhigh pH assuming leaching of concrete components

Setting of necessary DF considering particulate α nuclide
Target DF to reduce public exposure impact is set to 100, andannouncement density is the concentration reduction target for transport to concentration
the water treatment facility. (no progress)
Setting of concentration reduction goal of individual nuclides

Of issues organized in the previous PJ*, No.1 and No.4 were selected as issues to work on assuming control of nitrogen atmosphere.
It was suggested that the form of each α nuclide to be removed may be different in nitrogen atmosphere. Therefore, this PJ will evaluate potential impact onsolubility and adsorption phenomenon and study soluble nuclide removal system that can remove α nuclides in nitrogen atmosphere.
*FY2018 supplementary budget “Subsidy Project of Decommissioning and Contaminated Water Management (Development of Technology for Retrieval of
Fuel Debris and Reactor Internals)” (implemented in FY2019, FY2020)©International Research Institute for Nuclear Decommissioning3. Development results up to the previous PJ and remaining issues

No.5① Development of soluble α nuclide removal technology
No.

Item

Initiatives and results in the previous PJ

Remaining issues

Applicableadsorbent

As a result of evaluation of α nuclide adsorption performance by an
immersion test, active carbon, zirconium phosphate and titanic acid
were selected as candidate applicable adsorbents. Since it is at astage where adsorbents with large equilibrium adsorption were
Evaluation method of adsorption removal performanceselected through an immersion test, it is an issue to confirm
when applied to water flow treatment
performance when applied to water flow treatment. The evaluation
method is an issue since distribution tests handling large amounts of α
nuclides cannot be performed with laboratory tests.

Adsorption
tower
configuration

The tower configuration arranges zirconium phosphate, which
adsorbs Pu, Am, Cm and Np, and titanic acid, which adsorbs Np, etc.,
in a series with active carbon for removing colloidal α nuclides at thefront. Two towers are assumed for each tower assuming a merry-goround operation. It is necessary to conduct an impact assessment forcases when the respective nuclides exist in the solution together, andalso appropriately re-examine the tower configuration based on
removal performance and replacement frequency for water flow
treatment.

8

Replacementtiming /replacement
frequency

Since all adsorbents selected as applicable candidates have low Cs
adsorption performance, risk of replacement due to increased dose on
the tower surface is low. In general, frequency of replacement due todecreased adsorption performance is evaluated with a distribution
Evaluation method of replacement frequency of adsorbent
test, but the evaluation method is an issue since distribution tests
handling large amounts of α nuclides cannot be performed withlaboratory tests.

9

Replacement
methodWhen either replacing the entire adsorption tower or replacing only
adsorbents, equipment design is required in accordance with thetreatment policy of used adsorbents (long-term storage for the time
being).

Tower size

If the treatment flow rate is 10m 3/h, it requires the same size as theadsorption tower of the Advanced Liquid Processing System (outer
diameter approx. 1m, height approx. 2.5m).

6

7

10

Mutual impact of multiple α nuclides
Impact of adsorption inhibiting components

Re-examination of tower configuration based on removal
performance and replacement frequency for water flow
treatment

Replacement operation
Handling of used adsorbent

Design of adsorption tower inputting shielding design,
replacement methods, etc.

©International Research Institute for Nuclear Decommissioning3. Development results up to the previous PJ and remaining issues② Development of RO concentrated water treatment technology③ Development of secondary waste treatment technology
No.

: To be studied in this PJ

Initiatives and results in the previous PJ

Remaining issues

Assumed water quality of drain water and backwash water dispensed from theparticle removal system and input to the sludge collection system wasorganized based on the element test result and filter test result of the fueldebris retrieval method PJ.
However, since intermediate removal filter has not been selected among
particle removal equipment, it is necessary to select candidate equipment andevaluate the properties of generated liquid waste. Additionally, element tests up
to the previous fiscal year have only evaluated cases when spherical particles
have passed through water, so it is necessary to evaluate properties of filter
behavior and generated liquid waste when non-spherical particles passthrough water.

Select candidate intermediate removal filters based on literature review.
Confirm applicability through element tests and acquire properties ofgenerated liquid waste.

Treatment
flow rate

Treatment flow rate requirements of the sludge collection system were
evaluated based on the operation schedule of the fuel debris retrieval method,
filter test results, and the flow rate of the liquid system.

If the assumed input water quality of the liquid system is changed due tofuture studies, incorporate it into the treatment flow rate requirements ofthe sludge collection system.

3

Treatment
methodCoagulation-sedimentation treatment was selected as the method to treat filter
liquid waste and RO concentrated water, and usable coagulants were selectedbased on element tests. Additionally, the process of coagulation-sedimentation
treatment procedures using the sedimentation separation tank was studied.
It is necessary to study applicability to actual equipment regarding
coagulation-sedimentation treatment of RO concentrated water using thesedimentation separation tank. Additionally, it is necessary to study reductionof the generated amount of sludge and simplification of operation for thetreatment procedures.

Select adsorption treatment using power adsorbents and coagulationsedimentation treatment using coagulants, as methods to treat RO
concentrated water, and select usable powder adsorbents and coagulants.

4

Properties
andgeneratedamount of
sedimentati
on sludge

Generated amount of sedimentation sludge was estimated based on the
coagulation-sedimentation test. As a result, it was evaluated that thegenerated amount of sedimentation sludge of RO concentrated water islarge due to large treatment flow rate and large water content in sedimentation
sludge. Thus it is necessary to reduce the volume of sedimentation sludge
since there are concerns about complication of delivery treatment andincreased load on delivery destinations.

1

2

Item

No.6

Water
quality ofwater to betreatedEvaluate the impact on the particle removal system when treating nonspherical particles, through element tests.
When it is evaluated that there is a predominant influence on behavior of
equipment differential pressure increase, recovery rate, and amount ofgenerated liquid waste, incorporate it into the assumed input water qualityof the sludge collection system.

Consider revising each process of RO concentrated water treatment
procedures, in order to reduce the generated amount of sedimentation
sludge. Study reduction of added amount of powder adsorbents, reductionof added amount of coagulants and optimization of stirring conditions
through element tests.
Study application of dehydration treatment to sedimentation sludge, selectapplicable dehydration technology through literature review, and evaluate
applicability to actual equipment through element tests.

In the previous PJ, it was suggested that the amount of sedimentation sludge generated from treatment of RO concentrated water may be large. Thus,
this PJ shall conduct studies that contribute to reduction and volume reduction of sludge such as optimization of treatment methods (No.3) and sludge
dehydration (No.4). Additionally, dependency on particulate filters (No.1) shall be evaluated in this PJ, as the previous PJ only conducted the filter test
using spherical particles and did not conduct an evaluation.
©International Research Institute for Nuclear Decommissioning3. Development results up to the previous PJ and remaining issues

No.7② Development of RO concentrated water treatment technology③ Development of secondary waste treatment technology
No.

Item

5

Properties ofsupernatant

Amount of residual components and SS (suspended solids) insupernatant after solid-liquid separation was organized.

Organize acceptance requirements of soluble nuclide removal equipment
and existing water treatment facility, which are delivery destinations, andstudy necessity of neutralization treatment, removal of neutralizing salts andremoval of trace amounts of residual SS components.

Operation
methodThe coagulation-sedimentation treatment process for filter liquid
waste and RO concentrated water using the sedimentation
separation tank was studied based on the operation schedule ofthe fuel debris retrieval method as well as element test results. An
operation cycle that completes 1 batch treatment in 1 day wasstudied.

Incorporate the element test results of this term into the coagulationsedimentation treatment process studied up to the previous PJ.

Equipment
specifications

Operability of the sedimentation separation tank and eachconstituent element was verified with an equipment test.
Since an event of large particles biting in when opening andclosing the gate valve occurred, it is necessary to re-examine thespecifications of the gate valve.

Re-evaluate each mechanism of the sedimentation separation tank inelement tests using RO concentrated water treatment test equipment to be
conducted after this term.
Additionally, survey valves applicable to the sedimentation separation tank.

8

Delivery methodof sedimentation
sludgeDelivery method of sedimentation sludge using the sludge
collection container was studied, and it was confirmed with thetest equipment that it is applicable to actual equipment.
Piping was inserted from the top of the sedimentation separation
tank, and it was confirmed that sedimentation sludge can beextracted by pump suction.
In the future, acceptance requirements of delivery destinations of
sedimentation sludge shall be organized, and size of sludge
canister and application of dehydration treatment shall be studied.

When treating and disposing of sedimentation sludge as waste, it isnecessary to apply pretreatment in consideration of final disposal of
sedimentation sludge as it is delivered to the waste line. If a policy for the
final disposal method of sludge is adopted in the future, study necessary
pretreatment in the liquid system and incorporate it in the delivery method of
sedimentation sludge.
When storing sedimentation sludge in storage canisters, it is necessary toapply pretreatment so that the requirements for long-term storage aresatisfied as they are delivered to the storage canister line. Organize
requirements for long-term storage of sedimentation sludge, and study theshape and moisture content of the sludge collection container.

9

Handling of
sedimentation
sludgeAssumed properties and generated amount of sedimentation
sludge were organized based on the element test result and liquid
system requirements.

Share assumed properties of sedimentation sludge with waste PJ andstorage canister PJ, and identify issues in each line.

67

Initiatives and results in the previous PJ

Remaining issues©International Research Institute for Nuclear Decommissioning

No.8

4. Input and output information
This PJ was conducted in coordination with relevant PJ such as the Project of Development ofTechnology for Containing, Transfer and Storage of Fuel Debris and Project of R&D for
Treatment and Disposal of Solid Waste.
ID

Demand-side project

Provider-side project

Description (overview)

Usage of information

1

Development of Safety System(Liquid/Gas Phase Systems,
Criticality Control Technology)

Development of fuel debris
packaging, transport and storage
technology

Acceptance conditions ofstorage canister

Study of stabilization treatment
technology of secondary waste

2

Development of Safety System(Liquid/Gas Phase Systems,
Criticality Control Technology)

Research and development ofsolid waste treatment anddisposal technology

Acceptance conditions of waste Study of stabilization treatment
storage container
technology of secondary waste

3

Development of Safety System(Liquid/Gas Phase Systems,
Criticality Control Technology)

Development of analysis andestimation technology for fueldebris characterization

Study results of formation
behavior of fuel debris particles

Study of properties of particles
contained in circulation cooling
water

4

Development of fuel debris
packaging, transport and storage
technology

Development of Safety System(Liquid/Gas Phase Systems,
Criticality Control Technology)

Properties of sedimentation
sludge generated from theliquid system

Identification of issues inhandling slurry and sludge fueldebris5

Research and development ofsolid waste treatment anddisposal technology

Development of Safety System(Liquid/Gas Phase Systems,
Criticality Control Technology)

Properties of sedimentation
sludge generated from theliquid system

Identification of issues intreatment and disposal of slurry
and sludge radioactive waste©International Research Institute for Nuclear Decommissioning

No.9

5. Implementation schedule
Large
category
大分類

Small
category
小分類①溶解性α核種① Developmentof除去技術の開発
solubleα nuclideremoval technology

a. 概念検討／試験計画

a. Conceptual study /
test
plan
b. 試験準備・手配
b. Test preparation andc. 予備試験
arrangement

令和４年度
FY2022

Apr May
Jul Aug
Nov 12月
Dec JanOct 11月
4月
5月 6月
9月 10月
1月 Feb
2月 Mar
3月 4月
5月 Jun
6月 7月
8月 Sep9月 10月
1月 Feb
2月 Mar3月
Oct 11月
Dec Jan
Nov 12月
Apr MayJun 7月Jul 8月Aug Sep

Project steering meeting
プロジェクトステアリング会議

主要なマイルストン
Major
milestones(1)(１) Liquid/Gas
液体系・気体系
システム
Phase
Systems

令和３年度
FY2021

Implementing
operator
実施事業者(Also list the（外注先、試験場所がsubcontractor
and testing
site, if any)
ある場合は併記）(1) Hitachi GE（１）日立GE
Subcontractor: None
外注先：なしTesting site: Hitachi
試験場所：日立研究所
Research Laboratory

中間報告Interimreport中間報告
Interimプロジェクトステアリング会議
Project steering meeting
reportInterimプロジェクト
Projectsteering 中間報告
ステアリング会議
meeting report

最終報告
Final
ReportConceptual
study / test plan
概念検討／試験計画: Planned: Actual

Test試験準備・手配
preparation and arrangement

予備試験 test
Preliminary
Element
test
要素試験

c. Preliminary testd. 要素試験

d. Element test

まとめ
Summary

Interimsummary
中間まとめ

e. まとめe.
Summary

Conceptual study / test plan
概念検討／試験計画

概念検討／試験計画
a.a. Conceptual
study /

test planb. 試験準備・手配
b. Test preparation andarrangementc. 予備試験（２）東芝ＥＳＳ(2)
Toshiba ESS
外注先：NFD NFD
Subcontractor:
Testingsite: Toshiba ESS
試験場所：東芝ESS研究所、
laboratory,NFD
NFD

試験準備・手配
Test preparation
and arrangement
予備試験 test
Preliminaryc. Preliminary test

Elementtest
要素試験

d. 要素試験

d. Element test

② Development
of RO②RO濃縮水の処理
concentrated
技術の開発water
treatment technology

概念検討／試験計画
a.a.Conceptual
study /
test planb. 試験準備・試験装置試作
b. Test preparation andarrangementc. 予備試験（１）日立GE(1)
Hitachi GE

c. Preliminary test

Testing site: Rinkai Plant

d. Element test

(Hitachi City)d. 要素試験

Summary
まとめInterim summary
中間まとめ

まとめ
e.e. Summary

Conceptual
study / test plan
概念検討／試験計画

外注先：日立プラントコンス
Subcontractor:
Hitachi Plant

Test preparation and test
manufacturing of test equipment
試験準備・試験装置試作

トラクション
試験場所：臨海工場(日立市)

Construction

Preliminary test
予備試験
Element
test
要素試験

中間まとめ
Interimsummary

Summary
まとめ

まとめ
e.e.Summary③二次廃棄物処理③ Developmentof
技術の開発
secondarywaste
treatment technology(1) Hitachi GE

a.a.Conceptual
study /
概念検討
test plan
技術調査/試験計画
b.b.Test
preparation andarrangement（１）日立GE
Subcontractor: Hitachi Plant
外注先：日立プラントコンス
トラクション
Construction
試験場所：臨海工場(日立市)
Testing site: Rinkai Plant

c. Preliminary test

(Hitachi City)
外注先：みすず精工
Subcontractor: Misuzu Seiko
試験場所：みすず精工研究所
Testing site: Misuzu Seikoc. 試験準備／試験体製作

要素試験 test
d.d.Element

e.e.Summary
まとめ

Conceptual
概念検討 study
Technical study / test plan
技術調査／試験計画
Test preparationand test manufacturing of test equipment
試験準備／試験装置試作
Element
要素試験test

中間まとめ
Interimsummary

Summary
まとめ

Research Institute©International Research Institute for Nuclear DecommissioningNo.10

6. Implementation framework
Tokyo Electric Power Company
Holdings Inc.
➢ Various adjustments from theperspective of applicability in thefield

Technology Research Association International ResearchInstitute for Nuclear Decommissioning➢ Formulation of overall plan and coordination of technical supervision➢ Coordination of technical management such as progress of technicaldevelopment

Collaborating development
project teams

Hitachi-GE Nuclear
Energy, Ltd.

Toshiba Energy Systems& Solutions Corporation

Mitsubishi Heavy
Industries, Ltd.

Japan Atomic Energy Agency[Element test / technical development](1) Development of Liquid/Gas Phase
Systems① Development of soluble α nuclide
removal technology
・Study of soluble α nuclide removal test
assuming use of actual liquid (accumulated
water in reactor building)・Element test assuming fuel debris
retrieval work
② Development of RO concentrated water
treatment technology
・Selection of adsorbent and coagulant
・Study of applicability to actual equipment③ Development of secondary waste
treatment technology
・Inspection of stabilization treatment
technology
・Study of applicability to actual equipment[Element test / technical development](1) Development of Liquid/Gas Phase
Systems① Development of soluble α nuclide
removal technology
・Study of soluble α nuclide removal test
assuming use of actual liquid (accumulated
water in reactor building)・Element test assuming fuel debris
retrieval work

[Element test / technicaldevelopment](2) On-site operation methods forcriticality approach monitoring
technology and neutron absorbertechnology[Element test / technicaldevelopment](2) On-site operation methods forcriticality approach monitoring
technology and neutron absorbertechnology

・Formulation of criticality
approach monitoring procedures・Performance assessment ofsubcriticality measurement of
neutron detector

・Development of solidified
absorbent technology(2) On-site operation methods for criticality
approach monitoring technology andneutron absorber technology
・Formulation of criticality approachmonitoring procedures[Details of re-consignment]
・Subcriticality measurement test (Kyoto
University)[Main outsourcing contents]
・Operational test of non-soluble neutron
absorbent (HPC)・Water flow test of MF membrane and otherfilters (Misuzu Seiko)・Cartridge filter test (HPC)(2) On-site operation methods for criticality
approach monitoring technology andneutron absorber technology
・Formulation of criticality approachmonitoring procedures・Performance assessment ofsubcriticality measurement of neutron
detector・Formulation of operating procedures forabsorbents
・Study of impact on debris drying process[Details of re-consignment]
・Subcriticality measurement test (Kyoto
University)[Main outsourcing contents]
・Soluble α nuclide removal test (NFD)
・Corona-type detector (FEO/TENEX-J)
・Multicell-type detector (CETD)

Development of fueldebris retrieval method
Development of fueldebris packaging,transport and storage
technology
Research and developmentof solid waste treatment
and disposal
Development of analysis
and estimation
technology for fuel debris
characterization
Development oftechnology for detailed
investigation inside PCVDevelopment oftechnology forinvestigation inside RPV[Details of re-consignment]
・Subcriticality measurement test (Kyoto
University)[Main outsourcing contents]
・Assistance in subcriticality measurement
test (IPL)・Assistance in study of criticality approachmonitoring procedures (NS Engineering)[Details of re-consignment]
・Structural analysis of solidified absorbents(Shizuoka Institute of Science andTechnology)[Main outsourcing contents]
・Contract work for drying tests of test
specimens coated with solidified absorbents(Inspection Development Company Ltd.)

Development of analysis
and estimation
technology for fuel debris
characterization
Development of technology
to sort fuel debris andradioactive waste©International Research Institute for Nuclear Decommissioning7. Implementation items

No.11① Development of soluble α nuclide removal technology
・Development background• There is a plan to build a liquid system as a safety system for fuel debris retrieval. Exposure dose assessment of the liquid system using
tentative conditions found that site boundary exposure dose falls below the threshold both during normal times and during an accident but islargely affected by Pu system and its daughter nuclides (Pu-238, Pu-241, Am-241).
• Since α nuclides are not released to the environment at existing nuclear facilities, study and evaluation of effectiveness of the technology toremove α nuclides contained in contaminated water generated during fuel debris retrieval began in FY2017 from the standpoint of keeping
exposure caused by α nuclides as low as reasonably achievable.
• “Adsorption” and “RO membrane” are selected as candidate applicable technology to remove soluble α nuclides. “Adsorption” does not
require secondary waste treatment as α nuclides are captured with adsorbents, but “RO membrane” requires treatment of concentrated
water which is generated by separating α nuclides. This PJ conducted technical development of “adsorption” which requires smaller
equipment scale.

• Subsidy projects up to FY2020 conducted adsorption performance assessment of candidate adsorbents using five elements, Pu, U, Am,
Cm, and Np, which are α nuclides to be removed. This assessment was conducted under open-air conditions to screen many adsorbents.
• Impact on adsorption performance in the environment of actual equipment has not yet been grasped. Additionally, adsorption performanceassessment was conducted using immersion tests up to FY2020, and performance during water flow treatment has not yet been grasped forthe study of actual equipment.
• Thus, this PJ studied evaluation methods to grasp the impact on adsorption performance by simulating the environment of actual equipment
and grasp performance during water flow treatment for the study of actual equipment.
©International Research Institute for Nuclear DecommissioningNo.12

7. Implementation items① Development of soluble α nuclide removal technology
・Position of this PJ(Operator engineering)[Previous PJ (FY2019, FY2020)]

Resetting of system
configuration andrequirements ofsoluble α nuclide
removal equipment

Restrictions of arrangement ofequipment inside existingbuildings (space, dose, etc.)

Organization of liquid
phase system pattern

Preliminary engineering
・System configuration
・System requirements

Adsorption water flow
test
Candidate adsorbent

Adsorption immersion
test

Test plan simulating
PCV environment

Necessity of simulating
gas phase environment
Evaluation
methodNp)・Container adhesion impact
・Particles generation impact

Adsorption
immersion test
(Pu, U, Am, Cm,
Np)

Verification test using actual contaminated water[This project (FY2021, FY2022)]

・Test procedures (after revision)・Adsorbent to be evaluated(Pu, U, Am, Cm,

Equipment requirements(performance, flow rate, maintenance,
quality, etc.)

Policy on application toactual equipment

Test plan(revision of procedures,
etc.)

Preliminary test

[Draft development plan after FY2023]

Candidate applicable adsorbent

・Organization of PCV gas
phase environment
・Pre-test
・Test procedures・Adsorbent to be evaluatedPreliminary test
simulating PCV
environment
・Adsorption test

Adsorption
performance data, etc.

・Applicable adsorbent
・Water flow conditions, etc.
Conceptual study of
α nuclide adsorption
removal equipment

Form of existence anddissolved amount of αPreliminary test
・Contaminated water to be
used
・Test method, etc.

Test plan using actual
contaminated water

Necessity of actual contaminated water

Study and evaluationof applicability toactual equipment

Basic / detailed
designApplication for permission
and authorization

Component
manufacturing

Adsorption performance data

Conceptual studyof α nuclide
adsorption
removalequipment

Design issues

Study of evaluation
methods to solve

Issues in performanceassessment

issuesOrganization of water
quality data of actual
contaminated water

Study of method tosecure test solution using
actual contaminatedwater

On-site construction
・Trial run
Operation andmaintenance

Figure Overall flow of development of soluble α nuclide removal technology©International Research Institute for Nuclear DecommissioningNo.13

7. Implementation items① Development of soluble α nuclide removal technology
・Issues to be addressed in this PJ, and implementation items as well as goals
a.Study of soluble α nuclide removal test assuming use of
actual liquid[Issues]•
There is lack of data such as adsorbent life and impact ofcompeting ion, regarding α nuclide adsorbents selected in
the subsidy project conducted in the previous fiscal year
(FY2019-FY2020).
•
Method of performance assessment test of candidate
adsorbents using actual liquid (accumulated water in reactor
building) has not been established.
[Implementation items]•
Organization of data required for design and operation ofsoluble α nuclide removal equipment.
•
Study of test methods assuming use of actual liquid(accumulated water in reactor building).
•
Study of test equipment and plans assuming implementationof column water flow test with actual equipment.

Impregnatedactive carbon(Colloid)

Inlet water
After debris particle
removalIncludes soluble α
nuclideImpregnatedactive carbon(Colloid)

Remove colloidal α nuclide,U, Pu, etc.

Zirconium(Pu, Am,
Cm, Np)Zirconium(Pu, Am,
Cm, Np)Remove soluble Pu, Am,
Cm, NpTitanicacid(Np, Sr)

Outletwater
After debris particle
removalAfter soluble α nuclide
removalRemove soluble Np(when Np concentration is not
sufficiently reduced with zirconium)

Example of adsorption tower configuration using candidate adsorbents

R/B
Accumulated
water in PCV[Issue] The following items
have not been evaluated➢ Necessary number of
towers for each adsorbent➢ Performance maintenanceperiod of each adsorbent(generated amount of
secondary waste)

Accumulated waterof torus room
Sample ofactualliquidCandidate actual contaminated water to be
used in the test
Adsorbent

Filter[Goals]•Materialization of test methods assuming use of actual liquid(accumulated water in reactor building).
•
Materialization of test equipment and plans assuming use of
actual liquid (accumulated water in reactor building).

Titanicacid(Np, Sr)

Constant flow pump
ColumnConcentration
analysis

Adsorbent
Liquid waste
tankTreatment water

Sample ofreceiving tankactual liquid
Draft test plan using actual liquid (immersion test, water flow test)©International Research Institute for Nuclear DecommissioningNo.14

7. Implementation items① Development of soluble α nuclide removal technology
・Issues to be addressed in this PJ, and implementation items as well as goals

b. Element test assuming fuel debris retrieval work
[Issues]• Behavior of α nuclides in solution has not been confirmed underconditions simulating the environment inside PCV during fuel debris <Image of test equipment to simulate nitrogen injection
environment during fuel debris retrieval work>
retrieval.
• Impact of changes in solution water quality on soluble α nuclide
removal performance has not been evaluated.
Nitrogen[Implementation items]• Fabrication of test equipment to simulate theenvironment during fuel debris retrieval.
• Preliminary test to confirm dissolution behavior of α
nuclides under conditions simulating the environment
Simplified glove box
during fuel debris retrieval.
• Selection of applicable adsorbents upon conducting<Example of adsorption test>
Measure the substanceadsorption tests to confirm α nuclide removal① Immersion test
concentration with andwithout adsorbents at
・Establishmentofhighlyperformance under conditions simulating thethe end of the test toaccurate test procedures
measure the amount ofenvironment during fuel debris retrieval. At the same
・Implementation of tests withadsorption
time, also consider cases of changes in water quality.
Pu, U, etc.
[Goals]• Selection of candidate adsorbents upon acquiring α
nuclide adsorption performance data under anenvironment assuming fuel debris retrieval.
ShakerAdsorption test
Shaker(internal)
container• Setting of water quality adjustment policy of soluble α(external)
nuclide removal equipment.

Polyethylene
container

Adsorbent

Test solution©International Research Institute for Nuclear DecommissioningNo.15

7. Implementation items① Development of soluble α nuclide removal technology
・Development schedule of this PJ
Items to bestudiedMajor milestones

FY2021
4

5

6

7

8

9

10

11

FY2022
12

Interim report ▼

1

2

3

4

Interim report ▼

5

6

7

8

9

Interim report ▼

10

11

12

1

2

Final report ▼a. Study of tests using
actual liquid
b. Element test
1. Test plan

2. Test preparation(preparation of test
supplies)

Preparation of test equipment, etc.

3. Pre-test
4. Preliminary test /adsorption test
(Am, Cm, Np)(U, Pu)Preliminary test and adsorption test
using Cm, Am, Np
Preliminary test and adsorption
test using U, Pu

4. Summary

Remarks©International Research Institute for Nuclear Decommissioning3

No.16

7. Implementation items① Development of soluble α nuclide removal technology
・Setting of prerequisite system (Liquid/Gas Phase Systems)• The liquid system takes water from D/W (drywell) or S/C (suppression chamber) and water level is controlled to confine
contaminated water in PCV, and performs circulation cooling, which cools and injects water taken from D/W or S/C, to cool
fuel debris.
• In order to limit the range of α nuclides that migrate from fuel debris to liquid phase, α nuclides are removed according to its
form (particle removal filter / soluble nuclide removal adsorption tower) from water taken from PCV and reactor building
basement.
• The gas system controls negative pressure to confine dust, etc. in PCV, and injects nitrogen to scavenge hydrogen.
• Although air flows in due to negative pressure control, gas phase inside PCV is a nitrogen environment consisting mostly of
nitrogen.
Pretreatment
systemTo exhaustport

Gas phase filter

Boric acid (silo)

Exhaust fan
Equipment subject tothis project

Particle
removal filter
Fuel debris retrieval work water*Used only during regular use of boric acidCollected boric acidtank

CoolerDrain water /backwash water

Soluble nuclide
removal adsorption
tower

Tank
(including boric acidadjustment)

To existing
Soluble nuclide
water treatment
removal RO
facility
membrane
Concentrated Supernatant water
water

NitrogenParticle
removal filter
Drain water /backwash water

Soluble nuclide
removal adsorption
tower

Sludge collection
all filtering
equipment

To existing water
treatment facility

Supernatant water

Sludge collection Sludge
sedimentation
separator

Permeable water

Boric acidcollection
equipment*RO membrane: Reverse osmosis membrane
Supernatant water / permeable water

Ambient air

Groundwater

Collected boric acid solution

Sludge collection
sedimentation
separator

SludgeDehydration andstabilization
treatment system

Dehydration andstabilization
treatment system
To wastestorage

Supernatant water
Sludge (filter)

Sludge collection Sludge
sedimentation
separator

Dehydration andstabilization
treatment system

To canisterstorage* Used to collect drain water from rough particle removal treatment

Figure Liquid/Gas Phase Systems during fuel debris retrieval (subsidy project basis)
Note: This conceptual diagram is an example of equipment configuration (case with no leakage from PCV)

©International Research Institute for Nuclear DecommissioningNo.17

7. Implementation items① Development of soluble α nuclide removal technology
・Assumptions and action policy for soluble α nuclide removal system configuration••

Adsorption
tower

Adsorption
tower

The previous PJ studied equipment configuration consisting of receiving tank, adsorption tower, monitor tank,sampling equipment, etc., as a soluble α nuclide removal system when adsorption removal is applied.
Additionally, adsorption tests wereStirring pumpconducted using α nuclide elements to be
From particle
removalremoved in air atmosphere, and adsorption
equipment
tower configuration was proposed withTransfer
Monitorpumptank
impregnated active carbon expected to
Receiving
Dischargetankpumpmainly remove colloidal α nuclides at the
To existing
front and zirconium and titanic acidwater
treatment
Monitor
expected to remove soluble α nuclides atfacility, etc.
tankthe rear.
Sampling equipment•

This PJ shall conduct similar adsorption
tests in nitrogen atmosphere simulating gas
phase environment in PCV and evaluatesolubility of α nuclides and impact on
adsorption performance, thereby confirming
whether tower configuration and applicableadsorbents set in the previous PJ need tobe revised.

Impregnatedactive carbon

Impregnatedactive carbon

Remove colloidal αnuclide, U, Pu, etc.

Zirconium

Zirconium

Remove soluble Pu, Am,
Cm, NpTitanic
acid

Titanic
acid

Remove soluble Np

Figure Schematic system diagram and proposed tower configuration* of the soluble α nuclide removal system*Supplemental information on tower configuration• The number of towers in each adsorption tower is tentatively set at 2 to allow merry-go-round operations. The necessary number of towers of each adsorption tower needs tobe evaluated in the future.
• In order to reduce the number of adsorption towers as much as possible, adsorption tower of impregnated active carbon which has high removal performance of colloidal αnuclides in addition to soluble α nuclides and adsorption tower of zirconium which has high adsorption performance for multiple soluble α nuclides shall be installed. However,since Np adsorption performance of impregnated active carbon and zirconium is relatively high and concentration reduction during water flow treatment is insufficient, titanicacid adsorption tower shall be installed at the rear of zirconium.
©International Research Institute for Nuclear DecommissioningNo.18

7. Implementation items① Development of soluble α nuclide removal technology (b. Element test assuming fuel debris
retrieval work)
・Study of environmental conditions of actual equipment and policy on incorporation toenvironmental conditions of tests• Proposed management policy and proposed control values of PCV gas phase environment were organized in the below table based on
the status of operator engineering study.
• Oxygen concentration is an item related to the degree of air inflow. Since it is considered to set the control value below 4%, it shall beincorporated in the conditions in the test evaluation. (However, 4% is a tentative value and not a definitive value)• Since oxygen concentration in air is approx. 20%, atmospheric components in gas phase in PCV are estimated at concentrations lessthan 1/5 of that of air.
Table Proposed management policy and proposed control value of environment items during fuel debris retrieval (gas system)
No.

Environment items

Management
requirement

Proposed management
policyProposedcontrol value*1: Positive pressure management and equal
Undeterminedpressure management are options when negative*1
pressure management is not feasible

Remarks

PCV gas phasepressure

Gas phase confinement(leakage prevention)

Manage at negative
pressure by managing
exhaust flow rate2

Nitrogenconcentration

None(Management of nitrogen
supply amount)─

Hydrogenconcentration

Gas phase confinement(prevention of fire andexplosion)

Manage below the lower
explosive limit ofhydrogen by managing
nitrogen supply amountand exhaust flow rate

Less than 2%*2*2: Not a definitive value as it is a tentative value

Oxygenconcentration

Gas phase confinement(prevention of fire andexplosion)

Manage below Zr fireconditions by managing
nitrogen supply amountand exhaust flow rate

Less than 4%*3

Concentration increases due to inflow of air during
negative pressure management*3: Not a definitive value as it is a tentative value

Carbon dioxide
concentration

None

In the course of events─Concentration increases due to inflow of air during
negative pressure management

Dust concentration

Gas phase confinement(leakage prevention)

Manage below theprescribed activity
concentration by
managing exhaust flow
rate

3

4

5

6

Gas phase environment

1

Nitrogen concentration itself is not subject tomanagement

Undetermined *4: Assuming below concentration equivalent to*4
public exposure of 5mSv at total release©International Research Institute for Nuclear DecommissioningNo.19

7. Implementation items① Development of soluble α nuclide removal technology (b. Element test assuming fuel debris
retrieval work)
・Study of water quality conditions assumed for actual equipment• Gradual changes in water quality is expected before fuel debris retrieval work starts, due to circulation cooling operation with the liquid
system and PCV negative pressure management with the gas system.
• When fuel debris retrieval work starts, pH in liquid phase is expected to increase due to leaching of concrete components during cutting.
• Additionally, pH increase is expected to increase the amount of carbon dioxide in gas phase that dissolves in liquid phase.

Table Water quality of accumulated water in PCV and status of evaluation of α nuclide adsorbent performance in operational
states of liquid system / gas system and fuel debris retrieval
Operational state

STEP1

STEP2

STEP3

STEP4

Circulation cooling

×

○

○

○

PCV negative pressure
management

×

×

○

○

Fuel debris retrieval work

×

×

×

○

Negative pressure management of
Circulation of fresh water could
PCV increases carbon dioxidecause leaching of concrete
concentration of gas phase due tocomponents, etc. Since gas
inflow of air into PCV, which
phase is nitrogen atmosphere,
increases inorganic carbonic acid
there is no atmospheric influence.
concentration underwater.

pH is increased by increase of eluted amountof concrete components due to fuel debris
retrieval work. pH increase increases inorganiccarbonic acid concentration. Injection of boric
acid water is also assumed to prevent
criticality.

OutlineBasic water quality is thesame as that of fresh water,
since fresh water is injectedfrom outside and injectedwater leaks outside PCV.

Estimated pH

Around 7

Around 7-9

Around 6-9

Around 6-11

Inorganic carbonicacid* concentration(estimated)

Around 1ppm

Around 1ppm

Several ppm(influence of air inflow)

Several to several dozen ppm(influence of air inflow + pH increase)

Status of evaluation of α nuclide
adsorbent performance(Not applicable)(Not applicable)

Candidate adsorbents were selectedfrom adsorption tests up to FY2020

Not evaluatedWater
quality ofaccumulated
water in
PCV

* General term for constituent carbons of CO2, HCO3-, CO32- existing in liquid phase. Abundance ratiochanges according to pH of the solution.

©International Research Institute for Nuclear DecommissioningNo.207. Implementation items• In the previous PJ, thermodynamic equilibrium calculations were conducted forsystems with carbon dioxide in the gas phase. It was evaluated that under alkalineconditions, abundance ratio of dissolved α nuclides in forms of carbonate ion
coordination and hydroxide ion coordination increases compared to neutral
conditions.
• In nitrogen environment and atmospheric environment, chemical forms of α
nuclides may be affected under alkaline conditions since carbon dioxide
concentration changes.
• Additionally, under acidic conditions, abundance ratio of dissolved α nuclides inmetal ion forms (such as Pu3+) may increase compared to neural and alkalineconditions.
Sea water

Pu

Abundance ratio of dissolved
species (%)① Development of soluble α nuclide removal technology (b. Element test assuming fuel debris
retrieval work)
dioxide concentration in gas phase: 3ppm
・Inspection and evaluation regarding dissolved form of α nuclides (1) Carbon(simulated
nitrogen environment)

Percentage of formsof carbonate ion
coordination
increases above pH 9(2) Carbon dioxide concentration in gas phase: 350ppm(simulated atmospheric environment)

Sea water + CO2 (3ppm)
Sea water + CO2 (30ppm)

Hydrolysate (Pu(OH)4, etc.)

Abundance ratio of dissolved
species (%)

Pu concentration [mol/l]

Sea water + CO2 (350ppm)

Percentage of formsof carbonate ion
coordination
increases above pH 7

Molar concentration per
10Bq/ml of radioactive element

Figure 1 Evaluation result of Pu solubility by thermodynamic equilibrium calculations using carbon
dioxide concentration conditions in the gas phase as parameters (50 times diluted seawater)(Result of thermodynamic equilibrium calculations of previous PJ)

Figure 2 Evaluation result of Pu dissolved forms by thermodynamic
equilibrium calculations using carbon dioxide concentration conditionsin the gas phase as parameters (50 times diluted seawater)(Result of thermodynamic equilibrium calculations of previous PJ)©International Research Institute for Nuclear DecommissioningNo.21

7. Implementation items① Development of soluble α nuclide removal technology (b. Element test assuming fuel debris
retrieval work)
・Setting of element test conditions• Assumed water quality conditions of accumulated water in PCV were divided into patterns and organized in the following table from threeperspectives: ①management policy of each environment item during fuel debris retrieval, ②study result of water quality conditions
considered to predominantly change due to fuel debris processing, and ③analysis result of accumulated water in PCV conducted to date.
• For the evaluation assuming increase in pH due to elution of concrete components, the evaluation was conducted with around pH9, 10 whichassumes impact of pH decrease due to dissolution of carbonic acid to be minimal and sedimentation of carbonate not to be generated.
• Additionally, evaluation was also conducted assuming water quality adjustment (pH adjustment), as adsorption performance could be
reduced if water quality of water to be treated changes in actual equipment.
Scope of this PJ

Table Element test conditions in this PJ
Liquid conditionspH*

1

1000 times diluted seawater

6-7

2

200 times diluted seawater

6-7

No.

3

4
5

PurposeWater
qualityconditions
assumedfor actual
equipment

6

9

Conditions
assuming
water
qualityadjustment

Ambient air /nitrogenAround 9

Remarks

Regarding conditions that were evaluated up to the previous PJ, the policy is to diver the results of theprevious PJ, and tests shall be conducted with conditions whose results have not been acquired.
・Assume operation near the analyzed value of chloride ion concentration in accumulated water to date
(19ppm), due to the condition of low leaching of seawater component
・Assume increase in pH due to leaching of concrete components

Around 10
Nitrogen

200 times diluted seawater +Ca(OH)2
Sodium pentaborate water(7000ppm as B)

781000 times diluted seawater+ Ca(OH)2

Gas phase
conditions

Around 10
8-9Around 31000 times diluted seawater+ HCl・Assume operation near the upper control limit (100ppm), due to the condition of high leaching of seawatercomponent
・Assume increase in pH due to leaching of concrete components

Around 9

Ambient air /nitrogenAmbient air /nitrogenAround 5

Apply for impact assessment of injection of sodium pentaborate.

・In order to grasp the trend of which range, neutral, alkaline or acidity, is appropriate to be adjusted to,
conduct an evaluation under acidic conditions (pH3, 5) whose data has not been acquired to date.
・Tentatively set hydrochloric acid as the pH adjuster, in consideration of impact on rear equipment*: Set with initial pH©International Research Institute for Nuclear DecommissioningNo.22

7. Implementation items① Development of soluble α nuclide removal technology (b. Element test assuming fuel debris
retrieval work)
・Adsorption test procedures
•
•

•
•

Figure 1 shows the test system for evaluation simulating the environment inside PCV (nitrogen atmosphere).
Simplified glove box (simplified GB) is installed in the hood of the laboratory from the perspective of radiation control, and nitrogen
injection line is connected from the N2 cylinder to create nitrogen atmosphere inside GB. Oxygen concentration inside the simplified
GB is regularly checked, and nitrogen injection operation is accordingly performed so that oxygen concentration does not exceed
4%.
Temperature control inside the simplified GB is difficult due to space limitations. So, temperature is measured and recorded during
tests to collect data to consider the impact of differences in temperature if it emerges.
Figure 2 shows the image of adsorption test procedures. Amount of adsorption on the adsorbent is evaluated by shaking the test
solution containing adsorbent and α nuclide tracer (test container is PP (polypropylene) container) in the simplified GB (nitrogen
atmosphere) (21 days) and analyzing the α nuclide concentration in the solution after 0.1μm filtering.
Nitrogenatmosphere

Test container
N2 (nitrogen
atmosphere)

Laboratory

Concentration analysis(Evaluation of amount ofadsorption)

α nuclide(dissolved)

Hood

Test
solutionT
Adsorbent

Simplified glove box

Thermometer
Shaking for aprescribed period in
nitrogen atmosphereCON2cylinderTest specimen

Shakerα nuclide(dissolved)

2

CO2 densitometer

O2

Pass box

Figure 1 Test system simulating nitrogen atmosphereO2 densitometer

α nuclide(dissolved)
α nuclide(adsorbed)

0.1μm filter filtering(Solid-liquid separation)

Figure 2 Image of adsorption test
©International Research Institute for Nuclear DecommissioningNo.23

7. Implementation items① Development of soluble α nuclide removal technology (b. Element test assuming fuel debris
retrieval work)
・Adsorption test result: Alkaline conditions simulating concrete leached components• Immersion tests were conducted for zirconium and impregnated active carbon using Pu under alkaline conditions (initial pH adjusted to 9 or
10 for 1000 times diluted seawater) and nitrogen conditions.

• Although zirconium showed relatively high distribution coefficient (Kd), it was lower than that of impregnated active carbon, which made itdifficult to determine as an applicable adsorbent in alkaline conditions. There is possibility that adsorption performance may decrease underalkaline conditions.
• Kd of impregnated active carbon after 21 days of immersion is about the same as that of the previous PJ (liquid phase: around pH 6-7, 1000
times diluted seawater; gas phase: atmospheric environment), which indicates robustness to pH. However, pH decreased to 7 after
immersion, which shows the possibility that adsorption performance may decrease during water flow treatment with alkaline conditioned
solution.
• Based on this result, pretreatment to adjust pH upstream adsorption tower to a constant level could be effective, in order to demonstrate
stable removal performance even when the quality of water to be treated fluctuates in actual equipment.
1.E+06

Distribution
coefficient [ml/g]*1

Ci: Nuclide concentration of the sample added with adsorbent
after the immersion period
Cf: Nuclide concentration of the sample without adsorbent added
after the immersion period
V: Volume of immersed liquid phase [ml]m: Weight of adsorbent [g]*1: Distribution coefficient (Kd) is
calculated assuming decrease in Pu
concentration is entirely due toadsorption. If distribution coefficient ishigh, it is evaluated that equilibrium
adsorption is high and judged that
adsorption performance is high.

Kd [ml/g]

1.E+05

Range ofrelatively high Kd

7day

21day

ND

ND

1.E+04
1.E+03
1.E+02

1.E+01

ND: Kd is
calculated using
the lower limit ofquantification,
sinceconcentration after
immersion is
below thedetection limit

1.E+00
Zirconium
1000 times diluted
seawater
Initially pH 9

pH after 21 days of immersion (pH in parenthesis is
pH at the start of the test)
pH adjuster: Calcium hydroxide*27.8 (9.2)

Zirconium
1000 times diluted
seawater
Initially pH 10*2

8.3 (10.0)

Impregnated active carbon Impregnated active carbon
1000 times diluted seawater1000 times diluted
Initially pH 9seawater
Initially pH 10
6.8 (9.3)

7.2 (10.2)

Figure Pu adsorption test result (nitrogen environment / initial pH©International
alkaline)Research Institute for Nuclear DecommissioningNo.24

7. Implementation items① Development of soluble α nuclide removal technology (b. Element test assuming fuel debris
retrieval work)
・Adsorption test result: Range of pH adjustment upstream adsorption tower• To grasp the appropriate pH range for adjusting pH upstream adsorption tower at a constant level (Figure 1), Figure 2 shows
the result of distribution coefficient (Kd) of Pu and Cm to zirconium adsorbent in nitrogen environment at each pH.

• The adsorption test (immersion test) with solutions of initial pH 3 and 5 (1000 times diluted seawater added with hydrochloricacid) showed that Kd decreased by one order of magnitude than the condition with initial pH 9 and 10. Thus, it is possible that
adjusting the water quality to the acidic side could reduce adsorption performance.
• To summarize the results of immersion tests for initial pH to date, Kd is relatively high when the pH after immersion from initial
pH is ranged around pH 6.5-9. Since there is concern of decrease in Cm adsorption performance when pH becomes lower
than 6, the pH adjustment range was tentatively set around pH 7-9.

Water quality
adjustment
equipment

Water tobe treatedAdsorption
tower

Treatment
water

Adsorption
tower

…

Figure 1 Treatment flow of α nuclide adsorption removal assuming water quality adjustment
1.0E+05

ND

Kd（ml/g）Range ofrelatively high Kd

NDPu

1.0E+04

1.0E+03

1.0E+02
pH3

pH4-6

pH6.5-7.5

pH8-9

pH8-10Cm

ND: Kd is calculated using thelower limit of quantification, sinceconcentration after immersion is
below the detection limit*: The range in pH is due tofluctuations between initial pH
and pH after immersion, resulting
from the test environment andadsorbents.

Figure 2 Kd of Pu and Cm to zirconium adsorbent for each pH* (nitrogen environment)©International Research Institute for Nuclear DecommissioningNo.25

7. Implementation items① Development of soluble α nuclide removal technology (b. Element test assuming fueldebris retrieval work)・Selection of candidate adsorbents•••

Below figure shows the distribution coefficient (Kd) of candidate adsorbents subject to evaluation in this PJ, obtained
from adsorption tests (immersion tests) using Np, Pu, Am and Cm in nitrogen atmosphere around initial pH 6-9.
Pu, Am and Cm, which are considered to be contained in large amounts in water to be treated with actual equipment,
showed high distribution coefficient (Kd) with titanic acid, zirconium and impregnated active carbon. On the other
hand, silicotitanate was eliminated as a candidate since it may have poor Pu adsorption performance.
Therefore, titanic acid, zirconium and impregnated active carbon shall be set as candidate adsorbents, and Np, Pu,
Am and Cm shall be removed with combinations of titanic acid, zirconium and impregnated active carbon to reduce αnuclide concentration of water to be treated.
1.0.E+06

Kd（ml/g）

1.0.E+05

Np

*1

Pu

Am

ND

Cm

*1

*1: Result of 7 days of
immersion (basis is result of
21-day immersion test)

ND

1.0.E+04
1.0.E+03*1*1*2: α nuclide concentration inblank (no adsorbent) solution
drastically decreased(reference value)*1, 2

1.0.E+02*3: Result for U is omitted
since its priority for removal islow

1.0.E+01
1.0.E+00

Titanic
Titanic acidacid

Silicotitanate
Silicotitanate

Zirconium
Zirconium

Kd is calculated using the
Impregnated active
Impregnatedactive ND:
lower limit of quantification,
carbonsince concentration after
carbonFigure Kd of α nuclide to be removed to each candidate adsorbent *3

immersion is below thedetection limit(nitrogen atmosphere / around initial pH 6-9 / 1000 times diluted seawater)©International Research Institute for Nuclear DecommissioningNo.26

7. Implementation items① Development of soluble α nuclide removal technology (b. Element test assuming fuel debris
retrieval work)・Result of study of schematic system and identification of issues for application to actual
equipment

To particle
removalequipment
Fromparticle
removalequipment

P

ImpregnatedactivecarbonImpregnatedactivecarbonTitanic
acid

Zirconium

Titanic
acid

P
Stirring pump

pH adjuster

pH

Remove Pu, Am

Receivingtank

ZirconiumP

After fuel debris particle
removalIncludes soluble α nuclide
Around pH 6-11
(assumed)pH
adjustment
tank

P

Around pH 7-9
If precipitates that could
block the adsorption
tower are formed due
to pH adjustment

FilterP

Remove Pu, Am, Cm,
Np

Remove Pu, (Am), Np

・・・

Adsorption
tower

•

Below figure shows the proposed schematic system configuration when assuming adjustment of water quality upstream adsorption
tower, which was newly added to the development goal in addition to the conventional adsorption tower system configuration (areas
revised from the previous PJ is indicated in blue). A system that measures pH (assuming pH 5.5-11) of water to be treated in advance,
and feeds water to the adsorption tower after adjusting pH to 7-9 with the pH adjustment tank if water is alkaline (pH 10 or higher)
shall be assumed.
Three issues up to application of the system that assumes adjustment of water quality upstream adsorption tower to actual equipment
were identified.
Issue ①: Selection of pH adjuster (hydrochloric acid was tentatively set for the test, but pH adjuster shall be selected including othermeans)
Issue ②: Setting of range of pH adjustment (setting of target pH to be adjusted by chemical injection / setting of action policy for
deviation from target pH)
Issue ③: Formation of precipitates due to pH adjustment (determining whether filter is needed)

Adsorption
tower

•

Monitortank

Dischargepump

P

To existingwater
treatment
facility, etc.

Monitortank

Sampling equipment

Figure Proposed schematic system configuration for soluble nuclide removal (when assuming water quality adjustment)©International Research Institute for Nuclear DecommissioningNo.27

7. Implementation items① Development of soluble α nuclide removal technology (a. Study of soluble α nuclide removal test
assuming use of actual liquid)・Necessity of water flow test using actual liquid• By conducting immersion tests with conditions assuming changes in water quality during fuel debris retrieval, adsorbents withlarge equilibrium adsorption capacity for α nuclides are evaluated, and impregnated active carbon and zirconium are selected as
candidate adsorbents. However, since performance of candidate adsorbents during water flow has not been evaluated, it isnecessary to conduct water flow tests to evaluate applicability of the process to actual equipment and acquire necessary datafor design and operation.
• Water flow tests require much α nuclides, but it is realistic to conduct water flow tests using actual liquid (hereinafter actual
contaminated water) due to limitations of α nuclide tracer amount that can be handled at the laboratory. Therefore, the plan up to
the water flow test using actual contaminated water was studied.
Adsorption test
[previous PJ](under atmosphericenvironment)

Assumption of
prerequisite system
configuration

Candidateadsorbent

Candidate ①:
Adsorption tower

Design of actual
equipment

Scope of this PJ

Process conditions(water flow rate,etc.)

・Evaluation ofadsorbent forremoving Cs/Sr

Planning ofwater flow test

Planning ofadsorption test
Water flow test
Adsorption test
(under nitrogenenvironment)

There are adsorbents
with large equilibrium
absorption of α
nuclides and small
impact on changes inwater quality

Confirmation ofadsorbents with high α
nuclide adsorption
capacity
・Evaluation with liquid conditions
after water quality adjustment
・Evaluation of powder
adsorbents

Revision of
applicable
process (1)Candidate ②: Water quality
adjustment + adsorption tower

Candidate ①:
Adsorption tower

Candidate ③:
Adsorption tank

Confirmation ofadsorbents with high α
nuclide adsorption rate

α nuclideconcentration of outlet
water is below thetarget value* whenwater is fed to theadsorption tower

Candidate ④: Coagulationsedimentation + adsorption tower

Candidate ⑤: Adsorption
tower + RO

Revision of
applicable
process (2)Replacement
frequency ofadsorption tower isbelow the target value*
When development proceeds ideally
Selection of applicable
process

When the applicable process needs
to be revised*
Target α nuclide concentrationof outlet water:
Three conditions (tentative) of
4 Bq/L, 40 Bq/L and 400 Bq/L
were set in operator
engineering
Target replacement frequencyof adsorption tower:
Tentatively set at 1 month (30days) based on thereplacement frequency of
existing water treatment facility

Figure Flow up to selection of process applicable to actual equipment©International Research Institute for Nuclear DecommissioningNo.28

7. Implementation items① Development of soluble α nuclide removal technology (a. Study of soluble α nuclide removal test
assuming use of actual liquid)・How to proceed with water flow test

• It is necessary to verify removal performance of adsorption tower during water flow treatment and its adsorption performance maintenanceperiod, before starting operation of actual equipment. It was confirmed that such verification requires water flow test (verification test) in
1/10 scale of actual equipment through literature review.
• Since it is necessary to set adsorption tower configuration in order to set the specifications of verification test equipment, water flow test
shall be conducted using column of actual equipment length with the same water flow conditions (SV *1 or LV*2) as those of actual
equipment to confirm that the target removal performance and performance maintenance period of the set adsorption tower is satisfied.
• While water flow conditions can be matched with those of actual equipment for the test with column of actual equipment length, it requires alarge volume of test solution. Thus, water flow test shall be conducted matching SV of water flow conditions with that of actual equipment
and using mini column which requires a small amount of test solution, in order to confirm feasibility of adsorption tower configuration
selected in this PJ.
• Based on the above, a plan was developed to proceed with the water flow test up to application to actual equipment in the following order:
water flow test with mini column, water flow test with column of actual equipment length, water flow test (verification test) in 1/10 scale.
• The following page shows the study on the materialization of the plan of water flow test with mini column to be conducted in the first step.

✓✓

Water flow testwith minicolumnWater flow test withcolumn of actual
equipment length

Narrow down from multiple
candidate adsorbents
Narrow down from multiple
candidate adsorption tower
configurations✓✓

Set the adsorbent andadsorption tower configuration
Set the number of actual
adsorption towers

Water flow test in
1/10 scale of actual
equipment✓✓

Trial run and startof operation of
actual equipment

Verify removal performance during
treatment with actual equipment
Verify adsorption performance
maintenance period during
treatment with actual equipment

Figure Proposed way to proceed with water flow test up to application to actual equipment*1 SV: Space velocity *2 LV: Linear velocity©International Research Institute for Nuclear DecommissioningNo.29

7. Implementation items① Development of soluble α nuclide removal technology (a. Study of soluble α nuclide removal test
assuming use of actual liquid)・Study of test equipment configuration and test conditions of water flow test with mini column
•
•
•
•

•

Test equipment configuration and test conditions of water flow test with mini column were studied.
Water flow test solution is prepared in the inlet tank and fed to the test column. A filter is installed at the front of the column to capturesolids. Sample collection points are also secured to evaluate changes in concentration and pH before and after water flow.
In order to secure a certain period of water flow, several dozen liters (target) of test solution for several thousand BV* is secured.
Two cases, 1E+03BQ/L (same level as current actual contaminated water) and 1E+05Bq/L (assuming two-digit increase in concentrationcompared to the current actual contaminated water), are set for gross α concentration of water flow test solution to prepare test solutions
containing the same amount for the water flow test.
Currently, no actual contaminated water with gross α concentration exceeding 1E+05Bq/L after filter filtering has been confirmed. Thus, itis necessary to study solution preparation methods to secure appropriate concentration for the water flow test.
*BV: Bed Volume; water flow rate per adsorbent filling volume
Inlet tank
Outlet tank
Sampling
・・・
Water flow test
solution

0.1μm
FilterSeveral dozen litersof water flow test
solution are
required per test

P

Test column
Solution after
water flow test

Sampling

Sampling

Sampling Sampling

Sampling

Figure Schematic system configuration of test equipment of water flow test with mini column (draft)©International Research Institute for Nuclear DecommissioningNo.307. Implementation items① Development of soluble α nuclide removal technology (a. Study of soluble α nuclide
removal test assuming use of actual liquid)・Study of method to secure soluble α nuclide concentration in test solution• According to the disclosed analysis result of actual contaminated water, most α nuclides contained in contaminated water are present in
non-soluble form and the percentage of soluble form is small. Thus, it is necessary to perform acid dissolution of non-soluble α nuclides inorder to secure ionic α nuclide concentration. Dissolution behavior when adjusting to the water quality assumed for actual equipment shallalso be confirmed. (No.3-8 in below figure)• However, if acid dissolution and water quality adjustment are performed, ion amount in the solution may drastically differ from liquidconditions for treatment with actual equipment and preparation operation itself may be difficult. Thus, solution preparation methods with noacid dissolution and water quality adjustment shall be assumed. (No.1-2 in below figure)• Solution preparation methods for water flow test solutions shall be set by testing such solution preparation methods using actual
contaminated water and confirming soluble α nuclide concentration in the solution.
Concentration analysis
Actual contaminatedwater

No water quality
adjustment

Water intake with sludge

1.0E＋0-3Bq/L insolid-liquid
separation samples

Filter filtering
Preparation of test solution using actual contaminated water

Acid dissolution
Acidic water
quality conditions

Filter filtering

Ionic α nuclides
are dominant
Approx. 1.0E＋
5Bq/L in solid-liquid
separation samples

Adjusted to assumed
water qualitypH≒7, adjustment of Cl, Ca, B concentration

Filter filtering

Assumed water
quality conditions
for fuel debris
retrieval

Adjusted to assumed
water quality

Mixture of ionic andcolloidal α nuclides

Filter filtering*Indication of stirring andsettling is omitted

Figure Proposal of securing water flow test solution©International Research Institute for Nuclear DecommissioningNo.31

7. Implementation items① Development of soluble α nuclide removal technology (a. Study of soluble α nuclide
removal test assuming use of actual liquid)・Identification of issues for water flow tests using actual contaminated water• Below figure shows how to proceed with the water flow test before operation of actual soluble α nuclide removal system assuming theadsorption tower method, based on the study results so far.
• Simulated demonstration is required before operation of actual equipment, in which water flow test equipment in 1/10 scale of actual
equipment is required. As the application phase to set adsorbent / adsorption tower configuration and number of adsorption towers, waterflow test with column of actual equipment length and water flow test with mini column shall be conducted. Such tests shall be conductedupon confirming whether the test solution can be secured, as securing the test solution is an issue.
• Future issues of water flow tests using actual contaminated water were organized as follows.
Issue ①: Setting of policy to acquire data for setting conditions for operation of actual equipment
Issue ②: Setting of sampling specifications (target, location, method, etc.) for actual contaminated water
Issue ③: Materialization of water quality adjustment method of actual contaminated water (including whether it can be implemented at
the site)
Issue ④: Study of alternative means for cases when α nuclide concentration suitable for water flow tests cannot be stably secured
Issue ⑤: Study of materialization of water flow test equipment
Issue ⑥: Selection of areas to implement water flow test
Issue ⑦: Setting of analysis specifications (measurement/analysis items, analysis framework) for water flow test solution
Applied research(TRL 3)

Immersion
test

Application research(TRL 4)

Water flow
test
(mini column)

Water flow test(column of actual
equipment
length)Simulated demonstration(TRL 5)

Field demonstration(TRL 6)

Actual operation(TRL 7)

Water flow test
(1/10 scale ofactualequipment)

Verification of
performance
during trial run ofactual equipment

Operation of
actualequipment

Confirmation of whether test
solution can be secured in advance

Figure Flow of water flow test up to operation of actual soluble
α nuclide removal equipment (adsorption tower method)

TRL: Technology Readiness Level©International Research Institute for Nuclear DecommissioningNo.32

7. Implementation items① Development of soluble α nuclide removal technology
・Summary of issues of soluble α nuclide removal equipment using adsorption tower method•

Below table shows the status of efforts and progress on issues identified at the start of this PJ as well as remaining issues, towardapplication of soluble α nuclide removal equipment applied with adsorption tower method to actual equipment. (Issues requiring
test evaluation are indicated in red)
Table Status of efforts on issues for application of soluble α nuclide adsorption removal system to actual equipment (1/3)

No.

Item

Status of efforts and progress on issues

Remaining issues

1

At the initial stage of this PJ, 15 nuclides of 5 elements,U, Np, Pu, Am and Cm, were tentatively set as
nuclides to be used in exposure assessment and nuclides to be removed with the Advanced Liquid
Processing System.
Nuclides to be Analysis results of contaminated water in different PJ found that soluble α nuclides hardly exist in
removedaccumulated water in the building and that soluble α nuclides, mostly consisting of Pu-238, 239, 240 andAm-241, exist in accumulated water in PCV.
Additionally, tests conducted in this PJ simulating the environment and water quality of actual equipment
found that it is highly possible that U can be eliminated from nuclides to be removed, as filtrate after
0.1μm filter filtering hardly contained any U.

・Whether U (U-234, U-235, U-236, U238) can be eliminated from elements
to be removed

2

At the initial stage of this PJ, regarding water quality of water to be treated, increase in Ca concentrationand pH was assumed due to elution of seawater components and concrete components with chloride ion
Water quality
concentration below 100ppm.
of water to be
The test simulating the environment of actual equipment, conducted in this PJ, found that pH istreatedmaintained at a somewhat high level and pH tends to decrease due to inflow of air in nitrogenenvironment assuming actual equipment. It was also confirmed that there is no extreme decrease inoxidation-reduction potential.

・Impact assessment with water quality
conditions (e.g. case when liquid isreturned from the secondary waste
treatment facility) that have not beenstudied3

The range of treatment flow rate was tentatively set to 2-10m3/h, assuming operation of actual equipment.
Treatment flow ・2m3/h: Assuming installation of the system outside the circulation cooling equipment and treatment of・Setting of treatment flow rate
rate
considering nuclide removal efficiency
22m3/day・10m3/h: Assuming installation of the system inside the circulation cooling equipment

4

α nuclideconcentrationin assumed
water quality

In the initial stage of this PJ, α nuclide concentration of water to be treated was not set and the policy was
to conduct adsorption tests under the concentration condition of 10Bq/ml from the perspective ofanalytical precision during tests.
Assumed conditions in operator engineering were discussed, and a policy to set 100Bq/ml as the
maximum value for concentration condition of gross α was established, which was incorporated in the
study and plan.

・Setting of concentration conditions foreach nuclide
・Confirmation of α nuclide
concentration with draft procedures forwater flow test solution preparation(test using actual contaminated water)©International Research Institute for Nuclear DecommissioningNo.33

7. Implementation items① Development of soluble α nuclide removal technology
・Summary of issues of soluble α nuclide removal equipment using adsorption tower method• Below table shows the status of efforts and progress on each issue as well as remaining issues. (Issues
requiring test evaluation are indicated in red)
Table Status of efforts on issues for application of soluble α nuclide adsorption removal system to actual equipment (2/3)
No.

5

Item

Status of efforts and progress on issues
At the initial stage of this PJ, target DF to reduce public exposure impact was set to 100, andConcentration announcement density was the concentration reduction target for transport to the water treatment
reduction
facility. Assumed conditions in operator engineering were discussed, and a policy to set DF of gross α
requestto 10, 100 or 1000 according to inlet conditions was established, which was incorporated in the study
and plan. Additionally, the test plan was studied since water flow tests using actual contaminatedwater are required to verify DF.
Immersion tests were conducted using U, Np, Pu, Am and Cm, which are elements to be removed, toevaluate equilibrium adsorption of candidate adsorbents, and impregnated active carbon, zirconium
and titanic acid were selected as adsorbents with large equilibrium adsorption capacity and highprospects for application.
Such adsorbents have large equilibrium adsorption under neutral conditions, but since equilibrium
adsorption relatively decreased under alkaline conditions assuming concrete leached components, it
is inferred that appropriate water quality adjustment is necessary.

Remaining issues
・Setting of removal requirements foreach nuclide・Materialization of test plan using
actual contaminated water

・Setting of adsorbents that satisfy
removal requirements(including new candidate adsorbents)
・Setting of test methods that reducethe impact of fluctuations in water
quality during the test period
・Evaluation of generated amount of
secondary waste
・Investigation/evaluation of impact of
disposal of used adsorbents

6

Applicableadsorbent

7

Adsorption Confirmation by water flow tests is required to set the adsorption tower configuration, but since smalltower
scale tests in the laboratory is not sufficient for confirmation, confirmation methods with water flow
configuration tests using actual contaminated water were studied.

・Materialization of test plan using
actual contaminated water

Replacementtiming /replacement
frequency
Replacement
methodTarget replacement frequency was set at once a month, and a policy to confirm in the water flow test
using actual contaminated water that removal performance is maintained after feeding waterequivalent to one month of treatment was studied.

・Evaluation of removal performance
maintenance period・Materialization of test plan using
actual contaminated waterA replacement policy was set to replace the entire adsorption tower from the perspective ofconfinement.

None

8
9

10

Tower size

As the adsorption tower assuming treatment flow rate of 2-10m3/h, the same size as that of adsorption
tower of Advanced Liquid Processing System with system flow rate of 250m 3/day was tentatively set
None
as the study conditions for the water flow test using actual contaminated water.
©International Research Institute for Nuclear DecommissioningNo.34

7. Implementation items•① Development of soluble α nuclide removal technology
・Summary of issues of soluble α nuclide removal equipment for treatment process other thanadsorption tower method
Additional issues were organized based on results of study in this PJ.
Table Status of efforts on issues for application of soluble α nuclide adsorption removal system to actual equipment (3/3)
No.

Item

Status of efforts and progress on issues

Remaining issues

11

Performance assessment has been conducted up to the previous PJ in neutral conditions. In this PJ,
adsorption tests were conducted in acidic and alkaline conditions under nitrogen environment.
Water quality Equilibrium adsorption showed a decreasing trend in some conditions compared to neutral conditions. It
adjustment was inferred that pretreatment is effective so that pH of adsorption tower inlet water is constant near
neutral.
Accordingly, system configuration was studied with the assumption of water quality adjustment.

12

In July 2021, MSIV leakage water from PCV accumulated water was sampled and analyzed at Unit 3,
and it was confirmed that α nuclides in contaminated water are not only in ion forms but also include
particle and colloidal forms that remain after filter filtering (blue area in below figure). Since colloidal αnuclides may not be removed with the adsorption tower method, depending on the amount and water
quality, it was judged that it is necessary to set measures to remove colloids (water quality adjustment,
coagulation-sedimentation, batch treatment, etc.).

Colloidal
measures

・Selection of pH adjuster
・Setting of range of pH adjustment
・Formation of precipitates due to
pH adjustment

・Setting of policy to remove
colloidal α nuclides[Reference] Evaluation and analysis of α nuclides of samples of Unit
3 MSIV room◼
◼

Stepwise filters were installed for the test water, and collected materials and filtrate at eachfilter were analyzed.
Even with 0.02μm filtrate, Am and Pu are higher than the announcement density.

Test water
Particle size
10μm
filter

Filtrate
1μmfilter

Filtrate
0.1μm
filter

<0.02μm (filtrate)[Reference]
Particle size

Filtrate
0.02μmfilter

<0.02μm (filtrate)

Gross aSource: TEPCO HD “Progress oftreatment of accumulated water frombuildings March 31, 2022”

Filtrate©International Research Institute for Nuclear DecommissioningNo.35

7. Implementation items① Development of soluble α nuclide removal technology
・Summary[Results to date]No.
a-1

Implemented
item
Study ofsoluble α
nuclideremoval test
assuming useof actual liquid

a-2

b-1

b-2

Element test
assuming fueldebris retrieval
work

Goal

Development results

Materialization of test
methods assuming useof actual liquid(accumulated water inreactor building)

Regarding the development goal of soluble α nuclide removal equipment using the adsorption tower
method, a plan was developed to study steps for evaluating α nuclide adsorption removal performance andperformance maintenance period before application to actual equipment and proceed with water flow testwith mini column, water flow test with column of actual equipment length and water flow test in 1/10 scale.

Materialization of test
equipment and plans
assuming use of actual
liquid (accumulated
water in reactor
building)

Of water flow tests to be conducted before application to actual equipment, test equipment configuration
was materialized for the water flow test with mini column and the test solution preparation method using
actual contaminated water was studied.
Since the solution preparation method needs to be confirmed with actual contaminated water in order tosecure necessary α nuclide concentration for the water flow test, test details using actual contaminatedwater were studied and the test plan up to start of water flow test was materialized.

Selection of candidate
adsorbents uponacquiring α nuclide
adsorption
performance data
under an environment
assuming fuel debris
retrieval

Adsorption test (immersion test) was conducted for candidate adsorbents (impregnated active carbon,
zirconium, etc.) identified in the previous PJ, in nitrogen environment and alkaline conditions.
It was confirmed that impregnated active carbon achieves the same level of Pu removal rate in alkalineconditions as that in neutral conditions in the previous PJ. On the other hand, zirconium showed slightly
lower Pu and Cm removal rate in alkaline conditions, which confirmed decreasing trend of adsorption
performance in alkaline conditions for some adsorbents.
Thus, adsorption test assuming pH adjustment and adsorption tower system configuration was conducted,
and the same adsorbents as candidate adsorbents selected thus far were selected as candidates.

Setting of water quality
adjustment policy ofsoluble α nuclide
removal equipment

In adjusting pH at the front of the adsorption tower, adsorption tests were conducted in acidic conditions asevaluation to select the appropriate pH range. It was confirmed that α nuclide adsorption performance
becomes low in acidic conditions, and the policy to adjust pH near neutral was tentatively set.

©International Research Institute for Nuclear Decommissioning7. Implementation items

No.36① Development of soluble α nuclide removal technology
・Summary[Future issues to be studied]✓

Adsorption evaluation with water quality conditions that have not been studied thus far (impact assessment with return water fromsecondary waste treatment facility) shall be conducted based on studies conducted in other PJ.

✓

It is necessary to acquire necessary data for evaluation of applicability of the process to actual equipment as well as design andoperation (removal performance and adsorption performance maintenance period of candidate adsorbents during water flow), in
future water flow tests, and it is an issue to prepare a large amount of test solutions while securing the prescribed α nuclide
concentration. It shall be evaluated whether requirements of water flow test solutions can be satisfied by using the actual
contaminated water of Fukushima Daiichi.

✓

Additionally, materialization of the test plan using actual contaminated water shall be studied. It shall also be studied whether
necessary data can be acquired without using actual contaminated water, in case tests using actual contaminated water cannot beestablished.

✓

New candidate adsorbents that have not been evaluated thus far shall be evaluated, and disposal impact of used adsorbents shall be
investigated and evaluated regarding the candidate adsorbents.

✓

In addition to technical development with the adsorption tower method, which had been the development goal, applicability oftreatment processes such as water quality adjustment, coagulation-sedimentation and batch treatment shall be evaluated asmeasures against decreased treatment performance due to changes in water quality and measures for colloid components that could
not be removed with adsorption removal.

[Summary]
Since the possibility of changes in treatment performance due to changes in water quality was suggested from the element test
results, evaluation was conducted assuming water quality adjustment equipment, candidate adsorbents were selected on thepremise of pH adjustment, and schematic system configuration was tentatively set. A phased water flow test plan from small
scale to scale of verification of actual equipment shall be materialized in order to acquire necessary data for evaluation of
applicability of the process to actual equipment as well as design and operation.
©International Research Institute for Nuclear DecommissioningNo.37

7. Implementation items② Development of RO concentrated water treatment technology➢ Study flow
Technical development of other items (water treatment)[FY2019, FY2020 (previous PJ)]
Organization of liquid
system pattern(Operator engineering)・α nuclide adsorbent

・Necessity of RO membrane
・Necessity of RO concentrated water treatment technology

RO concentrated water
treatment
Selection of candidate
technology[FY2021, FY2022 (this PJ)]

・Selection of coagulationsedimentation treatment

Filter liquid waste
Technical development of
coagulation-sedimentation
treatment

Soluble nuclide
Development ofremoval technology

Preliminary engineering

・Coagulant

Technical development of other items(storage canister PJ)

Identificationof issues

Investigation of existing knowledge andorganization of issues for storage canisterstorage of sludge and slurry

・Increase in generated amount of sludge

Verification test using simulated liquid

Preliminary test

Verification test using
actual liquid

Technical developmentof sludge stabilization
・Optimization of treatment method
・Volume reduction due to dehydration, etc.

Adsorbent test
・Selection of powder adsorbent

Coagulant test
・Selection of coagulant
・Beaker scale test
Coagulationsedimentation test ・Equipment test

Technical
study

Adsorbent test
Coagulant test

・Result of selection of dehydration technology

Element
test

RO concentrated
water treatment test

・Treatment method
Drafting of RO
Conceptual
concentrated water
designtreatment method ・Equipment specifications

Detailed design©International Research Institute for Nuclear DecommissioningNo.38

7. Implementation items② Development of RO concentrated water treatment technology[Issues]•
Nuclide removal by coagulation-sedimentation is studied as a treatment method of RO concentrated water generated by RO
membrane, and it is necessary to establish treatment procedures.
•
As powder adsorbents and coagulants that can be used for RO concentrated water treatment have not been selected, its
data on nuclide adsorption performance and coagulation-sedimentation performance is insufficient.
Simulated RO
concentrated water(pure water +simulated nuclide)[Implementation items]•
Candidate powder adsorbents and coagulants shall be selected basedon literature review and actual use at existing water treatment facility.
•
Nuclide removal performance of powder adsorbents shall be
evaluated with adsorption tests.
•
Coagulation performance of coagulants on powder adsorbents shallbe evaluated with coagulation-sedimentation tests.
•
Applicability of the developed RO concentrated water treatment
procedures shall be evaluated with simulated RO concentrated water
tests in beaker scale and equipment scale.
[Goals]•
Evaluate nuclide removal performance of powder adsorbents andcoagulation-sedimentation performance of coagulants.
•
Select applicable candidates of powder adsorbents and coagulants.
•
Draft RO concentrated water treatment procedures.

Powder adsorbent
Coagulant
pH adjuster

Stirring

SludgeBeakerFigure Image of RO concentrated water
treatment element test

Before coagulationsedimentation treatment

After coagulationsedimentation treatment

Figure State of RO concentrated water treatment element test
©International Research Institute for Nuclear DecommissioningNo.39

7. Implementation items② Development of RO concentrated water treatment technology➢

Sludge collection system configuration
Collect solid components from liquid waste generated from non-soluble nuclide removal
equipment and RO membrane equipment

Soluble nuclide removal equipment

Non-soluble nuclide removal equipment
Accumulated
water in PCV

Cut particles suchas fuel debris(large particle)

Rough removal
systemIntermediate
removal system

Adsorption towerequipment

Final treatment
systemDrain /
backwash water

Backwash water

Extraction
Coagulant

To soluble
nuclideremovalequipment

Extraction
Coagulant

Natural sedimentation
separationor
Coagulationsedimentation separation

Powderadsorbent

To soluble
nuclideremovalequipment

SludgeCoagulationsedimentation
separation

RemovalExtraction

Coagulant

To watertreatment
facility

Supernatant

Supernatant

Cut particles suchas fuel debris(small particle)

To watertreatment facility

RO concentrated water

Backwash water

Supernatant

SludgeRO membraneequipment

Powder adsorbent
for nuclide
adsorption

RemovalSludgeCoagulationsedimentation
separation

Removal

To wastestorageTo
canisterstorageRoughly removed liquid waste treatment facility

Intermediate removal / final liquid
waste treatment facility

RO concentrated water
treatment facility

Figure Sludge collection system configuration concept©International Research Institute for Nuclear DecommissioningNo.407. Implementation items② Development of RO concentrated water treatment technology➢

Requirements of sludge collection system configuration
Liquids to be treated at each sedimentation separator andtreatment requirements are as follows.

*Estimated with water quality conditions set in thesubsidy project.
Underlined parts were revised incorporating the resultsof this PJ.

Table System requirements of each sedimentation separator
No.

Item

Roughly removed liquid waste
treatment facility

Intermediate removal / final liquid
waste treatment facility

RO concentrated water treatment
facility

1

Water to be treatedDrain water / backwash water
generated from rough removal
equipment

Backwash water generated fromintermediate removal equipment and final
treatment equipment

RO concentrated water generated from RO
membrane equipment

2

Contained particle
components(subject tocoagulationsedimentation)

Molten fuel, reactor internals, concrete
components, other solid components

Molten fuel, reactor internals, concrete
components, other solid components

Powder adsorbent for nuclide adsorption

3

Particle concentration

Several thousand to 10000 ppm

Several thousand to 20000 ppm

From filter element test results

From filter element test results

When adsorbent is added: Several hundred
ppm
After CF concentration: Approx. 10000 ppm(particle size 100 to several dozen μm)(particle size several dozen to 0.1μm)

From adsorbent element test results

45Treatment flow rate

Treatment flow raterequirement

Intermittent drain: 100 m3/year
From filter element test results

Intermittent drain: More than 0.7
m3/dayFlow rate to enable treatment at 80%
equipment availability

1.4μm MF membrane: 30 m3/year
0.1μm UF membrane: 10 m3/year
0.05μm UF membrane: 50 m3/year

8-11 m3/day

From filter element test results

Calculated assuming RO membraneconcentration factor of 2 to 3 times for debris
retrieval work water (22 m3/day)

1.4μm MF membrane: More than 0.2
m3/day0.1μm UF membrane: More than 0.063
m3/day0.05μm UF membrane: More than 0.4
m3/dayFlow rate to enable treatment at 80% equipment
availability

More than 14 m3/day

Flow rate to enable treatment at 80%
equipment availability

6

Operation method

Batch treatment
Assuming to complete treatment of 1 batch in one to several days

©International Research Institute for Nuclear DecommissioningNo.41

7. Implementation items② Development of RO concentrated water treatment technology[Issue] Amount of coagulating sediment (hereinafter sedimentation sludge) generated with the sedimentation
separation tank is large.
When assuming the sludge container container is the unit can size
(φ200×H400mm), the necessary number of times of discharge is estimatedas shown in the below table.

Sedimentation sludge properties
・Generated amount is approx. 10-15 vol% of
initial liquid volume
・Moisture content is around 90%
・High liquidity

Table Number of times of discharge of sludge collection container
Liquid waste to betreatedQuantity ofcollection container

Generated amount of
sedimentation sludge

Roughly removed
liquid waste

1 container/day

3m3/year

Final treatment liquidwaste<10 containers/day<30m3/year

RO concentrated
water

<120
containers/day➢➢<300m3/year

Filter liquid waste coagulationRO concentrated water
sedimentation test
treatment test
*SiO2 10000ppm*Magnetite 1000ppm

Figure Appearance of liquid waste after
coagulation-sedimentation treatment
The amount of sedimentation sludge generated from the sedimentation separation
tank for filter liquid waste treatment is relatively small, and it is inferred that the
number of times of discharge is an order applicable to actual equipment
The amount of sedimentation sludge generated from the sedimentation separation tank for RO concentrated water
treatment is massive, and the number of times of discharge needs to be revised as it is also massive

Items to be studied
①
②
③
④
⑤

Selection of adsorbents with higher removal performance
Study of optimization of addition amount of powder adsorbent
Study of optimization of addition amount of coagulant
Re-examination of sedimentation separation treatment procedures
and equipment
Study of volume reduction treatment of sedimentation sludge due todehydration①-④ ⇒ Study with element tests⑤ ⇒ Implement element test after literature review anddesktop study*Study in item ③©International Research Institute for Nuclear DecommissioningNo.42

7. Implementation items② Development of RO concentrated water treatment technology➢

Adsorbent test
[Purpose] Select adsorbents applicable to RO concentrated water treatment[Implementation items] Evaluate adsorption performance of powder adsorbents with stirring-typebatch test

Study of test system
In general, immersion-type batch test is conducted when evaluating
adsorption performance of adsorbents. However, since stirring
treatment is performed for RO concentrated water in thesedimentation tank, the test system*1 simulates the system of actual
equipment.

Figure State of adsorbent test (titanic acid)
Simulated liquid
waste(1000 times diluted
seawater, tracer)

Addition of powder
adsorbent

Sampling &analysis

Selection of powder adsorbent
Adsorbents to be used in this test were selected as shown in the following
table, with reference to actual usage at 1F and adsorption rate data acquired
through tests up to FY2020, based on literature review. Adsorbents reported
to have high adsorption performance for α nuclides to be removed (Pu, Am,etc.) were selected.
Table Result of selection of powder adsorbent*2

Figure Image of stirring-type adsorption test
*1 Since hot test is difficult with this test method,
all tests are conducted as cold test.
For α nuclides (Pu, etc.) that are difficult tosimulate with the cold tracer, item ①soluble adsorption rate data was used as
reference.

No.

Adsorbent typeParticle
size(average)

1

Titanic acid

4.08μmActual usage at the US Savannah River Site (batch treatment)

2

Titanium silicate

8.55μmActual use at Fukushima Daiichi

3

Active carbon

48.5μm4

Hematite

1.32μm5

Magnetite

1.96μm

It was also selected from adsorbents that
Actual use at Fukushima Daiichi
cannottake granular form and is difficult to fill
Iron oxide adsorbent. Has been reported of high Kd for α nuclides such
adsorption
towers.
as Pu.
Iron oxide adsorbent similar to hematite.

Remarks*2 Selected from α nuclide adsorbents.

©International Research Institute for Nuclear DecommissioningNo.43

7. Implementation items② Development of RO concentrated water treatment technology➢

Coagulant test
[Purpose] Select coagulants applicable to RO concentrated water treatment[Implementation items] Conduct coagulation-sedimentation test using simulated liquid
added with powder adsorbent
Powder adsorbent

Simulated liquid*1(1000 times
diluted seawater)

Coagulant (including pH adjustment*2)

Stirring blade

BeakerCreation ofsimulated liquid
waste

Sampling & analysis

Sludge components

Addition of
coagulant

Stirring

Settling

Figure Image of coagulant test
*1 Tracer removal rate is not evaluated, since whether coagulation
sedimentation of powder adsorbents can be performed is the main
item to be confirmed.Thus, tracers have not been added.

Figure Appearance of coagulant test equipment*2 Coagulants selected in this test were all acidic reagents, and since it needs to be
adjusted to the neutral range after addition, sodium hydroxide solution shall be used.

Study of test system
Stirring-type jar test similar to the adsorbent test shall be adopted.
A scaled-down system of treatment with an actual coagulation
sedimentation tank shall be simulated.

Selection of coagulant
Coagulants were selected as shown in the right table based on results ofliterature review and FY2020 test results.
In addition to coagulation sedimentation performance, redundancy for ionicstrength of liquid waste was also taken into consideration.

Table Result of selection of coagulant
No.
1

2

Coagulant

Remarks

High treatment performance has
High base PAC*3 been reported for liquid waste with
low ionic strength
High treatment performance forliquid waste containing boric acid
Aluminum sulfate was confirmed in tests of theprevious PJ*3 PAC: Polyaluminum chloride©International Research Institute for Nuclear DecommissioningNo.44

7. Implementation items② Development of RO concentrated water treatment technology➢

Result of selection of adsorbent and coagulant
Reagents were selected based on the results of adsorbent test and coagulant test.
Reagents for the RO concentrated water treatment test were set based on the
below selection result.
Titanic
Titaniumacid
silicatecarbon
ActiveHematite
Magnetite

Titanium silicate, impregnated active carbon, (titanic acid)
Powder adsorbents to be tests shall be titanium silicate and active
carbon among adsorbents that were evaluated to have high α nuclide
removal performance. One of each type was selected from adsorbents
for selective α nuclide removal and adsorbents for gross activity
removal.

Distribution coefficient kd [ml/g]

Powder adsorbent*High α nuclide removal performance was also confirmed with titanicacid, but it is not used in this test since the sample amount usablefor this test is small.

Figure Distribution coefficient in adsorbent test result

CoagulantHigh base PAC

Based on the coagulant test result, aluminum sulfate shall be selected whichshowed good results with redundancy to water and relatively low amount ofgenerated amount of coagulating sediments.
Tests using high base PAC were also conducted only under representative
conditions.

Aluminum sulfate

Agglomerate volume [vol.%]

Aluminum sulfate, (high base PAC)

Titanic acidTitanium silicate

ActivecarbonHematite

Magnetite

Figure Sedimentation sludge volume in adsorbent test result©International Research Institute for Nuclear DecommissioningNo.45

7. Implementation items② Development of RO concentrated water treatment technology➢ Selection of powder adsorbent and coagulant
Addition amount (concentration) of each powder adsorbent and coagulant was set as follows.
Powder adsorbent
Table Evaluation of required addition amount of powder adsorbent

Adsorbent
Additive
Eu (Am) concentrati
onDistribution coefficient [ml/g]
No.

Adsorbent

1

Titanic acid2
3

Titaniumsilicate
Impregnatedactive carbon

Cs

Sr

9.2.E+03

1.0.E+06

6.0.E+06<50ppm

1.6.E+06

1.3.E+05

1.1.E+06

50ppm

1.2.E+03

2.6.E+02

5.9.E+04

200ppmSetting of adsorbent additive concentration
Value that satisfied DF100 was set based on the
distribution coefficient obtained from the adsorbent
test. For the convenience of the test, margin that
does not fall under the lower limit of detection wasconsidered for analysis.

Coagulant
Table Evaluation of required addition amount of coagulant

Coagulant additive concentration [ppm]ActiveTitanic acid Titanium silicate
carbonNo.

Coagulant ＼Adsorbent

1High base PAC

100

1000

1000

2

Aluminum sulfate

20

200

50

Setting of coagulant additive concentration
In the coagulant test, additive concentration that
confirmed that SS concentration of supernatant canbe reduced below 20ppm was set.

©International Research Institute for Nuclear DecommissioningNo.46

7. Implementation items② Development of RO concentrated water treatment technology➢

Proposed RO concentrated water treatment procedures
RO concentrated water(Radionuclides including α nuclides are
dissolved)(1) Adsorption treatment
Adsorb dissolved nuclides by adding powder adsorbent andstirring for a certain period of time.

Start stirring
Addition of
powder adsorbent
Stirring (adsorption
treatment)
CF concentration
treatment

Powder adsorbent:
Titanium silicate
Active carbon, (titanic acid)
Additive concentration: 50-200ppm(1)

Stirring time: Less than 30minConcentrate SS components with filtration
SS concentration: >10000ppm(2)

Delivery of permeate (toexisting equipment)

Addition of
coagulant

Change to acidity ofaround pH3-4
Accordingly
measure pH(3)
Coagulant: Aluminum sulfate
Additive concentration: 2000-5000ppm

Addition of pH adjuster(NaOH)

pH adjustment range: pH7±1

Strong stirring

Stirring time: 5min

Weak stirring

Stirring time: 15min

Settling (stopstirring)

Settling time: More than 24h (depends
on amount of treated liquid)Supernatantseparation

Supernatant residual particle: 10-20ppm
Generated amount of sludge: Approx. 5vol.%Sludge dehydration(2) Concentration treatment
Concentrate solid components of liquid waste after
adsorption treatment, with the cross flow filter (CF).
*Repeat the process of reception of RO concentrated water→ (1) → (2) to secure a certain amount of CF concentrate.
(3) Coagulation-sedimentation treatment
Perform coagulation sedimentation of powder adsorbents byadding coagulant to CF concentrate, adjusting to the neutral
range with pH adjustment reagent and stirring for a certain
period of time. (generation of RO concentrated water sludge)(4) Delivery of supernatant and sedimentation sludge
Supernatant: Residual particles filtered and removed after
extraction.
Sedimentation sludge: Collected with sludge collection
container.
(4)(5) Sludge dehydration treatment
Perform dehydration treatment on sedimentation sludge withthe cartridge filter. Reduce moisture content to around 60%.

Delivery of supernatant (toexisting equipment)©International Research Institute for Nuclear DecommissioningNo.47

7. Implementation items② Development of RO concentrated water treatment technology➢

RO concentrated water treatment test
Treatment test of simulated liquid of RO concentrated water was conducted using sedimentation separation tank. Test
conditions were set based on the beaker-scale test result. This test confirmed feasibility of the developed RO
concentrated water treatment method.

Simulated target process
Powder adsorption additive concentration was set in two patterns, low concentration and high concentration, considering
whether concentration treatment is performed.
Adsorption
treatment

Coagulationsedimentation treatment
CFconcentration
treatment

Adsorption
treatment
Liquid waste

Powder adsorbent(Low/highconcentration)*

Coagulationsedimentation
treatment

CF concentration treatment was not performedin this test but simulated by adjusting toassumed SS concentration after concentration
treatment.
Supernatant
Analysis of residual amount oftracer, residual amount of SS andother components

Coagulant (+ pH adjuster)Stirring

Discharge due to extraction

Sedimentation
separation
tank
SludgeCollection container

Discharge by removing
collection container

Reception ofsimulated liquid
waste

Addition ofadsorbent

Addition of
coagulant

Stirring

Settling

Figure Image of test equipment (sedimentation separation tank) test procedures

Sedimentation sludge
Analysis of generation ratio, moisture
content, etc.

Data on nuclide removal performance with this treatment method as well as operability of each mechanism of thesedimentation separation tank and treatment time required for each process shall be acquired to be incorporated in
the study of operation of actual equipment.
©International Research Institute for Nuclear DecommissioningNo.48

7. Implementation items② Development of RO concentrated water treatment technology➢

Setting of test parameters
Parameters of the RO concentrated water treatment test were set as shown in the below table.
Table Test parameters in RO concentrated water treatment result

1

Item

Water quality conditions

No.

Setting
Cs

0.1 ppm

Sr0.1 ppm

Eu (Am)0.5 ppm

Nuclideconcentration(tracer concentration)

Study policy
Cs, Sr and Am are selected as representative nuclides contained in circulation cooling water.
Tracer concentration is the initial concentration that can evaluate DF100 (lower limit of detection: 1ppb).
However, Eu is set to 0.5ppm based on evaluation with pre-tests, since it has low stability unless it hascertain high concentration and significantly decreases due to precipitation.
Since Am is an α nuclide that is difficult to be handled, tests are conducted using Eu as a substitute due to its
chemical similarity.

Equivalent to 1000 times diluted
seawater

Chlorine concentration equivalent to accumulated water in PCV was set.

pH

7 (neutral range)

Initial value is neutral. In pH adjustment in coagulation-sedimentation treatment, pH is adjusted to pH
recommended for each coagulant.

Water temperature

25℃

Type

①
②

Additive
concentration

50-100ppm (no concentration
treatment),
10000ppm (with concentration
treatment)

Two conditions of low concentration and high concentration were set, assuming two patterns of case when
implementing concentration treatment with filter filtering and case without such concentration treatment.

Type

①
②

Selected from the viewpoints of coagulation performance for target particles and redundancy for ionicstrength.
Since there is concern of organic coagulants having an adverse effect on the treatment disposal process if
coagulating sediments contain organic materials, inorganic coagulants are selected to be studied.

Additive
concentration

20-100ppm (no concentration
treatment),<5000ppm (with concentration
treatment)

For liquid waste with SS concentration around 10000ppm, coagulant concentration of around 2000-5000ppm
is set as it was evaluated to require such value based on additional test results.

9

Stirring speed/time

Strong stirring: 157rpm/5min
Weak stirring: 52rpm/15min

Necessary stirring speed to achieve stirring capacity obtained under stirring conditions with 500mL beakerscale tests with the equipment scale (600L) is converted, evaluated and set. Two-stage paddle type structure
is adopted for the stirring blade.

10

Settling time

60min-24h

Set at 24h at maximum since 1 batch treatment in 1 day is the basis in the study

3

4

5

6

Powderadsorbent
conditions

Simulated liquidity

28

Coagulant

7

Titanium silicate (8.55μm)
Active carbon (48.5μm)

High base PAC
Aluminum sulfate

Water temperature is expected to be around 10-40℃ with actual equipment due to seasonal fluctuations, butsince the preliminary test of the previous subsidy project confirmed that temperature dependence is small,tests are conducted at 25°C normal temperature as a representative condition.
Adsorbents evaluated to have high α nuclide removal performance based on adsorption tests are selected.
Particle size is measured with the laser diffraction method.
*Titanic acid is not studied as only a small amount of samples for testing could be collected and a large
amount of industrial samples could not be collected. There is actual industrial usage at overseas sites.

©International Research Institute for Nuclear DecommissioningNo.49

7. Implementation items② Development of RO concentrated water treatment technology➢

Adsorption performance assessment result
Test conditions
Test solution: 1000 times diluted seawater, pH: 7, tracer: 0.1ppm (Cs, Sr), 0.5ppm (Eu), liquid volume: 600L, stirring
condition: 157rpm×10-120min
Coagulant: Aluminum sulfate (200ppm/50ppm), adsorbent: titanium silicate (50ppm), impregnated active carbon (200ppm)

Removal rate [%]Changes inremoval rate are
not observed overtime
Cs removal rate
Sr removal rate
Eu removal rate

Only Eu removal rate
is selectively highRemoval rate[%]

Removal rate is high for
all nuclides

Cs removal rate
Sr removal rate
Eu removal rate

Cs and Sr are
hardly adsorbed

Adsorption rate is extremely highfrom the initial stage of treatment
Elapsed time [min.](a) Titanium silicate (50ppm)

Elapsed time [min.](b) Impregnated active carbon (200ppm)

Figure Tracer in adsorption performance assessment result (removal rate)
•
•
•

•

Titanium silicate can remove Cs, Sr and Eu, and is superior in reducing gross activity concentration.
Impregnated active carbon does not remove Cs and Sr but can selectively remove Eu, and it is superior in reducing only α activity
concentration.
Both tracers have fast adsorption speed, and are considered possible to satisfy the prescribed removal performance (DF100) in a very
short treatment time. However, the adsorption test result shows that adsorption speed tends to slow down as addition amountdecreases, and a trend with longer required treatment time was confirmed.
Coagulation-sedimentation treatment did not show any influence such as breakaway of tracers adsorbed to adsorbent particles.
Removal performance and appropriate addition amount of adsorbents are observed from the viewpoints of equilibrium adsorption andadsorption isotherm.
©International Research Institute for Nuclear Decommissioning

No.507. Implementation items② Development of RO concentrated water treatment technology➢

Equilibrium adsorption evaluation

Since relative concentration ofadsorbents against tracer
concentration is large, it is observedthat equilibrium adsorption was
evaluated as low.

Adsorbent additive concentration Cm [ppm](a) Titanium silicate

Observed to be evaluated low
since equilibrium was notreachedEquilibrium adsorption W [mg/g]

Equilibrium adsorption W [mg/g]

Eu equilibrium adsorption of each adsorbent was evaluated based on the adsorption test result.
Adsorption amount at 120 min was used, assuming that adsorption equilibrium was reached at 120 min.

Assumed range ofequilibrium adsorption

Assumed value of Eu equilibrium
adsorption
Titanium silicate: 10-8.0mg/g
Impregnated active carbon: 2.52.0mg/g

Adsorbent additive concentration Cm [ppm](b) Impregnated active carbon

Figure additive concentration dependence of Eu equilibrium adsorption•••

Equilibrium adsorption of each adsorbent is considered to be in the range indicated in red. However, it is possible that low values were calculated
in this test due to dependence on adsorption speed and additive concentration.
Equilibrium adsorption showed little dependence on adsorbent concentration, and it was suggested that it may be nearly constant in a wide
concentration range.
Evaluation result of adsorption isotherm indicated the possibility of adsorption performance being maintained even when nuclide concentration
changes.
It was suggested that saturated adsorption of each adsorbent may be around the same level as equilibrium adsorption
confirmed in this test result.
©International Research Institute for Nuclear Decommissioning

No.51

7. Implementation items② Development of RO concentrated water treatment technology➢

Summary of adsorption performance assessment result
Eu removal performance of each adsorbent was evaluated as shown in the below table, based on this test result.
Table Evaluation result of Eu removal performance of each adsorbent
Eu distribution
coefficient*1
Titanium silicate 8.4E+04-3.8E+06ml/g
Adsorbent

Impregnatedactive carbon

2.7E+03-5.3E+05ml/g

Necessary additive
concentration*2

Eu equilibrium
adsorption

Treatmenttime*326mg/L (<50 mg/L)

10-8.0mg/g<30 min

190mg/L(<200mg/L)

2.5-2.0mg/g<30 min*1 Evaluation result in cases when lower limit may not have reached adsorption equilibrium.
*2 Evaluation using Kd on the upper limit side
*3 However, longer treatment time is required as adsorbent addition amount becomes smaller.

•

•

•

Eu distribution coefficient was evaluated as 1.0＋E04-06 for titanium silicate, and 1.0+E03-05 for impregnatedactive carbon. According to Eu distribution coefficient, it was evaluated that addition amount of more than
50ppm for titanium silicate and 200ppm for impregnated active carbon is required to meet DF100.
Additionally, if adsorbent addition amount is sufficiently large, adsorption speed is extremely fast, and it isthought to meet or exceed DF100 in a short treatment time. In this case, the required adsorption treatment time
is considered to be within 30min.
However, preliminary tests found that the time required to reach adsorption equilibrium tends to be longer if
adsorbent additive concentration is smaller. Regarding impregnated active carbon in particular, it wassuggested that treatment time of more than 120min. may be required if additive concentration is small.
*Since powder materials with small particle size were used, it is considered that surface area is large and adsorption speed is fast.

•

It was suggested that equilibrium adsorption may be nearly constant within the concentration range in this test
condition. However, dilute concentration range equivalent to actual liquid needs to be separately evaluatedsince equilibrium adsorption may decrease.

©International Research Institute for Nuclear Decommissioning

No.52

7. Implementation items② Development of RO concentrated water treatment technology➢

Coagulant performance assessment result (low SS concentration condition)
Here is the coagulant performance assessment result when coagulation-sedimentation treatment is performed for
liquid waste after adsorption treatment.
Test conditions (common)
Test solution: 1000 times diluted seawater, pH: 7, tracer: 0.1ppm (Cs, Sr), 0.5ppm (Eu), liquid volume: 600L,
stirring condition: 157rpm×10-120minNo.
SS concentration [ppm]

Initial SS concentration

Range ofconcentration
decrease due tocoagulant treatment

4-1
4-2

Table Test conditions (low SS concentration condition)
Coagulant
Adsorbent
Additive
AdditiveTypeType
concentration
concentration
Aluminum
50ppm
200ppmTitanium silicate
sulfateAluminum
Impregnated200ppm50ppm
sulfateactive carbon
Table Test result (low SS concentration condition)

It is difficult to
reduce SS
concentration toseveral ppm
Settling time [min.]Figure Transition of supernatant SS concentration incoagulant performance assessment

No.

SS
concentration
after treatment

Generatedamount ofsludgeSS
removalrate

Sludgemoisturecontent

4-1

18.1ppm

0.4 vol.%

63.8%

97.3 wt.%

4-2

12.5ppm

0.1 vol.%

93.8%

91.1 wt.%*Similar trend as that confirmed in beaker-scale test•••

It was confirmed that concentration of liquid, whose initial SS concentration of adsorbent was 50-200ppm, was reduced to 1020ppm with treatment using coagulants. Based on the reduction rate of SS concentration, it was suggested that several ppm ofparticles may remain in the supernatant after treatment.
If SS concentration requires more than DF100 (removal rate 99%), it is necessary to remove residual particles in supernatantwith filtration.
It was confirmed that sludge (sediment) generated from coagulation-sedimentation treatment has more than 90% moisture
content and its generated amount is around 0.1-0.4%. Since there is much water content, it is thought that sludge amount has
increased.
©International Research Institute for Nuclear Decommissioning

No.53

7. Implementation items② Development of RO concentrated water treatment technology➢

Coagulant performance assessment result (high SS concentration condition)
Here is the coagulant performance assessment result when coagulation-sedimentation treatment is performed for liquid wastethat had high SS concentration after CF concentration treatment.

Test conditions (common)
Test solution: 1000 times diluted seawater, pH: 7, tracer: not added, liquid volume: 600L, stirring condition: 157rpm×10-120min
Initial SS concentration (afterconcentration treatment)

Range of concentration decrease
due to coagulant treatment

Adsorbent

SS concentration [ppm]No.
Initial decrease
behavior is large

Table Test conditions (high SS concentration condition)

Type

Additive
concentration

4-3

Titanium silicate

10000ppm

4-4

Impregnatedactive carbon

10000ppm

Coagulant
AdditiveTypeconcentration
Aluminum
5000ppm
sulfateAluminum
2000ppm
sulfateTable Test result (high SS concentration condition)

No.

SS
concentration
after treatment

Generatedamount ofsludgeSS
removalrate

Sludge moisture
content

4-3

25.9ppm

4.1 vol.%

99.7%

83.9 wt.%

4-4

36.6ppm

3.8 vol.%

99.8%

81.3 wt.%

Settling time [min.]Figure Transition of supernatant SS concentrationin coagulant performance assessment*Similar trend as that confirmed in beaker-scale test•••

It was confirmed that concentration of liquid, whose initial SS concentration of adsorbent was 10000ppm, was reduced to around 20ppm withtreatment using coagulants. Sedimentation behavior immediately after starting treatment is especially noticeable, which decreased below
50ppm within 60min. Thereafter, sedimentation behavior hit the ceiling, and it was confirmed to asymptotically approach around 10ppm.
Similar to the test with low SS concentration conditions, it was confirmed that particles of around 10-20ppm remain in supernatant. It wassuggested that it may be difficult to reduce SS concentration below several ppm with coagulants independent of initial SS concentration.
It was confirmed that sludge (sediment) generated from coagulation-sedimentation treatment has more than 90% moisture content and its
generated amount is around 4%. It is considered that moisture content is slightly lower than that of sludge generated from treatment under low
SS conditions due to the consolidation effect.
©International Research Institute for Nuclear Decommissioning

No.54

7. Implementation items② Development of RO concentrated water treatment technology➢

Observations on removal performance of coagulants① SS concentration may be reduced to approx. 10ppm independent of initial SS concentration② Sedimentation performance of coagulants stabilizes as SS concentration becomes higher, and appropriate treatment ispossible within a somewhat wide range of additive concentration.
SS removal rate by coagulants can be simply evaluated with the below
equation.

SS concentration [ppm]

10000ppm

Range of concentration
decrease during high SS
concentration liquid waste
treatment

50-200ppm

10-20ppm(Lower limit that can be
reduced with coagulant
treatment)

Range of concentration
decrease during low
concentration liquid waste
treatment

𝑆𝑆𝑟𝑒𝑚𝑜𝑣𝑎𝑙𝑟𝑎𝑡𝑒[%]𝑖𝑛𝑖𝑡𝑖𝑎𝑙𝑆𝑆𝑐𝑜𝑛𝑐𝑒𝑛𝑡𝑟𝑎𝑡𝑖𝑜𝑛 ー 𝑙𝑜𝑤𝑒𝑟 𝑙𝑖𝑚𝑖𝑡 𝑜𝑓 𝑟𝑒𝑑𝑢𝑐𝑡𝑖𝑜𝑛10𝑝𝑝𝑚=
𝑖𝑛𝑖𝑡𝑖𝑎𝑙𝑆𝑆𝑐𝑜𝑛𝑐𝑒𝑛𝑡𝑟𝑎𝑡𝑖𝑜𝑛
× 100
Adsorption concentration satisfies nuclide removal requirements ataround 50-200ppm, but since efficiency of particle removal due tosubsequent coagulation-sedimentation treatment is low, it is considerednecessary to implement treatment that sufficiently increases SS
concentration.
Necessity of preliminary concentration treatment with CF filter, etc. is high.
(Organized as future development issues)

Settling time [min.]A system that implements concentration treatment for solid components prior to coagulation-sedimentation treatment is
considered more reasonable.

③ Amount of sludge generated from coagulation-sedimentation treatment tends to increase due to its extremely high moisture
content.
Coagulation-sedimentation treatment using coagulant may reduce sludge
moisture content, but 80% is considered to be the lower limit, and it isexpected to be accompanied by some amount of supernatant during
discharge.

It is difficult to
reduce moisture
content below 80%
with only treatmentusing coagulants.

Sludge moisture content[vol.%]

80vol.%

Low SS
concentration

High SS
concentration

Necessity of sludge dehydration treatment is high.
(Developed in development item ③ (Development of secondary waste treatment
technology))

It is considered necessary to implement sludge dehydration treatment to reduce its volume.
©International Research Institute for Nuclear Decommissioning

No.55

7. Implementation items② Development of RO concentrated water treatment technology➢

Observations on removal performance of coagulants
The cause of coagulation-sedimentation treatment resulting in SS concentration converging to around 10-20ppm
independent of initial SS concentration was observed.
Since a clear mechanism is unknown, the studied hypotheses is as follows.
When the number of coagulating particles and number of naturally dispersing particles
become balanced, growth of agglomerates apparently stops and decrease in SS concentration
due to coagulation sedimentation no longer occurs.

Coagulant neutralizes particle surface charges and cancels electrostatic repulsion between particles, which causes
coagulation of particles. However, the formed agglomerates are not bound together by strong forces, and certain
particles are expected to disperse over time. In order words, coagulation and dispersion occur in parallel, and if
coagulation has larger contribution rate, agglomerates grow large and sediment.
Particle coagulation occurs when particles collide with each other in liquid, and the probability of collision of particles
depend on the number of particles existing in liquid, or SS concentration. Therefore, contribution rate of coagulation
becomes smaller as SS concentration decreases, and it is assumed that coagulation and dispersion eventually
becomes balanced (equilibrium state).
The aforementioned state where coagulation and dispersion are balanced is considered to be 10-20ppm which is thelower limit of SS concentration that can be reduced as confirmed in this test.
High probability of collision between particles

Low probability of collision between particles
Sedimentation oflarge agglomerates

Adsorbent particle

SS concentration ≒10-20ppm
Coagulation
Dispersion

Coagulation Dispersion

Time lapse
Agglomerate
Sedimentation

Coagulation ＞ dispersion(a) High SS concentration (immediately after addition of coagulant)

Coagulation ≒ dispersion(b) Low SS concentration condition

Figure Image of coagulation and natural dispersion of particles due to coagulant©International Research Institute for Nuclear Decommissioning7. Implementation items

No.56② Development of RO concentrated water treatment technology➢

Analysis of properties of residual components after treatment
Assumed properties of supernatant and sedimentation sludge generated after RO concentrated water treatment wereorganized based on analysis data of each test.
Table Assumed properties of supernatant / sedimentation sludge generated after RO concentrated water treatment

Sedimentation sludge

SupernatantComponent / water
quality item

Initial concentration Concentration after[ppm]
treatment [ppm]Residual rate[%]

Remarks

Cl

17

17

―

Seawater component. Not measured in the test. Almost all amount is assumed toremain in liquid.

Al

9.21

0.067

0.7%

Coagulant component. Almost all amount is assumed to sediment as aluminumhydroxide, and Al concentration in liquid was less than 0.1ppm.

SO4

53

45

85%

Coagulant component. Sulfuric acid components tend to remain in liquid, butconcentration decreases by 10-20%.

Ca

2.72

2.32

85%

Concrete leached component. Concentration decreases by 10-20% due toprecipitation and coprecipitation, but it hardly changes from initial concentration.

B

201

199

99%

Boric acid solution component. Almost all amount is assumed to remain in liquid.
(In the case of 7000ppmasB, 5% at maximum precipitates; FY2019 element test
result)SS concentration

<200

<20

<10%

It was confirmed that concentration may be reduced to approx. 20ppm
independent of initial SS concentration

pH

(5-9)

7±0.5

―

Item

Assumed properties

Generated amountIn the case of low SS concentration: Approx. 1 vol.%In the case of high SS concentration: Approx. 4 vol.%Moisture content

90-99% (lower limit approx. 80%)

Components

Powder adsorbent, Al component (coagulant)

It is confirmed that there is no significant change from the pH adjustment range(pH7) after coagulation-sedimentation treatment.
Remarks
It is assumed that around 1 vol.% of sludge is generated against the treatment
liquid amount. However, approx. 4 vol.% is assumed for high SS concentration
liquid waste.
Many test results show moisture content of 80-90% or higher, independent oftreatment methods and water quality conditions, and the same is assumed withactual equipment.
Powder adsorbent is dominant, which contains aluminum hydroxide derived fromcoagulants. Coexisting ions contained in treatment liquid are assumed to remainin the liquid.

Concentration of Ca and B decreases due to precipitation and coprecipitation, but the amount of decrease is small, and almost all amount is delivered to the rear.
Additionally, Al components remain in sedimentation sludge and sulfuric acid components remaining in supernatant due to addition of coagulant, which isassumed to remain in liquid at SS concentration of around 20ppm.
©International Research Institute for Nuclear Decommissioning7. Implementation items

No.57② Development of RO concentrated water treatment technology➢

Study on delivery of supernatant to existing water treatment facility
Reception requirements of existing water treatment facility were studied based on
past records of reception of accumulated water of process main building (PMB).

*Set based mainly on FY2018 record of reception of PMB accumulated
water
Reception requirements of existing water treatment facility during thefuel debris retrieval work period is a future issue to be studied

Table Tentatively set values* of reception requirements of existing water treatment facility and assumed properties of supernatant
Item

Set value

Test result

Cl<700 ppm<20 ppm

Ca<50 ppm

2.3 ppm

Mg<30 ppm―

SO4<100 ppm

45 ppmB―

200 ppm

pH

7-8.5

7±0.5

SS
concentration<20 ppm

10-20 ppm

Gross αconcentration<7.4E+0 Bq/L

99% removal rate for
initial concentration(after 0.45μm filtering)

Remarks
If each ion concentration of circulation cooling water is below the assumed water quality condition, it isassumed that concentration in supernatant will not be higher than this item.
However, Cl concentration needs to be studied since accumulated water with high Cl concentrationsuch as torus room water may be treated.
If aluminum sulfate is used, a large portion of sulfate ions remains in supernatant. However, existingwater treatment facility has a record of receiving approx. 50ppm of sulfate ions on average and approx.
100ppm at maximum.
There is no record of being received. Upper limit of environmental release is 10mg/L. Thus, if boric acidwater is used, it is necessary to sufficiently reduce concentration with boric acid collection equipment at
the front.
Since pH of supernatant is adjusted to around pH7 in the process of coagulation-sedimentation
treatment, this item is assumed to be satisfied.
Since it is assumed that SS concentration can be reduced to 10-20ppm due to coagulationsedimentation treatment, reception requirements are assumed to be satisfied under SS concentration
criteria. Since particles remaining in supernatant have high activity, removal of DF100 or more isrequired or particle components. Thus, it is assumed that residual particles need to be removed with
UF (Ultra-Filtration) membrane.
Set based on maximum α concentration of inlet water of existing water treatment facility. It is organized
as a future issue to be studied.
It is necessary to evaluate it based on assumed gross α concentration of inlet water as well as nuclide
removal performance, addition amount and residual amount of powder adsorbents. Additionally,evaluation is necessary from the viewpoint of public exposure (DF100) and announcement density.

It is assumed that supernatant after coagulation-sedimentation treatment can be delivered to existing water treatment facility, considering
record of treatment at existing water treatment facility. However, it is considered that it is necessary to remove residual particles insupernatant with filtration, based on record of reception of existing water treatment facility with SS concentration and viewpoint of removalof activity adsorbed on adsorbents.
©International Research Institute for Nuclear Decommissioning7. Implementation items

No.58② Development of RO concentrated water treatment technology➢

Study of RO concentrated water treatment procedures
RO concentrated water treatment method was studied based on the development results.
Frequency of each treatment in the below table was tentatively set to simplify evaluation. It is necessary to set the actual treatment
frequency taking overall operability of the system into account.

No.

Treatment
procedures

Treatment
frequency

Generatedslurry/sludge

GeneratedamountSludge content

Remarks

1

Reception of RO
concentrated water―

RO concentratedwater

14 m3/day0%

Almost all solid components were removed with the particle removal
equipment at the front, and it is assumed that there is hardly any
solid components at the time of RO membrane treatment.

2

Nuclide adsorption
treatment(addition of powder
adsorbent)

1 batch / 1
dayAdsorption
treatment liquid

14 m3/day

0.005%(50ppm)Based on the test result of this subsidy project, addition amount of
powder adsorbent is assumed at 50ppm (0.005%; in the case oftitanic acid and titanium silicate).

Permeable liquid＜14m3/day

MinuteConcentration factor is assumed to be 100 times. It is assumed that
only a very small amount of particles pass through the cross flow
filter.

Cross flow
concentrate

0.14 m3/day

Approx. 1%(10000ppm)

It is assumed that sludge amount is concentrated to approx. 1%
with cross flow filtering. (concentrated water is generated)

Supernatant＜1.0 m3/week

0.002-0.005%(20-50ppm)

Based on the coagulation-sedimentation test result, residual particle
concentration is assumed at 20-50ppm. Residual particles are
removed with UF membrane.

Coagulationsedimentation
sludge0.1 m3/week<10wt.%

Coagulation-sedimentation treatment is performed for CF
concentrate generated in pretreatment to sludge nuclides. Sludge
content is assumed to be less than 10 wt.% and generated amountof sludge is assumed to be less than 10 vol.%.

Separated water＜0.4 m3/month

MinuteWhen applying cartridge filter filtering, a very small amount of SS isassumed to be contained in separated water.

3

4

5

6

Concentration
treatment with crossflow filtering

Coagulationsedimentation
treatment

1 batch / 1day

1 batch / 1
week

Sludge dehydration
treatment*Item ③

1 batch / 1monthLong-term storage―

Dry sludge

0.2 m3/month

Approx. 40wt.%

Sedimentation sludge is dehydrated to the target value of 60wt.%
water content (40wt.% sludge amount) by passing water through the
cartridge filter. However, water content may become larger
depending on the used powder adsorbent, and sludge content will
become approx. 20-30wt.%.

Dry sludge

0.2 m3/month

Approx. 40wt.%

Apply long-term storage with sludge content greater than 40wt.%.
However, stabilization treatment such as solidification needs to be
separately studied.
©International Research Institute for Nuclear Decommissioning

No.59

7. Implementation items② Development of RO concentrated water treatment technology➢

Study of RO concentrated water treatment flow
Here is the RO concentrated water treatment flow studied based on the development results.

Excess water 22m3/day
Circulation
coolingsystemConcentration rate 2 times
Soluble nuclide
removal RO
membrane

Permeable water 11m3/day

To existing water
treatment facility

RO concentrated water 14m3/day (margin rate 80%)
Adsorbent additive concentration
50ppm

Permeable water
14m3/dayCoagulant additive concentration
5000ppm

To existing water
treatment facility

Adsorption treatment liquid
Extraction frequency once/day
Generated amount 14m3/day

Powder adsorbent tank
Nuclideadsorption
Adsorption tankConcentration ofsolid components
Cross flow filter
Cross flow concentrate
Extraction frequency once/day
Generated amount 0.14m3/day

Coagulant tank
Sludging
Sedimentation
separation tankCoagulation-sedimentation sludge
Delivery frequency Once/week (tentative)
Generated amount 0.1m3/week

Supernatant water

Residual particle
separationUF filter
Separated water<1.0m3/week

Sludge dehydrationequipment*Item ③To existing water
treatment facility

Dry sludge
Delivery frequency Once/month (tentative)
To storage waste<0.2m3/month*≒3m3/year
storage line

*Assuming discharge with unit can size containers, the number of times ofdischarge is less than 20 containers per month (less than 1 container per day)To storage canisterstorage line

©International Research Institute for Nuclear DecommissioningNo.607. Implementation items② Development of RO concentrated water treatment technology[Results to date]✓As candidate powder adsorbents for α nuclide removal, impregnated active carbon was selected to selectively
remove α nuclides and titanic acid and titanium silicate were selected to remove gross activity including α
nuclides.

✓Aluminum sulfate was selected as the coagulant in order to perform coagulation sedimentation of powder
adsorbents. Additionally, data on generated amount and properties of supernatant and sedimentation sludge
generated as a result of coagulation-sedimentation treatment was acquired. A system applied with CF
concentration treatment was also drafted, based on effectiveness of coagulation-sedimentation treatment toward
high SS concentration liquid waste.

✓

Treatment procedures and treatment process for RO concentrated water were drafted based on test results.
Additionally, the equipment-scale test confirmed feasibility of the developed treatment procedures.

[Future issues to be studied]✓In this PJ, removal performance was evaluated using cold tracers of Cs, Sr and Eu. Regarding other nuclides
to be removed, it is necessary to evaluate whether the drafted treatment methods satisfy the prescribedremoval performance (evaluation of removal performance of gross activity and gross α in water quality of actual
liquid).

✓CF concentration treatment was not subject to development in this PJ, but its system was studied through casestudies and literature values. It is a future issue to conduct CF filter tests and evaluate whether concentrate
factor satisfies the prescribed required performance.

[Summary]
Methods to sludge and separate nuclides in ROC concentrated water through treatment using adsorbents andcoagulants were studied. Reagents were selected through element tests, required addition amount was evaluated,
and RO concentrated water treatment procedures were drafted. Feasibility of this treatment method was confirmed,
and conceptual study of the treatment process was conducted.

©International Research Institute for Nuclear DecommissioningNo.61

7. Implementation items③ Development of secondary waste treatment technology➢ Study flow
[FY2021, FY2022 (this PJ)]

[FY2019, FY2020 (previous PJ)]

Organization of liquid
system pattern

・Properties of generated
liquid waste

Organization of issues intechnical development of RO
concentrated water treatment

Coagulationsedimentation test
・Selection of coagulant
・Beaker scale test
・Equipment test

・Acquisition of operation data
・Selection of particle collector
・Properties of generated liquid waste

Identification ofissuesConceptual design

・Increase in differential pressure ofintermediate removal equipment
・Increase in generated amount ofliquid waste

Conceptual design

Technical developmentof sludge stabilization

・Coagulationsedimentation method

・3-system water flow method

Proposal of 2-systemwater flow method

Investigation of existing knowledge andorganization of issues for storage canisterstorage of sludge and slurry

Conceptual study ofsludge collection system

Conceptual study ofparticle removal system

Filter test

Technical development ofother items(storage canister PJ)

・Generated amount of sludge
・Sludge properties
・Equipment specifications of
sedimentation separationtank・Properties of generated liquid waste
・Draft system configuration anddraft operation
・Generated amount of liquid waste

Identification ofissues・Optimization of treatment method
・Volume reduction due todehydration, etc.

Setting of
dehydration goal
・Evaluation of generated
amount of hydrogen

Technical studyof dehydration
・Result of selection of
dehydration technology

Planning of element test

・Draft test
Evaluation of applicability
to actual equipment withelement test
・Dependence of particle form distribution
・Dehydration of sedimentation sludge

Dependence
evaluation of particleform・Reselection of intermediate
removal equipment
・Differential pressure behavior,
backwash performance
・Generated amount of liquid waste

Conceptual design
・Draft system
configuration

Detailed design©International Research Institute for Nuclear Decommissioning7. Implementation items

No.62③-1 Development of non-soluble nuclide removal technology➢

Water quality conditions of input water of particle removal system
Properties of accumulated water transferred to and treated with particle removal equipment were assumed asshown in the following table.
Table

Assumed water quality conditions of input water of particle removal system

No.

Item

Conditions

1

Temperature

Normal temperature

2

pH

5-9

3

Ionic strength

Seawater component, boric acid,
other

4

Particlecomponents

5SS
concentration<100ppm

Based on the processing test of simulated fuel debris conducted in the retrieval method PJ as well as
evaluation result of target throughput, SS concentration of particles transitioning from PCV to liquid
phase system was evaluated to be below 100ppm.

6Particle size

0.1-100μm

Particles transitioning from PCV to liquid phase system were evaluated to have particle size smaller
than 100μm, based on natural sedimentation speed and pump suction capacity.

7

Particle size
distribution

Unknown

Since it is unknown at present, broad distribution of 100-0.1μm is assumed.
Possibility of double-peaked distribution was suggested from the simulated fuel debris processing test
result of the characterization PJ.

Fuel debris, reactor internals,
concrete

Study status

Assuming seasonal fluctuation of 10-40℃.
Assuming neutral range. Basic water quality conditions of liquid phase system
Ion strength is assumed to be low in principle. However, ion strength is assumed to become relatively
high when there is intrusion of seawater components and when boric acid water is used.
At present, it is assumed that the three components on the left that have a large amount present in
PCV are dominant. However, the respective content ratio is unknown. It is also possible that thecomponents contain a very small amount of deposits (soil, sand, etc.) and precipitates.

8

Particle form

Spherical, non-spherical・Nearly spherical particles are generated in the case of heat input processing
・Flaky and block-like particles are generated in the case of machine processing
Tests using spherical particles were conducted in subsidy projects up to FY2020.
In this PJ, filter water flow test using non-spherical particles was conducted to evaluate the impact ofparticle forms.

9

Treatment flowrate

52 m3/day (subsidy project),
22 m3/day (engineering)Work is carried out at a rate of 10h/day with 2.2m3/h of debris processing work water and 3m3/h ofreactor injection water. System configuration studied in engineering only treats debris processing work
water.

©International Research Institute for Nuclear DecommissioningNo.63

7. Implementation items③-1 Development of non-soluble nuclide removal technology➢

Study of particle removal system configuration
Non-soluble α nuclide removal equipment (particle removal system)< Several
dozen μm

Roughremoval(large particle)

Accumulated
water in PCV
Contains chip powder
generated from fuel debris
processing

Intermediate < Several μm
removal(large to small
particle)

Drain water / backwash water

Backwash water

Final treatment(fine particle)< Several dozen to
several hundred nm

Soluble α nuclide
removal equipment

Backwash water

Drain water / backwash water treatment facility (sludge collection)
Figure Conceptual diagram of particle removal equipment system configuration➢

Result of selection of candidate particle collector
Applicability of particle collectors of each system were evaluated based on literature review result and test result up to FY2022
Table Result of applicability assessment of candidate particle collectors

No.

1

2

3

SystemParticle collector

Filter mesh size

Applicability
evaluation

RemarksHigh

Long-term operation is possible by recovering differential pressure
as appropriate with drain operation

Medium

It is determined that it is less applicable than auto strainers in termsof removal precision and generated amount of liquid waste

Auto strainer

50 μm

Liquid cyclone-

Auto strainer

5, 10, 20 μm

Low

Metal sintering filter

2, 5, 10, 20 μm

Low

Rough
removalIntermediate
removalFinaltreatment

Removal precision is high, but it is determined that applicability is
low since increase in differential pressure due to clogging issignificant

Bag filter

3 μm

MediumMF membrane (ceramic filter)*

1.4 μmMedium highTest conducted in this PJ. Result reported in this document.

UF membrane (ceramic filter)*

20, 50, 100 nm

High

It is evaluated that applicability is high. It is necessary to selectappropriate aperture.

*MF membrane: Micro-Filtration, UF membrane: Ultra-Filtration©International Research Institute for Nuclear DecommissioningNo.64

7. Implementation items③-1 Development of non-soluble nuclide removal technology➢

Development issues of particle removal system
Selection of candidate equipment of intermediate removal system•

Filter performance was evaluated as shown in the following table based on the test result of each filter equipment(previous subsidy project).
Table Summary of result of filter performance assessment (result of subsidy project FY2019-2020)

SystemRoughremovalRemovalefficiency

More than μm

More than %Intermittent drain

50

99

2

11

2.5

167

Permanent drain

50

99

2

11

Permanent drain

1300

40

80

-

-

Permanent drain

2000

Intermittent drain

20

96

2

11

502

33500

Permanent drain

20

96

2

11

118

9200

Metal sintering filter

2

99

18300

14650025.1

815

Bag filter

3

99

5

440

-

-

Particle collector

Auto strainer

Liquid cyclone
Auto strainer
Intermediate
removalFinaltreatment

Replacement
Generated
Number of times
Generated amountfrequency
amount ofofof liquid waste(life)
secondary waste drain/backwash
times/year
kg/yeartimes/daym3/year

Filtering
precision

UF membrane

0.1μm

0.1

99

8

314

0.17

5.5

UF membrane

0.05μm

0.05

99

3

152

0.11

48.2

Generated amount of liquid waste and generated amount of
secondary waste are evaluated to be extremely large•••

All equipment has extremely high particle removal rate, and was evaluated to appropriately remove particles in the collection range ofeach system.
In the rough removal system and final treatment system, each filter equipment selected was evaluated to have high applicability toactual equipment as increase in differential pressure due to clogging is gradual.
All filter equipment selected as intermediate removal equipment shows significant increase in differential pressure due to clogging, andgenerated amount of backwash liquid waste and generated amount of secondary waste are larger compared to other systems.
Since filter equipment applicable to intermediate removal system has not been selected, it is necessary to select candidate equipment.
©International Research Institute for Nuclear DecommissioningNo.65

7. Implementation items③-1 Development of non-soluble nuclide removal technology➢

Development issues of particle removal system
Impact assessment of particle form
Here are the particles used for the filter water flow test conducted in the previous subsidy project.
(×1,000)
No pretreatment ofreagents

SiO2 particle

SUS particle
Flaky particles were confirmed
but its abundance ratio is verysmall

WC particle*WC: Tungsten carbide

Particles used in this test are dominated byspherical particles

Figure SEM image of simulated particles used in the filter test of the previous subsidy project•••In the previous subsidy project, tests were conducted using simulated powder dominated by nearly sphericalparticles.
However, if mechanical methods such as AWJ and boring processing are adopted, generated chip powder is
expected to be dominated by flaky and needle-like forms.
If particles take non-spherical form such as flaky and needle-like forms, there is concern of increase in differential
pressure during water flow and impact on differential pressure recovery rate due to backwashing.

It is necessary to evaluate the impact of particle forms on each filter equipment.
In particular, it is necessary to acquire operation data of each filter when passing non-spherical particles through water, whichhave not been verified in the previous subsidy project.
©International Research Institute for Nuclear DecommissioningNo.66

7. Implementation items③-1 Development of non-soluble nuclide removal technology➢

Details of element test (filter water flow test)•

Evaluation of applicability of MF membrane to intermediate removal system
Applicability of 1.4μm MF membrane to actual equipment shall be evaluated as intermediate removal filter which was notselected in the previous subsidy project.

1.4μm MF membrane water flow test
The following items shall be evaluated by performing water flow treatment for simulated liquid of assumed input water
quality of intermediate removal system with 1.4μm membrane.
•
•
•

•

Collection performance for particles in the range of several dozen to several μm which is thecollection range of intermediate removal system
Trend of increase in equipment differential pressure during water flow and time available for water
flow
Trend of recovery of differential pressure due to backwashing when differential pressure increases

Evaluation of dependence of particle form of each filter
Impact on each filter when particle form changes shall be evaluated, which was not evaluated in the previous subsidy
project.

Evaluation test of dependence of particle form
Data, such as filter operation data, differential pressure behavior, particle collection rate and recovery rate ofdifferential pressure due to backwashing, when passing spherical particles and non-spherical particles throughwater, shall be acquired.
*Filters to be tested were selected among filters that were evaluated to have high applicability ineach system.
According to filter operation data acquired from the test, resistance of cake formed by spherical/nonspherical particles shall be evaluated and impact on filters attributable to particle form shall be evaluated.
©International Research Institute for Nuclear DecommissioningNo.67

7. Implementation items③-1 Development of non-soluble nuclide removal technology➢ Test system of MF membrane water flow test

MF membrane (ceramic filter)

Main line:
Backwashwater tank

Flushing pump

Flushing filter

Backwashpump
Particle
injection part

Filter element
Outlet water
receiving tank

Raw water Main pumptankParticlecollector

Y-shapestrainer
Drain water / backwash water receiving tank

Figure Test system configuration diagram of MF membrane water flow test
Filter element➢ Test case of MF membrane water flow test
Table MF membrane water flow test case table
Test
Case

Particle components mixing ratio
SS
Flow rate[wt.%]
concentration
Particle size [μm]Water flow time
[m3/h][ppm]
WC
SUS316
SiO2

1-1

0.1-101-2

0.1-101-3

0.1-10

1-4

1

10

100

33

33

33

100

58

30

12

500

33

33

33

100

33

33

33

Until allowabledifferential
pressure isreached

Housing*7 filter elements were loaded in the
test.
©International Research Institute for Nuclear Decommissioning7. Implementation items

No.68③-1 Development of non-soluble nuclide removal technology➢

MF membrane performance assessment parameter
Test parameters were set as shown in the following table based on assumed water quality condition.
Table MF membrane water flow test parameters

No.

Item

Conditions

Remarks

1

Particle collector

1.4μm MF membrane (ceramic filter)2

SS concentration

100ppm, 500ppm

100ppm, which is the assumed condition for the liquid phase system, is set as thebasic condition, and 500ppm is set as the high load condition

3

Particle size

0.1-10μm, 1μm

0.1-10μm, which is the range of particle size that could flow into the intermediate
removal system, is set as the reference condition, and only 1μm particles are set as
the high load condition

4

Particle form

Spherical particle

Spherical particles are used so that they can be studied in comparison with tests
using spherical particles conducted up to FY2020

5

Particle components(to be simulated)

Tungsten carbide (fuel debris),
SUS316L (reactor internals), silica
sand (concrete component)

Particles with true specific gravity close to their respective simulated components
are selected

6Particle mixing ratio① Equal mixing of three components
on a weight basis② Equal mixing of three components
on a volume basis

・① is set so that it can be studied in comparison with tests conducted up to
FY2020・② is set as the condition with equal number of particles of the three components

7

Flow rate

10m3/h8

Water flow time

Candidate equipment of intermediate removal system is selected

10m3/h, which is the basic system flow rate of the liquid phase system, is set

Water is passed through until backwash set pressure is reached. If long-term water flow is required, acceleration test is
conducted by increasing SS concentration.

Table MF membrane water flow test backwash conditions
No.

Item

Set value

1

Backwash flow rate

10 m3/h2

Backwashing pressure

0.10 MPa3

Backwash time

1 min/count4

Implemented number oftimes

Until recovering to initial differential pressure

Remarks

Conditions are set upon confirming the appropriate
differential pressure recovery trend through tests.

©International Research Institute for Nuclear DecommissioningNo.69

7. Implementation items③-1 Development of non-soluble nuclide removal technology➢

Test result of MF membrane water flow test
Here is the observation on behavior of increase in equipment differential pressure during water flow, using
representative examples.
Equipment differential pressure [kPa]

Increase in differential pressure changes according to particle size
distribution of particles. When conditions are the same for filter hole andparticle size (1-4), water flow time was 2/3 of that of the basic case.

•

•

•

Allowable differential pressure
Table Particle size range of simulated particles
Test case 1-1
Test case 1-4

Differential pressure after backwash

Test caseParticle size [μm]

1-1
1-4

0.1-101

Test case 1-4 has the particle size close to thefilter hole size.
Since filtering performance of MF membraneused in this test is 1.4μm which is the
manufacturer’s specification, substantial filter
hole size is assumed to be a little less than
1μm.

Initial differential pressure recovered due to backwashing.
However, there are 2-3% of residuals with only one
backwashing, so additional backwashing is required to reduce
residuals.
Elapsed time[min.]In test case 1-1 which is the basic condition, the time available for water flow until backwashing was evaluated to be 62min. It isevaluated to be highly applicable as increase in differential pressure is slow compared to candidate intermediate removal equipment
which was tested in the previous subsidy project. The number of elements was set to seven elements in this test, but since waterflow time until backwashing can be adjusted by adjusting the number of elements, it is desirable to design the number of elements
to allow operationally appropriate backwash frequency and equipment scale.
Test case 1-1 has broad particle size distribution, and test case 1-4 has the sharp peak of its particle size distribution at 1μm.
Differences in differential pressure increase considered to be attributable to differences in particle size distribution were confirmed,
and the time available for water flow in the condition where complete blockage is most likely to occur (test case 1-4) is considered to
be 2/3 of that of cases with broad distribution.
If backwashing operation is performed after differential pressure increases due to clogging, initial differential pressure largely
recovers, and it has been confirmed that it can be reduced to near its initial value. However, since particles that cannot bedischarged slightly remain, two or three additional backwashing is required in order to recover differential pressure to near its initialvalue. Necessity of complete recovery to initial differential pressure is low in the operation of actual equipment, and it is consideredthat operationally sufficient recovery rate can be secured with one backwashing.
©International Research Institute for Nuclear Decommissioning

No.707. Implementation items③-1 Development of non-soluble nuclide removal technology➢

Comparison of performance of candidate intermediate removal system filters
Candidate intermediate removal system equipment was compared and evaluated based on the element test result.
(including results of subsidy projects prior to FY2020)
Table Comparison of performance of candidate intermediate removal system equipment

Particle collector

Requirednumber ofelements

Piece•

Replacement
frequency

More than More thantimes/year
μm%

Generated
Number of
Generatedamount oftimes ofamount ofsecondarydrain/backwash liquid waste
waste

kg/yeartimes/daym3/yearApplicability
assessment
result

20

96

2

11

502

33500

Low

1

20

96

2

11

118

9200

Low-medium

Metal sintering filter

11

2

99

18300

146500

25.1

815

Low

Bag filter

122

3

99

5

440

-

-

MediumMF membrane

70

1.4

99

1

77

0.05

30

High

Result of evaluationaccording to the test resultof this PJ was added•

Removalefficiency

1

Auto strainer•

Intermittent
drain
Permanent
drain

Filtering
precision*Evaluation conditions: treatment flow rate: 65m3/day, SS concentration: 100ppm, particle collection rate: 33% in each system*Replacement frequency of MF membrane was conservatively assumed and set with the service life of sealing materials such as gasket

It was evaluated that MF membrane has the best performance from the viewpoint of generated amount of secondary waste andgenerated amount of liquid waste (concentration factor).
Compared to other candidate equipment, MF membrane increases differential pressure due to particle collection more gradually andcan collect more particles before backwash. It is considered to be a major factor that its effective filtering area is extremely large
compared to other intermediate removal filters.
Replacement frequency of ceramic filters such as MF membrane was set based on replacement cycle of sealing materials. It is
assumed that deterioration of sealing materials such as gasket is faster than increase in initial differential pressure and deteriorationof filter body due to repeating water flow and backwashing.
Based on this evaluation result, 1.4μm MF membrane was selected as candidate intermediate removal system equipment.
©International Research Institute for Nuclear Decommissioning

No.71

7. Implementation items③-1 Development of non-soluble nuclide removal technology➢

Parameters of evaluation test of dependence of particle form
Parameters for this test were set as shown in the following table.
The test system used similar equipment as that of MF membrane water flow test.
Table Test parameters of evaluation test of dependence of particle form

No.

Item

Conditions

Remarks

1

Particle collector

50μm auto strainer
1.4μm MF membrane
0.05μm UF membrane

Particle collector evaluated to have high applicability in each system is selected*Increase in differential pressure was not observed with auto strainer

2

SS concentration

100ppm, 500ppm

100ppm, which is the assumed condition for the liquid phase system, is set as thebasic condition, and 500ppm is set as the high load condition

3

Particle size

0.1-10μm, 1μm

0.1-10μm, which is the range of particle size that could flow into the intermediate
removal system, is set as the reference condition, and only 1μm particles are set
as the high load condition

4

Particle form

Spherical particle, non-sphericalparticle

Spherical particles and non-spherical particles are selected
Impact of particle form on filter performance is evaluated based on the test result ofboth forms.

5

Particle size
distribution

Double-peak distribution of 1μm, 10μm

6Particle components(to be simulated)

Silica sand (concrete component)

Silica sand is selected as the representative component.
With other component particles, it is difficult to produce non-spherical fine particles.

7

Flow rate

1m3/h, 10m3/h

10m3/h, which is the system flow rate of the liquid phase system, is set→UF membrane is tested with 1m 3/h and evaluated with proportional calculation.

8

Water flow time

Set with reference to cutting test of simulated fuel debris conducted in thecharacterization PJ.
*See next page

Water is passed through until backwash set pressure is reached.

This test evaluates the impact of particle form on filter performance from the viewpoint of differential pressure behavior andparticle collection rate.
©International Research Institute for Nuclear Decommissioning

No.72

7. Implementation items③-1 Development of non-soluble nuclide removal technology➢

Simulated particles used in this test
Here are the SEM image and particle size distribution of each particle used in this filter test

Frequency (%)

Accumulation of passed
through particles (%)

Spherical particle

Particle size (μm)

The mixing ratio of particles was
adjusted to achieve almost thesame particle size distribution

Accumulation of passed
through particles (%)

Non-spherical particle

Frequency (%)

Nearly sphericalparticles that have nocorners are dominant.

Block-like and plate-like particles
that have corners are dominant.
There are hardly any sphericalparticles.

Particle size (μm)

Adjustment was made so that there is no difference in particle size distribution of both types. (double-peak distribution)
Since other parameters are also the same, it is evaluated that the difference observed in the test result is attributable to particle form.
©International Research Institute for Nuclear Decommissioning7. Implementation items

No.73③-1 Development of non-soluble nuclide removal technology➢

Difference in behavior of differential pressure increase due to forms ofparticles passed through water

Non-spherical particles have slower
increase in equipment differential
pressure, but its difference is small.In
addition, the difference tends togradually increase.

3-1 (spherical particle)
3-2 (non-spherical particle)② Cake filtering stage① Cake formation stage

Equipment differential pressure [kPa]

Equipment differential pressure [kPa]

Behavior of increase in equipment differential pressure when passing spherical/non-sphericalparticles through water was compared and evaluated.
2-1 (spherical particle)
2-2 (non-spherical particle)② Cake filtering stage① Cake formation stage

Elapsed time [min.]Elapsed time [min.](a) 1.4μm MF membrane(b) 0.05μm UF membrane

Figure Differential pressure behavior in filter test result
Comparing overall behavior of equipment differential pressure, spherical particles show larger differential pressure than non-spherical particles. In
particular, difference starts to occur between the two types of particles approx. 30min. after starting water flow for the MF membrane and approx.
500min. after starting water flow for the UF membrane. The cause was observed as follows.

① Cake formation stage
Stage where particles are collected with filter filtering and cake is formed. Filter resistance is dominant, and impact of cakeresistance is minor. Thus, difference in behavior of increase in differential pressure is unlikely to occur due to particle form.

② Cake filtering stage
Stage where cake is formed and particles passed through water are cake-filtered. Impact of cake resistance appears in addition tofilter resistance. Cake resistance differs depending on particle filling density of cake. Spherical particles have larger cake resistancesince its filling density is large. Therefore, differential pressure is larger compared to the case when passing non-spherical particles
through water. The difference becomes larger as cake grows thicker with longer water flow time.
©International Research Institute for Nuclear Decommissioning7. Implementation items

No.74③-1 Development of non-soluble nuclide removal technology➢

Difference in backwash performance due to forms of particles passed through water

3-1 (spherical particle)
3-2 (non-spherical particle)

Elapsed time [min.](a) 1.4μm MF membrane

Equipment differential pressure [kPa]

Equipment differential pressure [kPa]

Differential pressure recovery behavior due to backwash operation when differential pressure increases by passing
spherical/non-spherical particles through water was compared and evaluated.
2-1 (spherical particle)

Differential pressure can berecovered close to initialdifferential pressure due tobackwashing for both sphericaland non-spherical particles

2-2 (non-spherical particle)

No particular difference indominance was confirmed

Elapsed time [min.](b) 0.05μm UF membrane

Figure Differential pressure behavior in filter test result

Observation on backwash performance••

It is evaluated that backwash performance is extremely high regardless of particle form, as differential pressure was recovered
close to initial differential pressure due to backwash for cases of passing both spherical and non-spherical particles throughwater.
Comparative evaluation shows a trend of lower differential pressure after backwash when using non-spherical particles, but thedifference between the two types of particles is minute and it is unclear whether the difference is attributable to particle form.

In general, backwash performance is known to improve when passing non-spherical particles that are close to fibrous form through water,
rather than spherical particles, as particles can be easily discharged by backwashing. The effect is the same when particles of several μm
order is dominant as in this test, but contribution is expected to decrease as the particle size decreases.

Regarding differential pressure recovery behavior during backwashing, difference attributable to particle form is minor,
and almost all particles are expected to be discharged by backwashing.
©International Research Institute for Nuclear Decommissioning

No.75

7. Implementation items③-1 Development of non-soluble nuclide removal technology
Evaluation using Lewis’ Equation of Filtration
Dependence of filtering performance on particle form was evaluated with average cake resistivity.

➢

Average cake resistivity 𝛼𝑎𝑣[𝑚/𝑘𝑔]
Physical quantity that indicates the resistance when
fluid passes through cake that was formed on thefilter surface due to filtering. In general, it is used todetermine filtering performance of the filter to be used
for liquid subject to filtering.

Table Evaluation result of average cake resistivity / average filtering resistivity(comparative evaluation of representative cases)
Test case

2-1

2-2

3-1

3-2

Filter type0.05μm UF
membrane

0.05μm UF
membrane

1.4μm MF
membrane

1.4μm MF
membraneParticle form

Spherical

Non-spherical

Spherical

Non-spherical

Reference judgment value [m/kg]

Particle collection rate [%]

99.9

99.9

69.7

71.5

𝛼𝑎𝑣 ≤ 1011 : Easy
1012 ≤ 𝛼𝑎𝑣 ≤ 1013 : Medium
1013 ≤ 𝛼𝑎𝑣 : Difficult

Average cake resistivity[m/kg]
Filter material resistance[1/m]

7.9.E+126.9.E+12

1.4.E+12

1.2.E+12-3.0.E+11-2.2.E+11

6.4.E+10

5.1.E+10*Judgment value is criteria based on general experience

•
•
•
•

*Phenomenologically, it cannot be a negative value, and so it is actually considered to be close to 0.
≒ When liquid not containing particles passes through water, equipment differential pressure hardly
increases.

In this test, average cake resistivity was evaluated as 1.0E+12 order with UF membrane and 1.0E+11 order with MF
membrane, regardless of spherical or non-spherical particles.
As comparative evaluation indicates that average cake resistivity tends to be larger with spherical particles but the differenceis small, it is considered that cakes with similar properties were formed.
Since it is assumed that particles have agglomerated to some extent and formed agglomerates of several μm or greater, it ispossible that dependence on particle form was small.
When treating liquid containing large amounts of particles with larger particle size (several mm order), dependence onparticle form is considered to be large. However, since cake resistance becomes smaller due to increase in porosity of cakes
as particle size becomes larger, and most particles larger than several mm are assumed to settle in PCV, it is consideredless necessary to evaluate the impact of particle form on filter performance when filtering large particles.
Based on this test result, it was judged that it is less necessary to design the particle removal system considering variations
in particle form.
©International Research Institute for Nuclear Decommissioning

No.76

7. Implementation items③-1 Development of non-soluble nuclide removal technology➢

Evaluation of impact of each parameter on particle removal system
Table Evaluation of impact of each parameter on particle removal system
Item

Degree of impact
on particle
removal system

1

Test results

Details of study

Particle components

Auto strainer: highCeramic filter: low

Highly feasible

Since auto strainer is a mechanism that separates particles also with centrifugal separation in addition
to filtration, separation performance is high for particles with higher specific gravity such as fueldebris. However, centrifugal separation hardly has any removal effect for particles with low specificgravity such as concrete particles.

2

SS concentration

High○Highly feasible

Since there is a significant impact on the speed of increase in differential pressure of each filter, itlargely affects the concentration factor.

3

Particle size range

High○Highly feasible

4

Particle size
distribution

High○Highly feasible

Particle size is considered to rate-control the percentage of the amount of particles collected in eachsystem. Therefore, concentration factor of the entire system greatly changes depending on theamount of particles in the particle collection range of each system.

5

Particle form

Low

○Highly feasible

Impact of particle form was evaluated to be small according to the filter test using spherical/nonspherical particles.

6

Particle strength

Low

×Low feasibility

Since particles generated in the fuel debris retrieval work are assumed to be basically solid materials
with a certain degree of hardness, it was evaluated that the impact is small.

7

Precipitate

Medium

×Low feasibility

Generation of gel-like precipitates containing much moisture causes significant clogging. Impact issmall if precipitates are highly crystalline.

8

Ionic strength

Low△Some may
not be feasible

If ion strength is high, apparent grain size becomes larger due to saltation effect. With water quality
considered in actual equipment, it is assumed that only almost negligible coagulation occurs.

9

Systemconfiguration

High○Highly feasible

Concentration factor of the entire system can be improved by changing the type, aperture and filtering
area of the filter to be installed, according to particle size distribution. Currently, a system that canhandle a wide range of conditions is being studied.

10

Removal range(lower limit)

Medium○Highly feasible

Impact is small if the required value of filtering precision is up to around 0.1μm. Impact such asincrease in generated amount of liquid waste and increase in equipment scale becomes apparent
from around 0.01μm.

11Concentration
reduction request

Medium△Some may
not be feasible

Impact is large since larger removal rate is required with larger reduction requirement of activity
concentration.

No.

Performance of particle removal system was evaluated using physical quantity with large impact (in red) as parameters.
©International Research Institute for Nuclear Decommissioning

No.77

7. Implementation items③-1 Development of non-soluble nuclide removal technology➢

Evaluation of generated amount of liquid waste of particle removal system
The amount of filter liquid waste generated during the fuel debris retrieval period was estimated from the filter test result.

Equation for evaluation of generated amount of filter liquid waste
3
𝑉𝑖𝑛 × 𝑣𝑛 × 𝑎𝑛 × 𝑐𝑖𝑛
𝑉𝑏𝑤 𝑐𝑖𝑛 , 𝑎𝑛 = ෍
𝑚𝑛
𝑛=1
𝑉𝑏𝑤 ：

•
•

➢

Table Set conditions of particle collector and number of elements
SystemParticle collector

Loaded number of
elementsRoughremovalIntermediate
removalFinaltreatment

50μm auto strainer

1 element

1.4μm MF membrane

70 elements*1

0.05μm UF membrane

70 elements*1

Annual amount of generated liquid waste [m3/year]

𝑉𝑖𝑛 :Annual treatment liquid amount [m3/year]𝑐𝑖𝑛 ：SS concentration of input water [kg/m3]

𝑣𝑛 ：Amount of backwash water per treatment of each filter
equipment [m3/count]𝑎𝑛 ：Particle collection rate of each filter equipment [%](particle size distribution)

𝑚𝑛 :Particle collection amount up to backwashing of eachfilter equipment [kg/count]*1 Number of elements for which backwash frequency is lessthan once a day and the filter line velocity is sufficiently small

Above parameters are set using studied values of the liquid system and element test result as boundary conditions
SS concentration 𝑐𝑖𝑛 and particle collection rate 𝑎𝑛 (particle size distribution) are set as variables and several patterns of water quality
conditions of input water are assumed

Basic model of this evaluation (particle removal system)
Rough removal system

Intermediate removal system

50μm
Auto strainer

1.4μm
MF membrane(ceramic filter)

Final treatment system
0.05μmUF membrane(ceramic filter)𝑛 =3

Accumulated
water in PCVSoluble nuclide
removal equipment

𝑣𝑛 ：Amount of backwash
water

𝑛 =2

𝑎𝑛 ：

Particle collection rate𝑚𝑛 :Particle collection
amount*This evaluation test does not take the effect of centrifugal separation of the auto strainer into
account, as only particle size and hole size were given as conditions to determine whether removal
is possible.

𝑛=1©International Research Institute for Nuclear Decommissioning

No.78

7. Implementation items③-1 Development of non-soluble nuclide removal technology➢

If rough removal system /intermediate removal system is notinstalled, it is assumed that particles
are collected with the subsequent
final treatment system

Particle removal system configuration
Particle removal system configuration was evaluated in the following 3 patterns.

1-system configuration

0.05μm UF membrane

2-system configuration

50μm
Auto strainer

3-system configuration

50μm
Auto strainer➢

0.05μm UF membrane

1.4μm MF membrane

0.05μm UF membrane

Evaluation result of generated amount of liquid waste
The generated amount of liquid waste when liquid waste with the assumed water
quality is treated was evaluated for each particle removal system configuration.

: 1.0 m3/year: 100 m3/year: 1000 m3/year

Table Evaluation result of generated amount of liquid waste of particle removal system [m3/year]
Particle
size range [μm]
粒径範囲[μm]
Roughremoval

粗取り50～100
33%
90%
80%
5%
10%
5%
10%

3-system
configuration
3系統構成

Final
treatment

SS concentration[ppm]
SS濃度[ppm]
中取り 最終処理
1～50
0.1～1
1
10 50 100 200
33%33%
3 27 135 271 542
5%5%
5 46 232 464 927
10%10%
4 43 215 430 859
90%
5%
1 10 50 99 198
80%
10%
1 13 65 130 259
5%90%
2 25 125 250 499
10%80%
3 25 127 253 507

Intermediate
removal

1
3
5
4
3
3
3
3

2-system
configuration
2系統構成

1-system
configuration
1系統構成

SS SS濃度[ppm]
concentration [ppm]

SS concentration[ppm]
SS濃度[ppm]

10
33
47
45
26
27
26
27

50
165
236
224
129
136
129
136

100
330
472
447
258
271
258
271

200
660
945
894
517
542
517
542

1
2
2
2
2
2
2
2

10
25
25
25
25
25
25
25

50
123
123
123
123
123
123
123

100
246
246
246
246
246
246
246

200
492
492
492
492
492
492
492

*Effect of centrifugal separation of auto strainer is not considered in this evaluation. Qualitatively, centrifugal separation effect is larger forlarger particles and particles with higher specific gravity.

Appropriate particle removal system configuration differs according to properties of input water.
Appropriate system configuration was studied for each assumed condition.
©International Research Institute for Nuclear Decommissioning

No.79

7. Implementation items③-1 Development of non-soluble nuclide removal technology➢

Basic system configuration of particle removal system
Rough removal system
50 μm
Auto strainer

Intermediate removal
system
1.4μm MF membrane(ceramic filter)

Accumulated
water in PCV

0.05μm UF membrane(ceramic filter)*1 Particles smaller than 50μm can be removed withcentrifugal separation when they have high specificgravity such as fuel debris

Soluble nuclide
removal equipment

Intermediate
removal system

Drain water / backwash water
100μm-50μm*1

Particle removal range

Particle size

Final treatment system

Large

100μm

50μm-1μm

10μm

Backwash water

1μm-0.1μm

1μm

Backwash water

Sludge collection
equipment(Adsorption tower, RO membrane)

0.1μm

Small
Colloid

Ion

Table Study of particle removal system configuration for each water quality
No.

Water quality conditions

Appropriate system
configuration proposal

Study of particle removal system configuration

1

Broad particle size
distribution

3-system configuration

3-system configuration system shall be applied so that particles can be collected with high efficiency in eachparticle size range. (basic configuration)

2

Many particles smaller than
1μm

1-system configuration

Particles smaller than 1μm are collected almost only in the final treatment system. Since only a few particles
are collected outside of the final treatment system, a system with only one system is highly applicable toreduce the equipment scale.

3

Many particles of several µm

3-system configuration

Configuration that collects particles of several µm with the intermediate removal system has the smallest
load. Thus, 3-system configuration is highly applicable.

4

Many particles with highspecific gravity

2 or 3-system configuration

Since particles with high specific gravity, such as fuel debris, are expected to be collected with the roughremoval system in high efficiency, a system with 2-system or 3-system configuration is desirable.

5

Low SS concentration(several ppm)

1-system configuration

If SS concentration is sufficiently small below 10ppm, there is no significant difference among all systems.
Thus, 1-system configuration is desirable which has smallest equipment scale.
©International Research Institute for Nuclear Decommissioning7. Implementation items

No.80③-1 Development of non-soluble nuclide removal technology➢

Summary of performance assessment result of each particle collector
Assumed specifications of candidate particle collectors in each system were evaluated based on test results acquired to date.
Table Summary of result of filter performance assessment (3-system water flow configuration)

SystemParticle collector

Auto strainer
RoughremovalIntermittent
drain
Permanent
drain

Liquid cyclone
Intermittent
drain
Permanent
drain
Metal sintering filter

Auto strainer
Intermediate
removalBag filter
MF membrane
UF
0.1μm
membraneFinaltreatment
UF
0.05μmmembrane➢Requirednumber ofunits

Requirednumber ofelements

Rated flow Filtering Removal
rate
precision efficiency

Approximate
dimension(per unit)

Required area
dimension(per unit)

Replacement
frequency(life)

Generatedamount ofsecondary
waste

Number of timesofdrain/backwash

Generatedamount ofliquid waste

More than More thanμm%

φ＿m×＿mH＿mL×＿mWtimes/year

kg/yeartimes/daym3/year

2.5

167

1 system
per unit

Piece

m3/h

1

1

10-30

50

99

φ0.5m×1.0mH

1.4mL×1.4mW

2

11

1

1

10-30

50

99

φ0.5m×1.0mH

1.4mL×1.4mW

2

11

1

1

10

40

80

φ0.4m×1.1mH

1.6mL×1.6mW

-

-

1

1

10-30

20

96

φ0.5m×1.0mH

1.4mL×1.4mW

2

11

502

33500

1

1

10-30

20

96

φ0.5m×1.0mH

1.4mL×1.4mW

2

11

118

9200

3

11

6.5

2

99

φ0.21m×1.2mH

3.0mL×1.6mW

18300

146500

25.1

815

4

122

10

3

99

φ0.7m×1.6mH

2.0mL×2.0mW

5

440

-

-

4

70

10

1.4

99

φ0.3m×1.4mH

1.4mL×1.4mW

1

77

0.05

30

2

33

6.5

0.1

99

φ0.21m×1.3mH

3.0mL×1.6mW

8

314

0.17

5.5

3

70

10

0.05

99

φ0.3m×1.4mH

1.4mL×1.4mW

1

77

0.11

50

Future development issues

PermanentdrainPermanent
drain

1300
2000

*Evaluation conditions: treatment flow rate: 52m3/day, SS concentration:
100ppm, particle collection rate: 33% in each system① Study of filter replacement method and evaluation of replacement cycle (life)
It is necessary to study remote replacement of filter elements and alternative replacement method when remote replacement is difficult. Additionally, filter
life data for setting the replacement cycle needs to be evaluated in the future since there is shortage of such data. Assuming ceramic filters are ratecontrolled by the replacement cycle based on service life of gasket materials (assuming 1-2 years), rather than replacement cycle attributable todeterioration of the filter body, it shall be considered to assume ceramic filters based on operation results of similar gasket materials of Fukushima Daiichi
filter equipment.
② Studyof optimization of loaded number of elements and water flow velocity
Since filter and cake resistance is reduced as line velocity becomes smaller with more loaded number of elements or slower flow velocity, the amount ofparticles that can be collected up to backwash differential pressure becomes larger. However, because of the tradeoff between increase in equipment scale andprolonged treatment time, it is necessary to optimize treatment conditions such as number of elements and flow velocity taking various factors into account.
©International Research Institute for Nuclear Decommissioning

No.81

7. Implementation items③-1 Development of non-soluble nuclide removal technology[Results to date]✓ 1.4μm MF membrane was selected as candidate equipment applicable to the intermediate
removal system.
✓ Dependence of particle form on candidate equipment of each system was verified. As a result, it
was confirmed that there is little need to incorporate particle form into the conceptual design ofthe particle removal system since impact of particle form is small.
✓ 3 patterns of particle removal system configuration, 1-system, 2-system and 3-system
configuration, were studied with 3-system configuration as the basis.
✓ Estimated amount of liquid waste generated from the particle removal system was evaluated, andsystem configuration considered to be dominant for each input water quality was organized.

[Future issues to be studied]✓ Fuel debris retrieval method shall be selected, and a more reasonable system shall be selectedusing the studied results when it becomes possible to estimate the properties (generated
amount, components, particle size distribution, etc.) of generated chip powder✓ Application of filter equipment is also studied for the soluble nuclide removal system and RO
concentrated water treatment system, and it is assumed to perform concentration treatment for
powder adsorbents with the cross flow method. It is necessary to study the applicability of crossflow filter when assuming such treatment.

[Summary]
Candidate filter equipment applicable to each system of particle removal system was selected, andperformance assessment was conducted. Appropriate system was studied according to properties of
input water quality with 3-system configuration as the basis.
©International Research Institute for Nuclear Decommissioning

No.82

7. Implementation items③-2 Development of sludge dehydration treatment technology➢

Evaluation of volume reduction effect by dehydration[Implementation items] Volume reduction of sedimentation sludge by dehydration was evaluated.
Behavior of volume reduction when removing moisture from assumed moisture content of sedimentation sludge ofapprox. 90vol.% was evaluated.

Areas dominated by reduction of upper and pore water
High
rate
of volume
Areas
dominated
byreduction
reductio of sludge

Volume after dehydration withinitial value set as 100 [-]High rate of volume reduction of sludge

When moisture content is below
60vol.%, dehydration is less likely tocontribute to the reduction of sludge
volume since the space occupied bymoisture becomes a void.

Areas where the decrease in pore water becomes voids
Small volume reduction rate of sludge

Initial moisture
content

Appearance of sedimentation sludgeUC

UC (unit can)Dehydration
60%

60%

Solid
components

Solidcomponents

Moisture

Moisture

Moisture content [vol.%]

Figure Relation between moisture content and sludge volume(a) Moisture content above 60%(b) Moisture content below 60%

Figure Image of behavior of reduction of sludge volume

If initial moisture content is above 90vol.%, volume can be reduced to nearly 20vol.%, by dehydration treatment up to 60vol.% moisture
content.
Since volume decrease hits the ceiling in the range below 60vol.% moisture content, effect of volume reduction due to dehydration is small.

Dehydration treatment technology shall be studied with target moisture content after dehydration treatment tentatively
set* to 60vol.%.
*Sludge content 40vol.%©International Research Institute for Nuclear Decommissioning

No.83

7. Implementation items③-2 Development of sludge dehydration treatment technology➢ Selection of candidate technology

Conceptual flow of the sludge dehydration treatment system was assumed as shown in the following figure.
Solid-liquid separation treatment
Water to betreated(After particle
removal)Nuclide removal
treatment(Adsorption
treatment)

Concentration
treatment of solid
components
Separatedwater

Sludgedehydration
treatment

SludgeSeparated water

Figure Conceptual flow of solid-liquid separation system of powder adsorbent

Solidification
treatment
Secondary
nuclide removal
treatment

Sludge after
dehydration
Treatment
water

Table Study of treatment method that can be applied with each solid-liquid separation technology
No.

Candidatetechnology

1

MF, UF membrane(DE) filtering

Candidate applicable
treatment

Approach
It is assumed to be applicable as technology that performs concentration treatment for liquid with SS concentration
below 100ppm to several thousand ppm.

Cross flow (CF)filtering
Coagulationsedimentation
treatment

Concentration
treatment of solid
components

It is assumed to be applicable as technology that concentrates liquid with SS concentration below 100ppm, to tens of
thousands ppm.
It is assumed to be applicable as technology that concentrates liquid with SS concentration of several hundred ppm totens of thousands ppm, to tens of thousands ppm, and makes them into sedimentation sludge. However, residual
amount of particles is relatively large, and particle collection rate is low.

4

RO concentration

It is assumed to be applicable for the removal of particles and ionic nuclides that are difficult to be filtered and removed.

5

Evaporator

Secondary nuclide
removal treatment
such as filter
permeable water

6

Cartridge filter

7

Depressurization /heatup dehydration

2
3

Sludge dehydration
treatment

It is assumed to be applicable for the removal of particles and ionic nuclides that are difficult to be filtered and removed.
It is assumed that adsorption treatment liquid / concentrate / sedimentation sludge can be filtered with cartridge filter,
dehydrated to solid concentration of around 40%, and converted into cake.
It is assumed to be applicable as technology that dehydrates concentrate / sedimentation sludge to solid concentrationof 40% or higher.

Applicability of each candidate technology to the sludge dehydration treatment system was studied.
©International Research Institute for Nuclear Decommissioning7. Implementation items

No.84③-2 Development of sludge dehydration treatment technology➢ Study on element test of sludge dehydration technology

Applicability of candidate technology selected based on literature review was verified through element tests.
No.

Requiredtreatment

Candidatetechnology

It is necessary to evaluate concentration limit according to slurry property. It is
MF, UF membraneassumed that it can be studied with reference to the filter test result of the(DE) filtering
previous year.

1

2

Concentration
treatment of solid
components

3
4
5Secondary nuclide
removal treatment
such as filter
permeable water

6

8

Element
test isrequired×Low
possibility

×Not
necessary

Cross flow (CF)filtering

It is necessary to evaluate concentration limit and throughput of treatment amountaccording to slurry property. It is assumed that it can be studied with reference tocases at overseas nuclear sites, but it may be difficult to evaluate throughput.

×Low
possibility△Necessary
with certain
conditions

Coagulationsedimentation
treatment

Implemented in item ② of this subsidy project×Low
possibility○ (item ②)Necessary

RO concentration

Not subject to element test as it was selected as nuclide removal technology

×Low
possibility―

Evaporator

Not subject to element test as it was selected as nuclide removal technology

×Low
possibility―

It is necessary to select appropriate filter elements (aperture, filtering area) andfiltering conditions, according to sludge property (particle size distribution,
moisture content, etc.).

〇Possible〇Necessary〇Possible

×Not
necessary△Possible
with certain
conditions―

Cartridge filter
Sludge dehydration
treatment

7

Items to be confirmed in element test

Can bedehydrated
to 60% oftarget value

It is assumed that dehydration treatment is possible up to the target value,
Depressurization /without largely depending on sludge properties (particle size distribution,
heatup dehydration
particle properties).
Solidification treatment is considered to be effective in terms of stabilization ofsludge waste, but final disposal concept of Fukushima Daiichi waste is(Solidification treatment)
undetermined. Since it is desirable to conduct solidification treatment according toburied body requirements, it is not studied in this development project, and longterm storage shall be assumed after dehydration treatment.

Operation data was acquired by conducting element test of cartridge filter.
Applicability of other candidate technologies was studied in desktop study.
©International Research Institute for Nuclear Decommissioning

No.85

7. Implementation items③-2 Development of sludge dehydration treatment technology➢ Cartridge filter test
Here are the test procedures of the cartridge filter test conducted.

Test condition (test case 1-1)
Test solution: 1000 times diluted seawater, pH: 7.1, adsorbent: titanium silicate, 50ppm, coagulant: not added, 0ppm,
Used amount of liquid: 13500mL, filtering pressure: 0.2MPa (pressurized filtering), filtering membrane: 1.2μm
membrane filter ×φ47mm
Pressurization

Main test procedures

Measurement items of cake

① Supply slurry of prescribed
properties to the slurry supply tank

Weight, moisture content, cake thickness② Pressurize with the compressor③ Cake sediments on filtering
membrane and permeable liquidleaches out over time(a) Filtering residue (cake)

Filtering membraneelement

Measurement items of filtrate④ Test ends when drops of liquid
are no longer confirmed

Weight (volume), residual particle concentration,
turbidity⑤ Conduct measurements andanalyses

Other measurement items
Figure Cartridge filter test equipment(during pressure filtering test)(b) Permeate (filtrate)

Filtering velocity (treatment time),
filtering performance, other visual checks

This test shall evaluate whether filtering velocity satisfies the target value, acquire data on properties of filtering cake,and evaluate applicability of cartridge filter as sludge dehydration technology.
©International Research Institute for Nuclear Decommissioning7. Implementation items

No.86③-2 Development of sludge dehydration treatment technology➢

Setting of supplied slurry/sludge properties
Table Organization of slurry/sludge that may be treated with cartridge filter equipment

No.
1
2
3

AssumedConcentration ofslurry/sludge
solid components
Adsorption treatment
Approx. 100ppm
water
CF concentrated
Approx. 10000ppm
waterParticlecomponents

Components

Remarks―

Liquid waste generated after adsorption treatment due topowder adsorbent
Concentrated water from concentration treatment of No.1
liquid waste with CF filtering
Sedimentation sludge generated after adding coagulant toNo.2 CF concentrated water and performing coagulationsedimentation treatment―

Coagulation<100000ppm(10%) Powder adsorbent
sedimentation sludge
Coagulant
component

Coagulation4
sedimentation sludge<100000ppm(10%)(addition)(no preliminary
concentration)➢

Sedimentation sludge generated after adding coagulant toNo.1 adsorption treatment water and performing
coagulation-sedimentation treatment

RO concentrated water treatment process proposal

In addition to RO concentrated water sedimentation sludge (No.3) that was initially assumed to be treated, tests were also
conducted for slurry/sludge assumed to be highly likely to be treated with the sludge dehydration treatment facility.
No.1 adsorption treatment water
Adsorption
treatment

RO concentrated
water

Adsorption
treatment

No.3 coagulation-sedimentation sludge
RO concentrated
water

Adsorption
treatment

CF concentration
treatment
CF permeable water

CF concentration
treatment

No.4 coagulation-sedimentation sludge (no preliminary concentration)
RO concentrated
water

Sludge after
dehydration

Cartridge filter
treatment

Sludge after
dehydration

CF permeable water

No.2 CF concentrated water
RO concentrated
water

Cartridge filter
treatment

Supernatant water

Coagulation-sedimentation
treatment

Cartridge filter
treatment

Sludge after
dehydration

Supernatant water

Adsorption
Coagulation-sedimentation
treatment
treatment
Figure Setting of water quality conditions of input water to cartridge filter equipment

Cartridge filter
treatment

Sludge after
dehydration©International Research Institute for Nuclear Decommissioning7. Implementation items

No.87③-2 Development of sludge dehydration treatment technology➢

Test parameters (setting of conditions of slurry/sludge)
Parameters concerning preparation of slurry/sludge are as follows.
Based on the RO concentrated water treatment test result, titanium silicate and impregnated active carbon were selected as the powder
adsorbent and aluminum sulfate was selected as the coagulant.
Table Parameters concerning slurry/sludge
Powder adsorbent*1

Coagulant

No.

Test solution conditions
Type

Concentration [ppm]

1-1

50

1-2

10000

1-3

Titanium silicate

Approx. 100000

1-5

200

1-6

10000
Impregnated active
carbon1-8
1-9*3
1-10*3

Titanium silicate
Impregnated active
carbonAluminum
sulfateApprox. 100000

1-4

1-7

Type

―

No.1 adsorption treatment water―

No.2 CF concentrated water / DE
filtering backwash water
No.3 coagulation-sedimentation
sludge5000
High base
PAC*2

Aluminum
sulfateApprox. 100000

2000

1000 times diluted
seawater

No.3 coagulation-sedimentation
sludge―No.1 adsorption treatment water―

No.2 CF concentrated water / DE
filtering backwash water
No.3 coagulation-sedimentation
sludge1000

Approx. 100000

High base
PAC*2

1000

50

High base
PAC

500

High base
PAC

500

200

(to be simulated)

Concentration [ppm]No.3 coagulation-sedimentation
sludge1000 times diluted
seawater

No.4 coagulation-sedimentation
sludge(no preliminary concentration)*1 High α nuclide removal performance was also confirmed with titanic acid, but it is not used in this test since the collectable sample amount is small.Test result oftitanium silicate which has similar particle size and specific gravity is used for the evaluation.
*2 Only basic conditions are applied also for high base PAC*3 Additional test: Simulation of treated sludge without concentration treatment and with coagulation-sedimentation treatment→It was confirmed that sludge becomes difficult to be filtered due to precipitation of gelatinous materials (explained later).
©International Research Institute for Nuclear Decommissioning

No.88

7. Implementation items
1-1③-2 Development of sludge dehydration treatment technology

1-2

1-3

➢ Applicability assessment of cartridge filter⚫ Observation of filtering property of each slurry/sludge(average filtering velocity)

1-4
1-9
1-5
1-6

Here is the evaluation result of average filtering velocity obtained in each
test.

*Filtering vellocity that can treat 14m3/day of RO concentrated water

Overall trend of filtering velocity

・Adsorption treatment water ≧ CF concentrated water ＞ sedimentation sludge
・Active carbon ＞ titanium silicate

•
•
•

•

1-8
1-10

Coagulant
―

―

―

―

Aluminum sulfate

5000ppm

High base PAC

2000ppm

High base PAC

500ppm――――Aluminum sulfate

1000ppm

High base PAC

1000ppm

High base PAC

500ppmAverage filtering velocity[ml/min]

Adsorption treatment water: 𝑉 = 28[ml/min]
CF concentrated water: 𝑉= 0.56[𝑚𝑙/𝑚𝑖𝑛]
Sedimentation sludge: 𝑉 = 0.11[ml/min]

Average filtering velocity [ml/min]

Filtering velocity decision threshold:
Average filtering velocity 𝑉 [ml/min]

Filtering velocity that satisfies the decisionthreshold is confirmed

1-7

Simulated material
Adsorbent
Adsorption
50ppm
treatment water
CF concentrated
10000ppm
water
Titanium
Sedimentation
Approx.
silicatesludge100000ppm
Sedimentation
Approx.
sludge100000ppm
Sedimentation
50ppm
sludgeAdsorption
200ppmtreatment water
CF concentrated
10000ppm
water
ImpregnatedSedimentation
Approx.
active carbon
sludge100000ppm
Sedimentation
Approx.
sludge100000ppm
Sedimentation
200ppmsludge(b) Impregnated active carbon(a) Titanium silicate
Figure Evaluation result of average filtering velocity

This test confirmed that filtering velocity exceeds the threshold in all tests.
Even if liquid waste of high SS concentration after CF concentration treatment is treated, decrease in filtering velocity is small
compared to liquid waste without concentration treatment (comparative evaluation of 1-1 and 1-2).
If coagulant is added, filtering velocity tends to become small compared to cases without coagulant addition. This is thought to bebecause cake thickness increases due to coagulant components. This trend was especially evident for cases treated under low SS
concentration conditions, as in additional test 1-9 and 1-10. This is thought to be because filtering membrane is clogged by
precipitates due to excessive addition of coagulant.
Based on the above, the prospect was obtained that sludge dehydration treatment with the cartridge filter is feasible in all treatment
patters set.
©International Research Institute for Nuclear Decommissioning

No.89

7. Implementation items③-2 Development of sludge dehydration treatment technology➢

Applicability assessment of cartridge filter⚫

Observation of filtering property of each slurry/sludge(average filtering resistivity)

1-1
1-2
1-3
1-4
1-9
1-5

Here is the evaluation result of average filtering resistivity obtained in each test.
No addition of coagulant: 1.0E+11～12[m/kg]
Addition of coagulant: 1.0E+10～11[m/kg]*Judgment value is criteria based on general experience

1-8

Average filtering resistivity [m/kg]

𝛼𝑎𝑣 ≤ 1011 : Easy
1012 ≤ 𝛼𝑎𝑣 ≤ 1013 : Medium
1013 ≤ 𝛼𝑎𝑣 : Difficult

1-7

1-10

Average filtering resistivity [m/kg]

Filtering property judgment value:
Average 𝑓𝑖𝑙𝑡𝑒𝑟𝑖𝑛𝑔 𝑟𝑒𝑠𝑖𝑠𝑡𝑖𝑣𝑖𝑡𝑦 𝛼𝑎𝑣 [m/kg]

1-6

Simulated material
Adsorbent
Adsorption
50ppm
treatment water
CF concentrated
10000ppm
water
Titanium
Sedimentation
Approx.
silicatesludge100000ppm
Sedimentation
Approx.
sludge100000ppm
Sedimentation
50ppm
sludgeAdsorption
200ppmtreatment water
CF concentrated
10000ppm
water
ImpregnatedSedimentation
Approx.
active carbon
sludge100000ppm
Sedimentation
Approx.
sludge100000ppm
Sedimentation
200ppmsludgeCoagulant
―

―

―

―

Aluminum sulfate

5000ppm

High base PAC

2000ppm

High base PAC

500ppm――――Aluminum sulfate

1000ppm

High base PAC

1000ppm

High base PAC

500ppm<1.0E+11[m/kg]

Overall trend of filtering resistivity
・ Adsorption treatment water ≒ CF concentrated water ＞ sedimentation sludge
・ Titanium silicate ＞ active carbon

•
•
•
•

(a) Titanium silicate(b) Impregnated active carbon

Figure Evaluation result of average filtering resistivity

Average filtering resistivity calculated from this test result was below 1.0E+12m/kg in all test cases, which show that filtering performance ofslurry/sludge is good (excluding 1-9 and 1-10).
As an overall trend, it was confirmed that filtering resistivity becomes small when coagulant is added and sedimentation sludge is created. However,since cake thickness (volume) becomes large due to coagulant components, filtering velocity becomes slow as indicated in the previous page.
Additionally, in conditions with no addition of coagulants (1-1, 1-5, etc.), active carbon has smaller filtering resistivity than titanium silicate. Since titanium
silicate carries surface charge, water permeability is thought to have been blunted by electrostatic force that grips water molecules. Since addition of
coagulants neutralize surface charge, filtering resistivity is thought to have decreased.
When coagulation-sedimentation treatment is performed in conditions with low concentration of adsorbents (1-9, 1-10), it was confirmed that sludge
becomes difficult to be filtered.
©International Research Institute for Nuclear Decommissioning

No.907. Implementation items

1-1③-2 Development of sludge dehydration treatment technology

1-2
1-3

➢

Applicability assessment of cartridge filter⚫

Observation on moisture content of filtering residue (cake)

1-4
1-9
1-5
1-6

Cake moisture content decreased when coagulant was added

1-7

Approx. 80%

Approx. 50%(a) Titanium silicate

Cake moisture content [vol.%]Cake moisture content [vol.%]

1-8
1-10

•
•

•

Coagulant
―

―

―

―

Aluminum sulfate

5000ppm

High base PAC

2000ppm

High base PAC

500ppm――――Aluminum sulfate

1000ppm

High base PAC

1000ppm

High base PAC

500ppmApprox. 70%
Cake moisture content is nearly
constant regardless of addition of
coagulant(b) Impregnated active carbon

Figure Measurement result of moisture content of filtering residue (cake)Titanium silicate

Simulated material
Adsorbent
Adsorption
50ppm
treatment water
CF concentrated
10000ppm
water
Titanium
Sedimentation
Approx.
silicatesludge100000ppm
Sedimentation
Approx.
sludge100000ppm
Sedimentation
50ppm
sludgeAdsorption
200ppmtreatment water
CF concentrated
10000ppm
water
ImpregnatedSedimentation
Approx.
active carbon
sludge100000ppm
Sedimentation
Approx.
sludge100000ppm
Sedimentation
200ppmsludge*Target value of moisture content after
dehydration is set to 60vol.% from theviewpoint of volume reduction treatment

When filtering is performed without adding coagulants, moisture content becomes high at approx. 80vol.%. When filtering is
performed by adding coagulants and creating sludge, target value is satisfied at approx. 50vol.%.
Since titanium silicate theoretically carries negative surface charge, moisture content in conditions with no addition of
coagulants is considered to have become high due to the effect of holding water molecules by electrostatic action. Since
crystal structure is three dimensional, it is considered unlikely that water will be held inside by swelling effect (it could
contain hydration water).
In conditions with addition of coagulants, moisture content is considered to have become relatively small, as surface charge
is neutralized and the aforementioned electrostatic impact has been mitigated. Water content of 40-50 vol.% is considered
to be pore water in the cake that is difficult to be dehydrated by filtering.
©International Research Institute for Nuclear Decommissioning

No.91

7. Implementation items
1-1③-2 Development of sludge dehydration treatment technology

1-2
1-3

➢

Applicability assessment of cartridge filter⚫

Observation on moisture content of filtering residue (cake)

1-4
1-9
1-5
1-6

Cake moisture content decreased when coagulant was added

Approx. 80%
Approx. 50%(a) Titanium silicate

Cake moisture content [vol.%]Cake moisture content [vol.%]

1-7
1-8
1-10

•
•
•
•

•

Coagulant
―

―

―

―

Aluminum sulfate

5000ppm

High base PAC

2000ppm

High base PAC

500ppm――――Aluminum sulfate

1000ppm

High base PAC

1000ppm

High base PAC

500ppmApprox. 70%
Cake moisture content is nearly
constant regardless of addition of
coagulant(b) Impregnated active carbon

Figure Measurement result of moisture content of filtering residue (cake)

Impregnated active carbon

Simulated material
Adsorbent
Adsorption
50ppm
treatment water
CF concentrated
10000ppm
water
Titanium
Sedimentation
Approx.
silicatesludge100000ppm
Sedimentation
Approx.
sludge100000ppm
Sedimentation
50ppm
sludgeAdsorption
200ppmtreatment water
CF concentrated
10000ppm
water
ImpregnatedSedimentation
Approx.
active carbon
sludge100000ppm
Sedimentation
Approx.
sludge100000ppm
Sedimentation
200ppmsludge*Target value of moisture content after
dehydration is set to 60vol.% from theviewpoint of volume reduction treatment

Moisture content was 60-70vol.% in all treatment conditions, regardless of whether coagulant is added or not.
It is difficult to perform dehydration treatment up to target value of 60vol.%, but it was confirmed that dehydration is possible to aclose level.
Since active carbon theoretically does not carry surface charge, water molecules are unlikely to be held by electrostatic action, andthere is also no swelling effect. Thus, it is thought that moisture content is mostly constant regardless of whether coagulant is added
or not.
Additionally, the reason why moisture content was high compared to the condition where titanium silicate created sedimentation
sludge (1-3, 1-4) is considered to be because water may be held in micropores as numerous micropores exist on the surface ofactive carbon particles.
Since cake density becomes small due to small true density of active carbon particles, mass water moisture content tends to becomelarge compared to that of titanium silicate.
©International Research Institute for Nuclear Decommissioning

No.92

7. Implementation items③-2 Development of sludge dehydration treatment technology➢ Applicability assessment of cartridge filter⚫ Observation on moisture content of filtering residue (cake)
Based on the test result, the following items are considered to be parameters that largely contributed tomoisture content of filtering residue (cake) remaining after pressurized filtering with the cartridge filter.
・Porosity of cake (particle filling density)
・Electrostatic properties of particle surface
・Crystal structure of particles(existence of swelling, existence of micropores, etc.)

This test result suggested the possibility that moisture held in cakepores is 40-50vol.% in terms of water content.
Additionally, moisture content of 20-30% is estimated to become
even greater, depending on physical properties of powder
adsorbents to be used.

Moisture content [vol.%]Cake moisture content increases due to surface properties andphysical properties of the particles themselves. However, if
coagulation-sedimentation treatment is performed under appropriate
conditions, impact of surface properties of the particles can be
mitigated, and dehydration is possible up to 60-70vol.%.
Changes in cake moisture content due to filtering conditions such asfiltering pressure, filtering time and filter aperture were minor.

② Other water content
Moisture held due to properties of adsorbents.
Held by electrostatic action, held bymicropores, etc. Estimated to be around 030% depending on adsorbent properties.

When cake moisture
content is large

When cake moisture
content is small① Pore water
Moisture held in cake pores. It isestimated that around 40-50vol.% will beheld regardless of adsorbent type.

*Since it is difficult to remove pore water with the pressurized filtering method, it is necessary to apply heated dehydration treatment if
dehydration treatment to a near-dry level is required.
However, in terms of hydrogen measures and volume reduction, dehydration treatment up to 60-70vol.% is considered to be enough.
©International Research Institute for Nuclear Decommissioning

No.93

7. Implementation items③-2 Development of sludge dehydration treatment technology➢ Observation on degraded filtering performance under sedimentation sludge(no prior concentration) conditions

Gelatinous sludge confirmed in test case 1-10

Since coagulant components were large,
cakes with high viscosity and high moisture
content were formed(a) Before drying(b) After drying

Figure Filtrate and filtering residue (cake)••
Figure Cartridge filter test equipment(during pressure filtering test)

Average filtering resistivity [m/kg]

In additional tests 1-9 and 1-10, CF concentration treatment was not performed after adsorption treatmentusing powder adsorbents but coagulation-sedimentation treatment was performed on liquid waste with SSconcentration of several dozen to several hundred ppm to simulate the case of coagulation-sedimentation
treatment. Later, cartridge filter filtering test was conducted using the generated sedimentation sludge.
As a result, significant decrease in filtering velocity was confirmed, and it was observed that it was causedby clogging by gelatinous materials.

Compared to other test cases, significant
increase in average filtering resistivity wasconfirmed.
Filtering velocity also dropped significantly.

Gelatinous materials and powder adsorbent particles are mixed and deposited, and it isconfirmed that sludge volume has increased.
As the test result (average filtering resistivity, etc.) indicates, filtering velocity significantly
dropped compared to other tests.

It is thought that permeation of filtrate is inhibited by clogging of filter holes due to sludge with high moisture content(gelatinous materials).
©International Research Institute for Nuclear Decommissioning

No.94

7. Implementation items③-2 Development of sludge dehydration treatment technology➢

Observation on cause of degraded filtering performance under low adsorbent concentration conditions
Gelation of coagulant components
In test cases 1-9 and 1-10, cakes difficult to be filtered are thought to have formed due to decreased filtering velocity
and increased average filtering resistivity. In test cases 1-9 and 1-10, it is possible that coagulants were excessively
added as adsorbent concentration was small at several dozen to several hundred ppm.
① Components precipitated on the surface
Addition amount of coagulant

•

①
①

•
•

② Components forming gel

②

•
•
•

②
Addition amount ofLarge
powder adsorbent (Test case 1-3, 1-8, etc.)

Components precipitated in narrow space or on chemically active
areas on solid phase surface.
Reaction proceeds preferentially.
Moisture content of precipitates is low.

Small

(Test case 1-9, 1-10)

Components forming colloid and solid phase in supernatant.
Excess components that can not precipitate on solid phase surface
are consumed in this reaction.
Moisture content of precipitates is high. Assumed to be amorphouswith small crystallinity.
*Since Al reagents area used for the coagulant, it is considered toprecipitate in the form of aluminum hydroxide.

As the amount of powder adsorbents becomes smaller, thepercentage consumed for precipitation on the surface becomes
smaller and percentage consumed for gel formation increases.
When adsorbent concentration is large (test case 1-3, etc.), because of the enormous surface area in liquid phase, almost all
coagulant components are consumed for precipitation on the surface and there are only a few excess components for forminggel. On the other hand, when adsorbent concentration is small (test case 1-9, etc.), because there is not enough surface area inliquid phase, there are excess components of coagulant. Excess components are considered to become gelatinous as they
precipitate containing a large amount of moisture in supernatant.
Gelatinous materials cause significant blockage of filter holes, which is considered to have caused predominant decrease in
filtering performance.

Figure Image of consumption of supersaturated components of coagulant©International Research Institute for Nuclear Decommissioning7. Implementation items

No.95③-2 Development of sludge dehydration treatment technology➢

Summary of element test result•••

In treatment conditions No.1-3, it was evaluated that cartridge filter satisfies the required throughput.
On the other hand, in treatment condition No.4, decreased throughput considered to have been caused by filter hole blockage
due to gelatinous materials was confirmed, and it was suggested that dehydration treatment using cartridge filter may bedifficult.
Moisture content of sludge due to the cartridge filter somewhat affects front treatment, but it can be reduced to around 6070%.

Treatment process evaluated to satisfy target throughput.
Filtering performance of generated slurry/sludge is good.
No.1 adsorption treatment waterRO
concentrated
water

Adsorption
treatment

Cartridge filter
treatment

Sludge after
dehydration

Cartridge filter
treatment

Sludge after
dehydration

Cartridge filter
treatment

Sludge after
dehydration

No.2 CF concentrated waterRO
concentrated
water

Adsorption
treatment

CF concentration
treatment

No.3 coagulation-sedimentation sludge
RO
concentrated
water

Adsorption
treatment

CF concentration
treatment

Coagulationsedimentation
treatment

Treatment process evaluated that it may be difficult to satisfy target throughput.
Generated slurry/sludge is difficult to be filtered.
Gelatinous
No.4 coagulation-sedimentation sludge (no preliminary concentration)
ROconcentrated
water

Adsorption
treatment

precipitates are
generated
Coagulationsedimentation
treatment

Filtering becomesdifficult due to clogging
Cartridge filter
treatment

Sludge after
dehydration

Based on the element test result, dehydration treatment is expected to be feasible in treatment process No.1-3.
It is necessary to rationalize the entire treatment process and study materialization of each equipment, based on the results.
©International Research Institute for Nuclear Decommissioning

No.96

7. Implementation items③-2 Development of sludge dehydration treatment technology[Results to date]✓ Regarding sedimentation sludge generated by RO concentrated water treatment, volume reduction effectand generated amount of hydrogen were evaluated and target value after dehydration was set to 60vol.%.
✓ Technology using the cartridge filter was selected as the dehydration treatment method of slurry/sludge,
and element test proposal was developed to evaluate applicability to actual equipment.
✓ Based on the element test result, adsorption treatment water, CF concentrated water and sedimentation
sludge all showed good filtering performance with the cartridge filter and were evaluated to satisfy
throughput.
✓ Possibility of low applicability was indicated for a system that performs coagulation-sedimentation
treatment without concentration treatment after adsorption treatment, as sedimentation sludge becomesgelatinous form that is difficult to be filtered.
✓ Moisture content of cake after filtering was 60-70vol.%, depending on treatment conditions. This indicatesthe possibility of volume reduction of the generated amount of sludge to the set target level.

[Future issues to be studied]✓ Equipment scale, throughput and operation pattern shall be studied for each treatment system
proposed, thereby selecting a more appropriate system.
✓ Different dehydration technologies shall be studied, and development issues considered necessary forthe study of system rationalization, such as discharge of cake by air backwashing, shall be organized.

[Summary]
Dehydration technology using the cartridge filter was selected as the sludge dehydration technology, andits applicability was evaluated with element tests. Based on element test results, multiple treatment
processes where dehydration treatment using the cartridge filter is feasible were selected, and conceptual
study of the sludge collection system was conducted.
©International Research Institute for Nuclear Decommissioning8. Specific goals for achieving the purpose of the project

No.97(1) Liquid/Gas Phase Systems① Development of soluble α nuclide removal
technology
・Study of soluble α nuclide removal test
assuming use of actual liquid

In order to verify the effectiveness and feasibility of soluble α nuclide removal
equipment, test methods, test equipments and test plans assuming use of actual
liquid are materialized.
(Target TRL at end: Level 3)・Element test assuming fuel debris retrieval
work

Regarding soluble α nuclide removal equipment, candidate adsorbents are
selected upon acquiring α nuclide adsorption performance data under anenvironment assuming fuel debris retrieval. At the same time, water quality
adjustment policy is also set.
(Target TRL at end: Level 4)
Regarding RO concentrated water treatment facility, powder adsorbents andcoagulants are selected upon evaluating nuclide removal performance of
powder adsorbents and coagulation-sedimentation performance of coagulants.
(Target TRL at end: Level 4)② Development of RO concentrated water
treatment technology
・Selection of adsorbent and condensing
agent・Study of applicability to actual equipment③ Development of secondary waste treatment
technology
・Inspection of pretreatment technology
・Study of applicability to actual equipment

Regarding concentrated water treatment facility, conceptual design of treatment
methods and equipment is implemented based on tests using the sedimentation
separation tank.
(Target TRL at end: Level 4)
Candidate stabilization treatment technology is selected for sludge generated
from coagulation sedimentation tank. At the same time, pretreatment for liquid
waste such as supernatant water is also selected.
(Target TRL at end: Level 3)
Regarding pretreatment technology, applicability to actual equipment is verified
based on element test results. Conceptual design is also established for
pretreatment equipment.
(Target TRL at end: Level 4)

Goals indicated in the above table have been achieved.

*TRL: Technology Readiness Level©International Research Institute for Nuclear Decommissioning

FY2021 Subsidy Project of Decommissioning and Contaminated Water
Management Development of Safety System(Liquid/Gas Phase Systems, Criticality Control Technology)

Final Report(2) Criticality Control Technology
June 2023
Technology Research Association International ResearchInstitute for Nuclear Decommissioning (IRID)
All rights reserved.
©International Research Institute for Nuclear Decommissioning (IRID)

0

No.1

Table of Contents
1 Purpose and goals of the subsidy project
2 Overview of subsidy project
3 Results of past projects and remaining issues
4 Input and output information
5 Implementation schedule
6 Implementation framework
7 Implementation items
7.1 Development of on-site operating procedures
7.2 Development of solidified absorbent technology (water glass)
7.3 Overall summary of criticality control
7.4 Organization of issues for the future
8 Specific goals for achieving the purpose of the project
Reference©International Research Institute for Nuclear Decommissioning1. Purpose and goals of “Development of Safety System(Liquid/Gas Phase Systems, Criticality Control Technology)”

No.2[Purpose of Development of Safety System (Liquid/Gas Phase Systems, Criticality Control
Technology)]
At Tokyo Electric Power Company Holdings Inc. Fukushima Daiichi Nuclear Power Station (1F), it is considered that
nuclear fuel melted with reactor internals and exists as fuel debris in the reactor pressure vessel (RPV) and primarycontainment vessel (PCV).
Fuel debris in RPV and PCV is currently considered to be in subcriticality, but since the plant itself is in an unstablestate which is different from the initial design, such as the reactor building (R/B), RPV, PCV, etc. being damaged dueto the accident, it is necessary to prevent the diffusion of radioactive materials and achieve a stable state byremoving fuel debris and maintaining subcriticality of fuel debris.
Against the above background, this project will conduct reviews to realize large-scale fuel debris retrieval task
under engineering and project management performed by Tokyo Electric Power Company Holdings Inc. (TEPCO) in
accordance with the “Mid-and-Long-Term Roadmap towards the Decommissioning of TEPCO's Fukushima Daiichi
Nuclear Power Station Units 1-4” (hereinafter Mid-and-Long-Term Roadmap). Development results of this project will
be utilized in engineering by TEPCO.
The purpose of this project is to facilitate decommissioning and contaminated water measures of 1F and raise thelevel of science and technology of Japan, by implementing the project that supports the development of technologies
that contribute to decommissioning and contaminated water measures of 1F in accordance with the Mid-and-LongTerm Roadmap and “FY2021 Decommissioning Research and Development Plan” (Decommissioning andContaminated Water Team Meeting / Secretariat Liaison Meeting (86th)).
In specific, technology related to on-site operation methods shall be developed regarding soluble α nuclide removal
technology for α nuclides considered to be eluted in circulation cooling water from fuel debris, RO concentrated
water (*) treatment technology, secondary waste treatment technology, as well as criticality approach monitoring
technology and neutron absorbent technology.

©International Research Institute for Nuclear Decommissioning1. Purpose and goals of “Development of Safety System(Liquid/Gas Phase Systems, Criticality Control Technology)”

No.3[Overall development goals]
Perform necessary element technology development and tests regarding system technology
and technology for ensuring safety, required to further expand the scale of retrieving fuel debris
and internal structures, in accordance with R&D results obtained up to date.

[Specific goals of development items]① Development of on-site operating procedures⚫ Develop on-site operating procedures for criticality approach monitoring using
neutron detectors and for non-soluble neutron absorbents, and provide information to
be utilized by the fuel debris retrieval method team for design of the retrieval systemand study of throughput.
⚫ Evaluate subcriticality measurement performance of the three new types of neutron
detector, and incorporate it in the on-site operation procedures.
② Development of solidified absorbent technology⚫ Acquire data on impact on the drying process of fuel debris with solidified absorbent(water glass) attached, and utilize it for ① Development of on-site operatingprocedures as well as study on equipment operation such as drying time of fueldebris.
©International Research Institute for Nuclear Decommissioning

No.4

2. Overview of subsidy project
Development of on-site operation methods for criticality approach monitoring technology andneutron absorbent technology
Regarding criticality control technology, criticality prevention, criticality approach monitoring,criticality detection and impact mitigation have been developed in the Government-led R&D
Program on Decommissioning and Contaminated Water Management up to the previous PJ.
Among these, study on criticality prevention and criticality approach monitoring, which haveissues remaining, was materialized based on past technical development results.
③ Detection ofcriticality

a. Limitation and managementof work procedures toprevent criticality② Criticality
approachmonitoring① Criticality
prevention
b. Injection of neutron absorbents
Solidified
Solidified

Scope of this project

Glass material Gd2O3
containing B・Gd particle

Neutrondetector

Gas
radiation
monitor

Emergency
boric acidinjection④ Mitigation of
impactWater glass /Gd2O3 granulated
powderOverall view of technical development for criticality control©International Research Institute for Nuclear Decommissioning2. Overview of subsidy project

No.5

Gradual criticality control⚫ Criticality control technology shall be gradually applied considering risks of criticality according to the retrieved
amount and retrieval method of fuel debris.
⚫ Since criticality risk relatively increases as the retrieval method has greater amount retrieved and larger changeof state, it shall be made possible to select diversified criticality control technologies.

Retrieval task

Trace amount(several g)

Retrieval method

Grabbing, suction,
etc.

Work restriction① Criticality
prevention② Criticality
approachmonitoring

Fuel debris retrievalInternal
investigation

Retrieved amount

Large

State of fuel debris

Small

Retrieval gradually expanding the scale
Small amount (several kg)
Grabbing, suction,
etc.

Core boring, etc.

・Limit of processing
Method that does
Method that does notamount per task
not makes changes makes changes to・Limit of retrieval
to fuel debris
fuel debris
position interval

Full-scale retrieval

Small amount (several
kg - several dozen kg)

Core boring, etc.

・Limit of processing
amount per taskーー

Neutron flux
monitoring

ー

ー

ー

ー

Non-soluble
neutron absorbent

ー

ー

ー

ー

Soluble neutron
absorbent

ー

ー

ー

ー

Criticality approachmonitoring③ Detection ofcriticality

Detection ofcriticality

PCV gas radiation
monitor④ Mitigation of
impactTermination ofcriticality

Injection of sodium Injection of sodium
pentaborate water pentaborate water

PCV gas radiation
monitorFull-scale(up to several hundredkg per day)Core boring, chisel, etc.

・Limit of processing
amount per task
(expanded)

Neutron flux monitoring Neutron flux monitoring
Subcriticality
measurement*1
Non-soluble neutron
absorbent*2

Scope ofthis project

Sodium pentaborate
water*3 (6000pp)

PCV gas radiation
monitorNeutron flux monitor

PCV gas radiation
monitorNeutron flux monitor

PCV gas radiation
monitorNeutron flux monitor

Injection of sodium
pentaborate water

Injection of sodium
38
pentaborate
water

Injection of sodium
pentaborate water*1, 2 and 3 are selected depending on criticality risk taking retrieval work details and field environmental conditions into account©International Research Institute for Nuclear Decommissioning3. Results of past projects and remaining issues

No.6

Development results of criticality approach monitoring technology and contents of this project
Final goals

Subcriticality
Establishment of
measurement
technology

Incorporation into
fuel debris
retrieval system

Goals to be achieved before
application to actual
equipment

Achievement status① Establishment of measurement
methods

・Selection of method combining reactor noise method and neutron source
multiplication method
・Study of detector sensitivity and arrangement to reduce measurement
error② Development of system
specifications

・Development of system specifications of detector, measurement circuit,etc.

③ Specifications of neutron detector

・Development of neutron detector specifications and selection of detector
for prototype④ System design and test
manufacturing

・Test manufacturing of system consisting of neutron detector,
measurement circuit and analysis PC, as test system⑤ Confirmation of feasibility of
subcriticality measurement

・Confirmation under uniform fuel debris simulation conditions (KUCA test
#1/#2)
・Confirmation under large-scale fuel debris simulation conditions (KUCA
test #3)
・Confirmation under non-uniform fuel debris / neutron absorbent simulation
conditions (KUCA test #4)① Confirmation of applicability ofneutron detector in the field・Organization of specifications of neutron detector unit for transfer withrobot arm
・Development of cable handling concept
・Test manufacturing of alternative neutron detectors (corona discharge
type, SiC semiconductor type, multicell He-3 type) and confirmation of basic
performance・Design and test manufacturing of neutron detector unit
・Evaluation of radiation resistance of preamplifier② Solenoid noise reduction method

Evaluation of impact of simulated noise, organization of draft response
measures③ Study of criticality approach
management procedures

Remaining issues / contents of this projectEvaluation of time required for measurement
Criticality approach monitoring procedures in each step of fuel debris
retrieval method

Applicability confirmation test of alternative neutron
detectors (corona discharge type, SiC semiconductor
type, multicell He-3 type) (KUCA test #5) (→7.1.2)
KUCA: Kyoto University Critical Assembly

Organization of measurement procedures before andafter processing, judgment of measurement target,
measurement time, work procedures of dailysubcriticality maintenance management, repair andmaintenance requirements, and request to processing
side (→7.1.1)©International Research Institute for Nuclear Decommissioning3. Results of past projects and remaining issues

No.7

Development results of non-soluble neutron absorbent and contents of this project
Final goals

Establishment ofcriticality
prevention
technology

Goals to be achieved before
application to actual equipment① Selection of candidate materials

・Selection of candidate non-soluble neutron absorbent upon
confirming basic characteristics, irradiation characteristics andnuclear characteristics (selection among 4 types of candidate
absorbent)② Confirmation of workability andspecifications of conditions formaintaining subcriticality

・Estimation of necessary amount of injection with the assumed
usage method
・Confirmation of workability of fuel debris crushed with chisel,
confirmation of absorbent distribution, and evaluation ofmaintenance of subcriticality (solid absorbent)③ Evaluation of corrosion impact

Remaining issues / contents of this project・Based on long-term irradiation test, hydrogen generation G
value (*) is below the design value and diluted seawater withleached absorbent components is pH6 or higher
・Evaluation of impact on rust inhibitor effect under irradiation
・Increase in the number of storage canisters and amount ofwaste by approx. 10% at maximum with solid absorbents andapprox. 40% at maximum with solidified absorbents

Impact of water glass on fuel debris drying(→7.2)① Study of injection equipment andoperation method for non-soluble
neutron absorbent

・Determination of absorbent injection method
・Confirmation of impact of chisel processing
・Confirmation that underwater injection is possible

Used amount of solidified absorbent
Application conditions for processing
scenesRecovery procedures from level 2 to level 1
Effectiveness of pre-injection(→7.1.3)② Design of absorbent injection
equipment

・Development of absorbent injection equipment concept basedon weight and dimensional constraints of absorbent injection
equipment and absorbent transport route constraints

Downsizing and weight reduction ofequipment (→7.1.3)③ Verification of combinations ofretrieval equipment

・Organization of applicable absorbent for fuel debris processing
method of each fuel debris position③ Evaluation of secondary effectIncorporation into
fuel debris retrieval
systemAchievement status(*) Hydrogen generation G value: Amount of hydrogen generated when absorbing 100eV of energy©International Research Institute for Nuclear Decommissioning

No.8

4. Input and output information
This PJ was conducted in coordination with the Project of Development of Technology for Further
Increasing the Scale of Retrieval of Fuel Debris and Internal Structures and Project of Developmentof Fuel Debris Collection, Transfer and Storage Technologies.

ID

Demand-side project

Provider-side project

Description (overview)

Methodofinterference
Development of technology
removal, method of access tofor further increasing the scale
fueldebris,methodofof retrieval of fuel debris andprocessing and collecting fuelinternal structures
debris and reactor internals

TimingUsage of information

1

Developmentof
SafetySystem (Liquid/Gas Phase
Systems, Criticality Control
Technology)2

Developmentof
SafetyDevelopment of technology
System (Liquid/Gas Phasefor further increasing the
Systems, Criticality Control
scale of retrieval of fuelTechnology)
debris and internal structures

On-site operating procedures
for
criticality
approachmonitoringwithneutronMarch 2023 Study of throughput of fueldetector, on-site operating
debris retrieval method
procedures for non-soluble
neutron absorbent

3

Developmentof
SafetyDevelopment of fuel debris
System (Liquid/Gas Phase
collection,transferand
Systems, Criticality Control
storage technologies
Technology)

Data on impact on the drying
process of fuel debris with March 2023 Study of storage
water
glass
(non-soluble
drying process
neutron absorbent) attached

June 2021

Study of on-site operatingprocedures
for
criticality
approachmonitoring
andabsorbent spraying

canister©International Research Institute for Nuclear Decommissioning

No.9

5. Implementation schedule
Criticality Control Technology
大分類
Large
category

Smallcategory
小分類

Implementing
operator
実施事業者(Also list the（外注先、試験場所がsubcontractor andある場合は併記）
testingsite, if any)

FY2021
令和３年度

Jun Jul
Dec 1月Jan FebMar
Oct 11月
Apr
Jun 7月Jul Aug
Dec 1月Jan Feb
Mar AprSep 10月
Nov 12月
Oct 11月
Nov 12月
4月 May
5月 6月
8月 Sep9月 10月
2月 3月
4月 May
5月 6月
7月 Aug
8月 9月
2月 3月

Major
milestones
主要なマイルストン
Project steering meeting
プロジェクトステアリング会議(2) On-site operation
methods for criticality(２) 臨界近接監視技術・中性
approach monitoring
子吸収材技術の現場運用
technology and neutron
方法
absorbent technology①現場運用手順の開発① Development of on-site(i) 臨界近接監視の運用手
operatingprocedures(1) 順の策定
Development ofoperating procedures ofcriticality approachmonitoring

a.a.Organizationof
前提条件の整理
prerequisites
b.b.Studyof
手順の検討
procedures
c.c.Evaluation/
評価・見直しrevision
d.d.Summary
まとめ

plan
中性子検出器の未臨界 a.a.Test
試験計画(ii)(ii)Subcriticality
度測定試験test ofmeasurement
b. Testneutron detectorb. 試験準備・手配
preparation
andarrangementc. 要素試験

c. Element test
まとめ
d.d.Summary

非溶解性吸収材の(ii(iii)
iDevelopmentof
運用手順の策定
operatingprocedures fornon-soluble absorbents

a. a.Organizationof
前提条件の整理
prerequisites
b. b.Studyof
手順の検討
proceduresc. c.Evaluation/
評価・見直しrevisiond. d.Summary
まとめ

FY2022
令和４年度

Interimreport中間報告

reportProject steering meeting Interim
プロジェクトステアリング会議
中間報告

Project steering
プロジェクトステ
meeting
アリング会議Interim report
中間報告

プロジェクトステ
Project steering 最終
Final reportmeeting 報告
アリング会議Organizationof prerequisites
前提条件の整理（１）日立GE(1) Hitachi GE
Studyof procedures
手順の検討（２）東芝ＥＳＳ(2) Toshiba ESS

Evaluation/ revision
評価・見直し（３）三菱重工(3) Mitsubishi Heavy(1) Toshiba ESS（１）東芝ＥＳＳ
Subcontractor:
外注先：FEO/CETD
FEO/CETD
試験場所：ＫＵＣＡ
Testing site:
名古屋大学
KUCA
（２）三菱重工
NagoyaUniversity
外注先：IPL(2) Mitsubishi Heavy
試験場所：ＫＵＣＡ

まとめ
Summary

Interimsummary
中間まとめ

Industries

Test plan
試験計画

Test
preparation and arrangement
試験準備・手配KUCA test
KUCA試験

Element
test
要素試験

Test plan
試験計画
Irradiation test
照射試験
分析・評価
Analysisand evaluation

Element
test
要素試験
Summary
まとめInterim summary
中間まとめ

Industries
Subcontractor: IPL
Testing site: KUCA

（１）日立GE

Organizationof prerequisites
前提条件の整理(1) Hitachi GE

Study手順の検討
of procedures（２）東芝ＥＳＳ(2) Toshiba ESS
Evaluation/ revision
評価・見直し（３）三菱重工(3) Mitsubishi Heavy

まとめ
Summary

Interimsummary
中間まとめ

Industries②固化型吸収材技術の開② Developmentof
発
solidified
absorbent
technology

試験計画
a.a.Testplanb. Testb. 試験準備・
preparation and
資材調達
procurement ofc. 要素試験
materials
d. まとめc. Element test
d. Summary(1) JAEA
Subcontractor: Inspection（１）JAEA
Development
Company
外注先：検査開発株式会社
Ltd.
試験場所：JAEA
Testingsite: JAEA
Consignee:
Shizuoka
委託先：静岡理工科大学
Instituteof Science and試験場所：静岡理工科大学
Technology
Testing site: Shizuoka
Institute of Science andTechnology

Test plan
試験計画
Test
preparation and procurement
試験準備・調達

Test preparationand procurement
試験準備・調達
Basic
test of parameters
パラメータ基礎試験

基礎試験
Basic test
Interimsummary
中間まとめ

Summary
まとめ©International Research Institute for Nuclear DecommissioningNo.10

6. Implementation framework
Tokyo Electric Power Company
Holdings Inc.
➢ Various adjustments from theperspective of applicability in the field

Technology Research Association International
Research Institute for Nuclear Decommissioning➢ Development of overall plan and coordination of technicalsupervision➢ Coordination of technical management such as progress oftechnical development

Collaborating development
project teams

Hitachi-GE Nuclear
Energy, Ltd.

Toshiba Energy Systems& Solutions Corporation

Mitsubishi Heavy
Industries, Ltd.

Japan Atomic Energy Agency[Element test / technical development](1) Development of Liquid/Gas Phase
Systems① Development of soluble α nuclide
removal technology
・Review of dissolved α nuclide removal
test assuming use of actual liquid(accumulated water in reactor building)・Element test assuming fuel debris
retrieval task
② Development of RO concentrated water
treatment technology
・Selection of adsorbent and coagulant
・Review of applicability to actual
equipment③ Development of secondary waste
treatment technology
・Inspection of stabilization treatment
technology
・Review of applicability to actual
equipment(2) On-site operation methods for criticality
approach monitoring technology andneutron absorber technology
・Development of criticality approachmonitoring procedures[Element test / technical development](1) Development of Liquid/Gas Phase
Systems① Development of soluble α nuclide
removal technology
・Review of dissolved α nuclide removal test
assuming use of actual liquid(accumulated water in reactor building)・Element test assuming fuel debris retrieval
task[Element test / technicaldevelopment](2) On-site operation methods forcriticality approach monitoring
technology and neutron absorbertechnology[Element test / technicaldevelopment](2) On-site operation methods forcriticality approach monitoring
technology and neutron absorbertechnology

・Development of criticality
approach monitoring procedures・Performance assessment ofsubcriticality measurement of
neutron detector

・Development of solidified
absorbent technology[Details of re-consignment]
・Subcriticality measurement test (Kyoto
University)[Main outsourcing contents]
・Operational test of non-soluble neutron
absorbent (HPC)・Water flow test of MF membrane and otherfilters (Misuzu Seiko)・Cartridge filter test (HPC)[Details of re-consignment]
・Subcriticality measurement test (Kyoto
University)[Main outsourcing contents]
・Soluble α nuclide removal test (NFD)
・Corona-type detector (FEO/TENEX-J)
・Multicell-type detector (CETD)(2) On-site operation methods for criticality
approach monitoring technology andneutron absorber technology
・Development of criticality approachmonitoring procedures・Performance assessment of subcriticality
measurement of neutron detector
・Development of operating procedures forabsorbents
・Study of impact on fuel debris drying
process

Development of fueldebris retrieval method
Development of fueldebris packaging,transport and storage
technology
Research anddevelopment of solid
waste treatment anddisposal
Development of analysis
and estimation
technology for fuel debris
characterization
Development oftechnology for detailed
investigation inside PCVDevelopment oftechnology forinvestigation inside RPV[Details of re-consignment]
・Subcriticality measurement test (Kyoto
University)[Main outsourcing contents]
・Assistance in subcriticality measurement
test (IPL)・Assistance in study of criticality approachmonitoring procedures (NS Engineering)[Details of re-consignment]
・Structural analysis of solidified absorbents(Shizuoka Institute of Science andTechnology)[Main outsourcing contents]
・Contract work for drying tests of test
specimens coated with solidified absorbents(Inspection Development Company Ltd.)

Development of analysis
and estimation
technology for fuel debris
characterization
Development of technology
to sort fuel debris andradioactive waste©International Research Institute for Nuclear DecommissioningNo.11

7. Implementation items
7.1 Development of on-site operating procedures
7.1.1 Operating procedures of criticality approach monitoring[Issues]
Criticality approach monitoring in rough work steps of top/side access method is studied in a relevant project (*)(Figure 1). In order to discuss operating procedures and throughput with the fuel debris retrieval method team, it isnecessary to further refine the work steps and materialize criticality approach monitoring procedures.
[Implementation items]⚫ Criticality approach monitoring procedures using neutron detectors shall beorganized in a step diagram for the fuel debris retrieval method andprocessing and collection equipment studied in the relevant project(*).
⚫ Neutron response to criticality approach during fuel debris processing shallbe studied, and procedures for response to determine criticality approach
shall be studied.
⚫ Organization of measurement procedures before and after processing,
judgment of measurement target, measurement time, work procedures ofdaily subcriticality maintenance management, repair and maintenancerequirements, and request to processing side shall be studied.
[Goals]⚫ Develop on-site operating procedures for criticality approach monitoring
using neutron detectors, and provide information to be utilized by the fueldebris retrieval method team for design of the retrieval system and study ofthroughput.

Criticality approach monitoring
before processing andcontinuous monitoring during
processing, using processing
tool juxtaposition type detector
during fuel debris processing

Processing tool juxtaposition
type detector

Processing
tool

Figure 1 Criticality approach monitoring during fuel debris
retrieval (example of side access PLAN-A)(*) Relevant project “Development of technology for further increasing the scale of retrieval of fuel debris and internal
structures” (FY2019, FY2020)©International Research Institute for Nuclear Decommissioning7. Implementation items

No.12

7.1 Development of on-site operating procedures
7.1.1 Operating procedures of criticality approach monitoringa. Organization of prerequisites
Functions that comprise criticality approach monitoring
Name

Subcriticality measurement detector

Criticality approach monitoring
detector(Note 1)

Continuous monitoring
detector(Note 1)

Usage purpose

Grasp pre-work conditions

Determine start of processing

Detect unexpected changes

Subcriticality measurement

Criticality approach monitoring before
and after processing

Continuous monitoring of neutron flux
during processing

Function

Measurement of absolute value ofmultiplication factor

Measurement of relative change of multiplication factor

Form

Arm-mounted type

Processing tool juxtaposition type

Platform-mounted type / processing tool
juxtaposition type

Weight30kg-100kg

30-50㎏Less than 30kg

Measurement time

Several hours to 1 week
(depends on site environment)

Around 10min.

Continuous

Measurement position

One point near the retrieval
position(Note 2)

Move as necessary according to theSame as left / location with an overhead
processing position(Note 2)view of the processing position(Note 1) “Criticality approach monitoring” and“continuous monitoring” may be usedconcurrently under conditions such as weak
gamma ray in the field and strong neutrons offuel debris.
(Note 2) Measurement positions for fueldebris are not fixed and may be changed
each time.

Platformmounted type

©International Research Institute for Nuclear DecommissioningNo.13

7. Implementation items
7.1 Development of on-site operating procedures
7.1.1 Operating procedures of criticality approach monitoringb. Study of procedures
Overview of implementation items of 7.1.1.b

10 items raised as
issues were studied,in discussing
operatingprocedures andthroughput with thefuel debris retrieval
method team.

- - - Top retrieval method - - -- - - - - - - - - - - - Side retrieval method - - - - - - - - - - -① Criticality monitoring
method during top retrieval(→No.14)② Criticality monitoring
method of side retrieval
PLAN-A method(→No.15)③ Criticality monitoring
method of side retrieval
PLAN-B method(→No.16-18)- Top access method for transferring
the unitized structure -③ Criticality monitoring
method of top accessmethod for transferring
the unitized structure(→No.16-18)④ Monitoring range of three types of criticality approach monitoring (→No.19)⑤ Organization of measurement procedures during fuel debris processing
and retrieval work (→No.20)⑥ Organization of roles of criticality approach monitoring (→No.21)⑦ Basic approach of criticality monitoring based on organized environmental
conditions of partial submersion processing and submersion processing(→No.22)⑧ Organization of criticality risk in submersion processing (→No.23-26)⑨ Monitoring procedures for representative processing examples (laser
processing) (→No.27)⑩ Maintenance of monitoring system(→No.28)©International Research Institute for Nuclear DecommissioningNo.14

7. Implementation items
7.1 Development of on-site operating procedures
7.1.1 Operating procedures of criticality approach monitoringb. Study of procedures① Criticality monitoring method during top retrieval[Study results]
Method of criticality approach monitoring infuel debris retrieval steps was studied
using the top retrieval method as arepresentative example.
Criticality approach monitoring consists ofthree types of neutron measurement, but it
is unnecessary to have all three types ofmonitoring in all work steps.

Platformmounted typedetector(continuous
monitoring)

Platform-mounted
type detector(continuous
monitoring)

Processing tool
juxtaposition typedetector(approach monitoring)

Processing tool
juxtaposition typedetector(approach monitoring)

Upper grid plate

Necessary monitoring functions shall be
used differently according to the assumed
criticality risk.
(3 types or 2 types)

Remaining fuel in the core

Platformmounted typedetector(continuous
monitoring)

Processing tool
juxtaposition typedetector(approach monitoring)

Arm-mounted typedetector (subcriticality
measurement)

Arm-mounted
type detector(subcriticality
measurement)

Processing tool
juxtaposition typedetector(approach monitoring)Core support plate

RPV bottom©International Research Institute for Nuclear DecommissioningNo.15

7. Implementation items
7.1 Development of on-site operating procedures
7.1.1 Operating procedures of criticality approach monitoringb. Study of procedures② Criticality monitoring method of side
retrieval PLAN-A method[Study results]
Criticality approach monitoring
method was studied for the side
retrieval PLAN-A method.
Of the three criticality monitoring
functions, criticality measurement
shall be performed with the armmounted type detector and criticality
approach monitoring and continuous
monitoring during processing shallbe performed with the processing
tool juxtaposition type detector,
thereby reducing the replacement
frequency of processing tools anddetectors and minimizing the impact
on fuel debris retrieval workthroughput.

Criticality monitoring method of side retrieval PLAN-A method
Subcriticality
measurement using armmounted type detector
before fuel debris
processing

Processing tool
juxtaposition typedetector

Confirmation of
conditions using armmounted type detector
before fuel debris
processing
Arm-mounted type detector

Arm-mounted typedetector
Criticality approachmonitoring before
processing and continuous
monitoring during
processing, using
processing tool
juxtaposition type detector
during fuel debris
processing
Processing tool
juxtaposition
type detector

Continuous monitoring
using tool juxtaposition
type detector during
interference objectives
processing

Processing tool

Figure Criticality monitoring during fueldebris retrieval (excerpt)

Processing tool
juxtaposition
type detector

Processing tool

Figure Criticality monitoring during
interference removal (excerpt)©International Research Institute for Nuclear DecommissioningNo.16

7. Implementation items

7.1 Development of on-site operating procedures
7.1.1 Operating procedures of criticality approach monitoring b. Study of procedures③ Criticality monitoring methodof side retrieval PLAN-B method / top access method for transferring the unitized structure⚫ Overview of side retrieval PLAN-B method / top access method for transferring the unitized structure
Secondary boundary

Carrying-in cell

Waste
carrying-out
cell

Fuel debris
retrieval cell

Passageway

Retrieval equipment

Work container

Maintenance cell

Additional building

PCVContinuous
sleeveFuel debris
carrying-out cell

Dual lid
TransporterWork
container

Passageway (access tunnel)

Dedicatedtransport cask

Moving arm

Work arm・Access from equipment hatch
・Use of multiple work equipment (some equipment is assembled
in PCV)・Integration and transfer of core internals including fuel debris as solidified materials
containing neutron absorbent
・Top access retrieval of fuel debris below pedestal (bottom right on previous page)©International Research Institute for Nuclear DecommissioningNo.17

7. Implementation items

7.1 Development of on-site operating procedures
7.1.1 Operating procedures of criticality approach monitoring b. Study of procedures③ Criticality monitoring methodof side retrieval PLAN-B method / top access method for transferring the unitized structure⚫ Regarding the work steps of the side retrieval PLAN-B method and top access method for transferring the unitized
structure, criticality approach monitoring procedures when applying the processing and collection method using thefollowing equipment were studied, and issues and risks were identified.

Loading and moving of
neutron detector unit
(simulated) with the heavy
weight object transfer
equipment(Side retrieval PLAN-B
method)Loading and moving of
crushing and suction
equipment with the heavy
weight object transfer
equipment(Side retrieval PLAN-B
method)Drawing of interference
removal work at core bottom(Top access method fortransferring the unitized
structure)

Drawing of fuel debris retrieval
at pedestal bottom after
carrying out bottom RPV
mirror and CRD housing(Top access method fortransferring the unitized
structure)©International Research Institute for Nuclear DecommissioningNo.18

7. Implementation items
7.1 Development of on-site operating procedures
7.1.1 Operating procedures of criticality approach monitoring b. Study of procedures③ Criticality monitoring
method of side retrieval PLAN-B method / top access method for transferring the unitized structure
No.

[Studied results]⚫ Issues and risks regarding work stepsof side retrieval PLAN-B method and
top access method for transferring theunitized structure were identified in thetable on the right.

Issues1

Can existence and amount of fuel debris be confirmed from camera footage, neutron flux distribution andcharacteristic gamma ray distribution?

2

Can the neutron detector be installed at an appropriate position and distance with the heavy weight object
transfer equipment which is used to transfer neutron detectors?

3

Can subcriticality be measured for fuel debris that is above criticality mass and weighs several dozen kg?

4

Can the neutron detector be installed if there is an inclined surface on fuel debris?

5

Can the neutron detector be installed at an appropriate position and distance by transferring the neutron
detector on the platform?

6

Can reactivity input be caused by crack propagation and water intrusion by directly installing the equipment onfuel debris?

7

How would it be managed if uncollectible lumps and puddles in depression are formed when lowering thewater level and processing and collecting sediments?

8

The cutting surface is predicted to become lower than the pedestal bottom, when removing the fuel debris
mixed with bottom concrete. Thus, measures against accumulation of cooling water discharged from theworker access tunnel are considered to be necessary.

9

Can the neutron detector be transferred between the shroud perimeter and RPV and installed at an
appropriate position and distance?

10

Reactivity may be input due to discharge of cooling water being prevented by measures to prevent outflow offilling solidification agent (transient state until solidification agent injection).

No.

Risk

1

Reactivity input due to crack propagation and water intrusion caused by equipment drop and impact on fueldebris.
Crack propagation due to falling of fuel debris, and excessive criticality mass, during processing.

2

3

Reactivity input due to crack propagation and water intrusion caused by detachment, falling and impact of fueldebris from solidified parts.
©International Research Institute for Nuclear DecommissioningNo.19

7. Implementation items
7.1 Development of on-site operating procedures
7.1.1 Operating procedures of criticality approach monitoringb. Study of procedures④ Monitoring range of three types of criticality approach monitoring⚫ Respective monitoring range of the three types of criticality approach monitoring functions wasorganized. (figure shows the example of top retrieval)

Function: Criticality approach monitoring
Function: Subcriticality measurement
before and after processing
Measurement point: 1 point near the retrieval
Measurement point: Move as necessary
start pointaccording to the processing position
Usage purpose: Grasp pre-work conditions
Usage purpose: Determine start of processing

燃料デブリ
Fuel
debrisMonitoringarea

Subcriticality
未臨界度測定器
measurement

Function: Continuous monitoring of
neutron flux during processing
Measurement point: Location with an
overhead view at a distance from theprocessing position(*1)
Usage purpose: Detect unexpected changes
臨界連続監視器
Continuous
monitoring

臨界連続監視エリア
Monitoring area

Criticality approachdebris 加工予定場所
Planned processing
燃料デブリ
Plannedmonitoring
Fueldebris 加工予定場所
臨界近接監視器 Fuel
燃料デブリ
locationprocessing
Monitoring area
location 臨界近接監視エリア(*1) Image of continuous monitoring when moving as necessary according to the processing position has the same monitoring area as criticality approach monitoring in the figure in the center.
©International Research Institute for Nuclear Decommissioning7. Implementation items

No.20

7.1 Development of on-site operating procedures
7.1.1 Operating procedures of criticality approach monitoring

Subcriticality measurementb. Study of procedures ⑤ Organization of measurement procedures during fueldebris processing and retrieval work
Light blue: Reactor noise method⚫ Measurement procedures during
fuel debris processing andretrieval work were organized.
・Measure neutron multiplication factor (keff) with thereactor noise method before starting fuel debris
retrieval work ①
This is the basis for the neutron source multiplication
method.
・Measure neutron counting rate before and after fueldebris processing ②・Perform continuous monitoring of neutron counting
rate during fuel debris retrieval processing
・Evaluate neutron multiplication factor (keff) with theneutron source multiplication method if the neutron
counting rate changes
・Stop work when the neutron multiplication factor
reaches the judgment criteria
・Use an absorbent
・Measure neutron multiplication factor (keff) with thereactor noise method after using the absorbent ①
This is a re-measurement of the basis for theneutron source multiplication method.
𝒌𝒆𝒇𝒇 : Neutron multiplication factor
𝒌𝒓𝒆𝒇 : Neutron multiplication factor at the time of
measurement with the reactor noise method
𝝓𝑟𝑒𝑓 : Neutron flux at the time of measurement with thereactor noise method
𝝓 : Pre-retrieval neutron flue

Yellow: Neutron source
multiplication method

Measurement of𝒌𝒆𝒇𝒇 with the reactor noise
method ① (up to 1h)
During pre-retrieval
inspection

Pre-work neutron flux,
𝒌𝒆𝒇𝒇 (𝝓𝑟𝑒𝑓 , 𝒌𝒓𝒆𝒇)

Measurement before retrieval ②(1-2m)

燃料デブリ
Fueldebris Monitoringarea

Pre-retrieval neutron flux (𝝓)Subcriticality
未臨界度測定器
measurement

Criticality approach monitoring
before and after processing

Evaluation of pre-retrieval keff with theneutron source multiplication method
𝐷𝑒𝑟𝑖𝑣𝑒 𝒌𝒆𝒇𝒇 𝑤𝑖𝑡ℎ 𝑡ℎ𝑒 𝑓𝑜𝑙𝑙𝑜𝑤𝑖𝑛𝑔 𝑒𝑞𝑢𝑎𝑡𝑖𝑜𝑛𝝓𝑟𝑒𝑓
𝟏 − 𝒌𝒆𝒇𝒇=𝝓𝒌𝒆𝒇𝒇𝟏 −𝒌𝒓𝒆𝒇Judgment ofsubcriticality wih
retrieval 𝒌𝒆𝒇𝒇Noprocessing location

Ok

Stopping of retrieval work

Fuel debris
processing andretrieval

Criticality
approach
臨界近接監視器
monitoring
臨界近接監視エリア
Monitoring area

Planned
Fueldebris 加工予定場所
燃料デブリ

Use of absorbent

Continuous monitoring of
neutron flux during
processing (constantly inoperation apart from above
figure)©International Research Institute for Nuclear DecommissioningNo.21

7. Implementation items
7.1 Development of on-site operating procedures
7.1.1 Operating procedures of criticality approach monitoringb. Study of procedures ⑥ Organization of roles of criticality approach monitoring

Criticality approach monitoring before and after processing⚫ Roles of the three types of monitoring functions whenapproaching criticality were organized.

Neutron multiplication factor

・Measure neutron multiplication factor (keff) with the reactor noise method before starting fueldebris retrieval work ①
This is the basis for the neutron source multiplication method.
・Measure neutron counting rate before and after fuel debris processing ②・Perform continuous monitoring of neutron counting rate during fuel debris retrieval processing
・Evaluate neutron multiplication factor (keff) with the neutron source multiplication method if
the neutron counting rate changes
・Stop work when the neutron multiplication factor reaches the judgment criteria
・Use an absorbent
・Measure neutron multiplication factor (keff) with the reactor noise method after using thePlannedabsorbent ①
Criticality approachprocessing
Fueldebris 加工予定場所
臨界近接監視器
燃料デブリ
・Determine work restart - restart work
monitoring
location
The aforementioned monitoring functions shall be used differently according to the situation in
臨界近接監視エリア
Monitoring area
the field, as indicated in sheets No.14-15.
Stopping of retrieval work
Continuous monitoring of→Use of absorbent
k=1.00 criticality
neutron flux during processing
Subcriticality measurement

燃料デブリ
Fuel
debrisMonitoring Subcriticality
未臨界度測定器
areameasurement

臨界連続監視器
Continuous
monitoringk-alarm

臨界連続監視エリア
Monitoring area

Measurement ②

Measurement ②

Measurement ①

Retrieval

Retrieval

Reactor noise
methodRetrieval work (time lapse)

Planned processing
Fuel燃料デブリ
debris 加工予定場所
location©International Research Institute for Nuclear DecommissioningNo.22

7. Implementation items
7.1 Development of on-site operating procedures
7.1.1 Operating procedures of criticality approach monitoringb. Study of procedures⑦ Basic approach of criticality monitoring based on organized environmental conditions of partial
submersion processing and submersion processing

Partial submersion processing⚫

⚫

Environmental conditions of partial submersion
processing and submersion processing, which are
prerequisites of criticality control, were organized.
Aerial criticality risk is small and does not require thesame kind of control as being underwater, but localized
attention such as puddles is necessary.

・Water table has not formed above fuel debris.
・There is possibility of puddle below fuel debris.
・Cooling water is free-flowing.
・Water drops are falling.
・Processing is performed while spraying wateraround the processing area to prevent dust
scattering.

Submersion
processing
・Water table has formed
above fuel debris.
・Processing is performed
underwater from start tocollection.

Sprayingof water
ダスト 飛散防to prevent dust
止用の散水scattering

Water
drops
水滴が落下しare falling.
ている

Possibility of puddle below
デブリの下に水溜まりの可能性
debrisFigure 1 Environmental conditions of partial submersion processing

Figure 2 Environmental conditions of submersion processing©International Research Institute for Nuclear DecommissioningNo.23

7. Implementation items
7.1 Development of on-site operating procedures
7.1.1 Operating procedures of criticality approach monitoringb. Study of procedures⑧ Organization of criticality risk insubmersion processing(Example of processing work) Laser gouging[Criticality risk assessment]⚫ Criticality evaluation result (Figure 1) indicates that criticalmass is around 40kg(*3), and it requires around 2 hoursof processing time to reach said amount.
⚫ Most chip powder settles out quickly (speed 40-80[cm/s])
even if they swirls up underwater, and critical mass will
not be reached.
⚫ There are hardly any criticality risks during processing, butattention is needed so that chip powder does notcollectively accumulate.
(*1) Development of Fundamental Technology of Retrieval of Fuel Debris and Internal
Structures FY2016 Final Report IRID(*2) Development of Fuel Debris Criticality Control Technology FY2014 Research ReportIRID(*3) Easing of restrictions is expected if knowledge of fuel debris properties isgained in the future.

State of laser gouging (figure on left) and simulated test body after
processing (figure on right)

Table 1 Particle size distribution of sediment (%)(*1)
気中－ガス
Aerial - gas

Underwater
- gas
水中－ガス

Underwater - water
水中－水流
flow

250μm
以上 500μm
Morethan 250μm,
less than 未満
500μm

13.0

12.5

3.9

More
than 500μm,
less than1mm
500μm
以上 1mm
未満

14.2

20.4

11.8

More
than以上1mm, 4mm
less than
4mm
1mm
未満

64.7

63.6

83.1

More
than以上
4mm
4mm

8.1

3.5

1.1

Water reflector
SUS reflector

Minimum criticalmass (kg)[Processing test result] Case of underwater to water flow
⚫ Processing speed
330g/min (maximum)⚫ Particle size distribution of chip powder (Table 1)
Particles greater than 1mm account for a large portion at
70-80%.

Minimum fuel composition per body

U/Pu concentration (g/L)

Figure 1 Evaluation of minimum critical mass whenfuel debris swirls up(*2)(*3)©International Research Institute for Nuclear Decommissioning7. Implementation items

No.24

7.1 Development of on-site operating procedures
7.1.1 Operating procedures of criticality approach monitoringb. Study of procedures⑧ Organization of criticality risk in submersion processing

Fuel debris processing method and overview of criticality evaluation(*1)
Fuel debris processing
Assumed criticality
methodscenario
Laser gouging
Cutter, wire saw, abrasive Multiple cracks form onwater jet (AWJ)
fuel debris and water
enters
ChiselCrushed and fragmented
fuel debris falls into waterCore boring

Water enters drilled holes

Illustration of events

When cracks change the volume by about
1vol% from the state of lump-like fuel debris
containing 20wt% structural material, reactivityof about 0.1%Δk(*2) is applied.
崩落による形状変化

加工による形状変化、水の浸入加工による形状変化、微粒子状
デブリの巻き上がり

Laser

Overview of criticality evaluation

Granular fuel debris cut
underwater swirls up

Laser
レーザ
ー

When falling and crushing change the volume
by about 10vol% from the state of lump-like fueldebris containing 20wt% structural material,
reactivity of about 1%Δk(*2) is applied.

When holes 7cm in diameter and 1m in depth
are drilled from the state of lump-like fuel debris
containing 20wt% structural material, reactivityof about 0.5%Δk is applied. Since processing
speed of core boring is around 1cm/min.,reactivity of about 0.005%Δk/min.(*2) is applied.
Since processing speed is slow and swirled up
fuel debris settles out before reaching criticalmass of 40kg(*2), criticality does not occur. (fromprevious sheet)(*1) FY2015 supplementary budget, Subsidy Project of Decommissioning and Contaminated Water Management, Development of Fuel Debris Criticality Control Technology, Research Report, March 2018, IRID(*2) Easing of restrictions is expected if knowledge of fuel debris properties is gained in the future.
©International Research Institute for Nuclear DecommissioningNo.25

7. Implementation items
7.1 Development of on-site operating procedures
7.1.1 Operating procedures of criticality approach monitoringb. Study of procedures⑧ Organization of criticality risk insubmersion processing⚫

Criticality risk of the processing method studied in the relevant project(*1) was
evaluated based on analysis.

Usage different from the originally intended
usage method

Attention is needed when
making multiple cuts at shortintervals (up to several cm)(*2).

Processing that could exceed the size of minimum
critical mass at optimum deceleration state
Attention is needed
when fuel debris
fine powder isswirled up in a wide
range underwater.
(Guideline: Morethan 46cm indiameter, morethan 71kg in
weight)(*2)

There is no criticality
risk in the work itself
that picks up small
pieces of fuel debris up
to 200mm (size thatcan go into storage
canister) underwater.
However, attention isneeded in work thatstirs and breaks up
small pieces of fueldebris with a robot
arm.

Attention isneeded when
dropping small
pieces of fueldebris andcollapsing thedeposited small
pieces of fueldebris.

Processing where injection reactivity could exceed 0.5%ΔkAttention is needed when drilling
multiple holes with a short pitch (upto around 100mm)(*2) between
holes by core boring.

Attention is needed when a large amount offuel debris fine powder deposits
underwater. (Guideline: More than 35cm inradius, more than 499kg in weight)(*2)(*1) Relevant project “Development of technology for further increasing the scale of retrieval of fuel debris and internal
structures” (FY2019, FY2020)

When crushing fuel debris
with a chisel, there is
criticality risk when theprocessing range exceeds
16×16×16cm(*2).
(*2) Easing of restrictions is expected if knowledge of fuel debris properties is gained in the future.

©International Research Institute for Nuclear DecommissioningNo.26

7. Implementation items
7.1 Development of on-site operating procedures
7.1.1 Operating procedures of criticality approachmonitoringb. Study of procedures

Table Necessity of criticality monitoring in each processing method
Category

Processing
methodCharacteristics / overview (*)⑧ Necessity of criticality monitoring in submersion processing
⚫

⚫
⚫

⚫

Necessity of criticality monitoring in each processing method wasorganized by partial submersion processing and submersion
processing. (However, criticality risks are not determined only by theprocessing method but largely depends on fuel debris properties.
Thus, need for monitoring does not necessarily mean criticality risks
are high.)
Even if criticality risks are not expected for the processing method
itself, attention is needed for secondary effects such as unexpectedusage and deposits in a location different from the processing point.
In the case of “Attention”, it shall be studied to perform criticality
approach monitoring or continuous monitoring in case of emergency,
but processing range limitations and subcriticality measurement may
be omitted by adopting appropriate processing procedures.
Since chisel processing instantaneously changes the form of fueldebris, it was determined that criticality monitoring is necessary. The
basis is processing range limitations and subcriticality measurement+ criticality approach monitoring + continuous monitoring, but
subcriticality measurement may be omitted if the condition of fueldebris is known.

Grabbing

Grab small pieces of fuel debris

Partialsubmersion
processing
No

BucketScoop up small pieces of fuel debris

No

Attention

SuctionSuctionSuction and collection of fuel debris fine powder

No

Attention

Cutting

Laser gouging Cutting of fuel debris surface with laser

No

Attention

Pick upCutting

Example of unexpected usage

Ultrasoniccore drill

Can process ceramics and metals at the same
time

Attention

Circular saw

Can process ceramics but tool damage issevere. Blade tip is jagged.

Attention

Cutting wheel

Can process ceramics and metals at the same
time

Attention

Hydraulic-driven scissors

Attention

Cutting with high-pressure water mixed withabrasive
Dispersion of cutting width to the surrounding
environment
Laser cutting
Thermal cutting by laser heat

Attention

AWJ

Crushing

Attention

Attention

HydrauliccutterDeposit in a locationdifferent from theprocessing point

Submersion
processing

Be careful if
・Punching process with hole saw
there are・Separate core cutting is necessary
・Water for discharging chips and cooling blades puddles
is necessary

Perforation Core boring

See explanationin previous pageNo criticality at theprocessing point

Necessity of criticality
monitoring

Hammer(chisel)Crushing by pressing the chisel against thetarget and striking blows with up-and-down
movement of piston

Attention

Yes

Example of secondary effect©International Research Institute for Nuclear Decommissioning(*) Relevant project “Development of technology for further increasing the scale of retrieval of fuel debris and internal structures” (FY2019, FY2020)

7. Implementation items

No.27

7.1 Development of on-site operating procedures
7.1.1 Operating procedures of criticality approach monitoringb. Study of procedures[Evaluation conditions]
Fuel debris composition: Concentration 4wt%
uranium(*2)
No structural material, FP, Gd⑨ Monitoring procedures for representative processing examples(laser processing)
Laser cutting, which is easy to understand, was used as arepresentative example to study the measurement procedures
during fuel debris processing. (*1)⚫

A condition of criticality approach due to a straight cutting line
forming on the fuel debris and water entering from there wasassumed.

⚫

Neutron multiplication factor of fuel debris before processing is nearkeff=0.95, and it is assumed that it barely does not exceed thejudgment criteria.

⚫

Measure neutron counting rate before processing.

⚫

When processing starts, reactivity of Δk=0.016(*2) is added in 10min.
of processing, and neutrons discharged from fuel debris increase by
1.5 times. (Figure)⚫

Measure neutron counting rate after processing. It is observedthat the neutron counting rate increased by 1.5 times, and it isestimated with the neutron source multiplication method that theneutron source multiplication factor changed from 0.95 to 0.967.

⚫

It is judged that the neutron multiplication factor exceeded thejudgment criteria, and retrieval work is stopped. Since increase ofneutron counting rate stops almost at the same time as whenprocessing ends (Figure), operators can make decisions with plentyof time.

(*1) There is no risk of criticality with one cutting line, but extreme conditions were set as aneasy-to-understand example.
(*2) Easing of restrictions is expected if knowledge of fuel debris properties isgained in the future.

Water is in the pores of the fuel debris, andkeff=0.95 of the initial condition before
processing.
Processing to insert a cutting line 2cm wide, up
to 60cm deep and up to 60cm long is assumed.
(*1)

Processing speed of laser cutting is 1[mm/sec],
and 10min. of processing is performed.
Effective multiplication factor was calculated with
the MVP code, and neutron response wascalculated with single-point reactor kinetics
analysis.
Neutron source strength 1[n/sec]
Neutron detector detection efficiency 100 [%]Criticality approach monitoring
before and after processing
Neutron counting rate [1/s]⚫

1.5 times

Processed length, effective
multiplication factor

Processing time [s]Figure Simulation of neutron response assuming laser cutting©International Research Institute for Nuclear Decommissioning7. Implementation items

No.28

7.1 Development of on-site operating procedures
7.1.1 Operating procedures of criticality approach monitoringb. Study of procedures⑩ Maintenance of monitoring system⚫ Approach on daily inspection, periodic
inspection and corrective maintenance wasorganized for maintenance of the criticality
approach monitoring system (Table).
⚫ In daily inspection, the detector unit is moved tothe maintenance area (assuming SFP, etc.(*)) in
the building(Note 1) to perform remote
operation(Note 2). (Figure)⚫ In some periodic inspection and corrective
maintenance when defects occur, the unit isdecontaminated and transferred outside thebuilding and worked on by workers.

Neutrondetector unit

Sliding lid
Neutronsource moving
equipment

NeutronsourceSource holder
elevator

Table Maintenance details of criticality approach monitoring system

Figure Maintenance area inside building

PurposeFrequency

Inspection
description

Inspection items

MethodPreventive
maintenanceDaily
inspection

System integrity,
noise impact

Confirmation ofoperation withneutron source,
backgroundmeasurement

Move the detector unit to the remote maintenance area, and perform indirect
maintenance of fuel debris processing and collection equipment in remote operation,using the neutron source equipment installed in said area. (required time is around 1
hour)
Replacement is necessary as neutron source itself also attenuates.

Periodic
inspection
1-2 times per
year

Integrity ofdetector /connector /cable, detector
sensitivity

Discrete plateau,insulation resistance,
capacitance, cablecharacteristic, source
calibration

If remote maintenance in the maintenance area is difficult, remove the detector unit
from the fuel debris processing and collection equipment, take it out of PCV, andperform direct maintenance after decontaminating it. (required time is around several
days)

Irregular

Disassembly
and inspection,
repair,replacement

Gas leak, decreased
insulation, poor
contact,
disconnection, circuit
diagnosis, etc.

If the system experiences defects and the detector cannot be restored inside PCV,
remove the detector unit from the fuel debris processing and collection equipment,take it out of PCV, and perform inspection with direct maintenance after
decontaminating it. (more than 1 week if disassembly and assembly are required)(Note 1) Moving the fuel debris
retrieval equipment loaded withthe detector unit to themaintenance area and putting it
on standby is a routine operation.
(Note 2) Method of remote
operation of neutron detector
calibration using neutron source
needs to be developed in theengineering stage.

Neutron source is exposed from the source moving
equipment to confirm operation of the neutron
detector

Corrective
maintenance©International Research Institute for Nuclear Decommissioning(*) Relevant project “Development of technology for further increasing the scale of retrieval of fuel debris and internal structures” (FY2019, FY2020)

No.29

7. Implementation items
7.1 Development of on-site operating procedures
7.1.1 Operating procedures of criticality approach monitoringd. Summary[Result]✓ Regarding criticality approach monitoring during fuel debris retrieval,

image of on-site operation of items, such as monitoring range ofneutron detector, measurement procedures, criticality risk during
processing and system maintenance, was materialized, andinformation was provided for the fuel debris retrieval method team toutilize it to study the retrieval system design and throughput.

©International Research Institute for Nuclear DecommissioningNo.307. Implementation items
7.1 Development of on-site operating procedures
7.1.2 Confirmation of applicability to subcriticality measurement of neutron detector[Issues]Three types of neutron detector (corona discharge type, SiC semiconductor type, multicell He-3 type),
which are candidates from the viewpoint of subcriticality measurement, are developed in a relevant
project(*) (Figure 1). In order to confirm that these detectors can be applied in the field, it is necessary toverify that subcriticality can be measured and evaluate the performance of criticality approach monitoring.
[Implementation items]
Structure a subcriticality system simulating fuel debris with Kyoto University Critical Assembly (KUCA),
and perform subcriticality measurement test using the developed neutron detector. (Figure 2)
Additionally, evaluate neutron
measurement performance undergamma ray environment through tests.
Preamplifier[Goals]⚫ Evaluation of subcriticality
measurement performance ofneutron detector⚫ Incorporation into operatingprocedures of neutron detector
assuming the field

Corona detector
Φ25.4×260mm
Figure 1 Appearance of B-10 corona
discharge type detector

Figure 2 Kyoto University Critical Assembly(*) Relevant project “Development of technology for further increasing the scale of retrieval of fuel debris and internal structures” (FY2019, FY2020)©International Research Institute for Nuclear DecommissioningNo.31

7. Implementation items
7.1 Development of on-site operating procedures
7.1.2 Confirmation of applicability to subcriticality measurement of neutron detectora. Test plan
Table Summary of KUCA tests conducted in relevant projects
up to previous PJ and test details of this project
Test 6

Neutron detector

PurposeResultKUCA test(1st time, 2nd time)

B-10 proportionalcounter

Confirm that subcriticality can be measured
in a small system using uranium fuel・Measurement error (2%-10%) of different
subcriticality, from near criticality of neutron
multiplication factor =0.95 to deep subcriticality of 0.7
・Impact on measurement error due to uncertainty(neutron energy spectrum) of mixing state of fuel debris
and water is small・Measurable distance between fuel debris anddetector is within 20cm underwater and within 35cm inair

KUCA test(3rd time)

B-10 proportionalcounter

Confirm that subcriticality can be measured
in a large (uniform) system using uranium
fuel

Confirm that local subcriticality near the detector canbe monitored (monitoring of the entire system is difficult)

KUCA test(4th time)

B-10 proportionalcounter

Confirm impact of large and uniform system
and arrangement of neutron absorbents

Subcriticality was measured with a system simulating
uneven fuel debris, and positional dependence wasconfirmed.
Test was conducted in a condition simulating surface
application of non-soluble neutron absorbent, and it was
confirmed that subcriticality measurement can be
performed. (-15～+14%)

KUCA test(5th time)

Corona discharge type
SiC semiconductor
type
Multicell He-3 type

Confirm that alternative candidates of theneutron detector can measure criticality

Implemented June 28 - July 9, 2021

Test target this time
©International Research Institute for Nuclear DecommissioningNo.32

7. Implementation items
7.1 Development of on-site operating procedures
7.1.2 Confirmation of applicability to subcriticality measurement of neutron detectora. Test plan
Table Candidate neutron detectors for subcriticality measurement
No.

Detector name1

Fission ionization chamber

2

Improved small-size B-10
proportional counter3

4B-10 proportional counter

Test implemented up to previous PJ
He-3 proportional counter(Multicell-type He-3 proportionalcounter)

Supplier

Category

PhotonisFissionionizationchamber

IRID／
Hitachi-GE
CETD

Gas
detector
CETD

Remarks

Neutron detector considered to be adopted in PCVinternal investigation(studied in relevant project (*))

Neutron detectors being studied in criticality
PJ(confirmed to be usable for subcriticality
measurement with the Feynman−α method)

Test target this time5

B-10 corona discharge type detector

RosRAO6

He-3 corona discharge typedetector

RosRAO

7

SiC semiconductor detector

IPL

8CMOS detector

IRID／
Hitachi-GE

Semiconduc
tor detector

Neutron detector considered to be adopted in PCVinternal investigation(neutron detector with radiation resistance, studied
in relevant project*)©International Research Institute for Nuclear Decommissioning(*) Relevant project “Project on Development of Fundamental Technology of Retrieval of Fuel Debris and Internal Structures (Development of Small-size Neutron Detector)”

No.33

7. Implementation items
7.1 Development of on-site operating procedures
7.1.2 Confirmation of applicability to subcriticality measurement of neutron detectorb. Test preparation and arrangement

Appearance of neutron detectors that underwent performance assessment in this KUCA test
Preamplifier

Corona detector
Φ25.4, ℓ 260mm

Corona discharge type neutron detector (B-10/He-3)

50×50× 540 mm
SiC semiconductor neutron detector

Sensitive
length 150mm

Multicell-type He-3 proportional counterΦ25.4mmConventional B-10 proportional counter(for reference; not used in this test)©International Research Institute for Nuclear DecommissioningNo.34

7. Implementation items
7.1 Development of on-site operating procedures
7.1.2 Confirmation of applicability to subcriticality measurement of neutron detector
c. Element test
KUCA test system in this project⚫ Test reactor is composed of fuel bodies combining enriched uranium and polyethylene (solid moderator)⚫ The neutron detector to be tested is arranged adjacent to the fuel bodies, and analysis is conducted with
the Feynman−α method by collecting time series data of neutron counts
1
り 9
ぬ 9
る 9
を 9

2
8
9
6
6

3
8
9
6
6

4
8
9
6
6

5
8
9
6
6

6
8
9
6
6

7
8
9
6
6

8
8
9
6
6

9 10 11 128 8 8 89 9 9 96 6 6 66 6 6 613 14 15 16 17
8 8 8 8 8
9 9 9 9 9
6 8 8 6 6
6 8 8 6 6

わ 9
か 9
よ 9
た 9
れ 9

6
6
6
6
6

6
6
6
6
6

6
6
6
6
6

8
8
6
6
6

8
8
6
6
6

6
6
6
6
6

6
6
6
6
6

6
6
6
6
6

6
7
7
7
7

6
7
7
7
7

6
7
7
7
D

6
7
7
7
7

6
7
7
7
7

6
7
7
7
1

6
7
7
7
7

6
7
7
7
7

Explanation
説明18 19 20 21 22 23 ID
24 25
26 27 28 29
8 8 8 8 8 8 18 3/8"p32EU燃料体
8 8 8 8 fuel9 り
3/8”p32EUbody
9 9 9 9 9 9 29 1/8"p72EU燃料体
9
9
9
9
9
ぬ
1/8”p72EU fuel body
6 6 6 6 6 6 6 6 6 6 6 9 る
3 1/8"p54EUEU燃料体
1/8”p54EUEU fuel body
6 6 6 6 6 6 6 6 6 6 6 9 を
4 3/8"p10EU燃料体
3/8”p10EU fuel body
6 6 6 6 6 6 8 8 6 6 6 9 わ
5 吸収材サンプルセル
Absorbent sample cell
7 7 7 6 6 6 8 8 6 6 6 9 か
6旧PEブロックセル
Former7 7 7 6 6 6 6 6 6 6 PE
6 block
9 よ cellNew
7 7 7 6 6 6 76 新PEブロックセル
6 6 PE
6 block
6 9 cell
た
E 7 7 6 6 6 86 ボイド,
6 6 KUCA
6 KUCA計装セル
6 instrumentation
9 れ
Void,
cell

そ 9
つ 9
ね 9
な 9
ら 9
む 9

6
6
6
6
6
6

6
6
6
6
6
6

6
6
6
6
8
8

6
6
6
6
8
8

6
6
6
6
6
6

6
6
6
6
6
6

6
6
6
6
6
6

6
6
6
6
6
6

7
7
7
7
7
7

7
7
7
7
K
K

7
E
7
7
K
K

1
1
1
C
7
7

1
1
1
1
F
7

1 1
1 1
1 1
1 1
G H
7 7

1
1
1
E
7
7

7
B
7
7
7
7

う 9
ゐ 9
の 9
お 9
く 9
や 8
1

6
6
6
6
9
9
2

6
6
6
6
6
9
3

6
6
6
6
6
9
4

6
6
6
6
6
9
5

6
6
6
6
6
9
6

6
6
6
6
6
9
7

6
6
6
6
6
9
8

6 7 7 76 6 6 66 6 6 66 6 6 66 6 6 69 9 9 99 10 11 12

7 7 7 7 7
6 6 6 6 6
6 6 6 6 6
6 6 6 6 6
6 6 6 6 69 9 9 9 913 14 15 16 17

7 7 6
8 10 6
7 7 6
7 7 6
7 7 6
7 7 6

6
6
6
6
6
6

6
6
6
6
6
6

6 黒鉛セル6 6 cell
6 6 9 そ
9
Lead
6
6
6
6rod
6 full9 つ
A Control
制御棒フルストロークstroke6 6 6 6 6 9 ね
C1 rod cell (adjusting rod)
B C1ロッドセル(調整棒)
6 6 6 6 6 9 な
C
C2
rod)
6 C2ロッドセル(調整棒)
6 6rod6cell
6 (adjusting
9 ら
6 C3ロッドセル(調整棒)6 6
6 (adjusting
9 むDC3rod6cellrod)

Safety(S4-S5)7 7 7 6 6 6 E6 安全棒セル(S4～S5)
6 6 6rod6 cell
9 う
6 6 6 6 6 6 F6 He-3
6 6 corona
6 6 9counterゐHe-3
コロナカウンタ6 6 6 6 6 6 G
6 IPL製SiC検出器
6 6SiC
6 detector
6 9 の
IPL
6 6 6 6 6 6 6 6 6 6 6 9 お
H B-10
B-10 corona
コロナカウンタ
counter
6 6 6 6 6 6 6 6 6 6 9 9 く
CETD
multicell
He-3
K
CETD製
マルチセルHe-3
9 9 9 9 9 9 9 9 9 9 9 8 やAmBesourceAmBe線源18 19 20 21 22 23 10
24 25
26 27
28 29

Layout of KUCA test core and detector (keff=0.95)

Centralframe
Control rod

Am-BesourceFissionChamber
CETD
multicell
He-3

B-10 coronacounter
SiC detector
He-3 coronacounterState of actual test
©International Research Institute for Nuclear Decommissioning7. Implementation items

No.35

7.1 Development of on-site operating procedures
7.1.2 Confirmation of applicability to subcriticality
measurement of neutron detector
c. Element test
Evaluation procedures of subcriticality measurement(Feynman−α method)

B-10 corona detector
Keff=0.95
60min. measurement

NeutronMeasurement time [s]FissionPlot variance-to-mean ratio (Y value) 1 − exp(− T ) 
Y (T ) = Y1 −

T



Image of data to be acquired (time list data)

B-10 corona detector
Keff=0.95
60min. measurement

Derive α (prompt neutron attenuation constant) by fitting (red line)with the theoretical equation.

𝑘𝑒𝑓𝑓 =

1 − 𝛼ℓ1−𝛽α: Prompt neutron attenuation constant(1/sec)β: Delayed neutron percentageℓ : Prompt neutron life (sec)𝒌𝒆𝒇𝒇 : Neutron multiplication factor
ΔT: Gate time width of neutron count (sec)
Y: Variance-to-mean ratio of neutron count

Derive keff with the measured α, and l and β calculated in advance.

Time interval (sec)Frequency distribution of time list data

keff obtained above and keff (reference solution)
calculated with the analysis code were compared.
©International Research Institute for Nuclear DecommissioningNo.36

7. Implementation items
7.1 Development of on-site operating procedures
7.1.2 Confirmation of applicability to subcriticality measurement of neutron detector
c. Element test
This KUCA test result (corona discharge type, multicell type)⚫ It was confirmed that the corona discharge type and multicell He-3 type detectors can measure subcriticality withthe same level of precision as the conventional B-10 proportional counter (Figure, Table).
⚫ With the multicell He-3 type detector, event where operation became unstable due to changes in count indication
value was confirmed.
B-10
proportionalcounterHe-3 corona
discharge
type
detector

B-10 corona
discharge
type
detector

Multicelltype He-3
proportionalcounter

Difference
fromreferencevalue+0.05%dk-0.39%dk-0.80%dk+0.08%dk

Uncertainty±0.09%dk±0.03%dk±0.12%dk±0.32%dk

Measurement
time(min.)

30, 3010, 10, 12, 33,
60

10, 10, 12, 30,
60

10Number of
data

2

5

5

1

1.2
Multicell-typeHe-3 proportional counter
マルチセル型He-3比例計数管

B-10コロナ検出器
B-10
corona detector

1.1

value)
Keff (measured
keff (測定値)

He-3
corona detector
He-3コロナ検出器

1.0

0.9

0.8

0.7

0.6
0.6

0.7

0.8
0.9
1.0
Keff (reference
value)
keff (参照値)

1.1

1.2

Table Measurement result of each detector (keff=0.95)

Figure Difference between measured value and reference value(corona discharge type, multicell type)
Since keff calculation of the KUCA test system using the MVP code tends to overestimate by around 0.5%dk at critical conditions, this is calibrated©International Research Institute for Nuclear DecommissioningNo.37

7. Implementation items
7.1 Development of on-site operating procedures
7.1.2 Confirmation of applicability to subcriticality measurement of neutron detector
c. Element test
This KUCA test result (SiC semiconductor type)⚫ Data obtained with the SiC semiconductor detector is processed with the Feynman−α method, and subcriticality
was evaluated. As the following Figure and Table show, it was confirmed that subcriticality can be measured with adifference of around 1%dk. If subcriticality is deep, difference from the reference value increased to around 4.8%dk,
but it was confirmed that subcriticality can be measured with the same level of precision as the conventional B-10
proportional counter.
KUCA subcriticality measurement result of SIC detector

Table Measurement result of each test system
Test core

Keff (measured value)

Case1-1(keff=0.95)
Case1-2(keff=0.97)
Case1-3(keff=0.98)
Case1-4(keff=0.70)

Referencevalue

Measuredvalue

Difference
(%dk)

0.948

0.944

-0.4

0.962

0.958

-0.4

0.979

0.968

-1.1

0.979

0.958

-2.1

0.979

0.981

0.2

0.681

0.729

4.8

Remarks

Measurementchanging
pulse
discrimination
parameters

Keff (reference value)

Reference value: Core analysis value by MVP

Figure Difference between measured value and reference value(SiC semiconductor neutron detector)©International Research Institute for Nuclear DecommissioningNo.38

7. Implementation items
7.1 Development of on-site operating procedures
7.1.2 Confirmation of applicability to subcriticality measurement of neutron detector
c. Element test
Overview of gamma-ray resistance test
[Test purpose]⚫ Check the upper limit of neutron-discriminable gamma dose
rate of the candidate neutron detector for subcriticality
measurement, to see up to how much strength of gamma dose
rate neutrons can be discriminated.
⚫ Check if measurement is possible with the Feynman−α method
in high gamma ray environment.

State of actual test

[Test location]
Nagoya University Co-60 irradiation room
[Test target]⚫ B-10 corona discharge typedetector⚫ He-3 corona discharge typedetector⚫ Multicell-type He-3 proportionalcounter⚫ B-10 proportional counterPb shielding
Evaluation of Pb
shielding thickness

PEmoderator

Co-60
sourceNeutron source

Neutron detector

Other neutron detectors have been
separately evaluated with vendor
tests, etc.

Co-60 source - neutron detector
Adjustment of irradiation dose bychanging the distance

Test system©International Research Institute for Nuclear DecommissioningNo.39

7. Implementation items
7.1 Development of on-site operating procedures
7.1.2 Confirmation of applicability to subcriticality measurement of neutron detector
c. Element test
Example of gamma-ray resistance test result(Wave height discrimination characteristics of B-10 corona discharge type detector)

Range where count fromgamma ray is significant

0Gy/h, no neutron
0Gy/h, with neutron
0.6Gy/h, with neutron
1.2Gy/h, with neutron
5.0Gy/h, with neutron
8.0Gy/h, with neutron
9.9Gy/h, with neutron
28.8Gy/h, with neutron⚫

600mV is set as the discretelevel that can separate gamma
ray and neutrons, based on themeasurement result in the figureon the left.

⚫

Decrease in neutron detection
sensitivity due to gamma ray
was 47% at 10Gy/h.

⚫

It was checked whether
Feynman−α measurement ispossible with this detector
sensitivity. (next page)

Range where neutrons are counted

Pulse wave height discrimination level of measurement circuit©International Research Institute for Nuclear Decommissioning7. Implementation items

No.40

7.1 Development of on-site operating procedures
7.1.2 Confirmation of applicability to subcriticality measurement of neutron detector
c. Element test
Gamma-ray resistance test result
1.2E-01

1.0E-01

8.0E-02

Y値
Y value

2.0E-02

0.0E+00
0.0E+00

γ ray
absorbed
dose rate[Gy/h]

0Multicell-type
He-3
proportionalcounterーHe-3 corona
discharge
type detector

B-10 corona
discharge
type detector

Y value: Reference
value==0.030
Y値：参照値
0.030

6.0E-02

4.0E-02

5.0E-03

B-10
proportionalcounter(182064)

0.5
1

○

○

(63948)(10599)○○(60396)(33043)×○(1697)(21944)

1.5E-022.0E-02

3.0E-022.0E-02(Note 1)

1.0E-02
ゲート幅
Gate
width(sec)
(sec)

Y値
Y value⚫
⚫

It is determined to be measurable if Y curve is observed.
Measurement is possible without shielding, up to 0.5Gy/h for the He-3 corona type and up to
10Gy/h for the B-10 corona typeIt was measured up to 30Gy/h with the conventional B-10 proportional counter
It was unmeasurable with the multicell-type He-3 proportional counter due to unstable operation

Y value: Reference
value =0.004
Y値：参照値= 0.004

1.0E-02

0.0E+00-1.0E-02
0.0E+00

5.0E-03

1.0E-02

1.5E-022.0E-02

ゲート幅
Gate
width(sec)(sec)

3.0E-025ー(Note 1)(9164)○○(228973)(13634)

2.5E-022.0E-0210△○(2604)(11876)

Y値
Y value⚫
⚫

Y value: Reference
value =0.006
Y値：参照値= 0.006

1.5E-02

1.0E-0230ー(Note 1)(74656)○

5.0E-03(16792)0.0E+00
0.0E+00

5.0E-03

1.0E-02

1.5E-022.0E-02

ゲート幅 (sec)

Number in parenthesis is the total count; no lead shielding©International Research Institute for Nuclear Decommissioning(Note 1) Unmeasurable unless operation is stabilized at high voltage

7. Implementation items
7.1 Development of on-site operating procedures
7.1.2 Confirmation of applicability to subcriticality measurement of neutron detector
c. Element test

No.41

Gamma-ray resistance performance verification test of SiC semiconductor detector (1/2)

Test details① Confirmation of γ ray removal
Conducted at NPL (National PhysicalLaboratory) which owns equipment that
can irradiate 60Co of 1.5×1013Bq. (seebelow figure)
Data of irradiation up to 100Gy/h wascollected, and pulse form of noise
attributable to neutrons and γ ray wasevaluated.

Verification items andjudgment criteria

ResultPulse discrimination that candiscriminate γ ray andneutrons can be set

It was confirmed with parameters such as pulse height,
width and rise time that pulse discrimination of γ ray andneutrons is possible.
As a result, the probability of erroneously recognizing γ ray
as neutron is around 2%, which indicate that there are no
problems with using the detector in the field.

Theratron

SiC detector

60Co source

behind shield©International Research Institute for Nuclear Decommissioning7. Implementation items
7.1 Development of on-site operating procedures
7.1.2 Confirmation of applicability to subcriticality measurement of neutron detector
c. Element test

No.42

Gamma-ray resistance performance verification test of SiC semiconductor detector (2/2)
Test details② Impact of cumulative dose of γ ray
Conducted at DCF (Dalton Cumbria
Facility of Univ. of Manchester) which owns
equipment that can irradiate 60Co up to
6540Gy/h.
Test board including the sensor chip andpreamplifier was irradiated. (installedinside chamber in below picture)
Cumulative dose in the usable area of the
facility is confirmed.

Verification items andResultjudgment criteria
Confirm that the detector
It was confirmed that the detector properly operates after
operates without any problem irradiation of total cumulative dose of 63kGy. (waveform
after irradiation.
was confirmed with α source, count rate was confirmed)
Trend of significant increase of leakage current was notconfirmed even after irradiation, and it was determined thatthe detector could operate without problem evenconsidering extrapolating such increase to 1MGy.
γ ray irradiation test result at DCF

Dose: Dose rate at irradiation,
Duration: Irradiation time,
Test Dose: Cumulative dose of each test
Accumulated Dose: Total cumulative dose
©International Research Institute for Nuclear DecommissioningNote) This test is a verification test conducted with PCV detailed inspection detector using the same SiC semiconductor detector.

No.43

7. Implementation items
7.1 Development of on-site operating procedures
7.1.2 Confirmation of applicability to subcriticality measurement of neutron detectord. Summary
Performance assessment (1/3)

Applicability to field environment

Category〇: Applicable, △: Conditionally applicable, ×: Not applicable

Item

Atmosphereenvironment

Requirement
specification

Underwateror 100%humidity

Fissionionizationchamber

Improved
B-10

B-10
proportionalcounter

Multicell He-3
proportionalcounter○
Secured bywater-sealing
case○
Secured bywater-sealing
case

B-10
coronaHe-3 corona

SiC detector

CMOS
detector○
IP68*1○
IP68*1○
IP68*1○
IP68*1○100 Gy/h○1000 Gy/h○
IP68*1○
IP68*1○650 Gy/h〇
30 Gy/h*3×*3(30 Gy/h*4)○
10 Gy/h*3(700 Gy/h*5)〇
Same asabove△
Respond withshielding (1cmthickness)

×*3△
Respond withshielding (2cm
thickness)△
Respond withshielding (4.5cm
thickness)〇
Same asabove〇
Same asabove○
1MGy*7△*8
0.5 MGy〇
0.64 MGy*9

No data
Estimated as thesame level as B-10
corona

○1 MGy *10

×0.013 MGy

○
50 ℃

○
50 ℃

○
50 ℃

○
50 ℃

○
50 ℃

△
40 ℃

Gamma raydose rate(averagevalue)

10 Gy/h○
1×104 Gy/h*2

Gamma raydose rate(maximumvalue)

100 Gy/h〇Same asabove

Cumulative
dose(usageperiod)More than
0.6MGy(8 months)○
1000 MGy

Atmospheretemperature

50 ℃

○
300 ℃

○
＞1MGy
(including
preamplifier)*6

○
100 ℃

△
0.5 Gy/h*3
Respond withshielding (2.5cm
thickness)*) See separate page for notes©International Research Institute for Nuclear Decommissioning7. Implementation items

No.44

Applicability to subcriticality measurement

Category

7.1 Development of on-site operating procedures
7.1.2 Confirmation of applicability to subcriticality measurement of neutron detectord. Summary〇: Applicable, △: Conditionally applicable, ×: Not applicable
Performance assessment (2/3)
Item

Requirement
specification

Fissionionizationchamber

Improved
B-10

B-10
proportionalcounter

Multicell He-3
proportionalcounter

B-10
coronaHe-3 corona

SiC
detector

CMOS
detector

Thermalneutronsensitivity*1

10 cps/nv

×(2.5 units)*12(5265cm3)*11〇(50 units)(192cm3)*11〇(6 units)(730cm3)*11〇(0.5 units)(642cm3)*11〇(6 units)(693cm3)*11〇(1 unit)(120cm3)*11〇(-)*13×(1250 units)*11(15708cm3)(Thermal neutron
sensitivity of single alone)

4 cps/nv*14@100Gy/h

0.2 cps/nv@100Gy/h

1.7 cps/nv@0Gy/h
0.8 cps/nv@28.8Gy/h*3

21 cps/nv@0Gy/h*4
0.2 cps/nv@0Gy/h*3

1.9 cps/nv@0 Gy/h
0.6 cps/nv@8 Gy/h *3

11 cps/nv@0 Gy/h
7.2 cps/nv@0.6 Gy/h*3

0.05cps/nv/c
hip@70Gy/h*13

0.008 cps/nv@100Gy/h(Size of detector alone)

Φ80mm
×419mm*14

Φ7mm
×100mmΦ25.4mm×245mm

Φ76mm
×283mmΦ25.4mm×260mmΦ25.4mm×260mm

－*13

Φ20mm
×40mm

〇

〇
1000 nsec(dependent on
preamplifier
method)○Below 400
nsec〇
Below 800 nsec〇
2000 nsec〇
1800 nsec〇*154000 nsec-○
100 nsec〇
100 nsec〇
100 nsec〇
100 nsec〇
100 nsec-○
10 nsec〇
10 nsec〇
10 nsec〇
10 nsec〇
10 nsec-

Pulse width(detector)

Below
2000 nsec

Identifiableadjacent pulseinterval (circuit
system)Below 100
nsec

〇

〇
1000 nsec(dependent on
preamplifier
method)Variation inpulse output
delay time(circuit system)

Below 10
ns

〇

To be
confirmed*) See separate page for notes©International Research Institute for Nuclear Decommissioning7. Implementation items

No.45

7.1 Development of on-site operating procedures
7.1.2 Confirmation of applicability to subcriticality measurement of neutron detectord. Summary
Performance assessment (3/3)

Applicability to remotehandling

Category〇: Applicable, △: Conditionally applicable, ×: Not applicable

Requirement
specification

Fissionionizationchamber

Improved
B-10

B-10
proportionalcounter

Multicell He-3
proportionalcounter

B-10
coronaHe-3
coronaSiC
detector

CMOS
detector

Unit
size

Less than
Φ700
mm×L1000mm〇
5265 *16cm3〇
192 *16cm3〇730 *16cm3〇642 *16cm3〇693 *16cm3〇
120 *16
cm3

〇*17

〇
15708 *16cm3

Unitweight50-100 kg○
Shieldingunrequired〇
Shieldingunrequired△
1cm
shieldingー△
2cmshielding×
4.5cm
shielding

Not applicable(due to low
thermal neutron
sensitivity)

Applicablefor
subcriticality
measurementin the field at
1F

Applicable forsubcriticality
measurementin the field at
1F

Evaluated as
not applicablein the current
situation(due to unstable
operation)

Applicablefor
subcriticality
measurement in the field
at 1F

Not
applicable(due toshieldingthicknessbeing toolarge)

Item

General evaluation*) See separate page for notes○
Shieldingunrequired*18

Applicablefor
subcriticality
measurementin the field at
1F

-

Not
applicable(due to low
thermalneutronsensitivity andlow cumulativedose)

It is evaluated as “Applicable” with 〇, △ and if there is no ×©International Research Institute for Nuclear DecommissioningNo.46

7. Implementation items
7.1 Development of on-site operating procedures
7.1.2 Confirmation of applicability to subcriticality measurement of neutron detectord. Summary
List of notes for the performance assessment*1 Indicator for waterproof performance in JIS/IES standards. IP68 indicates that it can continue to be used underwater.
*2 MIRION CFUG08*3 Measured value of gamma-ray resistance test
*4 Vendor evaluation*5 Not verified for gamma ray of fuel debris in Digital Signal Processing mode in the vendor evaluation*6 Irradiation test was conducted with dose at the preamplifier position at 2.3kGy/h using 60Co source. The results showed that it can be measurable at cumulative dose of 1MGy.
See FY2018 supplementary budget “Development of technology for further increasing the scale of retrieval of fuel debris and internal structures” Final Briefing Material No.840.
*7 Evaluation value of only the sensor, since the preamplifier is installed inside PCV with shielding. It is a value based on model specifications of the same type, and irradiation has
not been performed on the subject prototype.
*8 Evaluation value of only the sensor, since the preamplifier is installed inside PCV with shielding. It is a value based on model specifications of the same type, and irradiation has
not been performed on the subject prototype.
*9 Evaluation value of only the sensor, since the preamplifier is installed outside PCV without shielding. It is a value based on the irradiation test result of a prototype of the same
type in the relevant project “Project on Development of Fundamental Technology of Retrieval of Fuel Debris and Internal Structures (Development of Small-size Neutron
Detector)”, and irradiation has not been performed on the subject prototype.
*10 Vendor evaluation (evaluation from the result of factory inspection). Irradiation of 60Co up to 6540Gy/h was performed with irradiation equipment on the test board including
the sensor chip and preampliﬁer (no additional shielding).
Usability from change in behavior after irradiation of total cumulative dose of 63kGy to 1MGy was confirmed.
*11 In order to compare achievable sensitivity when assembling the detector unit, volume for the number of detectors that yield 10cps/nv is indicated.
*12 Size-up may be subject to nuclear material regulatory restrictions*13 Multiple sensor chips make up the detector, but sensitivity of each chip changes depending on composition. Typical values of the detector used in the KUCA test are indicatedhere.The total unit volume including shielding and moderator to obtain 10cps/nv is expected to be around 4000cm3.
*14 MIRION CFUG08*15 Since it is configuration where multiple detectors independently conduct measurement and superimposed pulses can be separated by pulse waveform analysis, pulse width
size does not have an adverse effect such as count omission.
*16 Shielding thickness, and moderator polyethylene size in the case of partial submersion processing are added, but requirement specifications are evaluated to be satisfied evenwhen taking such into account*17 In order to achieve 10cps/nv, total unit volume including shielding and moderator is expected to be around 4000cm3.
*18 Shielding is applied to the preamplifier as necessary, but the unit weight is expected to be less than 20kg even when taking such into account.

©International Research Institute for Nuclear DecommissioningNo.47

7. Implementation items
7.1 Development of on-site operating procedures
7.1.2 Confirmation of applicability to subcriticality measurement of neutron detectord. Summary[Result]✓ Subcriticality measurement test was performed with KUCA on the three types of new neutron detectors(corona discharge type, SiC semiconductor type, multicell He-3 type), and it was confirmed that subcriticality
can be measured with the same level of precision as the conventional B-10 proportional counter.

✓ Neutron measurement test was performed with the corona discharge type and multicell He-3 type in gamma
ray environment, and the upper limit of neutron-discriminable gamma dose rate was confirmed. Additionally, it
was confirmed that measurement is possible with the Feynman−α method in high gamma ray environment.
✓ Candidate neutron detectors to be used for subcriticality measurement in the field of fuel debris retrieval at

1F were tested and evaluated from the viewpoints of applicability to on-site environment, applicability to
subcriticality measurement and applicability to remote handling, and the following was selected.

➢ B-10 proportional counter (conventional / improved small-size)➢ B-10 corona discharge type
➢ SiC semiconductor type

©International Research Institute for Nuclear DecommissioningNo.48

7. Implementation items
7.1 Development of on-site operating procedures
7.1.3 Operating procedures for neutron absorbents[Issues]
Method of injecting non-soluble neutron absorbents to fuel debris is studied in a relevant project(*). (Figure 1) In
order to discuss operating procedures and throughput with the fuel debris retrieval method team, it is necessary tomaterialize the procedures for neutron absorbent spraying and injection equipment operation.

[Implementation items]⚫ Procedures to spray non-soluble neutron absorbents to fuel debris shall beorganized in a step diagram for the fuel debris retrieval method and processing
and collection equipment studied in the relevant project (*).
⚫ Conditions of application to fuel debris processing scene, use of different
absorbents, effectiveness of pre-injection, and downsizing and weight reductionof injection equipment shall be studied.
⚫ Operability and feasibility of overall procedures shall be evaluated along with theoperating procedures of criticality approach monitoring (7.1.1).

Glass material
containing B・Gd

Gd2O3
particle

Water glass /Gd2O3 granulated
powderExample of injection
equipment for solids(powder)[Goals]⚫ Develop on-site operating procedures for non-soluble neutron absorbents, andprovide information to be utilized by the fuel debris retrieval method team fordesign of the retrieval system and study of throughput.
Figure 1 Drawing of injection of absorbents to fuel debris(*) Relevant project “Development of technology for further increasing the scale of retrieval of fuel debris and internal structures” (FY2019, FY2020)©International Research Institute for Nuclear DecommissioningNo.49

7. Implementation items
7.1 Development of on-site operating procedures
7.1.3 Operating procedures for neutron absorbentsa. Organization of prerequisites
Method of use of different absorbents••

Various forms of fuel debris is assumed, such as rod-like and granular. Policies to use the optimum type of neutron absorbent for
the respective fuel debris form have been established.
Non-soluble neutron absorbents are divided between a solid type and a type that solidifies from liquid to solid over time (waterglass type). Since the water glass type has fluidity and viscosity immediately after injection, it is effective when residual fuel(stump fuel) is forested vertically (Figure 5), when gaps in fuel debris are small (Figure 3, 4) and when fuel debris surface islargely uneven (Figure 6).
DebrisAbsorbent(solid)Figure 1 Use of solid type ofabsorber on pebble-like fuel debris

Figure 3 Use of liquid to solidified type ofabsorber on pebble-like fuel debris

Figure 5 Use of liquid to solidified typeof absorber on rod-like fuel debris

Figure 2 Use of solid type ofabsorber on rock-like fuel debris

Figure 4 Use of liquid to solidified typeof absorber on rock-like fuel debris

Figure 6 Use of liquid to solidified type ofabsorber on fuel debris with large unevenness©International Research Institute for Nuclear Decommissioning

No.507. Implementation items
7.1 Development of on-site operating procedures
7.1.3 Operating procedures for neutron absorbentsa. Organization of prerequisites
Effect of mixing of absorbent and fuel debris

・Effect of absorbent when mixing fuel debris and absorbent by processing was analyzed and evaluated.
・Conditions were set for the retrieval area so that fuel debris (UO2) turns into grain due to processing and criticality occurs due
to water entering the gaps.
・Neutron multiplication factor becomes maximum when volume share of fuel debris is 20vol% and water is 20vol%. (Figure 1)
・When mixing absorbent (Gd2O3; grain), subcriticality occurs at only 3vol% (Figure 2).
・It was indicated that particle size of absorbent is more effect at 1mm than 1cm.
Complete reflection
Absorbent diameter

Retrieval area
(Fuel debris, water, absorbent)(Rectangular parallelopiped:
100×100×20cm)

90cm

110cm
Fuel debris (concentration 4wt%UO2)

200cm
Figure 1 Analysis model(*1) Easing of restrictions is expected if knowledge of debris
properties is gained in the future.

Neutron multiplication factor

Water

Absorbent volume occupancy

Figure 2 Analysis result (*1)
From the research report of FY2018 supplementary budget “Subsidy Project of Decommissioning andContaminated Water Management” Development of Technology for Retrieval of Fuel Debris and Internal Structures©International Research Institute for Nuclear Decommissioning

No.51

7. Implementation items
7.1 Development of on-site operating procedures
7.1.3 Operating procedures for neutron absorbentsa. Organization of prerequisites
Effect of arranging absorbent on fuel debris

・Assuming ring-like absorbent in 20cm width, analytical evaluation was conducted to see if it had an effect toexpand the retrieval area.
・When absorbents are arranged on the fuel debris surface in a collectable form, reactivity reduction effect isaround 2%Δk at most, and the effect is small although the retrieval area can be expanded by 2cm (from16cm to 18cm per side).

Complete reflection
Non-soluble
absorbent

Water

Retrieval area
(cube)

Fuel debris(only fuel, 1.0 volume share)

Drawing seen from above

Table 1 Size of retrieval area to limit addition reactivity to 0.1%Δk

90cm

110cm

Length of arrow is 20cm

Additionreactivity of
retrieval areaNoabsorbent

Length of 1 sideof retrieval area

16 cm(*1)

Collectable absorbent
Gd2O3
particle

Water glass(Gd)

18 cm(*1)

18 cm(*1)

200cm

Figure 1 Analysis model

From FY2017 supplementary budget “Subsidy Project of Decommissioning andContaminated Water Management” Sophistication of Method and System of Retrievalof Fuel Debris and Internal Structures (Technical development related to establishing(*1) Easing of restrictions is expected if knowledge of fuel debris properties is gained in the future. the criticality control method)©International Research Institute for Nuclear Decommissioning

No.52

7. Implementation items
7.1 Development of on-site operating procedures
7.1.3 Operating procedures for neutron absorbentsa. Organization of prerequisites
Entry of absorbent in gaps in fuel debris
・Solidified type (water glass / Gd2O3 granulated powder) of absorbent was dropped to the test body
(granular lava; several cm in diameter) simulating fuel debris, and it was confirmed that absorbent
penetrates deep into and adheres to gaps in fuel debris.
・Since particle size of solid type (granular) of absorbent is smaller than 1mm and its specific gravity is
sufficiently larger than water at 3-4, said absorbent is considered to penetrate gaps larger than 1mm.

Measurement result of weight of neutron absorbent
attached to granular lava
Neutron absorbent flowing underwater(injection amount 1000[g])Viscosity[mPa・s]Injectionamount [g]Granular lava

Numberof tests

Increased
weight [g]

1

109.2

2

105.4

1

19.5

2

23.4

1000

Increased
weight(average) [g]
107.3

2000
100

2cm

Figure 1 State of test of attachment to granular lava

21.5

From FY2015 supplementary budget “Subsidy Project of Decommissioning and Contaminated Water
Management” (Development of Fuel Debris Criticality Control Technology)©International Research Institute for Nuclear Decommissioning

No.53

7. Implementation items
7.1 Development of on-site operating procedures
7.1.3 Operating procedures for neutron absorbentsa. Organization of prerequisites

Mixing of fuel debris and absorbent by processing
Fuel debris processing tool: Chisel

B4C metal
sineratedmaterial・Underwater chisel processing test was conducted with solid type absorbent
sprayed on the test body (MCCI simulation) simulating fuel debris.
・Absorbent on the surface layer was removed after crushing, and it was
confirmed that absorbent had mixed into the area of crushed fuel debris
fragments.

Glass material
containing B・
Gd

5 mm

10.0
Solid-type absorbent(a) Underwater chisel processing test (simulated below pedestal)
Crushed fueldebrisBottom after removing
crushed fuel debris(kg)
Weight
重量（kg）MCCI simulated
fuel debris

Absorbent

Non-functional (surface layer / non-crushed part)
無機能（表層・未破砕部）
Not collected
未回収
Functional (debris mixing)
機能（デブリ混合）

8.0
6.0

No
デブリ
mixingと混合
withdebris
なし

4.0

(*1)

Not
未回収
collected

2.0

Required
必要量amountMixingデブリ
withと混合
debris0.0
B4C metal
B₄C金属焼結材
sinerated material

Glass material
B・Gd入ガラス材
containing B・Gd

吸収材
Absorbent
10 cm
Surface of
simulated fueldebrisAfter absorbent
application andprocessing

After removal ofabsorbent andcrushed fuel debris(b) Appearance before and after test (top)(*1) Easing of restrictions is expected if knowledge of fuel debris properties is gained in the future.

Figure 1 Mixed amount and required amount of absorbent

From FY2017 supplementary budget “Subsidy Project of Decommissioning andContaminated Water Management” Sophistication of Method and System of Retrievalof Fuel Debris and Internal Structures (Technical development related to establishing
the criticality control method)©International Research Institute for Nuclear Decommissioning

No.54

7. Implementation items
7.1 Development of on-site operating procedures
7.1.3 Operating procedures for neutron absorbentsb. Study of procedures
Overview of implementation items of 7.1.3.b
6 items raised as issues were studied, in discussing operating procedures and throughput
with the fuel debris retrieval method team.
- - - Top retrieval method - - - -- - - - - Side retrieval method (PLAN-A/B) - - - - - - - - - - -- Top access method fortransferring the unitized
structure - -① Judgment of use of absorbent, procedures for recovery to Level 1(→No.55)② Specifications of supply equipment of absorbent (→No.56～57)③ Operation process of absorbent (→No.58)④ Carry-in/out process
of absorbent (→No.59)⑤ Curing time of solidified absorbent (→No.60)⑥ Effectivenessevaluation of preapplication ofabsorbents (→No.61)©International Research Institute for Nuclear Decommissioning

No.55

7. Implementation items
7.1 Development of on-site operating procedures
7.1.3 Operating procedures for neutron absorbentsb. Study of procedures ① Judgment of use of absorbent, procedures for recovery to Level 1
Continuous monitoring of
neutron flux during processing
Criticality
continuous
臨界連続監視器
monitoring equipment⚫

Use of absorbent was determined, and procedures
for recovery to Level 1 were organized.

Subcriticality measurement

Criticality continuous
臨界連続監視エリア
monitoring area

Before fuel debris
processing

Continuous
monitoring isnormalConnect tosubcriticality
measurement
equipment

Approachmonitoring isnormal

Subcriticality
is normal

燃料デブリ
Fuel
debris

Subcriticality
未臨界度測定器
measurement equipment

Fueldebris Planned
processing location
燃料デブリ
加工予定場所Criticality approach monitoring
before and after processing

Fuel debris processing
Reset to initial valueof approachmonitoring

Supply of neutron
absorbent

Injection of neutron absorbent

Continuous
monitoring isnormalTo “Before fuel debris
processing”
Approachmonitoring isnormalFuel debris collection after
processing
Criticality approach
Fuel
debris Planned
臨界近接監視器
燃料デブリ
加工予定場所
monitoringprocessing
Criticality
continuous
location 臨界近接監視エリア
monitoring area

Fuel debris has approached criticality
at this point, but since it is not yet
critical, there is plenty of time to injectabsorbent
Reactivity suppression effect after injection
is quantitatively confirmed withsubcriticality measurement.

Fueldebris
燃料デブリ

Neutronabsorbent
中性子吸収材©International Research Institute for Nuclear Decommissioning

No.56

7. Implementation items
7.1 Development of on-site operating procedures
7.1.3 Operating procedures for neutron absorbentsb. Study of procedures ② Specifications of supply equipment of absorbent (simplificationof supply mechanism)
Table Required amount of absorbent⚫ Based on the required amount of absorbent(Table) revised in the relevant project(*1),
specifications of supply equipment of absorbent
were revised to downsize the volume by about
half. (Figure 1, 2)⚫ Supply equipment is different from the safetysystem, but since it is intended to maintain fuelΦ320
debris at subcriticality, high reliability equivalent
to that of fuel debris retrieval equipment isrequired.
Φ356

Non-soluble neutron
absorbent

Absorbent weight(*2)(*3)(kg/day)

Absorbent
capacity(*2)(*3)(liter/day)

Gd2O3 particle

6.4

1.5

Water glass / Gd2O3
granulated powder material

6.3

3.0

(*2) Fuel debris retrieval target per day is assumed at 300 kg.
It is assumed that 50% of the injected absorbent effectively functions.
(*3) Easing of restrictions is expected if knowledge of fuel debris properties isgained in the future (No.87).

容量：20L

Airエアーチューブ
tube

Φ300

ホッパー
Hopper
５L
5L

容量：5L 5L
Capacity:
Pressurized
加圧容器
container

672

Electric motor
電動モーター

360

392

スクリュー
Screw

Container
before
加圧前容器(5ℓ)
pressurization (5L)
Container
加圧後容器afterpressurization

Nozzlelid
ノズル蓋
開閉シリンダOpening/closing
cylinderNozzleノズル

ノズル
Nozzleバネ
SpringNozzleノズル
ノズル開閉蓋Nozzle

過年度検討
opening/closing lid

Figure 1 Supply equipment for solid type (Gd2O3 particle)

FY2021concept
2021年度構想

Air
cylinderエアシリンダーFigure 2 Supply equipment for solidified type(water glass / Gd2O3 granulated powder)(*1) Relevant project “Development of technology for further increasing the scale of retrieval of fuel debris and internal structures” (FY2019, FY2020)©International Research Institute for Nuclear Decommissioning

No.57

7. Implementation items
7.1 Development of on-site operating procedures
7.1.3 Operating procedures for neutron absorbentsb. Study of procedures ② Specifications of supply equipment of absorbent (confirmation of injection range)
・It was studied whether absorbent (glass material containing B・Gd) can be stably injected even if the injection mechanism is simplified byeliminating the motor and screw and directly attaching a valve to the hopper for the solid type absorbent injection equipment.
・Blockage was caused by hydrous absorbent with type with the smallest equipment outlet diameter, but it was confirmed that stable supply ispossible with other types. Additionally, it was confirmed that the diffusion range tends to expand with more distance from the target location of
input such as water depth becoming deeper and distance from equipment outlet to water surface becoming farther.
・It is considered better to have equipment outlet and injection location as close as possible to prevent increase in the amount of waste due toexcessive injection during actual operation, so it is considered necessary to study detailed design including the fact that the equipment operates
in an environment where the equipment itself is somewhat submerged.

①②③
Equipment outlet diameter, water depth anddistance from equipment outlet to water
surface are set as parameters.

Nominaldiameter: 1S
Inner diameter:
23.0mmOuter
diameter:
25.4mmNominaldiameter: 2S
Inner diameter:
47.8mmOuter diameter:
50.8mmNominaldiameter: 3S
Inner
diameter:
72.3mmOuter
diameter:
76.3mmFigure 1 Test equipment and used injection equipment type

Side

Top
Approx. 270mm

Approx.
45mm
Approx. 270mm

Side

Top
More than
300mm
More than 300mm*Diffusion throughout
the water tank

Approx. 5mm(deposit near center)

Water depth:
50mm
Distance fromwater surface:0mm
Equipment:③
Water depth:
600mm
Distance fromwater surface:0mm
Equipment:③

Case with actual scale of injection amount assumed at 5kg per injection and water
depth set as a parameter.

Figure 2 Distribution of absorbent after injection (example)©International Research Institute for Nuclear Decommissioning

No.58

7. Implementation items
7.1 Development of on-site operating procedures
7.1.3 Operating procedures for neutron absorbentsb. Study of procedures③ Operation process of absorbent⚫ The process assumed for on-site operation of absorbent
was studied (Table 1, 2).
⚫ In the case of solidified type absorbent (water glass),
there are more operation processes than the solid typesince raw materials need to be mixed and injection
equipment needs to be cleaned with water after use.

Table 1 Operation process of solid type
(Gd2O3 particle, glass material containing B・Gd)
No.
1

2
3

4

5

6

Table 2 Operation process of solidified type(water glass / Gd2O3 granulated powder)No.

Operation process

Details1

Storage of raw
materials

2

Weighing of raw
materials

Raw materials of water glass neutron absorbent are ① waterglass, ② gadolinia granulated powder, ③ cement, ④monosodium phosphate, and ⑤ water. Temporarily store
them at the place of production. ④ and ⑤ can be mixed andstored as aqueous solution.
Respectively measure raw materials required for the amountof application of absorbent. Another method is to inject theprescribed amount of raw materials stored in a tank into thesmelter using a dedicated squeeze feeder.

Mixing of raw
materials(*)
Loading of absorbent
to injection equipment

Mix the raw materials with the smelter and produce waterglass neutron absorbent.
Load the water glass neutron absorbent produced with thesmelter to the hopper of the injection equipment.

Operation
process
Storage ofabsorbent

Details
Glass material containing B・Gd or Gd2O3 particle isstored in a state that can be immediately used asabsorbent.

3

Weighing ofabsorbent
Loading ofabsorbent toinjectionequipment
Transfer of
injectionequipment to thefield
Injection ofabsorbent to fueldebrisMeasure the amount of absorbent required for work.

4

Load absorbent to the hopper of the injection
equipment.

5

Transfer of injection
equipment to the field

6

Injection of absorbent
to fuel debris

Connect the injection equipment loaded with water glass
neutron absorbent to the manipulator and move it directly
above the fuel debris retrieval site.
Inject water glass absorbent from directly above the fueldebris retrieval site to apply it onto the fuel debris surface.

7

Return of injection
equipment

After finishing application, return the injection equipment fromthe fuel debris retrieval site to the prescribed location.

8

Cleaning of injection
equipment(*)

Return of injection
equipment

After finishing injection, return the injection equipmentfrom the fuel debris retrieval site to the prescribedlocation.

Clean the used smelter and injection equipment with water towash away the adhered water glass absorbent. Collect andstore cleaning water.

Connect the injection equipment loaded withabsorbent to the manipulator and move it directly
above the fuel debris retrieval site.
Inject absorbent from directly above the fuel debris
retrieval site to spray it onto the fuel debris surface.

(*) Process in red is specific to the solidified type and is not required for the solid type.
©International Research Institute for Nuclear Decommissioning7. Implementation items

No.59

7.1 Development of on-site operating procedures
7.1.3 Operating procedures for neutron absorbentsb. Study of procedures④ Carry-in/out process of absorbent⚫
⚫

Example of solid type (Gd particle) of top retrieval method(mobile cell method)

Transport procedures from storage to use and collection were studied for the solid type absorbent (Figure 1, 2).
In order to save the limited space of R/B, absorbent is stored and prepared in a different building used for maintenance. Transport equipment(mobile cell) developed in the relevant project(*) is used to transfer the contaminated absorbent injection equipment between buildings. The mobile
cell is intended to transfer unit cans storing fuel debris.
The time required for transfer from preparing solid type absorbent to injecting it to fuel debris and collecting the injection equipment was estimatedat around 1.5 hours for outbound trip + 3 hours for return trip (due to cleaning and air replacement in the transfer area) = 4.5 hours in total.
If the injection equipment is prepared in the maintenance area in advance, injection can be performed within 1 hour.

⚫

⚫

Blue line:
Secondary boundary
青線：２次バウンダリ
パークセル（Ｃ４Ｍ）遠隔操作
Park
cell (C4M) remote operation

Maintenance cell (C4M)
メンテナンスセル（Ｃ４M）
Temporaryaccess by workers
作業員一時立ち入り

C4Mメンテナンスセル
C4M
maintenance cell

Workeroperation
作業員操作

Carry in mobile cell
モバイルセルを搬入

C3Transfer with unit can transport crane
ユニット缶搬送クレーンで移送

遠
隔
操
作

Remote operation

投入装置

Injectionequipment

搬送装置用
昇降機

Transfer with unit can hoisting equipment
ユニット缶吊り下げ装置で移送

Injection equipment is connected with
ロボットアームが投入装置を接続
robot arm
（ツールチェンジャ）(tool changer)
新オペフロ
荷重受け構台Injection equipment is moved to
ロボットアームが投入装置をデブリ現
debris site with robot arm
場へ移動

Fuel debris retrieval
燃料デブリ取り出し装置
equipment（プラットホーム下側）(lowerside of platform)

モバイルセル
Mobilecell

Figure 2 Carry-in route of supply equipment

Inject absorbent onto debris
吸収材をデブリ上に投下

モ
バ
イ
ル
セ
ル
の
搬
送

Transport mobile cell

Remoteoperation
遠隔操作
Separate injection equipment from
モバイルセルから投入装置を切離し
mobile cell

C4M
park cell
C4Mパークセル

構内搬送用
コンテナ

原材料の保管貯蔵
Storageof raw materials
Nomixing
混錬のプロ
process
セスなし

Weighing of raw
原材料の計量materials

Radiation
放射線管理区域
operation
作業員操作
controlled area Worker

投入装置
equipment
Injection

Loadingof absorbent to injection equipment
投入装置に吸収材を装填Store injection equipment in mobile cell
モバイルセルに投入装置を格納

R/B

Building used
for maintenance different from R/B
R/Bとは別の保守用建屋

Figure 1 Operation process and flow up to absorbent injection (Gd particle)(different from process to collect injection equipment and mobile cell after injection)©International Research Institute for Nuclear Decommissioning(*) Relevant project “Development of technology for further increasing the scale of retrieval of fuel debris and internal structures” (FY2019, FY2020)

No.607. Implementation items
7.1 Development of on-site operating procedures
7.1.3 Operating procedures for neutron absorbentsb. Study of procedures ⑤ Curing time of solidified absorbent⚫ Issues of solidified type absorbent (water glass / Gd2O3 granulated powder) became evident with respect to transport time.
⚫ It is necessary to adjust water glass absorbents so that they do not solidify before being injected (up to 2 hours). It is also
necessary to make adjustments so that the absorbents do not solidify before cleaning residuals remaining inside thecontainer after injection. (around 3 hours when cleaning with the dryer separator pit (DSP), around 6 hours when cleaningat the maintenance building)⚫ Range of viscosity that allows injection and cleaning is 0-5000[mPa・s].
⚫ Test was conducted changing concentration (0.95-1.87wt%) of curing retardant (Table), and time until 5000[mPa・s] isreached was measured.
20000⚫ Test results showed the prospect of delaying curing time by morethan 5 hours (Figure). It is slightly short of 6 hours, but there is achange of achieving it if the transport process is studied in detail.

Raw materials

Category

Main material

Unit 1 sodium silicate (Baumé specific gravity
55)

Curing agent

Ordinary Portland cement

Curing retardantMonosodium phosphate

Water

Ion exchange water

Neutronabsorbent

Gadolinium oxide

15000

Viscosity [mPa・s]

Table Raw materials of water glass / Gd2O3
granulated powder

TX-10(IRID, 2018)#1～5: Curing retardant differs

0.95wt%

1.5

1.7

1.4
1.3

1.87

10000
5000
Operation
range

0
0

60

120 180 240 300 360 420 480
Time [min.]Figure Test result of elapsed time and viscosity after
mixing of water glass / Gd2O3 granulated powder©International Research Institute for Nuclear Decommissioning7. Implementation items

No.61

7.1 Development of on-site operating procedures
7.1.3 Operating procedures for neutron absorbentsb. Study of procedures ⑥ Effectiveness evaluation of pre-application of absorbents⚫

Applicability of non-soluble absorbent was studied as a method to respond tocriticality risks outside the fuel debris processing area.Specific criticality risks
including falling of equipment and fuel debris⚫

Pre-injection is proposed since local subcriticality cannot be measured⚫

Effectiveness in preventing criticality risks due to falling of fuel inside the largetransport cask was evaluated as an example
Level 1 (Normal state)
Prevention of abnormality = criticality prevention

Level 2 (Abnormal state)

Level 3 (Accident)

Prevention of spread of abnormality =prevention criticality accident

Mitigation of impact ofaccident

Lid
Fuel debris

Solidifying
agentTransport
cask
Structure
fixingframe

Bottom RPV
mirror
CRD housing
Functional
requirement

Non-soluble
absorbent

Site boundary dose
(Maintenance of subcriticality) (Prevention of reaching criticality)
Subcriticality
is maintained
within thecontrol
standards

Prevention of falling
objects

Failure(criticality
occurs)

Maintenance ofsubcriticality with boric
acid water・Monitoring of boron
concentration

UO2 / free water falls(Option)

Detection of criticality(Gas sampling system)

Success (recovery to normal state by removing
critical areas)

Emergency boric acidinjection

Success (recovery to normal state by removing areas where criticality
occurred, after termination of criticality)

Failure (normal
level exceeded)

Additional injection of boric acidwater
Decreased water level due tofeedwater shutdown
Suppression of release ofradioactive gas

Prevention of
external impact
greater than thenormal level, bydetectingcriticality andpromptly
stopping
operation
Prevention ofoverexposureof generalpublic andworkers

Criticality control method during large-scale retrieval (falling of heavy weight object)→Referencematerials

UO2
Free water+
Non-soluble
absorbent

Drawing of storage inside transport cask
©International Research Institute for Nuclear DecommissioningNo.62

7. Implementation items
7.1 Development of on-site operating procedures
7.1.3 Operating procedures for neutron absorbents
d. Summary[Result]✓ Regarding criticality prevention during fuel debris retrieval, image of on-site operation of items, such as judgment of use of non-soluble
absorbent, procedures for recovery after use, supply equipment,
transport procedures and pre-application, was materialized, andinformation was provided for the fuel debris retrieval method team toutilize it to study the retrieval system design and throughput.

©International Research Institute for Nuclear DecommissioningNo.63

7. Implementation items
7.2 Development of solidified absorbent technology (water glass)[Issues]
Of non-soluble neutron absorbents developed in the relevant project(*1), water glass type absorbent (hereinafter
water glass material) is a viscous solidified material that adheres to fuel debris surface by covering it (Figure 1).
➢ When water glass material is covering the fuel debris surface (Figure 2), there is concern of inhibiting drying offuel debris.

[Implementation items]⚫ Changes in moisture content of the test body covered by water glassmaterial was measured, with reference to the conditions of fuel debris
drying process obtained in the relevant project(*2).
➢ Al2O3 porous ceramics which had the same drying behavior as UO2
was used for the test body.
➢ The test of FY2021 was conducted under conditions where thesurface in contact with outside air is completely covered.
➢ Target value of dry state is set to 0.5 vol% moisture content(equivalent to 0.1wt% fuel debris)[Goals]⚫ Acquisition of data on impact on the drying process of fuel debris withwater glass material covered

Figure 1 State of injection of waterglass type neutron absorbent tosimulated fuel debris

Realistically possible distribution

Fuel debris (containing water)

Porous simulated fueldebris(containing water)
Solidified neutron
absorbent

Stainless steel
crucible

Solidified neutron absorbent
・Adhered to part of debris
・Adhered to entire surface of debris(conservative assumption)

Figure 2 Estimation of state of fuel debris in the storage canister, andsimulated sample

Main results:⚫
Quantitative data on impact on the drying process of fuel debris with water glass (insoluble neutron absorber) attached is acquired.
⚫
Adjustment with each PJ (related to retrieval work and drying treatment) at present can support the study on applicability in the field,
which is necessary to grasp the impact of the retrieval work process and set drying conditions.
➢ By setting the target of the test body to “0.3wt% moisture content (zeolite equivalent)”, the evaluation is equivalent to the fuel debris
drying test conducted in the storage canister PJ and can feed back the results of this research.
(*1) Relevant project “Development of technology for further increasing the scale of retrieval of fuel debris and internal structures” (FY2019, FY2020)(*2) Relevant projects “Technology for identifying and treating properties of fuel debris” (FY2014), “Fuel debris characterization” (FY2015, FY2016)©International Research Institute for Nuclear DecommissioningNo.64

7. Implementation items
7.2 Development of solidified absorbent technology (water glass)a. Test plan
Overview of solidified absorbent⚫
⚫
⚫
⚫

No.1 sodium silicate (water glass), etc. mixed with Gd2O3 granulated powder which is a neutron absorbing material (Table 1).
Viscous liquid that solidifies over time.
Applicable for uneven surface and slanting surface which differ from flat surface. Can be adhered to fuel debris (Figure 1).
Neutron absorbent is used differently according to the various forms of fuel debris assumed such as rod-like and granular(Figure 2).

Table 1 Components of water glass typeneutronabsorbent
Component
ChemicalformulaWater glass material can be used regardlessof debris form
DebrisAbsorbent(solid)No.1 sodium silicate Na2O・nSiO2・xH2O
SiO2, CaO, Al2O3, Fe2O3,
CementCaSO4
Monosodium
NaH2PO4・2H2O
phosphate
Water
H2O
Gadolinium oxide

Exampleofofapplying
applyingsolid
solid
Exampleabsorbentfor
forpebble-likepebble-likeabsorbent
debrisdebrisExample ofof applying
applying solidified
solidified
Exampleabsorbent forfor pebble-likepebble-likeabsorbent
debrisdebrisExampleofofapplying
applyingsolidified
solidified
Exampleabsorbentfor
forrod-likerod-likedebris
debrisabsorbent

Gd2O3
Example of applying solid
absorbent for rock-likedebrisExample of applying
solidified absorbent for rockrock-likelikedebrisdebris

ExampleExampleofofapplying
applyingsolidified
solidifiedabsorbent for
debriswith large
unevenness
absorbent
for debris
with large
unevenness

Figure 2 Examples of different use of neutron absorbents

Figure 1 Example of water glass absorbent covering
uneven surface

When absorbent is covering the fuel debris surface,
there is concern of inhibiting drying of fuel debris.
©International Research Institute for Nuclear DecommissioningNo.65

7. Implementation items

7.2 Development of solidified absorbent technology (water glass) b. Test preparation
Overview of FY2021 test on impact assessment of fuel debris drying characteristics of solidified
neutron absorbent(1) Porous material
containing water(used as simulated fueldebris in past test)(2) Water glass material(alone)(3) Water glass material
covering porous
Covering
material
Porous Material
Containing water

Metal
container

water glass
underwater

Heat analysis
Drawing inside equipment

Drying characteristic curve (image)

Crucible
Thermocouple(temperature
measurement inside
heating reactor)
Thermocouple(temperature
measurement ofsample)Scale(measurement ofchanges in sample
weight)Porous material

Water glass material

Water glassmaterial + porous
material

External observation using a microscope, etc.

Water

Implementation items of FY2021•
Sample imaged with fuel debris covered with water glass material was prepared, and moisture and drying test system suitable for such
evaluation was rebuilt.
•
Test with the condition with water glass material entirely covering the sample (most conservative condition) was conducted to evaluate
whether it is affected by water glass material.
•
Property analysis of the water glass material itself was conducted for a quantitative evaluation of the impact of water glass material.
©International Research Institute for Nuclear DecommissioningNo.66

7. Implementation items
7.2 Development of solidified absorbent technology (water glass)
Reference: Past knowledge (investigation of drying behavior of fuel debris) (*1)

➢ In the past, a basic research on evaluation of drying behavior of fuel debris was conducted in cooperation with thecollection and storage PJ.
➢ In the case of fuel debris, since many parameters (form, material, drying temperature, etc.) are predicted to affectdrying behavior, focus was placed on the drying characteristic curve which enables uniquely centralized
expression and comparison.
➢ Since it is applied to fuel debris that has various specific gravity and is not comparable by moisture content to study the
optimum illustration method. , a method to evaluate the relationship between drying speed and residual water
was newly developed➢ The mechanism leading to drying differs between water glass material and porous material, but impact on drying
treatment can be inferred by clarifying the degree of impact on drying speed.

<Drying characteristic curve and drying theory>

C-D(Transition of dry
surface)

B-C(Evaporation ofwater from surface)

D-E(Evaporation ofresidual water)

Contrived
water
Suspended water(III) Decreased(II) Constant-rate (I) Preheating
rate drying period drying period
periodW/dt: Drying speed

A(Saturated
moisture state)
Sample containing
water (image)

Moisture content

Dry weight: W 1
Water content weight: W 2
Flow of process

Equilibrium moisture content: We
We: Equilibrium moisture
content
Wc: Limiting moisture
content
W: Moisture content → amount ofresidual water

Content at which drying does
not progress further

Limiting moisture content: Wc
Moisture content when transitioning
from constant-rate drying period todecreased-rate drying period©International Research Institute for Nuclear Decommissioning(*1) Relevant projects “Technology for identifying and treating properties of fuel debris” (FY2014), “Fuel debris characterization”(FY2015, FY2016)

No.67

7. Implementation items

7.2 Development of solidified absorbent technology (water glass) b. Test preparation
Impact assessment of fuel debris drying characteristics of solidified neutron absorbent
Aging curve of moisture content
Results and challenges of FY2021 test
Porous material + water glass material
Coating thickness 1mm, underwater curing 24 hours

The following was set as issues to be solved in FY2022⚫ Specific impact of water glass material on drying time of fueldebris cannot be observed just with pellet scale study➢ Impact of scale-up shall be inspected to make the data easier tobe incorporated in the study on drying process⚫ Since material balance of injection amount of water and weightafter drying does not match, presence of water needs to be
identified in terms of hydrogen generation➢ Composition of water glass material after drying shall be
elucidated for impact assessment of bound water, etc.
➢ Data on quantitative evaluation of drying time shall be acquired
to investigate impact of injection amount (percentage) of waterglass material*

Results and challenges⚫ It was confirmed that saturated porous Al2O3 (simulated fuel debris) will dry even when waterglass material is applied.
⚫ It was also confirmed that water glass material itself will dry to the point moisture is notreleased.
▲ Drying behavior of water glass behavior is different to that of porous Al2O3 alone, whichsuggests the impact of bound water, etc. in water glass material.
▲ It was confirmed that the injection amount (water content) of water glass material also affects
drying time.

0.5vol%Target moisture content of storage canister PJ (drying test of hydrous zeolitein actual equipment scale)⇓
0.1wt% of fuel debris equivalent ≒ 0.3wt% of zeolite equivalent ≒ 0.5vol%

Target moisture content: Less than 0.5vol%Side ×140

Top ×140

Cracks have formed on
the top

Gaps between particles
are confirmed©International Research Institute for Nuclear DecommissioningNo.68

7. Implementation items
7.2 Development of solidified absorbent technology (water glass)b. Test preparation
Overview of scale-up test⚫
⚫

Laboratory scale

Bench scale

Pilot scale

PurposeGrasp basic characteristics

Grasp scale-up factors

Grasp safety, efficiency andeconomyDrying testShape(simulated fueldebris)Internalstructure(porosity)

Internalstructure(pore
diameter)Material・Composition(Element・Chemicalform)

Hot air flux

Large

Large

From FY2014
Powdery

From FY2021

Decompressed exhaust(Internal pressure)

Radiant
heating

Oxide(Ceramic)
(UO₂)

Lumpy

Drying test(+ water glassmaterial)

*1

*1

Small

Small

heater(heating
temperature)

Simulated
MCCI
products(cementbase)

100mm

TMI-2
Debris,etc.

Al2O3 granule

Whether water glassmaterial is dry or not
Comparison of drying speedwith porous material

hot air(heating
temperature)
Chamber orstorage canister(Capacity)

Outer diameter
210mm

Conceptual study of drying equipment

Impact of injection amount (percentage)of water glass material

Impact of distribution (area / depth)of water glass material

Elucidation ofcomposition and reaction
mechanism of waterglass material

Ensure margin to accommodate minorchanges in drying equipment bygrasping basic characteristics andscale-up factors.
©International Research Institute for Nuclear Decommissioning(*1) Relevant projects “Technology for identifying and treating properties of fuel debris” (FY2014), “Fuel debris
characterization” (FY2015, FY2016)

Practical application

In the test scale of FY2021, the drying test was subject to small-scale (10g order) test bodies. Thus, scale-up is essential to gain
quantitative data.
Knowledge on scale-up factors shall be obtained by conducting bench scale test using the same evaluation method as that of laboratory
scale.
(Pilot scale test which assumes actual operation is unsuitable for identifying factors that lead to changes in drying time)⚫ Flow of development of fuel debris treatment technology (drying)7. Implementation items

No.69

7.2 Development of solidified absorbent technology (water glass)b. Test preparation
Overview of scale-up test
Impact of coating thickness<Drawing of water glass material adhering to fuel debris in storage canister>
Drawing of evaporation surface (conservative
assumption)

Realistically possible distribution

V=V’
Impact of evaporation area (percentage of coverage)

Solidified neutron absorbent

粒子状デブリの巻き上がり

Solidified neutron absorbent
・Adhered to part of fuel debris
・Adhered to entire surface of fueldebris(conservative assumption)

粒子状デブリの巻き上がり

Fuel debris (containing water)

Porous alumina containing water(simulated fuel debris)

φ9.5cm

FY2022: Drying test taking distribution (area, depth) of waterglass material into accountStainless steel crucible

φ45mm

Zeolite*
φ1.6mm*Same specifications
as the drying test of
storage canister PJ

1cm

FY2021: Drying test in the most conservative conditions (laboratory scale)

FY2022: Drying test in the bench scale covering previous test conditions

Test ① Simulated fuel debris drying test
➢ Investigation of drying behavior with the drying characteristic curve of zeolite applied to water glass material
and measurement of amount of change of water content©International Research Institute for Nuclear Decommissioning
Test ② Crystal structure analysis of water glass material➢ Investigation on moisture properties with composition analysis of water glass material alone

No.707. Implementation items

7.2 Development of solidified absorbent technology (water glass)c. Element test
Test ① Simulated fuel debris drying testTest result of zeolite + water glass material(5mm equivalent)
Zeolite alone

Zeolite + water glass material (5mm equivalent)

Change of water content over time

Change of water content over time

Appearance before and after drying test
(Zeolite, water glass material 5mm equivalent)
Drying characteristic curve

Property investigationof water glass material Drying characteristic curve
related to residual©International Research Institute for Nuclear Decommissioning
weight is conducted in
test ②

No.71

7. Implementation items

Water content [g]Hydrous zeolite with water glass material Φ45mm
Hydrous zeolite Φ45mm
Hydrous zeolite with water glass material Φ90mm
Hydrous zeolite Φ90mm

Same level of water content→Drying time does notchange(drying time depends on watercontent)

Drying end time [min]⚫

⚫

Hydrous zeolite (+ water glass material) volume [㎤]7.2 Development of solidified absorbent technology (water glass)c. Element test
Test ① Simulated fuel debris drying testEvaluation of impact on drying time
(Note) Drying end time
Time until drying speed reaches 0(mg/cm2・min)Hydrous zeolite with water glass material Φ45mm
Hydrous zeolite Φ45mm
Hydrous zeolite with water glass material Φ90mm
Hydrous zeolite Φ90mm
Increased drying time due toimpact of moisture associatedwith water glass material

Drying end time [min]It was confirmed that total amount of moisture contained in zeolite and moisture contained in water glass material
affect the drying end time(Note).
➢ There was no inhibition of drying treatment caused by zeolite pores and blockage of gaps by water glassmaterial.
➢ As water content of the test body increased due to injection of water glass material, drying end time tended toincrease compared to the condition without water glass material.
(around 100min. when approx. 50% of volume of hydrous zeolite is replaced with water glass material)
Surface area largely contributes to drying speed of the test body, which suggests the possibility of dependence on
container dimensions.
©International Research Institute for Nuclear Decommissioning

No.72

7. Implementation items
7.2 Development of solidified absorbent technology (water glass)c. Element test
Test ② Crystal structure analysis of water glass material
Structural analysis using X-ray diffraction analysis (XRD) andnuclear magnetic resonance analysis (NMR)

1h- heatup drying 01
24h- heatup drying 01

Evaluation of the drying characteristic curve indicates that the test bodyreaches a sufficiently dry state, but has confirmed residual weight ofmoisture.
Impact of residual weight of moisture shall be observed by clarifying the
crystal structure of water glass material after drying treatment.
⚫ For structural analysis of amorphous regions like water glass material, it
is effective to use XRD analysis.
➢ In the structural analysis of FY2022, measurement was conductedwith analysis equipment suitable for analysis of amorphous regions.
⚫ NMR, which is expected to evaluate electrochemical characteristics ofanalyzed sample surface, shall be used to observe the differences ofcrystal structure (network formation status of Si-Si bonding) that cannot
be analyzed with XRD.
⚫ Such structural analyses shall qualify and quantify amorphous regions inwater glass material and help understand the state of bond water ofwater glass material.
(re-commissioned to Shizuoka Institute of Science and Technology)⇒ Derived ratio of free water to bond water in water glass material anddrying behavior of bond water shall be incorporated in the bench scale
drying test results.

Ca3SiO5 :Monoclinic (00-055-0738)

Na2O・CaO・SiO2 :Cubic (00-024-1069)

SiO2 :Cubic (01-073-3133)<X-ray diffraction pattern of water glass material (FY2021)>©International Research Institute for Nuclear Decommissioning

No.73

7. Implementation items
7.2 Development of solidified absorbent technology (water glass)c. Element test
Test ② XRD analysis result of crystal structure analysis of water glass material
Gas phase curing 48h_before
drying_Gd2O3-1/8 times

Curing in water 48h, 80℃, 300min. drying_Gd2O31/8 times

Sum(Tobermorite)
Sum(Quartz_syn)
Tobermorite+Quartz_syn

Sample cured in gas phase for 48 hours

Sample cured underwater for 48 hours

Quartz: Compound similar to silica gel (SiO2· nH2O)
Tobermorite: Compound similar to tobermorite (5CaO・6SiO2·5H2O)⚫ It was revealed to contain structures similar to silica gel and tobermorite, which are known as amorphous components
existing in cement and concrete materials.
➢ The test result shows that the main components of water glass material do not form strong crystalline bodies in underwater
curing.
➢ The test result shows that there is very little crystalline water composition of water glass water material in underwater curing.
©International Research Institute for Nuclear Decommissioning

No.74

7. Implementation items

7.2 Development of solidified absorbent technology (water glass)c. Element test
Test ② Comparison of gas phase curing and underwater curing in NMR analysis of solids of crystal
structure analysis of water glass material
Qn species bonding of Si-Si bonds increases

Qn species

Gd2O3=1/8 times

Q0
SiO4 tetrahedral structureO
Underwater curing 1hUnderwater
curing 48h

O

Si

Si

Si

Si

Si

Q1

Si

Q2
Q3

O
Si

O

Si
Si

Q4

Q23

Q36

NMR analysis result of water glass material after drying treatment⚫
⚫

⚫

NMR analysis revealed that peak intensity of high-dimensional structure of SiO4 tetrahedral structure tends to increase.
Amorphous nature of solidified materials including neutron absorbents has a structure that does not hold water strongly as it does nothave a pore structure.
➢ Water existing in water glass material is crystalline water, and results suggest that this crystalline water affects residual watercontent.
➢ Crystalline water is known not to contribute to hydrogen generation under irradiation 1), 2).
➢ It agrees with the results of γ ray irradiation test of water glass material conducted in the past 3).
Existence of Gd2O3 does not affect the crystal growth process, and it was suggested that it is held in the crystal structure (maintains
neutron absorption effect).
©International
Institute for Nuclear Decommissioning
1) Masaomi Takahashi and Tomonari Fujita, “Effect of condition of the water in cement paste for hydrogen gas generated by gamma-ray irradiation”, CRIEPI
Report, L11020, Research
20132) Taisuke Tsukamoto, Hideaki Uchiyama, et al., “Development of Fuel Debris Canister (9) Measurement of the Quantity of Hydrogen by the Radiolysis of the Concrete Component Water under the γ-Ray Irradiation”, Atomic Energy Society ofJapan Fall Meeting, 2D18, 20173) Ryo Ishibashi, Toshiyuki Fujita, et al., “Criticality control technique development for Fukushima Daiichi fuel debris (25) Evaluation of Gamma-ray Irradiation Resistance of Insoluble Neutron Absorber”, Atomic Energy Society of Japan Fall
Meeting, 2H20, 2016

No.75

7. Implementation items
7.2 Development of solidified absorbent technology (water glass)d. Summary[Result]
Test ① Simulated fuel debris drying test (scale-up test)⚫
It was confirmed that total amount of moisture contained in zeolite and moisture contained in water glass material affect thedrying end time.
➢ There was no inhibition of drying treatment caused by zeolite pores and blockage of gaps by water glass material.
➢ As water content of the test body increased due to injection of water glass material, drying end time tended to increase
compared to the condition without water glass material.
Test ② Crystal structure analysis of water glass material⚫ XRD analysis revealed that water glass material contains structures similar to silica gel and tobermorite, which are known as
amorphous components existing in cement and concrete materials.
➢ It was revealed that the main components of water glass material do not form strong crystalline bodies in underwater
curing.
⚫ NMR analysis revealed that amorphous nature of solidified materials including neutron absorbents has a structure that does
not hold water strongly as it does not have a pore structure.
➢ Components that contribute to crystal water of water glass material used in an actual environment tends to be trace
amounts, and it is inferred that amount of water evaporation generated by heating of fuel debris due to such crystal wateris also negligible.
⚫ It was revealed that water glass material after drying has a composition with very little crystal water and also has a structurethat does not hold water strongly.
➢ It was confirmed that impact of radiolysis of water glass material on hydrogen generation is limited.

While water glass material does not inhibit drying of fuel debris, results showed that water glass material extends drying time.
Thus, injection amount of water glass material shall be minimized.
It was also revealed that water glass material after drying treatment has a composition that does not generate hydrogen.
©International Research Institute for Nuclear Decommissioning

No.76

7. Implementation items
7.3 Overall summary of criticality control

・Based on the aforementioned studies of 7.1-7.2, requirements and monitoring procedures wereorganized as the overall summary of criticality control.
・Procedures according to the fuel debris retrieval method need to be selected for the criticality control
method, but basic requirements that should be followed in all cases were organized to unify theapproach.
・Developed examples of a representative method are presented to confirm whether the requirements areappropriate.
・Specific procedures and approach on judgment criteria were organized as supplemental explanation ofcriticality monitoring procedures incorporated in criticality control procedures.
Organized items
Item1) Purpose of criticality control

2) Criticality prevention3) Detection and termination ofcriticality
4) Implementation conditions ofretrieval task5) Containment of retrieved material

Requirement
Purpose of criticality control
・Purpose and goals of criticality control
Risk identification
Maintenance of subcriticality
・Elimination of criticality approach factors
Prevention of reaching criticality
Criticality approach monitoring・Termination before reaching criticality even when criticality is(predictive monitoring)approached
Condition monitoring during work(constant monitoring)・Early detection of abnormality in deep subcritical state assumed Long-term monitoring
Detection of criticality
・Terminate the event before becoming accident level with early
Termination of criticality
detection and termination, even when criticality occurs.
Measures after termination ofcriticality
－

－

－

－
©International Research Institute for Nuclear Decommissioning

No.77

7. Implementation items
7.3 Overall summary of criticality control

Configuration
・Requirements common to all methods are organized (red box), and specific developmentof a representative method is shown on the right (examples shown in below table).
・Development to each method is indicated by each assumed risk
The lower (lateral) retrieval method is organized into three categories: no boric acidwater, normal use of boric acid water, and interference removal.
・Criticality control procedures are attached as supplemental explanationi) Subcriticality measurement, ii) Predictive monitoring, iii) Constant monitoring, iv)
Detection and termination of criticality
Requirements of criticality control
Item2) Criticality
prevention

Requirement2-1) Implement
subcriticality maintenancemeasures (criticality
prevention measures) for
identified risks.

Requirement (details)

Specific development: Lower lateral retrieval work (withoutnormal use of boric acid water)
Over-processing of fuel Accumulation of chip(omitted)debrispowder

2-1-1) (Criticality control personnel) Implement
Retrieval work is the direct
Retrieval work is the initiating
measures, ① Maintenance of subcriticality and ②
initiating event. Thus,
event, but a long period of time
Prevention of reaching criticality, for criticality prevention. implement measures for ①
is required for chip powder to
2-1-2) (Criticality control personnel) For ① Maintenance Maintenance of subcriticality
accumulate from the generatedof subcriticality, implement measures such as preventing and ② Prevention of reaching amount to critical mass, andevents that cause criticality approach, suppressing
criticality in 2-2) and 2-3),
multiple condition
excessive criticality approach (suppressing addition
respectively, considering thesuperpositions are required
reactivity) and maintaining subcriticality with neutron
occurrence frequency is large. such as concentrated
absorbents.
Selection of criticality control
accumulation in specific2-1-3) (Criticality control personnel) For ② Prevention of method for each target is shown locations. Thus, implement
reaching criticality, implement measures to monitor
in supplemental flow 1.
implement measures for only
① Maintenance of subcriticality
criticality approach and stop work before reaching
criticality. However, if criticality risk of retrieved objects isin 2-2), considering thedetermined to be small in ①, or when enhancing theoccurrence frequency is small.
method to maintain subcriticality with neutron
absorbents, it is acceptable to relax criticality monitoring.

(omitted)©International Research Institute for Nuclear Decommissioning

No.78

7. Implementation items
7.3 Overall summary of criticality control

Purpose of criticality control
・“Accident” is defined as an event that causes an external impact exceeding the normal level
・The purpose of criticality control is to “prevent criticality, and prevent criticality from spreading toa criticality accident in case it occurs.”
Identification of criticality risk
・Each method is required to analyze and identify risks and select the management method
according to risk management (characteristics and magnitude or risk).
Requirements of criticality control
Item

Requirement
Requirement (details)
1-1) The purpose of criticality control is to prevent
1-1-1) Definition of criticality accident
criticality associated with fuel debris retrieval work,
Criticality accident is defined as an event that causes an impact
mitigate the impact in case criticality occurs, and prevent exceeding the normal level due to additional release of radioactive
criticality accidents from occurring (prevent expanding to materials due to criticality.
criticality accident). Earthquakes and other natural
In “normal level”, exposure dose of the general public at site boundary
disasters are not subject to management under thisis 1mSv/year and exposure dose of workers is 50mSv/year (Note 1).
requirement.
1-1-2) Definition of criticality
1) Purpose of
Criticality is when effective multiplication factor (keff) =1.0. However,criticality control
considering uncertainty of criticality events, keff=0.95 is set as thecriticality control standard in studies of various measures (Note 2).
1-2) Analyze risks associated with fuel debris retrieval,
1-2-1) (Criticality control personnel) Identify criticality risks in subject
and implement measures to prevent criticality according work and methods.
to the identified risk and measures to prevent spread to 1-2-2) (Criticality control personnel) Implement independent criticality
an accident in case criticality occurs.
prevention measures for the identified risks and measures to mitigate
impact (prevent spread to criticality accident) when criticality occurs.

Note 1) The definition of normal level is an example that is currently being studied, and it shall be revised in accordance with the progress ofthe study on the concept of safety. Site boundary dose is targeted below 1mSv, even when considering increase in impact per event due
to fuel debris retrieval.
Note 2) In criticality evaluation, keff=0.95 is often set as the judgment criteria including various uncertainties.
©International Research Institute for Nuclear Decommissioning

No.79

7. Implementation items
7.3 Overall summary of criticality control

Level of defense in depth and approach of response measures
・In the approach on safety management, external impact is divided into levels (Note 1), in line with“accident” being defined as an event that causes an external impact exceeding the normal level. The
policy is to establish means to determine the level and establish measures to prevent transition to ahigher level and recovery measures in case of such transition.
Note 1) Since external impact may change due to various nuclear output even with the same criticality, accident standard is set as“event that causes an external impact exceeding the normal level” (that is to say, it does not mean that keff of the system is 1.0).
However, consideration shall be given so that worker exposure is not exceeded.

Since it has been confirmed in past studies that external impact does not exceed the normal level by promptly
detecting and terminating criticality, the goal of future equipment on-site procedures is to keep the level below thenormal level even during criticality.
Accident standard

・Above approach does not allow for criticality.
Multiple measures are implemented for criticality prevention, as
mentioned later.

Normal standard

Public dose (mSv)

Level 1: Normal state
Level 2: Abnormal state (physical criticality has occurred but
external impact is below the normal level)
Level 3:Criticality accident state (external impact has occurred)

When criticality occurs due tofalling of 450kg of fuel debris
attached to 1 CRD.
It is relatively large
supercriticality, but it is within
the normal level. (*1)
Time (h)(*1) Easing of impact is expected if knowledge of fuel debris properties is gained in the future.

Example of criticality exposure assessment (from FY2016 subsidy
project results)©International Research Institute for Nuclear Decommissioning7. Implementation items

No.80

7.3 Overall summary of criticality control

Case of normal use of boric acid water

Criticality control method during large-scale retrieval (no boric acid water) is shown in the reference material
Level 1 (Normal state)
Prevention of abnormality = criticality prevention

keff

Level 2 (Abnormal state)
Prevention of spread of abnormality= prevention criticality accident

Level 3 (Accident)
Mitigation of impactof accident

Functional
requirement

1.0

Site boundary dose
(Maintenance of subcriticality)

1mSv(Prevention of reaching criticality)

Retrieval work restriction

5mSv

Xe = 1 Bq/cm3 (*)*State positioned as “occurrence of criticality” in the current
implementation plan

Subcriticality is
maintained within
the control
standards

Failure(criticality approach occurs)
Criticality measurement

Predictive monitoring before processing, constant monitoring during processing

Criticality approach monitoring (neutron
detector, gas sampling system)

Stopping of retrieval work
Success (recovery to normal state by removing areas
approaching criticality, by deepening subcriticality withneutron absorbent)(According to the situation)
Transition to normal use of boric
acid water (**)

Failure (criticality occurs)

Detection of criticality(gas sampling system,
neutron detector)**Inject boric acid water in advance and perform retrieval
work in boric acid water

Emergency boric acid
Success (recovery to normal state by removing critical areas) injection
Success (recovery to normal state by removing areas where criticality occurred, aftertermination of criticality)

Failure (normal
level exceeded)

Prevention ofcriticality bydetecting criticality
approach morethan expected andstopping retrieval
work

Prevention of
external impact
greater than thenormal level, bydetecting criticality
and promptly
stopping operation

Prevention of
Additional injection of
boric acid water
overexposure of
Decreased water level due
general public
to feedwater shutdown
and Decommissioning
workers
Suppressionofreleaseof©International Research Institute for Nuclear
radioactive gas

No.81

7. Implementation items
7.3 Overall summary of criticality control

・Requirements corresponding to the level of defense in depth
Level

Required function

MethodLevel 1
(Normal state)[Maintenance of subcriticality] →No.84
Subcriticality is maintained within thecontrol standards[Subcriticality measurement] →No.84
Start of work upon quantitatively grasping margin until criticality[Retrieval work restriction]
Suppression of excessive reactivity addition[Maintenance of subcriticality with neutron absorbent] (option) →No.85
Normal use of boric acid water, non-soluble absorbent[Prevention of reaching criticality]
Prevention of criticality by detecting
criticality approach and stopping retrieval
work[Criticality approach monitoring] (predictive monitoring / constantmonitoring) →Reference
Monitoring of conditions during retrieval work to stop work before
reaching criticality[Detection and termination of criticality]
Prevention of external impact greaterthan the normal level, by detecting andpromptly terminating criticality[Detection and termination of criticality]
Detection of criticality, termination with boric acid water injection

Prevention of overexposure of generalpublic and workers[Securing of method to terminate criticality and method to mitigate
external impact]
Securing of supplementary method of termination and method tomitigate external impact, in case criticality is not terminated withboric acid water injection

Prevention ofabnormality= Criticality
prevention

Level 2
(Abnormal state)
Prevention of
spread ofabnormality= Criticality
accident prevention

Level 3
(accident)
Mitigation of impactof accident

Italic numbers indicate the location of relevant explanation©International Research Institute for Nuclear Decommissioning

No.82

7. Implementation items
7.3 Overall summary of criticality control

Maintenance of subcriticality
・Maintenance of subcriticality
Estimate current subcriticality, and start work upon confirming the margin until criticality
・Elimination of criticality approach factors
Fuel debris processing volume limitation (requirement for criticality monitoring every 4,000cm3 (*1))
Specific development: Lower lateral retrieval work (withoutnormal use of boric acid water)

Requirements of criticality control
Item

Requirement

Requirement (details)

Over-processing of fuel debris

2-2)
2-2-1) Maintenance of subcriticality
Maintenance Start work upon confirming that subcriticality isofmaintained even when considering criticality is
subcriticality approached due to the implemented work.
Eliminatefactors ofcriticality
2-2-2) Eliminate factors of criticality approach.
approach2)
Also select work methods that do not cause
Criticality and maintain excessive reactivity addition.
prevention subcriticality
during work.

Before starting work, measure subcriticality using subcriticality
measurement neutron detector loaded to the retrieval arm, andstart work upon confirming that criticality does not occur evenwhen considering reactivity addition associated with processing
work.
→Subcriticality measurement procedures
Perform criticality monitoring (predictive monitoring) at every(*1)
fixed processing volume (4,000cm3) . Take uncertainty ofprocessing volume into consideration for the judgment criteria ofpredictive monitoring, and confirm validity of the margin at thephase before large-scale retrieval.

2-2-2-1) Judgement criteria of excessive reactivity Addition reactivity of each criticality monitoring (predictive
are maintenance of subcriticality (criticality is not monitoring) is set at less than 0.5%Δk/k.
reached with one addition), and amount ofaddition reactivity that does not cause immediate
criticality in case criticality occurs.

Other requirements are organized in similar ways. (predictive monitoring, constant monitoring, long-term monitoring,means of criticality detection and termination, response after termination, retrieval work conditions)(*1) Easing of impact is expected if knowledge of fuel debris properties is gained in the future.
©International Research Institute for Nuclear Decommissioning7. Implementation items

No.83

7.3 Overall summary of criticality control
・Make specific developments for each method, based on basic requirements.
Here are developments to normal use of neutron absorbent (boric acid water) and removal of interference objectives (non-fuel debris)
Item

Lower lateral retrieval work(without normal use of boric acid water)

Requirement

Lower lateral retrieval work
(normal use of boric acid water)

1) Purpose of
Omittedcriticality control

CommonMajor risks:
2-1) Implement subcriticality maintenance Excessive fuel debris processing /measures (criticality prevention
accumulation of chip powder / falling ofheavy weight object / increase in watermeasures) for identified risks.
level2) Criticality
prevention

Interference removal inside pedestal

Since processing amount of fuelCriticality risk is reduced by boric aciddebris is small, risk is low, and falling
water, and decrease in boronof heavy weight object becomes theconcentration becomes the major risk.
major risk.

2-2) Maintenance of subcriticality
Eliminate factors of criticality approachand maintain subcriticality during work.

Confirmation of margin with subcriticality Maintenance of subcriticality withmeasurement
boric acid water of concentration(*1)
Limitation of fuel debris processing(6,000ppm)

2-3) Prevention of reaching criticality
2-3-1) Criticality approach monitoring(predictive monitoring)

Perform predictive morning at every fixed Boron concentration monitoring andvolume processing
maintenance2-3-2) Condition monitoring during work(constant monitoring)

Perform constant monitoring during
processing

2-3-3) Long-term monitoring of thecondition during retrieval

Confirm that more changes than expected－have not occurred before and after
processing
1/m, trend analysis of gas monitor
Same as left (implemented with lessindication value
frequency)

Detection of criticality using the PCV gas
3) Detection and 3-1) Detect criticality and promptly
termination of terminate criticality to prevent spread to a monitor, termination of criticality due tocriticality accident.
emergency boric acid water injection
criticality
4-1) Retrieval work must not be performed
Same as left
4) Implementation without above means functioning.
conditions of4-2) Confirm that detectors used for
retrieval task
Same as left
monitoring have been appropriately
calibrated and are operable.
When storing retrieved objects in thetransfer container (unit can, waste
The prerequisite is that appropriate5) Containment ofcontainer), no special controls (measures criticality prevention is implemented for
retrieved material
such as monitoring and absorbent
canisters with shape management.
addition) are implemented.

Not applied since changes due toprocessing are small
Not applied since criticality risk is
small

Perform continuous monitoring(constant monitoring) of neutron flux Perform constant monitoring in viewof unexpected situations.
as backup of boron concentration
monitoring.
－Same as left (implemented with lessfrequency)

Same as left (detection of criticality,
emergency injection of boric acidwater)

Same as left (detection of criticality,
emergency injection of boric acidwater)

Same as left

Same as left

Same as left

Same as left

Same as left

Same as left
©International Research Institute for Nuclear Decommissioning(*1) Easing of impact is expected if knowledge of fuel debris properties is gained in the future.

No.84

7. Implementation items
7.3 Overall summary of criticality control

Subcriticality measurement procedures

Item

Requirement

Approach

1) Purpose / function

・Quantitatively grasp margin until criticality
・Set initial values of predictive monitoring (refine monitoring)・Start retrieval work upon confirming enough margin
・Expand the operational range by refining predictive monitoring (keff monitoring)2) Measurementtarget・Fuel debris submerged on a potentially critical scale (deposit
inside pedestal, deposit of RPV lower head)・Can be handled with predictive monitoring and/or constant monitoring, if fuel debris
processing scale is small3) Measurement
method・Acquire neutron pulse time history data with the neutron
detector near fuel debris, and apply the reactor noise method(Feynman−α method)

・The Feynman−α method is a representative method for subcriticality measurement.

4) Measurement
position・Measure at least one point on a lump of fuel debris
Re-measure if the retrieval position moves to a different lump
・Re-measure for every 10-20cm of digging

・A lump of fuel debris has one subcriticality regardless of measurement position.
・Consider the possibility of composition and properties changing in the vertical direction
such as MCCI fuel debris
・Neutrons are not necessarily emitted equally in all directions. Considering the detector
could go behind the robot arm, simulate neutron measurement as necessary to
determine detector placement.

5) Measurement time・Assumed as 1 hour to several days depending on fuel debris
form (size of system)

・Estimate required measurement time based on neutron flux in the field and detector
sensitivity6) Uncertainty
expected in themeasurement
result・Expect uncertainty of subcriticality measurement (prompt
neutron attenuation coefficient α) and uncertainty of
conversion to keff (uncertainty of ℓ*)・Evaluate uncertainties with KUCA test results7) Method of setting
judgment criteria

・Confirm that criticality will not occur even when adding
addition reactivity of several times of processing to themeasured keff

・Confirm that criticality does not occur with the first processing (for second time andonward, monitor criticality approach with predictive monitoring)8) Measures whenmeasurement
cannot be
performed

・Perform re-measurement increasing the measurement time・Perform monitoring with 1/m instead of keff for predictive
monitoring.
In some cases, limit fuel debris processing volume to a small
amount, and gradually expand the scale while performing
predictive monitoring.

・There are cases when measurement cannot be performed due to the size of fuel debris
system and depth of subcriticality.
・A method is required to differentiate the causes of not being able to performmeasurement between deep subcriticality (outside measurement range) and otherreasons.

・It has been applied to Unit 1. It is desirable to perform measurement with the same

・Method of estimating subcriticality based on the percentage of method also at Units 2 and 3.
9) Alternative means
・Organize the position and necessity regarding the relationship with subcriticality
Xe/Kr measured value on the PCV gas monitor
Predictive monitoring procedures are shown in the reference materials

estimation with neutron detector.
©International Research Institute for Nuclear Decommissioning

No.85

7. Implementation items
7.3 Overall summary of criticality control
Procedures on injection of non-soluble neutron absorbent(Example of using near-critical fuel debris)
Item

Requirement

Approach

1) Purpose /function

・When it is confirmed by subcriticality measurement orpredictive monitoring that fuel debris has approached criticality, ・Prevention of abnormality of criticality control method (no boric acidwater) during large-scale retrieval is addressed.
deepen subcriticality to recover to a normal state.

2) Method

・Load non-soluble absorbent to the injection equipment,
transport it above the target fuel debris with robot arm, andspray absorbent to the surface of fuel debris from above.

・It order to use it locally, the prerequisite is that the scope of target fueldebris is known.

3) Injection
frequency

・When necessary. (irregular)

・It is not used on a daily basis but used only when criticality approachis confirmed.

4) Injection
position・Local fuel debris area which has been confirmed to beapproaching criticality

・The target range is equivalent to daily fuel debris retrieval target of
300kg.

5) Injection time・Around 1 hour

・It is assumed that fuel debris retrieval is performed after injecting
absorbent once a day.
・There is no urgency since it is Level 1 operation.

・Inject absorbent in the field upon setting the absorbent type・Issues are the approach of determining absorbent type, injection6) Method of(solid or solidified), injection amount and range.
amount and range, and the confirmation method. (→effectivenesssetting injection ・Confirm that the predetermined amount has been injected in the evaluation)predetermined range.
amount・Confirm the effect of injection with subcriticality measurement.
・Perform subcriticality measurement after injection.
・If the predetermined amount and range are not achieved,7) Response torepeatedly inject absorbent until they are satisfied.
shortage of・If
effects are not confirmed by subcriticality measurement,
injection amountchange the absorbent type.
8) Effectivenessevaluation

・Perform criticality evaluation by simulating the observed fueldebris form, and confirm effectiveness of the absorbent.

9) Response toequipment
failure・Position standby of non-soluble absorbent injection
equipment as a condition of retrieval work.

10) Other・Establish procedures and judgment criteria for restarting fueldebris retrieval work after injecting absorbent.

・If solid type absorbents do not enter gaps in fuel debris and effects are
not confirmed, effects may be obtained by injecting solidified typeabsorbents.
・It is recommended to simulate representative patterns of fuel debris
forms assumed in advance. This analysis will be used to determine theinjection amount in 6).
・Injection equipment failure itself is not a cause of criticality event.
However, since it cannot be recovered to the normal state, fuel debris
retrieval cannot be performed.
・Perform subcriticality measurement to confirm that keff is lower than
the judgment criteria. The judgment criteria is a value smaller than0.95, considering increase due to work.
©International Research Institute for Nuclear Decommissioning

No.86

7. Implementation items
7.3 Overall summary of criticality control
Summary・Basic requirements were organized for criticality control during large-scale fuel debris
retrieval.
Purpose of criticality control
Maintenance of subcriticality: Subcriticality measurement, retrieval work restriction,
normal use of neutron absorbent
Prevention of reaching criticality: Criticality approach monitoring (predictive monitoring /constant monitoring), long-term monitoring
Detection and termination of criticality: Detection of criticality, criticality termination due
to emergency injection of boric acid water・Criticality monitoring procedures were organized.
Subcriticality measurement, predictive monitoring, constant monitoring, detection andtermination of criticality

・Injection procedures for non-soluble neutron absorbent were organized.

©International Research Institute for Nuclear Decommissioning

No.87

7. Implementation items
7.4 Organization of issues for the future

Based on knowledge and issues obtained so far, issues to be addressed by full-scale
retrieval are shown below while taking into account materialization of retrieval methods,
internal information and new information of fuel debris forms.

(Table 1) Issues of methods⚫ It is necessary to take note that importance of issues to be solved changes according to the adopted method
Item

Issues・Identification of objects that could fall or collapse by confirming the situation inside. Considerations in the
Measures against
retrieval plan.
falling of heavy・Analysis of fuel debris properties (evaluation of criticality risk by grasping fragility “destructive energy”)
weight object
・Evaluation of external impact when criticality occurs in a representative event
・Accumulation potential evaluation by chip powder flow analysis based on generated amount of chip powder of
each processing method, water level during large-scale retrieval, and coolant circulation amountMeasures against・(As necessary) Study on criticality detection method in vent pipe, S/C and torus room, and boric acid water
chip powder
injection method
・Evaluation of external impact of chip powder accumulation criticality
Throughput
improvement

・Re-evaluation of criticality risk based on acquired information, study on simplification of management methods
・Study on possibility of acceleration of fuel debris retrieval with normal use of boric acid water

Other・Confirmation of opinions of experts regarding fuel debris retrieval without boric acid water・Criticality evaluation used as basis for limitations on methods, such as processing limits, water level limits,
minimum critical mass and handled amount, is performed under conservative conditions taking uncertainties offuel debris into account. If information such as fuel debris enrichment, heterogeneity and U content is obtained
through future internal investigations and gradual retrieval, limitations of methods are expected to be mitigatedby revising the excessively conservative conditions and re-studying criticality evaluation.
©International Research Institute for Nuclear Decommissioning7. Implementation items

No.88

7.4 Organization of issues for the future(Table 2) Issues of criticality monitoring
Item

Subcriticality measurement

Predictive monitoring

Constant monitoring

Detection and termination of criticality

Purpose /function

－

－

－

－

Method・Preliminary verification of subcriticality
measurement method (addition ofverification data, blind test)・Confirmation by pilot application in
gradual retrieval
・Confirmation by estimation with the PCV
gas monitor and by cross-checking withsampling analysis
・Judgment of whether the Feynman−αmethod should be applied (different use
from 1/m)－

・Identification of the location when acriticality event other than theprocessing area is detected with thenoble gas monitor, and proposed
methods of countermeasures(measures of recovery to Level 1)

－

－

－

－

Measurement
frequency－

Measurement ・Development of monitoring plan withposition
neutron flux distribution

Measurement ・Setting of required count to obtaintime
significant measurement results
Method ofsetting
judgment
criteria

・Confirmation of validity of margin until
criticality－

・Method of confirmation of processing
volume in the field・Confirmation of uncertainty ofprocessing volume in the phase before
large-scale retrieval
・Development of monitoring plan withneutron flux distribution
・Setting of measurement position
considering impact of higher modechanges in neutron flux distribution－－

・Reduction of uncertainties attributable toUncertaintyℓ (prompt neutron life)
expected in the Examples: Criticality monitoring by α・Impact after absorbent injection
measurement
eigenvalue(analysis evaluation of KUCA test data)
result・Impact after absorbent injection (analysis
evaluation of KUCA test data)

・Setting of average moving time
according to equipment noise impact－
and neutron flux in the field・Confirmation of preconditions(uncertainty of processing amount,・Criticality detectability of vent pipe,
processing volume, etc.) of
S/C (study in cooperation with theeffectiveness evaluation (confirmation
system side)in the phase before large-scale
retrieval)－－©International Research Institute for Nuclear Decommissioning7. Implementation items

No.89

7.4 Organization of issues for the future(Table 2) Issues of criticality monitoring (continued)
Item ()

Subcriticality measurement

Predictive monitoring

Constant monitoring

Detection and termination ofcriticality

Response todeviation from
judgment criteria－－

・Organization of procedures in
the field

・Incorporation of shutdown
capacity requirements into theboric acid water system design

Response to ・Method of remote calibrationequipment failure using neutron source

Other

・Study of confirmation method ・Study of confirmation methodof operation in the field andof operation in the field andmethod of sensitivity
method of sensitivity
calibration.
calibration.
・Method of remote calibration ・Method of remote calibration
using neutron source
using neutron source－

・Measures when measurement
cannot be performed
・Effectiveness evaluation
Method of confirming validity
Explanation of validity ofof measurement with the YPORCAS-F code(*)
value curve
・Measures when criticality is
Method of differentiating the
not terminated・Incorporate in a framework socauses of not being able to ・Incorporate in a framework so
To be studied with systemthat the following procedures
perform measurementthat the following procedures
and safety personnel
can be implemented without
between subcriticality depth
can be implemented without
including response withhaving criticality engineers.
(outside measurement range) having criticality engineers.
temporary equipment
and other reasons (not・Incorporate in a framework someasurable).
that the following procedures
can be implemented without
・Incorporate in a framework sothat the following procedures
having criticality engineers.
can be implemented without
having criticality engineers.

(*) In order to confirm validity of PORCAS-F code, comparisons have been made with criticality transient experiment simulating solution system (TRACY experiment) andcriticality transient experiment simulating fuel rod system (SPERT experiment). However, it is different from conditions with granular and lump-like forms, such as fuel debris,being cooled by natural convection. No cases of experiments simulating such criticality behavior have been found, and it is considered difficult to implement such experiment. A
method of confirming validity without experiments is the method of conducting sensitivity analysis based on PIRT analysis and evaluating uncertainties expected in analysis
results.
©International Research Institute for Nuclear Decommissioning7. Implementation items

No.90

7.4 Organization of issues for the future(Table 3) Issues of neutron absorbent
Item

Boric acid water (normal use, emergency injection)

Purpose / function ・For normal use and emergency. Development of usage method.

Non-soluble neutron absorbent
・Development of respective usage methods as multiple usage ispossible.
(hereinafter is an example of the “case of using near-critical fuel debris”)

Method・Materialization of required equipment for normal and emergency
・Confirmation of remote operation combined with robot arm andinjection
prototype of injection equipment
・Impact assessment of other equipment when using boric acid・Confirmation by pilot application in gradual retrieval
water, study of impact mitigation measures

Injection
frequency

・(Normal use) Setting of operation concentration taking into
account changes in concentration during operation
・(Emergency injection) Setting of representative concentration
required for termination of criticality event

Injection position

・Study of criticality prevention measures at a position where
coolant is not applied

Injection time・(Emergency injection) Study of equipment to set and realize
injection time that keeps criticality impact within the permissible
range

Method of setting ・(Normal use) Setting of method and interval of boron
injection amount concentration monitoring
Backup with neutron flux constant monitoring
Response toshortage of・Method of maintaining concentration (study of response to
injection amount decreased concentration)
Effectivenessevaluation

・Prepare the approach of determining absorbent type, injection amountand range for the fuel debris form to be processed.
None in particular(If the predetermined amount and range are not achieved, repeatedly
inject absorbent until they are satisfied.)

・Analyze criticality evaluation accordingly incorporating fuel debris composition and form gained through internal investigation andsampling, confirm effectiveness of absorbent, and revise required concentration and injection amountResponse to・Study of alternative means of concentration monitoring, andequipment failure boric acid supplementation method
Other

・In order to set the absorbent injection frequency less than once a day,confirm the processing range of fuel debris retrieval per day anddetermine the capacity of injection equipment for sufficient amount of
injection.
・Approach of determining absorbent type and injection amount for thefuel debris form in the processing range pre-planned per day
・Confirm that absorbent can be injected with the same positioning
accuracy as fuel debris processing equipment.
・Establishment of method of confirming the injected range by visualinspection
・Confirm the time until injection equipment is transported to the field by
remote operation using robot arm. Additionally, confirm the time fromcollecting equipment after injection to completing preparations to load
absorbents for next use.

・Method of confirmation of operation in the field
・Method of collecting failed injection equipment and replacing withspare equipment©International Research Institute for Nuclear Decommissioning
・Provision of information to waste PJ regarding waste after use

No.91

8. Specific goals for achieving the purpose of the project(2) On-site operation methods for criticality approach monitoring technology and neutron
absorbent technology
On-site operating procedures for criticality approach monitoring with
① Development of on-site operatingprocedures
neutron detectors are developed.
Subcriticality measurement performance of neutron detectors isevaluated and incorporated in the on-site operation procedures.
On-site operating procedures for non-soluble neutron absorbents
are developed.
(Target TRL at end: Level 4)
Data on impact on the drying process of fuel debris with solidified② Development of solidified absorbent
technology
absorbent (water glass) attached is acquired.
(Target TRL at end: Level 4)

Above goals have been achieved.

©International Research Institute for Nuclear Decommissioning

No.92

Reference
Reference 1 Criticality control method during largescale retrieval
Reference 2 Basic requirements of criticality control
Reference 3 Explanation of terms©International Research Institute for Nuclear DecommissioningReference 1 Criticality control method during large-scale retrieval
No.93(normal use of boric acid water)
Inject boric acid water in advance and perform retrieval work in boric acid water
Level 1 (Normal state)
Prevention of abnormality = criticality prevention

keff

Level 2 (Abnormal state)
Prevention of spread ofabnormality = prevention criticality
accident

Level 3 (Accident)
Mitigation of impactof accident

Functional
requirement

1.0
1mSv

Site boundary dose
(Prevention of reaching criticality)(Maintenance of subcriticality)
Drastically mitigated or eliminatedSubcriticality is
maintained
within thecontrol
standards

Retrieval work restriction
Maintenance ofsubcriticality with boric
acid water, and monitoringof boron concentration

Failure(criticality approach occurs)

Prevention ofcriticality bydetecting criticality
approach morethan expected andstopping retrieval
work

Only constant monitoring during processing
Criticality approach monitoring (neutron detector)

Failure (criticality occurs)
Stopping of retrieval work
Success (recovery to normal state by removing areas approaching
criticality, by deepening subcriticality with increased boron
concentration)

Success (recovery to normal state by removing critical areas)

5mSv

Detection of criticality(gas sampling system,
neutron detector)
Emergency boric acidinjection

Success (recovery to normal state by removing areas where criticality occurred, after termination of criticality)

Failure (normal
level exceeded)

Prevention of
external impact
greater than thenormal level, bydetecting criticality
and promptly
stopping operation

Additional injection of boric acid
Prevention ofwater
Decreased water level due tooverexposure offeedwater shutdown
general public
Suppression of release ofradioactive gas
and workers©International Research Institute for Nuclear DecommissioningReference 1 Criticality control method during large-scale retrieval(falling of heavy weight object)
Level 1 (Normal state)
Prevention of abnormality = criticality prevention

Level 2 (Abnormal
state)

Level 3
(Accident)
Mitigation of
impact ofaccident

Prevention of spread ofabnormality = prevention
criticality accident

keff

No.94

Functional
requirement

1.0

Site boundary dose
(Maintenance of subcriticality)
Prevention of falling
objects

1mSv

5mSv

(Prevention of reaching criticality)

Subcriticality
is maintained
within thecontrol
standards

Failure (criticality
occurs)

Maintenance ofsubcriticality with boric
acid water・Monitoring of boron
concentration(Option)

Detection ofcriticality(Gas sampling system)

Success (recovery to normal state byremoving critical areas)

Emergency boric
acid injection

Success (recovery to normal state by removing areas where
criticality occurred, after termination of criticality)

Failure (normal
level exceeded)

Additional injection of
boric acid water
Decreased water level
due to feedwater
shutdown
Suppression of releaseof radioactive gas

Prevention of
external impact
greater than thenormal level, bydetectingcriticality andpromptly
stopping
operation

Prevention ofoverexposureof generalpublic andworkers©International Research Institute for Nuclear DecommissioningReference 1 Criticality control method during large-scale retrieval(accumulation of chip powder)
Level 1 (Normal state)
Prevention of abnormality = criticality prevention

keff

Level 2 (Abnormal state)
Prevention of spread ofabnormality = prevention
criticality accident

No.95

Level 3 (Accident)
Mitigation of impactof accident

Functional
requirement

1.0

Site boundary dose
(Maintenance of subcriticality)

1mSv

Collection of chippowder

Subcriticality is
maintained within
the control
standards

Prevention of outflow ofcooling water bycontrolling water level

Failure(criticality approach occurs)

Maintenance ofsubcriticality with boric
acid water, andmonitoring of boron
concentration

Gas monitor / neutron monitor (constant monitoring)(Option)

5mSv(Prevention of reaching criticality)

Prevention ofcriticality bydetecting criticality
approach morethan expected andstopping retrieval
work

Criticality approach monitoring(neutron detector / PCV gas monitor)

Failure (criticality occurs)
Stopping of retrieval work
Success (recovery to normal state by removing areas
approaching criticality, by deepening subcriticality withincreased boron concentration)

Success (recovery to normal state by removing critical areas)

Detection of criticality(Gas sampling system)
Emergency boric acidinjection

Success (recovery to normal state by removing areas where criticality occurred, after termination of criticality)

Failure (normal
level exceeded)

Prevention of
external impact
greater than thenormal level, bydetecting criticality
and promptly
stopping operation

Additional injection of boric
Prevention ofacid water
Decreased water level due tooverexposure offeedwater shutdown
general public andSuppression of release of©International
Institute for Nuclear
Decommissioning
workers
radioactiveResearchgasReference 1 Criticality control method during large-scale retrieval

No.96

Supplemental explanation of defense in depth・Response when criticality is not terminated regardless of boric acid water injection
In past criticality control studies, criticality prevention as well as termination andmitigation of external impact of criticality were mainly studied, but it is desirable toprepare additional means for termination and mitigation of external impact whenassuming criticality is not terminated regardless of boric acid water injection (specificstudies are future issue).
In past discussions, there were comments saying to include mobility response
instead of only addressing all additional means with equipment. Thus, it is desirable
to analyze the possibility of criticality not being terminated upon studying equipmentthat achieves up to Level 2, and study measures including response using temporary
equipment (future issue).
・Measures after Level 3
Additional measures when criticality is not terminated regardless of boric acid water
injection are tentatively indicated as Level 3 measures, but essentially, specificmeasures should be discussed upon studying definition of levels and events to behandled as a whole (including criticality) (after Level 4, it will be primarily addressedby operators).
©International Research Institute for Nuclear DecommissioningReference 2 Basic requirements of criticality control

No.97

Prevention of reaching criticality
・Below is an excerpt of predictive monitoring procedures.
Confirmation of whether criticality is approached and determining whether processing can bestarted, based on neutron flux measurement before processing.
Requirements of criticality control
Item

Requirement
Requirement (details)2-3) Prevention of reaching criticality
No details
Even if criticality is approached, monitor
this and terminate the event before
reaching criticality.
2-3-1) Criticality approach monitoring
2-3-1-1) Have means to monitor criticality(predictive monitoring)approach during work
2-3-1-2) Measurement position of neutron
flux is in a position that can appropriately
monitor criticality approach.
2-3-1-3) Measure neutron flux after fueldebris processing, and start work upondetermining that criticality will not bereached in the next processing.

2) Criticality
prevention

2-3-1-4) Criticality approach monitoring
frequency
Measure intervals of addition reactivity
stipulated in 2-2-2.
2-3-1-5) Approach on judgment criteria of
criticality approach
Criticality control standards reduced by atleast one count of addition reactivity,
margin, measurement uncertain andevaluation uncertainty.
2-3-1-6) If criticality approach judgment
criteria is violated, perform cause analysis
without starting the next work. Implement
pre-developed procedures for work restart.
Support procedures for work restart.

(*) Since this table is an excerpt, 2-3-2 and 2-3-3 are omitted.

Specific development: Lower lateral retrieval work (without normal use of boric
acid water)
Over-processing of fuel debris
Even if criticality is approached, implement measures to monitor this and(*)
terminate the event before reaching criticality. Details stipulated in 2-3-1)-2-3-3) .

Use predictive monitoring neutron detector to be loaded to the retrieval arm.
Perform measurement on an “island” of fuel connected to the processing
position.
Measure neutron flux near the fuel debris processing position after fuel debris
processing. Derive post-processing keff based on initial keff (subcriticality
measurement result), and perform the next processing after confirming that it isbelow the judgment criteria. If keff cannot be measured, make the decision withinverse multiplication factor (1/m).
Perform criticality approach monitoring at every processing volume of 4,000cm3.

Judgment criteria is the value reducing criticality judgment criteria of keff (0.95)
by 5 times the addition reactivity and measurement uncertainty as uncertainty ofprocessing volume. Even in the case of 1/m, it is set from criticality control
standards.

If criticality approach judgment criteria is violated, do not start the next work.
Confirm whether criticality is approached by measuring subcriticality. Confirm
the margin until criticality. Based on the result, study continuation of retrieval
with smaller processing volume limit or switch to retrieval with normal use of
boric acid water.
©International Research Institute for Nuclear DecommissioningReference 2 Basic requirements of criticality control

No.98

Predictive monitoring procedures
Item

Requirement

Approach

1) Purpose /function・Detect criticality approach, and stop processing work to prevent criticality.
1) Detect excessive criticality approach (keff=0.95, etc.) and stop work2) Detect signs of criticality approach early on by grasping greater changes than
expected・2) is a preliminary analysis that captures greater changes than expected, in order toensure 1).

2) Method3) Measurement
frequency

4) Neutron
measurement
position・Measure neutron flux with the neutron detector near the processing position before
・Criticality approach evaluation is based on the neutron source multiplication method
fuel debris processing, extrapolate the amount of change from the previous
generally used for criticality approach monitoring.
measurement, and evaluate the possibility of reaching criticality in the next
・Perform monitoring with 1/m if initial keff is not provided.
processing to determine whether processing can be started.
・Perform monitoring for each processing equivalent to the volume equivalent toaddition reactivity 0.5%Δk/k. (for example, for every fuel debris processing volume
・Addition reactivity 0.5%Δk/k is ①sufficiently small so that criticality is not reached inof 4,000cm3)
one processing, and ②set to prevent prompt criticality even when fuel debris
・Set the processing volume taking into account uncertainties associated withprocessing is performed and reactivity is added from a critical state.
processing (set uncertainties to be expected based on various information obtainedin prior small-scale retrieval).
・Since the consecutive lumps of fuel debris have the same subcriticality and the same
relative change in neutron flux, measurement points can be fixed regardless of the・Lateral direction: Perform measurement within 50cm in the lateral direction (in theprocessing position.
case of partial submersion method) from the processing position with the “lump” ofDistance of 50cm between the processing position and measurement point is asame fuel debris as the fuel debris processing position. If lateral direction distancetentative value based on KUCA and other experiences.
from the measurement position increases due to progression of processing, re・Neutrons are not necessarily emitted equally in all directions. Considering the detector
measure the reference count rate and move the measurement point.
could go behind the robot arm, simulate neutron measurement as necessary to
determine detector placement.

5) Neutron
・Measurement time is set so that uncertainties to be expected in judgment criteria aremeasurement timesatisfied (assumed at around approx. 10min.)・Statistical uncertainties of neutron flux count is set from the allowable uncertainty of
the ratio to reference count rate with standard deviation σ=√ (count value).

6) Method of setting
judgment criteria

When using keff) Set by reducing addition reactivity (0.5%Δk/k×margin) and
uncertainty of the next processing from the criticality judgment criteria (0.95)
When using 1/m) Set by reducing uncertainty from 1/m=0.5

See left7) Method of setting
uncertainty

・Statistical uncertainty of count value, reproducibility of measurement point(misalignment), impact of noise, fluctuations and attenuation of detector sensitivity,
etc.

・Confirm the impact of noise in pre-mockup tests of retrieval equipment.

8) Response todeviation from
judgment criteria

・Stop processing work due to possibility of criticality approach.
・Re-measurement and subcriticality measurement extending the measurement time・Based on the result, study measures such as downsizing of processing size (shorteningof monitoring interval) and application of neutron absorbent

9) Monitoring
・Evaluate assumed changes according to initial keff.
method in normal ・Cause analysis when exceeding the normally expected magnitude of change・Work can be continued since criticality risks do not occur unless judgment criteria of 6)
state (deep・Convene criticality control experts and start analysis, while continuing retrieval work.
is not violated
subcriticality)10) Response toneutron detector
failure・Position proper operation of neutron detector as a condition of retrieval work.

・Neutron detector failure itself is not a cause of criticality event, but fuel debris retrieval
cannot be performed since monitoring is not possible in the situation.

11) Predictive
monitoring when ・Criticality approach prevention (maintenance of subcritical state) when using neutron ・It is better to have regular neutron flux level monitoring as backup of concentration
neutron absorbent
absorbent can be replaced with absorbent concentration management.
monitoring.
©International Research Institute for Nuclear Decommissioning
is applied

No.99Reference 3 Explanation of terms
Term

Explanation

Reactor noise method

When measuring neutrons from nuclear fission reaction with a detector, neutron count rate
within a certain time period is observed to fluctuate around the average value. Such temporalfluctuation in neutron count is referred to as reactor noise. The reactor noise method is amethod that analyzes such noise to obtain core characteristics of the system.

Feynman−α method

It is a type of reactor noise method. It is a method that investigates the average and dispersion
ration of the number of detected neutrons in a given time width T, and obtains prompt neutron
attenuation constant α by plotting against time width T. It was proposed by R.P.Feynman.

Non-soluble neutron
absorbent

Neutron absorbents have the property of absorbing neutrons to make it less likely to become
critical. Non-soluble neutron absorbents have “non-soluble” or “insoluble” properties, incontrast to soluble neutron absorbents such as boric acid water.

Neutron detector- Corona discharge type

It is a detector with B10 which reacts with neutrons applied inside the cylindrical tube, and ischaracterized by applying voltage between the center line and cylindrical tube and operating ina high voltage region where corona discharge occurs. It is a technology by Russian company
FEO (former RosRAO), and a prototype was made in a relevant subsidy project.

- SiC semiconductor type

It is a thin semiconductor detector doped with B which reacts with neutrons. Since it has low
gamma-ray sensitivity and can reduce the shielding body, compact and lightweight highsensitivity detector can be configured. It is a technology developed by British company IPL,
and a prototype was made in a relevant subsidy project.

- Multicell He-3 type

It has a structure with multiple anode and cathode cores arranged inside the cylindrical tube
filled with He-3 gas, and is designed aiming for high neutron detection sensitivity. A prototype
was made in a relevant subsidy project by domestic company CETD.

©International Research Institute for Nuclear DecommissioningNo.100

END©International Research Institute for Nuclear Decommissioning