Subsidy Project of Decommissioning and Contaminated Water Management

Development of Safety Systems(Liquid/Gas Phase Systems and Criticality Control Technology)
Accomplishment Report for FY2021(1) Liquid/Gas Phase Systems(2) Criticality Control Technology
August 2022
International Research Institute for Nuclear Decommissioning(IRID)All rights reserved.
©International Research Institute for Nuclear Decommissioning (IRID)0(1) Liquid/Gas Phase Systems

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

75

No.1

Table of Contents
1. Purposes and goals of subsidy projects
2. Overview of subsidy projects
3. Previous fiscal year project results and remaining issues
4. Input/output information
5. Implementation schedule
6. Project organization
7. Implementation details① Development of soluble α-nuclide removal technology② Development of Reverse Osmosis (RO)-concentrated water treatment
technology③ Development of secondary waste treatment technology
1) Characterization of liquid waste generated from particle removal system
2) Selection of dehydration technology for sedimentation sludge

8. Specific objectives for achieving the purposes of technologicaldevelopments©International Research Institute for Nuclear Decommissioning

1

No.2

1. Purposes and goals of subsidy projects[Purpose of Development of Safety Systems (Liquid/Gas Systems, Criticality Control Technology)]
It is assumed that nuclear fuel has melted along with the reactor internals at Tokyo Electric Power Company(TEPCO) Holdings, Inc. Fukushima Daiichi Nuclear Power Station (NPS) and exists in the form of molten fueldebris in the Reactor Pressure Vessel (RPV) and the Primary Containment Vessel (PCV).
The fuel debris accumulated inside the RPV and PCV is estimated to be currently in a sub-critical state;
however the plant itself is in an unstable condition unlike its initial design, since the Reactor Building (R/B), RPV,
PCV, etc. have been damaged due to the accident. Therefore, it is necessary to retrieve the fuel debris in order tomaintain the sub-critical state, and to prevent diffusion of radioactive materials.
Against this background, this project is intended to conduct studies based on the “Mid-and-Long-Term Roadmap Towards Decommissioning of TEPCO’s Fukushima Daiichi NPS” (hereinafter “Mid-and-Long-Term Roadmap”), aiming towards the implementation of large-scale fuel debris retrieval in coordination with theengineering and project management activities undertaken by TEPCO. The development results of this projectwill be used in TEPCO’s engineering activities.
The purpose of this project is to advance the level of Japanese science and technology by supporting thedevelopment of technologies that contribute to the safe decommissioning of the Fukushima Daiichi NPS andhandling of contaminated water, based on the Mid-and-Long-Term Roadmap and the “FY2021 Decommissioning
Research and Development Plan” (the 86th Secretariat Team Meeting for Countermeasures for
Decommissioning and Contaminated Water Treatment).
Specifically, technologies for removal of soluble α-nuclides that are considered to be eluted in circulating
cooling water from fuel debris, treatment of RO-concentrated water (*), treatment of secondary waste, and onsite operating methods for criticality approach monitoring and neutron absorber will be developed.
*Liquid waste with a high impurity concentration generated by Reverse
Osmosis filter (RO filter) removal of impurities from cooling water[Overall development goal]
Development goals are to develop element technology necessary for systems and to conduct tests fortechnology for ensuring safety for further increasing the scale of retrieval of fuel debris and the reactor
internals based on the R&D results obtained so far.
©International Research Institute for Nuclear Decommissioning2

1. Purposes and goals of subsidy projects

No.3

The material in this report deals with the following items.
The objective of each item’s implementation is shown.

(1) Liquid/gas systems① Development of soluble α-nuclide removal technology- Element test assuming fuel debris retrieval work
Selection of candidates for adsorbents that can remove α-nuclides (Pu, U, etc.) during fuel debris
retrieval work, taking into account the gas phase environment inside the PCV (nitrogen
atmosphere).
② Development of Reverse Osmosis (RO)-concentrated water treatment technology- Selection of adsorbent and condensing agent
Selection of powder adsorbents and flocculants needed to treat RO-concentrated water throughelement tests and devise a treatment method for RO-concentrated water.
③ Development of secondary waste treatment technology- Investigation of pretreatment technologies
1) Characterization of liquid waste generated from particle removal system
Performance assessment tests on the filter components of the particle removal system are
conducted. Then the properties of the filtered liquid waste discharged into the sludge collection
system are consolidated.
2) Selection of dehydration technology for sedimentation sludge
Applicable dehydration technologies for volume reduction treatment of sedimentation sludge
from sludge collection equipment are selected from the literature research. Then a sludge©International
ResearchandInstitutefor Nuclear Decommissioning
treatment method is planned based on the results of the literature
researchelement
tests.
3

No.4

2. Overview of subsidy projects

System configuration diagram of the liquid and gas systems under review for the subsidy project is shown.
In this year’s subsidy project, a soluble nuclide removal equipment and a sludge dehydration and stabilization
treatment equipment will be developed.
② Development of RO-concentrated water
treatment technology
a. Selection of adsorbent and condensing agent
b. Examination of applicability to actual
equipment① Development of soluble α-nuclide removal technology
a. Examination of soluble α-nuclide removal tests assuming use of actual liquid
b. Element test assuming fuel debris retrieval work
Pretreatment
facilityGas phasefilterTo exhaust port

Boric acid (silo)*Use only when using boric acid regularly

Exhaust ventilator: Development items foreach facility: Facilities for this project

Boric acidcollection tank

CoolerParticle
removalfilterA

Atmosphere

Soluble nuclide
removal adsorption
tower

Tank (including
boric acidadjustment)

Supernatant water/
permeable waterP
P

Particle
removal filter

P

Drain water/
backwash waterA*

B

To existingwater
treatment
facility

Soluble nuclide
removaladsorption tower

Concentrated Supernatantwater water
Sludge collection
sedimentationequipment

Supernatant water

Sludge collection
sedimentationequipment

SludgeSupernatant water

SludgeSludge (filter)
Sludge collection
collection totalC
sedimentationA
equipment
filtering facility*: Particle removal; used as drainage water collection from roughing treatment.

*RO filter: reverse osmosis
filterSoluble nuclide
removal RO filter

Nitrogen

Groundwater

Collected boric acid solution

Boric acidcollection
facilityB

Drain water/
backwash water

Debris retrieval water

Permeable water③ Development of secondary waste
treatment technology
a. Investigation of pretreatment
technologies
b. Examination of applicability to actual
equipment

To existingwater
treatment
facility

SludgeDehydration/
stabilization
treatment facility

Dehydration/
stabilization
treatment facility

C

B
SludgeC
To wastestorage

Dehydration/
stabilization
treatment facility

C

C

To
canisterstorage

Figure. Liquid and gas systems during fuel debris retrieval (based on subsidy projects)
Note: This conceptual diagram is an example of©International
facility configuration(in the caseof noDecommissioning
leakage from PCV)
Research Institutefor Nuclear

4

No.5

3. Previous fiscal year project results and remaining issues① Development of soluble α-nuclide removal technology
No.

1

Items
Nuclidestargeted forremovalEfforts and results of the previous project*: For examination in this project
Remaining issues

The effects of powder generation for 15 nuclides of 5 elements (U, Np,
Pu, Am, and Cm) were verified by tests. It was verified that the
Dissolution and adsorption behavior of α-nuclides for U,
necessity of adsorption and removal of U is low because its solubility
Np, Pu, Am, and Cm, respectively, assuming a nitrogennear neutral pH 5 to 9 under atmospheric conditions and theenvironment in the PCV
concentration in the PCV stagnant water are about two orders ofmagnitude lower than the announcement density.

2

The evaluation was conducted assuming elution of seawater
Quality of watercomponents, elution of concrete components, and injection of sodium
to be treatedpentaborate, a criticality inhibitor.

3

If an adsorption/removal system is installed in the facility that circulates
and cools the PCV stagnant water, a treatment flow rate of 10 m 3/hcan be set to maintain a lower concentration of α-nuclides. If an
Treatment flow
Setting of treatment flow rate considering facility operating
adsorption/removal system is installed outside of the facility thatrate
schedule
circulates and cools the PCV stagnant water, only the volume of water
increase from fuel debris retrieval work is treated so the water balancecan be maintained if the treatment flow rate is 3.5 m 3/h or higher.

4

The dissolution behavior of each α-nuclide was determined based on
the water quality that is assumed under conditions of elution ofConcentration
seawater components and the injection of sodium pentaborate, aof α-nuclides incriticality inhibitor. Regarding the effect of concrete elution, it was
assumed water
verified that evaluation in a nitrogen atmosphere similar to the PCV
qualityenvironment is necessary because in an atmospheric environment αnuclides co-precipitate with the formation of calcium carbonate.

5

Requirementsforconcentration
reduction

Effects of oil, coating, rust inhibitor, and non-soluble
criticality inhibitor components

Dissolution and adsorption behavior of α-nuclides at highpH assuming elution of concrete components

Necessary DF setting accounting for the concentration of
The target DF for reduction of public exposure effects is 100, and theparticulate α-nuclides
concentration reduction target for transferring to water treatment facility
Setting of concentration reduction targets for individualis the announcement density. (No progress)
nuclides* Subsidy Project of Decommissioning and Contaminated Water Management in the FY2018 Supplementary budgets“Development of Technology for Retrieval of Fuel Debris and Reactor Internals”(2) (ii) Technical development of treatment of fuel debris and deposit① Removal technology for soluble nuclides in circulating cooling water② Treatment technology for deposits collected from inside PCV©International Research Institute for Nuclear Decommissioning
Hereinafter referred to as “the previous project.”

5

No.6

3. Previous fiscal year project results and remaining issues① Development of soluble α-nuclide removal technology
No.

Items

Efforts and results of the previous project

Remaining issues

Adsorbent
application

As a result of the evaluation of α-nuclide adsorption performance byimmersion tests, activated carbon, zirconium phosphate, and titanicacid were selected as candidate adsorbents for application. At thisstage, since the adsorbent with a large equilibrium adsorption capacity Method for evaluating adsorption/removal performance
was selected through immersion tests, it is necessary to verify its
when applied to water flow treatment
performance when applied to water flow treatment. In addition, since it
is not possible to conduct high quantity α-nuclide circulation tests in alaboratory, the evaluation method is an issue.

Adsorption
tower
configuration

Each tower has a series of activated carbon for removing colloidal α- Interactive effects of multiple α-nuclides
nuclides, zirconium phosphate for adsorbing Pu, Am, Cm, Np, titanic
Effects of adsorption inhibitor componentsacid for adsorbing Np, etc., with activated carbon at the top. Assuming
merry-go-round operation, each tower is counted as two towers. The
tower configuration should be reviewed taking into account the impact Review of tower configuration based on removal
assessment of each nuclide simultaneously present in solution, as
performance and replacement frequency during water flow
well as the removal performance and replacement frequency during
treatment
the water flow treatment.

8

Replacementtiming/Replacement
frequency

All of the adsorbents selected as candidates for application have low
cesium (CS) adsorption performance, so the risk of replacement due
to an increase in the tower surface dose is low. Although it is typical toconduct a circulation test to determine the replacement frequency due Evaluation method of adsorbent replacement frequency
to the deterioration of adsorption performance, high α-nuclide quantity
makes such a test impossible in a laboratory setting. Thus, theevaluation method remains an issue.

9

Replacement
methodWhether replacing the entire adsorption tower or only the adsorbent,
Operation at the time of replacementin either case it is necessary to design equipment according to thetreatment policy of the used adsorbent (long-term storage for the time Handling of used adsorbents
being).

Tower size

At a treatment flow rate of 10 m3/h, the size is about the same as theadsorption tower of the multi-nuclide removal system (outside
diameter: approx. 1 m, height: approx. 2.5 m).

6

7

10

Adsorption tower design with input of shielding policy,
replacement method, etc.

©International Research Institute for Nuclear Decommissioning

6

No.7

3. Previous fiscal year project results and remaining issues② Development of RO-concentrated water treatment technology③ Development of secondary waste treatment technology
For examination in this project
No.

Items

Efforts and results of the previous project

Remaining issues
Candidate components for the intermediate filter will beselected by literature research. Applicability will be verified
through element tests and the properties of the generated
liquid waste should be determined.

1

Water quality fortreatment

The assumed water quality of the drain water and backwash water
discharged from the particle removal system, which is the input
water to the sludge collection system, was regulated based on the
results of element tests and filter tests of the fuel debris retrieval
method project.
However, since the intermediate filter of the particle removal
system has not yet been selected, it is necessary to selectcandidate components and evaluate the properties of thegenerated liquid waste. In addition, in the previous fiscal year’s
element tests, evaluation was only possible when sphericalparticles were passed through, and it is necessary to evaluate thefilter behavior and the properties of the generated liquid waste
when non-spherical particles are passed through.

2

Treatment flowrate

The treatment flow rate requirement of the sludge collection system
was assessed based on the operating schedule of the fuel debris
retrieval method, the filter test results, and the liquid system flow
rates.

If the assumed input water quality of the liquid systemchanges as a result of future studies, this will be reflected in
the treatment flow rate requirement of the sludge collection
system.

Treatment
methodCoagulation-precipitation treatment was selected as the treatment
method for filtered liquid waste and RO-concentrated water, andusable flocculants were selected through element tests. The use of
sedimentation tanks in the coagulation-precipitation treatment
procedures were also examined.
It is necessary to examine the applicability to actual equipment
for the flocculation sedimentation treatment of ROconcentrated water in the case of sedimentation tank usage. In
addition, it is necessary to investigate treatment procedures tosimplify operations and reduce the amount of sludge generated.

Adsorption using powdered adsorbent and flocculationprecipitation using flocculant as treatment methods for ROconcentrated water will be selected. Usable powderedadsorbents and flocculants will be selected.

Sedimentation
sludgeproperties andthe amount
generated

The amount of sedimentation sludge generated was estimatedbased on flocculation sedimentation test. The amount of
sedimentation sludge generated from RO-concentrated water
was rated high due to the high treatment flow rate and the highmoisture content of sedimentation sludge, and there were concerns
about the complexity of the discharge process and the loadincrease at the discharge point. Therefore, it is necessary to
reduce the amount of the sedimentation sludge.

3

4

Element tests will be conducted to assess the impacts on theparticle removal system when treating non-sphericalparticles.
If it is assessed that there is a predominating impact on the
behavior of component differential pressure increase,
recovery rate, or the amount of liquid waste generated, thiswill be reflected in the assumed input water quality of thesludge collection system.

In order to reduce the amount of sedimentation sludge
generated, each step of the RO-concentrated water
treatment procedures will be reviewed. Through element
tests, the effects of reducing the quantity of flocculant andpowder adsorbent, and optimization of stirring conditions,etc. will be examined.
The application of dehydration treatment to sedimentation
sludge will be analyzed, applicable dehydration technology
will be selected via literature research, and the applicability
to actual equipment
be assessed
element tests.
©Internationalwill
Research
Institute forthroughNuclear Decommissioning

7

3. Previous fiscal year project results and remaining issues

No.8② Development of RO-concentrated water treatment technology③ Development of secondary waste treatment technology
No.

Items

Efforts and results of the previous
project

Remaining issues

5Supernatantliquidproperties

The quantity of suspended solids (SS) and othercomponents remaining in the supernatant after solid-liquid
separation were sorted.

The required acceptance conditions of the soluble nuclide removal
equipment and the existing water treatment facility will be sorted
out, and the necessity of neutralization treatment and removal ofneutralized salt and trace amounts of residual SS components will
be examined.

Operatingmethods

The process of coagulation-precipitation treatment offiltered liquid waste and RO-concentrated water using asedimentation tank was examined based on the operatingschedule of fuel debris retrieval method and element test
results. An operating cycle in which one batch process iscompleted per day was examined.

The results of this year’s element tests will be applied to the
coagulation-precipitation treatment process studied in the previous
project.

Equipment
specifications

Equipment tests verified the operability of thesedimentation tank and each component.
Since large particles were caught when opening andclosing the gate valve, it is necessary to reconsider the
specifications of the gate valve.

In element tests using RO-concentrated water treatment test
equipment to be conducted from this fiscal year onwards, eachmechanism of the sedimentation tank will be re-evaluated.
In addition, valves that can be applied to sedimentation tanks will
be examined.

8

Methods fordischarge of
sedimentation
sludgeThe discharge method of sedimentation sludge using asludge collection container was examined, and through anequipment test confirmed that this method can be applied
to actual equipment.
It was verified that it is possible to insert piping from the
top of the sediment tank and remove sedimentation sludge
by suction pump.
In the future, the required acceptance conditions of thesedimentation sludge discharge point will be sorted out
and sludge canister size and applicability of dehydration
treatment will be examined.

When the sedimentation sludge is treated and disposed of as
waste, it is discharged to the waste line, so it is necessary toconsider the final disposal of sedimentation sludge and apply
pretreatment. In the future, if a policy for the final disposal of sludge
is presented, the necessary pretreatment within the liquid system
will be examined and applied to the sedimentation sludge
discharge method.
When sedimentation sludge is stored in canisters, it is discharged
to the canister line, so it is necessary to apply pretreatment tosatisfy the requirements of long-term storage. The requirements forlong-term storage of sedimentation sludge will be sorted out andthe shape and moisture content of the sludge collection container
will be examined.

9

Handling of
sedimentation
sludgeAssumed properties of sedimentation sludge and theamount generated were arranged based on the results ofelement tests and the liquid system requirements.

Assumed properties of sedimentation sludge will be shared to both
the waste project and the canister project. Issues will be identified
respectively to each project.

67©International Research Institute for Nuclear Decommissioning

8

No.9

4. Input/output information
This project will be proceeded in collaboration with related projects such as theDevelopment of Technologies for Containing, Transfer and Storage of Fuel Debris
project and the R&D for Treatment and Disposal of Solid Waste project.
ID

Demand-side project

Provider-side projectContents (overview)

Information use

1

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

Development of Technologies
for Containing, Transfer andStorage of Fuel Debris

Acceptance conditions forcanisters

Examination of stabilization
treatment technology forsecondary waste

2

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

R&D for treatment anddisposal of solid waste

Acceptance conditions for
waste storage containers

Examination of stabilization
treatment technology forsecondary waste

3

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

Development of Analysis andEstimation Technology for Fuel
Debris Characterization

Results of examination on
formation behavior of fueldebris particles

Examination of properties ofparticles in circulating
cooling water

4

Development of Technologies
for Containing, Transfer andStorage of Fuel Debris

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

Properties of sedimentation Identification of issues in thesludge generated from liquid handling of fuel debris ingas systemslurry or sludge form

5

R&D for treatment anddisposal of solid waste

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

Identification of issues in theProperties of sedimentation
treatment and disposal of
sludge generated from liquidradioactive waste in slurry or
gas systemsludge form

*Also planned to discuss collaboration with TEPCO in the future.

©International Research Institute for Nuclear Decommissioning

9

No.105. Implementation schedule
Subcategory

Category

Plant operator
FY2021
FY2022(If there is asubcontractor or test
Apr May Jun Jul Aug Sep Oct Nov Dec Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec Jan Feb Mar
location, it is also listed)

Major milestones(1) Liquid/gas
systems①
Development ofsoluble αnuclide removal
technology

a. Conceptual study/
test planb. Test preparation/
arrangementsc. Preliminary tests
d. Element tests
e. Summary

Project steering meeting

Project steering meeting Interim report

Interim report

Interim report

Final report

Conceptual study/test plan(1) Hitachi GE
Subcontractor: None
Test location: Hitachi
Research Laboratory: Line of planning
Test preparation/arrangements: Line of results
Preliminary tests① Development of soluble α-nuclide
removal technology
A similar test was conducted at Hitachi
GE/Toshiba ESS.
Allocated tests for each nuclide andadsorbent.

Element test
Summary

Interim summary

Conceptual study/test plana. Conceptual study/
test planb. Test preparation/
arrangementsc. Preliminary tests(2) Toshiba ESS
Subcontractor: NFD
Testing location:
Toshiba ESS
Research Laboratory

Test preparation/arrangements

Preliminary tests
Element test

d. Element tests
e. Summary② Development of
RO-concentrated
water treatment
technology

a. Conceptual study/
test planb. Test preparation/trial
manufacturing of test
equipmentc. Preliminary tests(1) Hitachi GE
Subcontractor: Hitachi
Plant Construction,
Ltd.
Test location: Rinkaifactory

Conceptual study/test plan
Test preparation/test equipment manufacturing
Preliminary tests
Element test

d. Element tests
e. Summary③ Developmentof secondary
waste treatment
technology

a. Conceptual study

Interim summary(1) Hitachi GE
Subcontractor: Hitachi
Plant Construction, Ltd.
Test location: Rinkaifactoryb. Technical study/test
planc. Test preparation/test
Subcontractor: Misuzu
sample manufacturing Seikod. Element tests
Test location: Misuzu
e. Summary

Summary

Interim summary

Seiko researchlaboratory

Summary

Conceptual study
Technical study/test plan
Test preparation/trial manufacturing of test equipment

Element test
Interim summary

Summary*Hitachi GE: Hitachi-GE Nuclear Energy, Ltd.
©International
Research
Institute forCorporation
Nuclear Decommissioning*Toshiba ESS: Toshiba Energy
Systems& Solutions

10No.11

6. Project organization
Tokyo Electric Power Company Holdings, Inc.

➢ Coordination for on-site applicability

International Research Institute for Nuclear Decommissioning(IRID)➢ Coordination of general planning and overall technology
management➢ Coordination of technology administration including
technology development progress management

Development project
teams incooperation with thisproject

Hitachi-GE Nuclear Energy,
Ltd.

Toshiba Energy Systems &Solutions Corporation

Mitsubishi Heavy Industries,
Ltd.

Japan Atomic Energy
AgencyDevelopment of Fuel
Debris Retrieval Method[Element test, technicaldevelopment](1) Liquid/gas system development① Development of soluble α-nuclide
removal technology- Examination of soluble α-nuclide
removal test assuming the useof actual liquid (stagnant waterin the reactor building)- Element test assuming fueldebris retrieval work
② Development of RO-concentrated
water treatment technology- Selection of adsorbents andflocculants- Examination of applicability toactual equipment③ Development of secondary waste
treatment technology- Investigation of stabilization
treatment technology- Examination of applicability toactual equipment[Element test, technicaldevelopment](1) Liquid/gas system development① Development of soluble α-nuclide
removal technology- Examination of soluble α-nuclide
removal test assuming the useof actual liquid (stagnant waterin the reactor building)- Element test assuming fueldebris retrieval work

[Element test, technicaldevelopment](2) On-site operating methods forcriticality approach monitoring
technology and neutron
absorber technology- Formulation of criticality
approach monitoring procedures- Performance assessment of
neutron detector sub-criticality
measurement[Element test, technicaldevelopment](2) On-site operating methods forcriticality approach monitoring
technology and neutron
absorber technology- Development of solidified
absorber technology

Development ofTechnologies for
Containing, Transfer andStorage of Fuel Debris(2) On-site operating methods forcriticality approach monitoring
technology and neutron
absorber technology- Formulation of criticality
approach monitoring procedures(2) On-site operating methods forcriticality approach monitoring
technology and neutron absorbertechnology- Formulation of criticality
approach monitoring procedures- Performance assessment of
neutron detector sub-criticality
measurement- Formulation of operatingprocedures for usage ofabsorbers- Examination of the impacts of thefuel debris on drying process

R&D on Solid Waste
Treatment and Disposal

Development of Analysis
and Estimation
Technology for Fuel
Debris Characterization
Development ofTechnology for Detailed
Investigation inside PCVDevelopment ofTechnology for
Investigation inside the
RPVDevelopment of Analysis
and Estimation
Technology for Fuel
Debris Characterization
Development ofTechnology for Sorting ofFuel Debris andRadioactive Waste©International Research Institute for Nuclear Decommissioning

11

No.12

7. Implementation details① Development of soluble α-nuclide removal technology (element test
assuming fuel debris retrieval work)- Development history• In the FY2016 subsidy project, “Upgrading of Methods and Systems for Retrieval of Fuel Debris and Reactor
Internals,” site boundary exposure doses for liquid system during normal and accident conditions were evaluated.
The site boundary exposure dose was below the threshold value in both normal and accident conditions, but inboth cases the effects of Pu-based nuclides and daughter product (Pu-238, Pu-241, Am-241) were significant.

• Alpha-emitting nuclides (hereinafter referred to as α-nuclides) at existing nuclear facilities are not nuclides that are
discharged into the environment and the exposure due to α-nuclides should be as low as reasonably achievable.
Therefore, from the subsidy projects (FY2017/2018) “Upgrading of Methods and Systems for Retrieval of Fuel
Debris and Reactor Internals,” examination and evaluation of the effectiveness of removing α-nuclides in liquid
system have started. Candidate technologies are selected for each type of α-nuclide (non-soluble α-nuclides andsoluble α-nuclides), and element tests are conducted to evaluate the feasibility of the selected technologies.
• Adsorption removal and RO filters have been selected as candidate technologies for removal of soluble αnuclides; RO filters are capable of removing α-nuclides in principle. On the other hand, although adsorption
removal is a proven technology for the Fukushima Daiichi NPS contaminated water treatment, its applicability
depends on the adsorbent selected and the form of α-nuclides present in solution.
• The previous project (subsidy project for FY2019 to 2020) dealt with the development of adsorption removal
technology and adsorption tests under open-atmospheric conditions were conducted on Pu, U, Am, Cm, and Np
α-nuclides (the 5 nuclides targeted for removal). Candidate adsorbents were selected, but their actual
effectiveness in the event of PCV water quality fluctuation has not yet been evaluated.
• In addition, in order to verify the adsorption removal technology’s efficacy in removing α-nuclides, a verification
test using actual contaminated water is required, but plans for such a test are still under discussion.
©International Research Institute for Nuclear Decommissioning

12

No.13

7. Implementation details① Development of soluble α-nuclide removal technology(element test assuming fuel debris retrieval work)- Mapping out of this project[Previous projects (FY2019/2020)]

Soluble α-nuclide
removal equipment
Systemconfiguration/Reesta
blishment of
requirements(Plant operator engineering)
Constraints on equipment placementin the existing building (space, dose,etc.)

Arrangement of Liquid
system pattern

Preliminary engineering- System structure- System requirements

End of FY2021 Subject of report[Development plan for FY2023 onwards]
Equipment requirements (performance,
flow rate, maintenance, quality, etc.)

Verification test using actual contaminated water

Policy for applicationof actual equipment[This project (FY2021/2022)]

Adsorption waterflow test
Candidate adsorbents

Test plan (procedure
review, etc.)- Test procedure (after revision)- Adsorbents to be evaluatedPreliminary test Gas phase environment(Pu, U, Am, Cm, Need for simulation
Np)- Impacts of adhesion tocontainer- Impacts of powder generation

Adsorption
Candidate adsorbents
immersion test for application(Pu, U, Am, Cm,
Np)

Adsorption
immersion test

PCV environment
simulation testing plan

Evaluation
methods- Organization of gas phase
PCV environment- Preparatory test
- Test procedures- Adsorbents to be evaluatedPreliminary andadsorption tests insimulated PCV
environment
Need for actual contaminated water

Adsorption
performancedata, etc.

- Adsorbent application- Water flow conditions, etc.
Conceptual study of
α-nuclide adsorption
and removal
equipment

Form and dissolved
amount of present α

Preliminary tests- Contaminated water to be
used- Test methods, etc.

Test plan using actual
contaminated water

Examination andevaluation of
applicability to actual
equipment
Basic design/detailed
designApplication for
permission andauthorization
Component
manufacturing

Adsorption performance data
Design issues
Conceptual study of
α-nuclide adsorption
and removal
Issues related to performanceequipment

assessment

Examination ofevaluation methods
for solving issues

Organization of dataon actual contaminatedwater quality

Examination of theeffectiveness of test
evaluation using actual
contaminated water

On-site construction
and trial runs

Operation/maintenance©International ResearchInstitute for Nuclear Decommissioning
Figure Overall flow of development of soluble α-nuclide removal
technology

13No.14

7. Implementation details① Development of soluble α-nuclide removal technology (element test
assuming fuel debris retrieval work)- Issues, implementation details and goals to be addressed in this project[Issues]• The behavior of α-nuclides in solution under conditions simulating
the PCV environment during fuel debris retrieval work has not beenverified.
• The impacts of changes in solution water quality on the soluble αnuclide removal performance has not yet been evaluated.
[Implementation details]• A test equipment was manufactured to simulate theenvironment during fuel debris retrieval work.
• Preliminary tests were conducted to verify the
dissolution behavior of α-nuclides under conditions
simulating the environment during fuel debris retrieval
work.
• Adsorption tests were conducted to verify the α-nuclide
removal performance under conditions simulating theenvironment during fuel debris retrieval work, andapplicable adsorbents are selected.
[Goals]• Data acquisition on α-nuclide adsorption performance inthe environment assuming fuel debris retrieval work, andselect candidate adsorbents.
• Establishment of a water quality adjustment policy forsoluble α-nuclide removal equipment.

<Illustration of test equipment for simulating anitrogenous environment during fuel debris
retrieval work>

NitrogenSimple glove box<Adsorption test>

α-nuclide(dissolved)
α-nuclide(adsorption)

View of glove box interior Container for adsorption testing
Measure the nuclide concentration in the liquid before
and after the test and assess the adsorption amount©International Research Institute for Nuclear Decommissioning

14

No.15

7. Implementation details① Development of soluble α-nuclide removal technology (element test
assuming fuel debris retrieval work)- Response flow up to element test parameter setting

After setting up the prerequisite system, the environmental and water quality conditions in the actual equipment were
examined. The test environment and test solution conditions were also examined based on the literature researchand calculation of thermodynamic equilibrium related to the dissolution state of α-nuclides. Then the test parameters
were set according to the adsorbent to be evaluated.
Prerequisite
system setup
System configuration andcontrol policy

Nuclides targeted
for removal

Examination ofwater quality
assuming actual
equipment

Selection ofelements to beevaluated

Water quality and environmental
conditions assuming actual equipment

Selection of tracer
for usage in testing
Candidateadsorbents

Examination ofadsorbents to beevaluated

Examination ofenvironmental
conditions assuming
actual equipment

Tracer specifications(nuclide concentration,
solvent concentration,
carrier concentration,
isotope ratio)

Literature research
related todissolution state

Thermodynamic
equilibrium calculations
related to dissolution
state

Adsorbents tobe evaluatedSetting of test parameters

Examination of test
solution and test
environmental conditions
α-nuclide solubility anddissolution stateaccording to water
quality andenvironment

Test solution
conditions
Test
environmental
conditions

Items re-examinedin this project
Items that will use
settings from theprevious project

Figure Response flow up to element test planning ©International Research Institute for Nuclear Decommissioning

15

No.16

7. Implementation details① Development of soluble α-nuclide removal technology (element test
assuming fuel debris retrieval work)- Setting of prerequisite system (liquid/gas systems)• In the liquid system, in order to control water level and confine the contaminated water in the PCV, water is taken fromthe D/W (dry well) or S/C (PCV suppression chamber) and fuel debris is cooled by injecting and circulating watertaken from D/W or S/C after cooling.
• The gas system manages negative pressure to confine dust and other substances in the PCV, and fills the PCV withnitrogen for hydrogen scavenging and other purposes.
• Although air flows in due to negative pressure control, the gas phase inside the PCV becomes mostly nitrogenous.
• Up until 2020 α-nuclide adsorption tests had been conducted in an atmospheric environment. Testing in a nitrogenenvironment is necessary in order to further investigate application to actual equipment.
Extension building

Reactor building

Pretreatment
facility

Primary
Containment
Vessel

PCV stagnant
water

Reactorpressure vessel

To ①

Exhauster

Filters

Particle
removalsystemTo exhaust
port

Soluble αnuclide removal
equipment

To existingwater treatment
facility

α-nuclide removal equipment

Fuel debris

CoolerAtmosphere
NitrogenOperationalwater

Pedestal

Stagnant water
inside the building

To existing water treatment facility

Nitrogenenvironment

Liquid system
Gas system©International
Research Institutebasis)
for Nuclear Decommissioning
Figure Schematic diagram of liquid/gas systems during fuel debris retrieval(engineering

16No.17

7. Implementation details① Development of soluble α-nuclide removal technology (element test
assuming fuel debris retrieval work)- Setting of prerequisite system (soluble α-nuclide removal system)

In the previous project, to study soluble α-nuclide removal system when adsorption/removal is applied, a facility
configuration consisting of receiving tank, adsorption tower, monitoring tank, sampling equipment, etc. was
examined.
Agitator pump•

Adsorption tests were conducted underFrom
particle
atmospheric conditions using α-nuclides
removalelements (the nuclides targeted for removal). system
Transferpump
ReceivingA tower configuration with attached activatedtankcarbon on top and zirconium phosphate/titanicacid in the second stage was proposed. The
former is assumed to remove mainly colloidal
α-nuclides while the latter is assumed toremove dissolved α-nuclides.

•

In this project, similar adsorption tests under anitrogen atmosphere (simulating the gas
phase environment in the PCV) wereconducted to evaluate the impacts on thesolubility of α-nuclides and adsorption
performance, and to investigate whether the
tower configuration and adsorbent used in theprevious project need to be altered.

Attached
activatedcarbonMonitoringtank

Adsorption
tower

Adsorption
tower

•

Dischargepump
To existingwater
treatment
facility, etc.

Monitoringtank
Sampling equipment

Attached
activatedcarbonRemoval of U, Pu, etc.

Zirconium
phosphate

Zirconium
phosphate

Removal of Pu, Am,
Cm, NpTitanic
acid

Titanic
acid

Removal of Np

Figure Schematic diagram of soluble α-nuclide removal systemand proposed tower configuration**Supplementary information on the tower configuration• The number of towers in each adsorption tower is provisionally set at 2 to enable merry-go-round operation. The number of required towers for each adsorption tower needs to beevaluated in the future.
• In order to reduce the number of adsorption towers as much as possible, attached activated carbon adsorption towers (which have high removal performance for colloidal α-nuclides in
addition to soluble α-nuclides) and zirconium phosphate adsorption towers (which have high adsorption performance for multiple soluble α-nuclides) are installed. However, since theNp adsorption performance of the attached activated carbon and zirconium phosphate is only relatively high and the concentration
reduction
duringwaterforflow
treatment
may be©International
Research
Institute
NuclearDecommissioning
insufficient, a titanic acid adsorption tower should be installed at the post zirconium phosphate stage.

17

No.18

7. Implementation details① Development of soluble α-nuclide removal technology (element test assuming fueldebris retrieval work)- Examination of environmental conditions assuming actual equipment and policy for
applying these conditions to the test environmental conditions• The following table summarizes the proposed control policy and control values for the gas phase environment of the PCV,
based on the analysis status in the plant operator engineering.
• Oxygen concentration is related to the degree of air inflow, and since it is under consideration to set the control value lessthan 4%, this will be reflected in the conditions in the test evaluation. (Note, however, that the 4% figure is tentative, and not adefinite value.)• The atmospheric component in the PCV gas phase is estimated to be at a concentration less than 1/5 that of theatmosphere, which has an oxygen concentration of about 20%. (Carbon dioxide concentration in the PCV is assumed to be
70 to 80 ppm or less)
Table Proposed control policy and values for environmental items during fuel debris retrieval (gas system)
No.

Environmental item

Control requirements

Proposed control policy

Proposedcontrol value

Remarks

PCV gas phasepressure

Gas phase confinement(leakage prevention)

Negative pressure ismanaged by controlling theexhaust flow rate

TBD

If negative pressure control is not feasible, thealternatives are positive pressure control orequalization pressure control.

2

Nitrogen concentration

No(Nitrogen supply control)─

Nitrogen concentration itself is not a control target.

Hydrogenconcentration

Gas phase confinement(fire/explosion
prevention)

Nitrogen supply andexhaust flow rate arecontrolled to keep hydrogen
below the lower explosive
limitLess than 2%This is tentative, not a definite value.

Oxygen concentration

Gas phase confinement(fire/explosion
prevention)

Nitrogen supply andexhaust flow rate arecontrolled to prevent theconditions of Zr fire
outbreak.

Less than 4%

Concentration increase due to air inflow during
negative pressure control.
This is tentative, not a definite value.

Carbon dioxide
concentration

No

Course of events─Concentration increase due to air inflow during
negative pressure control.

Dust concentration

Gas phase confinement(leakage prevention)

Exhaust flow rate iscontrolled to keep theactivity concentration below
the specified level

TBD

At the time of full discharge, the concentration isless than or equal to the public exposure dose of 5
mSv.

3

4

5

6

Gas phase environment

1©International Research Institute for Nuclear Decommissioning

18

No.19

7. Implementation details① Development of soluble α-nuclide removal technology (element test assuming fueldebris retrieval work)- Examination of water quality conditions assuming actual equipment and policy for
applying these conditions to the test solution conditions (1/3)• By the start of fuel debris retrieval work, incremental changes in water quality are assumed due to circulation cooling
operations in the liquid system and PCV negative pressure control in the gas system.
• Once fuel debris retrieval work begins, it is anticipated that the pH of the liquid phase will increase due to elution of
concrete during cutting.
• The increase in pH is also assumed to increase the amount of gas phase carbon dioxide that dissolves into the liquid
phase.
Table Evaluation status of water quality of PCV stagnant water and α-nuclide adsorption performance underliquid/gas systems and fuel debris retrieval operating conditions
Operating conditions

STEP1

STEP2

STEP3

STEP4

Circulation cooling

Not conducted

Conducted

Conducted

Conducted

Not conducted

Not conducted

Conducted

Conducted

Not conducted

Not conducted

Not conducted

Conducted

Overview

RO-treated water is injectedfrom the outside the PCV andthen leaks out, so the water
quality is basically the same as
RO-treated water.

Circulation of RO-treated water
may cause elution of concrete,
etc. Since the gas phase is in anitrogen atmosphere, there is
no atmospheric impacts.

EstimatedpH

Approx. 7

7 to 95 to 95 to 12

Approx. 1 ppm

Approx. 1 ppm

Several ppm(Impacts of air inflow)

Several dozen ppm(Air inflow + impacts of pH increase)(Not applicable)(Not applicable)

Selection of candidate adsorbents
through adsorption tests up to
FY2020

Not yet evaluated

PCV negative pressure
controlFuel debris retrieval
work

Quality of
PCV
stagnant
water

Inorganiccarbon*concentration (estimated)
α-nuclide adsorption
performance evaluation
statusBy managing the PCV with
Increased elution of concrete due to fuelnegative pressure, the carbonicdebris retrieval work causes a pH increase.
acid concentration underwater
Carbonic acid concentration is positively
increases due to gas phase carbon correlated with pH. Injection of boric aciddioxide concentration increase from solution as criticality prevention is also
air inflow into the PCV.
hypothesized.

©International Research Institute for Nuclear Decommissioning

19*Collective term for CO2, HCO3- and CO32-. The abundance ratio varies with the pH of the solution.

No.207. Implementation details① Development of soluble α-nuclide removal technology (element test assuming fueldebris retrieval work)- Examination of water quality conditions assuming actual equipment and policy for
applying these conditions to the test solution conditions (2/3)• The pH range required for estimation was studied because the amount of pH and carbon dioxide elution into
the liquid phase are assumed to increase due to elution of concrete components during fuel debris retrieval.
• When calcium from concrete is eluted, the pH of a saturated solution of calcium hydroxide is approximately
12.4, but in the high pH region, carbon dioxide in the gas phase is more soluble, and over time the pH drops
to about 9 to 10 due to carbonic acid dissolution.
• When the pH decreases due to carbonic acid dissolution, calcium carbonate precipitates and α-nuclide
concentration is assumed to decrease due to coprecipitation, in which case the need for adsorption removal
is reduced.
• In order to verify the dissolved α-nuclide adsorption performance, adsorption tests under alkaline conditions(pH 8 to 10) are planned under this project, in which the α-nuclide concentration in the liquid phase is notassumed to decrease significantly.
Implementation
conditions of theprevious project

Approx.
pH 5 to 7

Implementation
conditions of
this project

Concretecomponent
elutionα-nuclide (Pu)
Pu3+
Pu(OH)4

Approx.
pH 8 to 9
α-nuclide
Pu(OH)4
Pu(CO3)2(OH)2-

Concretecomponent
elutionApprox.
pH 9 to 10
α-nuclide
Pu(CO3)2(OH)2-

Concretecomponent
elutionpH decrease
due tocarbonic
acid
dissolution

Approx.
pH 12α-nuclide
Precipitates withcarbonate
precipitation andmetal hydroxides

Figure Water quality fluctuation scenario in case of concrete component
elution Research Institute for Nuclear Decommissioning©International

20

No.21

7. Implementation details① Development of soluble α-nuclide removal technology (element test assuming fueldebris retrieval work)- Examination of water quality conditions assuming actual equipment and policy for
applying these conditions to the test solution conditions (3/3)• There are three factors that contribute to changes in PCV stagnant water quality: elution of seawater components,
elution of concrete components, and injection of boric acid solution (sodium pentaborate). The water quality
conditions assuming actual equipment with each condition as a parameter are summarized below.

• Concrete component elution-mediated pH increase is assumed to alter the form of α-nuclides. Since it is necessary tounderstand the impacts of pH on adsorption performance, implementation under two conditions is planned.
Table List of PCV stagnant water quality conditions assuming actual equipment
Type oftest water

1
2-1

2-2

3-1

3-2

4-1

4-2

5

Expected conditions

Circulation with water of similar
quality to RO-treated water
Small amount of seawatercomponents eluted into thecirculating cooling water
Large amount of seawatercomponents eluted into thecirculating cooling water
Small amount of concrete
component eluted into water
quality 2-1
Large amount of concrete
component eluted into water
quality 2-1
Small amount of concrete
component eluted into water
quality 2-2
Large amount of concrete
component eluted into water
quality 2-2
Boric acid solution injected into
water quality 2-1 to prevent
criticality

Conditions
Boric acid
Seawater Concrete
Cl ionCa ionBoronsolution Temperaturecomponent component
concentration concentration concentration
component
[°C]
[ppm]
[ppm]
[ppm]

pH

InorganicSS
carbonconcentration
concentration
[ppm]
[ppm]

No

No

No

10 to 40

0

<1

<1

6 to 7

Approx. 1

<1

Yes

No

No

10 to 40

20

<1

<1

6 to 7

Approx. 1

<1

Yes

No

No

10 to 40

100

Approx. 1<1

6 to 7

Approx. 1

<1

Yes

Yes

No

10 to 40

20

Low

<1

8 to 9

Low

<1

Yes

Yes

No

10 to 40

20

High

<1

9 to 10

High

<1

Yes

Yes

No

10 to 40

100

Low

<1

8 to 9

Low

<1

Yes

Yes

No

10 to 40

100

High

<1

9 to 10

High

<1

Yes

Yes

Yes

10 to 40

20

Low

7000

8 to 9High<1©International Research Institute for Nuclear Decommissioning

21

No.22

7. Implementation details① Development of soluble α-nuclide removal technology (element test assuming fueldebris retrieval work)- Investigation and evaluation related to dissolution state

Seawater
Seawater ＋ CO2 (3 ppmn)

Pu(1) Carbon dioxide concentration in gas phase: 3 ppm
Dissolved species abundance ratio (%)• In the previous project, thermodynamic equilibrium calculations were
performed for a system in which carbon dioxide exists in the gas
phase. It was evaluated that the solubility form of coordinatedcarbonate ions (carbonate complexes) as the dissolved form of αnuclides increases in alkaline conditions.
• The increase in carbonate ion concentration accompanying the rise in
pH may cause precipitation formation with Ca (Ca carbonate), a
concrete component, accompanied by a decrease in concentration due
to coprecipitation of α-nuclides. However, this coprecipitation
phenomenon cannot be evaluated by equilibrium calculations, so
verification through test evaluation under simulated environment
assuming actual equipment is required.

At pH 9 or
higher, the ratioof carbonate
ion-coordinatingcomplexes
increases(2) Carbon dioxide concentration in gas phase: 350 ppm

Pu concentration [mol/l]

Seawater ＋CO2 (350 ppm)

Molar concentration
per 10 Bq/m ofradioactive elements

Figure Results of evaluating Pu solubility based on thermodynamic
equilibrium calculation with condition of carbon dioxide concentrationin gas phase as a parameter (50-fold diluted seawater)

Dissolved species abundance ratio (%)

Seawater ＋CO2 (30 ppm)

At pH 7 orhigher, the ratioof carbonate
ion-coordinatingcomplexes
increases

Figure Results of evaluating dissolved form of Pu based onthermodynamic equilibrium calculation with condition of carbon dioxide©International
Institute for
Nuclear Decommissioning
concentration in gas
phase asResearcha parameter(50-fold
diluted seawater)

22

No.23

7. Implementation details① Development of soluble α-nuclide removal technology (element test assuming fueldebris retrieval work)- Overall test plan••

In the previous project, the soluble α removal test was conducted in an environment simulating the air atmosphereinside the R/B, whereas in this project the test simulated the nitrogen atmosphere environment inside the PCV.
The adsorption test was conducted after completing the preparatory and preliminary tests. The outline andobjectives of each test are as follows.
① Preparatory test: Verification of feasibility of controlling the nitrogen atmosphere in the simple glove box(verification of feasibility of maintaining an oxygen concentration of 4% or less)② Preliminary test: Verification of the impacts of α-nuclide in the liquid phase adhering to containers as well as
the impacts of precipitation formation (verification of concentration reduction factors otherthan adsorbents)③ Adsorption test: Verification of soluble α-nuclide adsorption performance by immersion testing[This project] Simulation of PCV environment (nitrogen atmosphere)

Comparison of nitrogen atmosphere/air atmosphere[Previous project] Simulation of R/B environment(air atmosphere)

Test in air atmosphere① Preparatory test
① Preliminary test

② Preliminary test
③ Adsorption test

Use of a simple glove box (under anitrogen atmosphere) to simulate thePCV gas phase environment

Figure Overall test implementation flow

② Adsorption test
©International Research Institute for Nuclear Decommissioning

23

No.24

7. Implementation details① Development of soluble α-nuclide removal technology (element test assuming fueldebris retrieval work)- Test system and preparatory test results
•

•
•
•

The diagram below (Figure A) shows the test system for evaluating simulation of the internal PCV environment (nitrogen
atmosphere).
In order to investigate radiation control, a simple glove box (simple GB) was installed in the hood of the laboratory, and anitrogen-filled line from an N2 cylinder is connected to create a nitrogen atmosphere inside the simple GB.
By verifying the impacts of air inflow into the simple GB, it was confirmed that it takes about 10 days to exceed the 4%
oxygen concentration control target, and that the atmosphere can be maintained even when human work is unavailable(e.g., nights and holidays). (Figure B)
In addition to periodically checking the oxygen concentration in the simple GB and filling it with nitrogen as appropriate, it
was also decided to adopt a policy of adding nitrogen prior to collecting liquid from test samples, since the GB
atmosphere is strongly affected each time the lid of the test sample container is opened.
Temperature control in the simple GB is difficult due to space constraints. Therefore, temperature was measured andrecorded during the test and data necessary to investigate any impacts from temperature difference was collected.
NitrogenatmosphereLaboratory
Hood
Simple glove box
Test sample
N2cylinderShakerPass box

Oxygen concentration (%)•

Oxygenconcentration [%]Oxygen concentration
control target 4%

Number of days (day)Figure A Test system simulating a nitrogen atmosphere

Figure B Results of oxygen concentration changes©International Research Institute for Nuclear Decommissioning
verified in a simple GB24

No.25

7. Implementation details① Development of soluble α-nuclide removal technology (element test assuming fueldebris retrieval work)- Preliminary test procedures

The preliminary test procedures, established based on the preparatory test results, are shown below.
The test procedures are the same as in the previous project, and the test environment was performed
under a nitrogen atmosphere.
Assuming operation in
Tracer(undiluted
solution)

Shaking for a specified
period under anitrogen atmosphere

Test container
Pure
water

N2 (nitrogen
atmosphere)
Test solution
Ionic α-nuclides

Ionic

Adhesion towall surface

Colloidal(Reference
solution)

Precipitation
formation (solid
phase)

Before shaking

Without
filtration

Tracer(dilutedsolution)

After shaking

0.1 μmFiltration

Adjust the pH toabout 1 and alsomeasure the counts
when almost all αnuclides are
dissolved

a simple GB under anitrogen atmosphere

Preliminary test
procedures

Dissolved oxygen
concentration
measurement

pH adjustment with
NaOH and
Ca(OH)2

Ionic nuclides +colloidal nuclides

Figure Diagram of preliminary test

Adjust nuclide
concentration to
10 Bq/mLpH
adjustment

Shaking

FractionationpHmeasurement

Ionic nuclides +colloidal nuclides+ precipitation(solid phase)

Testsolution

Nuclideconcentration
analysis

Approx. 2
days

Fractionation

0.1 μmfiltration
Nuclideconcentration
analysis©International
Research Institute test
for Nuclear
Decommissioning
Figure Preliminary
procedures25

No.26

7. Implementation details① Development of soluble α-nuclide removal technology (element test assuming fueldebris retrieval work)- Adsorption test procedures

Assuming operation in
The adsorption test procedures, established based on the preparatory test results, are showna simple GB under abelow.
nitrogen atmosphere
The test procedures are the same as in the previous project, and the test environment was
performed under a nitrogen atmosphere.
Dissolved oxygen
Adsorption test
Test container

N2 (nitrogen
atmosphere)
α-nuclides(dissolved)

Concentration
analysis (evaluationof amount ofadsorption)

Test solution

Sample for verifying
reduction of nuclide
concentration via
adsorbent addition

Shaking for a specified
period under anitrogen atmosphere

Tracer(dilutedsolution)

pH adjustment
with NaOH and
Ca(OH)2

concentration
measurement

Adjust nuclide
concentration to
10 Bq/mLpH
adjustment

0.1 μm filtration

Adsorbent

Adsorbent*1
Shaking

1 day*2

Shaking

7 days*2

Shaking

21 days*2

α-nuclide(dissolved)

α-nuclide(dissolved)
α-nuclide(adsorption)

Test
solutionprocedures

0.1 μm filter filtration(solid-liquid separation)

Figure Diagram of adsorption test

Fractionation

Fractionation

Fractionation

0.1 μm filtration

0.1 μm filtration

0.1 μm filtration

Nuclideconcentration
analysis

Nuclideconcentration
analysis

Nuclideconcentration
analysispHmeasurement*1: Some
adsorbents are prewashed to reduce
pH fluctuations fromaddition ofadsorbents.
*2: Equilibrium
attainment isensured by setting a
3-pattern shaking
period.

Post-immersion concentration©International
Research Institute for Nuclear Decommissioning
Figure Adsorption
test procedures

26

No.27

7. Implementation details① Development of soluble α-nuclide removal technology (element test assuming fueldebris retrieval work)- Preliminary test parameters

Considering the PCV gas phase environmental conditions (nitrogen environment) and the alkaline conditions
during concrete cutting, the following test conditions were set.
Table Test parameters for preliminary test
No.

Test parameters

Setting requirements

Remarks

1

NuclideAm, Cm, Pu, Np, U

Based on the results of exposure assessment at the site boundary, it is highly necessary
to remove these particular α-nuclides

2

Nuclide concentration

10 Bq/mL3

Gas phase environment

Nitrogen atmosphere

4

Material of test
container

PP, PFA

5

Water quality conditions

See the table below

Particular concentrations that can be handled by the test facility and that allow foranalysis of removal rates
Required conditions given the anticipated gas phase environment inside the PCV during
fuel debris retrieval
Based on preparatory test results, the atmosphere is controlled using a simple GB with an
oxygen concentration of 4% as the reference value
Adsorption tests are conducted under two conditions, containers with a low container
adhesion impacts are used
PP: Polypropylene, PFA: Fluoropolymer.

Table Water quality conditions for preliminary test
No.

Water quality conditions

pH

Remarks

1

1000-fold diluted seawater

6 to 72200-fold diluted seawater

6 to 7

The policy is to use the results of condition implementation from the previous project, and toutilize as-yet unimplemented conditions.

3

1000-fold diluted seawater＋Ca(OH)2

Approx. 9 - Assumed to be operated near the analysis value of the chloride ion concentration in thestagnant water up to the present (19 ppm) due to the low elution of seawater components
Approx. 10 - Assuming a rise in pH due to elution of concrete components

200-fold diluted seawater＋Ca(OH)2

Approx. 9 - Assumed to be operated near the upper control limit (100 ppm) due to high elution ofseawater components
Approx. 10 - Assuming a rise in pH due to elution of concrete components

4

5
6
7

Sodium pentaborate water(7000 ppm as B)

8 to 9Conditions for evaluating the impacts of sodium pentaborate injection©International Research Institute for Nuclear Decommissioning

27

No.28

7. Implementation details① Development of soluble α-nuclide removal technology (element test assuming fueldebris retrieval work)- Adsorption test parameters

The proposed adsorption test conditions are shown in the table below. Based on the results of the preliminary tests
to be conducted in the future, water quality conditions, adsorbent conditions, and shaking period conditions to be
implemented will be selected.
Table Test parameters for adsorption test
No.

Test parameters

Setting requirements

Remarks

1

NuclideAm, Cm, Pu, Np, U

Based on the results of exposure assessment at the site boundary, it is highly necessary to
remove these particular α-nuclides

2

Nuclideconcentration

10 Bq/mL

Particular concentrations that can be handled by the test facility and that allow for analysis ofremoval rates3

Gas phaseenvironment

Nitrogen atmosphere4

Shaking period

1, 7, and 21 days

5

Material of test
container

PP or PFA

6

Adsorbent

see remarks

7

Water quality
conditions

See the table below

Required conditions given the anticipated gas phase environment inside the PCV during fueldebris retrieval
Based on preparatory test results, the atmosphere is controlled using a simple GB with an
oxygen concentration of 4% as the reference value
Evaluation of near-equilibrium state by shaking for 7 and 21 days. Judgment whether theadsorption rate is remarkably low by shaking for a short time.
Based on the preliminary test results, the type of container to be used is selected.
PP: Polypropylene, PFA: Fluoropolymer.
Adsorbents (attached activated carbon, activated carbon, titanic acid, silicic titanic acid, zirconium
phosphate) that have obtained high adsorption performance in the evaluation under atmosphericenvironmental conditions up to the previous project are targeted.

Table Water quality conditions for adsorption test*
No.

Water quality conditions

1
2
3
4

5

1000-fold diluted seawater＋Ca(OH)2

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

pH

Remarks

Approx. 9 - Assumed to be operated near the analysis value of the chloride ion
concentration in the stagnant water up to the present (19 ppm) due to theApprox. low elution of seawater components
10- Assuming a rise in pH due to elution of concrete components*Since all conditions cannot be
implemented due to restrictions inApprox. 9 - Assumed to be operated near the upper control limit (100 ppm) due tothe test process, the plan is tonarrow down the water quality
Approx. high elution of seawater components- Assuming a rise in pH due to elution of concrete components
10
conditions to be implemented
based on the preliminary test
8 to 9Conditions for evaluating the impacts of sodium pentaborate injection©International Researchresults.
Institute for Nuclear Decommissioning28

No.29

7. Implementation details① Development of soluble α-nuclide removal technology (element test assuming fueldebris retrieval work)- Preliminary test progress• Using Cm, 1000-fold diluted seawater and 200-fold diluted seawater were adjusted to pH 9 and 10 with calcium
hydroxide, and shaken for 2 days in two types of containers (PP, PFA).
• Since the counts are reduced by 10 to 30% without filtration, it is estimated that part of Cm adheres to thecontainer.
• In addition, it is estimated that part of the Cm changed into particles from the decrease in the count after 0.1 μm
filtration.
• Adhesion to containers and formation of particles were verified, but since Cm exists in a dissolved state to someextent, it was verified that an adsorption test was required. Since there is no big difference in the results between
PP and PFA, PP containers will be used in the adsorption test.
Without filtration
0.1 μm filtration
700
600

Cm additive concentration (approx. 10 Bq/mL)Counts per minute [CPM]

500
400
300
200
100
0
1000-fold diluted
1000-fold diluted
1000-fold diluted
seawaterseawaterseawater
Ca(OH)2 pH=9 PP Ca(OH)2 pH=9 PFA Ca(OH)2 pH=10 PP

1000-fold diluted
seawater
Ca(OH)2 pH=10
PFA

200-fold diluted
200-fold diluted
200-fold diluted
seawaterseawaterseawater
Ca(OH)2 pH=9 PP Ca(OH)2 pH=9 PFA Ca(OH)2 pH=10 PP

Figure Preliminary test results using Cm

200-fold diluted
seawater
Ca(OH)2 pH=10
PFA©International Research Institute for Nuclear Decommissioning

29

No.307. Implementation details① Development of soluble α-nuclide removal technology (element test assuming fueldebris retrieval work)- Preliminary test progress (2)• Based on 1000-fold diluted seawater and 200-fold diluted seawater, Am was used to prepare solutions adjusted
to pH 9 and 10 with calcium hydroxide and B=7000 ppm with sodium pentaborate, and shaken for 2 days.
• The counts were lower than that of the reference solution under the unfiltered conditions of 1000-fold diluted
seawater pH 9,10 and 200-fold diluted seawater pH 10, and it is estimated that part of Am adheres to thecontainer. In addition, it is estimated that part of Am changed into particles from the decrease in the count after
0.1 μm filtration.
• Adhesion to containers and formation of particles were verified for some liquids, but since Am exists in a
dissolved state to some extent, it was verified that an adsorption test was required.
Without filtration

0.1 μm filtration

700

600

Am additive concentration (approx. 10 Bq/mL)Counts per minute [CPM]

500
400
300
200
100
0
1000-fold diluted 1000-fold diluted 1000-fold diluted 1000-fold diluted 1000-fold diluted 200-fold diluted 200-fold diluted
seawaterseawaterseawaterseawaterseawaterseawaterseawater
pH=1 (reference
pH=6 to 7
pH=9
pH=10
B=7000 ppm
pH=1 (reference
pH=6 to 7
solution)solution)

Figure Preliminary test results using Am

200-fold diluted
seawater
pH=9200-fold diluted
seawater
pH=10200-fold diluted
seawater
B=7000 ppm©International Research Institute for Nuclear Decommissioning30

7. Implementation details

No.31① Development of soluble α-nuclide removal technology (element test
assuming fuel debris retrieval work)- Summary[Results for this year]✓ Based on the control policy for the gas phase environment inside the PCV during fueldebris retrieval, a policy was determined to control the oxygen concentration at 4% orless as a simulation of the test environment, and it was verified that is possible to control
the atmosphere inside the simple glove box during the preparatory test.
✓ Based on the gas-phase atmosphere control policy verified in the preparatory test, thetest procedures for the preliminary test and the adsorption test were devised in the simple
glove box with nitrogen atmosphere.
✓ Some preliminary tests have been started, and Am and Cm exist in a dissolved state tosome extent in a nitrogen atmosphere and alkaline conditions (pH 9, 10), so it wasverified that there is a high need for nuclide removal and an adsorption test.
[Future plans]✓ Preliminary tests for Pu, U, and Np will be implemented using the proposed test
procedures to verify the behavior of α-nuclides under each water quality condition.
✓ Based on the results of the preparatory and preliminary tests, conditions for adsorption
tests will be examined.
✓ An adsorption test will be conducted to evaluate the adsorption performance of eachcandidate adsorbent for α-nuclides.
✓ Based on the results of the adsorption test, the conceptual system design will be
reviewed and a test plan using actual contaminated water will
be devised.
©International
Research Institute for Nuclear Decommissioning31

No.32

7. Implementation details② Development of RO-concentrated water treatment technology➢ Flowchart of examination[FY2019/2020 (previous project)](Plant operator engineering)Technological development for other items (water treatment)
Development ofsoluble nuclide
removal technology

Preliminary engineering

Arrangement of Liquid
system pattern- α-nuclide adsorbent- Necessity of RO filter- Necessity of RO-concentrated water
treatment technology

Selection of candidate
technologies for ROconcentrated water
treatment[FY2021/2022 (this project)]- Select flocculation sedimentation
treatment

Verification test using
simulated liquid
Preliminary
tests
Adsorbent test
- Selection of powder
adsorbent

Items to bereported at theend of FY2021

Flocculant test
- Selection of flocculant

Flocculation
sedimentation test

- Beaker scale test
- Equipment test

Development of technology
for flocculation sedimentation
treatment of filter liquid waste- Flocculant

Technological development for otheritems (canister project)
Investigation of existing knowledge
and organization of issues in storageof sludge or slurry form in canisters

Issue
identification- Increased amount of sludge
generated

Development ofsludge stabilization
technology

Verification test using
actual liquid- Appropriate treatment method- Volume reduction by
dehydration, etc.

Adsorbent test

Technological
study

Flocculant test

- Dehydration technology
selection result

RO-concentrated
water treatment test

Element test
- Treatment method

Proposal of ROconcentrated water
treatment method - Equipment

specifications

Conceptual
designDetailed design©International Research Institute for Nuclear Decommissioning32

No.33

7. Implementation details② Development of RO-concentrated water treatment technology[Issues]• As a treatment method for RO-concentrated water generated from RO filters, flocculation
sedimentation is being examined, and it is necessary to establish treatment procedures.
• Powder adsorbents and flocculants that can be used for RO- concentrated water treatment have notbeen selected, and data on their nuclide adsorption performance and flocculation sedimentation
performance are insufficient.
Simulated ROconcentrated water(pure water +simulated nuclide)[Implementation details]• Candidates for powder adsorbents and flocculants wereselected based on literature research and past resultsof use in the existing water treatment facility.
• The nuclide removal performance of the powder
adsorbent was evaluated by adsorption test.
• A beaker scale flocculation sedimentation test evaluates
the flocculant performance of a flocculant on a powder
adsorbent.
[Goals]• Evaluation on the nuclide removal performance of thepowder adsorbent and the flocculation sedimentation
performance of the flocculant.
• Selection of applicable powder adsorbent and flocculant
candidates.
• Development of treatment procedures for ROconcentrated water.

Powder adsorbent
Flocculant
pH adjuster

Stirring

SludgeBeakerFigure RO-concentrated water treatment
element test image

Before flocculation
sedimentation treatment

After flocculation
sedimentation treatment

Figure RO-concentratedwater
element test
©International ResearchInstitutetreatment
for Nuclear Decommissioning

33

No.34

7. Implementation details② Development of RO-concentrated water treatment technology➢ System configuration of sludge collection system

Solid components were collected from liquid waste generated from non-soluble nuclide removal equipment
and RO filter equipment.
Soluble nuclide removal equipment
Non-soluble nuclide removal equipment
Insidethe PCV
Stagnant
water

Cutting particles
such as fuel debris(large particles)

Roughing
systemIntermediate
systemDrain/backwash water

Backwash water

Extraction

To soluble
nuclide removal
equipment

Flocculant

Adsorption towerequipment

Final treatment
systemExtraction

Natural sedimentationorFlocculation sedimentation

To soluble
nuclide removal
equipment

Flocculant

SupernatantCutting particles(small particles)
such as fueldebrisSludgePowder adsorbent
Extraction To water
Flocculant
treatment facility

Supernatant

Powder adsorbent
for nuclide
adsorption

Flocculation sedimentation
Removal

To watertreatment facility

ROconcentrated
water

Backwash water

Supernatant

SludgeRO filterequipment

RemovalSludgeFlocculation sedimentation

To wastestorageRemovalTo
canisterstorageRoughing liquid waste treatment equipment

Intermediate/final treatment liquid
waste treatment equipment

RO-concentrated water
treatment facility©International Research Institute for Nuclear Decommissioning
Figure System configuration concept of sludge collection system

34

No.35

7. Implementation details② Development of RO-concentrated water treatment technology➢ Sludge collection system requirements

The liquid to be treated and treatment requirements for each sedimentation equipment are shown below.
Table System requirements for each sedimentation equipment
No.

Items

Roughing liquid waste
treatment equipment

1

Water to be treatedDrain water and backwash water
generated from roughing
components

2

Contained particle
component(for flocculation
sedimentation)

Molten fuel, core structures,
concrete components, other solids

3

4

5

6

(Particle size 100 to several tens of μm)*Trial calculation based on the water quality
conditions set in the subsidy projectIntermediate/final treatment liquid
waste treatment equipment
Backwash water generated fromintermediate components and final treatment
components(It is assumed that the intermediate system will
be in bypass operation during fuel debris retrieval
work)

Molten fuel, core structures, concrete
components, other solids(Particle size several tens to 0.1 μm)

RO-concentrated water
treatment facility
RO-concentrated water generated
from RO filter equipment

Powder adsorbent for nuclide
adsorption

Particle
concentration

Thousands to 10000 ppm

Thousands to 10000 ppm

Hundreds to 10000 ppm

From filter element test results

From filter element test results

From adsorbent element test results

Treatment flow rate

Intermittent drain: 100 m3/year
Regular drain: 1300 m3/year

0.1 μm UF filter: 10 m3/year
0.05 μm UF filter: 100 m3/year

From filter element test results

From filter element test results

Fuel debris retrieval waterCalculated assuming an RO filter
concentration ratio of 2 to 3 times at (22m3/day)Intermittent drain: 1.0 m3/day or
more
Regular drain: 8.2 m3/day or more

0.1 μm UF filter: 0.063 m3/day or more
0.05 μm UF filter: 0.63 m3/day or more

Treatment flow raterequirements

Operating method

Flow rate to enable treatment at 80%
equipment availability

Flow rate to enable treatment at 80% equipment
availability

8 to 11 m3/day

14 m3/day or more
Flow rate to enable treatment at 80%
equipment availability

Batch treatment
Assuming completion of 1 batch treatment in 1 day©International Research Institute for Nuclear Decommissioning35

No.36

7. Implementation details② Development of RO-concentrated water treatment technology[Issue] A large amount of flocculated sediment (hereinafter referred to as sedimentation sludge)
are generated from a sedimentation tank.
Assuming that the sludge collection container is a unit can size (Φ200
x H400 mm), the required number of times of discharging is
calculated as shown in the table below

Sedimentation sludge properties- The amount generated is about 10 to 15
vol% of the initial liquid volume- The moisture content is about 90%- High fluidity

Table Number of discharge of sludge collection container
Liquid waste to
be treated

Number of collection
containers [pieces/day]Roughing liquid
waste

1

Final treatment
liquid waste

8

RO-concentrated
water

111

➢

➢

RO-concentrated water Filter liquid waste flocculation
treatment test
sedimentation test
*Magnetite 1000 ppm*SiO2 10000 ppm
Figure Appearance of liquid waste after flocculation sedimentation treatment

The amount of sedimentation sludge generated from the sedimentation tank for filter liquid waste
treatment is relatively small, and it is estimated that the number of times it is discharged is on the orderthat can be applied to the actual equipment
The amount of sedimentation sludge generated from the sedimentation tank for RO-concentrated
water treatment is enormous and needs to be discharged an enormous number of times, so
improvement is necessary

Proposed plan
①
②
③
④
⑤

Selection of adsorbent with higher removal performance
Examination of optimizing the amount of powder adsorbent to be
added
Examination of optimizing the amount of flocculant to be added
Review of sedimentation treatment procedures and equipment
Examination of volume reduction treatment of sedimentation sludge
by dehydration① to ④ ⇒ Examine via element tests⑤ ⇒ Perform element tests after literature research andtheoretical study*Examined in Item ③©International Research Institute for Nuclear Decommissioning36

No.37

7. Implementation details② Development of RO-concentrated water treatment technology➢ Adsorbent test
[Purpose] Selection of adsorbents applicable to RO-concentrated water treatment[Implementation details] The adsorption performance of the powder adsorbent wasevaluated by an agitated batch test.
Examination of test system
When evaluating the adsorption performance of adsorbents, it iscommon to conduct an immersion batch test, but since ROconcentrated water is agitated and treated in a sedimentation tank,a test system that simulates the actual equipment was adopted.

Figure Adsorbent test (titanic acid)
Simulated liquid
waste(1000-fold diluted
seawater, tracer)

Add powder
adsorbent

Sampling &Analysis

Selection of powder adsorbent
The adsorbents used in this test were selected as shown in thetable below, referring to the results of literature research, the resultsof use in the Fukushima Daiichi NPS, and the adsorption rate dataobtained in tests up to fiscal 2020. Adsorbents reported to have highremoval performance against α-nuclides (Pu, Am, etc.) wereselected.
Table Selection results of powder adsorbent*
No.

Adsorbent typeParticlesize

1Titanic acid

4.08 μm

2

Titanium silicate

8.55 μm

3

Activated carbon Several μm There are the use results in the Fukushima Daiichi NPSIron oxide-based adsorbent. There is a report of high Kd
Hematite
1.32 μm(partition coefficient) for α-nuclides such as Pu.
Magnetite
1.96 μm An iron oxide-based adsorbent similar to hematite.

4
5

Remarks
There is the use results at the Savannah River Site in the
United States (batch treatment)
There are the use results in the Fukushima Daiichi NPS

Figure Stirring type adsorption test Image*Since HOT test is difficult with this test method,
all tests are performed as COLD test.
For α-nuclides (Pu, etc.) that are difficult tosimulate with a COLD tracer, item ① soluble
adsorption rate data is referred.
*Selected from α-nuclide adsorbents
Selected from adsorbents that cannot be
formed into granules and are difficult to fill inadsorption towers*Particle size of activated carbon not yetmeasured(to beforimplemented
in the future)©International
Research Institute
Nuclear Decommissioning37

No.38

7. Implementation details② Development of RO-concentrated water treatment technology

Eu removal performance of
Kd=1E6 to 7 was verified bytitanic acid and titanium silicate.

➢ Adsorbent test results

Verified decrease in tracer
concentration even after 30
minutes under the condition thatthe amount of adsorbent added
was small
However, in almost all test
cases, verified that adsorption
equilibrium was reached within
30 minutes

Figure Adsorbent test results (titanic acid 20 ppm) Change
in partition coefficient over time

Titanic acidTitanium silicate
Activated carbon
Hematite
Magnetite

Partition coefficient Kd [mL/g]

Partition coefficient Kd [mL/g]

Test conditions
Test solution: 1000-fold diluted seawater, pH: 7
Tracer: Cs (0.1 ppm), Sr (0.1 ppm), Eu (0.5 ppm)
Adsorbents: right figure (particle size: several tens
to several μm)
Concentration of adsorbent added: 20 to 500 ppm
Stirring conditions: 150 rpm×30, 60, 120 min

Figure Adsorbent test results: Partition coefficient

Titanic acid has a Kd of about 1E+7 and has the highest Eu removal performance.
Titanium silicate has a Kd of 1E+6 to 7 and has the second highest Eu removal performance after titanic acid.
Activated carbon has a Kd of about 1E+5 to 6, and hematite and magnetite both have a relatively low Kd of about 1E+3 to 4.
In addition, the Kd of titanic acid with respect to Sr and the titanium silicate with respect to Cs and Sr are comparable(compared to the Kd of Eu), and the Kd with respect to Cs and Sr of activated carbon, hematite and magnetite are relatively
low.
The adsorption rate is fast and it is assumed that the adsorption equilibrium will be reached within 30 minutes.

©International Research Institute for Nuclear Decommissioning38

No.39

7. Implementation details② Development of RO-concentrated water treatment technology➢ Adsorbent test results
The additive amount required to satisfy the removal requirement (DF100) was evaluated.
Evaluation of required additive amount of powder adsorbent
No.

Adsorbent

1

Titanic acid2
3

Titaniumsilicate
ActivatedcarbonPartition coefficient [mL/g] *1
Cs

Sr

Eu (Am)Additiveamount*2

1.3.E+04

1.9.E+06

1.4.E+07<50 ppm

2.6.E+06

2.8.E+05

3.2.E+06

50ppm

2.5.E+03

5.8.E+02

2.2.E+05

200 ppm<

4

Hematite

3.5.E+03

9.1.E+02

5.2.E+03

1000 ppm<

5

Magnetite

1.1.E+03

6.0.E+02

5.9.E+03

1000 ppm<

Treatmenttime *3<30 min*1 Results with relatively high removal
performance are shown in red. No.3 is arelatively high result, but is inferior to No.1 and2, so it is shown in light red.
*2 In this test, the tracer concentration was set tobe higher than the actual water quality, so it ispossible that the amount of adsorbent added
can be reduced for liquid waste with lower
concentration. It is necessary to evaluate theadsorption isotherm separately and evaluate
the partition coefficient in each concentration
region.
*3 If the solution can be uniformly agitated, thetreatment time is assumed to be short(approximately 30 min) because the adsorption
rate in the initial stage is fast.

Titanic acid (No. 1) and titanium silicate (No. 2) have high removal performance against Eu, and the required additive
amount was evaluated as several tens of ppm.
Since it is possible to reduce it from the 1000 ppm that was assumed in the past, it is thought that the amount of sludge
generated can be reduced accordingly.
It was verified that the iron oxide-based (No. 4, 5) have a relatively low Eu removal rate.
→ Compared to titanic acid, etc., the amount to be added must be increased (several thousand ppm or more).
Since the removal rate for Cs and Sr is low, it is possible to selectively remove only α-nuclides

When selectively removing only α-nuclides → Select activated carbon (magnetite)
When removing total activity including α-nuclides → Select titanic acid and titanium silicate©International Research Institute for Nuclear Decommissioning39

7. Implementation details

No.40② Development of RO-concentrated water treatment technology➢ Adsorbent selection policy
In addition to the removal performance of α-nuclides, the removal performance of Cs
and Sr is considered to select the adsorbent.
Liquid system configuration plan 1

The liquid system shall be a system capable of selectively removing only α-nuclides and notremoving Cs and Sr as much as possible
•

•
•
•

The α-activity concentration shall be reduced to a level that can be discharged to theexisting water treatment facility
β and γ nuclides such as Cs and Sr are removed by the existing water treatment facility
If only α-nuclides are sludged and captured, the dose of sludge can be kept low
The shielding function required for sedimentation equipment and sedimentation sludge
treatment equipment can be kept low, contributing to the reduction of worker exposure

Candidate powder adsorbents: activated carbon, (titanic acid, magnetite)
Adsorbents with high removal rate of α-nuclide (Eu) and low removal rate of Cs and Sr are selected
Considering the Eu removal rate, there is a concern that the additive amount must be increased (several
hundred ppm)
→The amount of sludge generated increases
Even when retrieving fuel debris, if the Cs and Sr concentrations are at a level that can be treated with the existing water
treatment facility, the only activity that needs to be reduced in the liquid system is α-nuclides. If a powder adsorbent
capable of adsorbing and removing Cs and Sr is used for RO-concentrated water treatment, a strong shielding function
will be required for the equipment itself, leading to an increase in the scale of the equipment. For the above reasons, it isassumed that it is necessary to select a powder adsorbent capable of selectively removing only α-nuclides. As thepowder adsorbent for selective removal of α-nuclides, select a powder adsorbent that has high Eu removal performance©International Research Institute for Nuclear Decommissioning
and low adsorption performance for Cs and Sr.

40

No.41

7. Implementation details② Development of RO-concentrated water treatment technology➢ Adsorbent selection policy
In addition to the removal performance of α-nuclides, the removal
performance of Cs and Sr are considered to select the adsorbent
Liquid system configuration plan 2
The liquid system shall be a system capable of reducing the concentration of total
activity, including α-activity•••

The total activity concentration including α-activity concentration shall be reduced to a level thatcan be discharged to the existing water treatment facility
Since Cs and Sr, which are dominant as activity, are also sludged and captured, the dose of thesedimentation sludge will be highFrom the viewpoint of worker exposure prevention, the level of shielding function required for
sedimentation equipment and sedimentation sludge treatment equipment will increase

Candidate powder adsorbents: Titanic acid *, titanium silicate*Titanic acid has a low Cs removal rate

Adsorbents with high removal rates for α-nuclides (Eu), Cs, and Sr are selected
Due to the high Eu removal rate, the additive amount is evaluated to be relatively small (<several tens of ppm)
If the total activity concentration is set as a system requirement for being discharged to the existing water
treatment facility, it will be necessary to reduce the concentrations of Cs and Sr, which are assumed to be
predominantly radioactive, in addition to α-nuclides. Therefore, it is assumed that it is necessary to selecta powder adsorbent that can reduce the total activity. As a powder adsorbent for total activity removal,
select a powder adsorbent with high removal performance for all of Cs, Sr, and Eu.
©International Research Institute for Nuclear Decommissioning41

No.42

7. Implementation details② Development of RO-concentrated water treatment technology➢ Flocculant test
[Purpose] Selection of flocculants applicable to RO-concentrated water treatment[Implementation details] Flocculation sedimentation treatment for the simulated liquid towhich the powder adsorbent is added was performed.
Simulated liquid(1000-fold diluted
seawater)Powderadsorbent

Flocculant (including
pH adjustment*)

Stirring blade

BeakerPreparation ofsimulated liquid waste

Sampling & Analysis

Sludgecomponents

Addition offlocculant

Stirring

StaticFigure Flocculant test Image*The tracer removal rate is not evaluated because the main item to beverified is whether the powder adsorbent flocculates and settles.
Therefore, no tracer is added.

Figure Flocculant test
Appearance of equipment*All of the flocculants selected this time are acidic reagents, and it is necessary toadjust to the neutral region after addition, so sodium hydroxide solution is used.

Examination of test system
A stirring jar test similar to the adsorbent test is adopted.
It simulates a scaled-down system of treatment in an actual flocculation
sedimentation tank.

Selection of flocculant
Flocculant was selected as shown in the table on the right based on the resultsof the literature research and the results of FY2020 test.
In addition to the flocculation sedimentation performance, redundancy withrespect to the ionic strength of liquid waste was also considered.

Table Results of flocculant selectionNo.
1
2

Flocculant

Remarks

There are reports of high treatment
High base PAC performance for liquid waste with
low ionic strength
Preliminary project test verifies
Sulfuric acidhigh treatment performance for
Aluminum©International
Research
Institutefor Nuclear
Decommissioning
liquidwaste
containing
boric acid42

No.43

7. Implementation details② Development of RO-concentrated water treatment technology➢ Flocculant test results
The amount of sedimentation sludge generated after the flocculation sedimentation treatment wasless dependent on the type of flocculant, and largely dependent on the type of powder adsorbent.

Amount of sedimentation sludge generated
Conventionally, when the amount of powder adsorbent added was 1000 ppm, the amount of sludge
generated was assumed to be about 10 vol%Titanic acid, activated carbon, magnetite → All results were found in 0.2 to 0.3 vol%.
Titanium silicate, hematite →The results of order of several vol% were frequently found(varies depending
on test conditions).
Table Flocculant test conditions (adsorbent/flocculant)

Types

Concentration[ppm]

Titanic acid50

Titanium silicate

50

Flocculant*2
Types

High
base
PAC

Concentration
[ppm]

Types

Concentration
[ppm]

100

20

1000

200

Activatedcarbon200

Hematite

500

1000

1000

Magnetite

500

1000

50

1000

Aluminum
sulfate*1 Conditions set based on adsorbent test results*2 Conditions set based on the preparatory test

50

High base
PAC
Aluminum
sulfateSedimentation sludge generation
rate [vol%]Powder adsorbent*1

Titanic acidTitanium silicate

ActivatedcarbonHematite

Magnetite

Figure Flocculant test results Sedimentation sludge volume

There was no significant difference between high-base PAC and aluminum sulfate in terms of the amount©International Research Institute for Nuclear Decommissioningof sedimentation sludge generated.
43

No.44

7. Implementation details② Development of RO-concentrated water treatment technology➢ Flocculant test results•••

A large percentage of particles settle within 10 minutes after flocculation sedimentation treatment
It was evaluated that it is possible to reduce the SS concentration to about 10 ppm by adding a flocculant.
It was difficult to reduce the SS concentration to 1 ppm or less without depending on the amount of powder
adsorbent added.
*Because the turbidity measurement results were converted to SS concentration, there is a possibility that theregion of 10 ppm or less could not be evaluated appropriately•

It was verified that the less the SS components, the narrower the adjustment range of the flocculant
concentration.
→ It will be difficult to adjust the amount of flocculant added in actual equipment operation.
Aluminum sulfate/Titanate

Powder adsorbent*

1

Types

Concentration[ppm]

Titanic acid50

Titanium silicate

50

Activated carbon

200

Hematite

500

Magnetite

500

Flocculant
Types

Concentration
[ppm]

*2

Concentration
[ppm]

Types

100

20

1000
High
base
PAC

1000

200
Aluminum
sulfate

50

1000

1000

1000

50

*1 Conditions set based on adsorbent test results*2 Conditions set based on the preparatory test

Supernatant SS concentration [ppm]Table Flocculant test conditions (adsorbent/flocculant)

The figure on the rightshows the transition of SS
concentration

Aluminum sulfate/Titanium silicate
High base PAC/Titanic acid
High base PAC/Titanium silicate

It is possible to reduce theSS concentration of thesupernatant to about 10 ppm
by adding a flocculant

Static time [min]Figure Flocculant test results Transition of SS
concentration (titanic acid, titanium silicate)
No difference was observed in sedimentation behavior of particles between high base PAC and aluminum sulfate. Aluminum sulfate
was selected as the flocculant because it was verified to have high flocculation and sedimentation properties in the previous project,©International Research Institute for Nuclear Decommissioning
even when water quality fluctuated due to boric acid solution conditions.

44

No.45

7. Implementation details② Development of RO-concentrated water treatment technology➢ Examination on the discharge of supernatant to the existing water treatment facility
Acceptance requirements for the existing water treatment facility were
examined mainly based on the past results of accepting stagnant
water in the process main building (PMB).

*Primarily set based on the past results of accepting
PMB stagnant water in FY2018.
Acceptance requirements for the existing water
treatment facility during the fuel debris retrieval
work is an issue for future study*

Table Tentative set values for acceptance requirements for the existing water treatment facility
Items

Set value

Test results

Remarks

Cl<700 ppm<17 ppm

Ca<50 ppm-

Mg<30 ppm-

It is assumed that if the concentration of each ion in the circulating cooling water is below theassumed water quality conditions, the concentration in the supernatant water will not exceed thisitem.
However, regarding the Cl concentration, it is necessary to examine the possibility of treatinghigh Cl concentration stagnant water such as torus room water.

SO4

-

pH

7 to 8.5SS
concentration<20 ppm

Total αconcentration<7.4E+0 Bq/L

Scheduled to acquire
When aluminum sulfate is used, most of the sulfate ions remain in the supernatant.
data
In the process of flocculation sedimentation treatment, the pH of the supernatant is adjusted toabout pH 7, so it is assumed that this item will be satisfied.
Since it is assumed that the SS concentration can be reduced to about 10 ppm or less by
flocculation sedimentation treatment, the SS concentration reference is assumed to satisfy theacceptance requirements.
<10 ppm
However, since the particles remaining in the supernatant have a high amount of activity, it isnecessary to separately evaluate whether they meet the requirements for acceptance based on
the activity concentration.
Set from the maximum α concentration of the inlet water of the existing water treatment facility.
99% removal rate
Issues for future study. It is assumed that it can be evaluated from the total α concentration of theagainst initial
input water and the removal performance and additive amount of the powder adsorbent.
concentration
Evaluation is also necessary from the viewpoint of public exposure dose (DF100) and(after 0.45 μm filtration)
announcement density.
6.7 to 7.2

It is estimated that the supernatant water after the flocculation sedimentation treatment can be discharged to theexisting water treatment facility, considering the assumed treatment results of the existing water treatment facility.
However, removal of residual particles in the supernatant water, such as by filtration,
may be Researchnecessary.
©International
Institute for Nuclear Decommissioning45

No.46

7. Implementation details② Development of RO-concentrated water treatment technology➢

Examination of solid-liquid separation of powder adsorbent when the amount added is small
It was suggested that the amount of powder adsorbent added could be reduced from the conventional set value(1000 ppm). Since flocculants are not applicable to the treatment of liquid waste with a small amount of SS, a systemthat concentrates the SS components as a pretreatment was examined.

Concentration of powder adsorbent added

Solid-liquid separation technology

Thousands to hundreds of ppm → Solid-liquid separation by flocculant is effective
Hundreds to several tens of ppm → Solid-liquid separation by flocculant (concentration treatment if necessary)
Several tens of ppm or less→ Concentration treatment + flocculant*It is assumed that cross-flow filtration and dead-end
filtration are effective for concentration treatment
Lower limit that can be reduced by flocculant
Solid-liquid separation
technology Region requiring concentration by filtration● Cross-flow concentration +concentrated water treatment● Dead-end filtration (total filtration
method) + backwash water treatment

Powder adsorbent

Small

Titanic acidTitanium silicate

Region where flocculant is effective● Sedimentation by flocculant

Conventional set value

ActivatedcarbonHematite
Magnetite

Large
1000
1
10
100
Concentration of powder adsorbent added (supernatant SS concentration) [ppm]*Since the adsorbent test was
implemented at a higher tracer
concentration than the assumed
water quality, the amount of powder
adsorbent added may be less than
the set value

By concentrating the SS components through concentration treatment, solid-liquid separation using a flocculant is©International Research Institute for Nuclear Decommissioning
possible while reducing the amount of liquid waste
46No.47

7. Implementation details② Development of RO-concentrated water treatment technology➢

Examination of solid-liquid separation of powder adsorbent when the amount added is small

Application of concentration treatment to RO-concentrated water treatment
RO-concentrated water

Powder adsorbent
Flocculant
Permeate (SS concentration: very low)
pH adjustment Supernatant (SSSuspension (SSconcentration: low)
reagent
Cross-flow
concentration: low)
Concentrated liquid
Sedimentation sludge
Decrease in...
(SS concentration: high)liquid amountConcentrated liquid(SS concentration:
medium)Receiving ROconcentrated water

Adsorption
treatment

Cross-flow
concentration

Receiving
concentrated liquid(More than once)

Flocculation
sedimentation treatment(including pH
adjustment)

StaticFigure Proposed procedures for RO-concentrated water treatment method when cross-flow concentration is applied
If the concentration of powder adsorbent added is low (dozens of ppm or less), cross-flow filtration is performed after adsorption
treatment to separate moisture to reduce the liquid amount and concentrate SS components (approx. 1000 ppm or more).
After that, the concentrated liquid is subjected to flocculation sedimentation treatment to separate into supernatant and sedimentation
sludge.
In addition, since the liquid amount is reduced by the concentration treatment, it can be assumed that the concentrated liquid for
multiple times can be subjected to the flocculation sedimentation treatment at once. Therefore, the concentrated liquid is discharged
into a tank for flocculation sedimentation treatment, and the treatment is performed after securing a certain amount of liquid.

Mass balance (water volume, sludge volume) in the system using each concentration treatment©International Research Institute for Nuclear Decommissioning
will be evaluated in the future.
47

No.48

7. Implementation details② Development of RO-concentrated water treatment technology
RO-concentrated water Proposed flocculation
sedimentation treatment process
RO-concentrated
water(dissolved radionuclides including α-nuclides)

Start of stirring

Stirring speed: 150 rpm
Add powder
adsorbent

Stirring (adsorption
treatment)

Concentration
treatment
Changes toacidity of about
pH 3 to 4
Properly
measure pH①
Powder adsorbent: Titanium silicate
Titanic acid, activated carbon
Additive concentration: 50 to 200 ppm

Stirring speed: 150 rpm, time: 30 min
SS components are concentrated by filtration
SS concentration: 1000 ppm or more
Addition offlocculant

Addition of pH adjuster(NaOH)
pH adjustment range: pH 7±1
Stirring speed: 150 rpm, Time: 5 min

Low stirring

Stirring speed: 50 rpm, Time: 15 min

Supernatantseparation②③
Flocculant: Aluminum sulfate
Additive concentration: 200 ppm or moreHigh stirring

Static(1) Adsorption treatment
Soluble nuclides are adsorbed by adding a powder adsorbent
and stirring for a certain period of time.

(3) Flocculation sedimentation treatment
A flocculant is added to the concentrated liquid, adjusted to aneutral region with a pH adjusting reagent, and agitated for acertain period of time to cause flocculation and sedimentationof the powder adsorbent. (RO-concentrated water
sedimentation sludge generation)(4) Discharge of supernatant and sedimentation sludge
Supernatant: Filtration to remove residual particles after
extraction.
Sedimentation sludge: Extracted by pump or discharged in acollection container.

Static time: 60 min or more
Supernatant residual particles are 10 ppm or less
The amount of sludge generated is about 1 vol%against the initial liquid amount.

(2) Concentration treatment (adding reflecting theexamination results)
Powder adsorbents are concentrated, such as by cross-flow
filtration or dead-end filtration.
* Receiving RO-concentrated water → ① → ② are repeated tosecure a certain amount of concentrated liquid.

④

The volume reduction of sedimentation sludge by applying
dehydration treatment will be examined in the development item 3“Development of secondary waste treatment technology.”©International Research Institute for Nuclear Decommissioning48

No.49

7. Implementation details② Development of RO-concentrated water treatment technology- Summary[Results for this year]✓ Adsorbent tests were implemented, and powder adsorbents with high removal performanceagainst α-nuclides were selected. Activated carbon was evaluated as highly applicable forselective removal of α-nuclides, and titanic acid and titanium silicate were evaluated as highly
applicable for removing total activity including α-nuclides.
✓ A flocculant test was implemented, and aluminum sulfate was selected as a flocculant forflocculation and sedimentation of the powder adsorbent. In addition, data on the supernatantand the amount of sedimentation sludge generated and its properties were obtained.
✓ The water quality conditions under which solid-liquid separation by flocculant is effective wereorganized. As a treatment method for low SS concentration liquid waste, which is difficult toseparate solid-liquid using flocculant, a system applying concentration treatment was devised.
✓ Based on the results of this fiscal year, treatment procedures for RO-concentrated water weredevised.

[Future plans]✓ Beaker-scale RO-concentrated water treatment tests and equipment tests will be implemented
to obtain test data.
✓ Based on the results of each test, a method for treating RO concentrated water will be devised,and a conceptual system design for the system will be implemented.
©International Research Institute for Nuclear Decommissioning49

No.507. Implementation details③ Development of secondary waste treatment technology➢ Flowchart of examination[FY2019/2020 (previous project)]

Conceptual study ofparticle removal
systemArrangement of
Liquid systempattern[FY2021/2022 (this project)]Technological development
for other items(Canister PJ)

Investigation of existing knowledge
and organization of issues in storageof sludge or slurry form in canisters

Conceptual study ofsludge collection
systemDevelopment ofsludge stabilization
technology

Organization of issues in ROconcentrated water treatment
technological development- Flocculation
sedimentation method- Three-system water flow method

Filter test

- Properties of liquid
waste generated

Flocculation
sedimentation test
- Selection of flocculant- Beaker scale test
- Equipment test

- Acquisition of operational data
- Selection of particle capturecomponents- Properties of liquid waste
generated

Issue
identification

Conceptual
designDehydration
target setting- Evaluation of the amount ofhydrogen generated

Technological studyof dehydration

Conceptual
design- Differential pressure rise ofintermediate components- Increase in the amount of liquid
waste generated

Proposal of 2system waterflow method- Appropriate treatment method- Volume reduction by
dehydration, etc.

- Dehydration technology
selection result- Amount of sludge generated- Sludge properties- Sedimentation tankequipment specifications- Properties of liquid waste generated- System configuration plan andoperating plan
- Amount of liquid waste generated

Issue
identification

Evaluation of applicability
to actual machine byelement test

- Re-selection of intermediate
components- Differential pressure behavior,
backwash performance- Amount of liquid waste
generated

Conceptual
designElement test planning- Test plan

Particle shapedependence
evaluation

Items to bereported at theend of FY2021- Dependence of particle shape
distribution- Dehydration of sedimentation sludge- System
configuration plan
- Equipment
specifications

Detailed design©International Research Institute for Nuclear Decommissioning50

7. Implementation details

No.51③ Development of secondary waste treatment technology
1) Characterization of liquid waste generated from particle removal system➢ System configuration of particle removal system
The system configuration of particle removal system was examined based on the results of development up to
this point.
*1 Even particles with a size of 50 μm or less can

PCV stagnant
water

Roughing system

Intermediate system

Final treatment system

50 μm auto
strainer

1.4 μm MF filter *2

0.05 μm UF filter(ceramic filter)(ceramic filter)

Roughing
systemIntermediate
systemDrain water/backwash water
100 μm to 50 μm*1

Particle removal
range
Particle size Large

100 μm

50 μm to 1 μm

10 μm

*2 Performance assessment test for 1.4 μm MF
filter is scheduled to be implemented this fiscal
year

Soluble nuclide
removal equipment

Final treatment
system

Backwash water

Backwash water

1 μm to 0.1 μm

1 μm

be removed by centrifugation if they have a large
specific gravity such as fuel debris

Sludge collection
equipment(Adsorption tower, RO filter)

0.1 μmSmall

Colloid

Ion

Figure Conceptual diagram of the particle removal equipment system configuration
It is assumed that the fuel debris processing work performed in the PCV will atomize the processed matter, and particles ofapproximately 100 μm or less will float in the liquid and migrate to the liquid system.
Particles of 50 μm or larger are removed by the auto strainer in the roughing system. Particles with a high specific gravity(e.g., fuel debris, core structures) are also removed, where their centrifugal separation effect is significant. Small particles
that have passed through the roughing system are removed by the MF and UF filters in the intermediate and final treatment
systems, with the entire system removing particles down to 0.1 μm.
Captured particles are discharged by drain/backwash, and the liquid waste is treated in the sludge collection equipment.
©International Research Institute for Nuclear Decommissioning51

7. Implementation details

No.52③ Development of secondary waste treatment technology
1) Characterization of liquid waste generated from particle removal system: Roughing; 50 μm auto strainer➢ Examination of water flow cycle during fuel debris retrieval work

: intermediate; 1.4 μm MF filter

3-system configuration (Roughing→ Intermediate → Final treatment)
Backwashing (roughing)With particle
No particle treatment
treatment
Constant differential
Differential pressurepressure
increase

Backwashing (roughing/final treatment)
Backwashing (roughing/intermediate)With particle
With particle
treatment
treatment
No particle treatment
No particle treatment
Differential
Differential
Constant differential
Constant differential
pressure increasepressure increasepressurepressureDifferential pressure
recovers almost to theinitial value withbackwashing

Differential pressure ofcomponent

Drainage at regular intervals
and sequential recovering of
the differential pressure(intermittent drain): Final treatment; 0.05 μm UF filter

Since the auto strainer has a substantial
differential pressure recovery rate due tobackwashing, it is assumed that the initialdifferential pressure will not increase due
to water flow

MF and UF filters need to be replaced after acertain period of time as the initial differential
pressure gradually increases with repeatedwater flow and backwashing
Slight increase in initialdifferential pressure after
backwashing

Backwash once
every few cycles fordifferential pressure
recovery

Adjust the number of MF and UF
filter elements to optimize thebackwash and replacement cycles

Fuel debris
processing work
time
(10h)

From 1st day

Tool exchange, fueldebriscollection/storage/tra
nsferring time

Fuel debris
processing work
time
(10h)

Tool exchange, fueldebriscollection/storage/tra
nsferring time

From 2nd day

Work time

Fuel debris
processing work
time
(10h)

Tool exchange, fueldebriscollection/storage/tra
nsferring time

From 3rd day©International Research Institute for Nuclear Decommissioning52

7. Implementation details

No.53③ Development of secondary waste treatment technology
1) Characterization of liquid waste generated from particle removal system➢ MF filter water flow test
[Issue]

The intermediate filters such as an auto strainer, a sintered metal filter, and a bag filter wereselected and their applicability was evaluated by filter tests. However, both resulted in asignificant increase in the component differential pressure during water flow due to the small
effective filtration area and the small contribution of centrifugal separation effect to particles
of only a few μm. As such, a highly applicable intermediate filter component has not yet been
selected at this time.
[Purpose]
Evaluation of the applicability of a MF filter with pore size of 1.4 μm (ceramic filter) as anintermediate filter.
[Implementation details]
Implement a filter water flow test using a 1.4 μm MF filter to obtain operational data.
1.4 μm MF filter•••

It was verified that the UF filter selected as the final treatment filter has a very large effective filtrationarea, leading to a gentle increase in differential pressure
It is assumed that off-the-shelf identical ceramic filters with pore size of up to 1.4 μm can remove
particles in the intermediate filter range
The investigation confirmed that ceramic filters with pore sizes of 1.4 μm or larger are not available, orthat only tube types with a small filtration area are available©International Research Institute for Nuclear Decommissioning53

7. Implementation details

No.54③ Development of secondary waste treatment technology
1) Characterization of liquid waste generated from particle removal system➢ Particle shape dependence evaluation test
[Issue]
The effect of particle shape on filter performance has not been evaluated for the selectedfilter components in each system. All filter tests conducted up to the previous project usedspherical particles, and there is a possibility that the filter performance was not evaluated in aconservative manner.
[Purpose]
Evaluation of the particle shape dependence of the filter performance for the selected filter
components in each system.
[Implementation details]
A filter water flow test using spherical particles and non-spherical particles was
conducted to evaluate particle size dependence by comparative examination.
In addition, properties of the backwash water, which are the input conditions to thesludge collection equipment were examined.
Dominated by sphericalparticles
Flat particles were also observed,
but the abundance ratio wasinsignificant

Figure Particles used in the filter test of the previous project©International Research Institute for Nuclear Decommissioning54

7. Implementation details

No.55③ Development of secondary waste treatment technology
1) Characterization of liquid waste generated from particle removal system➢ Non-spherical particles used in filter test
Simulant particles were selected based on the assumption that the cutting debris generated bymachining process is passed through a filter.
Powders generated by machining process test of simulant fuel debris in theimplementation of the characterization project*

Simulant target

Relatively small particles: <1 μm
Relatively large particles: several μm

Particle size distribution
Formation of agglomerates of <1 μm powders on thesurface of a large particle of the order of several to 10 μm*FY2018 Subsidy Project of “Development of Analysis and Estimation Technology
for Fuel Debris Characterization”

Non-spherical particles used in thisstudy
Relatively large particles:
several μm

5 μm

Cumulative Passage (%)

Relatively small particles:<1 μm

Frequency (%)

Simulant particles

Figure. SEM image of a sample after cutting test of simulant
fuel debris implemented in the characterization project(excerpt)

Particle Shape
Dominated by Block-like particles

Particle size (μ/m)

Figure Simulant particles SEM image

Figure Simulant particles Particle size distribution

Particle size distribution and shape are
simulated based on the test report of thecharacterization project
Silica sand particles with a shape andparticle size that are relatively similar to
those used by particle manufacturers
are usedSizing to be performed as necessary©International Research Institute for Nuclear Decommissioning55

7. Implementation details

No.56③ Development of secondary waste treatment technology
1) Characterization of liquid waste generated from particle removal system➢ Test system for water flow test
For filter components, particle capture components of each system selected up to theprevious project were used.
The 1.4 μm MF filter, which has been evaluated as having high applicability based onliterature research, was only be used for the intermediate filter. Table list of particle capture components used
Main line:
No.

SystemParticle capturecomponent

Filter mesh size

1

Roughing

Auto strainer

50 μm

F

2

Intermediate

MF filter (ceramic filter)

1.4 μm

Backwash
pump

3

Final
treatment

UF filter (ceramic filter)

0.05 μm

P

Backwashwater tank

Flushing pump

Flushing filter

Table Evaluation Items
Particle inlet

P
F

Primary Main pumpwater tank

Receiving tank foroutlet water

P

No.

Items

1

Particle capture rate

2

Time of water flow until
backwash pressure isreached(Differential pressure
rising trend)3

Backwashing
performance(Differential pressure
recovery performance)

4

Backwash water
properties

Y

Particle capture component

Y-shaped strainer

Figure Filter test system configuration diagram

Drain water/backwash
water receiving tankIn this test, the performance of each filter was evaluated, and property data on©International Research Institute for Nuclear Decommissioning
the liquid waste was obtained.
56

7. Implementation details

No.57③ Development of secondary waste treatment technology
1) Characterization of liquid waste generated from particle removal system➢

Filter element replacement cycles
Replacement cycles for the particle capture components are assumed to be as follows
Roughing system components: 2 times/year → Rate-limiting corrosion ofmaterials due to chlorine*It is assumed that it can be operated
Intermediate system components: Evaluation for this year* → Rate-limiting
for a longer period of time because
increase in initial differential pressure
the pore size is larger than that of the
Final treatment system components: 5 to 8 times/year → Rate-limiting increase
final treatment system components
in initial differential pressure➢

Filter element replacement
It is necessary to evaluate worker exposure due to residual particles in the housing during the replacement
Fundamentally, a cartridge system will be adopted, and it is assumed that the entire housing will be replaced.
The housing should be sufficiently shielded to prevent exposure to workers.
Another possible method is to replace the elements remotely, but the feasibility of full remote control processing needs tobe studied.

➢Examination of the amount of filter liquid waste and secondary waste generated
This shows the filter specifications and approximate results of the amount of liquid waste and secondary waste generated
which was evaluated from the test results up to FY2020

Summary of filter performance assessment results based on previous element test results
SystemNumber of
requiredbaseParticle capture component

Auto strainer
Roughing

Intermittent
drain
Continuous
drain

Final
treatment

MF filter
UF filter

Rated flow

Base/1systemNumberm3/h

1

1

10 to 30

Removalefficiency

Approximate
dimensions(per reactor)

μm or more % or higher

Φ＿ m × ＿ mH

＿ mL × ＿ mW

Φ0.5 m × 1.0 mH

1.4 mL × 1.4 mW

Filteringaccuracy

50

99

Amount ofsecondarywaste
generated

Number oftimes
drained/backwashedAmount ofliquid waste
generatedtimes/year

kg/yeartimes/daym3/year

2

11

2.5

167
1300

11

Required area dimensions Replacement(per reactor)
frequency

1

1

10 to 3050

99

Φ0.5 m × 1.0 mH

1.4 mL × 1.4 mW

2

11

1

1

10

40

80

Φ0.4 m × 1.1 mH

1.6 mL × 1.6 mW--

Continuous
drain
Continuous
drain

0.1 μm

2

33

6.5

0.1

99

Φ0.21 m × 1.3 mH

3.0 mL × 1.6 mW

8

314

0.34

0.05 μm

3

46

<23

0.05

99

Φ0.3 m × 1.4 mH

1.4 mL × 1.4 mW
5
253
0.23
98.6
©International Research Institute for Nuclear Decommissioning

Liquid cyclone
Intermediate

Number ofelements
required

1.4 μm2000

Scheduled to be tested in FY2022

57

No.58

7. Implementation details③ Development of secondary waste treatment technology
2) Selection of dehydration technology for sedimentation sludge[Issues]• The amount of sedimentation sludge generated from the sludge collection equipment is highdue to the high moisture content, etc., so it is necessary to reduce the volume by dehydration.
• Since dehydration technology applicable to sedimentation sludge has not yet been selected, it
is necessary to select a candidate technology.
[Implementation details]• Assumed properties of sedimentation sludge andsupernatant were organized to select an applicabledehydration treatment technology through literature
research, etc.
• An element test plan was developed to evaluate theapplicability of the selected candidate technology for
actual equipment.
[Goals]• Selection of candidates for the sedimentation sludge
dehydration treatment technology.
• Planning of an element test for evaluation of applicability
to actual equipment.

Simulated liquid waste(Simulant particles + purewater)

Before flocculation sedimentation
treatment
Sedimentation sludge
Volume after flocculation:
Approx. 15 to 20%
Moisture Content: Approx. 90%
to 95%

After flocculation sedimentation
treatment

Simulantdrain water and backwash water©International Research Institute for Nuclear Decommissioning
flocculation sedimentation test58

7. Implementation details

No.59③ Development of secondary waste treatment technology
2) Selection of dehydration technology for sedimentation sludge
Dehydration technologies that can be applied to sedimentation sludge wereselected through literature research and element tests.
The applicability of the selected dehydration technologies for actual
equipment was examined through element tests.
The sedimentation sludge for the dehydration technology to be developed shall be thesedimentation sludge generated from the sedimentation tank for RO-concentrated water
treatment, where the amount generated is enormous.

[Implementation details]

Implementation items
1. Coordination with the canister project (connection conditions, etc.)
2. Consolidation of assumed properties of sedimentation sludge based on the previous project
results
3. Evaluation of the amount of hydrogen generated based on sedimentation sludge properties
4. Setting dehydration target (moisture content after dehydration)
5. Investigation of existing dehydration technologies and theoretical study of their applicability
6. Proposal of test plan that can evaluate the applicability to the actual equipment
7. Implementation of applicability evaluation test on the actual equipment
The following items were coordinated among project teams: Project of Development of Technology for Containing,
Transfer and Storage of Fuel Debris*1 and project of Development of Safety Systems (liquid and gas systems)*2.
〇 Liquid system - Canister storage line interface requirements〇 Sharing of development results*Canister PJ: Control policy for collection, transfer, and storage
Safety system (liquid and gas systems) PJ: Properties of sedimentation sludge at the time of discharge, etc.
〇 Classification of the future development contents of both projects*1 Project of Development of Technology for Containing, Transfer and Storage of Fuel Debris: Canister PJ©International Research Institute for Nuclear Decommissioning*2 Project of Development of Safety Systems (liquid and gas systems: Safety System PJ

59

No.607. Implementation details③ Development of secondary waste treatment technology
2) Selection of dehydration technology for sedimentation sludge[Implementation details] The volume reduction of sedimentation sludge due to dehydration was evaluated.
The volume reduction behavior of sedimentation sludge was evaluated when moisture was removed from theassumed moisture content of approximately 90%.

Volume after dehydration
when the initial value is
100 [-]In the region where supernatantwater and interstitial water decrease
is dominant, the sludge volume
reduction rate is large

In the region where the reduction ininterstitial water results in gaps, thesludge volume reduction rate issmall

Water content ratio below
60% is unlikely tocontribute much to thesludge volume reduction,
as the space occupied bymoisture results in gaps

Initial moisture
content

Appearance of
sedimentation sludge
UC (unit can)Dehydration

UC

60%

60%

Solid
content

Solid
content

Moisture(a) Water content ratioof 60% or more

Moisture(b) Water content
Water content ratio [vol%]ratio of 60% or lessFigure Relationship between moisture content
Figure Illustration of sludge volume reduction behavior
and volume of sludge
If the initial moisture content is 90 vol% or more, the volume can be reduced to nearly 20 vol% by dehydration treatment
to a moisture content of 60 vol%.
Volume reduction by dehydration has a small effect in the region where the water content is 60 vol% or less, because thevolume reduction reaches its limit.

Tentatively set the target value for the water content after dehydration treatment at 60 vol%*, and examine
dehydration treatment technology.
*Sludge content rate of
40 vol%Research Institute for Nuclear Decommissioning©International60

7. Implementation details

No.61③ Development of secondary waste treatment technology
2) Selection of dehydration technology for sedimentation sludge[Implementation details] The amount of hydrogen generated from sedimentation sludge during thetransferring period was evaluated.
The amount of hydrogen generated was evaluated based on the collection, transfer, andstorage conditions set in the canister PJ.
⇒ The results are shown on the next page

■ Collection, transfer, and storage conditions(1) Fuel debris: 100% UO2 with maximum burnup in the Fukushima Daiichi NPS(2) Waste filling rate: 30 vol% of fuel debris(3) During transferring- Transferring period: 1 week
- Gas vent (opening the coupler of the canister)Case 1 Canister: Vent, Transfer cask: Sealed
Case 2 Canister: Sealed, Transferring cask: Sealed(4) During storage- Gas Vent Canister: Vent

■ Assumed properties of sedimentation sludge(a) Sedimentation sludge generated from the sedimentation tank for filter liquid waste
treatment: fuel debris content 100%(b) Sedimentation sludge generated from the sedimentation tank for RO-concentrated water
treatment: fuel debris content 0.1%**Rate setting based on the removal rate of nuclides (DF100) in the particle removal equipment and soluble nuclide©Internationalremoval equipment, the assumed concentration rate (2 to 4 times) in the RO filter,etc. Research Institute for Nuclear Decommissioning

61

No.62

7. Implementation details[Implementation details] The amount of hydrogen generated fromsedimentation sludge during the transferring period was evaluated.
Transferring period

Water content ratio [vol%](a) Filter liquid waste
sedimentation sludge

Period until hydrogenconcentration reaches 4% [days]

Period until hydrogenconcentration reaches 4% [days]③ Development of secondary waste treatment technology
2) Selection of dehydration technology for sedimentation sludge

CanisterTransfercask

Period required for canister to
Closed Closed
reach 4% hydrogenconcentration
Period required for transfer
Opened Closed
cask to reach 4% hydrogenconcentration

Transferring period

CanisterGas phase: 0.0133 m3
Sludge: 0.0115 m3
Transfer cask
Gas phase: 1.82 m3
Sludge: 0.16 m3

Water content ratio [vol%](b) RO-concentrated water
sedimentation sludge

Figure. Evaluation of sludge moisture content and amount of hydrogen generated
In both cases, it was verified that the amount of hydrogen generated was the largest when the amount of water and fueldebris content were around 40% to 60%.
If the filter liquid waste sedimentation sludge is dehydrated from the initial moisture content of about 90%, there is aconcern that the risk of hydrogen explosion will increase unless the water content is reduced to a few percent or less.
It is unlikely that the RO-concentrated water sedimentation sludge will reach the explosive limit of hydrogen concentration
during the transferring period regardless of the water content.

Even if RO-concentrated water sedimentation sludge is stored for a long period, the amount ofhydrogen generated would not reach the explosion limit, so the need for dehydration treatment is low©International Research Institute for Nuclear Decommissioning
from the viewpoint of the amount of hydrogen generated.
62

No.63

7. Implementation details③ Development of secondary waste treatment technology
2) Selection of dehydration technology for sedimentation sludge[Implementation details] Proven technologies were selected as the sludge collection treatment
technology for overseas nuclear facilities.
Table Results of sludge dehydration technological study at overseas nuclear power sites
No.

Treatment method

Advantages

Disadvantages

Examples/Ready-made products

Requires Backwash andbackwash water
treatment.

1

MF, UF filters (DE) filtration High DF, low cost.

2

Cross-flow (CF) filtration

Suitable for radioactive
colloid removal.

Clogging due to fouling.
Used in recent PWR liquid waste treatment.
Requires concentrated
Equipment available for treatment at 300 GPM
water treatment.

3

RO concentration

Suitable forconcentration of CF
filtered water.

Clogging due to fouling. Used in recent PWR liquid waste treatment.

4

Evaporator

High DF.

Requires scaling
treatment, highequipment cost.

-5

Cartridge filter

Space-saving, low cost.

Low waste collection
efficiency.

Solids Collection Filter (SCF) with shielding
function for treating large volumes, is used toremove S/P pool sludge, etc.

6

Decompression/thermal
dehydration(In-Drum dryer)

Waste can be reducedby 85 to 95%.

Treatment speed
Mobile equipment
depends on drying level.

Cartridge filters, which are assumed to be able to dehydrate sedimentation sludge to the target
moisture content of 60%, as well as decompression and thermal dehydration will be selected as
the dehydration technology for sedimentation sludge.
For other candidate technologies, applicability as concentration treatment technology andsecondary nuclide removal technology, which are assumed to have high applicability for each
©International Research Institute for Nuclear Decommissioning
technology, will be examined.
63

No.64

7. Implementation details③ Development of secondary waste treatment technology
2) Selection of dehydration technology for sedimentation sludge[Implementation details] Proven technologies were selected as the sludge collection treatment
technology for overseas nuclear facilities.
For sludge collection in accident reactors and weapon legacy systems, the technology used is selectedaccording to the sludge content.
*UK: Atomic Weapons Establishment (AWE), US: TMI, Hanford Site, etc.

Table Investigation of cases of dehydration technologies applied by sludge content
No.

Amount ofsludge1

Miniscule amount

SNF pool water

Cartridge filter filtration

2

Miniscule amountRadioactive liquid waste

Cartridge filter filtration + evaporator, RO
concentration

3

Small amount(sub-percent)Sludge during reprocessing,saltcake

Cross-flow filter filtration

4

Medium amount(a few percent)Long-term corroded SNF
pool

Dehydration through filters

5Large amount(several tens ofpercent)

Research reactor waste
Concentrates No. 2 to 4

(Solidified as is)

Target caseSludge removal, dehydration, and drying
methods

When the sludge content is small, there are many cases where filtration with a cartridge filter is adopted. If the sludge content is slightly low,cross-flow filtration is used to concentrate it to a few percent. However, raising the concentration to several tens of percent is difficult.
If the sludge content is high, instead of dehydration treatment, solidification treatment such as cement solidification or vitrification is adopted
depending on the activity level and final disposal method. If the treatment/disposal method is undecided, it is stored for a long period of time,making use of facilities such as underwater storage.
In many cases of overseas sites, the sludge content is increased by cross-flow concentration, drying
treatment, etc., and then solidified.
©International Research Institute for Nuclear Decommissioning

64

No.65

7. Implementation details③ Development of secondary waste treatment technology
2) Selection of dehydration technology for sedimentation sludge➢ Examination of RO-concentrated water treatment system
The case results show that in many cases, appropriate dehydration methods have been adopted foreach sludge content region, and the sludge content has been concentrated to the target sludge
content.

Selection of RO-concentrated water sedimentation sludge dehydration technology
Assumed moisture content of ROconcentrated water sedimentation
sludge (up to 10%)

Concentration of powder adsorbent
added (initial value)(0.01% to 0.1%)

Assumed route for dehydration treatment so far

Target value after
dehydration
treatment● Dead-end filtration + back wash
● Cross-flow concentration

Long-term storage● Underwater storage● Cement solidification● Vitrification● Flocculation sedimentation treatment by addition of flocculant● Decompression/thermal dehydration● Cartridge filter

Small

0.1

1

10

40%

100

Large

Sludge content [%]The following two proposals were examined as treatment processes for RO-concentrated water.
Plan 1:① Flocculation
sedimentation treatment

Plan 2:① Flocculation
sedimentation treatment→② Cartridge filter filtration→③ Long-term storage→② Decompression/thermal
dehydration→③ Long-term storage* For plan 2, since it is difficult to treat large volumes with decompression/thermal dehydration equipment, a system that reduces the amount ofliquid by concentration treatment (cross-flow, etc.) as a pretreatment was also examined.
©International Research Institute for Nuclear Decommissioning

65

No.66

7. Implementation details③ Development of secondary waste treatment technology
2) Selection of dehydration technology for sedimentation sludge➢ Examination of RO-concentrated water treatment system① Flocculation
Plan 1: sedimentation treatment→→② Cartridge filter filtration③ Long-term
storage

Cooling water circulation system
To PCV

PCV stagnant
waterParticle
removalfilterNo solids
Contains
solids

RO filter equipment

CoolerSoluble nuclide
removaladsorption tower

Surpluswater

To existingwater
treatment
facility

Permeable water

Soluble nuclide
removal RO filter

RO-concentrated water
Powder adsorbent tankElementtechnology
required

Element technology to
be applied

Nuclide removal
technology

Adsorption treatmentusing powder adsorbent

Adsorbent solidliquid separation
technology

Sedimentation treatmentusing flocculant

Sludgedehydration
technology

Filter-based filtration(cartridge filter)

Supernatantwater

Flocculant tankParticle
removalUF filter

Permeable water

pH adjustment
reagent tank
Sludge collection
Sedimentationtank

To existingwater
treatment
facility

Sludge dehydration
Cartridge filter

Permeable water

Sedimentation sludge
UF filter
Housing +filter +sludgeTo wastestorage line

To canisterstorage line

©International Research Institute for Nuclear Decommissioning
RO-concentrated water treatment facility

66

No.67

7. Implementation details③ Development of secondary waste treatment technology
2) Selection of dehydration technology for sedimentation sludge➢ Examination of RO-concentrated water treatment system① Flocculation
Plan 1: sedimentation treatment
RO-concentrated Powder adsorbent
water
Flocculant→② Cartridge filter filtration

Supernatant water
To existing water
treatment facility

RO-concentrated water
Adsorbent suspension sedimentation sludge Replacementof the entire
Based on the investigation of item ②, the amount of ROhousing
concentrated water sedimentation sludge generated was→

Housing
Also be used as a storage
container

Extraction

Sedimentation tank

③ Long-term
storage
UF filter element
Since the average particle size of thepowder adsorbent is on the order ofseveral μm, the filter pore size of
about 1 μm is assumed to beappropriate

Permeable water
To existing water
treatment facility

Cartridge filter equipment

evaluated to be about 1 vol%, so it is assumed that extraction
by pump suction is more appropriate than discharging it in a
Sludgesludge collection container*Verified in the FY2020 test that it has fluidity due to its high moisture
content of about 90% and can be extracted with a pump

To waste storageor
To canister storage

The structure of cartridge filters can be designed to be replaced as a module.
By handling the housing as a storage container, long-term storage is possible as it is.
Since it is not necessary to retrieve the element, worker exposure can be reduced if the housing has a shielding function.
It is also simpler to operate than element replacement and can therefore be expected to be operated
remotely.
©International Research Institute for Nuclear Decommissioning

67

No.68

7. Implementation details③ Development of secondary waste treatment technology
2) Selection of dehydration technology for sedimentation sludge➢ Examination of RO-concentrated water treatment system① Flocculation
Plan 2: sedimentation treatment →② Decompression/thermal
dehydration→③ Long-term
storage

Cooling water circulation system
To PCV

PCV stagnant
water

RO filter equipment

CoolerParticle
removalfilterNo solids
Contains
solidsSoluble nuclide
removaladsorption tower

Surpluswater

Soluble nuclide
removal RO filter

To existingwater
treatment
facility

Permeable water

RO-concentrated water

Supernatant
water

Powder adsorbent tankElementtechnology
required

Element technology to
be applied

Nuclide removal
technology

Adsorption treatmentusing powder adsorbent

Adsorbent solidliquid separation
technology

Sedimentation
treatment using
flocculant

Sludgedehydration
technology

Decompression,
thermal dehydration

Flocculant tankpH adjustment
reagent tank

Sludge collection
Sedimentation
tank

Sedimentation
sludgeSludge dehydration
Decompression, thermal
dehydration equipment

Particle
removalUF filter

Vapor

Permeable water

Vaporcollection
Condenser

Condensed
water

Dried sludge (filled in container)

To existingwater
treatment
facility

To existingwater
treatment
facility
To wastestorage line
To canisterstorage line

RO-concentrated water treatment facility©International Research Institute for Nuclear Decommissioning
68No.69

7. Implementation details③ Development of secondary waste treatment technology
2) Selection of dehydration technology for sedimentation sludge➢ Examination of RO-concentrated water treatment system① Flocculation
Plan 2: sedimentation treatment
RO-concentrated Powder adsorbent
water
Flocculant→② Decompression/thermal
dehydration→③ Long-term
storage

Supernatant water
To existing water treatment facility
VacuumP
Extraction

Vapor

Condenser

Condensed water
To existing water
treatment facility

Heat up

Dried sludge...
Sedimentation tank
Adsorbent suspension

RO-concentrated water
sedimentation sludge

To waste storageor
To canister storage

Decompression, thermal dehydration equipment
Heating container

Based on the investigation of item ②, the amount of RO-concentrated water sedimentation
sludge generated was evaluated to be about 1 vol%, so it is assumed that extraction by pumpsuction is more appropriate than discharging it in a sludge collection container

Also be used as a storage container*Verified in the FY2020 test that it has fluidity due to its high moisture content of about 90% and can be extracted with a pump

RO-concentrated water sedimentation sludge is transferred to the container for dehydration treatment, and the water is evaporated bydecompression and heating.
The generated vapor is collected and condensed in the condenser and discharged to the existing water treatment facility.
Although dried sludge remains on the container side, it is assumed that the dead space inside the container is large due to the volume
reduction.
Therefore, the cycle of receiving sedimentation sludge→ dehydration treatment→ residual dried sludge is repeated to increase the fillingrate of dried sludge in the container. After it is filled to a certain level, the entire container is discharged and stored for a long period of time.
©International Research Institute for Nuclear Decommissioning
By giving the container a shielding function, worker exposure can be reduced.

69

No.707. Implementation details③ Development of secondary waste treatment technology
2) Selection of dehydration technology for sedimentation sludge➢

Examination of RO-concentrated water treatment system (when adopting cross-flow concentration)① Cross-flow② Flocculation③ Decompression,④ Long-term
concentration→ sedimentation treatment → thermal dehydration→
storage
Cooling water circulation system
No solids
To PCV

PCV stagnant
waterParticle
removal filter

RO filter equipment

Cooler
Surpluswater

Soluble nuclide
removal adsorption
tower

Soluble nuclide
removal RO filter

Contains
solidsTo existingwater
treatment
facility

Permeable water

RO-concentrated water

Elementtechnology
required

Element technology to
be applied

Nuclide removal
technology

Adsorption treatmentusing powder adsorbent

pH adjustment reagent
tank

Flocculant tank

Solid content
concentration
technology
Adsorbent solidliquid separation
technology
Sludgedehydration
technology

Cross-flow filtration
Sedimentation
treatment using
flocculant
Decompression,
thermal dehydration

Backwashwater tank

Powder adsorbent tankpHadjustment
reagent tankParticle removal
Cross-flow filter

Sludgecollection
ConcentratedAdsorption tankwater

Sedimentation
sludgeParticle removal
UF filter
Vapor

Permeable water

Condensed
Vapor collection water
Condenser

Sludge dehydration
Decompression, thermal
dehydration equipment Dried sludge (filled in container)

RO-concentrated water treatment facility

To existingwater
treatment
facility

Suspension (concentration treatment)

Supernatant water
Sludge collection
Sedimentation
tank

Permeable water

To existingwater
treatment
facility
To wastestorage line
To canisterstorage line
©International Research Institute for Nuclear Decommissioning70

No.71

7. Implementation details③ Development of secondary waste treatment technology
2) Selection of dehydration technology for sedimentation sludge➢ Examination on element test of sludge dehydration technology
The applicability of candidate technologies that were selected by the literature research
was verified through element test.
Sludge dehydration
technology

Cartridge filter

Decompression,
thermal dehydration(Solidification
treatment)

Concept

Items to be verified inelement test

Possibility of
dehydration up to thetarget value of 60%

Element
test
Necessity

Possible

Required

Possible

Not required――- Selection of filtration method
Depending on the sludge properties (particle sizeand conditions
distribution, moisture content, etc.), it is necessary to select- Dependence of dehydration
an appropriate filter element (pore size, filtration area),
performance on sedimentation
filtration conditions, etc.
sludge properties
It is assumed that the dehydration treatment can be
performed up to the target value without greatly depending―on the sludge properties (particle size distribution, particle
properties)
Solidification treatment is considered to be effective fromthe viewpoint of stabilization of sludge type waste, but theconcept of final disposal of waste in the Fukushima Daiichi
NPS is currently under examination. Since solidification―
treatment should be performed according to the burial
requirements, it is not examined in this development, andlong-term storage after dehydration treatment is estimated.

Element tests for cartridge filters were conducted to obtain operational data.
➢*System using decompression and thermal dehydration will also be examined with reference to literature data, etc.

Small-scale test (assuming 0.1 to 1L scale)
A filtration method which the sedimentation sludge moisture content after dehydration treatment satisfies the target valueof 60 vol% and the amount generated is at a level that can be discharged to a later stage equipment was selected.
Tests for Vacuum filtration and pressure filtration using RO-concentrated water sedimentation sludge simulants wereconducted. Information on average cake resistance, wet-dry mass ratio, etc. to evaluate whether the requirements for©International Research Institute for Nuclear Decommissioning
throughput and dehydration effect can be satisfied. (Test content under consideration)71

7. Implementation details

No.72③ Development of secondary waste treatment technology- Summary
1) Characterization of liquid waste generated from the particle removal system[Results for this year]✓ A 1.4 μm MF filter was adopted as the intermediate filter. Non-spherical particles were selected
to evaluate particle shape dependence of filter apparatus for each system. A test method offilter test was devised to evaluate the performance of each filter.
[Future plans]✓ A filter water flow test (both scheduled for FY2022) will be conducted to evaluate theapplicability of the 1.4 μm MF filter (intermediate component) and the particle shapedependence of each filter apparatus.

2) Selection of dehydration technology for sedimentation sludge[Results for this year]✓ The required dehydration performance was examined from the viewpoint of the amount ofhydrogen generated and volume reduction, and the sludge dehydration target was tentatively
set at a moisture content of 60 vol%.
✓ Based on literature research, cartridge filters and decompression and thermal dehydration wereselected as dehydration treatment methods applicable to sedimentation sludge.
[Future plans]✓ Element tests will be planned and conducted to examine the applicability of cartridge filters.
✓ The sludge dehydration treatment system that was selected candidate technologies (cartridge
filters, decompression, thermal dehydration) through literature research
will be studied.
©International Research Institute for Nuclear Decommissioning72

8. Specific objectives for achieving the purposes of technological developments No.73(1) Liquid/gas systems① Development of soluble α-nuclide removal
technology- Examination of soluble α-nuclide removal
test assuming use of actual liquid

In order to verify the effectiveness and feasibility of the soluble α- nuclide
removal equipment, the test method, test equipment and test plan that assume
the use of actual liquid shall be specified.
(TRL at the end of the project: Level 3)- Element test assuming fuel debris retrieval
work

Regarding the soluble α-nuclide removal equipment, the α-nuclide adsorption
performance data in the environment assumed during fuel debris retrieval work
shall be obtained, and candidate adsorbents shall be selected. At that time, awater quality adjustment policy shall be established.
(TRL at the end of the project: Level 4)② Development of RO-concentrated water
treatment technology- Selection of adsorbent and condensing
agent

Regarding the RO-concentrated water treatment facility, the nuclide removal
performance of the powder adsorbent and the flocculation sedimentation
performance of the flocculant shall be evaluated and the powder adsorbent andflocculant shall be selected.
(TRL at the end of the project: Level 4)- Examination of applicability to actual
equipment③ Development of secondary waste treatment
technology- Investigation of pretreatment technologies- Examination of applicability to actual
equipment

Regarding the RO-concentrated water treatment facility, the conceptual designof the treatment method and equipment shall be implemented from the test using
the sedimentation tank.
(TRL at the end of the project: Level 4)
A candidate stabilization treatment technology shall be selected for the sludge
generated from the flocculation sedimentation tank. At that time, pretreatment
technology for liquid waste such as supernatant water shall also be selected.
(TRL at the end of the project: Level 3)

The applicability of the pretreatment technology to the actual equipment shall beverified based on the results of element tests. In addition, the conceptual designof the pretreatment equipment shall be established.
(TRL at the end of the project: Level 4)*TRL: Technology readiness©International
levels Research Institute for Nuclear Decommissioning73

No.74End©International Research Institute for Nuclear DecommissioningSupplementary Budget for Government-Led R&D Program on
Decommissioning and Contaminated Water Management“Development of Safety Systems (Liquid/Gas Systems, Criticality Control
Technology)”

Accomplishment Report for FY2021(2) Criticality Control Technology
August 2022
International Research Institute for Nuclear Decommissioning(IRID)All rights reserved.
©International Research Institute for Nuclear Decommissioning (IRID)

No.1

Table of Contents
1 Purposes and goals of subsidy projects
2. Overview of subsidy projects
3. Previous fiscal year project results and remaining issues
4 Input/output
5. Implementation schedule
6. Project organization
7 Implementation details
7.1 Development of on-site operating procedures
7.1.1 Operating procedures for criticality approach monitoring
7.1.2 Verification of applicability of neutron detectors to sub-criticality
measurement
7.1.3 Operating procedures for neutron absorbers
7.2 Development of solidified absorber technology (water glass)

8. Specific objectives for achieving the purposes oftechnological developments©International Research Institute for Nuclear Decommissioning1. Purposes and goals of “Development of Safety Systems (Liquid/Gas
Systems, Criticality Control Technology)”

No.2[Purpose of Development of Safety Systems (Liquid/Gas Systems, Criticality Control Technology)]
At the Fukushima Daiichi Nuclear Power Station (NPS) of Tokyo Electric Power Company Holdings, Inc.
(TEPCO), it is believed that nuclear fuel melted together with the reactor internals and remains as fuel debris
inside the Reactor Pressure Vessel (RPV) and the Primary Containment Vessel (PCV).
The fuel debris in the RPV and PCV is likely present in a sub-criticality state. However, due to damage to theReactor Building (R/B), RPV, PCV and other structures, the entire nuclear power plant is in an unstablecondition deviant to the original design. In order to stabilize the plant, the fuel debris must be retrieved while
maintaining its sub-criticality state and preventing radioactive material diffusion.
This project will be examined within the above-stated background context and based on the “Mid-and-Long
Term Roadmap towards the Decommissioning of TEPCO’s Fukushima Daiichi Nuclear Power Station”(hereinafter referred to as the “Mid-and-Long-Term Roadmap”), with the goal of realizing large-scale fuel debris
retrieval work under TEPCO engineering and project management. The development results of this project will
be utilized in the engineering conducted by TEPCO.
The purpose of this project is to advance the level of Japanese science and technology by supporting thedevelopment of technologies that contribute to the safe decommissioning of the Fukushima Daiichi NPS andhandling of contaminated water, based on the Mid-and-Long-Term Roadmap and the “FY2021 Decommissioning
Research and Development Plan” (the 86th Secretariat Team Meeting for Countermeasures for
Decommissioning and Contaminated Water Treatment).
Specifically, development will be implemented on the technology to remove soluble α-nuclides that areconsidered to be eluted in circulating cooling water from fuel debris, treatment technology of RO-concentrated
water, treatment technology of secondary waste, and on-site operating methods for criticality approachmonitoring and neutron absorber technologies.

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

No.3[Overall development goal]
Applying the R&D results obtained thus far, the element technologies necessary for systems
and safety assurance technologies for further increasing the scale of retrieval of fuel debris
and the reactor internals will be developed and tested.

[Specific goals of development items]① Development of on-site operating procedures⚫ Development of on-site operating procedures for criticality approachmonitoring using neutron detectors to share among the fuel debris retrieval
method team.
⚫ Evaluation of the sub-criticality measurement performance of the three new
types of neutron detectors to reflect it in the on-site operating procedures.
⚫ Development of on-site operating procedures for non-soluble neutron
absorbers to share among the fuel debris retrieval method team.
② Development of solidified absorber technology⚫ Acquisition of data on the impacts of fuel debris with solidified absorbers(water glass) on the drying process to utilize it for ① Development of on-site
operating procedures as well as examination of equipment operation suchas fuel debris drying time.
©International Research Institute for Nuclear Decommissioning

No.4

2. Overview of subsidy projects
Development of on-site operating methods for criticality approach monitoring andneutron absorber technologies

As for the criticality control technology, each element of criticality prevention, criticality
approach monitoring, criticality detection, and impact mitigation has been developed in the
Project of Decommissioning and Contaminated Water Management by the previous fiscal year.
Of these, criticality prevention and criticality approach monitoring, which are still the issues, will
be examined and specified based on the results of technological development up to theprevious fiscal year.
③ Criticality
detection② Criticality
approach monitoringa. Restrict and control
procedures toprevent criticality① Criticality
prevention
b. Injection of neutron absorbers
Solid type

Solidified type

Targetrange ofthis project
Gd2O3
B4C sintered
B/Gdmetal material containing particles
glass material

Waterglass/Gd2O3
granulated
powderNeutrondetector

Gas
radiation
monitor

Emergency
boric acidinjection④ Impact
mitigation

Overview of technological development for
criticality control©International Research Institute for Nuclear Decommissioning3. Previous fiscal year project results and remaining issues

No.5

Development results of criticality approach monitoring technology and details of this project
Final target

Establishmentof subcriticality
measurement
technology

Goals to be achieved until
application to actual
equipment

Status

① Establishment of
measurement method- A method that combines the reactor noise method and the neutron
source multiplication method was selected.
- Detector sensitivity and placement to reduce measurement errors② Development of system
specifications- System specifications for detectors, measurement circuits, etc. weredeveloped.

③ Specifications for neutron
detectors- Specifications for neutron detectors were developed and a detector
for prototype was selected.

④ System design and test
manufacturing⑤ Verification of feasibility of
sub-criticality measurement

Remaining Issues/Details of this Project- A system consisting of a neutron detector, measurement circuit, andan analysis PC was manufactured as a test system.
- The feasibility was verified under simulated conditions ofhomogeneous fuel debris (KUCA test #1/#2).
- The feasibility was verified under large-scale fuel debris simulated
conditions (KUCA test #3)- The feasibility was verified under simulated conditions of nonhomogeneous fuel debris / neutron absorber (KUCA test #4).

Applicability verification test foralternative neutron detectors (coronatype, SiC-type, multi-cell He-3 based)(KUCA test #5) (→ Section 7.1.2)
KUCA: Kyoto University Critical Assembly- The specifications of the neutron detector that will be transferred by
the robot arm were organized.
handling concept was developed.
① Verification of applicability of -- Cable
Alternative neutron detector prototype (corona-type, SiC- based,
neutron detectors to the onmulticell He-3 type) was manufactured and the basic performancesite
was confirmed.
- Neutron detector units were designed and test-manufactured.
- Radiation resistance of preamplifier was evaluated.
Incorporation
Electromagnetic noise
into fuel debris ②reduction
methodretrieval system③ Examination of criticality
approach control procedures

The impacts of simulated noise were evaluated and countermeasures
for the noise impact were examined.
Measurement procedures before andafter processing, determination of
measurement targets, measurement
The time required for measurement was evaluated.
daily work procedures for subProcedures for criticality approach monitoring at each step of the fuel time,
criticality
maintenance control,
debris retrieval method
repair/maintenance requirements, andorganizationof forrequirements
to the©International
Research Institute
Nuclear Decommissioning
processing side (→ Section 7.1.1)

3. Previous fiscal year project results and remaining issues

No.6

Development results of non-soluble neutron absorbers and details of this project
Final target

Establishment ofcriticality
prevention
technology

Goals to be achieved until
application to actual
equipment① Selection of candidate
materials- Candidate materials for non-soluble neutron absorbers wereselected after verification of basic characteristics, irradiation
characteristics, and nuclear characteristics (4 types ofcandidate materials were selected).

② Verification of constructability
and specifications of subcriticality maintenance conditions- The required amount of input was estimated according to
the assumed usage.
- Workability of chisel for crushed fuel debris was verified.
The distribution of absorber was confirmed and sub-criticality
maintenance was evaluated (solid absorber).

③ Evaluation of corrosion effects

Incorporation into
fuel debris
retrieval system

StatusRemaining Issues/Details of this Project- From the long-term irradiation test, the hydrogen generation
G value (*) was less than the design value, and the diluted
seawater with eluted absorbent components had a pH of 6 orhigher- The impacts of irradiation on rust inhibitor effect were
evaluated.

③ Evaluation of secondary
impacts- The number of canisters and the amount of waste increased
Impacts of water glass on fuel debris
by up to about 10% for solid absorbers and up to about 40%
drying (→ Section 7.2)
for solidified absorbers① Examination of injecting
equipment for non-soluble
neutron absorbers andoperating methods- A method for inserting absorbers was developed.
- Impacts of chisel processing was verified.
- Availability of underwater insertion was confirmed.

Amount of solidified absorbers used
Applicable conditions for processing
scenes Level 2→ Level 1 returnprocedures
Effectiveness of pre-insertion(→ Section 7.1.3)② Design of absorber injecting
equipment- The concept of the absorbent injecting equipment wasdeveloped based on the weight and size restrictions of theabsorbent injecting equipment and the absorbent
transportation path restrictions.

Downsizing and weight reduction of theequipment (→ Section 7.1.3)③ Verification of combination
with retrieval equipment- Applicable absorbers for fuel debris processing methods for©International Research Institute for Nuclear Decommissioning
each fuel debris location were examined.
(*) Hydrogen generation G value: Number of molecules of hydrogen generated when absorbing 100 eV energy

No.7

4. Input/output

This project is being conducted in collaboration with the Project of Development of Technology
for Further Increasing the Retrieval Scale of Fuel Debris and Reactor Internals, and the Project ofDevelopment of Technologies for Containing, Transfer and Storage of Fuel Debris.

ID

Demand-side project

Provider-side project

Development ofTechnology for Further
Increasing the Retrieval
Scale of Fuel Debris andReactor Internals

1

Development of Safety
Systems (Liquid/Gas
Systems, Criticality
Control Technology)2

Development ofTechnology for Further
Increasing the Retrieval
Scale of Fuel Debris andReactor Internals

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

3

Development ofTechnologies for
Containing, Transfer andStorage of Fuel Debris

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

Contents (overview)
Methods of removing
interfering objects,
methods of accessing fueldebris, methods ofprocessing and collecting
fuel debris and thereactor internals
On-site operatingprocedures for criticality
approach monitoring
using neutron detectors,
on-site operatingprocedures for nonsoluble neutron
absorbers
Data on the impacts offuel debris with waterglass (non-soluble
neutron absorber) on thedrying process

TimingInformation use

June
2021

Examination of on-site
operating procedures forcriticality approachmonitoring and absorbent
dispersion

March
2023

Examination ofthroughput of fuel debris
retrieval method

March
2023

Examination of the drying
process of canisters©International Research Institute for Nuclear Decommissioning5. Implementation schedule

No.8

Criticality control technology
Category

Subcategory

Major milestones(2) On-site operating
methods of criticality
approach monitoring
technology and neutron
absorber technology① Development of on-site
operating procedures(i ) Formulation ofoperating procedures forcriticality approachmonitoringa. Organization of
preconditions

FY2021

Plant operator(If there is asubcontractor or test
location, it is also listed)

Apr May Jun

Aug Sep Oct Nov

Project steering meeting(1) Hitachi GE

Dec Jan Feb Mar

Interim report

Apr May Jun

Interim reportJul

Aug Sep Oct Nov Dec Jan Feb Mar

Interim report

Final report

Organization of preconditions
Examination of procedures(2) Toshiba ESS
b. Examination ofproceduresJul

FY2022(3) Mitsubishi Heavy
Industries

Evaluation / review

c. Evaluation /review

Summary

Interim summaryd. Summary
Test Plan(ii) Neutron detector subcriticality measurement
test

a. Test planb. Testpreparation/
arrangements
c. Element test
d. Summary(iii) Formulation ofoperating procedures
for non-soluble
absorbersa. Organization of
preconditions(1) Toshiba ESS
Subcontractor: FEO /CETDTest location: KUCA
Nagoya University

Test preparation/arrangements
KUCA test(2) Mitsubishi Heavy
Industries
Subcontractor: IPL
Test location: KUCA

(1) Hitachi GE

Interim summary

Organization of preconditions
Examination of procedures(2) Toshiba ESS
b. Examination ofprocedures

Element test

Evaluation / review(3) Mitsubishi Heavy
Industries

Summaryc. Evaluation /review

Interim summaryd. Summary② Development ofsolidified absorbertechnology

Test Plan
a. Test planb . Test
preparation/material
procurementc. Element test
d. Summary(1) JAEA
Subcontractor: Inspection
Development Company
Ltd. Test location: JAEA

Test preparation / procurement

Test preparation / procurement
Parameter fundamental test

Fundamental test

Interim summary

Summary©International Research Institute for Nuclear Decommissioning

No.9

6. Project organization
Tokyo Electric Power Company Holdings, Inc.

➢ Coordination's for on-site applicability

International Research Institute for Nuclear
Decommissioning (IRID)➢ Coordination of general planning and overall technology
management➢ Coordination of technology administration including
technology development progress management

Partnerdevelopment
project teams
Hitachi-GE Nuclear Energy,
Ltd.

Toshiba Energy Systems &Solutions Corporation

Mitsubishi Heavy Industries,
Ltd.

[Element test, technicaldevelopment](1) Liquid/gas system development① Development of soluble α-nuclide
removal technology- Examination of soluble α-nuclide
removal test assuming the useof actual liquid (stagnant waterin the reactor building)- Element test assuming fueldebris retrieval work
② Development of RO-concentrated
water treatment technology- Selection of adsorbents andflocculants- Examination of applicability toactual equipment③ Development of secondary waste
treatment technology- Investigation of stabilization
treatment technology- Examination of applicability toactual equipment[Element test, technicaldevelopment](1) Liquid/gas system development① Development of soluble α-nuclide
removal technology- Examination of soluble α-nuclide
removal test assuming the useof actual liquid (stagnant waterin the reactor building)- Element test assuming fueldebris retrieval work

[Element test, technicaldevelopment](2) On-site operating methods forcriticality approach monitoring
technology and neutron absorbertechnology- Formulation of criticality
approach monitoring procedures- Performance assessment of
neutron detector sub-criticality
measurement(2) On-site operating methods forcriticality approach monitoring
technology and neutron absorbertechnology- Formulation of criticality
approach monitoring procedures(2) On-site operating methods forcriticality approach monitoring
technology and neutron absorbertechnology- Formulation of criticality
approach monitoring procedures- Performance assessment of
neutron detector sub-criticality
measurement- Formulation of operatingprocedures for usage ofabsorbers- Examination of the impacts of thefuel debris on drying process

Japan Atomic Energy Agency[Element test, technicaldevelopment](2) On-site operating methods forcriticality approach monitoring
technology and neutron absorbertechnology- Development of solidified
absorber technology

Development of Fuel
Debris Retrieval Method
Development ofTechnologies for
Containing, Transfer andStorage of Fuel Debris
R&D on Solid Waste
Treatment and Disposal

Development of Analysis
and Estimation
Technology for Fuel
Debris Characterization
Development ofTechnology for Detailed
Investigation inside PCVDevelopment ofTechnology for
Investigation inside the
RPVDevelopment of Analysis
and Estimation
Technology for Fuel
Debris Characterization
Development ofTechnology for Sorting ofFuel Debris andRadioactive Waste©International Research Institute for Nuclear DecommissioningNo.107. Implementation details
7.1 Development of on-site operating procedures
7.1.1 Operating procedures for criticality approach monitoring[Issues]In the related project (*), a method of criticality approach monitoring in the rough work steps of the top/side-entry
method is being examined. (Figure 1) In order to discuss the operating procedures and throughput with the fueldebris retrieval method team, it is necessary to further refine the work steps and materialize the criticality
approach monitoring procedures.

[Implementation details]⚫ Organization of the procedures for criticality approach monitoring
using neutron detectors in a step diagram for the fuel debris
retrieval method and processing/collection equipment studied in
the related project (*).
⚫ Study on the neutron response when approaching criticality
during fuel debris processing, and the procedures fordetermining criticality approaching.
⚫ Study on the measurement procedures before and after
processing, determination of measurement targets,measurement time, daily work procedures for sub-criticality
maintenance control, repair/maintenance requirements, andorganization of requirements to the processing side.
[Goals]⚫ Development of the on-site operating procedures forcriticality approach monitoring using neutron detectors toshare them with the fuel debris retrieval method team.

During fuel debris
processing
Pre-processing criticality
approach monitoring andcontinuous monitoring
during processing using a
detector juxtaposed to theprocessing tool

Detector
juxtaposed toprocessing tool

Cutting tool

Figure 1 Criticality approach monitoring during
fuel debris retrieval (example of side-entry)©International Research Institute for Nuclear Decommissioning(*) Related project “Development of Technology for Further Increasing the Retrieval Scale of Fuel Debris and Reactor Internals” (FY2019/2020)

No.11

7. Implementation details
7.1 Development of on-site operating procedures
7.1.1 Operating procedures for criticality approach monitoringa. Organization of preconditions
Development of criticality control involving the retrieval step
Step

Sub-stepArmmaximumloadingweightPossibility
Processingof prerestrictions/criticality insertionofcontrol methods
absorbers

Criticality approach monitoring
Criticality
approach
Other criticality
Continuous
Sub-criticality
monitoring forprevention measuresmonitoring
duringmeasurement
each constantprocessing
amount ofprocessing

Remarks (*2)

2. Removal of interfering objects in the PCV/preparation for fuel debris removal
No deposits presumed
to be fuel debris: No― (*1)
restrictions
With deposits (small
amount of
deformation):
Restrictions on the―(Loading into amount of cutting③ Removal 4)
PCV) up to powder falling onto theofProcessing/
100 kgbottom sedimentsinterfering cutting of(Mounted on With deposits (large
objects
interferingarm) up to 30 amount ofin the PCV objects
deformation):kg
Possible- Processing whileWhen
performing criticality
approach monitoring approachin- Restrictions on theg criticality
amount of cutting
powder falling onto thebottom sediments

Criticality control
Step division of
methods, processing
retrieval method
restrictions, etc.
(Insert sub-step on
Potential step tocriticality side)
apply non-soluble
Constraints of detectors andabsorbers
non-soluble absorbent
injecting equipment

―

―

―

―

―

- Prevention of cut
pieces from falling
onto sediments inside Deformation criterion
the pedestal
is minimum critical――- Prevention of heavy weightweight objects from (30 to 60 kg)
falling onto the top
sediments in thepedestal (measures to
Neutron flux
prevent contact with
Continuous
measurement/critic
monitoring during CRD, etc.)
ality approachprocessing may be - Measures to prevent
monitoring before
the cutting tool fromrequired as a
and after
falling)
backupprocessing/cutting

Discuss specificprocedures forcriticality approachmonitoring.

*1 “―” means not applicable.
*2 Left blank if there are no
special notes.
Points to note from criticality control(mainly countermeasures to prevent theoccurrence of accompanying events)©International Research Institute for Nuclear Decommissioning7. Implementation details

No.12

7.1 Development of on-site operating procedures
7.1.1 Operating procedures for criticality approach monitoringa. Organization of preconditions

Functions to configure criticality approach monitoring

Name

Detector for sub-criticality
measurement

Criticality approach monitoring
detector (Note 1)

Continuous monitoring
detector (Note 1)

Intended use

Understanding the state before
work

Decision to start processing

Detection of unexpectedchanges

Sub-criticality measurement

Criticality approach monitoring Continuous monitoring of neutron
before and after processing
flux during processing

Absolute value measurement of
neutron multiplication factor

Measurement of relative change in neutron multiplication factor

Function

Form

Mounted on arm

Juxtaposed to processing tool

Installed on platform /juxtaposed
to processing tool

Weight30 kg to 100 kg

30 to 50 kg

30 kg or less

Measurement time

From several hours to a week
(depending on-site environment)

Approx. 10 min

Continuous

Measuring position(Note 2)

Same as left/a place that can be
One point near the retrieval start Moves appropriately according toviewed from a distance from thepositionthe processing position
processing position(Note 1) Depending on the conditions
such as weak gamma rays at on-site andstrong neutrons in the fuel debris, it ispossible that both “criticality approachmonitoring” and “continuous monitoring”
can be used.
(Note 2) It is not fixed and may bechanged from time to time.

Installed on
platform©International Research Institute for Nuclear DecommissioningNo.13

7. Implementation details
7.1 Development of on-site operating procedures
7.1.1 Operating procedures for criticality approach monitoringb. Examination of procedures
Outline of implementation of 7.1.1.b- - Top access retrieval method - -While discussing
the operatingprocedures andthroughput withthe fuel debris
retrieval methodteam,examination of 10
items that wereraised as issues
wasimplemented.

① Criticality monitoring
method during topaccess retrieval(→ No. 14)- - - - - - - - - - - - Side access retrieval method - - - - - - - - - - -② Criticality monitoring
method during sideaccess retrieval(→ No. 15)③ Criticality monitoring
method for side-entry
PLAN-B method (→No. 16 to 18)- Method for transferring the
integrated unit -③ Criticality
monitoring method
for top-entry integral
transfer plan (→No. 16 to 18)④ Monitoring range of three types of criticality approach monitoring (→ No. 19)⑤ Organization of measurement procedures during fuel debris processing andretrieval work (→ No. 20)⑥ Organization of the roles of criticality approach monitoring (→ No. 21)⑦ Basic concept of criticality monitoring based on environmental conditions foraerial and underwater processing (→ No. 22)⑧ Organization of criticality risks in underwater processing (→ No. 23,24)⑨ Monitoring procedures in typical processing example (laser processing) (→ No.
25)⑩ Monitoring system
maintenance (→ No. 26)©International Research Institute for Nuclear DecommissioningNo.14

7. Implementation details
7.1 Development of on-site operating procedures
7.1.1 Operating procedures for criticality approach monitoringb. Examination of procedures ① Criticality monitoring method during top-access retrieval[Results of the study so far]
Taking the top-access retrieval
method as a typical example, themethod of criticality approachmonitoring in the fuel debris
retrieval step was investigated.
Criticality approach monitoring
consists of three kinds of neutron
measurements (← No.12), but it is
not necessary to perform three
types of monitoring in all work
steps. (← No. 11)

Detector installedon platform(continuous
monitoring)

Detector
juxtaposed toprocessing tool(proximity
monitoring)

Detector
juxtaposed toprocessing tool(proximity
monitoring)Remaining fuel in core

Upper grid plate

Detector installedon platform(continuous
monitoring)

Depending on the assumed
criticality risk, use the necessary
Detectormonitoring function properly.
juxtaposed to(3 kinds or 2 kinds ofprocessing tool(proximity
measurements;)
monitoring)
Detector mounted
on arm (subcriticality
measurement)

Detector installedon platform(continuous
monitoring)

Detector
juxtaposed toprocessing tool(proximity
monitoring)Core support plate

Detector mounted
on arm (subcriticality
measurement)

RPV bottom©International Research Institute for Nuclear DecommissioningNo.15

7. Implementation details
7.1 Development of on-site operating procedures
7.1.1 Operating procedures for criticality approach monitoringb. Examination of procedures② Criticality monitoring method during side-access retrieval[Results of the study so far]
A criticality approachmonitoring method was
examined for the side-accessretrieval method.
Of the three criticality
monitoring functions, subcriticality measurement isperformed by a detector
mounted on the arm, whilecriticality approach monitoring
and continuous monitoring
during processing are
performed by a detector
juxtaposed to the processing
tool, thereby reducing thefrequency of processing tooland detector replacement andminimizing the impacts on thefuel debris retrieval workthroughput.

Criticality monitoring method during side-access retrieval
Before fuel debris
processing
Sub-criticality
measurement using a
detector mounted onarm

Detector
juxtaposed toprocessing tool
Before processing
interfering objects
Status verification
using a detector
mounted on arm

Detector
mounted on arm
During fuel debris
processing
Pre-processing
criticality approachmonitoring andcontinuous monitoring
during processing
using a detector
juxtaposed to theprocessing tool
Detector
juxtaposed toprocessing tool

Cutting tool

Figure Criticality monitoring during
fuel debris retrieval (excerpt)

Detector
mounted onarmDuring processingof interfering
objects
Continuous
monitoring using a
Detectordetector
juxtaposed tojuxtaposed to tool
processing tool

Cutting tool

Figure Criticality monitoring during
removal of interfering objects(excerpt)©International Research Institute for Nuclear DecommissioningNo.16

7. Implementation details

7.1 Development of on-site operating procedures
7.1.1 Operating procedures for criticality approach monitoring b. Examination of procedures③ Criticality monitoring methods for side-entry PLAN-B method/top-entry integral transfer plan
[Details of study]⚫ For the work steps of the side-entry PLAN-B method and the modified method 2 for top-entry integral transfer plan B ,the criticality approach monitoring procedures when applying the processing and collection methods using thefollowing equipment were examined, and issues and risks were identified.

Loading and moving aneutron detector unit
(simulant) using a heavy
weight object
transportation equipment(Side-entry PLAN-B
method)Loading and moving
crushing and suction
equipment by heavy weight
object transportationequipment(Side-entry PLAN-B
method)Illustration of removal ofinterfering objects at thebottom of the reactor(Modified method 2 for topentry integral transfer plan B)

Illustration of fuel debris
retrieval from the bottom of thepedestal after transferring the
RPV bottom head and CRD
housing(Modified method 2 for topentry integral transfer plan B)©International Research Institute for Nuclear DecommissioningNo.17

7. Implementation details

7.1 Development of on-site operating procedures
7.1.1 Operating procedures for criticality approach monitoring b. Examination of procedures③ Criticality monitoring methods for side-entry PLAN-B method/top-entry integral transfer plan
⚫ Outline of the side-entry PLAN-B method and the modified method 2 for top-entry integral transfer plan B
Secondary boundary

Carry-in cell
Wastetransfer cell

Fuel debris
retrieval cell

Connecting
passage

Retrieval equipment
Task
container
Extension building

Maintenance cell

Connecting
sleeveFuel debris
transfer cell
Passage (Access tunnel)

Double lid
Transfer
carrier
Task
container

Transportation
container

Extendable
legs

Operationarm- Access from the equipment hatch- Use multiple operation equipment (some equipment are assembled in
the PCV)- Reactor internals including fuel debris are integrated with solidified material containing neutron
absorbers and transferred- Fuel debris at the bottom of the pedestal is retrieved by top-entry (lower right on the previous
page)©International Research Institute for Nuclear DecommissioningNo.18

7. Implementation details

7.1 Development of on-site operating procedures
7.1.1 Operating procedures for criticality approach monitoring b. Examination of procedures③ Criticality monitoring methods for side-entry PLAN-B method/top-entry integral transfer plan[Study results]⚫ Issues and risks were identified in
the table on the right for the work
steps of the side-entry PLAN-B
method and the modified method
2 for top-entry integral transfer
plan B.

[Future study]⚫ The optimum criticality monitoringprocedures for the identified
issues and risks will be examinedin collaboration with the project ofDevelopment of Fuel Debris
Retrieval Method.

No.

Issue

1

Can the presence and amount of fuel debris be confirmed from camera images, neutron flux
distribution, and characteristic gamma- ray distribution?

2

Is it possible to install neutron detectors with appropriate positional relationship and distancefrom the heavy weight object transportation equipment used to transport neutron detectors?

3
4

Is it possible to measure the degree of sub-criticality for fuel debris with a critical mass ormore and several tens of kilograms?
Is it possible to install neutron detectors when the fuel debris has an inclined surface?

5

Is it possible to transport a neutron detector on the platform and install the neutron detector
with appropriate positional relationship and distance?

6

By installing the equipment directly on the fuel debris, is there any reactivity input due to crack
growth or water intrusion?

7

When processing and collecting sediments by lowering the water level, how to manage lumps
that cannot be collected or accumulation of water in depressions?

8

It is assumed that the cutting surface will be lower than the bottom of the pedestal whenretrieving bottom concrete-mixed fuel debris. Therefore, it is considered necessary to takemeasures to prevent stagnation of cooling water that had been discharged from the worker
access tunnel.

9

Can neutron detectors be transported between the outer periphery of the shroud and the RPV,
and installed at an appropriate positional relationship and distance?

10
No.
1
2
3

Measures to prevent the outflow of the filling solidifying agent may prevent the cooling waterfrom being discharged, causing reactivity input (transitional state until the solidifying agent is
injected).
Risk
Reactivity input due to falling of equipment onto fuel debris, crack growth due to impact, andwater intrusion into the crack.
During processing, crack growth due to falling fuel debris, etc., exceeding critical mass.
Detachment of fuel debris from the solidified
part,Researchfall, crackdue
to impact, and©International
Institutegrowthfor Nuclear
Decommissioning
reactivity input by water intrusion into the crack.
18

No.19

7. Implementation details
7.1 Development of on-site operating procedures
7.1.1 Operating procedures for criticality approach monitoringb. Examination of procedures ④ Monitoring range of three types of criticality approach monitoring⚫ The monitoring ranges for each of the three types of criticality approach monitoring
functions were organized. (The figure shows an example of top-access retrieval)

Function: Sub-criticality measurement
Measurement point: One point near the
retrieval start position
Intended use: Grasping the state before
work

Function: Continuous monitoring of
neutron flux during processing
Function: Criticality approach
Measurement point: A place that can bemonitoring before and after processing viewed from a distance from the
Measuring point: Moves appropriately
processing position
according to processing position
Intended use: Detection of unexpected
Intended use: Decision to start processing changes
Continuous
Monitoring area
monitoring

Fuel debris

Monitoringarea

Sub-criticality
measurement

Fuel debris Planned processing location
Fuel debris Planned processing location
Criticality approach ©International Research Institute for Nuclear Decommissioning
monitoring
Monitoringarea

No.207. Implementation details
7.1 Development of on-site operating procedures
7.1.1 Operating procedures for criticality approach monitoring

Sub-criticality measurement

b. Examination of procedures ⑤ Organization of measurement
procedures during fuel debris processing and retrieval work Light blue; reactor noise⚫ The procedures for measurement
during fuel debris processing
and retrieval work was examined.
- Tthe neutron multiplication factor (keff) ismeasured by reactor noise method before
starting fuel debris retrieval work ①
This is the criterion for the neutron source
multiplication method.
- The neutron count rate is measured before
and after fuel debris processing. ②- Neutron count rate is continuously monitored
during fuel debris retrieval processing.
- The neutron multiplication factor is evaluated(keff) by the neutron source multiplication
method when the neutron count ratechanges.
- The work is suspended when the neutron
multiplication factor reaches the criterion- Use absorbers.
- The neutron multiplication factor is measured(keff) by reactor noise method after using
absorbers. ①
This is a re-measurement of the neutron
source multiplication criterion.

BeforeretrievalDuringinspection

Measurement of𝒌𝒆𝒇𝒇 by the reactor
noise method ① (upto 1 h)methodYellow; neutron source
multiplication method

Neutron flux before
work, 𝒌𝒆𝒇𝒇 (𝝓ref , 𝒌𝒓𝒆𝒇 )Sub-criticality
Monitoring
Measurement before
Fuel debris
measurement
area
retrieval ②
Criticality approach monitoring(1 to 2 m)
before and after processing
Neutron flux before retrieval(𝛷)Evaluation of keff before
retrieval by neutron source
multiplication method

Judgment of subcriticality by
retrieval 𝒌𝒆𝒇𝒇

Find 𝒌 by the relation
𝝓ref
𝟏−𝒌
=
𝝓
𝟏 − 𝒌𝒓𝒆𝒇No

Ok
Fuel debris
processing/retrieval

Suspension of
retrieval work

Criticality approach

Fuel debris Planned
monitoringprocessing
Monitoringarea
location

Continuous monitoring of
neutron flux during processing(always on, apart from the above
diagram)Use of absorbers©International Research Institute for Nuclear DecommissioningNo.21

7. Implementation details
7.1 Development of on-site operating procedures
7.1.1 Operating procedures for criticality approach monitoringb. Examination of procedures ⑥ Organization of the roles ofcriticality approach monitoring

Criticality approach monitoring
before and after processing⚫ The role of the three monitoring functions during criticality
approach was examined.

Sub-criticality measurement

Fuel debris

Monitoringarea

Sub-criticality
measurement

Neutron multiplication factor- The neutron multiplication factor (keff) is measured by reactor noise method before starting fueldebris retrieval work. ①
This is the criterion for the neutron source multiplication method.
- The neutron count rate is measured before and after fuel debris processing. ②- The neutron count rate is continuously monitored during fuel debris retrieval processing- The neutron multiplication factor (keff) is evaluated by the neutron source multiplication method
when the neutron count rate changes- The work is suspended when the neutron multiplication factor reaches the criterion- Use absorbers- The neutron multiplication factor (keff) is measured by reactor noise method after using
absorbers ①
Fuel debris- Decision to resume work - resumption of work
As described in No.14 to 15 sheets, the above monitoring functions are used according to theSuspension ofsituation of on-site.

retrieval work →Use of absorbers

k=1.00 criticality

Criticality approach

Plannedmonitoringprocessing
Monitoring area
location

Continuous monitoring of
neutron flux during processing
Continuous Monitoring area
monitoringk-alarm
Measurement ②
Measurement ②
Measurement ①

RemovalRemovalReactor noise
method©International Research Institute for Nuclear Decommissioning

Retrieval work (lapse of time)

Fuel debris

Planned processing
locationNo.22

7. Implementation details
7.1 Development of on-site operating procedures
7.1.1 Operating procedures for criticality approach monitoringb. Examination of procedures ⑦ Basic concepts of criticality monitoring based on theenvironmental conditions for aerial and underwater processing⚫ The environmental conditions for aerial
processing and underwater processing, which
are prerequisites for criticality control, wereorganized.
⚫ The risk of criticality in the air is small, and thesame control as underwater is not required, but
local attention such as accumulation of water isrequired.

Aerial processing

Underwater
processing- There is no water level above fuel debris.
- There is a possibility of accumulation of water underfuel debris.
- Cooling water is flowing.
- Water droplets are falling.
- Processing is performed while water is sprinkled aroundthe processing area to prevent dust from being
dispersed.

- Water level is formed above
fuel debris.
- From the start of processing tocollection, the process isperformed underwater.

Sprinkling water toprevent dust frombeing dispersed

Water droplets
are falling

Possibility of accumulation of water
under debris

Figure 1 Environmental conditions for aerial processing

Research Institute for Nuclear Decommissioning
Figure 2 Environmental©International
conditions
for underwater processing

No.23

7. Implementation details
7.1 Development of on-site operating procedures
7.1.1 Operating procedures for criticality approach monitoringb. Examination of procedures⑧ Organization of criticality risks in underwater processing⚫ The criticality risks of the processing method examined in therelated project (*) was evaluated based on the analysis, etc.
Use other than intended method
There is no criticality
risk in picking up
small pieces of fueldebris with a size ofup to 200 mm (sizethat fits in a canister)underwater.
However, attention isrequired when fueldebris retrieval workby using the robot
arm including stirringand crushing small
pieces of fuel debris.
Attention isrequired whenfalling small
pieces of fueldebris andcollapsing thedeposited small
pieces of fueldebris.

Processing where reactivity input may
exceed 0.5% dk
Attention is required when
performing multiple cuts atshort intervals (up to approx.
a few cm).

Processing that may exceed theminimum critical mass under the
optimum deceleration condition
Attention isrequired when
hoisting powdery
fuel debris over awide area
underwater.
(Indication: 46 cm
or more indiameter, 71 kg
or more inweight)Attention is required when
opening multiple holes withshort pitches between holes (upto approx. 100 mm) by core
boring.

Attention is required when a large amountof powdery fuel debris are accumulated
underwater. (Indication: radius of 35 cm or
more, weight of 499 kg or more)

When crushing fueldebris with a chisel,
there is a risk ofcriticality if theprocessing area
exceeds 16 x 16 x 16
©International Research Institute forcm.
Nuclear Decommissioning(*) Related project “Development of Technology for Further Increasing the Retrieval Scale of Fuel Debris and Reactor Internals” (FY2019/2020)

No.24

7. Implementation details
7.1 Development of on-site operating procedures

Table Criticality risk for eachprocessing method

7.1.1 Operating procedures for criticality approach monitoring
Classifi Processing
Properties/outlines (*)
b. Examination of procedures
cation methods⑧ Organization of criticality risks in
Pinch small pieces of fuel debris
Picking Grabbing
underwater processingup⚫ The criticality risk for each processing
method was organized separately foraerial processing and underwater
processing.
⚫ Even if there is no risk of criticality in theprocessing method itself, it is necessary
to pay attention to unexpected usage
and secondary effects other than theprocessing point.

Clamshell
bucketSuction Suction

Underwater
processing

No

CautionScoop up small pieces of fuel debris

No

CautionSuction/collection of fuel debris in powder form

No

CautionNo

Caution- Sampling process with a hole saw- Separate cutting of the core is required- Water is required to discharge chips and coolbladesUltrasonic core Capable of simultaneous processing of
drill
ceramics and metals

Cutting Circular saw

Ceramics can be processed, but tool damage issevere. The cutting edge is serrated.

Cutting wheel

Examples ofsecondary effects

Be careful if
there isaccumulation ofwater

CautionCautionCautionCapable of simultaneous processing of
ceramics and metals
Hydraulic cutter Hydraulic-driven scissors

CautionAWJ

Cutting with high-pressure water mixed withabrasives
Cutting width is dispersed to the surrounding
environment
Thermal cutting with laser heat

CautionPress the chisel against the object and crush itby hitting it with the up-and-down movement of
the piston

Yes

Laser cutting

Examples of
unexpected usage

Criticality risks
Aerialprocessing

Cutting Laser gouging Laser cutting of fuel debris surface

Opening Core boring
holes

From the previous
seat No.23

Crushing Hammer(chisel)CautionCaution©International Research Institute for Nuclear Decommissioning(*) Related project “Development of Technology for Further Increasing the Retrieval Scale of Fuel Debris and Reactor Internals” (FY2019/2020)

7. Implementation details

No.25

7.1 Development of on-site operating procedures
7.1.1 Operating procedures for criticality approach monitoringb. Examination of procedures⑨ Monitoring procedures in typical processing
example (laser processing)

⚫

⚫
⚫
⚫

⚫

⚫
⚫

The procedure of measurement during fuel debris processing
was examined using easy-to-understand laser cutting as a
typical example.
A condition was assumed in which a straight cut line generated
in the fuel debris, and water entered the cut line, causing it toapproach criticality.
It is assumed that the neutron multiplication factor of the fueldebris before processing is around keff=0.95, which does notexceed the criterion.
The neutron count rate before processing should bemeasured.
When processing starts, the reactivity of Δk=0.016 is added in
10 minutes of processing, and the neutrons discharged fromthe fuel debris increase by a factor of 1.5. (Figure)
The neutron count rate was measured after processing,and it was observed that the neutron count rate increased by afactor of 1.5. The neutron multiplication factor is estimated tohave increased from 0.95 to 0.967 by neutron source
multiplication method. (→ Sheet No. 20)
It is judged that the neutron multiplication factor has exceeded
the criterion, and the retrieval work is suspended (sheet No.
21).
Procedures for other processing methods will also be
examined in the future.

(Note) Although there is no risk of criticality due to a single cut, extreme condition was set as
an easy-to-understand example.

Criticality approach monitoring
before and after processing

a factorof 1.5

Processing length, effective
multiplication factor

Neutron count rate [1/s]⚫

[Evaluation conditions]
Fuel debris composition: 4 wt% enricheduranium structural material, no FP, Gd
The air cavities in the fuel debris are filled withwater, and keff=0.95 in the initial condition
before processing.
Assuming processing to insert a cutting linewith a width of 2 cm, a depth of up to 60 cm,
and a length of up to 60 cm. (Note)
The processing speed of laser cutting is 1
[mm/sec], and processing is performed for 10
minutes.
The MVP code was used to calculate theeffective multiplication factor, and the onepoint reactor kinetics analysis was used to
calculate the neutron response.
Neutron source intensity 1 [n/sec]
Neutron detector detection efficiency 100 [%]Processing time [s]©International Research Institute for Nuclear Decommissioning

Figure Neutron response simulation assuming laser cutting7. Implementation details

No.26

7.1 Development of on-site operating procedures
7.1.1 Operating procedures for criticality approach monitoringb. Examination of procedures⑩ Monitoring system maintenanceExposing the neutron source from thesource transfer equipment and verifying
the operation of the neutron detector

Neutron⚫ Regarding the maintenance of the criticality
detector
approach monitoring system, the concept ofunit
daily inspection, periodic inspection, andcorrective maintenance was organized (Table).
Slide lid
⚫ Daily inspections are performed by moving the
Neutrondetector unit to a maintenance area (assuming
sourceNeutrontransfersource
SFP, etc. (*)) in the building (Note 1) and by remoteequipment
control (Note 2). (Figure)
Lifting and lowering of⚫ Part of the periodic inspection and corrective
source holder
maintenance in the event of a problem are
Figure Maintenance area inside the building
decontaminated and transferred out of thebuilding, where workers carry out the work. Table. Maintenance items of the criticality approach monitoring system

Goal

Frequency

Inspection
details

Inspection items

MethodPreventive
maintenance

Daily check

System integrity,
noise impacts

Operation verification
by neutron source,
backgroundmeasurement

The detector unit will be moved to the remote maintenance area, and theneutron source equipment installed there will be used to perform indirect
maintenance of the fuel debris processing/collection equipment by remotecontrol. (It takes about 1 hour)

Periodic
inspection
1 to 2times/year

Detector/connec
tor/cable
integrity,
detector
sensitivity

Discriminator,
insulation resistance,
capacitance, cablecharacteristics, source
calibration

If remote maintenance in the maintenance area is difficult, remove thedetector unit from the fuel debris processing/collection equipment, lift itout of the PCV, decontaminate it, and then perform maintenance directly.
(It takes several days)

Irregular

Overhaul,repair,replacement,

Gas leak, insulation
deterioration, poor
contact, disconnection,
circuit diagnosis, etc.

If a system malfunction occurs and the detector cannot be restored in the
PCV, remove the detector unit from the fuel debris processing/collection
equipment, lift it out of the PCV, decontaminate it, and then perform
maintenance directly. (one week or more if disassembly/assembly isrequired)(Note 1) It is a routine operation
for the fuel debris retrieval
equipment with the detector
unit to move to themaintenance area and wait.
(Note 2) It is necessary todevelop a method for remotely
calibrating neutron detectors
using neutron source.

Corrective
maintenance©International Research Institute for Nuclear Decommissioning

26(*) Related project “Development of Technology for Further Increasing the Retrieval Scale of Fuel Debris and Reactor Internals” (FY2019/2020)

No.27

7. Implementation details
7.1 Development of on-site operating procedures
7.1.1 Operating procedures for criticality approach monitoringd. Interim Summary[Results of this year]
While discussing operating procedures and throughput with the fuel debris retrieval
method team (top-access retrieval/side-access retrieval/integral transfer), the following
items were specified and shared.
⚫
⚫
⚫
⚫
⚫
⚫

Three types of criticality approach monitoring roles and monitoring range
Measurement procedures during fuel debris processing and retrieval work
Concept of criticality monitoring in aerial processing and underwater processing
Criticality risks in underwater processing
Monitoring procedures when approaching criticality, using laser processing as a typical example
Monitoring system maintenance

These will be the preconditions to be considered when the fuel debris retrieval method is
to be implemented in the future.
[Future plans]
Based on the above procedures, the range of study will be expanded, such as
development of processing methods other than laser, to evaluate the impacts on thethroughput of fuel debris retrieval, aiming to optimize the operating procedures.
©International Research Institute for Nuclear Decommissioning

As of the middle of March 2022

No.28

7. Implementation details
7.1 Development of on-site operating procedures
7.1.2 Verification of applicability of neutron detectors to sub-criticality measurement[Issues]Three types of neutron detectors (corona-type, SiC-type, and multicell He-3 type) have beendeveloped in the related project (*) from the viewpoint of sub-criticality measurement (Figure 1). To
verify the on-site applicability of these detectors, it is necessary to validate the ability to measure
sub-criticality and evaluate the performance of criticality approach monitoring.
[Implementation details]
A sub-criticality system simulating fuel debris at the Kyoto University Critical Assembly (KUCA) wasestablished. A sub-criticality measurement test using the developed neutron detectors was
conducted. (Figure 2)
In addition, the neutron measurement
performance was evaluated undergamma-ray environment.
Preamplifier[Goals]⚫ Evaluation of sub-criticality
measurement performance ofneutron detectors⚫ Reflection in operating procedures
for neutron detectors assuming onsite conditions

Corona detector
Φ25.4×260 mm

Figure 1 Appearance of Corona-based
detector (10 B)

Figure 2 Kyoto University
Critical Assembly©International Research Institute for Nuclear Decommissioning(*) Related project “Development of Technology for Further Increasing the Retrieval Scale of Fuel Debris and Reactor Internals” (FY2019/2020)

No.29

7. Implementation details
7.1 Development of on-site operating procedures
7.1.2 Verification of applicability of neutron detectors to sub-criticality measurement
a. Test plan
Table. Summary of KUCA tests conducted in related projects up to the previous fiscal year and test items
Test 6

Neutron detector

Goal

KUCA tests(1st, 2nd)

B10 proportionalcounter

Verify that sub-criticality can bemeasured in a small system using
uranium fuel

KUCA tests(3rd)

B10 proportionalcounter

Verify that sub-criticality can bemeasured in a large (uniform) systemusing uranium fuel

KUCA tests(4th)

B10 proportionalcounter

Verify the effects of large, non-uniform
systems and neutron absorberplacement

KUCA tests(5th)Corona-based
Verify that alternative candidates for
SiC-type
neutron detectors can measure subMulticell He-3 based criticality

Results- Measurement error (2% to 10%) for different
degrees of sub-criticality, from near-criticality withneutron multiplication factor=0.95 to deep subcriticality with 0.7- Uncertainty of the mixing state of fuel debris
and water (neutron energy spectrum) has little
impacts on measurement errors- The measurable distance between the fueldebris and the detector is within 20 cm
underwater and within 35 cm in air
Verified that local sub-criticality near the detector
can be monitored (Monitoring of the entire
system is difficult)Measured sub-criticality in a system simulating
non-uniform fuel debris, and verified position
dependence.
Verified that the sub-criticality can be measured
by conducting a test in a state simulating thesurface application of non-soluble neutron
absorbers. (-15 to +14%)
June 28–July 9, 2021 Implemented

Target of this test

©International Research Institute for Nuclear DecommissioningNo.307. Implementation details
7.1 Development of on-site operating procedures
7.1.2 Verification of applicability of neutron detectors to sub-criticality measurement
a. Test plan
Table. Neutron detector candidates for sub-criticality measurement
No.

Detector name

ProviderClassific
ation
Fissionionizationchamber

1

Fission ionization chamber

Photonis

Tested2

Improved small-sized B-10
proportional counter

IRID/
Hitachi-GE

3

B-10 proportional counter

CETD

Target ofthis test4

He-3 proportional counter(Multicell-based He-3
proportional counter)

CETD

5

B-10 corona detector

RosRAO6

He-3 corona detector

RosRAO

7

SiC detector

IPL

8CMOS detector

IRID/Hitachi-GE

Gas
Detector

Semicond
uctor
detector

Remarks

Neutron detectors being considered foruse in the investigation inside the PCV(considered in the related project (*))
Neutron detectors that have been
investigated for the criticality project(verified that these can be used for subcriticality measurement by the Feynman-αmethod)Neutron detectors being considered foruse in the investigation inside the PCV(highly radiation-resistant neutron
detectors being considered in the related
project*)©International Research Institute for Nuclear Decommissioning

30(*) Related project “Development of Fundamental Technology for Retrieval of Fuel Debris and Reactor Internals (Development of Small Neutron Detectors)”

No.31

7. Implementation details
7.1 Development of on-site operating procedures
7.1.2 Verification of applicability of neutron detectors to sub-criticality measurement
b. Test preparation/arrangements

Appearance of neutron detectors whose performance was evaluated in the KUCA* test
Preamplifier

Corona detector

50×50×540 mm
SiC semiconductor-type neutron detector

Corona-based neutron detector (B-10/He-3)

Sensitive length 150 mm

Multicell-based He-3 proportional counter

Conventional B-10 proportional counter(Reference; not used for testing this time)©International Research Institute for Nuclear Decommissioning

31*KUCA: Kyoto University Critical AssemblyNo.32

7. Implementation details
7.1 Development of on-site operating procedures
7.1.2 Verification of applicability of neutron detectors to sub-criticality measurementc. Element test
KUCA* test system⚫ The test core was composed of a fuel assembly combining enriched uranium and polyethylene (solid
moderator)⚫ Neutron detectors to be tested were placed adjacent to the fuel assembly, time-series data of neutron
counts is collected, and analysis was performed by the Feynman-α method
Description

a
b
c
d
e
f
g
h
i

3/8” p32 EU fuel assembly
1/8” p72 EU fuel assembly
1/8” p54 EUEU fuel assemblyCentermount

3/8” p10 EU fuel assemblyAbsorbent sample cell
Old PE block cell
New PE block cell

Control rod

Void, KUCA instrumentation cell

j
k
l
m
n
o

Graphite cell

p
q
r
s
t
u

Safety bar cell (S4 to S5)Control rod full stroke

Am-BesourceC1 rod cell (adjustment rod)
C2 rod cell (adjustment rod)
C3 rod cell (adjustment rod)He-3 corona counter
IPL SiC detector
B-10 corona counter
CETD multicell He-3

FissionChamber

B-10 coronacounter

CETD
multicell
He-3

SiC detector

He-3 coronacounter

AmBe source

The KUCA test core and detector layout (keff=0.95)*KUCA: Kyoto University Critical Assembly

Actual test

©International Research Institute for Nuclear Decommissioning32

7. Implementation details

No.33

7.1 Development of on-site operating procedures
7.1.2 Verification of applicability of neutron detectors to
sub-criticality measurementc. Element test
Evaluation procedure for sub-criticality measurement (Feynman-α method)

B-10 corona detector
Keff=0.95
60 minutes measurement

Neutron* Measurement time [s]FissionPlot the variance-to-mean ratio (Y value)
Conceptual image of data to be acquired(time-list data)
B-10 corona detector
Keff=0.95
60 minutes measurement 1 − exp(− T ) 
Y (T ) = Y1 −



T


α (prompt neutron attenuation constant) is
determined by fitting (red line) with a theoreticalformula.

𝑘𝑒𝑓𝑓 =

1 − 𝛼ℓ1−𝛽Keff is determined from the measured α and precalculated l and β.
Time interval (sec)Frequency distribution of time-list data

The keff obtained above was compared with the keffcalculated by the analysis code (reference solution).
©International Research Institute for Nuclear DecommissioningNo.34

7. Implementation details
7.1 Development of on-site operating procedures
7.1.2 Verification of applicability of neutron detectors to sub-criticality measurementc. Element test
Results of KUCA* test (Corona-based, Multi-cell based)

Difference from reference solution ofeffective multiplication factor [% dk]⚫ The availability of corona-type and multicell He-3 type detectors was verified to measure subcriticality with the same level of accuracy as the conventional B-10 proportional counter (Figure andTable).
⚫ Regarding the multi-cell He-3 type detector, the event in which the counting value fluctuated and theoperation became unstable was verified.

Difference
fromreferencesolution

B-10
proportionalcounterHe-3
coronadetector

B-10
coronadetector

Multicellbased He-3
proportionalcounter+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

Figure Difference between measured value and reference
solution (keff=0.95)

Measurement
time
(min)

30, 3010, 10, 12,
33, 6010, 10, 12,30, 6010

Number of
data

2

5

5

1

Table Measurement results for each detector (keff=0.95)©International Research Institute for Nuclear Decommissioning

Since the keff calculation of the KUCA test system by the MVP code tends to overestimate by about 0.5% dk in the criticality state, this is corrected*KUCA: Kyoto University Critical AssemblyNo.35

7. Implementation details
7.1 Development of on-site operating procedures
7.1.2 Verification of applicability of neutron detectors to sub-criticality measurementc. Element test
Results of KUCA* test (SiC-type semiconductor)⚫ The degree of sub-criticality was evaluated after the data obtained by the SiC semiconductor-type detector wasprocessed by the Feynman-α method. As shown in the table below, it was verified that sub-criticality can bemeasured with a difference of about 1% when keff is around 0.9. When the sub-criticality is deep, the difference
from the reference value increases to approximately 7%, but it was verified that the sub-criticality can be measured
with the same accuracy as the conventional B-10 proportional counter.
SIC detector KUCA sub-criticality
measurement result

Table Measurement results for each test system
Test corekeff (measured value)

Case1-1(keff=0.95)
Case1-2(keff=0.97)
Case1-3(keff=0.98)
keff (reference value)

Case1-4(keff=0.70)

Reference Measured Difference
(%)
value
value

0.948

0.944

-0.4

0.962

0.958

-0.4

0.979
0.979

0.968
0.958

-1.1
-2.1

0.979

0.981

0.2

0.681

0.729

7.0

Remarks

Measurementswith different
pulse shapediscrimination
parameters

Reference value: Core analysis value by MVP

Figure Difference between measured and reference values(SiC semiconductor-type neutron detector)*KUCA: Kyoto©International
University
Critical
Assembly
Research
Institutefor Nuclear Decommissioning35

No.36

7. Implementation details
7.1 Development of on-site operating procedures
7.1.2 Verification of applicability of neutron detectors to sub-criticality measurement
Summary of results[Results of this year]✓ Sub-criticality measurement tests for three types of new neutron detectors(corona-type, SiC-type, and multicell He-3 type) were conducted at the Kyoto
University Critical Assembly (KUCA).

✓ It was verified that sub-criticality can be measured by using corona-type,multicell He-3 type, and SiC semiconductor-type detectors with the sameaccuracy as the conventional B-10 proportional counter.

[Future plans]✓ Neutron measurement tests for the corona-type and multicell He-3 typedetectors will be conducted under gamma-ray environment.
✓ The sub-criticality measurement performance will be comprehensively
evaluated, taking into account the effects of the assumed on-site gamma-ray
environment.

©International
Research
for Nuclear 2022
Decommissioning
As of themiddleofInstitute
March

36

No.37

7. Implementation details
7.1 Development of on-site operating procedures
7.1.3 Operating procedures for neutron absorbers[Issues]In the related project (*), a method of inserting non-soluble neutron absorbers into fuel debris is being
examined. (Figure 1) In order to discuss operating procedures and throughput with the debris
retrieval method team, it is necessary to specify the procedures for neutron absorber dispersion andinjecting equipment operation.
[Implementation details]⚫ For the fuel debris retrieval method and processing/collection
equipment examined in the related project (*), the procedures fordispersing non-soluble neutron absorbers to fuel debris wereconsolidated in a step diagram.
⚫ Study of applicable conditions for fuel debris processing scenes,
proper use of absorbers, effectiveness of pre-insertion, andreduction in size and weight of the injecting equipment.
⚫ Evaluation of the operability and feasibility of the entire
procedures together with the criticality approach monitoring
operating procedures (item 7.1.1).

[Goals]⚫ Development of on-site operation procedures for nonsoluble neutron absorbers and sharing them with the fueldebris retrieval method team.

Glass material
containing B/Gd

Gd2O3
particles

Water glass/Gd2O3
granulated powder

Example of injecting
equipment for solids(powder)

Figure 1 Illustration of inserting absorbers
into fuel debris(*) Related project “Development of Technology for Further Increasing the Retrieval Scale of Fuel Debris and Reactor Internals” (FY2019/2020)©International Research Institute for Nuclear Decommissioning37

No.38

7. Implementation details
7.1 Development of on-site operating procedures
7.1.3 Operating procedures for neutron absorbers
a. Organization of preconditions
Appropriate use of absorbers••

Various shapes such as rods and particles are assumed for fuel debris. So far, it has been shown that the optimum
type of neutron absorber should be used according to the shape of fuel debris.
Non-soluble neutron absorbers are classified into solid type and type that solidifies from liquid to solid over time (waterglass type). Since the water glass type has fluidity and viscosity immediately after insertion, the type is effective when
the residual fuel (stump fuel) stands vertically (Figure 5), when the gap between fuel debris is small (Figures 3 and 4),
or when the debris surface has large unevenness (Figure 6).
DebrisAbsorber (solid)

Figure 1 Using a solid typeabsorber for pebble-likefuel debris

Figure 2 Using a solid typeabsorber for bedrock-likefuel debris

Figure 3 Using a liquid →solidification type absorber for
pebble-like fuel debris

Figure 4 Using a liquid →solidification type absorberfor bedrock-like fuel debris

Figure 5 Using liquid →solidification type absorber for
rod-shaped fuel debris

Figure 6 Using a liquid → solidification typeabsorber for fuel debris with large unevenness©International Research Institute for Nuclear Decommissioning7. Implementation details
7.1 Development of on-site operating procedures
7.1.3 Operating procedures for neutron absorbers
a. Organization of preconditions Positioning of defense in depth⚫

Since the use of non-soluble neutron absorbers affects the throughput of fuel debris
retrieval, the policy is to use them at level 2 for the retrieval processing position.

Level 1 (normal state)

Level 2 (deviation state)

Retrieval work
restrictions (Note)(Note) This table covers events that become criticality due to fuel debris
retrieval work when boric acid solution is not used at all times. See No. 46
for events caused by falling fuel debris, etc.

Level 3 (abnormal state)

Criticality approach monitoring

Success(Termination of
approaching criticality andrestoration of state before
reaching criticality)

Functional
requirements

Criticality shall be
prevented by neutron
absorbers and retrieval
work shall besuspended by criticality
approach monitoring

Failure(Occurrence of criticality approach)

Retrieval work suspended

No.39

Failure (Occurrence of criticality)

Event progress stops

Non-soluble neutron
absorberCriticality detection
Neutron detector/Gas
sampling system

Emergency boric acid

Success (Criticality terminate) solution injection

Success (Criticality terminate/suppression of radioactive materials discharge to the outside)

Failure (Criticality
continuation)

The nuclear reaction
shall be suspended
when criticality occurs.
The impact ofradioactive materials
generated during
criticality shall be
reducedAdditional boric acid
Prevention ofsolution injection
excessive exposure of
Water level drop due tofeedwater shutdown
general public andSuppression ofradioactive
gas Researchdischarge©International
Instituteworkers
for Nuclear Decommissioning

39

No.407. Implementation details
7.1 Development of on-site operating procedures
7.1.3 Operating procedures for neutron absorbers
b. Examination of procedures
Outline of implementation details of 7.1.3.b

While discussing the operating procedures and throughput with the fuel debris retrieval
method team, six items identified as issues were examined.
- - Top access retrieval method - -- - - - -- - - - Side access retrieval method - - - - - - - - -- Method for transferring the
integrated unit -① Decision to use absorbers, procedure for returning to level 1 (→ No. 41)② Absorbent supply equipment specifications (→ No. 42)③ Absorbent operating processes (→ No.43)④ Absorber carry-in and –outprocess (→ No.44)⑤ Hardening time of solidified absorbers (→ No. 45)⑥ Evaluation ofeffectiveness of predispersion of absorbers(→ No. 46)©International Research Institute for Nuclear Decommissioning40

No.41

7. Implementation details
7.1 Development of on-site operating procedures
7.1.3 Operating procedures for neutron absorbers

b. Examination of procedures ① Decision to use absorbers, procedures for returning to Level 1
Continuous monitoring of
neutron flux during processing
Criticality continuous
monitor

Criticality continuous
monitoring area

Fuel debris Planned processing
location

Criticality approach monitoring
before and after processing⚫ Organized the decision to use absorbers and theprocedures for returning to level 1

Sub-criticality
measurement

Before fuel debris
processing

Continuous
monitoring normal

Sub-criticality
counterconnection

Proximity
monitoring
normalSub-criticality
normalFuel debris
processing
Continuous
monitoring normal

Proximity
monitoring initialvalue reset

Fuel debris

Neutron absorbersupplied

Sub-criticality
counter

Neutron absorber
insertion

Go to “Before fuel debris
processing”
Proximity
monitoring normal

Fuel debris collection
after processing
PlannedCriticality approach
Fuel debris processing
monitoring
location
Criticality approachmonitoring area

The reactivity
suppression effectafter insertion isquantified andverified by subcriticality
measurement.
Neutron absorption©International ResearchInstitute
Fueldebris for Nuclear Decommissioning
material41

No.42

7. Implementation details
7.1 Development of on-site operating procedures
7.1.3 Operating procedures for neutron absorbers
b. Examination of procedures ② Absorbent supply equipment specifications
Table Required amount of absorber⚫ Based on the required amount ofabsorber (Table) reviewed in the related
project (*), the specifications of theabsorbent supply equipment wasreviewed and the volume was reduced
to about half. (Figures 1 and 2)

Non-soluble neutron
absorber

Absorbent weight(kg/day)

Absorbent
capacity(liters/day)

6.4
6.3

1.5
3.0

(Note 1)

Gd2O3 particles
Water glass/Gd2O3
granulated powder

Note 1: It is assumed that the fuel debris retrieval target per day is 300 kg.
Assume that 50% of the inserted absorber functions effectively (safety factor doubled)

Air tube

Hopper 5L

Capacity: 5 L
Electric motor

screw

NozzlePressurized
container
Container before
pressurization ( 5 ℓ)Nozzle lid
opening/closi
ng cylinder

Container afterpressurization

NozzleSpringNozzleNozzleopening/closing lid

Figure 1 Supply equipment for solid type (Gd2O3 particles)

Concept for FY2021
Air cylinder
Figure 2 Supply equipment for solidified type (water glass/Gd
2O3
granulated powder)©International Research Institute for Nuclear Decommissioning(*) Related project “Development of Technology for Further Increasing the Retrieval Scale of Fuel Debris and Reactor Internals” (FY2019/2020)

No.43

7. Implementation details
7.1 Development of on-site operating procedures
7.1.3 Operating procedures for neutron absorbers
b. Examination of procedures ③ Absorbent operating procedures⚫ The assuming processes when using absorbers at
the site was considered (Tables 1 and 2).
⚫ In the case of the solidified type absorber(waterglass), since it is necessary to knead the raw
materials and to wash the injecting equipment withwater after use, the operating process is more
complicated than that of the solid type.
Table 1 Operating process of solid type (Gd2O3 particles, glassmaterial containing B/Gd)
No.
1

2
3

4

5

6

Operatingprocess
Storage ofabsorbers

Description

Weighing ofabsorberLoading absorber
into injecting
equipment

Weigh the amount of absorber needed for thework.
Load the absorber into the hopper of the injecting
equipment.

Transferring
injectingequipment to thesite
Insertingabsorber into fueldebrisConnect the injecting equipment loaded withabsorber to the manipulator and transfer it todirectly above the fuel debris retrieval site.

Return of
injectingequipment

After the completion of insertion, return theinjecting equipment to the predetermined position
from the fuel debris retrieval site.

Glass material containing B/Gd or Gd2O3
particles are stored ready for use as absorbers.

Drop the absorber from directly above the fueldebris retrieval site to be applied to the surface offuel debris.

Table 2 Operating process of solidified type (waterglass/Gd2O3 granulated powder)No.

Operating process

Description

1

Storage of raw
materials

2

Weighing of raw
materials

3

The five raw materials for the water glass neutron
absorber are ① water glass, ② gadolinia granulated
powder, ③ cement, ④ monosodium phosphate, and ⑤
water. Temporarily store these at the production site. ④
and ⑤ can be mixed and stored as an aqueous solution.
Weigh each of the raw materials required for the amountof absorber to be applied. There is also a method of
inserting a predetermined amount of raw material stored
in tanks into a kneader using a dedicated squeeze
feeder.
Knead the raw materials in a kneader to produce a waterglass neutron absorber.
Load the water glass neutron absorber prepared by thekneader into the hopper of the injecting equipment.

Kneading of raw
materials (*)
Loading absorber
into injecting
equipment
Transferring injecting Connect the injecting equipment loaded with water glass
equipment to the site neutron absorber to the manipulator and transfer it todirectly above the fuel debris retrieval site.
Inserting absorber
Drop water glass absorber from directly above the fuelinto fuel debris
debris retrieval site to be applied to the surface of fueldebris.
Return of injecting
After the completion of application, return the injecting
equipment
equipment to the predetermined position from the fueldebris retrieval site.
Cleaning of injecting Wash the used kneader and injecting equipment withequipment (*)
water to wash off adhering water glass absorbers.
Collect and store the wash water.

4
5
6
7
8

©International Research Institute for Nuclear Decommissioning(*) Processes in red are specific to the solidified type and are unnecessary for the solid type

7. Implementation details

No.44

7.1 Development of on-site operating procedures
7.1.3 Operating procedures for neutron absorbers⚫
⚫
⚫

For the solidified type, which has many operating processes, the transfer procedures from storage to use and collection were
examined (Figures 1 and 2)
To save the limited space in the R/B, the absorber will be stored and prepared in a separate maintenance building. In order to transfer
the contaminated absorbent injecting equipment between buildings, the transportation equipment (mobile cell) developed in the related
project (*) will be used. The mobile cell is for transferring unit cans collecting fuel debris.
The time required for transportation from preparing the water glass absorber, inserting it into fuel debris, and collecting the injecting
equipment is about 2 hours for the outward trip + about 4 hours for the return trip (for cleaning and exchanging air in the transfer area)= Estimated 6 hours total.
The reactivity suppression effect after insertion will be verified by sub-criticality measurement using neutron detectors. (No. 41)
Blue line: Secondary boundary
Park Cell (C4M) Remotely Operated

Maintenance cell (C4M)
Temporary entry of workers

C4M Maintenance cell

Carrying in mobile cell

Operation
by workers

Remotecontrol
Detaching injecting equipment frommobile cellContainer fortransport inside
the premise

Storage of raw
materials
Weighing of
raw materials

C4M Park cell

Elevator fortransportationequipment

Input
equipment

Transferring by unit can hanging
equipment

Robot arm connects injecting
equipment(Tool changer)
New operation floorload bearing gantry

Fuel debris retrieval
equipment(Lower side of platform)

Robot arm moves injecting
equipment to the fuel debris site
Mobile cell

Figure 2 Carry-in route of supply equipment

Dropping absorber on fuel debris

Remote control

Transferring by unit can transfer craneTransport of mobile cells⚫

Example of solidified type (waterglass/Gd2O3 granulated powder) for topaccess retrieval method (mobile cell method)b. Examination of procedures④ Absorber carry-in and -out process

Kneading of
raw materials
Radiationcontrolled area

Input
equipment

Operation
by workers

Loading absorber into injecting
equipment
Storing injecting equipment in mobile
cell

Maintenance building separate from R/B

Figure 1 Operating process flow up to insertion of absorber (water glass/Gd2O3
granulated powder)©International Research Institute for Nuclear Decommissioning(The process of collecting injecting equipment and mobile cell after insertion is separate)(*) Related project “Development of Technology for Further Increasing the Retrieval Scale of Fuel Debris and Reactor Internals” (FY2019/2020)

No.45

7. Implementation details
7.1 Development of on-site operating procedures
7.1.3 Operating procedures for neutron absorbers
b. Examination of procedures ⑤ Hardening time of solidified absorbers

Table Raw materials of water glass/ Gd2O3
granulated powder
Raw material
Main material

Classification
No. 1 Sodium Silicate (Baume specificgravity 55)

Hardener

Ordinary Portland cement

Retardant

Monosodium phosphate

Water

Ion-exchange water

Neutronabsorption
material

Gadolinium oxide

Viscosity [mPa·s]⚫ The issue of the solidified type absorber (water glass / Gd2O3 granulated powder) was clarified with respect to thetransportation time.
⚫ It is necessary to adjust the water glass absorber so that it does not solidify before insertion (up to 2 hours). In
addition, it is necessary to adjust so that the residue remaining inside the container after insertion does not harden
before cleaning. (Approx. 3 hours for cleaning in the dryer separator pit (DSP), approx. 6 hours for cleaning in themaintenance building)⚫ The range of viscosity that can be inserted and cleaned is 0 to 5000 [mPa·s].
⚫ The test was performed by changing the concentration of the retardant (Table) (0.95 to 1.87 wt%), and the time toreach 5000 [mPa·s] was measured.
20000⚫ As a result of the test, the prospect of delaying the hardening#1 to 5: Different amount of retardants
TX-10
time by more than 5 hours was obtained (Figure). Although it is
1.5
1.7(IRID, 2018)
1.4
a little short of 6 hours, it is thought that there is a possibility
15000
1.87
that it will be established if the transportation process is
0.95wt%1.3
examined in detail in the future.
10000
5000
Operating
range

0
0

60 120 180 240 300 360 420 480
Time [minutes]

Figure Elapsed time and viscosity test results after
mixing water glass/Gd2O3 granulated powder©International Research Institute for Nuclear Decommissioning7. Implementation details

No.46

7.1 Development of on-site operating procedures
7.1.3 Operating procedures for neutron absorbers
b. Examination of procedures ⑥ Evaluation of effectiveness of pre-dispersion of absorbers⚫
⚫
⚫
⚫

The applicability of non-soluble absorbers were examined to response to a criticality
risk outside of fuel debris processing locations. Specific criticality risks include the fallof equipment and fuel debris
Fuel debris
Absorbers should be inserted in advance since the local sub-criticality cannot be
RPV bottom
measured.
head
The effectiveness of preventing the criticality risk associated with falling fuel inside
CRD
housinga large transport cask was examined as a reference.
The necessity and applicability of measures against falling during work inside thepedestal will be examined.
Level 1 (normal state) Level 2 (deviation state)

Level 3 (abnormal state)

FP gas < Criticality criteria*1
Proposed application of DCP- Application of fall prevention equipment for heavy weight
objects above ●●- Suppression of reactivity insertion by pre-inserting nonsoluble absorbers

Causeelimination

When double contingency occurs- A heavy weight object of ○○ kg or
more is dropped.
- Small amount of absorber

FP Gas > Criticality
criteria
Emergency boric acidsolution injection

Cause
elimination- Local, such as falling
Undecided cause
elimination method
outside the range ofmonitoring

It is difficult to identify the falling point due to detector performance(detection distance is short). Therefore, DCP prevention is applied.

Lid
Solidifying
agentTransport
cask
Structure
fixed frame

Functional
requirements

Sub-criticality
within control
standard shallbe maintained

Non-soluble
absorber

UO2 / free water falls

Criticality approach
shall be detected
and criticality shallbe prevented bysuspending theworkLong boric acidarrival time
Determined thatcriticality maintains
Additional boric acid solution
injection
Water level drop
Suppression of radioactivegas discharge(Confinement (circulation,etc.))

Criticality shallbe detected
and terminatedquickly

Prevention ofexcessive
exposure ofgeneral public
and workers*1) It is difficult for gas monitors to monitor the proximity of
local criticality due to fall events

Criticality control method for reactivity input due to events
other than fuel debris processing locations

UO2
Free water+
Non-soluble
Illustration of storage inside absorber

the transport cask
©International Research Institute for Nuclear DecommissioningNo.47

7. Implementation details
7.1 Development of on-site operating procedures
7.1.3 Operating procedures for neutron absorbers
d. Interim Summary[Results of this year]

While discussing operating procedures and throughput with the fuel debris retrieval method team (topaccess retrieval/side-access retrieval/integral transfer), the following items were specified and shared.
⚫ Procedures for deciding necessity of use of absorber by criticality approach monitoring⚫ Procedures for returning to Level 1 after using absorber⚫ Size reduction of absorbent supply equipment specifications⚫ Transfer procedures of absorber into core and time required for transfer⚫ Effectiveness of pre-dispersion of absorber as a measure against criticality caused by fall of fuel

These items will be the preconditions to be considered when the fuel debris retrieval method is specified
in the future.
[Future plans]
Based on the above procedures, study of the impacts on fuel debris retrieval throughput will be

expanded, aiming to optimize the operating procedures.

As of the middle of March 2022
©International Research Institute for Nuclear DecommissioningNo.48

7. Implementation details
7.2 Development of solidified absorber technology (water glass)[Issues]Among the non-soluble neutron absorbers developed in the related project(*1), the water glass typeabsorbent (hereafter referred to as water glass material) is a viscous solidified material that covers
and adheres to the fuel debris surface (Figure 1).
➢ If the surface of the fuel debris is covered with water glass material (Figure 2), concerns havebeen pointed out that it will hinder the drying of the fuel debris.

[Implementation details]⚫ With reference to the conditions in the fuel debris drying
process obtained in the related project(*2), the change in thewater content of the test sample covered with water glassmaterial was measured.
➢ For the test sample, a porous ceramic of Al2O3 that had thesame drying behavior as UO2 was adopted.
➢ In this year’s test, the test was conducted under thecondition that the surface in contact with the outside air
was completely coated.
➢ Set the dry state target value to 0.54 vol% moisture
content (equivalent to 0.3 wt% hydrous-bearing zeolite)[Goals]⚫ Acquisition of data on the impacts of fuel debris coated withwater glass material on the drying process

Realistic possible distribution

Figure 1 Insertion of water glass
type neutron absorber into
simulant fuel debris
Porous simulant fueldebris (hydrous state)

Fuel debris (hydrous state)

Solidified neutron
absorberStainless steel
crucible

Solidified neutron
absorber- Attached to some debris- Attached to the entire debris
surface (conservative
assumption)

Figure 2 Estimation of fuel debris state in canister and simulated sample

Major achievements:⚫ Acquisition of quantitative data on the impacts of fuel debris with water glass on the drying process.
⚫ Coordination with the current projects (retrieval work and drying treatment related) enables cooperation in studies for on-site
applicability, which are necessary for understanding the impacts of the retrieval work process and setting drying conditions.
➢ By setting the target moisture content of the test sample to 0.3 wt% (in terms of zeolite), the evaluation is equivalent to thefuel debris drying test conducted in the canister project, and the results of this research can be fed back.
©International Research Institute for Nuclear Decommissioning(*1) Related project “Development of Technology for Further Increasing the Retrieval Scale of Fuel Debris and Reactor Internals” (FY2019/2020)(*2) Related projects “Development of Technology for Fuel Debris Characterization and Treatment (FY2014)”, “Fuel Debris Characterization” (FY2015/2016)

No.49

7. Implementation details
7.2 Development of solidified absorber technology (water glass) a. Test plan
Outline of solidified Absorber- No. 1 sodium silicate (water glass) mixed with Gd2O3 granulated powder, which is a neutron absorbing material(Table 1).
- It is a viscous liquid that solidifies over time.
- It can be applied to uneven surfaces and slopes that are different from flat surfaces. It can be attached to fuel debris.
- Various shapes such as rods and particles are assumed for fuel debris, and neutron absorbers are used accordingly.
Table 1 Components of water glass type neutron absorber
Component

Chemical formulaNo. 1 Sodium
Silicate

Na2O·nSiO2·xH2O

Cement

SiO2, CaO, Al2O3, Fe2O3, CaSO4

Monosodium
phosphate

NaH2PO4·2H2O

Water

H2O

Gadolinium oxide

Gd2O3

Water glass material can be used regardless ofdebris shape
DebrisAbsorber(solid)Figure 1 Example of applicationof solid type absorber topebble-like debris

Figure 4 Example of applying
solid type absorber tobedrock-like debris

Figure 2 Example of applicationof solidified type absorber topebble-like debris

Figure 5 Example of applicationof solidified type absorber tobedrock-like debris

Figure 3 Example of applicationof solidified type absorber torod-shaped debris

Figure 6 Example of application ofsolidified type absorber to debris withlarge unevenness

Figure Examples of proper use of neutron absorbers

Figure An example of covering unevensurface with water glass absorber

It has been pointed out that if the surface of fuel debris iscovered with absorber, it may hinder the drying of the fueldebris.
©International Research Institute for Nuclear Decommissioning49

7. Implementation details

No.50

7.2 Development of solidified absorber technology (water glass) a. Test plan
Impact assessment of solidified neutron absorbers on fuel debris drying characteristics
Outline of implementation
Result of basic characteristics
grasp test
Evaluation of drying behavior of
porous media in the case of fueldebris➢ Establishment of evaluation
method by drying
characteristic curve

Drying rate(III) Falling rate
drying period(II) Constant rate
drying period(I) Preheating
period

Shape

Lumps

Internalstructure(Void ratio)

Large

Bench scale

Material/composition
Internal(Element/chemical
structureform)(Diameter of
air cavities)
Oxide(ceramic)

Suction exhaustion(inner pressure)
Radiation
heating

Flow rate ofwarm air

Large

Heater(Heating
temperature)

Simulated
MCCI product(Cement
paste)
TMI-2
debris, etc.

Powder

Small

Pilot scale

Practical
application

Basic characteristics grasp test

Hot air(Heating
temperature)
Chamber orcollection(Capacity)

Al2O3 powder

Small

Internal
diameterConceptual study of drying equipment

Fundamental test based on proven porous media

Process flow

Both porous media andneutron absorbers
contribute to drying

Standardization of the impactsof water glass materials
i.e., Amount of change indrying rate (time)

Bench scale

Water in porous ceramics

Equilibrium
moisture content
Critical moisture
content

Water glass type neutron absorber

W: Water content ratio →Residual amount of water

The test conditions were determined with reference to test conditions of past projects (*1). Drying tests wereconducted in an experimental system that can be compared with the results of the past projects to understandthe differences in drying behavior. The impacts of water glass material on drying were clarified.
➢ Small-scale tests using proven porous media as a base material were conducted.
(*1) Related projects “Development of Technology for Fuel Debris Characterization and Treatment (FY2014)”, “Fuel Debris Characterization” (FY2015/2016)©International Research Institute for Nuclear Decommissioning50

No.51

7. Implementation details
7.2 Development of solidified absorbent technology (water glass) b. Test preparation

Impact assessment of solidified neutron absorbers on fuel debris drying characteristics Outline of implementation(1) Hydrous state
porous media(Used as simulant fueldebris in past tests)(2) Water glass material(single material)(3) Water glass-coatedporous media
Porous media ina hydrous state

Coating waterglass
underwater

Thermal analysis
Image inside equipment

Drying characteristic curve (image)

Crucible
Thermocouple(Temperature
measurement inside
the heating furnace)
Thermocouple(Temperature
measurement of sample)

Mass balance(Measurement ofweight change ofsample)Porous media

Water glassmaterial

Water glass material +porous media

Appearance observation using a microscope, etc.

Metal
container

Water

In order to obtain test data that can prove that the water contained in the porous media coated with the water glass material evaporates anddries,① Manufacture a test sample in which the route during evaporation of water is restricted to one direction so that the evaporated water fromthe porous media would be discharged after passing through the water glass material② In order to understand the impacts of the presence of water glass material on the drying behavior, obtain data on the drying of three typesof test samples and compare with each other➢ Data required for the drying characteristic curve (drying rate, etc.), change in moisture content of the test sample over time
③ Observe changes in water glass material coating conditions before and after drying treatment©International Research Institute for Nuclear Decommissioning51

7. Implementation details

No.52

7.2 Development of solidified absorber technology (water glass)
Reference: Past knowledge (investigation of drying behavior of fuel debris) (*1)
➢

➢
➢
➢

In the past, in cooperation with the collection & storage project, fundamental research was conducted on the evaluation of thedrying behavior of fuel debris
In the case of fuel debris, many parameters (shape, material, drying temperature, etc.) were assumed to affect the drying
behavior, so the drying characteristic curve, which allows unified expression and comparison was focused on
In order to apply it to fuel debris with various specific gravities that cannot be compared by moisture content, a new evaluation
method was devised based on the relationship between drying rate and residual water, and an optimum diagram method wasstudied
Although the mechanisms leading to the drying of water glass material and porous media are different, by clarifying the degreeof impact on the drying rate, it is possible to infer the impacts on the drying treatment◆ Drying characteristic curve and drying theory
A
Between B-C(Saturated moisture state) (Evaporation of waterfrom the surface)
Water-containing sample (image)

Between C-D(Transition of dry
surface)Drying rate(III) Falling rate
drying period(II) Constant rate drying
period(I) Preheating
periodDry weight: W 1
Wet weight: W 2
Process flow

Between D-E(Evaporation of residual
water)
Equilibrium
moisture content
Critical moisture
content

Funicular
water

Suspended
water

Water content ratio

W: Water content ratio →Residual amount of water

Equilibrium
moisture content
Water content ratio at which
drying does not proceed any
further
Critical moisture
content
Water content ratio whentransitioning from constant rate
drying period to falling rate drying
period©International Research Institute for Nuclear Decommissioning(*1) Related projects “Development of Technology for Fuel Debris Characterization and Treatment (FY2014)”, “Fuel Debris Characterization” (FY2015/2016)

No.53

7. Implementation details
7.2 Development of solidified absorbent technology (water glass) b. Test preparation
Fundamental test conditions and parameters⚫ Test under basic conditions (FY2021)

Test setting based on the fuel debris drying test*1 conducted in the past in the fuel debris property project✓ Simulated material: Al2O3 (Reference conditions for past tests)✓ Drying conditions‐ Temperature: 200°C*2 (reference conditions for past tests, basic conditions for drying equipment)‐ Atmosphere: inert gas (gas flow) (reference conditions for past tests)✓ Hydrous state: Saturated moisture state (reference conditions for past tests)✓ Coating method of water glass: fixed or no coating⇒ The drying behavior for the above test and the water glass material alone were investigated and determine thepossibility effect on the water glass material was the drying treatment⚫ Test with changing parameters (FY2022)✓ Examination of changing the simulated material: Zeolite (currently used for examination of the drying process) +Al2O3 for comparison✓ Coating method of water glass material‐ Change the quantity of coating: Change the thickness of the coating by changing the quantity of coating‐ Change the coating area: Change the coating ratio for the surface✓ In addition, treatment conditions such as drying temperature can be changed according to requirementsfrom other projects and decommissioning operators

Fiscalyear

Simulatedmaterial

2021

Al2O3

Sample size

Φ10×h10 mm
2022

Al2O3, Zeolite

Water glass material

Hydrous state

Drying conditions

No coating

Saturated moisture
state

200°C*2
Ar gas atmosphere

Amount of coating, coating area: fixed
Amount of coating, coating area:
parameters

Examine based on the results of the previous
year*1 Related projects “Development of Technology for Fuel Debris Characterization and Treatment (FY2014)”, “Fuel Debris Characterization” (FY2015/2016)*2 The temperature for the drying treatment when examining the drying process in the collection & storage project©International Research Institute for Nuclear Decommissioning7. Implementation details

No.54

7.2 Development of solidified absorbent technology (water glass) c. Element test
Drying test of water glass material alone
Thickness of water glass material: 1 mm, Underwater curing: 24 hours
Reference: Drying characteristic curveof hydrate-bearing porous media
Thermogravimetricanalysis result

Dryingcharacteristic curveHigh fluctuation

Evaporation of waterfrom surface to center

x140

Moderate fluctuation

Evaporation ofsuspended water

Evaporation ofwater near thesurface

Observation photographafter drying test (side ofwater glass material)• Thermogravimetric analysis results show that the water glass material alone can be dried to a state where no
moisture is discharged from it.
• Visual observation after drying test can confirm the existence of spaces between particles, and it is inferred that the
moisture in the test sample can pass through.
• The drying rate of the water glass material was of the same order as that of the porous media alone.
• The drying characteristic curve of the water glass material showed a drying behavior similar to that of the hydratebearing porous media as a whole, but showed a different trend in the fluctuation of the drying rate.
➢ It was suggested that the bound water of water glass material and the stagnant water (free water) existing
between the particles of water glass material may have different drying rates.
➢ Future studies will also focus on the changes in the crystal structure of the water glass material and proceedwith the investigation.
©International Research Institute for Nuclear Decommissioning7. Implementation details

No.55

7.2 Development of solidified absorbent technology (water glass) c. Element test
Correlation between water glass material thickness and amount of water⚫ The thickness of the water glass material was calculated
from the input amount and density of the water glassmaterial and the crucible diameter(Evaluation of coating thickness before drying test)⚫ The thickness of the water glass material was measured
with a microscope after the drying test
(Evaluation of coating thickness after drying test)
1h underwater curing - water glass material only

24h underwater curing - water glass material only
24h underwater curing - water glass material only 5 mm
48h underwater curing - water glass material only
72h underwater curing - water glass material only

Coating thickness of water glass material [mm]Amount of water in water glass material [mg]Amount of water in water glass material [mg]1h underwater curing - water glass material only

24h underwater curing - water glass material only
24h underwater curing - water glass material only 5 mm
48h underwater curing - water glass material only
72h underwater curing - water glass material only

Coating thickness of water glass material [mm]

⚫ The change in weight determined by thermal analysis was evaluated as the amount of water.
⚫ Both methods clarified a correlation between coating thickness (mean value) and amount of water.
➢ Accuracy needs to be improved by accumulating data, but it is possible to quantitatively evaluate the amount ofwater in water glass materials⚫ In underwater curing for 1 hour, it was judged that it was difficult to accurately evaluate the amount of water, and was
excluded from the quantitative evaluation.
©International Research Institute for Nuclear Decommissioning7. Implementation details

No.56

7.2 Development of solidified absorbent technology (water glass) c. Element test
Drying characteristic curve (Coating thickness 1 mm, Underwater curing 24 hours)
Drying characteristic curve (porous
media only)

Drying characteristic curve (porous
media + water glass material)

②

②

③

①

①

③
*

Reference: Drying characteristic curve(water glass material)③*⚫ Comparison of drying characteristic curves① The sample coated with water glass tend to dry faster during the preheating period.
② The drying rate of the sample coated with the water glass material is slightly faster than that of the porous media
alone.
➢ ① and ② suggest the possibility that the moisture in the water glass material (bound water) and the free waterthat has moved from the porous media into the water glass material evaporate at the same time
③ The sample coated with the water glass material shows a slight decrease in the drying rate near the drying end
point.
➢ It is inferred to be dominated by the effect of moisture generated by the intermittent thermal decomposition of thewater glass material
Overall, the change in drying rate showed behavior similar to the combination of porous media and water glass material.
*Water content ratio: A value obtained by calculating the volume ratio of “the amount of water present in the test sample” to “the volume of the porous media.”
The amount of water in the water glass material is the value obtained by the calibration curve added to the “the amount of water present in the test sample”©International Research Institute for Nuclear Decommissioning7. Implementation details

No.57

7.2 Development of solidified absorbent technology (water glass) c. Element test
Water content ratio change curve over time
Porous media + water glass material
Coating thickness 1 mm, Underwater
curing 24 hours

Porous media + water glass material
Coating thickness 5 mm, Underwater curing
24 hours***

0.54 vol%Porous media + water glass material
Coating thickness 1 mm, Underwater
curing 72 hours

0.54 vol%It was possible to dry the porous media until the water content wasless than the target even for the sample with the coating thickness of
the water glass material of 5 mm.
It was also verified that curing time underwater does not affect drying
behavior within the range of 24 to 72 hours.

0.54 vol%Target moisture content of the canisterproject (drying test of hydrous-bearing
zeolite on actual scale)⇓
0.3 wt% ≒ 0.54 vol% in terms of zeolite
Target moisture content: less than 0.54
vol%*Water content ratio: A value obtained by calculating the volume ratio of “the amount of water present in the test sample” to “the volume of the porous media.”
The amount of water in the water glass material is the value obtained by the calibration curve added to the “the amount of water present in the test
sample”©International Research Institute for Nuclear Decommissioning7. Implementation details
7.2 Development of solidified absorbent technology (water glass) c. Element test
Observation of water glass material-coated porous media after the test (Coating
thickness: 1 mm, Underwater curing 24 hours)
Top surface ×20

side ×30

Top surface ×140

side ×140

No.58

Cracks wereconfirmed on
the topsurface.
Overallshrinkage,
gaps aroundthe perimeter

Verified gaps
betweenparticles⚫ It was observed that the water glass material itself shrank due to the heating and cracks were generated.
➢ Investigation of the behavior changes until dry, such as composition analysis focusing on amorphous materials, isunder consideration.
©International Research Institute for Nuclear Decommissioning7. Implementation details

No.59

In order to specify the impacts of the presence of waterglass material on the drying operation, the drying rate foreach sample prepared in this test was compared.
⚫ The time when the target moisture content (less than
0.54 vol%) was reached was defined as the drying end
time, and the average drying rate per unit time wasobtained.
⚫ The drying rate under each test condition was comparedwith respect to the total volume of water content in the
porous media and the water glass material of the test
samples.
The drying rate (amount of evaporation per unit time) does
not change with or without the water glass material.
➢ It is inferred that when the coating thickness of the waterglass material is in the range of 1 to 5 mm, the boundwater in the water glass material and the moistureretained in the porous media can be treated equally.
➢ The time required to complete drying tends to depend on
the total amount of water derived from each of theporous media and the water glass material.

Average drying rate per minute [mg-H 2 O/min]

7.2 Development of solidified absorbent technology (water glass) c. Element test
Comparison of drying rate for each sample

Porous ceramics
Porous ceramics + water glass material ( coating thickness 1 mm, underwater curing 24 hours)

Porous ceramics + water glass material (coating thickness 5 mm, underwater curing 24 hours)
Porous ceramics + water glass material ( coating thickness 1 mm, underwater curing 48 hours)
Porous ceramics + water glass material (coating thickness 1 mm, underwater curing 72 hours)

Water content in sample [mg]©International Research Institute for Nuclear Decommissioning7. Implementation details

No.60

7.2 Development of solidified absorber technology (water glass)d. Interim Summary
In order to evaluate whether the porous media would reach a dry state even with the water glass material attached, andto evaluate the effect of the drying treatment, a drying test was conducted 結合水using porous Al2O3 as simulant fueldebris.
• The preparation method of test samples and the test procedures were examined, and a basic evaluation method forthe drying of the water glass materials was established.
• Drying characteristics of water glass material alone
It was suggested that heating at a temperature of 200°C would dry to a state where no moisture is discharged.
• Drying characteristics of porous media coated with water glass material
The water glass material had cracks during the drying process, and no events that interfered with the evaporation ofwater in the simulant fuel debris were confirmed.
➢ It was verified that even when covered with a water glass material, the volumetric moisture content of less than 0.54 [vol%] (≒ moisture content by weight of 0.3 wt% (converted to zeolite)), which is the target for achieving the drying
treatment, was reached.
➢ There was no significant change in the drying rate of simulant fuel debris before and after coating with water glassmaterial.
➢ It is possible to treat the bound water in the water glass material in the same way as the water retained in the porous
media.
[Future plans]
Tests in the next fiscal year will be conducted based on parameters such as the material of the porous media and thecoating amount of the water glass material.
Quantitative evaluation on the degree of impact of drying caused by water glass material will be performed by comparing
the obtained results with the test results of this year.
➢ The test results including the result of next fiscal year will be shared with the project of “Development of Technology
for Drying Fuel Debris” , to cooperate in considering how to reflect them in the drying process conditions.

As of the middle of March 2022
©International Research Institute for Nuclear DecommissioningNo.61

8. Specific objectives for achieving the purposes of technological developments(2) On-site operating 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 shall be formulated.
The sub-criticality measurement performance of neutron detectors
shall be evaluated and reflected in on-site operating procedures.
On-site operating procedures for non-soluble neutron absorbers
shall be formulated.
(TRL at the end of the project(*): Level 4)② Development of solidified absorbertechnology

Data on the impacts on the drying process of fuel debris attachedwith solidified absorber (water glass) shall be obtained.
(TRL at the end of the project(*): Level 4)(*) TRL; Technology readiness levels©International Research Institute for Nuclear DecommissioningNo.62End©International Research Institute for Nuclear Decommissioning