Current status of R&D for
Fuel Debris Retrieval

Apr. 25, 2014

International Research Institute for Nuclear Decommissioning(Plant information included in this document is taken from TEPCO official website.)©International Research Institute for Nuclear Decommissioning

1

Overview of Units 1-4
 The state of progress for decommissioning varies with each unit.
 Removing spent fuels from SFP at unit 4 started from November 18.
Spent Fuel Pool
(SFP)

Building cover

Blowout panel(closed)Reactor building (R/B)
Primary
Containment
Vessel(PCV)
ReactorPressure
Vessel(RPV)

Cover for fuel removal

構台

Transferred fuel (assemblies)
福島第一
安全第一
福島第一
福島第一
安全第一
安全
第一

ｸﾛｰﾗｸﾚｰﾝ726/1533（as of 4/23）

Water
injection

Water
injection

Water
injection

Fuel debris

福島第一
安全第一

Vent pipe
Torus room
Suppression chamber(S/C)

Unit 2

Unit 1

Electrical output
Date of commercial
operation

460MW
1971/3784MW
1974/7

Unit 3

Unit 4784MW
1976/3784MW
1978/10©International Research Institute for Nuclear Decommissioning

Outline of Mid and Long Term Roadmap2

 Mid-to-long term roadmap was revised in June 2013.
 Phased approach was confirmed.
 Fuel removal from unit 4 SFP started from November 2013.
December 2011
(Step 2 Achieved)

Efforts to stabilize
plant condition<Cold shutdown
achieved>•Achieve cold
shutdown•Significantly
reduce radiation
releases

1st half of 2020
(fast case)

November 2013

30 to 40 years in
the future

Phase 1

Phase 2Phase 3

Period up to thecommencementof the removal of
the fuel from thespent fuel pool(within 2 years)

Period up to thecommencement of theremoval of the fuel debris(within 10 years)

Period up to thecompletion of
decommissioning
measures (30 to 40
years in the future)

Started first fuel assembly removalfrom Unit 4 SFP on Nov. 18, 2013
“Mid-to-long term roadmap on Decommissioning of Fukushima Daiichi NPS”
was revised on June 27, 2013.
©International Research Institute for Nuclear Decommissioning

Work image for removing the fuel debris■The approach of removing the fuel debris submerged in water is the safest
approach from the standpoint of minimizing exposure of workers.
■The primary containment vessel (PCV) will be examined and repaired forfilling the PCV with water. Furthermore, R&D for the removal and storage offuel debris will be implemented.
■RFI’s were executed to solicit information to overseas countries forcontaminated water countermeasures and innovative approach for fuel debris
retrieval.

removing the fuel debris （image）

Overhead craneRepair of lower part of PCV（water stoppage）～water filling （image）

Spent Fuel
Pool (SFP)

Spent Fuel
Pool (SFP)

Reactor Pressure Vessel

Filling water

Building cover

Reactor Pressure Vessel

PCV

Stopping water
Torus room

PCV

Torus room

PCV

Torus room
©International Research Institute for Nuclear Decommissioning3

Major Challenges in the Existing Decommissioning Procedures
 Final goal is to defuel from the Reactor Building (R/B).
 Defueling procedure would be much more complicated than TMI-2 case due to• differences like:
TMI-2

Fukushima Daiichi

R/B Damage

LimitedDamaged by H2 explosion(Units 1,3,4)Water Boundary

RV remained intact

Both RPV/PCV have been damaged(Units 1-3)

Fuel Debris Location

Remained in RV

Possibly fallen out from RPV

Bottom of the Vessel

No structural components

Complicated structure with Control
Rod Drives

 TMI-2 experience can be utilized more efficiently for post-defueling procedures indecommissioning.
©International Research Institute for Nuclear Decommissioning

4

Fuel removal from spent fuel pool and storage
Removal of spent
Storage of spent
fuel etc.
fuel etc.
(FY2013-)(FY2013-)
Assessment oflong-term integrity (1-1)(-FY2017)

Flowchart of R&D on decommissioning /contaminated water
Mar. 27, 2014
<Legend>
R&D
programs
subsidized
countermeasures for Fukushima Daiichi NPS
by Agency for Natural
Resources and Energy

Reprocessing ofspent fuel etc.

R&D programs founded bysupplementary budget ofAgency for Natural Resources
and Energy started from 2014

Treatment ofdamaged fuel (1-2)(-FY2017)

Conducted by TEPCO(utilizing R&D results)

Securing work environment inside the reactor building (dose reduction etc.)
Develop remotedecontamination system (2-(1) 1a)(-FY2015)(2-(1) 1b)

PCV/RPVstructural integrity
assessment (2-(1) -8)(-FY2014)
Complete reduction of dose
inside the reactor building.
(-FY2019)(-FY2016)

Supply power source,
communication means, water, etc.

Anti-corrosion
measurement(FY2017-)

Input information

The numbers in the brackets, such as(2-(1) -9) indicates project No.

Establish does reduction plan

Introduction of decontamination
system to the site.

(FY2014-)

Secure stable state of the site

Flow of project

Debriscriticality control(2-(1) -9)(-FY2019)

Establish measures
to prevent criticality(FY2019-)<submersion method (fuel debris retrieval underwater)>(1) Full submersion method (if water can be filled up to the upper portion of PCV)
(2) Partial submersion method (if water cannot be filled up to the upper portion of PCV but handling of fuel debris will be carried out underwater.)
Bottom of PCV(vent tube, S/C, torus room etc.)
Development ofinvestigation system(2-(1) -2)(-FY2014)
Development of
repair method(2-(1) -3)

Detection of radioactive
material, such as, in S/C.
(2014-FY2015)(FY2017-)

Upper portion of PCV(hatch, pipe penetration, cooling system etc.)

Development ofinvestigation system(2-(1) -2)(-FY2017)
Development of
repair method(2-(1) -3)

Investigation on
relevant portions(FY2016-)(-FY2016)
Debris characterization
Characterization bysimulated debris (2-(3)-1)(-FY2015)

Repair construction(-FY2018)(-FY2017)<Method applied when submersion method is not applicable(retrieval in the air)>
RFI for technology
information

Detection of debris
in the reactor(2014-FY2016)
Upgrading of accident
progress analysis(2-(2)-1)

Treatment of fuel debris(2-(3)-3)(-FY2019)

FS for innovative
approach(FY2014)

Shielding walls on theland side byground freezing method(-FY2014)
ALPS

Maintenanceand control(FY2015-2020)
Start operation(-FY2014)(FY2014-)

Internal PCV
investigation (2-(1) -4)(-FY2016)
Investigation inside
the reactor(FY2016-)

Removal of fuel debris/reactor internals(FY2019)(FY2020-)(FY2014-2019)

Storing ofgenerated waste(FY2011-)Retrieval of fuel debris(FY2020-)
Debrispackaging/transfer/storage(2-(1) -7)(-FY2019)

Determinationof actual debris(2-(3)-2)(-FY2019)

RFI for technology
information

Storing of fuel debris(2-(1) -6)

Internal RPV
investigation(2-(1) -5)(-FY2019)
Debris sampling(as part of internal RPV
investigation project)(2-(1) -5)

Waste treatment and disposal*Retrieval method will be determined by 2018.

(debris location, amount etc.)

Repair construction(-FY2016)

Fuel debris retrieval

Identify conditions inside the reactor

Investigation on
relevant portions(FY2014-)

FS for contaminatedwater measurement

Waste treatment
and disposal(3)Carrying-out offuel debris(FY2011-)(FY2020-)

Material accountancy(conducted by JAEA withsubsidies granted for
operating expenses)(-FY2019)

Stable storage of debris
and waste materials*S/C: Suppression Chamber

S/C: Suppression Chamber(FY2014)・Seawater purification©International Research Institute for Nuclear Decommissioning
・Capturing radioactive substances from soil.
・Decontamination of contaminated water tank storage
・Unmanned BoringProgress status of
R&D Project in FY 2013
○ Projects related to fuel debris retrieval from PCV/RPV
as well as internal PCV/RPV investigation(2-(1)-2, 3) Development of technology for investigation and repair(water stoppage) for PCV water(2-(1)-4) Development of internal PCV investigation technology(2-(1)-5) Development of internal RPV investigation technology(2-(1)-7) Development of fuel debris packing/transfer/storage technology(2‐(1)‐9) Development of criticality control technology for fuel debris(2-(2)-1) Identify condition in the reactor by upgraded analysis technology for accident progression(2-(3)-1, 3)Development of characterization using simulated debris and debris disposal technology©International Research Institute for Nuclear Decommissioning

6

Technology development for investigation and repair (water stoppage) in preparation of water filling of the PCV

7・Investigation: For bottom section, manufacture device design , manufacturing and plant mock-up test facility and device performance test and actual device applicability
evaluation will be completed. For upper section, device design and manufacturing and performance check will be conducted for each investigation portion.
・Repair : for bottom section, in preparation of device design and manufacturing, repair method and detail verification and element test for water stoppage material will becompleted. As for (water stoppage ) upper section , in preparation of manufacturing of repair device applied for the portion which is highly likely damaged, results of test etc.
will be reflected to the detail verification and design for the water stoppage material.

Contents of implemented measures
1. Development of PCV investigation technology
1.1 Development of PCV bottom section investigation equipment
・ Manufactured investigation equipment for the leak location from PCV
bottom section investigation equipment ・reactor building to the
Floor surface
adjacent building. Manufacturing of plant mock-up test facility
travelling robot
and device performance check will be completed.
Mock-up facility
・Actual device applicability evaluation(field validation)plan and field
validation will be completed.
Under water swimming device
1.2 Development of PCV upper section investigation equipment
Fig.1
S/C
outer
bottomsectionFig.2 Underwater equipment/ Fig. investigation equipment for・As for PCV upper section investigation equipment ,D/W outer opening portion andinvestigationequipment
andplant mock-up test facility
device design and manufacturing and performance check will beplant mock-up test facility.
plant
mock-uptest
facility
conducted for each investigation portion.
(Leak detection device of investigation equipment of Dry-well(D/W)
outer opening section will be of basic small type for small diameter
penetration)
・Establish actual device applicability evaluation(field validation)plan.
Field validation will be planned in 2015.
2. Development of PCV repair (water stoppage ) technology
2.1 Development of PCV bottom section repair device
・In preparation of design and manufacturing of repair device forboundary structure such as by vent piping and suppression chamber,
Fig.6 Water stoppage test
verification of repair method in detail (detailed verification of water
Fig.4 Auxiliary material test condition
Fig.5 test device for water
status for water stoppage
stoppage material such as by water stoppage test and optimization offor closure of water stoppage of PCV
stoppage of PCV upper section
for PCV upper section
closure auxiliary material etc.) will be completed.
bottom section
2.2 Development of PCV upper section repair device
Issues and next plan direction
・ In preparation of manufacturing of repair device applied for the portion Overall scenario of repair including water filling level needs to be reflected to the design
which is highly likely damaged (hatch flange, penetration bellows,of technology development device by promoting the collaboration and linkage with otherelectric penetration), results from water stoppage test will be reflected projects.
to the detailed verification and design.
©International Research Institute for Nuclear Decommissioning

8

Development of internal PCV investigation technology• Investigation equipment for unit 1 was produced and its functioning test completed as for pre-survey of outside of the pedestal (image of PCV, dose, temperatures
etc., obtained). Equipment manufacturing for removing shielding block of Unit 2 and its verification test is planned to be completed in the preliminary investigation
inside the pedestal.
• Basic verification and element test for additional investigation equipment for the accessing point will be completed to be prepared for full scale investigation
in/outside of the pedestal where debris may be existing (distribution state of fuel debris and measurement of shape).
Contents of implemented measures
1. Development of equipment for preliminary investigation of internal PCV:
Sample of equipment development is shown in the figure on the right.
Development of equipment below is ongoing for demonstration test to be
conducted next year.
(1) Investigation equipment inserted from X-100B (Unit 1)
Completed manufacturing of equipment and function verification test.
Improvement items extracted by function investigation will be conducted by
FY2014.
(2) Equipment for removing X-6 shielding block(Unit 2)
Manufactured components (manipulator, end effector etc.) of equipment, andequipment assembly is ongoing. Measures on the handling objective with large
weight found from the results of on-site investigation is under the verification to
be reflected to the development plan.
(3) Investigation equipment inserted from X-6 penetration (Unit 2)
As for the results obtained from the investigation conducted through X-53 in theprevious year and issues, they are to be verified for changes in the equipment
structure of transfer mechanism, and reflected to the development.
Manufacturing of equipment / function verification test is planned to be
conducted by FY2014.
2. Development of access method and equipment(access equipment in/out side the pedestal )Verified concept of access equipment for inside/outside of the pedestal, andestablishment of specification of element making is ongoing. Also, verified isconcept for access equipment required for prevention of dispersion ofradioactive material when sending equipment into PCV.
Element making/test done by FY 2017.
3. Development of inspection equipment and technology(debris measurement apparatus)
Established equipment system structure for technology of measuring shape by
light cutting method. Also, element test for measurement simulating
disturbance environment (spray, rain etc. ) inside the PCV is ongoing.

Location to put
investigationequipment inserted
from X-6 of Unit 2.

X-100B

Location to put
equipment forremoving Unit 2
X-6 shielding
block.

Transformation

X-6
Grating:Access
route

Location to put
investigationequipment inserted
from X100B of Unit 1.

Development
example: investigationequipment inserted
from Unit 1X-100B.

Sample of location to put equipment and its development

Issues and direction of next plan
・Correspond to new issues found in the demonstration test and site investigation
results in the previous year (existence of unexpected obstacles and its large weightetc.) and, address the improvement for verification test.
・In the next plan, conduct the demonstration test, and promote equipment
development .
©International Research Institute for Nuclear Decommissioning

9

Development of internal RPV investigation technology

Established Technology development plan to conduct investigation inside the RPV by verifying the methods of accessing to the investigation location, investigation, andsampling and arranged investigation technology under the high dose environment inside the RPV(provisional value 1,000Gy/h), in order to obtain the location of fueldebris inside the RPV, damaged state of reactor internals, temperature inside the RPV, and dose.
(FY2015/FY2017: Technology to investigate through the system piping, FY2018: Investigation technology for drilling on RPV upper section, FY2019: Technology toinvestigate after opening of reactor).
Contents of implemented measures
1. Planning of internal RPV investigation
 Verification on major investigating item and investigation period
After selecting the investigation items for internal RPV investigation,verified investigation items and investigation period, and establisheddebris plan described in No.2 below (FY 2015/FY2017: Technology toinvestigate through the system piping, FY 2018: Investigation
technology for drilling on RPV upper section, FY 2019:Technology toinvestigate after opening of reactor. ).
 Verification on access route
Selected candidate as a route to investigate inside of the RPV from themethods of accessing : from piping, by creating hole on the upper
portion of RPV upper section, by opening up the reactor and by
evaluating its accessibility. (Fig. 1)
2. Planning of R&D plan
 Access technology
Conducted investigation on the existing technology based on theverification results of access route inside the RPV and extracted issues
on the development of issues such as technology to penetrate into
obstacles.
 Investigation technology(radiation resistant camera , dosimeter etc.)
After investigating existing technology, verified applicability andextracted issues regarding measures on radiation resistance etc.
 Sampling technology
Investigated existing technology, verified concept of sampling method,summarized those issues and established development plan.

Access fordrilling on
the upper
sectionpiping access

Access after
the opening ofreactorSelection of route for accessing thereactor core via PCV manhole-RPV
head spray nozzle-steam dryer baseplate drain pan.
Fig. 1 Verification for the access route inside the RPV (sample of access by drilling on the upper section)
Table 1 Development plan of access technology (sample of access by drilling on the upper section)No

Development technology
element

Issues

2014

2015

2016

2017

Creating hole for the steam dryer, andseparator
Tube expansion for the hole diameter ofsteam dryer and separater

1

Boring technology

2

Tube expansion technology

3

Remote control technology

Monitoring the passing on the curve andnarrow part , and operating condition

4

Boundary forming technology

Boundary reforming on the operation floor(sealed plug)

Issues and direction of next plan
Need to conduct equipment design and element making /test based on the plan fortechnologies for access, investigation, and sampling investigated this year.
©International Research Institute for Nuclear Decommissioning2018

Development of fuel debris packing/transfer/storage technology

10

Conducted investigation on the overseas technologies such as those applied for the accident occurred in the Unit 2 of Three Mile Island (TMI) Nuclear
Power Plant in US. Extracted issues on the development of canister, and established future development plan for the purpose of determining thespecification of canister, such as shape , for packing/transfer/storage of fuel debris.
Contents of implemented measures
1. Investigation on the transfer and storage of damaged fuelConducted investigation on overseas information on damaged fuel(including leak fuel) transportation/storage , and collected
information to be utilized for design of canister for
packing/transferring/storing the fuel debris, including transfer andstorage of fuel debris of TMI-2 in US.
2. Verification on the storage system
Investigated storage system of spent fuel, such as concrete cask etc.,
which we never experienced in Japan, and collected information to
be utilized for the selection of storage system for fuel debris.
3. Extraction of issues and planning of general plan
(1) Collaborative work with other R&D
Collaborating with related project, summarized information requiredfor canister design regarding criticality control and basic physicalproperty of debris.
Also, established plan of the basic process flow up to storing the fueldebris, and extracted issues and required tech development items.
(2) Selection verification for fuel debris storage method
Extracted technological issues and problems when serving as fueldebris storing in canister, and made a comparison for those.
(3) Established general plan
Established future R&D plan based on the investigation andverification above.
*: From DOE/SNF/REP-084 TMI Fuel Characteristics for Disposal Criticality Analysis(2013)

Sample of canister for fuel debris used in TMI-2 in US. (Reference)*

Issues and direction of next plan
Additional overseas investigation and collection of canister concept based on theinvestigation results obtained in FY2013 is required. In FY 2014, reflect those anddevelop safety analysis method required for design of canister.
©International Research Institute for Nuclear Decommissioning

Development of criticality control technology for fuel debris

11

In order to develop criticality control method during the fuel debris retrieval by 2019, as a element technology, completed evaluation of criticality scenario
in each process up to fuel debris retrieval, manufactured and verified criticality detector as prototype, manufactured and narrowed down the candidate
material for insoluble neutron absorber, and summarized soluble absorber issues in FY2013. (re-criticality detector in the reactor is ongoing).
These results will be integrated and criticality control method during PCV water filling and fuel debris retrieval which are major processes will beestablished in FY 2014.
Contents of implemented measures
1. Criticality evaluation･ Completed preparation of criticality scenario for each process from PCV water
filling to fuel debris retrieval, and summarized transitions of state wherere-criticality may be happened (Table 1).
Completed criticality evaluation by representative case including interaction ofdebris with concrete.
･ Completed improvement of behavior evaluation model at criticality by adding
handling of multiple debris with different properties, thermal-hydraulic model forfuel debris that can handle coolant boiling, and model for evaluation of FP nuclide
generation. (to be utilized for planning of criticality control method for next year).
2. Sub-criticality control technology for liquid waste treatment /cooling system･ Manufactured sub-criticality monitor (Fig. 1) experimentally, confirmed
detectability of approaches to criticality by critical assembly, and confirmed
system feasibility. (The development is planned to be done this year.)
3. Technology of re-criticality detection in the reactor･ Verified and designed neutron detector system specification, and procured its
prototype system.
･ Verified improvement of FP γ-ray detector system for gas sampling line andprocured prototype equipment for feasibility check test to detect re-criticality on
early stage.
･ Feasibility check tests above will be conducted. (Apr-May, 2017).
4. Criticality prevention technology･ Manufactured candidate material for insoluble neutron absorber experimentally(Fig. 2), obtained basic physical property data (Table 2) , and completed
narrowing down the candidates on the first phase (After next year, final
candidate will be determined by radiation resistant test and confirmation of
nuclear characteristics, and be applied for debris retrieval )･Completed in summarizing issues when applying soluble neutron absorber, andextraction of required verification items, such as corrosion test. (After verifying
issues in the next year, determine application method of absorber).

Table 1 Scenario for criticality status when retrieving fuel debris

Fig. 1 Display for sub-criticality monitor
Table 2 Performance evaluation check items

Fig.2 Sample of prototype of insoluble neutron absorber(gadolinia/slurry )Issues and direction of next plan
Integrate technologies developed by this year, and promote establishing the method ofdevelopment of criticality control collaborating with method verification such as of fuel debris
retrieval. Also, start developing in-core sub-criticality monitor for the purpose of detection on
the early stage for criticality control reasonably.
©International Research Institute for Nuclear Decommissioning

12

Identifying condition inside the reactor by upgraded technology for accident progression analysis

Completed upgrading of accident progression analysis technology (improvement of core damage progression model and behavior model of debris inside lower plenumetc.) for estimating condition in the reactor of fuel debris location. Utilized the results of upgraded accident progression analysis technology and conducted verification
to identify condition in the reactor. In consideration of latest information obtained from site operation, ratio of debris fallen into PCV was found as follows: amount of
Unit 1 is maximum , Units 2 and 3 are equal and less than Unit 1.
Damage on SRM

SRV leakmeasurement
value

Pressure in RPV (Pa)Fuel rod burst
Damage on the bottom
section of RPV.

Time after scram (h)Fig. 1 Improvement of MAAP model
Simulated debris

Fig.2 Prediction of pressure change inside
the reactor of Unit 1 by SAMPSON.
濃度測定値評価値*
200 NaCl
0%(純水)

壁面熱流束[kW/m2]

Contents of implemented measures
1. Confirmation of validity for code improvement and model change
Re-evaluated and revised the priority rank for the PIRT
(Phenomena Identification and Ranking Table) established in
FY2012 by sensitivity analysis
2. Improvement and upgrading of analysis code
Improved analysis code (MAAP, SAMSON) (Fig. 1) based on
information obtained from site operation, results of current
simulated test, latest findings which is resulted from PIRT and
improved accuracy of analysis.
3. Analysis by improved code(MAAP, SAMPSON (Fig.2))
Conducted analysis of accident progress/condition in the reactorof Unit 1-3 based on the improved latest code and constructeddata base, and confirmed impact of the model improvement.
4. Individual event analysis by CFD
Conducted debris flow analysis as trial using casting simulation
code, and confirmed applicability of debris flow behavior
evaluation by full-scale system (Fig.3).
5. Mock-up test
Conducted simulated test etc. (seawater heat transfer test,
behavior test for fallen molten fuel) contributing to detailed
analysis of progress of severe accident event, and confirmed
applicability of current heat transfer evaluation method when
injecting seawater in the reactor(Fig.4).

150

3.5%
7%
10%
15%
20%

100
*Sieder-Tate 式

500Fig.3 Reproduction of flow cessation test for simulated debris

質量流束 = 550 kg/m2s, 入口温度= 60

0

10
20
30
40
DTw (=壁温－流体温度) [K]

Fig.4 Comparison of experimental value of
seawater heat transfer and evaluation value.

Issues and direction of next plan
In collaboration with other projects ,make output for the Decommission project and input
for this project effectively by enhancing information sharing.
©International Research Institute for Nuclear Decommissioning

50

Study on characterization of debris and development of debris disposal technology using simulated debris 13
Manufactured simulated debris and obtained data such as hardness in order to estimate the property of actual debris to investigate the fuel debris
retrieval technology. Also, in order to verify disposal scenario after retrieving fuel debris, extracted applicability and technology issues on existing fueltreatment technology and compared options to be taken, and clarified those advantage and disadvantage.

Contents of implemented measures
Fe-Cr-Ni
Fe-Cr-Ni(Fe,Cr,Ni)2 B
Study on characterization of debris(2-(3)-1)(1) Verification of physical property required for fuel debris retrieval
・Identified the level of impact to the machinability of physical property suchas hardness, for each type of simulated material.
・Assuming the incorporation of metallic components in the reactor,
measured mechanical characteristics of (U,Zr)O2 in high Zr area and Fe
アー ク溶解酸化雰囲気で 焼鈍(Zr,U)O2 (Zr-rich)contained simulated debris, and reflected the measurement value to the ZrB2（箔状又は (Fe,Cr,Ni)2(Zr,U)（合金及びZrB2 酸化）（Ar-0.1%O
2,1500℃）(Fe 2 Zr型立方晶）
estimation of physical property distribution for each chemical system.
平板状結晶）
制御材（ B4 C+SUS）との反応 （溶融固化物断面観察像の例）(2) Determination of reaction specific to 1F accident(Obtainedknowledge
regardingthe composition of solidified material generated when control rod and fuel is melted・Confirmed the possibility of generating alloy phase and boride by thereaction with control material. Also, confirmed that trend that oxide(vitreous oxides) and alloy layer were separated by the reaction withconcrete (MCCI). The hardest substance was estimated to be boride.
生成した各相のビッカース 硬さ・Confirmed Gd was contained in the some of the fuels, and its impact andarea on the thermal properties of simulated debris((U,Zr)O2) of oxide.
(3) Estimation of actual debris characteristics
・Established debris property list (provisional version) from the results above.
Development of debris disposal technology. (2-(3)-3)(1) Arrangement of technology requirement for fuel debris disposal scenario
verification.
・Compared options for the disposal scenario for the retrieved fuel debris
and clarified advantage and disadvantage.
(Estimate hardness distribution for each chemical system of debris (boride, oxide, metal))
・Evaluated applicability of existing spent fuel transport cask. Found that thewater content of fuel debris etc. which has impact on the storage has highpriority.
Direction of next plan
(2) Verification of element technology for debris analysis
Identify the physical property such as hardness using simulated debris, and evaluate
・ Obtained basic data of melting process, which is a pretreatment
technology for analysis of each simulated debris including MCCI product.
the reactivity with materials in/outside the Pressure Vessel, characteristics evaluationsuch as of MCCI product material. In 2-(3)-3, verify and evaluate the water content of(3) Applicability verification of existing fuel treatment technology.
fuel debris that effects storage technology while continuing the development of・ Obtained basic date for the applicability of simulated debris to the wet
analysis element technology.
process and dry process.
©International Research Institute for Nuclear Decommissioning

14

New R&D project for FY2014
New R&D projects including Fuel debris/reactor
internal structure retrieval technology and Detection
technology for fuel debris in the reactor are planned
focusing on filling water method in FY2014 related tothe current project.
(Reference: Decommissioning, and Contaminated
Water Response Team on March 27th.)©International Research Institute for Nuclear Decommissioning