IRID Annual Symposium 2014

Status of R&D Projects
Related to Fuel Debris Retrieval
July 18, 2014
International Institute for Decommissioning*The contents of this presentation include the results of “Establishment of basic technology for decommissioning and safety of nuclear reactors for power generation in 2013(technological study and research concerning forming an idea for processing and disposing of radioactive waste resulting from the accident)“, a project commissioned by the
Ministry of Economy, Trade and Industry, and the 2014 subsidiary for decommissioning and contaminated water measures (development of technologies for processing anddisposing of waste resulting from the accident).
*Plant information included in this document is taken from TEPCO’s official website.

無断複製・転載禁止 技術研究組合 国際廃炉研究開発機構©International Research Institute for Nuclear Decommissioning

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

Reactor building (R/B)

Blowout panel(closed)Cover for fuel removal

構台

Transferred fuel (assemblies)

ｸﾛｰﾗｸﾚｰﾝ

Primary
Containment
Vessel(PCV)

福島第一
安全第一
福島第一
福島第一
安全第一
安
全
第
一

ReactorPressure
Vessel(RPV)

Water
injection

Water
injection

Removed building rubble
Under removing spent fuelWater
injection

FueldebrisVent pipe

福島第一
安全第一

Torus room
Suppression chamber(S/C)

Unit 1

Electrical output
Date of commercial
operation

Unit 2460MW
1971/3

Unit 4

Unit 3784MW
1974/7784MW
1976/3784MW
1978/10©International Research Institute for Nuclear Decommissioning2

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

1st half of 2020
(fast case)

November 2013

30 to 40 years
in the future

Efforts to stabilize
plant condition

Phase 1

Phase 2Phase 3

<Cold shutdown
achieved>

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

Period up to thecommencement of theretrieval of the fueldebris (within 10 years)

Period up to thecompletion of
decommissioning
measures (30 to 40
years in the future)•Achieved cold
shutdown•Significantly
reduced radiation
releases

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 Decommissioning3Concept image of work steps for fuel debris retrieval■The approach of retrieving the fuel debris submerged in water is the safest
approach from the standpoint of minimizing exposure of workers.
■Technologies for investigations and repairing methods for filling the PCV
with water has been conducting. Furthermore R&D for the retrieval, packing
and storage of fuel debris will be implemented.
■Request for information(RFI) on alternative methods for the fuel debris
retrieval was conducted.
Retrieval of fuel debris
Repair of lower part of PCV
Water stoppage →Water filling

Overhead crane

Building cover
Fuel debris can container
RPV lid

Spent Fuel
Pool (SFP)

Spent Fuel
Pool (SFP)

Reactor RPV

Filling water

Retrieval

Reactor RPV

PCVStopping water
Torus room

PCV

Torus room

PCV

Torus room
©International Research Institute for Nuclear Decommissioning4

Major Challenges in the Existing Decommissioning Procedures
 Final goal is to remove all fuels from Reactor Building (R/B).
 Removing 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 are damaged(Units 1-3)

Fuel Debris Location

Remained in RV

Possibly fallen out from RPV

Bottom of the Vessel

No structural components

Complicated structure including
Control Rod Drives

 TMI-2 experience can be utilized more effectively for procedures after
removal of fuels in decommissioning.
©International Research Institute for Nuclear Decommissioning

5

Fuel removal from spent fuel pool and storage
Removal of
spent fuel etc.
(FY2013-)

Storage of spent
Reprocessing offuel etc.
spent fuel etc.
(FY2013-)
Assessment ofTreatment oflong-term integrity
damaged fuel etc. (1-2)(-FY2017) (1-1)(-FY2017)

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

Securing work environment inside the reactor building (dose reduction etc.)
Establish does reduction
Develop remoteplan
decontamination system(2-(1) 1b)

Secure stable state of the site
PCV/RPVstructural integrity
assessment(2-(1) 1a)(-FY2015)(-FY2014)
Complete reduction of dose
inside the reactor building.
(-FY2019)
Supply power source,
communication means, water, etc.

Introduction of decontamination
system to the site.

(FY2014-)

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

Anti-corrosion
measurement(FY2017-)(2-(1) -8)(-FY2016)

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

Establishmeasures toprevent
criticality(FY2019-)

Conducted by TEPCO(utilizing R&D results)

Flow ofproject

Input
information

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

<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 of
Investigation on
investigation
relevant portions
system (2-(1) -2)(FY2014-)(-FY2014)
Development of
repair method(2-(1) -3)(-FY2016)

Repairconstruction(FY2017-)

Upper portion of PCV(hatch, pipe penetration, cooling system etc.)
Development of
Investigation on
investigation
relevant portions
system (2-(1) -2)(FY2016-)(-FY2017)

Fuel debris retrieval

Identify conditions inside the reactor(debris location, amount etc.)

Internal PCV
Detection of
investigation (2-(1) -4)
radioactive material,(-FY2016)
such as, in S/C.
Investigation
inside(2014-FY2015)
the reactor
Detection of debris(FY2016-)
in the reactor (2-(2)-2)
Internal RPV(2014-FY2016)
investigation(2-(1) -5)
Upgrading of accident(-FY2019)
progress analysis*Retrieval method will be determined by 2018.

(-FY2016)
Debris sampling
Debris
Debris characterization (as part of internal
RPV investigation
packaging/
RepairDevelopment of
Characterization by
project) (2-(1) -5) transfer/storage
repair method
construction
simulated debris(2-(1) -3)(2-(1) -7)(FY2019)(2-(3)-1)(-FY2018)(-FY2015)(-FY2019)(-FY2017)
Determination<Method applied when submersion method is not applicable Treatment of fueldebris (2-(3)-3)
of actual debris(retrieval in the air)>

RFI fortechnology
information

FS for innovative
approach(FY2014)

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

Maintenanceand control(FY2015-2020)
Start operation(-FY2014)(-FY2019)(-FY2019)(FY2014-)

RFI fortechnology
information

FS for contaminatedwater measurement

Removal of fuel debris/reactor internals (2-(1) -6)(FY2014-2019)

Retrieval of fueldebris(FY2020-)
Carrying-out offuel debris(FY2020-)
Material accountancy(conducted by JAEA withsubsidies granted for
operating expenses)(-FY2019)

Waste treatment and disposal
Storing of fuel debris(FY2020-)
Storing ofgenerated waste(FY2011-)
Waste treatment
and disposal(3)(FY2011-)

Stable storage ofdebris and wastematerials*S/C: Suppression Chamber

S/C: Suppression Chamber(FY2014)・Seawater purification©International
・Capturing radioactive substancesfrom soil.Research Institute for Nuclear Decommissioning
6
・Decontamination of contaminated water tank storage
・Unmanned BoringResearch & development projects related to removal of fuel debris
1. Research and development related to removal of fuel from spent fuel pool
(1-1) Assessment of long-term structural integrity of fuel assemblies removed from spent fuel pool
(1-2) Study of methods to process damaged fuel removed from spent fuel pool
2. Research and development related to preparations for retrieving fuel debris(2-(1)-1) Development of technology for remote decontamination of inside of reactor building(2-(1)-2 & 3) Development of technologies for investigation and repair (stopping water) toward filling
PCV with water(2-(1)-4) Development of technologies for investigating inside of PCV(2-(1)-5) Development of technologies for investigating inside of RPV(2-(1)-6) Development of technologies for retrieval of fuel debris/core internals(2-(1)-7) Development of technologies for fuel debris containing/transfer/storage(2-(1)-8) Development of technologies for assessing structural integrity of RPV/PCV(2-(1)-9) Development of technologies for criticality management of fuel debris(2-(2)-1) Analysis of core status by advanced technologies for analyzing progression of accident(2-(2)-2) Development of technologies for detecting fuel debris in reactor(2-(3)-1 & 3)Fuel debris characterization by using simulated debris and development of fuel debris
processing technologies©International Research Institute for Nuclear Decommissioning71. Research and Development related to removal of fuelfrom spent fuel pool
Assessment of long-term structural integrity of fuelassemblies removed from spent fuel pool (1-1)©International Research Institute for Nuclear Decommissioning8

(1-1) Assessment of long term structural integrity of fuel assemblies removed from spent fuel pool
○ Objective of research
It was confirmed by investigating new fuels in the spent fuel pool (SFP) and simulation of water
quality conducted prior to removing fuel from Unit 4 that the peculiar environment (seawater
injection and fall of rubble) did not influence the fuel removal work. The objective of this researchand development is to study methods to assess and store the structural integrity of fuel assemblies
over a long time, taking into consideration the peculiarity such as seawater injection and mixing of
rubble.
○ Problems for assessment of long-term structural integrity of fuel assemblies at 1F site
The following peculiarity should be considered when assessing the structural integrity of the fuelassemblies at 1F:• Changes in quality of water in SFP due to injection of seawater• Mixing seawater components such as chloride ion and increase in electric conductivity• Mixing of rubble into SFP• pH increase (alkalization) , damage of fuel assemblies (such as scratch and deformation)

Long-term structural integrity of 1F fuel will be assessed by gathering demonstrative data throughexperiments using actual fuels, specifying conditions along with existing data.
In addition, irradiation effects and acceleration test methods will be studied through the basic test.
©International Research Institute for Nuclear Decommissioning9

(1-1) Assessment of long term structural integrity of fuel assemblies removed from spent fuel pool

○Adaptabiity of store of fuel assembly in the interim storage / treatment facility forlong term after transferred from the spent fuel pool to the common pool.
Structural integrity ⇒ Load transfer paths meet the required structural strength
Fuel cladding scalability ⇒ Effect evaluation caused by fission products(FP)leaking from fuel cladding（For example, below the criteria of dry cask storage.)
Upper tie-plate

Structural integrity•Handle•Upper Tie-Plate•Fastening of Tie Rod
Bolt

Channel fastener
Channel box
External spring
Uranium dioxide pellet

Sealability of fuel cladding•Fuel cladding

Water rod
Spacer

Structural integrity•Lower Tie-Plate

Lower tie-plate©International
Institute for Nuclear Decommissioning
BWR fuelResearchassembly10

(1-1) Assessment of long-term structural integrity of fuel assemblies removed from spent fuel pool (2013 result - 1)
Objective
Since there were concerns that seawater and concrete components dissolved from rubble affect the quality of water in the channel box
locally, tests to assess the influences on the water quality of such components collected from Unit 4 when new fuels had been checked wereconducted.
Immersion test using rubble mixing with fuel in SFP of Unit 4
 Test conditions
Shape of rubble: Stone or sand shape
Quantity of solution: 100 mL
Immersion temperature: 60C
Immersion time: 1,000 hr (solution to be sampled
after 100, 300, and 500 hr)
Atmosphere: Released to atmosphere
Measuring method: Ion Chromatographs, inductively
coupled plasma mass spectrometry (ICP-MS)

HeaterStirrer
Block

Outline of
immersion test Test result Test result

Total Cl- flowing out per equipment
mass (mg)

Total quantity flowing out per
equipment mass (mg)0.8

rubblematerials

 Test conditions
Simulated rubble: rubble of commercially available concrete immersed in simulated seawater
solution of a chloride ion concentration of about 6,000 ppm and a temperature of 90C
Shape of rubble: Stone or sand shape
Quantity of solution: 100 mL
Immersion time: 1,000 hr (solution to be sampled after 100, 300, and 500 hr)
Immersion temperature: 40, 60, and 80C
Measuring method and atmosphere are equivalent to those of the SFP fuel rubble immersion
test.

8

Sand rubble ●: Cl-, ▲: Ca2+, ■: Na+
Stone rubble ○: Cl-, △: Ca2+, □: Na+

0.9

Immersion test with simulated rubble

0.7
0.6
0.5
0.4
0.3
0.2
0.1
0

7

Sand rubble●: 80oC●: 60oC●: 40oC
Stone rubble○: 80oC○: 60oC○: 40oC

6
5
4
3
2
1
0
0

0

200

400
600
Time (h)

800

1000

Total quantity of Cl-, Ca2+, and Na+ dissolved from rubble at 1F
 Seawater components Cl- and Na+, and concrete component Ca2+ were detected.
 Chloride ion concentration tended to saturate with immersion time.

200

400

600

800

1000

Time (h)Total Cl- concentration in solution in which simulated rubble is immersed
 Total chloride ion concentration tended to saturated with immersion time.
 No dependency on temperature of the quantity of chloride ion was observed.
©International Research Institute for Nuclear Decommissioning11

(1-1) Assessment of long-term structural integrity of fuel assemblies removed from spent fuel pool (2013 result - 2)
Objective
Data for comparison were collected for future research by using spent fuel rods stored in a common pool and measuring theiroxide film thickness and inspecting their appearance. Fuel rods were selected taking into account the types and the use history of
the fuel rods, and production conditions of the cladding.

Oxide film thickness measurement

Selection of fuel rod to be checked
type(GWd/t)

F2RN1

NFI

STEPⅡ

F6N94

GNF-J

F6M40
F5D38
F5C23

Cy

ation

45.1

4

1586

High burn-up

STEPⅡ

46.3

5

1688

High burn-up

GNF-J

STEPⅡ

44.9

6

2118

About 2,000Irradiation days

GNF-J

8x8

27.8

4

1339

GNF-J

8x8

27.8

Remarks

days

5

1659

Film thickness (μm)

cturerFuel tested (8 x 8 fuel, 1F3)
Shimane fuel (8 x 8 fuel, S1)
High-performance fuel (8 x 8 fuel,
1F3)
Test data (8 x 8 fuel, 1F3)

8 x 8 zirconium liner fuel (8 x 8 fuel, 1F3)
High burn-up fuel for verification (high burn-up 8 x 8 fuel,2F2)
9 x 9 fuel (type A) (9 x 9 fuel, 2F1)
9 x 9 fuel (type B) (9 x 9 fuel, 2F1)
Test data (high burn-up 8 x 8 fuel, 1F2/1F6)

High burn-up
Equivalent burn-up range but

many irradiation days

Appearance offuel assembly

Average fuel element burn-up (GWd/t)

Average fuel element burn-up (GWd/t)

Dependency on number of Irradiation days

Fuel tested (8 x 8 fuel, 1F3)
Shimane fuel (8 x 8 fuel, S1)
High-performance fuel (8 x 8 fuel,
1F3)
Test data (8 x 8 fuel, 1F3)

Number of irradiation days

8 x 8 zirconium liner fuel (8 x 8 fuel, 1F3)
High burn-up fuel for verification (high burn-up 8 x 8 fuel,Maximum oxide film thickness (μm)

Burn-up

Film thickness (μm)

Fuel

IrradiMaximum oxide film thickness (μm)

ManufaFuel No.

Dependency on burnup

2F2)
9 x 9 fuel (type A) (9 x 9 fuel, 2F1)
9 x 9 fuel (type B) (9 x 9 fuel, 2F1)
Test data (high burn-up 8 x 8 fuel, 1F2/1F6)

Number of irradiation days

Japan Nuclear Energy Safety Organization, “2005 Report on High Burn-up 9 x 9 Type Fuel Reliability
Demonstration Result”, July 2006
Japan Nuclear Energy Safety Organization, “2006 Report on High Burn-up 9 x 9 Type Fuel Reliability
Demonstration Result”, December 2007

・The fuel rods appeared to be covered with uniform brown clad and no special corrosive behavior was observed.
Nor was local collection of rust suggesting crevice corrosion due to increase in chloride ion observed.
・Result of measuring the thickness of the oxide film was compared with existing data but the measurement©International
Research
Instituteforconfirmed.
Nuclear Decommissioning
result was within the range of the existing data and no special
corrosive
behaviorwas12(1-1) Assessment of long-term structural integrity of fuel assemblies removed from spent fuel pool (2013 result - 3)
Objective(1) To assess influences of the environment history of the pool at the initial stage of the accident on the mechanical strength of cladding by conducting a ring type tensile strength testof cladding immersed in seawater at 80C(2) To check influences of oxide film of pitting corrosion on used cladding by measuring pitting corrosion potential in artificial seawater(3) To assess influences of radiation field (highly oxidized environment) on pitting corrosion of unused cladding by measuring pitting corrosion potential with gamma ray irradiation

Tensile strength (MPa)±2σUnusedImmersion time (h)UsedUnused(3) Influence of gamma ray irradiation (water qualitychange) on pitting corrosion potential of unused cladding in
diluted artificial seawater

Pitting corrosion potential (V vs. AgAgCl electrode)

Used(2) Pitting corrosion potential of unused and used
cladding in diluted artificial sea water

Pitting corrosion potential (V vs. AgAgCl electrode)(1) Tensile strength of used cladding immersed inseawater at 80℃ for 1,000 hours

Room temperature
Dose rate

Chloride ion concentration (ppm)
Chloride ion concentration (ppm)

Test result (outline)(1) History of immersion in high-temperature seawater for a short time
does not affect the strength characteristics(2) Existence of oxide film suppresses generation of pitting corrosion of
cladding.
(3) Pitting corrosion potential of cladding does not drop in radiation field.

Summary
It is considered that possibility of pitting corrosion on fuelcladding is low in the spent fuel pool environment and cladding
not damaged keeps integrity from the viewpoint of corrosion.
©International
Institutefor Nuclear
Decommissioning*Conductedby JAE with Researchsubsidiaryfrom theeducation
ministry13

(1-1) Assessment of long-term structural integrity of fuel assemblies removed from spent fuel pool (2014 plan)
Main goal of 2014
Reflecting the results of the projects in 2013, a plan for post-radiation tests using the materials of fuel removed from SFP will be made. Corrosion test andstrength test will be conducted on simulated fuel materials not irradiated to select Items to assess influences of rubble on corrosion and judge long-term
structural integrity of the materials when they are stored in a wet environment (water quality data, etc.). These items for assessment will also be checked forvalidity by assessing with underwater camera the fuel assemblies transferred from Unit 4 to the common pool, and the scale of influence of each item will be
confirmed through test. In addition, seawater component transfer assessment and corrosion test using spent fuel cladding will also be conducted as basic tests
related to long-term structural integrity assessment.
Objectives in 2014
1. Development of technologies for assessing long-term structural integrity of fuelassemblies(1) Study of test conditions for long-term structural integrity assessment• A plan of a test using the materials of fuel taken out of SFP will be made. The
results of water quality influence assessment and corrosion tests conducted in2013 will be reflected. In addition, a plan for transporting fuel materials will be
studied and incorporated into the above plan.
(2) Development of technologies for assessing long-term structural integrity of fuelstructures• Corrosion and strength tests using non-irradiated specimen simulating thestructure of fuel will be conducted to assess influences of rubble to be brought
into the common pool on corrosion and influences of damage to cladding on
corrosion.
• Assessment of 1F-4 spent fuel stored in the common pool will be evaluated andthe thickness of oxide film will be measured to assess corrosion of the fuel.
• Tests will be conducted to assess influences of damages from falling rubble andwater content, on the assumption that spent fuel taken out of the SFP will be
stored in a dry environment.
2. Basic tests related to long-term structural integrity assessment(1) A chloride ion transfer test using simulated cladding will be conducted to assess
absorption of seawater component at the surface cladding of spent fuel.
(2) Tests to check corrosion of the cladding of spent fuel will be conducted with solution
containing seawater and rubble components and with gamma ray irradiated, toassess influences on corrosion of local changes in water quality.

Schedule (2014)
実施工程（平成２６年度）
2014(first half)2014(前)2014(second half)2014(後)Planning test試験計画及び試験条件策定
and specifying test conditions
Non-irradiated material corrosion test and strength test in
simulated common pool environment
共用プール模擬環境下での未照射材腐食試験、強度試験

Examination of removed
fuel assemblies in common pool
共用プールでの取出し燃料集合体調査
Dry storage assessment test

乾式貯蔵評価試験

Seawater海水成分クラッド移行試験
component cladding transfer test

Corrosion
test with gamma ray irradiation
ガンマ線照射下腐食試験©International Research Institute for Nuclear Decommissioning

14

1. Research and Development related toremoval of fuel from spent fuel poolStudy of methods to process damaged fuelremoved from spent fuel pool (1-2)

15©International Research Institute for Nuclear Decommissioning(1-2) Study of method to process damaged fuel removed from spent fuel pool (2013 result – 1)
○ Objective of research
It is planned to store spent fuels removed from the spent fuel pool in the common pool of the power plant for thetime being, but it is necessary to decide a policy toward the future processing and disposal of the fuel. Therefore,technical problems over reprocessing and solutions to them will be organized by examining cases where damaged
fuels, were handled, both domestically and internationally. In addition, information necessary for organizing indexes
to decide on implementing processing and disposal and problems will also be organized.
○ Actions(1) Studying cases of handling of damaged fuels both domestically and internationally
Domestic cases where damaged fuel was handled will be referred to from materials and documents made public, such as reports onlaws and regulations, in order to identify problems of transporting and storing damaged fuels in a reprocessing facility and find solutions
to them. In addition, overseas cases will also be studied by checking documents on damaged fuels from international organizations suchas IAEA and the database of the International Nuclear Information System (INIS).
(2) Study of requirements and criteria for handling damaged fuel, etc., in other countries
Other countries’ documents on check items for identifying fuel damage status, criteria, fuel investigation method, etc., will be studied.
(3) Study of methods and cases of handling damaged fuel, etc., in reprocessing facility
Descriptions on handling spent fuels under the current license of a domestic reprocessing facility will be organized.
(4) Organizing decision criteria and technical problems toward reprocessing
Based on the above results, information necessary for preparing indexes to make decision on whether reprocessing can be
implemented, problems, and technical issues related to handling damaged fuels will be removed and countermeasures will also be
investigated.

©International Research Institute for Nuclear Decommissioning

16

(1-2) Study of method to process damaged fuels removed from spent fuel pool (2013 result – 2)

○ Decision indexes and technical problems toward reprocessing
 Major factors making it difficult to handle damaged fuels at a reprocessing facility
・Leakage of radioactive substances: Contamination of pool water・Deterioration of mechanical strength: Influences on removing and handling channel box
・Deformation:
Interference with channel box removal and equipment
・Accompanying impurities:
Influences on chemical processing, etc.
 Possible countermeasures (Depends on condition and degree of damage.)・Can container (sealed (exhaust/drain functions)/unsealed)・Repair and reinforcement
・ReassemblyMajor technical problems
・Influences on handling (including influences on channel box, can container, repair, andreinforcement)
・Influences on chemical processing, etc. (corrosion, products, waste, process operation)
 Possible decision indexes (permissible range at reprocessing facility)
・Ratio of leakage of radioactive substances
・Degree of deformation
・Quantity of accompanying impurities
・Structure, dimensions, etc., of can container©International Research Institute for Nuclear Decommissioning17

(1-2) Study of method to process damaged fuel removed from spent fuel pool (2014 plan)
Objectives in 2014
・To conduct corrosive tests on cans containing liquid waste at high concentration and liquid waste tanks to assess the concentration of impurities permissible for both thecans and tanks
・To analyze influences of impurities on U and Pu products and anion on U and Pu removal through removal test under FP and impurity coexisting conditions
・To complete analysis of influences of impurities on basic glass property values (such as glass transfer temperature) by producing glass specimen that takes impurities
into consideration
・To comprehensively extract and sort out the influences on processing damaged fuel, etc., at a reprocessing facility
Schedule (2014)

Actions in 2014
Study of influences of damaged fuels on chemical processing(1) Assessment of influences of impurities on corrosion of reprocessing equipment s
Corrosion tests (immersion test and electrochemical corrosion test) using
simulated solution that takes FP components into consideration will be
conducted on high-level liquid waste tanks, the representative equipment that
handles high-level liquid waste, to assess influences of impurities on corrosion.
(2) Assessment of in-process behavior of impurities
Impurities will be removed under condition where FP coexists to confirm
transfer of the impurities to the U and Pu products. U and Pu will be removed
under condition where anion coexists, to confirm the influences of the
impurities on U and Pu removal.
(3) Assessment of influences of impurities on liquid waste
To assess influences of impurities on vitrified objects, glass specimen will becreated by using powder materials based on the composition of high-level liquid
waste to gather glass property values such as density, glass transfer
temperature, and expansion coefficient.
(4) removal and outlining other influences
Influences expected at reprocessing facility will be comprehensively removed
and necessary research elements will be decided.

2014 (first half)2014 (second half)

Preparation

High-level liquid waste storage tank
Immersion/electrochemical test

Preparation

High-level liquid waste concentration can
immersion/electrochemical test

Summary

Preparation FP coexisting condition
Preparation Anion coexisting condition
Summary

Preparation Glass specimen creation

Gathering glass property
valuesremoving influence common andpeculiar to facility

Summary

Putting in order and summary©International Research Institute for Nuclear Decommissioning18

2. Research and development related to preparations for
retrieving fuel debris
Development of technology for remote decontamination of
inside of reactor building (2-(1)-1)©International Research Institute for Nuclear Decommissioning19

(2-(1)-1) Development of technology for remote decontamination of inside ofreactor building (2013 result – 1)
Main goals of 2013
(1) To confirm contamination situations on the upper floors (second and upper floors of the reactor building excluding the floors damaged by explosion) and at thehigh places on each floor (dose rate, radiation source, and contamination distribution).
(2) To share and study the specifications of remote equipment for upper floors for common use, and design and produce remote decontamination equipment for
the high places on each floor(3) To create shield necessary for the hot spots on the first floor of the reactor building and verify to confirm if it can be remotely installed
Actions taken in 2013

1. Gathering basic data on contamination conditions
Dose rate was measured and contamination distribution, surface contamination, contained radiation source, and penetration of contamination werechecked mainly on the upper floors and at high places of each floor of the reactor buildings of Units 1 to 3. To check penetration of contamination,
penetrated contamination (concrete core) samples collected were analyzed at the site to assess the amount of radioactivity. Some samples weretransported to JAEA for detailed analysis of penetration contamination.

Unit

Floor/area

Dose rate(dose rate
meter)

Contamination
distribution(γ camera)

1

First floor/both sides
First floor/high place
Second floor/whole area -1
Second floor/whole area -1
First floor/high place
Second floor/whole area -1
Second floor/whole area -1
Fifth floor (operation floor)/whole area -1
First floor/high place
Second floor/whole area -1

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○
○
○
○
○
○
○
○
○

Item to be checked
Surfacecontamination (β-ray
dose rate meter oraccumulative dose
rate meter)
○ -2
○ -2
○
-

○

○

-

2

3

Contained radiation
source(accumulative dose
rate meter)○ -2-

Penetration ofcontamination(core analysis)

Remarks○○
Second floor cannot be accessed
because rubble piles up on the stairs.

*1: Investigation into small rooms was not included.
*2: Surface contamination of core sample was investigated. *3: To be implemented at the northwest corner *4: Sample sent to JAEA for analysis

2. Adjusting decontamination technologies and study of concept of decontamination
Based on the situation of contamination investigated in 2012, decontamination technologies suitable for decontamination of the upper floors were selected (byreviewing the result of narrowing down on decontamination technologies, which was conducted in 2012), and a basic policy for decontaminating the upper floors
and the high places on each floor was considered.
3. Design and production of remote decontamination equipment and verification of remote decontamination
Specifications for common use of a remote decontamination equipment that would be applied to the upper floors were considered and designed. In addition, aremote decontamination equipment that would be applied to decontamination of the high places on each floor was designed and manufactured. The systemdemonstrated in 2012 was remodeled to verify its application to the real plant.
©International Research Institute for Nuclear Decommissioning
4. Verification of shielded installation of real plant
20
Necessary shields were created for the hot spots on the first floor of the reactor building and verification was conducted to confirm that the shields could beremotely installed.

(2-(1)-1) Development of technology for remote decontamination of inside ofreactor building (2013 result – 2)
[Schedule for collecting basic data]

Unit 2, 5th floor
Unit 1, 2nd & 3rd
floorsUnit 2, 2nd & 3rd
floorsUnit 3, 2nd floor
Low
Units 1 to place3, 1st floor High
place

○

Contamination
distribution
check
○

○

○

-

-

Decontamination system for upper floors

○

○

-

-

Decontamination system for upper floors

○

○

-

-

Decontamination system for upper floors

●

●

●

●

Decontamination system for low places

○

○

○

-

Decontamination system for high places

Unit 1, south

○

○

○

○

Decontamination systems for low and highplacesBasement

-

-

-

-

Place checked

Dose rate
check

Surfacecontamination
check○

Penetrated
contamination
check○

Decontamination system for upper floors

Applicable decontamination system●: Investigated in 2012○: Investigated in 2013
[Plan to apply result obtained by collecting basic data]
Development of decontamination system for low places
Dose rate data
Contamination
distribution
Surfacecontamination
data
Estimatingroute oftransfer ofcontamination

Surfacecontamination data
Penetrated
contamination data

Dose rate data
Contamination
distribution
Surfacecontamination data
Penetrated
contamination data

Simulation ofcontamination
Decontamination
test of simulated
contamination

Study ofapplication of
decontamination
technology

Development of remote decontamination system

Development of decontamination system for high places
and upper floors
Dose rate data
Contamination
distribution
Surfacecontamination data

Estimatingroute oftransfer ofcontamination

Surfacecontamination data
Penetrated
contamination data

Dose rate data
Contamination
distribution
Surfacecontamination data
Penetrated
contamination data

Planning for integrated reduction of dose
Dose rate data
Contamination
distribution
Surfacecontamination data

Assessmentof radiation
sourcedistribution

Simulation ofcontamination
Decontamination
test of simulated
contamination: Basic data of 2012 will be used.
：H24年度基礎データを活用

Study ofapplication of
decontamination
technology

Development of remote decontamination system

Dose rate data
Contamination
distribution

Assessment ofeffect ofmeasures tolower dose ineach area(decontamination, shielding,removal)Considerationof method toinstall shield©International Research Institute for Nuclear Decommissioning
Planning for integrated reduction of dose21

Development of technology for remotely decontaminating floor on the wall (lower
portion) and the first floor of reactor building – 1[Improvement of dry ice blast decontamination equipment](1) Improvement of decontamination time
To extend decontamination time by increasing dry ice blocks to be filled
Dry ice block compartment(2) Improvement to safely increase moving speed (to enhance
efficiency in operating camera and employ optical LAN)
◆ New around-view camera is mounted so that interference from all
the directions of a moving cart can be viewed on a single screen (or on
multiple screens by increasing the number of surveillance cameras).

Image from bird’s-eye-view camera
Tripling blast time
Blast time was 30 minutes before, which wasincreased to 1.5 hour after remodeling.

Direction ofbeam fromlight source

Line of sight ofsurveillancecameraDecontamination
unit

Direction ofbeam fromlight source

Line of sight ofsurveillancecameraLAN distributor
Size andinstallation
position pending

Bird’s-eye-view camera (leftside)pending

The number of cameras increased from fourto nine per cart.
©International Research Institute for Nuclear Decommissioning

22

Development of technology for remotely decontaminating floor and wall (lower portion) on first floor ofreactor building – 2[Improvement of blast/collecting equipment]
Improvement of operability♦ Improvement of screen (introduction of around-view monitor)♦ Increasing size of monitor♦ Improvement of software (reuse of arm teaching data)

Improvement of routing of hoses and cables

Production of collecting
decontamination head
Wide head for decontamination witha rotary brush was produced.

Improvement of visibility
Placement of cameras and lighting instruments wasreviewed.

♦ Adjustments were made to decrease the distance
between casters of the cart so that hoses andcables do not interfere with corners of pillars andothers.
♦ The specifications of the first cart on casters werechanged to improve the follow-up performance.

©International Research Institute for Nuclear Decommissioning

23

Development of technology for remotely decontaminating ceiling and wall (upper portion) on
first floor and roof of reactor building – 3[Result of investigation at south side of Unit 1 (1)]investigation with gamma camera into contamination condition
Date: December 22 to 24, 2013 (3 days)
Area: South area of first floor of reactor building of Unit 1
Operations: (1) Dose rate measurement (1,500 mm and 50 mmabove floor)(2) Shooting with gamma camera

Table: Result of dose rate measurement

Measurement points●: Only dose rate was measured.
X: Measurement with gamma camera(dose rate also measured9←: Measurement with gamma camera(Angle of elevation changed only formeasurement in the direction
indicated by the arrow.)▲: Measurement with gamma camera(Measurement focused on thesouthern direction only at elevation
angle of 0 (or 10 degrees in somecases).

Measurement
point

Dose rate measurement (mSv/h)Height: 1,500 mm

Height: 50 mm

LegendBlue: Measured on December 22
Orange: Measured on December 23
Red: Measured on December 24

North

MS tunnel
room

West

East
TIP room

Entrance forlarge objects

South
Figure: Position of investigation of contamination conditions at the south
side of Research
first floor©International
Instituteof
for Unit
Nuclear 1
Decommissioning24

Development of technology for remotely decontaminating ceiling and wall (upper portion)
on first floor and roof of reactor building – 4[Result of investigation at south side of Unit 1 (2)]Result of investigation by gamma camera at south side of first floor of Unit 1 (assessment of outline of inert gas
pipingPiping in the southeast area was confirmed by the photographs shot by the gamma camera to be highly contaminated.
(Vapor went through the piping because the PCV was vented at the time of accident.)
It was estimated through assessment that the dose rate at a position 50 cm from the surface of the inert gas piping was about 900
mSv/h.

North

東

West
TIP room

Inert gas piping route(Installation height: 2 m)
Figure: Route of inert gas piping at south side of first floor and shooting position of gamma camera

南©International Research Institute for Nuclear Decommissioning

Figure: Shooting position (1)(from above)

Figure: Shooting position (1)(wall side)

Figure: Shooting position (2) (wall
side)

25

Development of technology for remotely decontaminating ceiling and wall (upper
portion) on first floor and roof of reactor building - 5[Result of investigation at south side of Unit 1 (3): Core sampling investigation]


The following three positions where high dose was observed
were selected for sampling.
Sampling
locationCore size: Φ45 mm  60 to 70 m
Situation of sampling location

Sample (1)Sample (2)Sample (3)Around X-6 penetration(between penetration andpiping)Around water mark atbottom of AC piping

Around left wall of TIP
room door

0.10 mSv/h

0.27 mSv/h

0.09 mSv/h

Sampling situation

Sampleappearance

AroundX-6
penetration

Surface dose
rate (coating
side)
Sampleshape

Aroundopeningof inert
gas piping

Remarks

About 45 mm dia. x 60 to 70 mm (partially diagonal)
Surface partially missing(about 5 mm  20 mm)

Core sampling equipment
Left wallof TIP
room
door



A core drill was mounted to the left arm of MHI-MEISTeR, which was used as a robot fordecontamination, and a core recovery handle was mounted to the right arm.
Power and communications cables were supplied and recovered by a cable drum.

MEISTeR access
route

Sample(1)Sample(2)ShieldSample(3)©International Research Institute for Nuclear Decommissioning

26

Development of technology for remotely decontaminating ceiling and wall (upper
portion) on first floor and roof of reactor building – 6[Basic policy of high-place decontamination equipment] Possible sources of radiation at high places are radioactive vapor adhering to the duct, piping, supports, and electric appliances, and radioactive substances that havepenetrated into the concrete on the ceiling and wall. They are considered to be contaminated by straying, adhering, and penetrating radioactive substances.
 High-place decontamination equipment of different functions, such as to remove radioactive substances adhering to complicated shapes and chip away concrete at highplaces, needs to be developed.
 Based on the above, the concept of equipment that combines “suction/blast”, “high-pressure water”, “dry ice”, and a high-place work cart was designed.

High-pressure water

Concept of equipment

High pressure
water nozzle
Elastic arm

Dry ice blast

Absorption/blast

Tip tool mantled arm

Highpressur
e water
cart

Collecting nozzle

6m elastic instrument
Decontamination
methodSubstances adhering to it are removed byspraying high-pressure water to the object.

Dry ice blocks are shaved and shaved dry
ice is sprayed onto the surface of the object
to be decontaminated to remove
contaminant from the surface.

Hose reel sending blast
to high position[Suction] Contaminant is removed and collectedby the collecting mechanism mounted to thedecontamination head.
[Blast] Abrasives are sprayed by compressed air
to the object to be decontaminated to remove©International Research Institute for Nuclear Decommissioning
27
contaminant from the surface of the object.

27

(2-(1)-1) Development of technology for remote decontamination of inside ofreactor building (2014 plan)
Major goals of 2014
(1) Method to decontaminate portions immersed in contaminated water will be considered. In addition, measures forpreventing dispersion of dust when those portions dry up will also be studied.
(2) Specifications to produce and share remote decontamination equipment for the upper floors will be studied and designed.
Equipment that decontaminates the high places on a floor will be designed and manufactured.
Actions in 20142014 (first half)

1. Gathering basic data on contamination conditions
Considering the conditions of contamination of portions immersed in contaminatedwater and the portions to be decontaminated, measures for preventing dispersion ofdust when those portions dry up will be considered. In addition, a test to confirm theeffect of suppressing generation of dust will be conducted by simulating
contamination.
2. Adjusting decontamination technologies and studying concept of
decontamination
Concept of decontaminating portions immersed in contaminated water will be
studied. In doing so, specific portions of the real plant will be assumed.
3. Designing and producing remote decontamination equipment and verifying
remote decontamination2014 (second half)

1. Gathering basic data on contamination
conditions• Study of dust prevention measures• Confirmation of effect of dust prevention
measures
2. Putting decontamination technology inorder and designing decontamination
concept• Study of concept of decontaminatingparts immersed in contaminated water
3. Designing and producing remotedecontamination unit and verifying
remote decontamination
3.1 Development of high-placedecontamination unit
• Factory mock-up test
• Verification with actual machine
3.2 Development of upper-floor
decontamination unit
• Production of unit
• Factory mock-up test

3.1 Development of decontamination equipment for high places
Mock-up of the high-place decontamination equipment produced in 2013 will betested in factory. Part of a high place of the real plant will be simulated for thefactory mock-up test. In addition, verification with the real plant will be conducted athigh places on the first floor of the reactor building of Unit 1 to 3 at Fukushima
Daiichi nuclear power plant.
3.2 Development of upper-floor decontamination equipment
Mock-up of the upper-portion decontamination equipment designed in 2013 will betested in factory.

©International Research Institute for Nuclear Decommissioning28

2. Research and development related to preparations for removing
fuel debris
Development of technologies for investigation and repair (stopping
water) toward filling PCV with water(2-(1)-2 & 3)（

©International Research Institute for Nuclear Decommissioning29

(2-(1)-2 & 3) Development of technologies for investigation and repair (stopping water)
toward filling PCV with water・Investigation: For the lower portion, equipment has been designed and mock-up for factory test has been created, and equipment performance test and assessment ofapplicability of the real plant have been completed.
For the upper portion, equipment has been designed, produced, and checked for performance, for each portion checked.
・Repair: For the lower portion, repair technique and water-stopping materials have been studied in detailed and tested toward designing and production of equipment.
(Stopping water) For the upper portion, results of tests, etc., have been reflected on the detailed designing of water-stopping materials toward production of repair
equipment that will be applied to places heavily damaged.

Actions
1. Development of technology for investigating PCV
1.1 Development of equipment to inspect lower portion of PCV・Equipment that inspect the lower portion of the PCV and detect leakage ofwater from a reactor building to the adjacent buildings are produced.
Factory mock-up test equipment has been created and its performance has
been confirmed.
・A plan to assess applicability of the real plant (on-site verification) has beendrawn up and on-site verification has been completed.
1.2 Development of equipment to inspect upper portion of PCV・PCV upper portion investigation equipment has been designed, produced,
and checked for performance for each portion to be inspected. Basic type,small-diameter penetration has been conducted for a device that identifies
leakage of the dry well outer opening investigation equipment.
・A plan to assess applicability of the real plant has been drawn up. On-siteverification is slated for 2015.
2. Development of PCV repair (water-stopping) technology
2.1 Development of PCV lower portion repair equipment
・Details of the repair technique is studied toward designing and production of
repair equipment that makes up a boundary by using a vent tube and a
suppression chamber.
Completion of (detailed study of water-stopping materials through waterstopping test and optimization of auxiliary closing materials)
2.2 Development of PCV upper portion repair equipment
・Results of water-stopping test will be reflected on detailed study and designof water-stopping materials in order to produce repair equipment to be
applied to places that are highly likely damaged (hatch flange, bellows of
opening, and electric penetration) may have been damaged.

[D/W outer narrow part]
Long equipment[D/W outer opening]
Stretchable lifter[S/C lower outer side]
Magnetic crawler[Vent tube – D/W joint]
Magnetic crawler[S/C upper part]
Catwalk running unit
[Torus wall side]
ROV and floor running unit

Figure: Location where investigation/investigation equipment is
used©International Research Institute for Nuclear Decommissioning

30

Current status of lower
part of PCV 【Unit 1】
LeakageVacuumdestructionline

Ｄｒｙ ｗｅｌｌ（D/W）

【ｐｉｃ①】①Confirm Leakage from thebellows portion of
vacuum break line
(See pic①)

R/B １ｓｔ OP.10200

Ｗａｔｅｒ ｌｅｖｅｌ
OP.9000（Ｄｅｐｔｈ 3m）

１

Vacuum break line

Suppression
chamber（S/C）②Confirm running water fromthe sand cushion drain line
(See pic②）

Fuel debris(Place , distributionand , shape areunknown２

Sand cushion drain
line
Confirm
runningwater【ｐｉｃ②】→To pump
Torus room

（R/B basement FL)

Strainer

TEPCO data

©International Research Institute for Nuclear Decommissioning

31

Current status of lower part of PCV 【Unit ２】

Water level

R/B 1st FL OP.10200

OP.5824（Depth 30cm）

300A-AC-10

Dry well（D/W）

Not confirm running waterleakage from
①Vent tube sleeve（See pic①）②Sand cushion drain line
（See pic②）③ Bellows cover（See pic③）

Fuel debris
500A-AC-1(Place , distributionand , shape areunknown

Suppression
chamber（S/C）
１
３

４

２

④Confirm that the waterlevels of inside and outside
S/C are almost the same

Strainer【Pic①】→To pump（R/B basement FL)【Pic②】

Torus room
【Pic③】TEPCO data
32

©International Research Institute for Nuclear DecommissioningCurrent status of lower part of PCV 【Unit ３】
Main steam pipeＣ

Dry well（D/W）

bellowsＣ

Water level１

R/B １ｓｔ OP.10200

OP.12000（Depth 6.5m）
Assumed from pressure【Pic①】

difference between
D/W and S/C

Suppression

Leakage①Confirm leakage frombellows of the main steam
pipe in the room of main
steam isolation valves( See pic①）

chamber（S/C）
Fuel debris(Place , distributionand , shape areunknown

Strainer

Torus room

→To pump（R/B basement FL)

Sump Pit
TEPCO data

©International Research Institute for Nuclear Decommissioning33

Development of investigation and investigation equipment (1)(portions subject to construction)
Details of portions to be constructed

Opening outside D/W

７．

Narrow portion
outside D/W

Checking leakage by
raising to aroundopening

Opening outside

Upper portion of
S/C
Operation rack

Inserting long arm through drilledhole to observe opening withcamera

X105B

X106

X54
X53

Equipment
hatch

Reactor building
cross section

Vent tube-D/W joint
Confirming joint by absorbing to vent tube from hole on firstfloor and moving

Torus room wall

Vent tubeD/W joint

S/C

Checking waterleakage by indirect
visual investigation

Vent tube
bellows

Confirming lower portion of S/C
by absorbing it to shell

Upper portion of S/C

Vacuum breaker

S/C
S/C manhole

X2

Checking opening on wall of torus room by moving in the water
Opening on
torus room wall

Checking opening
on torus room
wall with camera
for damage

Checking leakage
by detecting
tracer withultrasonic wave

Water staying

Visually checking
damage thatprevents flooding

Water
staying

X100C X100B X100A

Torus room wall andtriangular corner

Outer surface atlower portion of
S/C

Outer surface atlower portion of S/CDrilling hole on ceiling of small
room and observing opening
through hole

R/B, 1st floor: Position of drilling

R/B, 1st floor

Vent tube-D/W joint

Narrow portion
outsideD/W・エアロック室Checking water at upper portion of S/C
by moving on catwalk

Triangular corner

Checking leakage at
triangular corner

Catwalk

Checking object at
Inserting long arm
upper portion of S/Cthrough drilledwith camera on top ofmask that stretches inhole to checkvertical direction
damagewith©International Research Institute for Nuclear
Decommissioning

Water
staying

camera34

Development of investigation equipment (2)(vent tube-D/W joint and S/C lower portion investigation equipment)
Vent tube & S/C upper portion simulator
and tester

S/C lower portion investigationequipment

Magnet wheel (x4)PCVシェルPCV shell

Front & rear cameras,side cameras (left &right)

S/C
shell
S/CシェルS/C outercircumferenceTop

Robot runningroute
S/C inner
circumference

Anti-seismic
supportMarkerVent tube

Bottom
Columnsupport

Wheel lift

Vent tube-D/W joint
investigation equipment
Magnet wheel (x4)

Simulated
模擬ｺﾝｸﾘｰﾄｽﾃｰｼﾞ
concrete stage
andwater tank 1
兼水槽1

Water水槽2
tank 2

S/C shell and water tank assembly
statusSimulated biologicalshieldingconcrete
模擬生体遮へいｺﾝｸﾘｰﾄ

Underground pin
installation status

Rear camera
Vent tube

Robot suction position
Robot
Robot running route

Vent tube and concrete sleeve(vertical installation)

Wheel lift

Front camera
・The equipment can suck the vent tube and S/C at any position.
・It can get over a difference in step of about 5 mm in height and 15 mm in width.
・The equipment can move at any position and posture without wheel lifting.

©International Research Institute for Nuclear Decommissioning

35

Development of investigation equipment (3) (Torus room wall, S/C upper portion, and D/Woutside narrow portion investigation equipment
V1 to V4 & H1 to H2
Four vertical thrusters
and two horizontal
thrusters

Camera can beextended 3 m
or more.

When
extended

Thermomet
erhygromet
ermicropho
ne equipped

SurveillancecameraCable reel
Elevatingmast

Pan/tilt
zoom
cameraWhen
withdrawn

7 m strokeWhen
withdrawn

Whenextended

Crawler

Elevatingmast
Elevating mast

Horizontal floor
removing equipment

Underwater swimming
equipment

Horizontal floor moving
equipment

Underwater swimming
equipment

・Small investigation equipment thatcan move on floor and move underwater
・Equipped with sonar and can checkleakage under water・Thruster has long life and easy tomaintain by using magnet coupling.
・Camera can move up/down andleft/right about 90 to have wide view
angle.

7 m max.

Pan/tilt zoom
cameraS/C upper portion
investigationequipment

装置
S/C upper portion
investigation equipment

D/W outer narrow portion
investigation equipment

・Camera can be extended 3 m or more to・Camera is extended (up to 7 m) andパンチ
inspect structure at upper portion of S/C.
inserted
into room to check objects
・Equipment measures 600 mm square and caninside ルト©International
Research
InstituteforNuclearDecommissioning
move on catwalk at upper portion of S/C.
・The equipment
canvoluntarily moveズーム
36
with camera
withdrawn.
カメラDevelopment of investigation equipment (4) (D/W outer opening investigationequipment)
investigation method
Typicalpenetration to becheckedImage of investigation at base ofpenetration

Checking leakage by accessing PCV
penetration at high place

PCV shell

Inserting leakage
detector to baseof penetration

Detecting flow of nitrogen gas inatmosphere

Enlarged view ofinside of penetration
ペネ内部拡大図

Biological shielding
barrier

Inserting leakage detector

CameraLeakage detector

Biological shielding
barrier

CRDM macaroni
pipingOverview of equipment (Red: Range of production)
7-axis manipulator

HeadLeakage detector

Leakage detector

CameraMicrophone

Tool changer

Drum rotation
axis

Biologicalshielding barrier

Actual machine simulation test
(under preparation)

Design and production status of equipment

Unloading module

PCV

Leakage detection
performance test

Control board box
Motor for feeding axis
Pipe insertion blockchecking camera

Drum

Minute particle
spray nozzle

Leakage detector feeder
Feeding axis

Head
ヘッド部

Driving performanceverification test

Unloading module(under production)
Cart and manipulator are of NEDO.

Leakage detector (head)

Cart©International Research Institute for Nuclear Decommissioning37

Development of PCV repair technology（１）(Part for repair)
Penetration partProtective cover

Cable connector
seal part

Penetration sleeve

PCV upper part
Bellows
Joint box

Electric wiring penetration

Bellows of tube penetration

PCV lower part

Part for repair of PCV upper part
Flange
PCV upper part

Part for repair of PCV lower part
Drywell (D/W)

Vacuum brake line (only UNIT-1)
Vent tube

Suppression chamber

Main steam safety valve vent line

Quencher

PCV wall

Apparatus hatch

Water path
Stopping water part©International Research Institute for Nuclear Decommissioning38

Development of PCV repair technology（2）(PCV lower part)
Result of 2013 (example)

Concept of construction methodof PCV lower part repair

Bent tube

Auxiliary material

Manipulator

Stopping water material

Construction method of stopping water at vent tube
Drywell(D/W)

Vacuum breakline (only Unit-1)

Concept of PCV lower part repair equipment

Bent tube
Suppression chamber (S/C)
Quencher
Main steam safetyvalve vent line

QuencherWater path
Stopping water part

Stopping water material

Construction method of stopping water at quencher seal Element test of stopping water material at vent tube
Manipulator

seal

Grout material
Guide pipe

seal

Reinforcement of suppression chamber

Construction method of stopping water at vacuum break line

39

of auxiliary
maerial©International Research Element
Institute fortest
NuclearDecommissioning

Development of PCV repair technology（3）(PCV upper part)Concept of construction methodof PCV upper part repair

non-cement material

Concrete wall

PVC inside

Concrete wall

PVC shell
non-cement material

Established seal

PVC inside

bellows
Joint box

protective cover

PVC shell
Penetration sleeve

Penetration sleeve

Protective coverPouring noncement material

PCV upper part

Pouring noncement material

Joint box

Construction method of repair at the open part of D/W outside
Element test of stopping water material
at the open part of D/W outside
Punching position

Repair equipment

Cement material(burial)Cement material(floodgate)

MS penetration bellows

Element test of stopping water material
at the narrow part of D/W outside

Construction method of repair at the narrow part of ©International
D/W outside
Research Institute for Nuclear Decommissioning40

(2-(1)-2 & 3) Development of technologies for investigation and repair (stopping water) toward filling PCV
with water (2014 plan draft)
Major goals of 2014
[Development of PCV investigation technology]• Of the PCV upper portion investigation equipment, study of improving the specifications of the dry well outer narrow portion investigation equipment will be completed
based on the record of accessing the targeted portion last year. Study of improving the specifications of the dry well outer opening investigation equipment will becompleted based on the result of a factory mock-up test. Study of the concept of the leakage detector of the dry well outer opening investigation equipment will becompleted as a series of equipment improvement study work because of the special penetration (penetration substantially deviating from the large-diameter
penetration and positioned at the center of the penetration group).
• Study of the concept of equipment for inspecting new portions (knuckle portion of the PCV) will be completed.
[Development of PCV repair (water stopping) technology]• As water-stopping techniques that can be applied at the site, (1) detailed designing of the PCV lower portion repair (water stopping) equipment and drawing up anelement test method will be completed, and (2) study of improving the specifications of the PCV upper portion repair (water stopping) and drawing up an element test
method will be completed.
７．

Actions in 2014
[Development of PCV investigation technology]
1. Development of lower portion check/investigation equipment• Development was completed by works in 2013
2. Development and improvement of upper portion check/investigation equipment• Study of improving the specifications of the dry well outer narrow portion investigation equipment will
be completed based on the result and record of accessing the targeted portions last year. Study of
improving the specifications of the dry well outer opening investigation equipment will be completed
based on the result of a factory mock-up test. Basic design of a device for special penetration will becompleted.
• Study of the concept of the PCV knuckle portion investigation equipment, which was newly needed,
will be completed.
[Development of PCV repair (water stopping) technology]
1. Development of PCV lower portion repair (water stopping) technique and equipment• Detailed designing of a repair equipment for boundary configuration with a vent tube, quencher,down-comer, suppression chamber, S/C joint tube, and others will be completed. Applicability ofwater-stopping materials that are used to repair these portions will be confirmed by conducting awater-stopping test in 1/2 scale. Portions of the torus room wall, triangular corner, and sleeves
between buildings where water must be stopped and water stopping methods will be considered andan element test method will be created.
・Specimen for a mock-up test and test equipment will be designed and produced.
2. Development of PCV upper portion repair (water stopping) technique and equipment• A water-stopping test in about ½ scale will be conducted to check portions that may have been heavily
damaged (such as hatch flange, opening bellows, and electric penetration) and the applicability will be
confirmed. As necessary, an element test method will be created.

Schedule (2015)2014 (first half)
D/W outer narrowportion2014 (second half)

Reactor building cross section
D/W outer opening

Study of improving equipment
Test plan
Specification
study

PCV knuckle
portionEquipment basic design
Test plan

Study
DesignFactory mock-up test
Concept study

Equipment detailed design
Water-stopping test

Crafting element test method
Production

Test plan

Water-stopping test

Method study

Crafting element test method©International Research Institute for Nuclear Decommissioning41

2. Research and development related to preparations
for retrieving fuel debris
Development of technologies for investigating insideof PCV(2-(1)-4)©International Research Institute for Nuclear Decommissioning42

Overall plan (Objective and goal of internal PCV investigation technology)[Objective of internal PCV investigation technology]
To develop the investigation technology to grasp internal PCV status required forfuel debris retrieval[Goal of PCV investigation]
As Fuel debris is assumed to exist
inside PCV, development of device
to take picture and detect meltagethat may include fuel debris, andthe device to approach to the survey
target area are set as a goal.
The ultimate goal is;
1. Develop the measuring device todetect meltage
2. Develop the device to approachinside the pedestal
3. Develop the device to approachoutside the pedestal
4. Implement feasibility tests on theabove devices

PedestalPedestalentrance

Approach from
X-6
Penetration the access

Accessentrance
to thebasementFL

entrance into
pedestal

Travel on thegrating at 1st FL

Investigationof basement
FL

Access route inside pedestal
Access route outside pedestal
Image of approach for investigation : *1
©International Research Institute for Nuclear Decommissioning*1: This access route may bechanged depending on the future examination

43

Development plan for investigation method and device
Set the development plan based on estimated condition of RPV and PCV of Unit 1 to Unit 3 (*1)
Unit 1

Unit 2・Almost all of melted fuel havebeen fallen down to the bottomof RPV plenum and little fuelhave left in RPV.
Development plan
・There is a possibility that fueldebris exists even outside of thepedestal, and investigation
outside the pedestal should be
conducted as priority.

Unit 3・While some part of melted fuels has fallen down to the bottom of RPVlower plenum and PCV pedestal, the other part may have been left
inside RPV.
・Presumed that more fuel than having estimated may have fallen
down to PCV in Unit 3.
Development plan・As the possibility that fuel debris spread outsides the pedestal islower compare with Unit 1, investigation inside the pedestal should be
developed as priority.
・As in Unit 3, the water level inside the PCV is high, penetration which
will be used in Unit 1 and 2 must be submerged, other methods should
be examined.

©International Research Institute for Nuclear Decommissioning*1：[Reference] TEPCO’s webpage Dec. 13, 2013 “The first progress report related to estimated state of reactor core and RPV of Fukushima
Daiichi NPS Unit 1, 2 and 3, and unsolved issues”

44

(2-(1)-4) Development of technology for investigation of inside of PCV(2013 result）• Investigation equipment for unit 1 was produced and its function verification test was completed as for pre-survey of outside of thepedestal (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 forfull 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:
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 removed 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, and equipment 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 the previous year and issues, they are to be verified forchanges 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, and establishment of specification of element making is ongoing.
Also, verified is concept for access equipment required for prevention of dispersion of radioactive material when sending equipment into
PCV.
Element making/test done by FY 2017.
3. Development of investigation equipment and technology（Debris measurement apparatus）©International
Institute for Nuclear
Decommissioning
Established equipment system structure for technology of measuring shape by optical-cutting
method.ResearchAlso, element
test formeasurement
simulating disturbance environment (spray, rain etc. ) inside the PCV is ongoing.
45

Development Steps of Each Unit (Unit 1)[Investigated area] : Outside the pedestal on the basement Fl. Near the access entrance[Investigation and steps to develop equipment](1) Investigation from the X-100B penetration （～FY2015）
As high dose rate was detected near the X-6 penetration, investigation is planned from penetration X-100B(Φ100mm) .
①To obtain information on the grating inside the PCV 1st Fl. (ex. Investigation on usability of CRD rail) : B1
②Based on the result of investigation in the torus room implemented in Nov. 2013, investigation to obtain picture
outside the pedestal on the basement Fl. is planned(near the access entrance and around the vent piping) : B2
(2) Investigation from X-6 (FY2016～FY2017)①Further investigation to gather information using debris shape measurement apparatus outside the pedestal on thebasement Fl. will be conducted ：B3

Examine necessity of additional investigation
B1．Investigation on the grating outside
the pedestal on first Fl.
（Planned in later FY2014) : use X-100B

B2．Investigation outside thepedestal on the basement Fl.
（Planned in FY2015）: X-100B
X-100BPenetration

Grating
opening on
1st Fl.

Depending on resultof B2 investigation, B3
test may be
conducted.

B3. Investigation on outsidepedestal on the basement Fl. And
workers entrance(Planned in FY2016～17) : Use X-6

X-6Penetration

Investigation inside the pedestal may
be conducted depending on theinvestigation of Unit 2.
46©International Research Institute for Nuclear Decommissioning

Workers entranceStatus of Development (Equipment for outside pedestal)
3.3.1 Investigation equipment of grating on the 1st floor outside pedestal(1)Overview of equipment
Shape-changing crawler equipment is inserted from the narrow access entrance (X-100B penetration opening：
φ100mm) and safely travel on grating stably.

(2) Image of investigation route and equipment
X-100BD/W 1st Fl. grating

Traveling through piping

Existing guide piping

Traveling direction

X-100B

Board camera*Use whentraveling
throughpiping.

Existing guide piping

Composite cable
Crawlertransformation
Traveling on grating
Thermometer*Installedinside cover

D/W basement Fl
ＰＣＶ
X-6：Investigation route (Plan) （※１）

[Cross Section of PCV]

*１ Investigation route and area are images andmay change depending on the condition of the PCV.

High dose
rate【Ｄ/W 1st Fl.】Internal PCV investigation route (Plan)

Camera forinvestigation

Features of equipment(3) Status of development①Extract improvements from stand alone/combination test held in the year before and to be applied②Toward the function verification test scheduled in Feb., form a test plan anddesignResearchand Institutemanufacture
full scale©International
for Nuclear Decommissioning
47
model.

Additional Study based on result of investigation from a boat at Unit 1
・When Unit 1 was investigated from a boat at Unit 1, water leakage from the sand cushion
drain pipe was found out. As early investigation on debris inside PCV should be required,additional investigation is planned.
・Equipment is inserted from the X-100B penetration to the basement Fl. outside pedestal,and approached around workers access entrance which is assumed that debris had spread.
・Plan to modify the shape-changing crawler equipment which is under development toinvestigate outside pedestal on the 1st Fl. And it can be applied to basement Fl.
Guide piping：Access route

X-100B

Workers
Access
Entrance

Grating
Accessopening on
basement Fl.
Basement Fl.
Whole picture of access route

Vent Piping
X-5E

Developing
shape-changing crawler equipment©International Research Institute for Nuclear Decommissioning48

Development Steps of Each Unit (Unit 2)[Investigated area] : On the platform(Upper surface of platform, Bottom of CRD housing) andbasement (Basement Fl.)
[Investigation and steps to develop equipment](1) Investigation from X-6 penetration （Φ115mm）（within FY2014）
・Plan investigation of inside pedestal platform from X-6 in later FY2014 : A2
(2) Investigation from the X-6（Enlarge hole or penetration）（FY2015～2016）：A3～A4
・Insert debris visualization system, investigate inside pedestal.

A1. Investigation on
CRD rail(Conducted inAug.2013)

A2. Investigation on platform
inside pedestal (later FY 2014）

Step to use X-6 penetration
・Remove shield in front of penetration
・Pierce a hole to penetration hatch・Remove inclusions inside penetration

A3. Investigation on bottom of
CRD and platform（later FY2015）
ＲＰＶ

A4. Investigation of pedestal
on the basement (later
FY2016)

ＣＲＤ
Pedestal
Opening
CRD railX-6
penetration

Platform

ＰＣＶ

Traveling on
grating of 1st Fl.
©International Research Institute for Nuclear Decommissioning

Based on the result of internal investigation from A2 to A4, investigation outside pedestal may be conducted.

49

Development status (Fuel debris measuring equipment)(1) Objective of development
To develop measuring equipment that analyzes the position and distribution of fuel debris and expected moltenmaterials(2) Actions
・Basic design of the equipment will be decided based on the optical-cutting method selected in 2012.
・The performance will be confirmed by element tests under the operating conditions to assess thevalidity of the basic design.
・ Additional examination of measuring techniques other than the optical-cutting method(such as component measurement) will be conducted and feasibility of equipment using thosetechniques will be assessed.
No.
1

Image

2
Cameraoptical axis
centerSlit lightsourceCameraSlit lightsource3-dimensional measurement by
triangulation
Rotating
mirror

3
4
5
6
7
8

Appearance of equipment

Depth map

Principle of optical-cutting measurement

9

Category
Temperature
distribution
measurement
Dose
measurement
Component
measurement(direct
measurement)
Component
measurement(sampling)

Measuring technique
Probe thermometer
Radiation thermometer
Dosimeter
Gamma ray detector
Compton camera
Laser-induced fluorescencemethodLaser-induced breakdown
emission spectrographs
Gas sampling
Solid sampling

Measuring techniques other than those formeasuring appearance and shape(3) Elements of development
・ Measurement performance and resistance to radioactivity in foggy atmosphere,
rain drop-containing atmosphere, and water©International Research Institute for Nuclear Decommissioning
・ Small and lightweight so that the equipment can be mounted on access equipment,
50
while maintaining high measurement performances(2-(1)-4) Development of technologies for investigating inside PCV (2014 plan draft)
Major objectives of 2014
A verification test with actual investigation device for advance investigation (to obtain pictures, dose rate and temperature data) of inside PCV
under an environment of high dose rate 100 Gy/h will be completed. In addition, basic design of investigation device that will access for fullscale investigation the inside and outside pedestal in PCV where fuel debris is expected to exist and production of some parts will becompleted.

Actions in fy2014*1 and 2 are for pre-investigation (to obtain images, dose rate, temperature andobstacles) and 3 and 4 are for full scale investigation (to measure distribution andshape of fuel debris).

1. Development and verification test of device to investigate situations of theplatform inside pedestal(1) Development of the device will be completed based on development of lastyear and by conducting verification tests and making improvements.
(2) Verification tests through X-6 penetration of Unit 1 will be conducted.
(The device includes investigation device which remotely removes theshielding block in front of X-6 penetration, device which drills a hole on thehatch of X-6 penetration, and device which accesses inside PCV.)
2. Development and verification test of device to investigate outside pedestal(1) Development of the device will be completed based on development of lastyear and by conducting verification tests and making improvements.
(2) Verification tests using X-100B penetration of Unit 1 will be conducted.
(The device accesses inside PCV through X-100B.)
3. Development of device toward further investigation inside and outsidepedestal
Basic design of device that accesses portions where fuel debris is expected toexist (on platform, pedestal basement, and near the worker access openig) will
be completed and production of some equipment will be started (completion isslated for March 2016).
4. Development of debris measurement device
Study of mounting fuel debris measurement device that uses a measurement
technique such as optical-cutting method to device in 3 above will be conducted
and design and production will be started.

Schedule (2014)2014 (first half)2014 (second half)

Device design/production/verification/
improvement
Verification test (Unit 2 X-6)

Device design/production/verification improvement
Verification test (Unit 1 X-100B))

Device design/production

Device design/production©International Research Institute for Nuclear Decommissioning51

2. Research and development related to preparations
for retrieving fuel debris
Development of technologies for investigating
inside of RPV(2-(1)-5)©International Research Institute for Nuclear Decommissioning52

Project overview (Objective of investigation of inside RPV)
[Objective of investigation of inside of RPV]
The purpose of investigation of inside RPV is to identify the position offuel debris inside RPV, estimate damage of core internals, and measuretemperature and dose rate inside RPV.
In this project (to investigate inside RPV), technologies that enable theabove investigation will be developed.

[Examples of expected results]
Information to be obtained inside RPV and expected results arelisted below.
Information to be obtained

Expected result• Images of fuel debris and coreinternals• Dose rate of reactor atmosphere andfuel debris• Temperature of reactor atmosphereand fuel debris

The situation inside the plant can be directly
confirmed, which can be fed back to a fuel debris
removal plan and device development.

• Sampling of fuel debris

At present, details of fuel debris are not known.
However, by sampling and analyzing a small
quantity of fuel debris, the composition of the fueldebris can be confirmed.

CS system

Feed
Water
systemStatus of Unit 1

CS system

Feed
Water
systemStatus of Unit 2

Feed
Water
systemCS system

Status of Unit 3[Estimated damage to reactor]©International Research Institute for Nuclear Decommissioning53

(2-(1)-5) Development of technologies for investigating inside of RPV(fy2013 results)
To identify the position of fuel debris in RPV and damages to core internals, and to measure the dose rate and temperature in RPV,
methods of accessing to the investigation portions, investigating, and sampling have been considered and technologies for inspectingin RPV, where dose rate is high (preliminarily estimated at 1,000 Gy/h), have been arranged, and development of technologies forinvestigating inside RPV have been planned (to develop technologies for investigating via piping systems in fy2015 and fy2017, thosefor drilling a hole on top of RPV for investigation in fy2018, and those for investigation in fy2019 after the reactor opened).

Actions
1. Planning for investigation of inside RPV
 Selection of major items to be investigated and period of investigation
Items to be investigated inside RPV were selected and the period of investigation was considered. As a result, the plan for technological development (todevelop technologies for inspecting via piping systems in fy2015 and fy2017, those for drilling a hole on top of RPV for investigation in fy2018, and thosefor investigation in fy2019 after the reactor opened) has been drawn up.
 Selection of access route
Candidate access routes to investigate inside RPV, such as a route to access through piping, one to drill a hole on top of RPV to gain an access, and one
for accessing after the reactor opened, were selected. Accessibility of each route was assessed and a candidate route was decided .

2. Planning development of technologies
 Access technology
Based on the result of selection of a route to access inside RPV, the existing technologies were examined and issues to development, such as how to go
through obstacles, were removed.
 investigation technology (radiation-resistant camera, dosimeter, etc.)
Applicability was considered by examining the existing technologies and issues such as how to resist radiation were removed.

 Sampling technology
Concept of sampling technique was examined by surveying the existing technologies, issues were organized, and a development plan was formulated.

©International Research Institute for Nuclear Decommissioning

54

Study of access route to inside RPVTable Development plan of access technology (sample of access by drilling on the upper section of RPV)No

Development technology
element

1

Boring technology

2

Tube expansion technology

3

Remote control technology

4

Boundary forming technology

Issues

2014

2015

2016

2017

2018

Creating access hole for the steam dryer andseparator
Tube expansion for the hole diameter ofsteam dryer and separator
Monitoring the passing on the curve andnarrow part of access route, and operating
condition
Boundary reforming on the operation floor(sealed plug)

Access through piping systems

Access via drilling hole on
the upper section of RPV

Access after RPV opened©International Research Institute for Nuclear Decommissioning55

Shield Plug

Well

Access route
ManholeN-8A
RPV head spray nozzle

N-10
Vent nozzle

Piping Access

Access via drilling
Hole on the upper
sectionAccess after
RPV opened

N-8B
RPV head spray nozzle
Lagging material

Steam dryer Bottom
plate drain piping port

Drilling hole on
Dryerbottom platedrain panStand pipe of
Separator (Inlet)

Drilling hole on
Dryer aroundbottom plate

Drilling hole on
Dryer
Side plate

Drilling hole onside of
Separator
stand pipe
©International Research Institute for Nuclear Decommissioning

56

Access route to investigate inside RPV
Planning investigation of inside RPV (routefor accessing via piping systems)
Feed water system A(Feed water supply
nozzles N4A and B)
OP26601

Result of selection (example of Unit 3)•As systems that have opening to RPV, the jet pumpinstrument line, feed water piping, core spray piping andothers were removed.
•The challenges are at bending of the system piping,
opening or closing of the valves of access piping, andsecuring dimensions that allow passage in accordancewith design concept of equipment, among others, andwill be further narrowed down in accordance with thedesign of equipment.

Core spray (CS) system A(CS nozzle N5A)
OP26131

Jet pump instrumentation
piping (jet pumpinstrumentation nozzle N8)
OP17673

OP14320（Vessel0）©International Research Institute for Nuclear Decommissioning57

(2-(1)-5) Development of technologies for investigating inside of RPV (fy2014 plan)
Major objectives of fy 2014
• To develop technologies for investigating inside RPV (to develop technologies for investigating via system piping in fy 2015 and fy2017, those for drilling a hole on
top of RPV for investigating in fy2018, and those for investigation in fy 2019 after the reactor is opened), system design and basic design of access devices,
investigation devices, and sampling devices that takes photographs and measure the dose rate and temperature inside RPV (access device) where dose is as highas 1,000 Gy/h (preliminarily estimated value) will be completed.
• Trial production and test operation of element technologies by which to produce the access device, investigation device, and sampling device will be completed.
Actions in 2014
1. Development of access device(1) Technologies for accessing inside the RPV via piping will be completed
by designing the system of device and experimentally producing andtesting element technologies that constitute the device.
Toward early realization of access to the inside RPV, basic design anddetailed design of the device will be completed, using the dimensions,weight, etc., of the investigation device to be developed in 2 below as
input.
(2) System design of device will be performed by using technologies togain an access from the top of RPV (technologies to drill a hole on topof RPV and make an access after the reactor is opened), andexperimental production and test of element technologies that
constitute the device will be completed. In addition, basic design anddetailed design of the device will be completed, using the dimensions,weight, etc., of the investigation device to be developed in 2 below as
input.
2. Development of investigation device
System design of equipment, such as a camera, will be completed, taking into
consideration the dimensional constraints of the investigation device andaccess route. Basic design of the investigation device will be performed andinput information that contributes to study of the access device in 1 above,
such as dimensions and weight, will be proposed.
3. Development of sampling device
System design and basic design of the sampling device will be performed,
taking into consideration the methods to process and recover sampled fueldebris.

Schedule
Item/year

1st half of2014

2nd half of2014

1. Development of access technologies
Piping access・System design of device
・Basic design of device
・Detailed design of device
・Trial production and test of elements
RPV upper portion access・System design of device
・Basic design of device
・Trial production and test of elements

2. Development of investigation technologies
・System design of device
・Basic design of device
・Trial production and test of elements

3. Development of sampling technologies
・System design of device
・Basic design of device
・Trial production and test of elements©International Research Institute for Nuclear Decommissioning58

2. Research and development related to preparations for
retrieving fuel debris
Development of technologies for retrieval of fuel debris/coreinternals (2-(1)-6)©International Research Institute for Nuclear Decommissioning

59

Overall plan (objective and system)
Major objective in fy 2014
Purpose: To estimate situations of inside RPV and dispersion of fuel debris to contribute todecommissioning of Fukushima Daiichi nuclear power plant

Estimate of inside RPV situationand field of analysis
Actual measurement of plant situation and remote visualization (TEPCO)[Project commissioned by Agency of Natural Resources and Energy]

Project for preparing technological base for safely decommissioning reactor for
power generation, etc.
(Estimating situation of inside RPV by using severe accident analysis code)
Improvement of analysis code and analysis of actual plant accident
・MAAP Group (Toshiba, Hitachi GE)
・SAMPSON Group (Institute of Applied Energy)
Assessment of individual events by heat flow analysis, etc.
International cooperation: OECD/NEA benchmark analysis PJ (BSAF), etc.
JAEA: Mock-up test (such as seawater heat transfer test, from fy 2013)
AESJ: “Severe accident assessment” research expert committee©International Research Institute for Nuclear Decommissioning60

(2-(2)-1) Estimating situation of inside RPV by advancing technologies for accident progression
analysis (fy 2013 result)
Accident progression analysis technologies for estimating the reactor conditions, such as the position of fuel debris, were advanced(by improving a reactor damage progression model, fuel debris behavior model in a lower plenum and others). Study to estimate theconditions of the reactor was conducted by using the result of the advanced accident progression analysis technology and based on
new information obtained from on-site operations, and it was estimated that the ratio of debris that fell into the PCV was the
maximum at Unit 1 and lower than or equal to that at Units 2 and 3

Actions
1. Validity Confirmation of improvement of code and addition of model
Ranking of the importance of PIRT (Phenomena Identification and Ranking Table) completed in fy2012 was newly confirmed andrevised. Its importance rank by sensibility analysis.
2. Improvement and advancement of analysis codes
Analysis codes (MAAP and SAMPSON) were improved to enhance the analysis accuracy, based on information gathered throughoperations at the site, result of existing simulation test, and newest information and opinions.
3. Analysis with improved codes (MAAP and SAMPSON)
Based on the latest version of the improved codes and database built up, accident progression and reactor conditions of Units 1 to
3 were analyzed to confirm the influences of improvement of the model.

4. Analysis of individual phenomena by CFD
A test to analyze spreading fuel debris was conducted by using cast simulation codes to confirm applicability of the codes to fueldebris diffusion behavior assessment with the real plant.
5. Mock-up test
Simulated tests contributing to detailed analysis of severe accident phenomenon progression (such as seawater heat transfer test
and molten fuel drop behavior test) were conducted to confirm applicability of the existing heat transfer assessment equations at atime of seawater injection.
©International Research Institute for Nuclear Decommissioning

61

Improvement and addition of model of MAAP
Enhancing assessment accuracy in estimating fuel debris position and plant behavior by improving MAAP code
■ Improvement and verification of MAAP5 code
Note)
Top: Before Improvement of code was commissioned to EPRI in the U.S.
improvement
Bottom: After
based on the following improvement items and theirimprovement
advancement specifications:• Improvement of reactor damage progression model(Migration pathways of molten material were considered.)• Improvement of lower plenum fuel debris behavior model
Reactor core
(Deposit form, interaction of structures)
Molten core
• Improvement of PCV fuel debris behavior modelCore plate(Spreading behavior, concrete interaction)
 Verification of improved code
Particle• Model of individual phenomenon was verified by element test, shaped
debris
Metal layeretc.
• Overall plant behavior was verified by actual-plant test, etc. Molten fuel pool

Single transfer path
Multi paths
ShroudRPV wallJet pump
Fuel support

CRD guide tube

Solidified crust

Added route (1)Fuel debris fell into holeof fuel supported.

Added route (2)Fuel debris falling into ahole into which a control
rod was inserted (crossshaped).

PCV floor
Uniform fuel debris deposit

Fixed fuel debris spreading

Consideration ofasymmetry/details of interaction
with structures

Modeling spreading behavior
based on physicality©International Research Institute for Nuclear Decommissioning62

Addition and improvement of model of SAMPSON(a) Modeling thermal stratification phenomenon in PCV suppression chamber・Three-dimensional flow model to solve natural circulation of water was developed(coordinate system: Descartes/cylindrical) ）・ Analysis capability of the model when RCIC operated was verified by analysis of Unit 2.
(b) Improvement of model of outflow to lower plenum and interaction model between moltenmaterials and structures/coolant at lower plenum
・ Route for molten core materials to flow out to lower plenum was studied through analysis
and model was improved.
・ Interaction model between molten materials and structures/coolant was developed andits function was verified.
(c) Models of eutectic reaction and oxidation reaction at hightemperatures were improved.
・ Models of B4C and ion oxidation reaction was added.
・ Models of B4C and ion eutectic reaction was developed
and their function were verified.
(d) Model of interaction between molten materials andstructures at lower part of RPV・Interaction model was developed and its function wasverified.

(c)
(b)

(d)
(a)
©International Research Institute for Nuclear Decommissioning

63

Accident analysis at Unit 1 by MAAP code
Event

Time of occurrence

Reactor water level ≦ Top of active fuel2011/3/11 around 17:50

Core damage (fuel rod maximum temperature≧1,200C)

2011/3/11 around 18:40

Core melting (fuel rod maximum
temperature≧2,200C)

2011/3/11 around 18:50

RPV damage

2011/3/12 around 10:00

Quantity of fuel debris (t)On pedestal floor: 91 tons
Transfer of moltenmaterials to lower
plenumCore region

On dry well floor: 89 tons

RPV damage
Pedestal floor

Dry well floor
Lower plenum(Note) Quantity of fuel debris at around 13:00,
March 12, 2011 (fuel and structures)©International Research Institute for Nuclear Decommissioning

64

Accident analysis at Unit 1 by SAMPSON code
Event

Time of occurrence

Reactor water level Top of active fuel2011/3/11 around 17:50

Core damage (fuel rod maximum temperature≧1,200C)

2011/3/11 around 19:40

Core melting (fuel rod maximumt
temperature≧2,200C)

2011/3/11 around 22:00

RPV damage

2011/3/11 around 22:10

Quantity of fuel debris (ton)Reactor core
instrumentation
tube damaged

Fuel debris transfer to lower plenum

Reactor core (including coreplate and shroud)
Pedestal floor

RPV damage

Reactor core:72 tons

Pedestal floor: 88 tons

UO2, U3O8
Zr, ZrO2
Fe, Fe2O3
B4C

33.6 wt%
34.2 wt%
31.3 wt%
0.9 wt%Lower plenum(Note) Quantity of debris at around 13:00, March 12, 2011 (fueland structures)©International Research Institute for Nuclear Decommissioning

65International cooperation: OECD-NEA BSAF project
Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF)
Period: November 2012 to October 2014
Objectives: (1) To obtain information and opinions on accident progression and reactor situations by gathering knowledge and opinionsfrom experts(2) To contribute to advancement of analysis techniques and codes
Official in charge ofinternational cooperation
at Agency of Natural
Energy and Resources

Operation organization:
JAEA(Cooperation)

TEPCO
Institute of Applied Energy
Toshiba
Hitachi GE

Participants:
Japan: JAEA, JNES, Central Research Institute of Electric Power Industry, Institute of Applied Energy
USA: USNRC and others France: IRSN, CEA
Germany: GRS
South Korea: KAERI
Russia: ROSATOM, IBRAE Switzerland: PSI
Spain: CSN

Information sharing through international project conference* and website (https://fdada.info/index)* ・Preparatory conference: June 18-20, 2012 in Paris (NEA headquarters)

・First conference: November 6-8, 2012 in Tokyo
・Second conference: October 15-17, 2013 in Paris (NEA headquarters)
・Third conference: June 2014 in Tokyo・Fourth conference (final): October 2014 in Paris (NEA headquarters)©International Research Institute for Nuclear Decommissioning66

2. Research and development related topreparations for retrieving fuel debris

Development of technologies for detecting fueldebris in Reactor(2-(2)-2)©International Research Institute for Nuclear Decommissioning

67

Overview of project● Identifying position and quantity of fuel debris in
RPV (1) and lower part of PCV (2) is important.
 Damage to fuel assemblies
 Inflow of molten fuel to narrow space Detailed distribution such as density of fuel debris
 Specific retrieval procedures and techniques

High dose rate inside
RPV makes difficult toaccess.

Early supply of fuel debris
distribution information to IRID
by using see-throughtechnology using muon

(1) Inside RPV(2) Lower part of
PCV

Object to beobserved: Position of fuel debris©International Research Institute for Nuclear Decommissioning

68

Overview of project (muon observation technology)
Permeation method

Scattering method
Muon

Muon

Muon detector
Incidence
angle

Muon
detector
Object to bemeasured

Object to bemeasured

Permeation ratio is measured.

Emitting
angle: Detecttion element

Scattering angle is measured.

Presence of matters in coming direction(two-dimensional)

Presence of matters at scattering position(three-dimensional)

Identifying ability (fuel debris): About 1 m

Identifying ability (fuel debris) ： About 30 cm

One small-size muon detector (applicable early)

Two large-size muon detectors(need to be developed)

Can identify heavy elements such as uranium.
©International Research Institute for Nuclear Decommissioning69

Outline of past result (muon permeation method)
Results of measurement with equipment in the same scale at Tokai No. 2 nuclear power plantof the Japan Atomic Power Company:
・No fuel in nuclear reactor
・Fuel existed in spent fuel poolMark-II PCV
and reactor inbuilding

Reactorbuilding

Detector 2-2

Detector 3

Detector 2-1

Detector 2-1

Measurement positions at Tokai No. 2
nuclear power plant:(3 points)

Detector 2-2

Detector 3

Measurement result at each point
High-density substance existed outside bell-shaped PCV
KEK(High Energy Acceleration
Research
Organization)data©International
Research Institutefor Nuclear Decommissioning70

Past result (muon scattering method 1/3)
・Uranium identifying test in research reactor in scale of 1/10 of Fukushima Daiichi
nuclear power plant
・Simulation and experiment value matched within 3%.
・Aimed at achieving identifying ability of 30 to 40 cm at Fukushima Daiichi
Cylindrical
reactor core
Fuel rod 2
Fuel rod 1

Muon detector 2

Fuel rod 1

Thin fuel rod
confirmed at left

Muon
detector 1

Fuel rod 2

Muon
detector 2

Concrete

Ion blockThin fuel rod
confirmed at right

Measurement systemMuon
detector 1
Obstacles place in front wereconfirmed.

Image of each cross section©International
measuredResearchthree-dimensionally
Institute for Nuclear Decommissioning
TOSHIBA Data

71

Past result (muon scattering method 2/3)

Fukushima Daiichi Unit 2

Muon detector 2: 7  14ｍ2

Muon detector 1: 1 4 7ｍ2

Numerical simulation was conducted in actual scale of Unit 2.

TOSHIBA Data

©International Research Institute for Nuclear Decommissioning72

Past result (muon scattering method 3/3)
(Result of numerical simulation)
Simulation of reactor core molten 0 to 100%, measured for 90 days
0%

10%

30%

50%

70%

100%

R=10cm

R=15cm

R=20cmFuel debris

Simulation of reactor core molten 50%, measured for 10 to 150 days
10 days

20 days

30 days

60 days

120 days

150 days©International Research Institute for Nuclear Decommissioning

TOSHIBA Data73

2. Research and development related to preparations
for retrieval of fuel debris
Development of technologies for fuel debris
containing/transfer/storage (2-(1)-7)©International Research Institute for Nuclear Decommissioning74

Overall plan (objective and goal)[Purposes of development of technologies for fuel debris packing, transfer, storage ]1F has the following differences in condition fromTMI-2:
 Fuel debris of 1F-1 to 3 exists at lower portion of RPV and in PCV. Its position and property are unknown.
 Dose rate is high in the building and make access difficult.
 Fuel is higher in burn-up and concentration than that of TMI-2 and difficult for fuel debris packing, transfer,storage.
 There is a concern that corrosion has progressed by injection of seawater into reactor.
Exclusive canisters were developed for TMI-2, which were used for fuel debris packing, transfer, storage. For fueldebris packing, a work block was installed on top of RPV and fuel debris was packed in canister in RPV.
This research is for development of technologies to create fuel debris canisters suitable for the situation of 1F andhandle the canisters, by referring to the records of TMI-2.

[Goals of development]
Fuel debris must be safely retrieval while keeping it well below critical point, prevented from scattering,appropriately shielded, cooled, and stored in accordance with the 1F execution plan, and safety of workers andinside and outside the site must be ensured.
 Deciding specifications required for canisters suitable for 1F fuel debris packing, transfer, and storage, andselecting their materials
 Development of technologies for assessing the safety of the fuel debris canister Design and production of canister for a mock-up test for retrieval of fuel debris©International Research Institute for Nuclear Decommissioning75

Research on transfer and storage of damaged fuel (1)Fuel canister
handling crane(shielded)

Fuel canister handlingcrane (shielded)

Packing fuel debris
Closing cover
Plugging(Overview of fuel debris handling at TMI-2)

Fuel canister
handling
crane(shielded)

Decontaminati
on of fuelcanister surface
by spraying
boron water

INL work

TMI-2 work

Transportation cask

Draining water
Injecting inert gas
Plugging

Drain/vent hole

Attaching venttube
Injectingdesalinatedwater

BufferFuel canisterO-ringWork block(Level allowing drainage)

RecoverytankFule canister

Vent tubes

Fuel canisters

Canisters
Shielding coverTransportation cask

Fuel canisters

Transportation cask
Dual container(Sealing canister was notexpected but dual can was
required.)

DebrisFuel canisters

Fuel canister(100Ｗ/can)Reactor vessel

Fuel pool
Fuel canalPacking debris (in water, remote
Loading fuel canister to
Taking out fuel canister
Moving fuel canister Making fuel canister
1979: Accident occurred.
cask(in air, remote)(inair,
remote)(in water, remote) (in water, remote) Moving fuel canister
1985-1989: Fuel taken out(in air, remote)

Fuel container for heating andvacuum drying

Transporting out of site
(Railroad & vehicle)

Taking out canister (in air, remote) & wet storage
1985-2001: Wet storage

Cover
Vacuum drier

Fuel container forheating and vacuum
dryingCanister

FilterHorizontal silo

Canister cover

CanisterCanister coverTransported
Vent tube

Canisters(70Ｗ/can)Sampling hole

Fuel container
for heating andvacuum drying

Fuel canisters
Canister

Fuel canisters
Fuel canisters

Retrieval of fuel canister(in air, remote)1999-2001: Taking out

Fuel canisters

Fuel canisters
Fuel canisters

Heating & vacuum drying

Loading fuel canister to canister(in air, remote)Welding primary & secondary
coversDry storage (horizontal silo)
1999: Dry storage started©International Research Institute for Nuclear Decommissioning76

Research on transfer and storage of damaged fuel (2)Outline of result research on transfer and storage of damaged fuelExample at TMI-2(fuel debris storage)

Example at Paks

Transfer of damaged
fuel in France

Objectcontained

Fuel debris(molten fuel)Same as left

Damage tocladding(fuel not molten)

Pin-hole leakWet ordry

Semi-dry (Only water was
drained and drying
processing was not
performed.)

Dry

Wet

Dry

Fuel
canister

Used

Used

Used

Not used (Fuel debris
transported in same way
as sound fuel)

Fuel canister wasconstructed so that hydrogengenerated is always vented
through screen filter on coverwhile can was in storage.
Cask for transporting can issealed and generation ofhydrogen was controlled byhydrogen recombination
medium.

Fuel canister had no
sealing function.
Hydrogen generated
was vented through
HEPA filter attached tocanister.

Fuel canister wasalways vented
through cover
having ventmechanism.

Canister was not used.
Cask was selected byanalyzing andassessing hydrogenconcentration duringtransportation.
Hydrogen concentration
was measured before
transportation.

Shielding was secured byperipheral equipment.
Thickness of canister wasdetermined by factors suchas structure.

Same as left

Same as left

Secured by cask

Quantity of fuel debris to be
stored in fuel canister wasconservatively decided by
optimum deceleration (debris
of one fuel assembly or lesswas stored in can. Operation
is managed in same way.)

Criticality wasassessed with 12 fuelcanisters storing new
fuel of maximum
concentration to full
capacity loaded tocanister.

Conservative
setting of optimum
deceleration incanisterSame as sound fuelAssessed by maximum
quantity of debris stored in
one fuel canister

Same as left

Only heat removal
capacity wasdescribed (details
unknown).

ShieldingCriticality
Heatremoval

Basic safety design

Containment

Example at TMI-2(fuel debris transfer)

Assessed with soundfuel

Example of fuel debris canister used at TMI-2, USA(reference)
Source: DOE/SNF/REP-084 TMI Fuel Characteristics for
Disposal Criticality Analysis (2013)

Fuel debris storage system at TMI-2, USA (reference)
Source: Andrew P. Szilagyi, Three Mile Island Unit 2
Overview and Management Issues, OECD-Nuclear Energy
Agency – 12th Meeting of the WPDD, France (2011i)©International Research Institute for Nuclear Decommissioning

77

Study of storage system
Storagemethod

Wet

Dry

Pool

Metal cask

Concrete shielding (including horizontal silo)

Vault

Overview

From HP of Recyclable Fuel Storage Company(2) Concept of concrete cask (U.S.A.: VSC24)

Figure 13. Concept of horizontal silo and concrete cask
[Source] Takuya Yanagishita, Tosiari Saegusa, Chihiro Ito, Shinichi Komatsu, Eiji Yoshimura: Overseas
Trend of Spent Fuel Intermediate Storage Technology, Atomic Energy Society of Japan magazine, Vol.
42., No. 11, 72 (2000)

Figure 8. Metal cask to be used at Recycle Fuel Storage Center[Source]

Merits & demerits

Ensuring safetyfunction

Sealing

・Metal gasket of primary and secondary
coversScreening ・Pool water and reactor building
・Metal cask (combination of steel andneutron screening materials) and building
Criticality
・Geometrical arrangement of fuel rack (fuel ・Geometrical arrangement of metal cask
prevention
rack materials as necessary)basket (basket materials as necessary)
Heat・Circulative cooling by pool water・Natural cooling of metal cask surface
removalTransporta ・Object to be stored must be stored in・Transportation and storage cask wastiontransportation cask.
developed and cask can be transported
as is.
Easy tocheck

・Pool water and reactor building

・Object to be stored can be easily takenout and checked for status because it isstored in water as is.

Needed・Storage density is high and installationarea
area may be small.
Expandabil ・New building must be constructed toityincrease capacity, needing large-scale
operation.
Maintenan ・High cost as cooling function and waterce costquality management are needed
Laws &
・There are domestic records and laws andregulations
regulations are established.

・Because object to be stored is stored incask, object must be checked in pool or
hot cell, which is more disadvantageousthan pool.
・Storage density is lower than pool.
・Cask, which accounts for large part ofcost, can be increased, so that system
can be gradually expanded.
・Low cost as only maintenance of electric
system is needed・Same as left

Jenny Morris, Contingency Options for the
Dry Storage of Magnox Spent Fuel in the
UK, ICEM 09/DECOM(2009)

・Welded primary and secondary covers ofcanister・Concrete cask (combination of steel andconcrete)・Geometrical arrangement of canister andbasket (basket materials as necessary)・Natural cooling of canister surface

・Same as left

・Canister must be moved from storage cask
to transportation cask (transportation andstorage canister may be used in somecases).
・Object to be stored is stored in canister
sealed by welding and must be checked inpool.

・Canister must be put in transportation
cask.

・Storage density is lower than metal cask
(installation area for storage is large).
・Same as left

・Storage density is higher than cask.

・Same as left
・There are overseas records but notdomestic records, and laws andregulations are not established.

・Building (concrete)・Same as left・Same as left・Same as left

・New building must be constructed toincrease capacity, needing large-scale
operation.
・Same as left・Same as left

©International Research Institute for Nuclear Decommissioning78

2. Research and development related to preparations for
retrieval of fuel debris

Understanding of characteristics by using simulated fueldebris and development of fuel debris processing
technologies (2-(3)-1 & 3)

©International Research Institute for Nuclear Decommissioning79Outline of fuel debris characteristics study
Management technology
study(storage, disposal,
processing)

Fuel Debris
management
technology

Management offuel debris

Study ofmaterial
accountancy

Temporary storage

Study of storage
method andcanister
Fuel Debris
characteristics
information

Criticality
safetymanagementProject for development of technologies toanalyze and mange fuel debris characteristics

Study of method
and tool for debris
retrieval

Retrieval of fuel debris

RPV(Lightly
damaged fuel)Core spray
pipingFeed water
sparger
Fuel canister
Core spray sparger

Core shroud
Spade basket

Top guide
BlockerCore support
plate

Jet pump(Fuel debris)

Loose debris
Molten solidified
matterShroud support ring

Stump-shaped fuelShroud support
Neutronmeasurement guidetube

Control rod driving
mechanism housing
Neutronmeasurement
housing, etc.

1F In-core status (assumption)

Lower head debris(Example of retrieval jig)

(MCCI product)
Fuel Debris Retrieval of TMI-2©International Research Institute for Nuclear Decommissioning

80

Purpose of analyzing fuel debris characteristics(Reflecting physical property information on other projects)
Development of debris retrieval system)

・Study of influences on cutting and drilling
・Reflecting information of physical property
data on cold simulation specimen

Storage &preservation
project

Criticality
control
project

Material
Accountancy
project<Analyzing characteristics with simulated debris>

Fuel Debris retrieval PJ(Physical property data)・Mechanical characteristics: Hardness, elasticity,
fracture toughness, etc.
・Thermal characteristics: Melting point, thermal
conductivity, specific heat, etc.
・Others: Particle diameter, shape, porosity, density,
chemical form, etc.

・Study of debris retrieval method
・Development of retrieval jig, etc.

Gathering physical property information oneach part(Image of drilling element test)

Single phase specimenorMetallic compound (SUSZr-U alloy, ZrB2, etc.)

Oxide (U,Zr)O2, etc.)
Crystal of
boridePellet of notmolten fuelCold specimen

Specimen of composite
materials
Study of cold simulation specimen

Compoundspecimen in
Kazakhstan also
used forreference

Structurematerial(metal)Example of actual fuel debris
sample (image)

Reactant withconcrete (silicatecompound, etc.)©International Research Institute for Nuclear Decommissioning81

Purpose of development of fuel debris management technology
Storage/preservation
収納・保管PjPj

Fuel Debris Retrieval Pj
デブリ取出しPj
Retrieval
取出In-water storage/transfer
水中収納/移送

Reacto炉心r core• Denaturation•空冷時の変性during cooling•含水・浸出••温度・ガス
Water content,
infiltration• Temperature,gas

Preprocessing
前処理
Crushing/
粉砕/

収納容器
Canister

乾燥等
drying,etc.

Tasks assumed
移送・保管にduring transfer
関して想定さandstorage

れる課題• Element•元素浸出
infiltration
••液性
Fluid
• Watertemperature
•水温上昇
rise

• Degree of•必要な前処necessary
preprocessing
理の程度❐ 収納・保管pjでの懸念事項➔ニーズ
Concerns
about storage/preservation Pj  Needs
Item
Quantity of water absorbed項
to fuel目debris

燃料デブリへの水の吸収量
Quantity(speed) of water released when debris is driedat room temperature
常温で乾燥させた場合の水の放出量（速度）Quantity (speed) of water released when debris isheated and dried
加熱して乾燥させた場合の水の放出量（速度）Quantity(speed) of radioactive substances flowing out
when
debris is submerged
浸水時の放射性物質の溶出量（速度）Quantity (speed) of halogen element flowing out when
debrisis submerged
浸水時のハロゲン元素の溶出量（速度）Quantity of radioactive substances released when debris
大気中保管時の放射性物質の放出量
is stored
in atmosphere
Quantity of radioactive substances released when debris
真空中の放射性物質の放出量
is stored
in vacuum

Category
Water
content,
カテゴリー
dehydration
含水･脱水
Water
content,
dehydration
含水･脱水
Water content,
dehydration
含水･脱水
Denaturation(inwater)
変性(水中)
Denaturation (inwater)
変性(水中)
Denaturation(in air)
変性(気中)
Denaturation(in air)
変性(気中)

Needs 適宜、現場作業や他
to be reflected asnecessary depending on
Pjの進捗に応じてprogressof work at site
ニーズを反映してゆく。
and other
projects

Midto
中・長期
long-term
保管storage••発熱
Heating• Pulverization••粉体化
Generation of gas
as H2•Hsuch等ガス発生
2

Needs generated
during
収納・保管Pjで生storage/preservatio
じるニーズを随時
n Pj to be reflected
反映してゆく。
as necessary

Debris Pj (2014)
デブリPj（H26年度）[Details of study]【検討内容】• Assessment of factors influencing on tasks and concerns••Study課題・懸念事項に影響する因子の評価
of assessment method (samples, parameters, assessment•items)評価方法の検討（試料、パラメータ、評価項目）••Examining
known document information and similar examples in
既知の文献情報、一般産業界の類似例の調査
general industries【開発項目】・
・・現状で想定される開発課題[Development
item]… Problems of development anticipated at➡
含水・脱水挙動評価
present
 Assessmentof water content and dehydration behavior（形状既知の試料による評価）(Assessmentwith sample whose shape is known)➡ 水中での浸水・特性評価
 Assessment of characteristics when submerged in water（使用済燃料の浸水試験等の結果参照、あるい(Referring to result of soaking test of spent fuel, etc., or hot test)
はホット試験） Denaturation
test in gas phase➡ 気相中での変性試験©International Research Institute for Nuclear Decommissioning82

(2-(3)-1 & 3) Analysis of characteristics by using simulated fuel debris and development offuel debris processing technologies
Data on fuel debris, such as the hardness, were gathered by producing a material simulating the characteristics of actual debris (simulated fuel debris)to estimate the actual debris characteristics so that technologies for taking out fuel debris can be considered. In addition, to study a scenario tomanage the fuel debris after retrieval, applicability and technical issues of the existing fuel processing technologies were removed and alternatives
available were compared to clarify gains and losses.

Actions
Analysis of debris characteristics (2-(3)-1)(1) Study of physical properties necessary for developing technologies for fuel debris retrieval
・Influences of various simulated materials, such as machinability and hardness, were analyzed.
・Mechanical characteristics of simulated fuel debris containing (U, Zr)O2 in a high Zr region and Fe were measured and each chemical form of
the measured values was reflected on estimation of physical property distribution, on the assumption that metallic materials had mixed inreactor.
(2) Analysis of reaction specific to accident at 1F
・Possibility of generation of alloy phase and boride through reaction with control materials was confirmed. Trend that oxide (glassy) andalloy layer separate in reaction (MCCI) with concrete was also confirmed. The hardest substance was estimated to be boride.
・Influences of Gd contained in some fuel on the thermal characteristics of oxide simulated debris ( (U, Zr)O2) and their range were confirmed.
(3) Estimating actual debris characteristics
・From the above results, a list of characteristics of fuel debris (preliminary version) was created.
Development of fuel debris management technologies (2-(3)-3)(1) Deciding technical requirements toward creating fuel debris management scenario
・Alternatives of scenario for managing fuel debris retrieval that out was compared to clarify gains and losses.
・Applicability of the existing spent fuel transportation canister was evaluated. It was judged that water content of fuel debris, which
influences storage, was important.
(2) Study of element technologies needed for analyzing fuel debris
・Basic data of fusion process, preprocessing technology of analysis, were collected from each type of simulated fuel debris containing MCCI
product.
(3) Study of applicability of existing fuel processing technology
・Basic data on applicability of simulated fuel debris to wet and dry processing were gathered.
©International Research Institute for Nuclear Decommissioning83

(2-(3)-1) Analysis of characteristics by using simulated fuel debris (how toproceed with research and development)
Basic policy● Estimating chemical form of fuel debris in reactor, reflecting theresults of the latest plant data at 1F and event progression
analysis● Reviewing items to meet needs by exchanging information withsite and fuel debris retrieval project (same applies to (2) to (5))● Reviewing necessary items in light of needs of other projects
under adjustment (storage/preservation, criticality safety,
measurement management)● Reflecting information and opinions from overseas (SA research,
TMI-2, etc.)● Studying physical properties heavily influencing necessary
methods (such as for drilling) and selecting physical properties
that serve as criteria for selecting simulated fuel debris suitable
for methods● Collecting mechanical property data from major materials offuel debris, by using simulated fuel debris● Gathering wide range of data including composition andinfluences of impurities● Collecting wide range of data in advance, taking various
possibilities into consideration, as many data are needed inearly stage● Assessing influences of reaction specific to 1F by using simulated
fuel debris● Collecting wide range of data, taking various possibilities into
consideration, as analysis of situations in reactor proceeds inparallel

How to proceed with research(1) Examining and organizing TMI-2 information,
SA research information, and Fukushima
information
Selection of fuel debris whose characteristics
are to be analyzed(2) Study of physical property ofdebris needed to be sampled andretrieval
Characteristics of fuel debris influencing eachn method

Study of items of fuel debris
characteristics list (database)Simulated fuel debris characteristics data

Fuel debris characteristics list format(3) Analysis of reaction specific to 1F accident
Data on fluctuation due to reaction specific
to 1F

Estimating and accessing fueldebris characteristics

Fuel debris characteristics
estimating method(4) Analysis of property of MCCI product
Data on characteristics of MCCI product

Estimated fuel debris
characteristics value● Getting information on MCCI phenomenon and products byeffectively using information and opinions from overseas andinternational cooperation(5) Comparison with TMI-2 debris● Obtaining physical property data by using fuel debris from TMI-2

Physical property data of TMI-2 debris

Creating fuel debris characteristics
list (database)Comparison of characteristics ofgeneral materials

Fuel debris characteristics list●

Estimating fuel debris characteristics based on data andinformation gathered to create a characteristics list

● Proposing simulated fuel debris materials for each method andtechnology by selecting general materials with similar
characteristics that have heavy influence

Proposal of simulated fuel debris ©International Research Institute for Nuclear Decommissioning
material for each method andReflecting on
84debris retrieval Pj, etc.
technology

(2-(3)-1) Analysis of characteristics by using simulated fuel debris (Result of study in FY 2013)
Fe-Cr-Ni(Fe,Cr,Ni)2B

ZrB2（Foil-like or (Fe,Cr,Ni)2(Zr,U)flat crystal structure) (Fe2ZrType Cubic）

Arc melting

Oxidative atmosphere(annealing)（ Ar-0.1%O2,1500℃）

Fe-Cr-Ni(Zr,U)O2 (Zr-rich)（Alloy andZrB2Oxidized）

Reaction with control material（ B4 C+SUS) （Example of the fusion solidification abstaining from side observation image（ Obtained knowledge regarding the composition of solidified material generated when control rod and molten fuel ）Vickers hardness of each formed aspect

BorideGlassinessOxidizer(Sintering)(Sintering,15-75%U)

Metal
Vickers hardness（Estimate hardness distribution for each chemical system of fuel debris (boride, oxide, metal）©International Research Institute for Nuclear Decommissioning85

(2-(3)-1) Analysis of characteristics by using simulated fuel debris (plan for 2014)- Production and assessment of characteristics of metal/ceramic molten solidified substance  To test and assess the characteristics of materials for producing a mock-up for developing equipment and decides a method to
produce a mock-up, by creating metal/ceramic uneven molten solidified substance considered difficult to be processed by retrieval
equipment[Major result] Macro/micro observation and chemical analysis of boundaries between metal and ceramic, and gathering data such as on
hardness and fracture toughness

Background

MethodTMI-2: Formation of crust layer of solidified metal/ceramic[1,2] Difficult crushing and cutting due to compound of metal with hightoughness and hard but fragile oxide
 At 1F, it is estimated that molten fuel fell on metal structures,
partially melting them, and then solidified and generated
compound mixing oxide and metal
 It is necessary to analyze the characteristics of the unevencompound.
Debris distribution at TMI-2
Crust layer- Metal + ceramic unevenly
solidified

Air conditioner

Upper debris layer
Crust layer
Molten solidified layer
Fuel layer keeping shape

Lower debris layer

Metal part
Cross-sectional
structure

UO2(1) Metal/ceramic molten solidified substance production
test
Falling molten UO2＋Zr compound (60 kg) onto
stainless structure.
Producing molten solidified substance of metal andceramic unevenly and rapidly cooled(2) Material assessment
Conducting macro/micro observation of cross section,
chemical component analysis, and measurement ofhardness and fracture toughness by focusing on
boundaries between metal and ceramic and theirperimeter(3) Designing mock-up of metal/ceramic molten solidified
substance
Deciding on production specifications by selecting
materials for and designing mock-up for developing
equipment
Setting production conditions from result of estimating
debris characteristics at 1F based on existing informationand opinions[1] Research activity report by Japanese Committee of US-Japan WR Research (March 1991)[2] Tokai et al., Nuclear Magazine, Vol.32 No.4 (1990)388.

Reflection on retrieval
fuel debris

UO2+Zr

Stainless steel
plate
Stainlesssteel saucer

Specimen©International Research Institute for Nuclear Decommissioning862. Research and development related to preparations
for fuel debris retrieval
Development of the criticality management of fueldebris(2-(1)-9)©International Research Institute for Nuclear Decommissioning

87

Overall plan (objective and goal)[objective of development of criticality control technology]
It is believed that fuel debris has not reached criticality at present but technologies forcontrolling and monitoring criticality will be developed so as to prevent re-criticality even if thefuel debris shape or water quantity changes while fuel is being retrieved.
[Goal of development of criticality control
technology]
Fuel debris is considered to exist in PCV via
RPV. Development of technologies to access andmonitor criticality and neutron absorbent toprevent re-criticality when fuel debris isremoved will be aimed.

Reactorcore
炉心部
RPV RPV下部
lower portion

PCVペデスタル
PCV pedestal
PCV drywell
PCVドライウェル
Cooled liquid waste
冷却廃液処理設備
processing facility

Roadmap: Following technologies will be developed with an eye toretrieve fuel debris in the first half of 2020:(1) Criticality assessment technology(2) Liquid waste/cooling facility sub-criticality control technologies(3) Reactor re-criticality detection technology(4) Criticality prevention technology

Assumed position of fuel debris©International Research Institute for Nuclear Decommissioning88Outline of monitoring sub-criticality and detecting re-criticality
Gas processing FP
gamma detection
systemXe, KrReactor re-criticality
detection system using
neutron

R/B

PCV gasprocessing
systemReactor re-criticality
detection systemusing neutron
Suction fan

Processbuilding
T/B Workers who maintain facilities for treating liquid
waste and cooling outside PCV should be protectedfrom risk of exposure due to criticality.
 sub-criticality monitoring

Detector
Criticality

Staying water
Present

 Since requirements for controlling criticality differs
between inside and output PCV, sub-criticality
monitoring or re-criticality monitoring is used
depending on the purpose (before fuel debris is
retrieved).

FutureEquipment monitoring subcriticality of liquid waste treatment
and cooling facility

U, Pu

格
本設水処理

Water treatment
facility of this system
Criticality
臨界 Even if criticality is reached in
PCV/RPV, risk of exposure is
extremely low because radiation
does not directly reach the outerwall of PCV. On the other hand,
however, monitoring thesituation in a relatively wide
range is important.
 Re-criticality detection

設備

Neutron beam distribution incase criticality in PCV©International Research Institute for Nuclear Decommissioning89

Development of technologies to detect re-criticality in reactor
Objective of development: To develop a re-criticality detection system using neutron and FPγ ray
to detect a case where fuel debris has reached re-criticality in RPV or PCVResult of developing in 2013 [Re-criticality detection system using neutron]・Consideration of specifications of re-criticality detection system- Specifications of a system that detects neutron to be emitted when fuel debris reaches re-criticality in RPV or PCV were drawnup.
- Based on the result of analysis of neutron dose distribution inside and outside of PCV and on the assumption that the system
will be installed in PCV:・Detector was selected (10B proportional counter tube, fission ionization chamber for reactor).
・ System was designed, trial system was manufactured, and feasibility test was conducted.
- Feasibility was assessed through γ resistance test of 10B proportional counter tube and system test [by using a facility of
National Institute of Advanced Industrial Science and Technology].
Development was completed in 2013. The system was tested for its performance to detect re-criticality when PCV wasfilled with water, and was used for criticality management.
Result of development in 2013 [Re-criticality detection system using FPγ ray]・Kr-87/88, as well as Xe-135, could not be measured as a nuclide to be monitored, to shorten the detection time.
・Optimum system configuration was considered according to conditions because Kr-87/88 substantially differs in concentration
depending on the location where the system was to be installed.
・It was confirmed that Xe-135 (nuclide that had been measured) could also be measured while Kr-87/88 was being measured.
・Effect of simultaneous counting, a method to improve the sensitivity, was verified.
・Recommended system configuration was studied based on the estimated concentration of spontaneous fission at 1F-1.
・Effect of shortening the time to fill water was evaluated.
・Sub-criticality estimation method using a Kr/Xe ratio was studied.
・Including the system in the criticality control procedure was studied.

©International Research Institute for Nuclear Decommissioning90

Development of re-criticality prevention technology
Objective of development
・To develop soluble and insoluble neutron absorbent as a means to prevent re-criticality when fuel debris is
retrieved
Results of development in 2013 [Insoluble neutron absorbent]・Study of basic characteristics test items of candidates (absorbent, binder)・Trial production/procurement of candidate materials. Their applicability to fuel debris removal work wasassessed based on the basic characteristics. Candidate materials had no problems in mechanical and thermal
characteristics and were screened by solubility.
・Deciding on future development policy (planning test to confirm the resistance to radiation)
Second screening by conducting radiation resistance test. An application method was studied to secure
homogeneity when the material was applied and the result was reflected on the criticality control method.
Results of development in 2013 [Soluble neutron absorbent]
・Putting in order problems and making a study plan on soluble neutron absorbents (such as sodium
pentaborate)- Study of influences on corrosion resistance of materials in reactor , selection of contents to be added forgalvanic corrosion tests- Irradiation test (gamma ray irradiation test at JAEA Takasaki laboratory) was conducted to confirm
influences on hydrogen generation of radiation decomposition accompanying mixing of boron andchlorine.
- Influences on nuclide removal performance of liquid waste treatment methods and methods to separate
or recover neutron absorbent were studied.
In additional corrosion test was conducted to decide on a method of applying the soluble absorbent andthe result was reflected on the criticality control method. Influences on the fuel debris canister were
evaluated.
©International Research Institute for Nuclear Decommissioning91

Development of criticality prevention technology(example,Insoluble neutron absorbent)
Result of screening candidate materials by solubility
Category

Neutron absorbent candidate

Radiation
resistance test

PolicyB4C/sintered metal material

Done

Base metal selected according tocost as necessary

B or Gd-contained glass material

Done

Zinc borate with low solubility
selected

Hollow boron

Pending*

Solubility needs to be improved.
(Candidate material for specificmethod (underwater drift))

Gd2O3 particle

Done

Cement/ Gd2O3 particle

Done

Water glass/ Gd2O3 particle

Pending*

Solubility needs to be improved.

B4C gel material

Pending*

Solubility needs to be improved.

B4C/sintered metal material

Solid

Liquid → Solid

LiquidB/Gd-contained glass

10mm

Slurry/Gd2O3 particle
Slurry/ Gd2O3 particle

Done©International Research Institute for Nuclear Decommissioning

Pending*: Radiation resistance test to be conducted after characteristics are improved

Figure: Candidate
materials screened922. Research and development related topreparations for retrieval of fuel debris
Development of technologies for assessing structural
integrity of RPV/PCV(2-(1)-8)©International Research Institute for Nuclear Decommissioning

93

Overall plan (objective and goal)○ Quantitative data on corrosion rate will be collected for RPV ,PCV, RPV pedestal, and reactor water injection
pipe after the severe accident, in order to evaluate aseismatic strength, taking into consideration long-termwall thinning by corrosion.
○ Corrosion inhibition to secure the integrity of structural over long time will be studied and its effect will be
confirmed. Applicability of the method to the real plant will be assessed. to contribute to keep the structural
integrity components.
RPV stabilizer
RPV stabilizer
RPV lower mirror
Lined used to inject water

Support skirt
Reactor shielding wall

Lined used to inject water

Piping that serves as PCV
boundary when PCV isfloodedD/W shell

Piping that serves as PCV
boundary when PCV isfloodedMounting bolt
RPV pedestal

Ring garter

From CS system

PCV penetration
PCV penetration＊
bellows*

From HPCI
systemVent tube
From FDW
systemS/C shell
From RHR system＊
Vent tube bellows*

Column support

Anti-seismic support：RPV-related facility：PCV-related facility* Because these are made of stainless steel and may already be
cracked by SCC, they will not be evaluated and a method to
exclude them from repair or flooding boundary is under study.

Line used to inject water at
the accident

Bold line range: Reactor
water injection pipe

Example of portions to be checked for structural integrity (remaining life assessment)©International Research Institute for Nuclear Decommissioning

94

Overall structural integrity assessment flow
Building behavior analysis*(building damage simulation)(a) PCV flooding status (example)(assessed in 2012)

Model creation*Input condition proposedby Tepco.

Building-equipment
coordination behavior
analysis・Water level・Basis ground motion Ss
・Damage to equipment
・Influence of corrosion

Putting in order items to be checked
and results expected by conducting
examination when conditions
leading to improvement of accuracyof remaining life assessment, such as
damage of buildings and current
water level (coordination with otherprojects such as to inspect inside
PCV is needed.)

Assessment condition

Organizing loading conditions

・Allowable value・Corrosion wall

Equipment strength assessment

・High-temperaturestrength deterioration(test data)

Listing & updating list ofequipment with low
permissible value

Application of corrosion
inhibition(b) Plant condition assumed atpresent (example)

Study of scenario and risk
assessment in case equipmentswith low allowable value bedamagedNo effect(c) Plant condition assumed until
PCV flooding (example)***Stopping water (repair
with grout)

Inputting information to otherprojects (requesting study ofinternal investigation andrepair method**)**Study by experts in equipment design and structure isindispensable and current research and development
system should be reviewed.

: Done in 2012

: Assessed in 2013

thinning (test data)***Information input from other project “development of technologies for stopping water leaking fromreactor building and repairing lower portion of PCV”

: Conducted or to be conducted from 2012 to 2015

Outline flow of remaining life assessment (example)

Water level in PCV estimated from present
situation and until PCV is flooded (example)©International Research Institute for Nuclear Decommissioning95

Development of corrosion inhibision - Result of test to add sodium nitrite Purpose
To gather quantitative data by conducting corrosion tests by using sodium nitrite, which has a been used for coolant bearings at a real plant,
in order to check the effect of suppressing corrosion of PCV (carbon steel SGV480 for PCV). In addition, additive density consideredappropriate for application to the real plant will be studied based on the gathered data.

Sodium nitrite addition test matrix(50C, seawater diluted 200-fold, saturated atmosphere)
Sodium nitrite
density (ppm)

Gas-liquid
phase
environment

Density (1) (200)Liquid phase

Density (2) (400)Liquid phase

Density (3) (2000)

Liquid phase

Density (2) (400)Gas-liquidinterface

Test time

50h
●
●
●
●

100h

500h

2000h

●
●
●
●

● (1)

－
●
－
－

● (2)
●

● (3)

Without anti-rust agent(liquid phase)(1) 200 ppm added(liquid phase)

Gas
phase

Liquidphase(2) 400 ppm added(liquid phase)

Test result
・Corrosion in liquid phase was prevented when sodium nitrite was added 400 ppm or more.
・Corrosion in liquid phase was prevented on gas-liquid interface but corrosion in gas phase wasobserved.

(3) 400 ppm added(gas-liquid interface)Appearance ofspecimen after test
(example)©International Research Institute for Nuclear Decommissioning96

Policy for evaluating aseismatic strength
Strength of each unit of equipment was evaluated from the load of seismic response analysis in plantconditions (three cases at each plant) assumed to have a higher priority when methods of fuel debris
retrieval was studied.
Plant/
case

1F-1

1F-2

1F-3

○ Present assumption• Building damaged• D/W water level:
About 2.9 m• S/C: Full of water• Vent tube: Full ofwater• Vacuum destruction
tube: full of water• Torus room waterlevel: OP3680○ Present assumption• Building sound• D/W water level:
About 0.6 m•S/C: OP3100• Vent tube: Water
flowing at bottom• Torus room waterlevel: OP3200○ Present assumption(3 years after accident)• Building damaged• D/W water level:
About 6.5 m• S/C: Full of water• Vent tube: Full ofwater• Torus room waterlevel: OP3200

H25-2○ Torus room waterlevel control• Building damaged• Equipment on stairs ofoperation floor:
About 5,100 t• D/W water level:
About 2.9 m• S/C: Full of water• Vent tube: Full ofwater• Vacuum destruction
tube: Full of water• Torus room waterlevel: OP-300○ PCV partially flooded• Equipment on stairs ofoperation floor: About
5,500 t• Small room buried• D/W water level:
About 5 m
• S/C: Concrete OP1900• Vent tube: Repair
considered• Torus room waterlevel: OP-300○ PCV partially flooded• Equipment on stairs ofoperation floor:
About 3,900 t• Small room buried• D/W water level:
About 6.5 m• S/C: Concrete OP1900• Vent tube: Repair
considered• Torus room waterlevel: OP-300

H25-3○ S/C reinforced• Building damaged• Equipment on stairs ofoperation floor:
About 5,100 t• D/W water level:
About 0.3 m• S/C: Concrete OP3570• Vent tube: Air• Vacuum destruction
tube: Air• Torus room waterlevel: OP-300,
concrete OP-485○ PCV flooded• Equipment on stairs ofoperation floor:
About 5,500 t• Small room buried• D/W water level:
About 35 m
• S/C: Concrete OP1900• Vent tube: Repair
considered• Torus room waterlevel: OP-300○ PCV flooded• Equipment on stairs ofoperation floor:
About 3,900 t• Small room buried• D/W water level:
About 35 m
• S/C: Concrete OP1900• Vent tube: Repair
considered• Torus room waterlevel: OP-300

H25-1

Seismic response
analysis

Modeling building-equipment
coordinated analysis (example)©International Research Institute for Nuclear Decommissioning97

Result of evaluation of structural integrity of RPV structures (example)
RPV stabilizer

Support skirt (primary stress)(MPa)

Unit 1
Case
25-1
25-2
25-3

Lower mirror plate
RPV support skirt

Stress58
59
59

Unit 2
Permissible
value

360

Stress39
40
71

Unit 3

Permissible value

360

Permissible
valueStress
38
39
61

360

Support skirt (compression)
Unit 1

Case
25-1
25-2
25-3

Buckling*

0.188
0.192
0.191

Unit 2
Permissible
value

1

Buckling*

Unit 3

Permissible value

Buckling*

1

0.112
0.115
0.182

0.113
0.118
0.213

Permissible
value

1

*Left member of inequality expression of buckling: = αB(P/A)/fc +αB(M/Z)/fb (source: JEAC4601-2008)

Lower mirror plate (primary stress)
Unit 1

Case
25-1
25-2
25-3

RPV support skirt

Stress
195
227
227

(MPa)
Unit 2

Permissible
value

540

Stress105
110
104

Unit 3

Permissible value

540

Stress130133
124Permissible
value

540

*: Strength of stress in this table indicates a value multiplied by a coefficient in accordance with the stress strength
required for filing an application for factory license and taking the plant conditions into consideration.

Mounting bolt

Ring garter

Stress generated at all portions assessed was equal to or lower than the allowable stress.

Anchor bolt

Figure: Parts evaluated©International Research Institute for Nuclear Decommissioning98

Result of evaluation of structural integrity of PCV structures (example)
Evaluation result (Unit 1)
Conditions:• Seismic condition: Ss wave
• Temperature: 50C
• Service condition: Ds• Corrosion wall thinning: Seetables on the right.

Estimated thinning of Unit 1 dry well(single side) (mm)After accident
10 years
15 years
Estimated
2.15
2.44
thinning
Estimated thinning of Unit 1 suppression
chamber (single side) (mm)After accident

10 years

15 years

Estimatedthinning

1.09

1.36

(1)-1 Result of primary stress evaluation of D/W shell(part evaluated: Sand cushion) (thinning on both sides
considered)
StressPermissible
strengthvalue
Case
Tolerance
(MPa)
(MPa)
25-1
124
423
3.41

<Evaluated cases>
Case 25-1: Equivalent to corrosion after 10 years
Case 25-2: Equivalent to corrosion after 15 years
Case 25-3: Equivalent to corrosion after 15 years(8) PCV stabilizer

Evaluation result:(7) Reactor shielding wall(2) Result of evaluation of primary stress on venttube (part evaluated: Joint between vent tube andD/W body(6) RPV pedestal(1) D/W shell

Case

25-1

Stressstrength(MPa)
117

Permissible
value
(MPa)
423

25-2

142

423

2.97

25-3

144

423

2.93

140

423

3.02

25-3

128

423

3.30

Note: Allowable stress (fc and fb) of column support
was evaluated with value F assumed to be value
Su at room temperature (value F was 1.2Sy atcompression).
(9) Result of evaluation of primary stress oncolumn support (part evaluated: Outer column
support)(4) PCV penetration(2) Vent tube

25-2(1) S/C shell

Tolerance
3.61

25-1

Combined stress(compression +bend)
Σc/fc + cσb/fb
1.083

25-2

0.946

1

1.05

25-3

2.354

1

0.42

Case

(9) Column support

Equipment evaluated (example: case 25-1)

Permissible
value

Tolerance

1

0.92

Generated stress of the parts evaluated as PCV boundary function was lower than the allowable stress but that of the suppression
chamber support structures was higher. Countermeasures for detailed evaluation and reinforcement (such as burying the torus
chamber with cement materials, etc.) will be studied in the future.
©International Research Institute for Nuclear Decommissioning99

Result of evaluation of structural integrity of RPV pedestal (example)
・Generated stress and distortion on all items evaluated were less than the evaluation criteria.
・At present, however, influences of corrosion by molten fuel debris are not taken into account andfurther study is needed, including how corrosion influences should be considered.

Evaluation result (Unit 1)

Maximum tensile distortion: 85 μ
Maximum :+2.550e-04
Facter:PART-1-1.25
A plane:1

Maximum :+8.458e-05
Facter:PART-1-1.1632
A plane:1

Minimum:-2.077e-4
Facter:PART-1-1.1
A plane:1

Minimum:-3.643e-5
Facter:PART-1-1.2114
A plane:1

Evaluation result (case H25-2)

Maximum tensile
distortion: 256 μ

Horizontal reinforcement
distortion distribution

Vertical reinforcement
distortion distribution
Maximum :+2.948e-01
Facter:PART-1-1.14
A plane:1

Maximum :+1.511e-05
Facter:PART-1-1.1615
A plane:1

Minimum:-1.621e-1
Facter:PART-1-1.34
A plane:1

Minimum:-2.882e-4
Facter:PART-1-1.24
A plane:1

Maximum compression
distortion: 289 μDistribution of major
distortion of concrete

Item

Generated
stress/distortion

Criteria*1

Concretedistortion289  10-6

3000  10-6

Reinforce
ment
distortion

256  10-6

5000  10-6

Antiplaneshear
stress354N/mm

1823N/mm*1: Criteria were calculated based on “Standards of
Nuclear Facilities for Power Generation – Concrete
Reactor PCV Standards JSME S NC1-2003” of Japan
Society of Mechanical Engineering.

Distribution of antiplane
shear stressMaximum antiplane shear stress: 0.30
N/mm2©International Research Institute for Nuclear Decommissioning100

Actions in 2014 to 2015
Flow of feasibility assessment of PCV flooding method(1) Preparing repair (reinforcement) measures in case tolerance was found not tobe secured by equipment assessment in2013

Seismic response analysis case conducted in 2013
Plant
/
case
○ Present
assumption• Building
damaged
• D/W waterlevel: About 2.9m• S/C: Full ofwater• Vent tube: Fullof water• Vacuumdestruction tube:
full of water• Torus room
water level:
OP3680○ Torus room
water level
control• Building
damaged• Equipment
on stairs ofoperation
floor: About
5,100 t• D/W waterlevel: About
2.9 m• S/C: Full ofwater• Vent tube:
Full of water• Vacuumdestruction
tube: Full ofwater• Torus room
water level:
OP-300[Equipment with no or lower tolerance]
Consultation with developer of PCV repair technology
・1F-1: Column support
・1F-2/3: Reactor shielding wall, column support, earthquake-resistance support(2) Creating and analyzing seismic response analysis model (building-equipment
coordination) with new plant information such as repair (reinforcement) and
fuel retrieval equipment added[Conditions of seismic response analysis]
・Two cases based on H25-2 and H25-3 of 2013
・Basic ground motion Ss (conventional Ss: Maximum acceleration of 600 Gal)・Reactor building seismic response analysis model: Model used in 2013 to be reused○ S/Creinforced• Building
damaged• Equipment
on stairs ofoperation
floor: About
5,100 t• D/W waterlevel: About
0.3 m• S/C: Concrete
OP3570• Vent tube:
Air• Vacuumdestruction
tube: Air• Torus room
water level:
OP-300,concrete OP485○ Present
assumption• Building sound• D/W waterlevel: About 0.6m•S/C: OP3100• Vent tube:
Water flowing atbottom• Torus room
water level:OP3200○ Present
assumption(3 years after
accident)• Building
damaged
• D/W waterlevel: About 6.5m• S/C: Full ofwater• Vent tube: Fullof water• Torus room
water level:OP3200○ PCV partiallyflooded• Equipment on
stairs ofoperation floor:
About 5,500 t• Small roomburied• D/W waterlevel: About 5 m• S/C: Concrete
OP1900• Vent tube:
Repairconsidered• Torus room
water level: OP○ PCV flooded
300• Equipment on
stairs ofoperation floor:
About 5,500 t• Small roomburied• D/W waterlevel: About 35 m
• S/C: Concrete
OP1900• Vent tube:
Repairconsidered• Torus room
water level: OP300○ PCV partiallyflooded• Equipment on
stairs ofoperation floor:
About 3,900 t• Small roomburied• D/W waterlevel: About 6.5m• S/C: Concrete
OP1900• Vent tube:
Repairconsidered• Torus room○
PCVlevel:floodedwaterOP•300Equipment on
stairs ofoperation floor:
About 3,900 t• Small roomburied• D/W waterlevel: About 35 m
• S/C: Concrete
OP1900• Vent tube:
Repairconsidered• Torus room
water level: OP300：RPV-related
facility：PCV-related
facility

RPV stabilizer(3) Evaluation of earthquake resistance of equipment under calculated loading
conditions[Equipment to be evaluated]
・Equipment to maintain functions of PCV flooding boundary and water-injection coolingequipment
・Excluding equipment to be repaired by development of PCV repair technology

RPV stabilizer

RPV lower mirror
Support skirt

Reactor shielding wall
D/W shell
Mounting bolt

RPV pedestal

Ring garter

PCV penetration
PCV penetration
bellows(4) Details of evaluation of equipment with low tolerance[Detailed evaluation method]
・Evaluation simulating real plant by finite-element method, etc.
・Criteria (setting of allowable stress, etc.): Criteria that allow achieving maintenance of thefunctions required of equipment, without emphasizing current codes & standards, will be
set up.

Vent tube

S/C shell
Vent tube bellows
Column support
Anti-seismic support

Equipment to be evaluated (example)©International Research Institute for Nuclear Decommissioning101

2.Research and development related to preparations
for retrieving fuel debris
Development of technologies for fuel debris
containing/transfer/storage (2-(1)-6)©International Research Institute for Nuclear Decommissioning

102

Explanatory CG for
Submersion Method for Fuel
Debris Retrieval©International Research Institute for Nuclear Decommissioning103

(2-(1)-6) Overview of development of technologies for retrieving fuel debris and coreinternals◎ Technical method study
Based on required information on the plant status,
development of related technologies, and investigation
results, etc.., to study total scenario.
→ Many information will be comprehensively evaluatedand unclear points should be assumed.
→ To be studied by the national project , including
existing technical investigation.

Cell, etc.
Work cart◎ Problems to development of technologies toretrieve fuel debris
Problems common to retrieval of fuel debris, regardless ofthe technical methods, are as follows:

SFP debris
canister

1) Cutting fuel debris
2) Remote operation
3) Prevention of expansion of contamination
4) Shielding5) Criticality prevention(1) Cutting fuel debris
1) Test to cut ceramic specimen
2) Production of specimen(2) Remote operation
1) Long arm control technology
2) Test models production of remote operated arm(including in cell)(3) Prevention of expansion of contamination
1) Selection of isolation film sheet

DSP
Long arm
Cutter, camera,
handling tools◎ Element test
Of the above problems, elements for cutting fuel debris,
remote operation, and prevention of expansion ofcontamination to be tested.

Isolation film sheet
Shielding

Remotely operated arm（

Image of Retrieval method©International Research Institute for Nuclear Decommissioning(2-(1)-6) Overview of development of technologies for retrieving fuel debris and coreinternals (2014 plan)
Major goal for 2014
To assess and plan a fuel debris retrieval technologies and plan research and development after examining and putting in order existing
technologies for retrieving fuel debris from RPV or PCV up to 35 m beneath from the operation floor. Development of device will be started if
possible.

Actions in 2014
1. Examining existing technologies
Examining and putting in order existing technologies (including equipment that experienced
at TMI) necessary for retrieving fuel debris while referring to existing catalogs
2. Planning fuel debris retrieval technique and development
Techniques for retrieval of fuel debris from RPV and PCV (such as a system to cool water
when fuel debris is retrieved, radiation shielding, radioactive substance dispersion prevention
measure, and method to contain fuel debris in canisters and transport the canisters, etc.) will
be assessed and a development plan will be made.
In making the plan, image processing systems and electronic devices will be evaluated,taking into consideration the fact that high radiation is emitted from fuel debris.
* Research and development will be performed by eagerly adopting proposals on usefultechnologies and techniques, by using a mechanism to invite proposals on retrieving fueldebris from both Japan and abroad.
* Techniques to retrieve fuel debris and core internals in water/air while taking into
consideration the assumed environmental conditions, including the present plant conditions
and the status of filling water at the upper part of PCV.
3. Development of equipment to retrieve fuel debris/element test
Device test and device development will be started, reflecting the result of survey of existing
technologies and development plan above. A test plan to process ceramic specimen
simulating fuel debris will be made with technologies used at TMI to process and cut fueldebris and core internals, and a test to process ceramic specimen will be conducted.

2014 (first half)

実施工程（平成２６年度）
Survey of existing technology

2014 (second half)
Arrangements

Finding problems

Makingdevelopment
plan

Assessing extraction technique

Makingdevelopment
plan

Employment of technologies at home and abroad as necessary

Making plan

Test

©International Research Institute for Nuclear DecommissioningRequest for provision of information on
alternative methods for fuel debris retrieval
from home and abroad©International Research Institute for Nuclear Decommissioning

106

Contents of RFI
Topic A: Internal PCV/RPV investigation

Topic B: Fuel debris retrieval

A-1:Conceptual study of method (following are samples)

B-1:Conceptual study of method (following are samples)

1. Method of inserting investigation device such as
cameras inside.
a. Utilize current throughbore such as piping/penetration.

b. Create new throughbore .

1.
2.
3.
4.

Access to fuel debris from the top of PCV underwater
Access to fuel debris from the top of PCV in the air*1
Access to fuel debris from the side of PCV in the air*1
Access to fuel debris from the bottom of PCV in the air*1*1 including partial submergencec. Methods of Shielding penetrations and of
equipment operation in terms of reduction of radiation
exposure.

B-2:Required technologies (following are samples)
1. Technology regarding fuel debris retrieval (cutting,suction).
2. Method of detecting fuel debris location by
2. Equipment/device such as remote control manipulator,
measurement outside, etc.
with superior control capability from long distance.
3. Technology of shielding against fuel debris withhigh radiation.
A-2:Required technologies (following are samples)
4. Device and equipment under the high radiation
1. Advanced measurement technology (camera,
environment.
dosimeter, thermometer etc. )5. Equipment/device to create a borehole on thea. High performance optical equipment(camera etc. ), otherbuilding concrete and PCV to access from the side
measurement technology (ultrasonic, laser etc.)
or bottom of the PCV.
b. Control technology of measuring instrument, and
6. Technology to store fuel debris safely in PCV/RPV
information transmission technology .
before retrieving.
2. Technology to detect whether the substance in thereactor is fuel debris or not.
©International Research Institute for Nuclear Decommissioning107

Results of RFI
About 60% of information was provided from Japan, and about 40% from overseas countries.

Total

JPN

US

Breakdown by country
GER
UK
FRA
BEL

A-1：Conceptual
study for the method.

33

20

7

3

-

2

-

1

-

A-2：Required
technology

58

32

6

10

6

2

2

-

-

B-1：Conceptual study
for the method

43

23

8

3

2

5

-

1

1

B-2：Required
technology

60

41

7

3

4

2

2

-

1

Total (No. of items
of information)

194

116

28

19

12

11

4

2

2

•Field of information for RFI

CAN

RUS

Topic A
Internal PCV/RPV
investigation
Topic B
Fuel debris
retrieval

About 60% of responses comes from inside Japanand 40% of responses from overseas©International Research Institute for Nuclear Decommissioning

108

Comparison of Submersion and Dry method
Submersion method Method to retrieve fuel debris in a submerged condition in which fueldebris is cut and stored under water.
Dry method Alternative method, in which fuel debris is cut and stored in air, or cut underwater
but stored in air, against the Submersion method.

Examplesbarrier

Submersion method

100t polar cranecontainer
Operation floor

Canistermanipulator

ＤＳP

SFP

・cut/store
equipment
・grabbingequipment etc.

barriermanipulator
・cut/store
equipment
・grabbingequipment etc.

RPV

PCV

Torus room

ＤＳP

SFP

Canister

RPV

PCV

Torus room

Fuel debris cut and stored under water
100t polar cranecontainer
Operation floor

barrierDry method

manipulator
・cut/store
equipment
・grabbingequipment etc.

100t polar cranecontainer
Operation floor

ＤＳP

SFP
RPV

PCV

Torus room

Fuel debris stored in air

Canister

Fuel debris cut and stored under water (PCV partial submersion)
100t polar cranecontainer
Operation floor

barriermanipulator

・cut/store
equipment
・grabbingequipment etc.

ＤＳP

SFP
RPV

Canister

PCV

Torus room

Fuel debris cut and stored in air©International Research Institute for Nuclear Decommissioning

109

Categorizing Provided Information
A: PCV/RPV internal investigation
A-1: Conceptual studyst

1 tierNo.
classification

2nd tier classification

From top (through new
boreholes)
From side (through2
existing penetrations)
From side (through new
3boreholes)
1

Placingequipment in
PCV/RPV

Measurement 4 From sidefrom outside
5 From bottomof PCV

Number of
responses

1
3

3
2
1

＊Category of topics may be different from the one
registered by applicant since IRID individually re-evaluates
the information. The number in the bracket is the one of
RFI submitted for Contaminated water issue, which is reevaluated from the viewpoint of RFI for fuel debris.

st

A-2: Support technology

1 tierNumber ofNo.
2nd tier classification
responses
classification
6 Radiation
8
Direct7 Thermal
1
measurement
8 Element analysis
2 (2)9 Camera10
10 Fiber scope
2
Direct11 Ultrasonic waves
6
observation
12 Laser scanner2
13 Others
1
14 Muon4
Indirect
15 X-rays, γ-rays and neutron3
measurement
16 AE method
1
Criticality control &17
5
exposure simulationWork2
environments 18 Water level
maintenance 19 Hot cell1
20 Lighting
1
21 Manipulator
5
22 Robot (underwater)
2
23 Robot (land)
6
Access24 Robot (amphibious)
6
technology
25 Cutting and boring tools
6
Radiation-resistant26©International Research Institute for Nuclear Decommissioning
10
components
110Categorizing Provided Information
B: Debris retrieval from PCV/RPV / C：Others
B-1: Conceptual studyst

1 tierNo.
2nd tier classification
classification
Submersion 27 From top
method28 From bottom
29 From top
From top - from side30
(combined approach)Dry method
31 From side
32 From bottom
33 Chemical method
Others34 Other than PCV/RPV
35 Others

Number of
responses

5
2
7

st

1 tierNo.
2nd tier classification
classification
36 Mechanical
Debris cutting 37 Thermal (plasma)
38 Thermal (laser)

7
7
4 (1)
3
2
2 (6)

C：Others

Debrisretrieval

39 Sorting
40 Category
Container (for temporary
41
storage)
42 Exposure management

43 ShieldingWork
environments 44 Decontamination ofcontainers
maintenance
45 Water stoppage
46 Water treatmentst

1 tierNumber ofNo.
2nd tier classification
responses
classification
Outside the
50 Outside the scope of RFI
7
scope of RFI

B-2: Support technology

Accesstechnology

Number of
responses

5 (2)
1
8 (1)

4
1
4
2
5 (1)
3
2 (2)
3

47 Manipulator

1248 Robot (for debris retrieval)

3 (1)49 Cutting and boring tools

1• Submersion method (both processes of cutting off fuel debris and storing into storage container are carried out underwater).
• Dry method (either or both processes of cutting and storing fuel debris is carried out in the air).
©International Research Institute for Nuclear Decommissioning

111

Example of RFP Responses-

Representative Method-©International Research Institute for Nuclear Decommissioning112

Internal Investigation
Shield plug

Barrier

Operation floor

Major issues
・Ensure work space
・Ensure boundaries

DS pitSpent fuel pool

Transport equipment

Guide tube

RPV

Viewing equipment
Monitoring equipment etc.

PCV
Torus room

Fig. Method to drill from the top (from shield plug)©International Research Institute for Nuclear Decommissioning113

Internal Investigation
Work platform

Shell wall

Barrier

Major issues
・Ensure boundaries
・Location ofdrilling/drilling
technology

DS pitSFP

Transport equipment

Operation floor
Guide tube

RPV

Existing openings etc

Viewing equipment
Monitoring equipment etc

PCV

Torus room

Fig. Method to drill from the top (from SFP)
©International Research Institute for Nuclear Decommissioning

114

Internal Investigation
Operation floor

DS pitSpent fuel pool

Major issues
・Access distance・Access method to RPV

Vent tube
Drilled opening/Existing opening

RPV

Barrier

PCV

Torus room

Guide tube

Viewing equipment
Monitoring equipment etc

Transport equipment

Fig. Method to survey lower PCV from vent©International Research Institute for Nuclear Decommissioning

115

Fuel Debris Retrieval
Operating device(Shielding plug)

100t class polar craneBarrierContainer

Ventilationequipment

Operation floor
Operating device(telescope etc)

Spent fuel pool

DS pitOperatingdeviceCutting device
Visual device
etc.

RPV

Ventilation equipment

PCV
約35m

Extension/contraction

Major issues
・Ensure boundaries
・Setting precision of therotating plug・Control of repulsive force
during cutting

Torus room

Research InstituteFig. Method to retrieve fuel debris in air by©Internationalrotating
plugfor Nuclear Decommissioning

116

Fuel debris retrieval
100t class polar craneOperating device(gate type crane etc. )Ventilationequipment

Container
Operation floor
Spent fuel pool

Barrier with shielding
function

DS pitCutting device
OperatingdeviceVisual device
etc.

Descening platform

RPV
Ventilation equipment

PCV

Major issues
・Ensure boundaries
・Radiation shielding
during operations

Torus room

Fig. Method to retrieve fuel debris in air by descending
work platform©International Research Institute for Nuclear Decommissioning
117

Fuel debris retrieval
100t class polar crane

Major issues
・Ensure boundaries
・Radiation shielding
during operations
・Location of entrance
opening

Container
Operation floor

DS pitSpent fuel pool

Existing equipment hatch/new
opening

Operating device
Cutting device
Visual device etc.

Barrier with shielding
function

RPV

Ventilationequipment

PCV

Torus room

Ventilationequipment
Operating device
Operating device

Access routeFig. Method to retrieve fuel debris in air from
theside©International
Research
Institute for Nuclear Decommissioning
118

RFP for Conceptual Study of Innovative Approach for
Fuel Debris Retrieval and Feasibility Study of Essential Technology
RFP has been processing by MRI since 27 June.
Fuel debris investigation team of IRID is performing technical support.

Description of the projects
Project 1: Conceptual Study ofInnovative Approach of Fuel Debris
Retrieval

To conduct conceptual study for the method to retrieve the fueldebris in the air in a safe and stable condition

To conduct feasibility study of technology to realize practical
Project 2: Feasibility Study of Visual application of compact and light-weighted visual equipment withand Measurement Technology forlighting function, and measurement equipment to characterize theInnovative Approach
properties of the objects both of which can be used under thevery high radiation environment
Project 3: Feasibility Study of Fuel
To conduct feasibility study of technology to realize practical
Debris Cutting and Dust Collection
application of the equipment capable of cutting the fuel debris
Technology for Innovative Approach (with dust collection function)
Remarks) Proposal of technologies for transfer debris or cutting the concrete / steel wall
surface will be requested as necessary when the specific method is adopted.

©International Research Institute for Nuclear Decommissioning

119Finally, - Toward retrieving fuel debris  Retrieving fuel debris of Fukushima Daiichi is expected to be more difficult
than TMI-2. It is therefore necessary to gather knowledge and information
domestically and internationally for developing overall strategy for fuel debris
retrievable.
 To retrieve fuel debris, it is necessary to make a plan best-suited for theentire related projects and to develop technologies flexibly while making
clear the purpose and goal of each project.
 To formulate a strategy, it is important to consider the end state (whatshould be done in the end) and study various feasible options. As a result,
it is important always to prepare alternatives, as well as the first idea.

©International Research Institute for Nuclear Decommissioning120