Development of Fuel Debris Retrieval
Technology at IRID
The 1st International Forum on the Decommissioning of the
Fukushima Daiichi Nuclear Power Station
Spa Resort Hawaiians
April 11, 2016
Kenro Takamori
International Research Institute for Nuclear Decommissioning(IRID)The results explained here have been obtained under subsidies to development costs for nuclear decommissioning
and safety technologies, granted by the Ministry of Economy, Trade and Industry.
Unauthorized duplication and transcription prohibited Technology Research Union International Research Institute for Nuclear Decommissioning©International Research Institute for Nuclear Decommissioning

Outline of IRID

1. NameInternational Research Institute for Nuclear Decommissioning (IRID)
2. Establishment
August 1, 2013: Establishment of IRID was approved by the Minister of Economy

2. Location of Headquarters
5F 3Toyo Kaiji Building, 23-1 Nishi-shinbashi 2-chome, Minato-ku
Tokyo 105-0003, Japan
TEL: +81 3 6435 8601 (representative)
website: http://www.irid.or.jp/en
3. Membership (18)
Research Institutes: Japan Atomic Energy Agency (JAEA),
National Institute of Advanced Industrial Science and Technology
Manufacturers, etc.: TOSHIBA Corporation, Hitachi-GE Nuclear Energy, Ltd.,
Mitsubishi Heavy Industries, Ltd., ATOX Co., Ltd.
Electric Utilities, etc.: Hokkaido Electric Power Co., Inc., Tohoku Electric Power Co., Inc.,Tokyo Electric Power Co., Inc., Chubu Electric Power Co., Inc.,
Hokuriku Electric Power Company, Kansai Electric Power Co., Inc.,The Chugoku Electric Power Co., Inc., Shikoku Electric Power Co., Inc.,
Kyushu Electric Power Co., Inc., The Japan Atomic Power Company,
Electric Power Development Co., Ltd., Japan Nuclear Fuel Limited

2©International Research Institute for Nuclear Decommissioning

Scope of Business
IRID gathers knowledge and ideas
from around the world for the
purpose of R&D in the area of nuclear
decommissioning under the
integrated management system.

R&D for
Decommissioning

R&D projects:･Fuel Removal from spent fuelpool
･Preparation for fuel debris
retrieval･Treatment and disposal ofradioactive waste

Promotion ofCollaboration on
Decommissioning
with Domestic andInternational
Parties3

Developmentof Human Resource
for R&D©International Research Institute for Nuclear DecommissioningRole of IRID
IRID focuses on a current urgent challenge—R&D for decommissioning of the Fuskushima Daiichi NPS—
with a view to strengthening the foundation of nuclear decommissioning technologies.

Tokyo Electric Power Co.
(D&D Company)

Decommissioning ofFukushima Daiichi NPS

Progress managementof decommissioning
Report on progress status

Government (METI)
Establishment of Mid-and-long Term
Roadmap (Including R&D programs)

ReportProvision of
R&D results

Entrusting ofimportant issues

R&D needs

NDF

IRID
Safetyregulation,
approval

Report,application

Presentation ofstrategies and R&D
plans

R&D
implementation

Grant• Strategy planning ofimportant issues• R&D planning

ReportSupport as needed

Fund fortechnical development

Nuclear Regulatory Authority
Approval of safety regulations andimplementation plans(Management office:Mitsubishi Research Institute)4©International Research Institute for Nuclear Decommissioning

IRID’s R&D Projects
R&D for Fuel Removal from Spent Fuel Pool
Evaluation of Long-term Structural Integrity ofFuel Assemblies Removed from the Spend Fuel Pool

Decontamination/Dose Reduction

Debris Retrieval<Securing of work environment>
Remotely Operated Reliability Evaluation
Decontamination of Remotely-operated
Decontamination
Equipment
Equipment

Repair and Water Leakage
Stoppage of PCV
Repair andWater Stoppage
Technology of PCV

Full-Scale
Test<Securing of stable condition>

Criticality Control <Fuel debris retrieval>
PCV/RPVin Fuel Debris
Integrity Evaluation
RetrievalUpgrading of Retrieval
Method for Fuel Debris& Reactor Internals

Investigation/Analysis in the Reactor
Investigation
Inside PCVInvestigation
Inside RPV
Identifying
Properties ofFuel Debris

Detection of
Fuel Debris

Accident
Progression
Analysis

5

Fundamental
Retrieval Technology
for Fuel Debris &Reactor Internals

Collecting,Transferringand Storing ofFuel DebrisRadioactive Waste
Treatment/Disposal

Solid Waste
Treatment
and Disposal©International Research Institute for Nuclear Decommissioning

Development of Technology for Remotely
Operated Decontamination

support to reduce the work exposure in the investigations andthe decommissioning works

6©International Research Institute for Nuclear Decommissioning

Development of Inspection/Investigation Equipment
Device for Investigation of the S/C upper part

Device capable of underwater
swimming and floor travelling

Investigation of leakage from the S/C upper structures

Investigation of areas penetrating
walls underwater and in muddy water

3.8 m

Gengo ROV

TrydiverTracerdischarge

Telerunner
0.8 m

TracerReflection
from a pipe

Reflection froma tracer

Sonar

Device for Investigation of the joint part
between the vent tube and the D/W

Device for Investigation of
the S/C lower part

Investigation of water leakage at the bottom part ofthe vent pipe, using a camera

Investigation of the existenceof a hole with a diameter of
30 mm or more in asubmerged part

SC-ROV

VT-ROV©International Research Institute for Nuclear Decommissioning

http://www.irid.or.jp/en/

7

Development of Technology for Remotely Operated Decontamination
Composition of an air dose

Contamination forms
Loose, fixed and permeated contamination forms are combined.
Loose contamination
Fixed contamination
Concrete(Permeatedcontamination)

70%

Surface coating(Epoxy)Air dose
20%

Approaches to decontamination areas
Air doses in thereactor building(example)

For high places

For low places(the floor, the lower surfacesof walls)

Extension

Decontamination device forcontamination forms

Revolvingarm

For upper floors

Decontamination
unit cart

Compressor cart

100 m

High pressure
water jet

Dry ice blast

40 m

20 m

Work cart

Blast(suction)©International Research Institute for Nuclear Decommissioning

8

Development of remote decontamination technology:
Examples of tie-ups with universities
Examples oftechnical issues

 Operating a robot using only data from a normal robot-mounted camera makes it
difficult to understand the robot’s surroundings, and operations are not easy.
 Avoiding interferences when using a multi-jointed manipulator in confined spaces isdifficult and requires complex operations.

Understanding the vicinity 1“Tokyo Uni. Yamashita Lab”
 Revised images from multiple robotmounted cameras, and images
looking down at the robot from above(quasi-overhead view) facilitate
understanding the condition of theenvironment
 Development of techniques to allow
Changes in camera type, attachment
position and direction to be handled
flexibly, and easily adjust the amountof image correction

Understanding the vicinity 2“Tsukuba Uni. Tsubouchi Lab”

Improved Operability“Kobe Uni. Yokokohji Lab”

 Development of a system that maps
3D measurement data to the robot
surroundings, thus facilitating
judgements.
 Considering applicability to robots,
development of a system thatappropriately selects still or moving
images of the required definition to
enable flexible responses even when
signal speed is low.

 While manipulators with multidegrees of freedom can avoid
obstacles and have advantages in
confined areas, they are also difficult
to operate.
 As operability should be made lesscomplex, the development of aneasy to use interface that allows
instinctive self-motion * movement
commands＊: Movements which change the entire
shape of the manipulator while the tip of its
claw and base are fixed in position
Shape changes(Image of shoulder joint moving
forwards and backwards)

Camer
a

Tip of claw isfixed in position

CameraQuasi-overhead images
taken from Super Giraffe

Quasi-overhead images
taken from MEISTeR

3D mapping screen (currently
under development) of datataken with Super Giraffe

Image showing understanding ofsurrounding environment using
3D sensor data and camera
images

最大牽引力3400N

Example of self-motion movements of manipulator
with 9 degrees of freedom©International Research Institute for Nuclear Decommissioning

9

Development of Technology to Repair the PCV
Set up of the basic configuration of the debris removal works
and reduction of the additional RI release risk

10

©International Research Institute for Nuclear Decommissioning

Development of Technology to Repair the PCVAreas that need to be studied for repair in order to submerge fuel debris
in the PCV with water or to form a PCV boundary
Study on the circulation cooling system

Water stoppage technology forleakage from the sealing section
Water stoppage technology forleakage from the pipe bellows
The inside of the small chamber

Water stoppage technology forleakage from the sealing section
Water stoppage technology forleakage from the pipe bellows
Open section

Water stoppage technology
involving injection of sealingmaterial into the vent pipe
Vent pipes, quenchers and strainers
Water stoppage technology forleakage from the vacuum break line
(Unit 1)

Construction of boundaries for PCV
connector pipes
Pipes installed in the torus room
which are the D/W connector pipes

Technology to stop water leakage fromclearances in the reactor building

Technology for strengthening the
suppression chamber support columns

D/W shell repair technology

Target areas for Development of Repair and Water Leakage Stoppage Technology©International Research Institute for Nuclear Decommissioning

11

Development of Technology of Repair of the Primary Containment Vessel (PCV)
Check of water leakages

Method to repair openings outside
the D/W
Injection of non-cement Protective cover
Injection
based material

Water leakage points

Expansion joint on main steam piping
Main steam
piping3
Expansion
joint

Joint box

1

Bellows

Method to repair narrow spaces
outside the D/W

1
The protective cover of the expansion joint
on the vacuum breaker line
22

Expansion
joint
Basic test of a water stoppage
method in which the vent pipe is filled
up with water stoppage material.

Method to repair PCV lower
part

Sand cushion drain pipe

provided by TEPCO©International Research Institute for Nuclear Decommissioning

12

Development of Technology to Repair the PCV
Stopping water leakage from the vent pipes

Stopping water leakage from the downcomer

Repairing materials

Guide pipe

used for vent pipe expansion flattery andthe measure against degradation ofareas water leak blockage part

Stoppage of waterleakage from thevent pipe

Sub inflatablesealing bag

Installation of
an inflatablesealing bag

Possibleminor waterleakage

Stoppage of waterleakage from thevacuum breaker

Inflatablesealingbag
BalloonPossible waterleakage to S/Cconnection
systemPossible waterleakage to S/C shell

suppresses a water flow amount at thetime of junction (welding) to S/C, grout
injection and construction

Repairingmaterials
Stoppage of waterleakage from thedowncomer
Mainly grout

Stoppage of waterleakage from thequencherWater stoppage material
Mainly grout.

Stoppage of water leakage from the downcomer and vacuum breaker

S/C (suppression chamber) cross-section

Stoppage of waterleakage from the strainer©International Research Institute for Nuclear Decommissioning

13

Mock-up Test
Facilities for the development andvalidation of remotely-operated
devices and equipment

It is necessary to prepare an environment to repeatedly test and verify devices
and equipment and avoid checking the applicability of the devices andequipment on site in a trial-and-error manner.

Equipment for Investigation of the S/C lower part

Method to repair the PCV lower partFull-scale test for repair and water leakage
stoppage technology for leakage points inside the PCV

1/1(1/8) Sector Suppression Chamber Mock up

JAEA, Naraha Remote Technology
Development Center Test building

Interior of the torus room

Interior of the S/C©International Research Institute for Nuclear Decommissioning

14

Probe for fuel debris
Investigation inside the PCVInvestigation inside the RPV
Muon technique
Identification and Analysis of Conditions inside the Reactor
Severe Accident Progression Analysis Technology

15©International Research Institute for Nuclear Decommissioning

Investigation inside the PCVDevelopment of robots to investigate inside the PCV● A linear shape that allows the robot to go through a guide pipe with a diameter of about 100 mm and stable movement in the PCV● Use of robots under severe environments (high radiation dose, darkness, steamy atmosphere, etc.) , and collection of "image","temperature" and "dose rate" data

Shape-changing robot

Camera& lighting

Thermometer
CameraScorpion robot
Camera& lighting

Crawler

In going through a pipe

Self-righting
mechanism

Dosimeter
Shape change In moving on aplane surface

The robot goes into the PCV through a 115 mm-diameter penetration.

CameraRequirements forradiation resistance
100 Gy/hCrawler* CRD: function todrive control rods

The robot goes into the inside of the pedestal from the CRD* rail.

Cumulative 1 KGy[Unit 1]
X-100B
Penetration(115 mm indiameter)[Unit 2]
X-6
penetration
CRD rail©International Research Institute for Nuclear Decommissioning

16

Internal inspection of reactor containment vessel (PCV)Internal inspection of pedestal

Expectations for internal inspection of PCV①Location of fuel debris

・Consideration of access, unloading route, cutting method, basic data required for design②Fuel debris distribution, composition (properties)
・Data required for decision on technical framework to enable risk management and keep reactor safe
・Data required for design that maintains environment and allows provision of cutting so that work proceeds steadily③Damage and other conditions
・Required data to enable water to be shut-off and allow continuation of removal work that employs water・Required data to allow debris removal to be completed smoothly, and retain the pressure vessel with certainty

Investigation Steps
Survey of CRD rail
condition（A1）Performed Aug 2013

Survey of condition of pedestal
inner platform / lower CRD / slot
opening (A2）Survey for deciding fueldebris removal method(A3)

Survey of area under
platform (A2)

RPV
Issues in A2: Steps up to survey of inner PCV

・Decrease dose around penetration
・Make a hole in penetration hatch (Φ115ｍｍ）
・Secure access route to platform

Issues in A3: Steps up to survey of inner PCV・Opening the hatch
・Avoiding/removing penetration inclusions※・Avoiding/removing suspended tools on CRD
rail※※Judgement will be made depending on the
result of A2 survey

X-53
penetration

CRD
Pedestal
openingA1

X-6
penetration

A2

Platform
CRD Rail
A2’

PCV

Unit 2: Proposed access route to inside containment vessel and pedestal (below pressure vessel)©International Research Institute for Nuclear Decommissioning

17

Investigation inside the PCV
Future plan

①Development of shielding, isolation and boring device for the X6
penetration of unit #2, Pedestal platform investigation②Investigation of fuel debris below the pedestal platform of unit #2③Investigation of PCV lower floor of unit #1 to utilize video cameraand dose rate meter from the upper floor④Investigation of inside the pedestal of unit #3 by underwater ROV

18©International Research Institute for Nuclear Decommissioning

Investigation inside the Reactor Pressure Vessel (RPV)
Purpose of investigation inside the RPV
Obtain data on the locations of fuel debris inside the RPV, the damaged condition of reactor internals,
temperature and dose inside the RPV, etc.
● Study methods to access, investigate and sample investigation targets● Access after the reactor is opened

Study of access methods
The following is considered● Doses and the installability of shield● Accessibility of investigationequipment● Observability
Feedwater A system● Access to drill the upper part

Reactor core spray

Device to drill the RPV head

● Access to system piping
Electromotive travelling device
Traveling

Horizontal and vertical
Elbow and deformed
pipes

Grabbing

Position andposture holding

Bifurcation

Passage of T-shapedbifurcation
Directional control

Jet pump
instrumentation Hydraulic type travelling device©International Research Institute for Nuclear Decommissioning

19

Visualization Technology

Visualization of fuel debris in the
RPV

3D imaging of the situation inside thereactor building
3D laser scanner
Data of 40,000,000 points per
10 minutes(Horizontal plane)

Development of Technology for Detection of Fuel Debris
in the Reactor

A route for lowering
devices from the ceiling(the floor surface on thefirst floor) was found.

(Upper surface)

Simulation(90 days)μμMuon detector

Visualization of air dose

Gamma raydetector

Gamma camera

Hot spots exist in
pipes.

CameraMuon detector

Collimator

XY detection unitμ1mShutterμ1m©International Research Institute for Nuclear Decommissioning

20

Installation of Detectors Using Transmission Method
Muon Detectors were installed at the north andnorth-west corners of Unit 1 reactor
building (late January, 2015)
タービン建屋N
 Measured from February through May Detector
 Detectors installed in the front of thereactor building were shielded by 10 cm
thick iron plates

GroundLevel

Detectors
設置予定位置N

Reactor Building
原子炉建屋

Detectors

3 layered XY detector unit
（１m × 1 m × 1 cm）

Iron Plate shield

Network switch
UPS

R/B Cover
カバー建屋
Detector
Y-direction

Detector
X-direction©International Research Institute for Nuclear Decommissioning

21

Estimation of Fuel Debris Location Based on Comparison
between Design Image and Measurement (Detector 1)

Boundary of PCVBoundary of reactor
Boundary of core

 Measured data, though it does not clearly
indicate, shows that equipment, etc. aredetected at locations where they are supposed
to exist based on the design documents
 The boundaries of the PCV and the RPV in theimage acquired from measurement matches
those in the image drawn from design data.
 High density material (fuel) is not detected at
the area where the reactor core was originallylocated.

©International Research Institute for Nuclear Decommissioning

22

Identification and Analysis of Conditions inside the Reactor
Severe Accident Progression Analysis CodeTEPCO: Remote visualization of actual plant status

MAAP (Modular Accident Analysis Program)

Understanding reactor internal status using
severe accident analysis code●”Simplified model” using test result-based correlation formula allows highspeed calculation●Improvement of Analysis Codes &Accident analysis using actual data

●Can perform parametric analysis of phenomenon of large uncertainty

・MAAP ／ SAMPSON●Individual event assessment through thermal flow analysis●International Cooperation: OECD/NEA benchmark analysis（BSAF）SAMPSON (Severe Accident analysis code with Mechanistic,
Parallelized Simulations Oriented towards Nuclear fields)●Uses theoretical models and “mechanistic model” with sophisticateddescriptions of physical and chemical phenomena●Multidimensional analysis of space distribution-related phenomena is possible

MAAP Schematic model of reactor pressure vessel interior
Heat Transfer to
Isolation Condenser
Steam Dryers
RPV Heat Loss to Drywell
Steam Separators

Core Sprays
SafetyInjectionLossthroughBreak
Zr/H2O Reaction

Stream &H2 Production from Core
Release throughSafety/Relief Valves
Main Steam Flow
Flow to HPCI/RCICHeat Transfer to Gas in Primary System
Main Feed Water Flow
Fission Products Decay
Core Heatup & Melt
Progression

JAEA: Mock-up Test (Seawater heat transfer test, etc.)

Relation between analysis item & analysis code
Event description/analysis
・Initial event
・Estimate of plant status(temp, press, PCV, RPV, water lev.)●Severe accident analysis
code MAAP, SAMPSONPCV leak / RPV damage location & area
・Assessment of PCV leak condition
・Assessment of RPV damage condition

Location & amount of debris

・Debris temperature・Analysis of core concrete reaction
・Re-criticality●Simple heat transfer flow
calculated: Energy balance
Thermal conduction,
Radiation, Convection●Transient Analysis CodeTRACG, etc.

FP amount in PCV & RPVHeat Transfer to Heat Sinks

Prediction of water feed amount,pressure & temperature

RPV Heat Loss to Pedestal

Heat Transfer to
Fluid Loss though RPV Penetration
Water in Core
Leakage from CRD
Heat Transfer from Debris in Lower Plenum
Provided by EPRI
Fission Products Decay

H2 density (upper limit)

・Prediction of amount of N2 suppliedpressure & temperature・Prediction of N2 & H2 densities●Universal heat flow
analysis code
“Overview of improvements inassessing conditions inside thereactor through technical progress inaccident development analysis” The
Institute of Applied Energy 22/10/2012©International Research Institute for Nuclear Decommissioning

23

Severe accident development analysis technology
Melted Core - Concrete Interaction (MCCI) Assessment
Example of MCCI assessment performed on the Unit 1 actual sump system with eroded concrete convection
flow/spread model added to the DSA module which has a high generality in the SAMPSON codes
Unit 1 actual sump system measured results

Addition of convection flow/spread of eroded concrete to debris

Analysis carried out until 12 hours after RPV damage

3.95 mEroded concrete convection flow/spread model

Model:
Cross-section of concrete floor【モデル：コンクリート床断面図】
Concretedensity
コンクリートdifference
濃度差t=3600 st=36000 s

B-B’ Vertical section heat
distribution

Sump pit

Concrete
コンクリートB

0.5 m

Primary
cavity wall
キャビティ初期壁面

B’

Pedestal wall

Sump pit■Melted debris ■Solid debris■Concrete

Debris surface

t=4200 s

Horizontal section ofliquid-solid distribution

Eroded face

0.2 0.4 0.6 0.8 1.0
Concrete density (-)

14.2 mFinal surface form of debris predicted within an accuracy of 13%

0

2000

B-B’ Vertical section
heat distribution

Validation through OECD/MCCI CCI-2Test Results:

Eroded
侵食箇所area

1000
Temp (K)

Erosionboundary
侵食境界

Melteddebris
溶融デブリ

0White line: Test observation data

■Melted debris ■Solid debris■Concrete

24(1) Condition wheredebris spread

B-B’ Vertical section ofliquid-solid distribution(2) Long-term
erosion behavior©International Research Institute for Nuclear DecommissioningSevere accident development analysis technology
Summary of overall analysis & assessment results of fuel debris distribution

Fuel debris assessment from severe accident development analysis codes and other survey results

Item

Unit 1

Unit 2

Accident
development
analysis

Most fuel debris
transferred to PCV side

Fuel debris distribution
greatly depends on amountof water pumped in by
firefighters

Most fuel debris
transferred to PCV side

Assessment ofheat balancemethods

Few heat sources inside
RPV

Constant rate in both RPV
and PCV

Constant rate in both RPV
and PCV

Myuon
Measurement

Almost no high density
material (fuel) in reactor
core

Almost no large pieces offuel debris in reactor core

Not measured

Internal
Inspection of
PCV

No large scale damage
to PCV wall in confirmed
scope

No large scale damage toexternal area surrounding
lower part of RPV

No damage to internal
PCV structure in
confirmed scope

Overall
Assessment

Most fuel debris
transferred to PCV side

Constant rate in both RPV
and PCV

Most fuel debris
transferred to PCV side

25

Unit 3©International Research Institute for Nuclear Decommissioning

Development of Fuel Debris Retrieval
Fuel Debris Characterization
Collection, Transfer and Storage of Fuel Debris
Structural Integrity Evaluation
Assessment of Seismic Resistance
Fuel Debris Retrieval Method

26©International Research Institute for Nuclear Decommissioning

Fuel Debris Characterization
Generation of mock debris

Metallic
compounds●Estimate of oxidization, metals produced➜Thermodynamic equilibrium calculated(fuel distribution, oxygen density,
temperature inside reactor)(SUS-Zr-U compound,
ZrB2 etc.)●High temperature reaction
with concrete (MCCI *)*Molten Core Concrete Interaction(metal)Reactants withconcrete(hydrated silica,
salt compounds)

Fuel debris from TMI-2

Investigation of physical properties required forextraction & sampling of fuel debris●Physical Features(shape, size, density/porosity, hardness,elasticity, fracture toughness)●Thermal Features(specific heat, thermal conductivity,
melting point)
Estimate of distribution of hardness for eachchemical system of debris

Boride●Comparison with TMI-２debris
Hardness of mock debris isalmost identical to TMI-2

Types of extraction tools(boride, oxide, metal)

Product composition varies with concrete
composition, melting temperature, time

ZrB2, (Fe, Cr)2B, (Cr, Fe) B

Cutting

Collection

Principle:
ImpactPrinciple:
Pick & place

Hydrated silica glass

Oxide
C, T-(Zr, U) O2(alloy oxidization)
SS-Si-Zr
Fe-Cr-Ni
SS-Si-AlSS-Si

Multiple layers of oxidized material
between eroded concrete surfaces●High temperature reaction withseawater sodium

Boridecrystallization
Structural material

Image of actual debris

Understanding distinctive reactions at
Fukushima Daiichi

Boride derived from B4C control
material is remarkably hard, and may
burden cutting tools((U, Zr)O2 etc.)

Unmeltedfuel pellet

Oxidation: （U, Zr）O2
Metals: Zr(O), Fe2（Zr, U）●Boron-generated reaction products

Oxidizedmaterial

05

C-(Zr, Ca, Al, Fe, Cr) O2
C-(U, Zr, Ca, Al) O2

Metal

10
15
20
Granular hardness(Gpa)

25

Cutting

SuctionPrinciple:
Shear

Principle:
SuctionCutting

Core-boring

Principle:
FusingPrinciple:
Grind &
compress©International Research Institute for Nuclear Decommissioning27Collection, Transfer and Storage of Fuel Debris
Collection, transfer and Storage of TMI-2 Fuel
DebrisCanisterhandling crane(shield)

Debris loaded
Lid closed
Close plug attached

Canister handling crane(shield)

Cover

FilterVacuumdryingequipment

Canister Lid

Transported
Heatingvacuum drying
container
Canister

Vent pipe

Transported
off-site(rail, road)

DebrisFuel canister

Storage Cans

Transportation
cask

Heated vacuum
dryingHorizontal silo

Fuel canister

Primary and secondary
lids welded
CanisterCanister lid

Reactor vessel

Fuel debris loaded

Fuel canalCanisterloaded to cask

Extraction of canister(Dry, remote)・Wet storage

Sampling
hole

TMI-2 storage pipe

Fukushima Daiichi Issues●High degree of burn-up and concentration⇒radiation, large disintegration, high reactivity●Generation of concrete-related melt product expected⇒Concerns of hydrogen emission via radioactive
breakdown of water in concrete●Seawater injection, melting of instrument cables⇒Effects of sodium in fuel debris, introduction ofmultiple impurities

Fuel canister

Dry storage (horizontal silo)©International Research Institute for Nuclear Decommissioning28

Development of Technology for Criticality Control in Fuel Debris Retrieval
Objectives of criticality control technology development
It is believed that currently the fuel debris is not in state of criticality, but alongside future work to retrieve fueldebris, criticality control methods and monitoring will be developed to prevent any re-criticality occurring evenwith changes to fuel shape and amount of water included.
Technical development point

PCV interior

Reactor building●Changes to fuel debris shape●Changes to amount of water included (waterlevel) due to submersion
Small risk of exposure, but important to monitor
conditions over a relatively wide area

TurbineBuilding⇒Development of re-criticality
occurrence detection methods
Prevent re-criticality from occurring

Coolingwater⇒Development of neutron-absorbing material toprevent criticality

PCV exterior
Retained water●Powder produced when cutting fuel debris has a riskof escaping and collecting in waste water

Currently

treatment/cooling equipment
Need to preemptively prevent exposure of equipment
maintenance workers

In future
Water treatment
equipment⇒Development of sub-criticality
monitoring Technologies©International Research Institute for Nuclear Decommissioning

29

Structural Integrity Evaluation
Residual life assessment outline flow
Need a means to maintain reactor structural
integrity for a long time until the fuel debris is
retrieved from reactor core
●Effect of high temperature during fuel melt

Buildingequipment
coordination
behavior
analysis

Building behavior
analysis(Building
damagesimulation)●Effect of corrosion due to inclusion of seawater and foreign particles●Effect of equipment damage due to hydrogen explosion●Effect of debris retrieval method●Effect of further earthquakes

Model Creation

Assessment Conditions

・Basic ground motion Ss

・Corrosion wall thinning (test data)

・ Influence of corrosion

D/W shell
Reactor shieldwall

Low marginequipment listedup・Allowable values

Anticipated plant condition

PCV Stabilizer

Equipment
strengthassessment

・Water level

・ Damage to equipment

Example of soundness assessment point

Organizing
loading
conditions

・High-temperature strength degradation(test data)

Increased anticipatedlong-term corrosion wastage

Submerged
PCV penetration

RPV Pedestal

S/C shell

Vent pipe

Water stopped(grout fixed)

Long-term
corrosion test

ColumnsupportComparison of
corrosion control
measures©International Research Institute for Nuclear Decommissioning

30

Methods of Fuel Debris Retrieval
Diagram below taken from NDF “Technical Strategic Plan 2015”

Selection options for method of fuel debris retrieval
a. Submersion – top entry methodb. Partial submersion – top entry method c. Partial submersion – side entry method

Techniques required for method feasibility study✔ Techniques to prevent contamination spread
during retrieval of large structure✔ Techniques to prevent contamination spread
during retrieval of fuel debris from inside RPV✔ Techniques to access fuel debris✔ Techniques to retrieve fuel debris throughremote operations✔ Techniques for cutting, collecting, monitoring,and measuring during fuel debris retrieval

System Concept, Investigation of Method Feasibility
Fuel debris retrieval device
Radioactive
dust treatment
systemPoints to consider regarding fuel debris retrieval equipment✔ Equipment resistance to radiation & maintainability

Internal gas management system(negative pressure control)✔ Improved efficiency in fuel debris retrieval method✔ Compatibility with storage cans and other equipment

Criticality control
systemCirculation,
feed water,
cooling system

Purification
System✔ Equipment to collect radioactive dust from around device
Reference:
Guidelines for solicitation of entities to implement with subsidies “Subsidy to Project ofDecommissioning and Contaminated Water Management (Project of Upgrading
Approach and System for Retrieval of Fuel Debris and Internal Structures)” and “Subsidy
to Project of Decommissioning and Contaminated Water Management (Project ofDevelopment of Fundamental Technologies for Retrieval of Fuel Debris and Internal
Structures)” in the FY2014 Supplementary
23 June 2015

©International Research Institute for Nuclear Decommissioning

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Summary
1.

Development of Fuel Debris Removal Technology
 Project of Fuel Debris Retrieval Method Development
 Projects to support fuel debris retrieval development・decontamination, investigations, accident analysis, repair and waterleakage stopping, structural integrity evaluation, criticality control
・collection, transfer and storage, waste management2.

Key technologies
 Challenges in the extremely high dose environment
・radiation shielding, mitigation of RI release・remote control technology
・visualization technology

 Established reliable technology and flexibility of introduction of newtechnique3.

To ensure the safety: Nuclear safety is the maximum value Introduction of debris handling experiences, performance and lessons
learned from overseas countries
 Design to support the risk management in the site Establishment of safety design plan and dialogue with the regulator
 Verification by examinations and mock up tests
 Enough trainings

32©International Research Institute for Nuclear Decommissioning