Subsidy Project of Decommissioning and Contaminated Water Management

Upgrading of the Comprehensive
Identification of Conditions inside Reactor
Accomplishment Report for FY2017

June 2018
International Research Institute for Nuclear Decommissioning (IRID)
The Institute of Applied Energy (IAE)All Rights Reserved. International Research Institute for Nuclear Decommissioning©International Research Institute for Nuclear Decommissioning

1

Contents of the Report
1. Planning for the Project of Identifying of the Conditions
inside the Reactors and the Project Structure
2. Project Achievements for FY2017
3. Research Details and Results Summary for FY2017©International Research Institute for Nuclear Decommissioning1-1. Project Needs⚫

⚫

For the decommissioning initiative for the Fukushima Daiichi Nuclear Power Station, it is crucial to estimate andunderstand the conditions of fuel debris and fission products (hereinafter FPs) inside the reactor pressure
vessels (hereinafter RPVs). Direct observation inside the RPVs is still challenging due to the highly radioactive
environment.
A practical alternative measure is to comprehensively advance analysis and evaluation techniques to estimate
the conditions inside the RPVs and the primary containment vessels (hereinafter PCVs) through accident
progression analysis and evaluation of various measurement data and information, such as those collected onsite. This knowledge is expected to be used for the decommissioning process. To this end, initiatives forunderstanding the conditions inside the reactors have been pushed forward since FY2011, centering on
improved severe accident analysis codes; the Modular Accident Analysis Program (MAAP) and the Severe
Accident Analysis Code with Mechanistic (SAMPSON).

⚫

Since FY2016, comprehensive activities have been carried out to advance identifying of the conditions inside
the RPVs and PCVs. Not only the analytical results, using the accident progression analysis codes, but also
plant parameters obtained at the time of the accident and other actual measurements will be used. Fact-based,
on-site information collected through the internal PCV investigations, or dose investigations inside the R/Bs, inaddition to the material science knowledge enhanced by recent studies will also be used.

⚫

Amid such circumstances, the following technical development was conducted in FY2017.
• In FY2016, the review was pushed forward with no improvement of the analysis codes based on theunderstanding of their capabilities and limits. In the end, it was necessary to upgrade the analysis codes toenhance the analysis results’ reliability. Therefore, it was decided to make partial improvements to the analysis
codes in FY2017.
• Estimating the conditions inside the RPVs and PCVs by using the results obtained up to FY2016 wereadvanced. The study items selected for further research to reduce uncertainties of the estimated conditions
inside the RPVs and PCVs were also analyzed and evaluated.

©International Research Institute for Nuclear Decommissioning2

3
1-2. Application of Project Achievements and Project Target⚫

The aim of the project is to estimate the conditions inside the RPVs and PCVs throughcomprehensive analysis and evaluation of the study on issues to be solved. Specifically, major
outputs of the project are shown in the below diagram; fuel debris and FP distributions that were
estimated based on comprehensive analysis and evaluation. The diagrams were occasionally
updated to reflect the latest results and provide them for other related research projects.

⚫

In other words, the purpose of the project is to provide information that will be useful for fueldebris retrieval and decommissioning work.
Development of Technology for Investigation inside PCVDevelopment of the Technology to Detect Fuel
DebrisFuel debris
distribution

Pedestal, presence/absence of fueldebris Damage conditions inside
the PCVs, etc.

Temperature, pressure, and actual
dose distribution Location of coolingwater leakage, etc.

The Project of Upgrading the Comprehensive
Identification of Conditions inside Reactor
Fuel debris distribution, quantity,
composition, etc.

Development of Technology for Criticality Control Methods
Development of Technology for Collection, Transfer andStorage of Fuel Debris

Other data and information

Fuel debris distribution, quantity, composition, FP
distribution (dose), temperature/pressure at thetime of the accident, etc.

Development of Technology for Investigation inside PCV
Fuel debris
distribution, quantity,
composition, FP
distribution (dose),damage condition
inside the reactors,
etc.

Development of the Technology to Repair/Stop Water
Leakage from the PCVs

Development of the Technology to Inspect inside
Pressure Vessels
Development of the Technology to Evaluate the
RPV/PCVs’ Integrity

Fuel debris retrieval, decommissioning and selection of the fuel debris
retrieval methods, etc.
©International Research Institute for Nuclear Decommissioning

4

1-3. Overall Plan: Implementation Schedule
FY2011a) Analyzing plant behavior at
the time of the accident

FY2012FY2013

FY2014FY2015FY2016

Conduct accident progression analysis using MAAP and SAMPSON

FY2017

Conduct data analysis and accident
progression analysisb) Upgrading severe accident
codes
Improve MAAP
and SAMPSON

Upgrade MAAP and SAMPSONc) Evaluating by means ofsimulation tests, etc.

d) Collecting and analyzing
relevant information(international cooperation)e) Estimating fuel debris and
FP behaviorf) Comprehensively analyzing
and evaluating theconditions inside thereactorsg) (Core team’s) Extracting
study items for further
research and continuously
reviewing the plan

Conduct simulation tests

BSAF Phase 1

Conduct tests for model
development*

BSAF Phase 2

Estimate fuel debris and FP distributions

Estimate the conditions inside the RPVs and PCVs

(Core team’s) Progress management* To be determined after a
discussion on the necessity©International Research Institute for Nuclear Decommissioning

5
1-4. Project Overview and Structure for FY2017
 Project Overview
Identifying of the conditions inside the reactors has been sought by gathering knowledge from Japan and abroad andcooperating with overseas institutions. Accident progression analysis technology, analyzed actual data and on-site investigation
results were used, as well as material science knowledge, etc. By using the highest level of technology available, the fueldebris and FP distributions inside the RPVs and PCVs as an input for the decommissioning task was estimated.

 Structure of the ProjectThis project is conducted under the cooperative framework between IRID and IAE, in which IRID manages and executes theoverall project in cooperation with IAE. IRID conducts the project in cooperation with the IRID member organizations, which are
JAEA (CLADS), Toshiba Energy Systems & Solutions, and Hitachi-GE Nuclear Energy. In addition, the project was promoted
by closely sharing information with relevant organizations, such as NDF and TEPCO, and thereby obtaining cooperation fromthem.
Mitsubishi Research Institute (MRI)(Secretariat)

International Research Institute for Nuclear
Decommissioning (IRID)
IRID Head Office
Organizers: Japan Atomic Energy Agency(JAEA)/Collaborative Laboratories for
Advanced Decommissioning Science (CLADS)

Communication via
the core teamToshiba Energy
Systems & Solutions

Cooperation andinformation sharing

JAEA (CLADS)

Hitachi-GE Nuclear Energy

The Institute of Applied Energy (IAE)

Cooperation andinformation sharing
Relevant organizations including
the Nuclear Damage Compensation
and Decommissioning Facilitation
Corporation (NDF) and TEPCOCore team: An organization consisting of project participating organizations. Responsible for project policymaking and estimating conditions inside the RPVs and PCVs.
©International Research Institute for Nuclear Decommissioning

6

1-5. Implementation Items and Schedule for FY2017First half of FY2017

Second half of FY2017(1) Comprehensively analyze and evaluate the conditions inside thereactors(1) Comprehensively analyze and evaluate based on the Units data andother R&D project results[1]-1 Extract issues for further research to reduce uncertainties of estimatedconditions inside the reactors

Present study items for further research by using database functions

Extract issues for further research by reviewing the FY2016 results, measurement data, and on-site inspection results

Understand the conditions inside the RPVs and PCVs and estimate the fuel debris and FP distributions[1]-2 Comprehensively analyze and evaluate based on the Units data and other
R&D project results

Review

Summarize

Review

Summarize

Other project results, test data, etc.
Collect and organize information[2] Establish a database for comprehensive analysis and evaluation
Enhance and improve the DB(2) Estimate and evaluate fuel debris behavior and fission product behavior
and their characteristics for comprehensive analysis and evaluation

Preparation
Analyze the accident scenario[1] Reduce uncertainties by using analytical techniques, etc.
Evaluate the analysis codes
Analyze based on the plant data at the time of the accident
Enhance necessary tests and data, and verify the analysis modelPreparation[2] Evaluate FPs’ chemical characteristics

Analyze the accident scenario
Evaluate the analysis codes
Enhance necessary tests and data, and verify the analysis model

BSAF Phase-2[3] International joint research to use knowledge inside and outside Japan©International Research Institute for Nuclear Decommissioning

7

2. Project Achievements for FY2017©International Research Institute for Nuclear Decommissioning2-1. Overall Structure of the Project for Identifying the Conditions
inside the Reactors(1) Comprehensively analyze and evaluate the conditions inside the reactors[1] Comprehensively analyze and evaluate based on the Units data and other R&D project results
1 Extraction of issues for further research to reduce uncertainties about the estimated conditions inside the reactors
2 Comprehensive analysis and evaluation based on the Unit data and other research development results[2] Establishment of a database for comprehensive analysis and evaluation (IAE)(2) Estimation and evaluation of fuel debris behavior, and fission productbehavior/characteristics toward comprehensive analysis and evaluation[1] Reduce uncertainties by using analytical techniques, etc.
1 Evaluation of the concrete erosion at the pedestal using sensitivity analysis (Hitachi GE)
2 Inverse problem evaluation using virtual reactors and a compiled database (JAEA)
3-(a)-1 Estimation of the amount of fuel debris that remains in the pedestal space obstacles (IAE)
3-(a)-2 Detailed evaluation of debris behavior inside the post-damage lower head (IAE)
3-(a)-3 Evaluation of the condensation behavior of water vapor containing hydrogen in the S/P (IAE)
3-(a)-4 Event sequence analysis at the time of core material slumping (JAEA)
3-(a)-5 Three-dimensional evaluation of MCCI behavior (IAE)
3-(a)-6 Determination of factors influencing debris distribution (IAE)
3-(a)-7 Analysis of each unit over three weeks after the accident (IAE)
3-(a)-8 Detailed accident progression analysis using accident progression analysis codes (Toshiba
Energy Systems & Solutions)
6 Simulated fuel assembly degradation test (JAEA)©International Research Institute for Nuclear Decommissioning8

2-1. Overall Structure of the Project for Identifying the Conditions
inside the Reactors(2) Estimation and evaluation of fuel debris behavior, and fission productbehavior/characteristics toward comprehensive analysis and evaluation[2] Evaluation of FPs’ chemical characteristics (JAEA, IAE)
1 Reaction between cesium and steel materials, and re-evaporation
2 Evaluation of particulate cesium compounds
3 Optimization of the cesium compounds evaluation model
4 Analysis and evaluation of the units using the upgraded model
5 Analysis of samples collected at Fukushima Daiichi Nuclear Power Station (NPS)[3] International joint research to make use of knowledge inside and outside Japan (IAE)©International Research Institute for Nuclear Decommissioning

9and downs

Identifying reactor
conditions

Fell below the saturation
temperature after water injection
from the CS

Temperature (C)

Approach promoting the estimate bydeepening the Identifying of thephenomenon using data analysis andinverse problem
analysis
Containment pressure (D/W)(actual meas.)

Approach by means ofenhanced reliability ofanalysis code evaluation
and accident progression
scenario analysis

Containment pressure [MPa (abs)]

Containment pressure (S/C)(actual meas.)
Reactor pressure (actual meas.)

Scram(success)

Reactor pressure [MPa (abs)]

Probability density

200°C and subsequent ups

Severe accidentCore damage

RPV integrity
RCIC water injection(maintained/damaged)
DC power(loss)(loss)RCIC water injection
Keep SRV open
(success)(success/failure)

Probability density

Condition severity

High temperature overComplementary relationshipAC power(loss)

Time axis of accident progression

10

2-2 Three Approaches for Further Research

Date/Time

Information obtained byＲＰＶ
on-siteinspection and estimate about theＣＲＤ
surroundingarea[Sv/h-γ]

1000
mm 300mm
350mm

PCV Wall

350
mm

0.3

0.6

0.6

0.3

0.3
0.5

1500
mm

1.0

0.7

1.4
1.4
1.1
1.0

0.7

)

Platform

PCV water levelX-6
penetration.

CRD rail

ＰＣＶ

Opening ofPedestal

X-6 penetration flange(After steel plate removal)
290mm
約50mm

600
mm

0.8

7.2
8.0

1.2

>10 >10Groove after
9.4 8.0

block removal

Red：surface of the flange
Blue：roof, wall

Green：Floor
melted stuff Decommissioning©International Research Institutefor and
Nuclear
Unit 5

2-3 Relationships between Implementation Items 112)-[1]-3-(a)-4 Event sequence
analysis at the time of corematerial slumping

Information
regarding thedispersion of fueldebris(2)-[1]-3-(a)-5Three-dimensional
evaluation of MCCI
behavior(2)-[1]-3-(a)-8 Detailed accident
progression analysis using theaccident progression analysis
codes

Study onthe FP
behavior
modeling(2)-[2]-1 Reaction between
cesium and steel materials, andre-evaporation(2)-[2]-5 Analysis of samples
collected at Fukushima Daiichi
Nuclear Power Station

Knowledge formodel optimization

Information aboutcore melt behavior
in the process
toward slumping

Knowledge of insolublecesium particles

Knowledge of coredamage progression(2)-[2]-2 Evaluation of particulatecesium compounds
Information about thereactor core state at thetime of slumping

Share accident
scenario analysis(2)-[1]-6 Simulated fuelassembly degradation
test

Provide the initial conditions for
MCCI

Knowledge formodel optimization(2)-[1]-3-(a)-7 Analysis of each unit
over three weeks after the accident(2)-[2]-3 Optimization of the cesium
compounds evaluation model

Provide theboundary
conditions for
MCCI(1)-[1] Comprehensive analysis andevaluation

Migration behavior of fueldebris to the lower head

Analysis using the optimal model• Estimate diagram of fuel debris
distribution• Estimate diagram of FP distribution(2)-[1]-3-(a)-2 Detailed
evaluation of debris
behavior inside the postdamage lower head(2)-[2]-4 Analysis and evaluation of
the Fukushima units using the
upgraded model

Information sharing

Study on FP distribution

Analysis conditions
and results(2)-[3] International
joint research use
knowledge inside andoutside Japan(2)-[1]-2 Inverse problem evaluation
using virtual reactors andcompilation of a database

Effect of fuel debris
depositsion conditions
Refinement of the PCV pressure behavior
evaluation(2)-[1]-3-(a)-3 Evaluation ofcondensation behavior of water vapor
containing hydrogen in the S/P(2)-[1]-1 Evaluation of theconcrete erosion at the pedestal
using sensitivity analysis

Migration behavior offuel debris to thepedestal

Evaluation of
the fuel rod
collapse
phenomenon

Timely information sharing(2)-[1]-3-(a)-6 Determination offactors influencing debris
distribution(2)-[1]-3-(a)-1 Estimate of theamount of fuel debris remaining in
the pedestal space obstacles

■JAEA
■IAE
■Hitachi GE■Toshiba Energy
Systems &Solutions

Study on fuel debris distribution©International Research Institute for Nuclear Decommissioning

122-4. Results of Comprehensive
Analysis and Evaluation©International Research Institute for Nuclear Decommissioning

Unit 1: Diagram of Summarized
Information
PCV top head

Reactor building (1F-4F)
•
•
•
•
•
•

•
•

13

Operation floor
•
•
•

•

The sealing has degraded, and a leakage opening
exists (from which vapor and FPs were discharged).

The ambient dose rate is several to tens of mSv/h.
There is high radiation around the RCW piping
There is significant contamination of 100 mSv/h or greater in the TIP room near the containment
penetration of the PCV
The inspection on the 1st floor of the reactor building revealed high radiation around the AC piping(presumably due the effect of the vent).
The southwest section of the 4th floor is heavily damaged, while the damage is minor on the eastside. The ceiling (5th floor) of the northwest section has collapsed. The major damage wasassumed to be affected by blast and rubble caused by the hydrogen explosion on the 5th floor.
Hydrogen of 10,000 ppm or greater was detected from the PCV spray piping (Sep 2011)
In the MSIV room, the max dose was 1,096 mSv/h near the top plate and duct of HVH (Nov 2015).
The source is estimated to be the upper duct of HVH.

•
•
•

•

•
•

60 mSv at 2-2.5 m above the shield plug (2011)
The roof collapsed, almost keeping its flat shape.
Significant contamination with max 121 mSv/h at approx. 1 m above the roof right on
the top of the reactor (Aug 2015)
The temperature right above the reactor was higher than in any other areas (Oct •2011), possibly due to the effect of heat from the FPs attached around the shield plug.
•
The top and middle layers of the shield plug were lifted.
In addition to the top and middle layers of the shield plug, the bottom one had •shifted from its original position.
The top north section of the shield plug had shifted 720 mm to the west. Also, the •top layer sagged down by 84 mm at its north section and by 155 mm at its middle•
section.
The vertical dose distribution around the shield plug indicated that the dose rises
toward the gap below the top layer of the shield plug (max 565.8 mSv/h)(2017). The
horizontal dose distribution inside the gap between the top and middle layers of theshield plug indicated that the dose is higher toward the center of the plug.
The surface dose rate on the shield plug was max 200 mSv/h, and the rate was higharound the center of the plug. (Jun 2017)
Gamma-ray spectra measurement detected only two nuclides, which were Cs-134
and Cs-137, in the gap and on the plug surface. (Possibly other nuclides were notdetected because Cs was a too strong source)(Jun 2017)

Upper part of the pressure vessel
The temperature reading was almost the same at
the upper and lower parts of the RPV.
A large amount of FPs have been attached.
The chemical forms of the attached FPs are
unknown (ex. solubility/insolubility, etc.).
The degree of re-evaporation of the attached FPs is
unknown.
Assume a leak opening was created near the upper
part of the pressure vessel (MS piping, for example)•••

A heat source is estimated to exist near the CRD piping
based on the variation in the HVH temperature.
There is a correlation between the feedwater system’swater injection flow rate and a part of HVH
temperature behavior.
(A heat source is estimated to exist near the CRD piping in
the northwest, northeast, and southwest.)(Debris is estimated to exist around the control rod drive
piping because the HVH temperature’s increment rate waspartially greater than the readings of other PCV
thermometers when the volume of water injection into thereactor was reduced from Dec 2016 to Jan 2017).
The contact condition of the control rod positionindicator was checked. There is no clear tendency
allowing us to estimate the lower RPV section’s
condition.

•

In-core structures
Analysis results indicate the structures’ downward
shift because of creep and support deformation.
Large amounts of FPs are assumed to have been
attached on the separator and dryer.
Cesium has possibly been taken into the steel
materials’ oxide layers.
Cesium may have been combined with molybdenum,
boron, and silicon.

••

Reactor core

Reactor well•

Measurement results and observation data available
Estimated based on the measurement results andobservation data
Accident analysis or qualitative estimation

•

•

CRD

•
•

•

•

High radiation is expected, presumably because
leakage exists in the PCV top head section.

•

The remaining volume seems to be almost zero,as shown by the analysis results.
Muon measurement was performed (Feb-May2015). No significant fuel mass was detectedin the reactor core section.
The water level in the reactor core area waszero

IC
RPV lower head••

The PCV temperature is almost equal to the
RPV temperature, but the PCV temperature
is partially higher.
A ruptured hole is suspected since no waterlevel is observed in the RPV.

•FDW/CS piping•

Body-side water level: A system 65%, B system
85% (Oct 18, 2011)
On the equipment outside the containment andthe piping, no damage that could lead to coolant
leakage was observed.

•

Slow temperature response to the changes in
the water injection status(from T/B)

RPV lower plenum
PCV vent/exhaust stack••Significant contamination of up to 5,000 mSv/h in front of the
SGTS room on 2F (Aug 2011)
High contamination exceeding 10 Sv/h was observed at theexhaust stack used for both Units 1 and 2 (near the SGTS
piping joint)(Aug 2011). The dose dropped to 2 Sv/h (Oct2015).
Deformation and fracture were found on a part of the brace(support) used for the exhaust stack for both Units 1 and 2.
The damage was probably caused by a hydrogen explosion.
Contamination with several Sv/h at the SGTS in Unit 2 ispresumably attributed to Unit 1 based on the estimation that there
was no damage to the rupture disk of Unit 2.
Cs134 (8.26 kBq/cm3) and Cs137 (51.9 kBq/cm3) were detected from theaccumulated water inside the pit (Sep 12, 2016).

•

•

•

•

Torus room
Drywell
•

•
•
•

HPCI
•

A white, powdery deposits exists around the
HPCI’s vapor line penetration X-53 (on the bellows
cover, floor, and wall) (sample collected). A leaktrace exists in the gap between the HPCI piping
and bellows cover as well as at the root portion of
the bellows cover and biological shielding wall.
The root portion has the highest radiation (max 7Sv/h).

MS piping

PLR•

CS pump

The PCV temperature is almost equal to the
RPV temperature, but the PCV temperature ispartially higher.

The inspection on the 1F grating outside the pedestal indicated no major structural damage,although there were some fallen objects. The average dose was several Sv/h. There is amass, seemingly a piping shield, around the PLR piping. (Presumably a molten lead mat)Slow temperature response to the water injection changes. Regardless of water injection,some PCV thermometers have shown higher readings due to the effect of nitrogen injection,
etc.
The measured temperature is lower than the estimated decay heat.
The pressure suggests some leakage in the gas phase of the containment (incl. the PCV top head).
Water-filling conducted in the water level gauge piping (the water level gauge piping is presumablysound).

RPV pedestal••

Due to the dropped fuel debris, damages are
estimated on the grating, TIP piping, CRD
exchanger, etc.
Water injected into the RPV is presumably falling
into the pedestal.

•
•
•
•
•

The water level has been maintained below O.P.3,200 toprevent water from flowing into the turbine building.
Water is leaking from the vacuum break line bellows andthe sand cushion drain line.
Dose on the catwalk: 200–2,400 mSv/h (May 2014)
There is some kind of penetration exists between the R/B andturbine building.
There is significant contamination in the northwest
section, with max 920 mSv/h in air and max. 800 mSv/h
underwater

Turbine building
•
•
•

•

The ambient dose rate was several μSv/h toseveral-hundred mSv/h.
High radiation was detected in theunderground floor.
Significant contamination of up to 5,000
mSv/h in front of the SGTS room on 2F (Aug2011)
Highly contaminated water was assumed to haveflowed out of the vacuum break line bellows andsand cushion.

Drywell floor
•
•
•
•

•

•
•

A water level of approx. 3m (leakage was found from the vacuum break line bellows,which is at about the same water level)
A camera inspection observed a deposits on the floor, and the second inspection indicated increased volume.
Analysis indicated that the debris had spread on the D/W floor, reaching the PCV shell.
At the observation point, the dose in the water in the PCV tends to rise toward the bottom, and it reaches the level ofseveral Sv/h to more than 10 Sv/h at about 0.3-1.6 m above the floor surface. The dose near the pedestal opening (atapprox. 1 m from the floor surface) was up to 9.4 Sv/h (2017).
Simplified fluorescent X-ray analysis of the deposits detected U, as well as Fe, Ni, Zn and Pb, which probably originatedfrom the structural materials and paint inside the containment. Although various gamma-ray analyses identified Cs-134,
137, Co-60, and Sb-125, small amounts in the deposits are estimated, given that none of them was detected in the simplified
fluorescent X-ray analysis.
Based on the captured image and the length of the wire sent, the height of the deposits was estimated to be approx. 0.3
m or lower on the opposite side of the opening and approx. 1.0 m or lower near the opening.
No major damage was observed on the sump pump, valve, etc. on the opposite side of the opening.

RPV pedestal floor••

The high radiation observed near the RCW piping in the reactor building
was it guessed to be caused by debris which damaged the equipment
drain sump in the pedestal and allowed radioactive materials to enter
the RCW system. Therefore, the pedestal is guessed to have debris.
The dropped debris are guessed to be eroding the pedestal floor andsump to some extent.

Suppression chamber••

The S/C is almost filled up with water (a slight amount of gas
generated in the early phase of the accident still exists; nitrogenshielding is underway).
The S/C is estimated to be almost sound.
From the results of the dose inspection at the torus room, it is estimatedthat the Cs concentration level is high at the building and S/C walls or in thestagnant water inside the S/C.

©International Research Institute for Nuclear DecommissioningEstimate Diagram of the Debris Distribution in Unit 1 andIts RPV and PCV Conditions••

From the muon measurement and analysis results, andthe fact that the water level is zero, it is presumed mostfuels have melted ,and there are no remaining fuel rods.

A molten pool may have been formed inside
the reactor during the accident. (generalassumption)

•
•

•

•

(actual measurement and analysis)

The shroud may be deformed, damaged orbuckled. (general assumption and analysis)
The shroud may be damaged, possibly
causing molten fuel enter the down-comer areaand damaging the jet pump. (general assumption)•

The CRGT may remain un-melted if the heattransfer from high-temperature fuel debris issmall. (general assumption)•

It is presumed that the remaining amount of debris issmall, given the fact that the cooldown was successfully
performed before starting the CS water injection (Dec 10,2011). (presumption based on the actual measurement)
Presumably general oxide debris, a result of molten fuelsolidification. (general assumption)••
There are ruptures in the lower plenum
because the water level cannot reach thereactor core. (presumption based on the actual

Particulate debris, if any, may accumulate in a stagnant
spot. (general assumption)
From the HVH temperature, the presence of debris near
the CRD is presumed. (presumption based on the actual
measurements and analysis)•

The drain, etc., at the bottom of the lower
plenum are fragile; they may be damaged.
(general assumption)•

The fuel fallen to the lower plenum may still
exist at the bottom of the RPV. (general

Since a certain HVH thermometer showed a significant
temperature increase when the FDW flow rate was
reduced, debris possibly exists near the CRD in the
peripheral area (but there is no telling whether it is
attached to the outer surface or it is inside), and there may
be RPV ruptures immediately above it. (presumption based on••
A pool of water on the PCV floor, if any, may
have formed particulate debris.
Particulate debris, if any, may accumulate in astagnant spot. (general assumption)••

Particulate debris
Fuel debris (rich in metals)*

The pedestal wall near the sump, and theconcrete below it, may have a larger erosion
area than other portions. (analysis)

The fuel debris that caused MCCI is mixed withconcrete. (general assumption and analysis)
In our estimation, the RCW piping in theequipment drain sump is damaged, allowing
radioactive materials to enter the RCW system.

CRGT
Damaged CRGT
CRD
CRD (with debris inside)

ShroudBecause of damage to the CRGT and CRD housing, asmall amount of fuel debris or molten metal may haveentered the CRD housing. (general assumption and testing)Damaged shroud

Pellet*Ruptured hole of the RPV••

Oxide debris (porous)

the actual measurement)

assumption)•

Remaining fuel rod and its debris*

Concrete mixed debris

measurement)•

14
Legendss

The results of the internal PCV investigation of Unit 3
suggest damage to the CRD housing, the platform, andthe RPV bottom. (presumption based on the actual measurement)

Upper tie plate*deposits (material unknown)•

Part of the fuel debris may have been solidified without
MCCI. (general assumption)•

Debris may have spread on the D/W floor through thepedestal opening. (general assumption and analysis)

Ballooning fuel*Oxide debris*
Heavy metal debris*(presumption based on the actual measurement)

Particulate pellet*•

There are mounds of deposits on theD/W floor, and the closer they are to thepedestal opening, the higher they tend tobe. (actual measurement)

Cladding residue*
Molten reactor internals*
Solidified B4C*•
Approx. 2.6 m

In our estimation, the PCV is damaged
because water is leaking from the sandcushion drain pipe. (actual measurement andControl rod mixed melt*

analysis)* Not used in the estimate diagram of Unit 1
There is a concrete layer, measuring approx. 7.6 m, below the PCV.

The dash-dotted line shows the inside of the containment is asymmetrical with the pedestal.

©International Research Institute for Nuclear Decommissioning

Unit 1: Estimate Diagram of FP Distribution*The estimate was made focusing on cesium, a major source nuclide

Rubble• A hydrogen explosion occurred on the Unit 1 operation floor,
and high radiation, exceeding 100 mSv/h, was measured on the
rubble. Therefore, in our estimation, FPs discharged through thefollowing route: containment -> containment top head flange ->reactor well -> shield plug -> operation floor. FPs are also
figured to exist unevenly along the FP migration route.

•<Chemical forms and characteristics of major types of cesium estimated>
 Cesium iodide, Cesium hydroxide, Cesium chloride• Easily evaporate and tend to escape from the pressure vessel as vapor due to adifference in pressure or concentration.
• Have a hydration property and tend to move with moisture by, for example, water
vapor condensation or dew condensation (on the wall surface).
 Cesium molybdate, Boric acid cesium
Have a lower vapor pressure compared to the afore-mentioned substances and,hence, tend to stay inside a pressure vessel.
Cesium molybdate may form because of a reaction between cesium and steel
material containing molybdenum.
 Compound of silicon and cesium- Insoluble cesium particles (amorphous particles including cesium with its
main component of silicon oxide)
Believed to be difficult to hydrate, and particles the size of a few microns tend to soarup.
- Products of reaction with steel material (crystalline substances)
Believed to be difficult to hydrate, and tend to stay in the oxide layer on the steel
surface when in the 800–1,000°C temperature range.

Cesium-bearing particles, ranging from tens to hundreds of microns in
size, which seem to have originated from Unit 1, were observed in theenvironment. Those particles are figured to have been generated by thereaction between cesium and the operation floor wall materials (such as
Glasron Wool), and it is guessed that similar particles are scattered on theoperation floor.

• Hydrated cesium compounds may have penetrated through theconcrete surface in the form of cesium hydroxide.

•• In a high temperature, the upper-structure may have moved
downwards from the original locations due to creep
deformation, etc.

• It is likely that the D/W pressure rose to a high level after
the accident due to the direct leakage from the pressure
vessel to the D/W; therefore, it is guessed the D/W also hasa large amount of FPs.

•

In the early stage of the core heating, insoluble cesium particles weregenerated by coagulation reaction between silicon oxide and cesiumhydroxide in the gas phase; however, it is estimated that the amount
generated is smaller than that in Unit 2.

••High contamination is estimated on
the building wall, the S/C wall, or thestagnant water in the S/C.

•

Cs134 (19 kBq/cm3) and Cs137 (35 kBq/cm3) were detected in
the samples collected from the stagnant water in the PCV. (Oct 913, 2012)• Cesium possibly remains inside the fueldebris although its amount is expected
to be small.
• FPs may unevenly exist in fuel debris.

Depending on the conditions of the gas phase, such as the water
vapor/hydrogen ratio, cesium molybdate or boric acid cesium may
have served as major chemical species, suppressing their discharge
to the outside of the pressure vessel. Also in Unit 1, water vapor loss
may have occurred earlier than in other units, which may have helped
suppress cesium molybdate generation.

Depending on the temperature/atmosphere history, some cesium inthe vapor phase inside the pressure vessel may have penetrated theoxide layer of the steel material due to the reaction with silicon oxideon the steam-oxidized steel surface. Considering this factor, the highradiation may have been caused at the separator and dryer becauseof their large surface areas.

• In Unit 1, the leakage in the S/C is located at a high level; therefore, FPs
released into the S/C may still be trapped in the stagnant water.
• A huge amount of FPs may still exist, given the large volume of stagnant
water in the S/C.

©International Research Institute for Nuclear Decommissioning

15

Unit 1: Estimate Diagram of DoseDistribution

16
Rubble• A hydrogen explosion occurred on the Unit 1 operation floor,
and high radiation, exceeding 100 mSv/h, was measured on
the rubble. Therefore, in our estimation, FPs discharged
through the following route: containment -> containment top
head flange -> reactor well -> shield plug -> operation floor. It
was also figured that FPs exist unevenly along the FP
migration route.
•

Based on the estimated FP migration route mentioned above, it is presumed
the gaps between shield plugs have high radiation levels. It is also presumed
the entire reactor well, located on the migration route and upstream of theoperation floor, is also highly contaminated.

• In the pressure vessel, FPs are believed to have attached to thestructures and wall surfaces; therefore, high radiation is expected.

• The dose near the pedestal opening (at approx. 1 m fromthe floor surface) was within 9.4 Sv/h (Mar 20-22, 2017)• On the 1F grating: 4.7 Sv/h 9.7 Sv/h (Apr 10, 2015)•

It is presumed that the S/C innerwall is highly contaminated since
the area in the vicinity of theatmospheric control piping, which
was used for PCV ventilation, ishighly contaminated.

• The dose increased toward adeposits and reached about
several Sv/h to more than 10
Sv/h at a location about 0.3–
1.6 m above the floor (Mar 18,
19, 22, 2017)• The max. dose was measured at the upper waterlevel in the torus room, at 920 mSv/h. (Feb 20,
2013)• High contamination is estimated on the building
wall, the S/C wall, or the stagnant water in the S/C.
©International Research Institute for Nuclear Decommissioning

Unit 2: Diagram of
Summarized Information
Reactor well
•
•
•

•
•

Likewise in Unit 3, the gap between the top andmiddle layers of the shield plug may have highlycontaminated areas.
In our estimation, the lower layer and the entire
reactor well are also highly contaminated.
Particulate FPs may exist.

•••

The sealing degraded, resulting in a leak
opening (discharging vapor and FPs).

Upper part of the pressure vessel•

Max dose: Approx. 880 mSv/h (Jun 30,
2012)
The dose is distributed along the direction
from the well to the blowout panel.

••

Reactor building (1F-4F)•

PCV top head•

Operation floor•
The ambient dose rate was several to tens of mSv/h(Apr 2011~ Feb 2014).
High radiation was locally observed at thepenetration.
No significant contamination was detected in the
TIP room, at several mSv/h (Mar 21, 2012).

At some location, the temperature ishigher than in the lower section of the
RPV.
In our estimation, a lot of FPs areattached.
The chemical forms of the attached FPs
are unknown (ex. solubility/insolubility,etc.).
The degree of re-evaporation of theattached FPs is unknown.

•

In-core structures
•
•
•
•
•
•

The PLR system pressure rose when the FDW flow rate increased, presumably because the water leveloutside the shroud rose.
It is presumed that the shroud still exists because the temperature dropped as a result of CS water injection
and also because the water level rose outside the shroud when the amount of water injection was increased.
Analysis results indicate that severe deformation is likely to occur at 1,000°C or greater.
FPs are figured to be attached to the separator and dryer.
Cesium may have been absorbed into the oxide layer of the steel materials.
Cesium may have combined with molybdenum, boron and silicon.

Reactor core••

There is an obstacle inside the TIP piping.
SEM analysis on the obstacle in the TIP piping detected Fe, Cr, Ni and Mn,
which are the elements of iron and steel materials, apart from Zr, anelement of the in-core structures and fuel cladding.
Zero water level in the reactor core area.
Analysis often finds the remaining volume to be zero.
Debris may get solidified at the upper portion of the speed limiter inside the CR
guide tube and may still exist there (if the CRD piping is sound).
Muon measurement indicates there may be fuel in the peripheral area.

RPV lower head
•
•
•

•
•

•
•

Because the CRD and cables remain in the peripheral area of the lower head,
there is no massive damage such as collapse of the entire lower head.
No water level is formed in the RPV, indicating ruptures.
A thermometer was installed through the SLC piping. (The measured
temperature was higher than the injected water’s temperature; it is presumed
that the temperature inside the RPV rose due to fuel debris.) (Since themeasured temperature > injected water temperature, the temperature is assumed
to have risen due to the fuel debris inside the RPV.)The ruptures presumably exist at the center and the peripheral of the lower head.
Temperature response to the changes in the water injection status was relatively
fast, from which it is presumed that the amount of retained water is small.
The RPV and PCV’s temperature responses to the reduction in the water injection
volume in Mar-Apr. 2017 indicate fuel debris may exist at the bottom of the RPV.
From the muon measurement results, it is presumed that high-density substances,seemingly fuel debris, exist.

•

Measurement results and observation data
available
Estimated based on the measurement results andobservation data
Accident analysis or qualitative estimation

•
•
•
•

FDW/CS piping
•

(from
T/B)

•

A temperature decrease was observed at the start of CS
water injection.
Existence of debris is estimated along the route from the CS
through the reactor core to the lower plenum.

RPV lower plenum•

As a result of water injection using the FDW system alone,
the temperature tended to increase by a few degrees C atsome part of the pressure vessel bottom.

•
•

CRD
•
•
•

The CRD and cables still exist in the lower
head periphery.
There was a deposits on the CRD
replacement rail.
There was no visible damage to the CRD
housing support.

PCV vent/exhaust stack
•

•

•
•

High contamination, ranging from several totens of Sv/h, was detected in the joint
exhaust stack for Units 1s and 2, the ventline for Unit 2, and the SGTS for Unit 2 (Mar26-27, 2012).
High contamination, ranging from several
tens of mSv/h to about 1 Sv/h, was detectedin the lower part of the exhaust stack, andfrom 500 mSv/h to 2 Sv/h in the SGTS piping
joint (Aug 6, 2014 to Oct 21, 2015).
Assuming no damage to Unit 2’s rupture disk, it
is likely to have originated from Unit 1.
Contamination stays at the same low level bothat the front and back of the rupture disk. Hence,
it is figured that the ventilation was
unsuccessful.

•••

RCIC/HPCI•

PLR

Drywell
•
•
•
•
•

•
•

The ambient dose rate is higher compared to other units(70 Sv/h).
Fallen powdery matter exists.
The Cs concentration level in the stagnant water in the
PCV is lower than that in the stagnant water in thebuilding, while the Sr and H3 are at the same level.
From the pressure and oxygen concentration levels, it ispresumed that leakage in the gas phase is relatively small.
The max reading of the D/W CAMS was 138 Sv/h at 16:10
on Mar 15, 2011. The trend of CAMS readings suggests thatthe maximum dose of 138 Sv/h occurred because thepressure vessel was damaged and fuel debris fell into thecontainment.
Near the CRD rail, high radiation, ranging from approx.
below 10 to 80 Gy/h, was observed (Jan 26 to Feb 9,2017).
Near the CRD replacement rail and the pedestal opening,a dose ranging from approx. 24 to 36 Sv/h was observed(Aug 12, 2013).

MS piping

RPV pedestal••

Torus room

The CRD and cables still exist in the lower head periphery.
There was no visible major damage to the existing structure (the•CRD exchanger)•
Water injected in the RPV is falling to the pedestal.
There is defective grating on the platform in the pedestal•(possibly due to the heat effect from fuel debris). There areattached depositss at various spots.
•
Near the pedestal inner wall, the dose is below approx. 10 Gy/h•(Jan 30, 2017).
There is no visible major damage to the pedestal inner wall•
surface
There is no visible change near the pedestal inner wall (on the
CRD rail side) at the height from the platform to an area about 2
m below it, at the dose rate level of 7–8 Gy/h, and thetemperature at about 21°C.

•
•
•
•
•

RPV pedestal floor
Drywell floor

CS pump

An obstacle exists in the SLC piping inside the RPV.
A thermometer was installed through the SLC piping. (Since themeasured temperature > injected water temperature, thetemperature is assumed to have risen due to the fuel debris inside
the RPV.)
The heat balance evaluation indicated that about 30% still exists.

No debris was found during the internal PCV investigation (except for asmall number of depositss).
Water level: Approx. 30 cm (located at the lower edge of the PCV ventpipe)
Because no water has been accumulated in the S/C, no water accumulation is
expected in the D/W.

••

Part of the fuel assembly was found to have fallen to thebottom of the pedestal, and the depositss found in its surroundingarea are assumed to be fuel debris.
depositss detected on the pedestal floor may contain fueldebris.

The S/C and torus room have almost the same water level.
The difference in the water levels comes from the difference
between the S/C pressure and atmospheric pressure.
There is some kind of penetration exists between the R/B andturbine building.
The inspection found air dose of 4.3–134.0 mSv/h and thewater dose of 18.7–23.7 mSv/h (Apr 2013).
Cs134 and Cs137 were detected from the collected stagnant
water. On the order of 104Bq/cm 3 (Apr 12, 2013).
The inspection found no water leakage near the vent pipe.

Turbine building•••

The ambient dose rate wasseveral tens of μSv/h (Apr2011 through Feb 2014).
High radiation was detected in
the underground floor (Apr2011 through Feb 2014).
Highly contaminated water isassumed to have flowed inthrough the rupture at the bottomof the S/C.

Suppression chamber•••

The S/C and torus room have almost the same water level.
There is a rupture at the lower level of the S/C (estimated to be O.P.512 or less).
After the core damage occurred in Unit 2, forced depressurization was performed
using the SRV, and a large amount of FPs are assumed to have migrated to the
S/C, as observed by the S/C CAMS. Nonetheless, the FPs stuck in the pooled
water are believed to have migrated to the building through a rupture, which isassumed to exist at the S/C’s lower level.

©International Research Institute for Nuclear Decommissioning

17

Estimate Diagram of Debris Distribution and the RPV andPCV Conditions in Unit 2•

By estimating the energy amount, based on the increased
PCV pressure from hydrogen generation, most fuel isassumed to have been damaged or collapsed. (actual
measurement and analysis)•

Since the temperature decreased during the CS water
injection, fuel is assumed to exist in the reactor core
periphery, where low-flow water injected from the CS could
reach (yet the detailed location of debris cannot be
estimated because a similar behavior is expected if molten
fuel fell onto the fuel support metal fittings or CRGT and was
solidified, which could also act as a heat source). (actual•

Because the PLR system pressure rose when the FDW flow rateincreased, a water level may be formed outside the shroud. (actual
measurement)•

Presumably there is no major damage to the shroud because thetemperature dropped by CS water injection, and the water levelrose outside the shroud when the amount of water injection wasincreased. (actual measurement)••

Muon measurement indicates fuel may exist in the reactor
core periphery. (actual measurement)
Even if some fuel rods are present, they should be just asmall portion of the peripheral area. (general assumption)
It is estimated to be general oxide debris from molten fuelsolidification. (general assumption)

The upper tie plate fell onto the pedestal periphery,
indicating a pressure vessel rupture in the peripheral area.
(presumption based on the actual measurement)••

It is at least presumed that there is a rupture large enoughfor the upper tie plate to fall through. (actual measurement)
The CRGT and CRD in the peripheral area may havepartially melted and collapsed due to fuel debris
accumulated at the pressure vessel bottom. (presumption based(general assumption, testing, and analysis)Concrete mixed debris

The CRGT would remain un-melted with little heattransfer from high-temperature fuel debris. (general•If particulate debris or pellets are present, they may
accumulate in a stagnant spot. (general assumption)•

The holes (which are not so big) in the RPV are estimated toexist in the central and peripheral areas because the CRD in
the peripheral was detected during the internal PCV
investigation and also because of the damage to the grating.

•

The observation revealed that water drops fell all over thepedestal floor even though its severity varies by location;
therefore, small ruptures may exist near the CRD housing at
the pressure vessel bottom. (presumption based on the actual
measurement)•

Some parts of the debris that fell through the ruptures are
assumed to stick to the CRD housing, etc. (general assumption)measurement)•

The fallen upper tie plate was found on the pedestal floor,
and the depositss in its surroundings are assumed to be fueldebris that fell through the RPV ruptures, just as the tie plate
did. (actual measurement)･As a result of damage to the CRGT and CRD housing, a small
amount of fuel debris or molten metal may have entered the
CRD housing. (general assumption and testing)•

There is no observable variation in the dose andtemperature in the area from the pedestal floor to theplatform, and no notable damage has been found in thepedestal sub-structure; therefore, the fuel debris on thepedestal floor seems to have relatively low dose and decay
heat, and may contain large quantities of metals. (presumption•

MCCI would have been limited as there are no signs ofdamage to the PCV shell (no leakage from the sand cushion
drain pipe). (actual measurement)
MCCI may have been limited since the pedestal wall, the cabletray, and the pillar of the CRD exchanger stay un-melted.

based on the actual measurement)

•

•
•
･

depositss containing fuel debris are assumed to havespread all over the pedestal bottom. (actual measurement)
A pool of water on the PVC floor, if any, results in particulate
debris formation. (general assumption)
Particulate debris, if any, is likely to be found in a stagnant
spot. (general assumption)

Damaged CRGT
CRD
CRD (with debris inside)(presumption based on the actual measurement)

Muon measurement showed shadows of high-density
substances, which appear to be fuel debris, located at thepressure vessel bottom. The fuel dropped to the lower
plenum may still exist at the bottom of the RPV. (actual

Particulate debris

CRGT•

on the actual measurement)•

Oxide debris (porous)

Since the increase in fuel temperature might not be so great in the
peripheral area, fuel rod debris and pellets may still exist there.

assumption)•

Remaining fuel rod and its debris

Fuel debris (rich in metals)•

measurement)•

Legendss

ShroudDamaged shroud*
PelletRuptured hole of the RPVUpper tie platedeposits (material unknown)•(actual measurement and general assumption)•

Fuel debris may have solidified, in most cases, without causing
MCCI. (assumption and analysis based on the actual measurements)
The fuel debris that caused MCCI may have been mixed withconcrete. (general assumption)

Ballooning fuel*Oxide debris*
Heavy metal debris*
Particulate pellet*
Cladding residue*
Molten reactor internals*
Solidified B4C*
Control rod mixed melt*

The dash-dotted line shows the inside of the containment is asymmetrical with the pedestal.

* Not used in the estimate diagram of Unit 2©International Research Institute for Nuclear Decommissioning

18

Unit 2: Estimate Diagram of FP Distribution•

Gas may have leaked from the pressure vessel to the D/W during theaccident progression, thereby allowing FPs to easily migrate directly to
the D/W side. This would explain the dose measurement data in the
D/W and S/C in Unit 2.

•

On the Unit 2 operation floor, high radiation was detected at the shieldplug. In addition, the pictures at the time of the accident show a large
amount of vapor was released from the blowout panel. Therefore, FPs
are assumed to have been discharged through the following route:pressure vessel -> containment -> pressure vessel top head flange ->reactor well -> shield plug -> operation floor. It was also figured that FPs
exist unevenly along the FP migration route.

•

Insoluble cesium particles, which seem to have originated from Unit 2,
were detected in the environment (the particle diameter is about a fewμm). In the sample analysis of the protective covering just above theshield plug, particles appearing similar to the insoluble cesium particles
were found; based on this, the reactor well may have been one of theroutes through which the insoluble cesium particles were discharged.
From the protective covering, multiple types of uranium-bearing
particles were detected.

*The estimate was made focusing on cesium, a major source nuclide<Chemical forms and characteristics of major types of cesium estimated>
 Cesium iodide, Cesium hydroxide, Cesium chloride• Easily evaporate and tend to escape from the pressure vessel as vapor due to adifference in pressure or concentration.
• Have a hydration property and tend to move with moisture by, for example, water
vapor condensation or dew condensation (on the wall surface).
 Cesium molybdate, Boric acid cesium
Have a lower vapor pressure compared to the afore-mentioned substances and,hence, tend to stay inside a pressure vessel.
Cesium molybdate may form because of a reaction between cesium and steel
material containing molybdenum.
 Compound of silicon and cesium
Insoluble cesium particles (amorphous particles including cesium with its main
component of silicon oxide)
Believed to be difficult to hydrate, and particles the size of a few microns tend to soarup.
• Products of reaction with steel material (crystalline substances)
Believed to be difficult to hydrate, and tend to stay in the oxide layer on the steel
surface when in the 800–1,000°C temperature range.
••

Hydrated cesium compounds may have penetrated
through the concrete surface in the form of cesiumhydroxide.

•

Although the fuel melt caused a high temperature inside thepressure vessel, the separator and dryer are assumed tohave maintained their shape based on the muon
measurement results of Unit 2.

•

The internal PCV investigation
detected localized high radiation,suggesting an FP concentration
mechanism.

••

Cs134 and Cs137 were detectedfrom the collected stagnant waterin the torus room. (Apr 12, 2013)
Cs134: 1.3E + 04 Bq/cm3
Cs137: 2.4E + 04 Bq/cm3••

Based on the accident progression in Unit 2, alarge number of FPs are assumed to havemigrated to the S/C via the SRV. Nonetheless,most of the FPs stuck in the pooled water are
estimated to have moved out of the containment
through the lower section of the S/C, and it ispresumed that they scarcely exist in the S/Cwater.

••••

Cesium possibly remains inside the fueldebris although its amount is expected tobe small.

••FPs may unevenly exist in fuel debris.

As the fuel temperature increased, boric acid in the control rod material was
possibly taken into the steel melt as a result of the eutectic reaction withsteel materials, suppressing the generation of boric acid.
Generally, as the fuel temperature goes higher, molybdenum evaporation
becomes significant, and cesium molybdate would be generated; as aresult, the ratio of cesium staying inside the RPV would increase.
Nonetheless, in Unit 2, cesium possibly moved out of the pressure vessel
more easily because stable cesium chloride was produced by an injection ofseawater. This could be one reason for the high radiation currently
observed in the containment.

In the early stage of the core heating, insoluble cesium particles weregenerated by a coagulation reaction between silicon oxide and cesiumhydroxide in the gas phase. The produced amount would be limited to theorder of 10 kg.
Some part of the insoluble cesium particles may still exist inside thecontainment or the reactor building.
Depending on the temperature/atmosphere history, some cesium in thevapor phase inside the pressure vessel is thought to be taken into the oxide
layer of the steel material due to the reaction with silicon oxide on thesteam-oxidized steel surface. Considering this factor, the high radiation may
have occurred at the separator and dryer, which have large surface areas.

In Unit 2, the water inside the D/W has leaked out of the containment
through the lower section of the S/C. Cs concentration in the water inside
the D/W should be low because the water is circulating.
Actual measurements (Aug 7, 2013)
Cs134: 2.14E + 03 Bq/cm3
Cs137: 4.38E + 03 Bq/cm3©International Research Institute for Nuclear Decommissioning

19

20
Unit 2: Estimate Diagram of
Dose Distribution•

401 mSv/h–880 mSv/h on the shield plug on theoperation floor (Jun 13, 2012)• On the Unit 2 operation floor, high radiation was detected at
the shield plug. In addition, the pictures at the time of theaccident show a large amount of vapor was released from theblowout panel. Therefore, FPs are assumed to have beendischarged through the following route: pressure vessel ->containment -> pressure vessel top head flange -> reactor well-> shield plug -> operation floor. It was also figured that FPs
exist unevenly along the FP migration route.
•

Based on the estimated migration route mentioned above, it ispresumed that the gaps in the shield plugs have a high radiation level
and that the entire reactor well is also highly contaminated since it islocated upstream of the operation floor on the migration route.

•

In the pressure vessel, FPs are believed to be
attached to the structures and wall surfaces;
therefore, a high dose is expected.

••

31.1 Sv/h–72.9 Sv/h in the gas
phase in the D/W (Mar 27, 2012)• 24 Sv/h–36 Sv/h on the CRD
replacement rail (Aug 12, 2013)

The following doses were detected during
the internal PCV investigation.
• Approx. below 10 Gy/h ~ approx. 70
Gy/h at the space in the upper portionof the CRD rail (Jan 26, 2017)• Approx. 10 Gy/h or below near thepedestal inner wall (Jan 30, 2017)• Around 7–8 Gy/h near the pedestal
inner wall (on the CRD rail-side), fromthe platform level to about 2 m below
it (Feb 2018)• Approx. 70 Gy/h to approx. 80 Gy/hon the CRD rail (Feb 9–16, 2017)•

In Unit 2, the water inside the D/W has leaked outof the containment through the lower section of the
S/C. The dose in the D/W water is presumably low
because the water is flowing through the building’s
main circulation line.
Actual measurement: 1.0 mSv/h or less (Aug 7,
2013)

•
•

4.3–134 mSv/h in the gas phasein the torus room (Apr 11, 2013)
18.7–23.7 mSv/h in the liquid
phase in the torus room (Apr 11,
2013)©International Research Institute for Nuclear DecommissioningUpper part of the pressure
vesselUnit 3: Diagram ofOperation floor
Summarized Information•
The max dose is approx. 2 Sv/h on the shield plug (2013)•

Reactor well•••

A highly contaminated area is assumed to exist
in the gap between the top and middle layers of
the shield plug.
High contamination is also estimated at its
bottom layer as well as the whole reactor well.

•

•
•
•
•

Reactor building (1F-4F)••••

A steam-like substance was observed at the shield plug gap. The
temperature there was about 10°C higher than in its surroundings.
The gate of the spent fuel pool is partially deformed. High radiation wasobserved in its vicinity.
Airborne radiation in the shield plug gap and joint on the operation floor:
Approx. 200–300 mSv/h (2013–2014)
The CUW F/D hatch fell into the pool (possibly due to a hydrogen
explosion).
The center section of the shield plug is deformed (approx. 300 mm).
Radionuclides contributing to the dose: Cs-137 and Cs-134.
Cs-bearing particles were observed during the dust filter inspection.
From the results of the elemental composition analysis, it us presumed
that they originated from Unit 2.

The ambient dose rate was several tens of mSv/h.
Local high radiation was observed at the equipment hatch.
The door of the TIP room was blown off.
The shielding plug at the equipment hatch shifted, presumably
due to propagated pressure from a hydrogen explosion in thebuilding.
The building’s 4th floor was damaged due to a hydrogen
explosion (the northwest section suffered significant
damage).

•••A lot of FPs are believed to have attached.
Chemical forms of the attached FPs are unknown (ex.
solubility/insolubility, etc.).
The degree of re-evaporation of the attached FPs is unknown.

•

•

•
•
•

Analysis results indicate the structures’ downward shift
because of creep and support deformation.
FPs seem to be attached to the separator and dryer.
Cesium has possibly been taken into the steel materials’
oxide layers.
Cesium may have been combined with molybdenum, boron,and silicon.

•
•

•
•

PCV vent/exhaust stack
•

•
•
•

The dose at the SGTS filter train was
several mSv/h (Dec 2011). Contamination ispresumably low, and there would be nobackflow through the filter train.
The estimate suggests that, due to the effectof the vent in Unit 3, hydrogen flowed into
Unit 4 and caused the explosion in the Unit 4building.

•

•
•
•

CS

RCIC/HPCI•Cs concentration in thestagnant water in the HPCI
room (Feb 2017)
Cs-134: 6.8E + 7 [Bq/l]
Cs-137: 4.3E + 8 [Bq/l]•••

The contact behavior of the control rod Position
Indicator Probes (PIPs) was checked, and there
were no PIPs that showed sound conduction
reaction (Sep 2011).
The CRD or cables are presumably severely
damaged.
A deposits of solidified molten material wasobserved on the metal fittings to support the CRD
housing.
A structure that appears to be the CRGT was
observed.

•

•

•
•

FDW/CS piping(from T/B)

•
•
•
•

The water level was approx. 6.5 m from theD/W floor (leakage has occurred from theMSIV room).
From the water level behavior in the D/W, it isfigured that there has been almost no leakage
from the lower part.

•

RPV lower plenum•

The measured temperature was higher than the injectedwater’s temperature; it is presumed that the temperatureinside the RPV rose due to fuel debris.
The temperature at the RPV lower part was higher than
the PCV temperature in the past, but currently theirreadings are almost at the same level.

•

CRD

Due to the dropped fuel debris, damages are estimated on the grating, TIP piping, CRD exchanger, etc.
Water injected into the RPV is presumably dropping into the pedestal.
Damaged multiple structures were observed inside the pedestal, with some substances thatseem to be solidified molten material (Jul 2017).
Structures that are expected to be, but not limited to, a speed limiter or upper tie plate were
found.
Coarse surfaces were observed on some walls inside the pedestal.
A substance that is seemingly a result of molten material solidification was observed near apedestal wall.
Cables of some thermometers at the RPV bottom were found to be broken. Those thermometers
are believed to be measuring PCV temperatures.

Torus room••••

The water level in the torus room was about
O.P.3000.
No corrosion was observed on the catwalk
handrails.
Water vapor leaking into the torus room is•presumably minor.
Given the stagnant water level, there may besome kind of penetration between R/B and•turbine building.

•••Hydrogen caused by MCCI may have contributed to the hydrogen explosion in Unit
3, and fuel debris may have dropped to the pedestal floor.
Unlike Unit 1, no contamination has been detected in the RCW system.
Multiple fallen objects and pebble or sand depositss were detected in thelower part of the pedestal.
A deposits was observed near the workers’ access opening in the pedestal.

•

Leakage occurred around theextension joint of the main
vapor pipe D.
The leak location is presumably
the main vapor pipe D only,
according to the camera
inspection and to the observationof the water flow on the floor.

Turbine building
Ambient dose rate was several-hundred
μSv/h to several mSv/h. • There was highradiation in the underground floor.
Highly contaminated water presumably flowed
into the turbine building from the MSIV room
during the early phase of the accident.

RPV pedestal floor•

MS piping•

PLR

Drywell

RPV pedestal
•
•
•

•

・ CS water injection started (in Sep 2011)
・ Water injection from the CS system wasstopped for 15 days (Dec 2013).
・ The volume of water injection was
reduced (Feb-Mar 2017).

The internal investigation showed no damage to the structures around the X-53 penetration.
The airborne radiation in the D/W was relatively low (approx. 1 Sv/h).
Cs concentration in the stagnant water in the PCV is lower than that of the stagnant water in the building, while Sr andH3 are at the same level. An alpha nuclide was detected (Oct 2015).
Based on the containment oxygen concentration (approx. 8%) and the D/W pressure, Unit 3 is assumed to have the highest
leak level in the PCV’s gas phase among other units (presence of air in-leak is assumed).
The temperature measured during the internal PCV investigation was approx. 26–27°C in the gas phase and approx.
33–35°C in the water.• Cs134 concentration [Bq/cm3]: 4.0E + 2 (near the water surface), 2.3E + 2 (approx. 0.7 m below
the water surface)Cs137 concentration [Bq/cm3]: 1.6E + 3 (near the water surface), 9.4E + 2 (approx. 0.7 m below the water surface)21

A temperature decrease was observed at the start of CS water injection (Sep2011).
Existence of debris is estimated along the route from the CS through the reactor
core to the lower plenum.
Meanwhile, no temperature increase occurred when the CS system wasstopped for 15 days (Dec 2013); instead, the FDW system alone was used forwater injection.
The temperature increase occurred when the amount of water injection wasincreased in May 2011.
No water level has been detected in the RPV.
Debris may have solidified and, hence, still may exist in the upper portion of the
speed limiter inside the CR guide tube (if the CRD piping was sound).
The muon measurement results (interim report) indicate part of the fueldebris still may exist, but no presence of huge, high-density material has
been detected in the reactor core (Jul 2017).
The muon measurement results indicate there may be no large mass of fueldebris in the original reactor core area.

Drywell floor

CRD
•

•

•

The sealing has degraded, and a leakage opening exists (fromwhich vapor and FPs were discharged).

Slow temperature response to the changes in the water injection status
The muon measurement results (interim report) indicate part of the fuel debris
still may exist, but no presence of huge, high-density material has been detected
at the pressure vessel bottom (Jul 2017).
The muon measurement results suggest that part of the fuel debris may exist at
the bottom of the RPV; however, the possibility is uncertain.
Inside the pedestal, water surface fluctuations were observed in the center andperipheral of the RPV; therefore, ruptures may exist in the center and peripheral of the
RPV.

•

Measurement results and observation
data available
Estimated based on the measurement
results and observation data
Accident analysis or qualitative estimation

Reactor core

RPV lower head•••

In-core structures

PCV top head••

Suppression chamber•

No water level has been
detected in the S/(but it ispresumed to be almost full).

©International Research Institute for Nuclear DecommissioningEstimate Diagram of Debris Distribution and the RPV andPCV Conditions in Unit 3
•
•

•

•

•
•

•

•

•
•

From an estimate energy amount based on the PCV pressure
increase caused by the generated hydrogen, it is presumed thatmost of the fuel melted. (actual measurement and analysis)
There was no temperature increase in each section of the RPV
when the CS system was stopped from December 9 through24, 2013 (the flow rate from the FDW was increased, and thetotal volume of water injection had remained stable). Based on
this finding, there should be little fuel debris in the reactor core(less than that in Unit 2). (actual measurement)
In addition to the above finding, the temperature at the lower
part of the RPV decreased when the total water injection
volume increased as water injection from the CS system began(on Sep 1, 2011); hence, fuel debris is probably in the lower
plenum. (actual measurement)
The muon measurement results indicate there may be no large
mass of fuel debris in the original reactor core area. (actual
measurement)

Because a structure, seemingly CRGT, fell out of thepressure vessel, it is presumed there is a rupture large
enough for the CRGT to fall through. (actual measurement)
Inside the pedestal, there were water surface fluctuationsin the center and periphery of the RPV; therefore,
ruptures may exist in the center and periphery of the
RPV. (actual measurement)
Since there is a gap between the flange faces at thelower part of the CRD housing, part of the welds at the
CRD housing and the pressure vessel bottom may not be
firmly fixed. (presumption based on the actual measurement)

The internal PCV investigation revealed that the damage
inside the pedestal had progressed further than in Unit 2,
suggesting that the unit’s PCV has more dropped fueldebris than Unit 2’s. (actual measurement)
Damage to the platform is presumably due to the drop of
high-temperature debris. (actual measurement)
Damage and depositss (the latter seemingly a result ofsolidified molten material solidification) were observed on
the metal fittings to support the CRD housing; fuel debris
may exist above and below it and also in its surroundings.

•(general assumption and analysis)•

A pool of water on the PCV floor,
if any, may have formed
particulate debris. (generalassumption)･

Particulate debris, if any, is likely
to be found in a stagnant spot.
(general assumption)

Oxide debris (porous)
Particulate debris

Since the increase in fuel temperature might not beso great in the peripheral area, fuel rod debris andpellets may still exist there. (general assumption, testing,
Even if some fuel rods are present, they should bejust a small portion of the peripheral area. (general•

It is estimated to be general oxide debris from molten
fuel solidification. (general assumption)•

Part of the CRGT remains un-melted when there is alittle heat transfer from high-temperature molten
debris. (general assumption)•

If particulate debris or pellets are present, they may
accumulate in a stagnant spot. (general assumption)•

The muon measurement results suggested that partof the fuel debris still may exist at the bottom of the
RPV; however, the possibility is uncertain. (actual
measurement)••

As a result of damage to the CRGT and CRD housing,a small amount of fuel debris or molten metal may
have entered the CRD housing. (general assumption and

Fuel debris (rich in metals)*

Concrete mixed debris
CRGT
Damaged CRGT
CRD
CRD (with debris inside)
ShroudDamaged shroud*
Pellettesting)

Ruptured hole of the RPVPart of fuel debris may have solidified without MCCI.

Upper tie plate*(general assumption)•

Considering that D/W was sprayed for more than anhour, from 7:39 on March 13 as part of the accident
response, water may have accumulated on the D/Wfloor when the pressure vessel suffered damage, andit may have prevented fuel debris from spreading
further. (actual measurement and general assumption)
Fuel debris spread out of the pedestal through thepedestal opening, but presumably no shell attack
occurred. (actual measurement and analysis)

deposits (material
unknown)*

Ballooning fuel*Oxide debris*
Heavy metal debris*
Particulate pellet*(actual measurement)･Remaining fuel rod and its debris

assumption)•
The lower part of the pedestal has some substances,seemingly a result of molten material solidification, andother dropped objects such as grating and depositss.

22

and analysis)•(actual measurement)•

At present, both possibilities exist for the
shroud: it could be sound or damaged.

Legends•

An explosion occurred in Unit 4
and then in Unit 3, possibly due
to hydrogen generated by
MCCI. (actual measurement)

Cladding residue*
Molten reactor internals*

Solidified B4C*
Control rod mixed melt*
The dash-dotted line shows the inside of the containment is asymmetrical with the pedestal. * Not used in the estimate diagram of Unit 3©International Research Institute for Nuclear Decommissioning

Unit 3: Estimate Diagram of
FP Distribution*The estimate was made focusing on cesium, a major source nuclide• The gamma-ray spectra measurement on the operation floor in Unit 3
indicated high dose levels in the shield plugs’ gaps and joints. The
pictures at the time of the accident also show a large amount of vapor
was discharged from the damaged reactor building.
• Therefore, it is presumed that FPs were discharged through thefollowing route: pressure vessel -> containment -> containment top
head flange -> reactor well -> shield plug -> operation floor. It was also
figured that FPs exist unevenly along the FP migration route.
• Hydrated cesium compounds may have penetrated
through the concrete surface in the form of cesiumhydroxide.
• Although the fuel melt caused a high temperatureinside the pressure vessel, the separator and dryer are
highly likely to have retained their shapes, according to
the muon measurement results of Unit 3.

<Chemical forms and characteristics of major types of cesium estimated>⚫ Cesium iodide, Cesium hydroxide, Cesium chloride•Easily evaporate and tend to escape from the pressure vessel as vapor due to adifference in pressure or concentration.
•Have a hydration property and tend to move with moisture by, for example, water
vapor condensation or dew condensation (on the wall surface).
⚫ Cesium molybdate, Boric acid cesium
Have a lower vapor pressure compared to the afore-mentioned substances and,hence, tend to stay inside a pressure vessel.
Cesium molybdate may form because of a reaction between cesium and steel
material containing molybdenum.
⚫ Compound of silicon and cesium- Insoluble cesium particles (amorphous particles including cesium with its
main component of silicon oxide)
Believed to be difficult to hydrate, and particles the size of a few microns tend tosoar up.
- Products of reaction with steel material (crystalline substances)
Believed to be difficult to hydrate, and tend to stay in the oxide layer on the steel
surface when in the 800–1,000°C temperature range.

• Depending on the conditions of the gas phase, such as the water
vapor/hydrogen ratio, cesium molybdate or boric acid cesium may have
served as major chemical species, suppressing their discharge to theoutside of the pressure vessel.

• The level of activity at 0.7 m below the surface of thestagnant water in the PCV was lower than that near thewater surface. The specimens collected at the two locations
near the water surface of the stagnant water in the PCV, andabout 0.7 m below the water surface, contained Cs134,
Cs137, tritium, and Sr90.(Oct 30, 2015)• Cs134 concentration [Bq/cm3]: 4.0E + 2 (near the water
surface), 2.3E + 2 [7] (approx. 0.7 m below the water
surface)• Cs137 concentration [Bq/cm3]: 1.6E + 3 (near the water
surface), 9.4E + 2 [7] (approx. 0.7 m below the water
surface)•

In the early stage of the core heating, insoluble cesium particles
were generated by coagulation reaction between silicon oxide andcesium hydroxide in the gas phase; however, it was estimated thatthe amount generated is smaller than that in Unit 2.

•• The FPs that migrated to the S/C and got stuck in
the pooled water may still exist in the stagnant
water.
• A huge amount of FPs may still exist, given thelarge volume of stagnant water in the S/C.

23

• Cesium possibly remains inside thefuel debris although its amount is
expected to be small.

Depending on the temperature/atmosphere history, some cesium in thevapor phase inside the pressure vessel is thought to be taken into theoxide layer of the steel material due to the reaction with silicon oxide on
the steam-oxidized steel surface. Considering this factor, the highradiation may have occurred at the separator and dryer, which havelarge surface areas.

• FPs may unevenly exist in fuel debris.

©International Research Institute for Nuclear Decommissioning

Unit 3: Estimate Diagram of Dose Distribution•

24

Airborne radiation near the gaps and joints of theshield plugs on the operation floor: approx. 200–300
mSv/h• The gamma-ray spectra measurement on the operation
floor in Unit 3 indicated high dose levels in the shield plugs’
gaps and joints. The pictures at the time of the accident also
show a large amount of vapor was discharged from thedamaged reactor building.
• Therefore, it is presumed that FPs were discharged throughthe following route: pressure vessel -> containment ->containment top head flange -> reactor well -> shield plug > operation floor. It was also figured that FPs exist unevenly
along the FP migration route.

• Based on the estimated migration route mentioned above, it ispresumed that the gaps between shield plugs have a high radiationlevel.
• It is also presumed that the entire reactor well, located on the
migration route and upstream of the operation floor, is also highly
contaminated.

•• In the pressure vessel, FPs are believed to be attachedto the structures and wall surfaces; therefore, a highdose is expected.

The dose in the PCV gas phase was approx. 1
Sv/h near the PCV wall surface, and approx.
0.75 Sv/h at a point about 550 mm away fromthe X-53 penetration exit. (Oct 30, 2015)• The FPs attached to the D/W wall surface may havefallen along the wall into the stagnant water in the
PCV; therefore, a high dose could be expected near
the water surface.

• Because the leakage of the PCV of Unit 3 is locatedaround the extension joint of the main vapor pipe D,
the dose in the lower part of the PCV’s stagnant
water could be higher than that on the water surface.
©International Research Institute for Nuclear Decommissioning

25(1)-[2] Establish a Database Necessary for Comprehensive Analysis and Evaluation[Overview and Objectives]•Promote the enrichment of the database by maintaining and managing it and updating its contents in a continuous and timely
manner so that the latest measurement data and research results can be viewed, analyzed, and downloaded from the database.
Also, upgrade the display functions to present the analysis results in a more user-friendly manner.
•The website for the OECD/NEA BSAF project (English site) is run separately; however, since the Japanese and English sites
have a lot of information in common, the two sites will be managed in an integrated manner. By doing so, information sharing in
English will be enhanced.
On the “Document Search” window, your search can be narrowed by data type, location, etc., as the same as the[FY2017 Research Details]
Japanese version.
• Improve the operability of the graph display functions(including the zoom function)• Integrated management of the English and Japanese databases• Timely update and management of the contents[Final Report]• Detailed display range settings for the graph display function wereintroduced. A display selection function to provide an easy-to-view
display of the measurement data collected within the first three
weeks after the accident was also introduced.
•The English version of the document search function to enable thesearch of English documents on the internal PCV investigations,etc., was developed.
•The timely update and management of the database contents wasintroduced.
•By implementing the afore-mentioned measures, making thecomprehensive analysis and evaluation mentioned in (1)-[1] more
efficient was contributed to.
[Database URL]
https://fdada.info/

The availability of the English/Japanese documents and links
will be shown.

Figure: Developing the English version of Document Search©International Research Institute for Nuclear Decommissioning26(2)-[1]-1 Sensitivity Analysis of the Concrete Erosion at the Pedestal
Drywell floor[Overview and Objectives]
Based on the results of the investigation inside Unit 1, a sensitivity
analysis of MCCI was conducted to help estimate the fuel debris
conditions in the pedestal.

Case 1(Spreading to the D/W)

PCV shell

Sump pit[Final Results]• The pedestal erosion tendency for each fuel debris fall locationand deposits state was grasped. The tendencies indicated thepossibility that, even in a situation where no water is injected froma fire engine, no major pedestal erosion occurs, although it may
depend on the concrete’s properties.
• The sensitivity analysis results were reviewed, including thoseobtained up until the last fiscal year, and reflected the pedestal
erosion tendencies in the fuel debris distribution diagram.

Fuel debris
Case 2(Eccentric drop)

Case 3(Central drop)Figure 1. Estimated condition of fuel debris depositss

3000
Temperature (K)[FY2017 Research Details]• Since the internal PCV investigations indicate that the pedestal isstanding on its own, the key parameters that affect concrete
erosion using the MAAP codes were studied.
• The significant effect of a concrete melting when no water was
injected from a fire engine was verified and reflected the finding in
the analysis conditions for long-term erosion.
• By using three-dimensional codes, long-term erosion analysis was
performed to consider the differences in the falls of fuel debris andtheir resultant deposits formulation.

Pedestal wall

Concrete melting point

2500

2000

Max/average debris temperature

1500

Erosionstopped1000

Debris surface
temperature

500

Structure material
temperature

0

0

2

4

6

8

Time after the fuel debris started to drop (d)Figure 2. Temporal change in fuel debris temperature and structurematerial temperature (Case 2)©International Research Institute for Nuclear Decommissioning27(2)-[1]-1 Sensitivity Analysis of the Concrete Erosion at the Pedestal[Application to the estimate diagram of fuel debris distribution] Fuel debris shape and concrete erosion behavior[Solid-liquid distribution][Temperature distribution]

t = 0.1 hourt = 0.1 hour[1][2]Sump pitt = 7 dayst = 7 days

Cavitation due to remelting➢➢

Even if fuel debris depositss solidify, they may remelt becauseof decaying heat and flatten.
It is possible that no major pedestal erosion occurs, even
without water injection from a fire engine, although it depends
on the concrete’s melting point. (Case 1 and 3 showed the sametendency.)
Molten debris

Concrete floor

Crust

Concrete outside the shell

500

1,000

1,500 2,000 2,500
Temperature (K)Figure 3. Case of eccentric fall and depositsion (Case 2)[1] t = 7 days, 0.45 m
below the floor➢[2] t = 7 days, 0.15 m
below the floor

Erosion of the pedestal wall is minor in
the areas away from the sump pit.

Eroded area (concrete -> fuel debris)
Pedestal wall/base and PCV shell

Figure 4. Review on the damage condition of the pedestal base and wall (Case 2)©International Research Institute for Nuclear Decommissioning

28(2)-[1]-2 Inverse Problem Evaluation Using Virtual Reactors and Compilation of a Database[Overview and Objectives]
From right after the accident until today, huge and various kinds of data were released to the public, including the temperatureand pressure data inside the PCV. By solving inverse problems from those data, fuel debris locations are estimated.
[FY2017 Research Details]
Virtual reactors were created by making use of three-dimensional CFD codes (StarCCM+). In addition to the creating more
RPV models, the temperature distribution from one month to six months after the accident was evaluated, with the forward
problem analysis approach, and the fuel debris locations inside the plant was estimated.
[Final Results][Unit 1]
Almost no heat source inside the RPV is
expected. Meanwhile, given the leakage ofsuperheating vapor from the RPV near thesafety valve, the presence of a heat sourcein the RPV needs to be assumed.

Estimated heatsource distribution
Apr 6, 2011
Estimatedtemperature
distribution[Unit 2]For an assumed situation where most of
the heat source exists in the RPV andsome 0.8MW (about 25% of the estimatedtotal heat generated) is present in thepedestal, the temperature distributions inthe steady and transient states weresuccessfully simulated.

Estimated heat source
distribution
Apr 1, 2011
Estimated temperature
distribution[Unit 3]
The temperature distribution suggested thepossibility that the heat source of about 0.25–0.3
MW (about 25–30% of the estimated total heatgenerated) is present in the lower part of the
RPV. According to a calculation, a significant
amount of heat, which is estimated to be 0.8MW(81% of the estimated total heat generated) or
greater, was removed by feedwater.

Estimated heat source
distribution
Jun 30, 2011
Estimated temperature
distribution(above the water surface)©International Research Institute for Nuclear Decommissioning

(2)-[1]-3-(a)-1 Estimate of the Amount of Fuel Debris Remaining in the Pedestal Space Obstacles

29[Overview and Objectives]
Fuel debris may have attached to structures, such as the grating and the penetration pipe, in the pedestal space.
From the Unit 2 internal PCV investigation results, thermal balance data, etc. the amount of fuel debris stuck in those obstacles
are estimated.
[FY2017 Research Details]
Through comprehensive analysis of temperature distribution and temperature behavior upon changes in the water injection
operation, the presence or absence of fuel debris in the pedestal space was evaluated. A heat transfer analysis was also
conducted by referring to the results of the internal PCV investigation held in January and February 2017 and to the accident
scenario, and how much fuel debris may have got stuck in the structures near the CRD housing was evaluated.
[Final Report]
Since the temperature in the lower part of the CRD housing is almost the same as that of the D/W (Figure 1), it is estimated that
there are no large amount of fuel debris near the CRD housing. Possible locations on the structures where fuel debris may havegot stuck (Figure 2) were extracted and a heat transfer analysis (Figure 3) was performed. The results indicate the possibility thatthe structures did not melt and that a small amount of fuel debris still exists.
A

Lower part of the RPV

Melting point of thestructure material

Lower part of the CRD housing
Return air drywell cooler
Supply air drywell cooler

Max structure material temperature [°C]

CRD
replacement
rail

Example ofanticipated
fuel debris

No temperature
difference between
the lower CRD
housing and theD/W

Figure 1. Temperature distribution near
the pedestal (Jun 1, 2015)

B

C
Figure 2. Structures near
the CRD housing

Initial conditions of fuel debris
Temperature: 1,500C–2,500C
Thickness: 3cm–7cm
Not reach the melting point in all cases

Time [s]Figure 3. Example of heat transfer analysis(hypothesized fuel debris in Figure 2’s A)©International Research Institute for Nuclear Decommissioning

(2)-[1]-3-(a)-2 Detailed Evaluation of Debris Behavior inside the Lower Head after Damage of the Lower Head

30

Penetration depth of fuel debris in the CRD[m][Overview and Objectives]
It is important to find out whether fuel debris still exists inside the CRD housing in the lower part of the RPV for debris removal
purposes.
[FY2017 Research Details]
Based on the specimen cutting inspection in the CRD housing melting test, the model parameter was optimized and the fueldebris behavior in the lower head (Figure 1) was analyzed. To evaluate the presence of fuel debris near the CRD and based on
the KAERI test results, a sensitivity analysis was performed with the detailed meshes showing fuel debris temperatures, thepresence/absence of water in the lower head, decay heat, etc. (Figure 2 and 3).
[Final Report]
The amount of fuel debris penetrating the CRD housing largely depends on the fuel debris temperature when melting occurredin the upper part of the CRD housing and, hence, involves significant uncertainties. The detailed analysis results showed that ahigh temperature persisted near the weld, even when water was present in the lower head, indicating that fuel debris may haveflowed out through the gap under the weld of the RPV and CRD housing.

DebrisHigh temperaturelevel near thewelding part.

Temperature (K)Water

Temperature (K)CRD housing

RPV wallArea 1

Area 2

Area 3

Area 4CenterFigure 1. Analysis of fuel debris
behavior in the Unit 3 lower head

Inside the CRD
housing

Periphery

Figure 2. Initial fuel debris
distribution according to thedetailed analysis

Fuel debris
possibly flowed out

CRD housing melted,
allowing fuel debris
penetrate inside

Figure 3. Fuel debris distribution
after the melt of CRD housing©International Research Institute for Nuclear Decommissioning

(2)-[1]-3-(a)-3 Evaluation of Condensation Behavior of Water Vapor Containing Hydrogen in the S/P

31[Overview and Objectives]
Vapor condensation behavior (a phenomenon of incomplete condensation) in the S/C is important in identifying the accident
progression. Especially, the behavior of mixed gas of hydrogen and vapor blowing into the S/C is important for evaluating the
PCV/RPV pressures. Through evaluating this behavior, and projecting the amount of generated hydrogen, etc., a reduction in
the the uncertainties of fuel debris migration behavior is sought.
[FY2017 Research Details]
By combining SAMPSON and the CFD tool POOL-3D with the temperature distribution and the condensation experiment on
steam containing non-condensable gas (LINX), the condensation behavior of water vapor containing hydrogen in the S/C isstudied. Based on the study results, the distribution of S/C water temperatures are provided, a key parameter regarding
condensation behavior, in the three-week analysis “(2)-[1]-3-(a)-7.”[Final Report]
The three-dimensional temperature distribution inside the S/C during the period in which vapor and hydrogen were discharged
from the SRV, was evaluated. It was found that, in Unit 2, there was a temperature difference of approx. 30°C up until the RCIC
stopped (70 hours after the scram). After forced depressurization (75 hours after the scram), the temperatures inside the S/Cequalized; however, it was found that temperature stratification occurred again due to vapor and hydrogen erupting from theSRV. It was found that, in Unit 3, temperature stratification occurred while the RCIC was in operation and it persisted while the
HPCI was working. It as also found that, after the HPCI stopped, temperature stratification was resolved by a huge amount ofhydrogen generated during the progression of core melt.

Unit 3

Unit 25 h (RCIC and SRV operation)

70 h (Stop of RCIC)

75 h (RPV depressurization)

80 h (Beginning of 2nd peak)

42 h (ADS activation)

30 h (steam injection)

35 h (steam + H2 injection)

Figure 1. Unit 2 temperature stratification at indicative event

Figure 2. Unit 3 temperature stratification at indicative event©International Research Institute for Nuclear Decommissioning322)-[1]-3-(a)-4 Event Sequence Analysis at the Time of Core Material Slumping[Overview and Objectives]
In all units of Fukushima Daiichi NPS, the core materials are believed to have relocated to the lower plenum (slumping) and to the pedestal.
The core state during the slumping affects the RPV failure mode and has a significant impact on the eventual material distribution and theirproperties. The corresponding accident progression is therefore determined by conducting the SA code analysis and using the plant data(such as pressure) to improve the accuracy of the evaluation of the final materials distribution and their properties.
[FY2017 Research Details]
A detailed MELCOR code analysis (Waseda Univ.) of the accident progression in Units 2 and 3 was performed, with a focus on theconsistency with the plant data. The most probable transitions and uncertainties was also evaluated by comparing the results with the
MAAP (EPRI), SAMPSON (IAE) and SCDAPSIM (JAEA) results all the while shedding light on core energy.
*Core energy: decay heat + oxidation reaction heat - heat transfer to the coolant or structure material
Major slumping in Unit 3

Core energy in Unit 2(the best estimate)Core energy in Unit 3(the best estimate)

Estimated initiation of core damage

Increase in core energy (GJ)

[Final Results]
The following knowledge was obtained and provided
in the estimate diagram of fuel debris distribution.
➢ Based on the evaluation of core energy, it was
concluded that most of core fuel did not melt whenslumping occurred in Unit 2 (the best estimate). It was
concluded, meanwhile, that in Unit 3, where it took along time from the initiation of core damage toslumping, a considerable amount of fuel melted due toaccumulated energy (the best estimate) (Figure 1).
➢ Based on the energy state described above, the effecton the heat transfer behavior of fuel debris andstructural materials in the lower plenum (Figure 2)
was assessed and a result matching the internal
investigation’s results was obtained.
Reflection in the estimate diagram of fuel debris
distribution
The combined knowledge with the results of the “simulated
fuel assembly degradation test” indicates a considerable
amount of un-melted fuel pellets may still exist in thereactor core or at the lower plenum area in Unit 2. The
knowledge indicated a relatively large mass of solidified
fuel may still exist in the pedestal or CRD area in Unit 3.
The Unit 3 data analysis results also indicate thepossibility that MCCI was mitigated by the water in the
PCV.

Estimated time forthe core material
to migrate to thepedestal in Unit 3

Estimated time for the core material to migrate to
the pedestal in Unit 2 (with great uncertainty)

First slumping in Unit 2

Estimated time elapsed after core voiding (hr)Figure 1. Comparison of the accumulated core energy (the best estimate) and timing of core material
migration to the lower plenum and pedestal in Units 2 and 3(Imaged) Heat transfer from slumping fuelto structure material in Unit 2(Imaged) Heat transfer from slumping fuelto structural material in Unit 3

Legends
Molten metal

Liquid phase water existed initially
Solid fuel pieces

Pool ofmolten metal

Solid fuel block(containing molten metal)

Liquid phase water
Liquid phase water existed
initially

Vapor bubbles
Solid fuel block(containing molten metal)

Solid fuel pieces
Heating up and damaging the
RPV side due to convective
heat transfer

Melting structure material and downward
fuel migration due to heat transfer©International Research Institute for Nuclear Decommissioning

Figure 2. Effect on the heat transfer behavior of fuel debris and structure material estimated according to the energy during slumping

(2)-[1]-3-(a)-5 Three-Dimensional Evaluation of MCCI Behavior

33[Overview and Objectives]
A quantitative evaluation of the state of concrete erosion caused by MCCI that is assumed to have occurred in the pedestal was
performed.
[FY2017 Research Details]
In the last fiscal year, a sensitivity analysis on Unit 1 was conducted, finding that the volume and temperature of dropped fuel debris
are the key parameters of fuel debris diffusion. In this fiscal year, with regard to Units 2 and 3, a three-dimensional evaluation on the
MCCI behavior under a condition where water remains on the pedestal floor was performed. A sensitivity analysis using parameterssuch as the remaining water volume was also conducted.
[Final Report]⚫ It was verified that, in Unit 2, the whole fuel debris would remain in the pedestal if the dropped debris was 20 tons (12.8 tons offuel debris) or 40 tons (25.6 tons of fuel debris) and concrete erosion stopped within 10 hours.
⚫ With regard to Unit 3, an analysis for a condition where the dropped debris was 140 tons (89.4 tons of fuel debris) was
performed, by setting the sensitivity parameters at the initial water level and the drop velocity of fuel debris. The results of theanalysis with SAMPSON/DSA showed the initial water level had a limited impact on the diffusion of fuel debris, while the drop
velocity of debris had a large impact on fuel debris diffusion.
Drop volume: 20 tons
Drop velocity: 300 kg/s

Drop volume: 40 tons
Drop velocity: 300 kg/s

Figure 1. Sensitivity evaluation of dropped fuel debris
volume (Unit 2)

Initial water level: 18 cm
Drop velocity: 450 kg/s
Water injection volume: 1.5 kg/s

Initial water level: 36 cm
Drop velocity: 450 kg/s
Water injection volume: 1.5 kg/s

Figure 2. Sensitivity evaluation of initial waterlevel (Unit 3)

Initial water level: 36 cm
Drop velocity: 450 kg/s
Water injection volume: 5 kg/s

Initial water level: 36 cm
Drop velocity: 10 kg/s
Water injection volume: 5 kg/s

Figure 3. Sensitivity evaluation of drop
velocity (Unit 3)©International Research Institute for Nuclear Decommissioning

(2)-[1]-3-(a)-6 Confrimation of Factors Influencing Debris Distribution

34

Evaluation of the two-dimensional temperature distribution based on the SAMPSON
three-week analysis
1.
The fuel rod temperature increased as shown in the figure; however, it ismore likely that the supporting strength in the lower part of the fuel rods
was lost before the fuel failure occurred.
2.
When the emissivity and the amount of hydrogen generation to an extent
consistent with the knowledge and actual measurements in the past waschanged, a result in which part of the fuel at the outermost ring collapsed wasgotten.

Surface temperature of fuel rod (K)[Overview and Objectives]
A quantitative evaluation of the effect on fuel debris distribution is performed, assuming the fuel rods collapsed before melting, toensure reliability of the fuel debris distribution information that could be obtained through the accident progression analysis, thus
reducing uncertainties.
[FY2017 Research Details]
A model representing the pre-melt collapse of fuel rods was introduced and a quantitative evaluation on the effect of thisphenomenon on the fuel debris distribution was conducted. The fuel debris distribution was comprehensively evaluated while
taking into account the results of the two-dimensional temperature distribution evaluation that studied the output per fuelassembly.
[Final Report]• In the VERCORS test, a fuel rod collapsed at approx. 2,600 K, meaning that it would take about 100 hours for a Larson-Miller
parameter model to collapse when the fuel rod temperature is around 2,300 K. Therefore such factors as the loss of metal
cladding, due to melting and oxidation, and the vibration of fuel rods, due to the sudden generation of vapor, were taken into
account and a model using the fuel rod temperature and time history for the SAMPSON code was introduced to simulate a fuelrod that collapses within a few hours even at 2,300 K.
• To evaluate the possibility of localized collapse of a fuel rod that cannot be evaluated through the SAMPSON analysis, thetemperature distribution at the horizontal section of Unit 2’s core was evaluated. For the latter evaluation, factors such asradiation and the Zr-water vapor reaction were taken into account, getting the following results.

Temperature at which
the strength of the lower
structure decreases

One-sided
radiation, k=1
Two-sided
radiation, k=1

Time after scram (h)Presumably the support structure at the lower part of a fuel rod starts losing
its strength before the fuel rod starts collapsing.
Also presumably, a fuel rod collapses downwards when the fuel falls toward
the center; or collapses onto the shroud when the fuel falls toward the
shroud.

Area 1(Core center)

Area 2

Area 3

Area 4(Outermost peripheral
part of the core)

Evaluation results of two-dimensional temperature distribution (at 82 hrs)
©International Research Institute for Nuclear Decommissioning

(2)-[1]-3-(a)-7 Analysis of Each Unit over Three Weeks after the Accident

35[Overview and Objectives]
Analysis was performed for up to three weeks, by which time the MCCI was expected to have reached a completely steadystate, in order to evaluate the final fuel debris distribution all the while taking uncertainties into account.
[FY2017 Research Details]
Sensitivity parameters affecting the distribution, composition, and properties of fuel debris were extracted and a sensitivity
analysis was performed. A three-week analysis was performed with each sensitivity parameter group that matches the actual
measurements obtained from the units and the results of the on-site inspections including the internal PCV investigations.
[Final Report]⚫ To reduce uncertainties in the accident progression scenario, which is the prerequisite for evaluating fuel debris migration
behavior and the final fuel debris distribution, accident progression scenarios to the most likely ones were narrowed down.
This process was completed by comprehensively evaluating results of studies conducted inside and outside Japan, such as
internal investigations using muon and robots, measurements of pressure and water levels in the reactors, the knowledge
and scientific considerations gained or made until today, and BSAF.
⚫ This initiative was conducted by IAE, the implementation organization for this particular study item, and in cooperation withthe parties involved in the Project (JAEA, Hitachi GE, Toshiba, and TEPCO) through discussions at study meetings (whichwere held six times).
⚫ The SAMPSON analysis results, which were obtained based on this accident progression scenario, matched well with theactual measurement results, significantly contributing to reducing the uncertainties for the final evaluation of the fuel debris
distribution.

RPV pressure behavior (Unit 1)

RPV pressure behavior (Unit 2)

RPV pressure behavior (Unit 3)©International Research Institute for Nuclear Decommissioning

(2)-[1]-3-(a)-7 Analysis of Each Unit over Three Weeks after the Accident

36[Final report (evaluation results of Unit 1)]⚫ From the discussions made at the accident-scenario study meetings and the SAMPSON analysis results, it was presumed thatthe fuel rods started fracturing about 4 hours after the scram and started melting in about 5 hours. Based on thesepresumptions, fuel debris fell on the lower plenum in about 11 hours and further down to the pedestal in about 15 hours. It was
concluded, therefore, that the fuel rods and fuel debris (containing fuel and structure materials) stayed in the reactor core area
for about 5 hours in the high-temperature state due to decay heat and oxidation reaction.
⚫ The VERCORS and CMMR tests indicate the possibility that the fuel rods maintain their column shape up to around 2,600 K.
Considering the melting point of the channel box covering the fuel rods is 2,100 K (or 1,500K or even lower due to eutectic
formation), the shroud would be directly exposed to radiation from fuel rods when their temperature far exceeds the melting
point of 1,700 K. For Unit 1, no water injection was conducted while core damage progressed; therefore, it was concluded thatthe water in the down-comer may have fallen to the BAF or lower in 5 to 7 hours after the scram.
⚫ The SAMPSON analysis reached the same conclusion, indicating that the shroud would fracture significantly and melt.
Considering the phenomenon in which a shroud loses its structural strength at around 1,000 K before it melts, it was concluded
that the shroud is deformed, damaged or buckled by the weight of the structures at the upper portion (such as the separator).
⚫ It was also concluded that the structures in the lower plenum (such as the CRGT) are damaged or melted further than theircounterparts in Units 2 and 3. It was also presumed that the concrete erosion by MCCI, even the largest one, will stop near the
PCV shell.

Shroud temperature history

Shroud temperature (K)Melting point of the
shroud5 hours after scram(before fracture of the channel box)

Shroud damage
5.5 hours after scram(after fracture of the channel box) was detected

ShroudCore support plate

Time after scram (h)©International Research Institute for Nuclear Decommissioning

(2)-[1]-3-(a)-7 Analysis of Each Unit over Three Weeks after the Accident

37[Final report (evaluation results of Unit 2)]⚫ From the discussions made at the accident-scenario study meetings and the SAMPSON analysis results, it was concluded
that the fuel rods started fracturing about 77 hours after the scram and started melting in about 78 hours. Then, it waspresumed that about 100 tons of fuel debris (containing fuel and structure material) fell on the lower plenum in about 80
hours and some part of fuel debris fell further down to the pedestal in about 92 hours.
⚫ The PCV pressure presumably rose between Hour 80 and 81, generating hydrogen and accelerating damage to, andmelting of, the fuel rods. The LINX test results indicate the possibility that non-condensable gases, such as hydrogen, inhibit
the condensation of water vapor. As is shown by actual values, the S/C water temperature rose to a high level during thehour, indicating our need to take incomplete condensation into account when evaluating the amount of hydrogen generated.
⚫ A SAMPSON analysis was performed that took account of the actual measurement data and the phenomenon, getting a
result showing that at least nearly 300 kg of hydrogen was generated. The analysis results showed that peripheral fuel rods
may not melt completely under some conditions.
⚫ A two-dimensional analysis ((2)-[1]-3-(a)-6) of the in-core temperature distribution was conducted, taking into account suchfactors as the oxidation reaction heat resulting from hydrogen generation, water injection and vapor temperature that was
obtained from the SAMPSON analysis, to closely evaluate the fuel rod temperatures. A presumption was reached that the
support structure of the peripheral fuel rods nearly rose to the melting point and that almost all of them eventually collapsed.
However, some fuel rods may still exist, leaning on the shroud side.

Containment pressure [MPa (abs)]

Sensitivity analysis of PCV pressure in
the 2nd peak period

Case of 291 kg of hydrogengenerated during the 2nd RPV
peak period

Peripheral fuelrods still exist

Case of 365 kg of hydrogengenerated during the 2nd RPV
peak period

Peripheral fuel rods
also collapsed

D/W actual meas.

Time after scram (hr)©International Research Institute for Nuclear Decommissioning

(2)-[1]-3-(a)-7 Analysis of Each Unit over Three Weeks after the Accident

38[Final report (evaluation results of Unit 3)]⚫ From the discussions made at the accident-scenario study meetings and the SAMPSON analysis results, it was concluded that
fuel debris (fuel and structure material) slumping occurred at 43 hours and 45 hours after the scram. It was concluded that thePCV pressure continued to rise after the PCV vent closed, but that increase stopped as a result of balancing with the leakage
from the D/W. Given the fact that the water injection from the fire engine was limited, it was concluded that the PCV pressure
finally started decreasing when the inside of the RPV dried out. It was concluded that, during that period, water started
migrating from the S/C to the D/W, creating, along with the water due to the D/W spray, some pools of water on the D/W floor.
⚫ It was concluded that the lower head damaged 50 hours after the scram and that fuel debris dropped to the pedestal. From theanalysis results showing almost zero residue of fuel debris in the reactor core, it was concluded there are about 30 tons ofcrusted fuel debris in the lower plenum and about 120 tons of fuel debris in the pedestal.
⚫ A relatively large amount of water was present when the fuel debris migrated to the pedestal. Considering the possibility thatwater seeped in the cracks of the crust on the surface, it was concluded that the fuel debris cooled down quickly, resulting inalmost zero concrete erosion.

PCV pressure behavior

Schematic diagram of crustal
fracture

Progression state of MCCI

©International Research Institute for Nuclear Decommissioning39(2)-[1]-3-(a)-8 Detailed Accident Progression Analysis Using the Accident Progression Analysis Codes[Overview and Objectives]•
In the detailed accident progression analysis conducted in FY2016 by using the MAAP code, the accident progression
scenario and the fuel debris distribution was studied while taking into account the code’s limit and for a physical phenomenon
model. The issues of reduction and quantification of the fuel debris distribution were also extracted.
•
In FY2017, some of the physical phenomena models of the code were changed. The renovated models in the accident
progression analysis were used to reduce the uncertainties involved in estimating the accident progression scenario and thefuel debris distribution.
[FY2017 Research Details]•
Improvement of the MAAP physical phenomena models➢ Model of fuel debris outflow when the RPV wasdamaged➢ Model of B4C oxidation of the residual heat control
rods➢ Model of core blockage and corium migration➢ Model of fuel debris flowing into the CRD housing•
Detailed evaluation of fuel debris dispersion behavior forthe study on MAAP boundary conditions•

Estimation of the accident scenario and quantification of
the fuel debris distribution based on the accident
progression analysis on Units 2 and 3 using the improved
code[Final Results]•
Improved the MAAP physical phenomena models andtesting its operation and validity•

Performed the detailed accident progression analysis in
Units 2 and 3 using the improved code•

Updated the accident progression scenario and theestimated fuel debris distribution

Example of model improvement (core blockage and corium migration model)
The model initially did not return to Type 2 or 3 once it reached theblockage node (Type 4); a core damage progression model was madeallowing such return in order to simulate the accident that presumably
occurred at Fukushima Daiichi. The impact on the amount of hydrogengenerated in the core damage process was particularly significant, and the
reproducibility of the PCV pressure behavior was improved.

©International Research Institute for Nuclear Decommissioning40(2)-[1]-3-(a)-8 Detailed Accident Progression Analysis Using the Accident Progression Analysis Codes

Unit 2
Area

Results of improved MAAP analysis

Estimated fuel debris distribution based on the analysis results

Reactor corearea

In the reactor core, 14.5% of theinitial UO2 mass still exists.

Most parts of the reactor core fell to the lower plenum.
The outermost peripheral area of the reactor core with low output may still exist on the core
support plate because of the long-term water injection from the fire engine.

Lower plenum

In the lower plenum, 65% of theinitial UO2 mass still exists asparticulate fuel debris.

Due to the localized damage to the weld of the instrumentation guide tube, and solidificationon the RPV sub-structure surfaces, the molten core may have gradually fallen to the lower
part of the PCV.
A rupture may exist at a location closer to the side of the RPV boom than to the center,
possibly limiting the fall of the molten core only from a position at or higher than the height of
the rupture.

Lower part of
the PCVOf the initial mass of the reactor
core, 20% falls to the pedestal.

At around 14:00 on March 15, the molten core, presumably part of the molten core in thelower plenum, possibly fell to the lower portion of the PCV. Concrete erosion may haveoccurred to a very limited extent.

Area

Results of improved MAAP analysis

Estimated fuel debris distribution based on the analysis results

Reactor corearea

Fuel no longer exists in thereactor core.

Most parts of the molten core may have fallen, in a large scale, from the reactor core area tothe lower plenum; as a result, the amount of the fuel that still exists in the reactor core area
may be limited.

Lower plenum

Of the initial mass of the reactor
core, 20% still exists in the lower
plenum.

Due to the major damage at the RPV bottom, the amount of the molten core that still exists inthe lower plenum may be small.

Lower part of
the PCVOf the initial rector core mass,
80% fell to the pedestal and thenpart of it flowed into the drywell.

Due to the massive molten core falling to the lower part of the PCV, the molten core may
have flowed from the pedestal to the drywell, but in a limited manner. There may be more
concrete erosion toward the floor.

Unit 3©International Research Institute for Nuclear Decommissioning41(2)-[1]-6 Simulated Fuel Assembly Degradation Test

[Overview and Objectives]
Regarding the core melting and migration behavior at the time of the accident on the 1st floor, it might have had a residual, heat-type progression,different from the TMI-type. Factors such as (1) gas permeability in the high-temperature core, and (2) entry of decayed fuel into the lower support
structure area substantially affect the presence or absence of molten fuel in the core. These have greatly contributed to the uncertainties in evaluating
the accident progression and the fuel debris distribution and characteristics. To reduce such uncertainties, a specimen simulating the residual heattype reactor core and lower support structure was heated up with a plasma torch and the decay, melt, and migrating behavior of the core substancewas observed.
Two planes of channel boxes (Zr)

Simulated fuel rod (ZrO2)  48
pcs
Control rod blade

CMMR4

Improved the system so it would allow
easier heating.
*No change to the max output of the plasma heatup.

Heated up the upper part to2,600 K or higher.
Observation angle

Fuel top

Observation angle of the image
below

Partial blockage
A small molten pool formed
due to the melt of oxides (thelower power blocked due tosolidification)
Space in which thecontrol rod blade
melted down
deposits and (partial) blockage
formed by molten metal (Fe or
Zr) including boron

Fuel rods do not decay
if exposed to hightemperature©International Research Institute for Nuclear Decommissioning

37 cm

30 cm

[FY2017 Research Details]
The specimen was heated up simulating the residual heat system up
to the assumed temperature in the lower core area at the time thecore substance in Unit 2 slumped (migrated to the lower plenum) andthe results were evaluated.
[Final Results]
The following knowledge was obtained and provided it in the
25 cm below
the fuel top
estimate diagram of fuel debris distribution.
[1] High-temperature core fuels form partial blockage; however,residue fuel columns tend not to fuse with each other andpresumably have macroscopic permeability against the gas
phase (and the liquid phase) including when they havedecayed (similar to the loose core debris in TMI-2).
[2] High-temperature fuels would maintain their column shape up
to nearly the melting point, and they hardly move in the form ofa pellet. If fuels in units decay due to a load, etc., the decayedhigh-temperature fuel would stay in the reactor core (with nochange to the axial position), presumably allowing only part of
the high-temperature fuels to enter the support structure in theprocess leading to the core melting.
Reflection in the estimate diagram of fuel debris distribution➢Combined with the knowledge obtained in the “event sequence
analysis at the time of core material slumping,” the results wereused to estimate the core sub-structure fracture mode in Unit 2 andthe properties (mainly solid) of the fuels that still exist in the lower
plenum.
➢The knowledge that would enable more accurate evaluation of the
RPV fracture mechanism, the fracture mode, the fuel debris
properties when the RPV was damaged, and the outflow property.

CMMR32.b) Evaluation of Chemical Characteristics of FPs

42

PurposeTo decommission the Fukushima Daiichi Units 1 to 3, it is essential to identify the distribution and chemical
properties of cesium accounting for a major part of the gamma source. Under this project, apart from the typicalchemical species in the evaporation, migration, and condensation processes (CsI: cesium iodide, and CsOH: cesiumhydroxide), a study on the possible presence of compounds that are believed to be relevant to evaluating Fukushima
Daiichi (molybdate compound, silicate compound, boronic acid compound, etc.) and their impact levels is conducted.
If there is uneven distribution of cesium with poor solubility --- a factor that would negatively influence the top-entry
method --- in the upper reactor core region due to the reaction with the upper-structure material (steel), was also
tried to be found out. Furthermore, a study was conducted on the mechanism by which insoluble cesium
particles/powder dust are generated and their migration routes.

Gaseous FPs
Particulate FPs
ShieldPlugOperation floor

Nuclear reactor
Well

Unit 2 Operation floor©International Research Institute for Nuclear Decommissioning43(2)-[2]-1 Reaction with Cesium and Steel Materials, and Re-evaporation[Overview and Objectives]
Overview: Conduct a test on Cs adsorption in steel materials and its re-evaporation, using parameters such as temperature andatmosphere to enhance knowledge of the Cs chemisorption behavior, thereby developing a Cs adsorption model thattakes major compounds generated by adsorption into account.
Objectives: Verify the existing model through a replication experiment on Cs adsorption and evaluate the chemical andthermodynamic properties of relevant compounds to develop a model that takes various chemical conditions, suchas re-evaporation and Cs concentration variation, into account.
Chemisorption model developed in this study(*): The adsorption volume was calculated using:

vd: reaction rate constant [cm/s]
Cg: CsOH concentration in the gas
phase [μg/cc]
CB: Si concentration in steel [wt.%]

Amount of Cs
chemisorption
to dryer/separator [kg][FY2017 Research Details](i) Evaluation test on adsorption and re-evaporation behavior
An experiment on Cs absorption was performed, usingparameters such as temperature and atmosphere, and theeffects of various chemical conditions was evaluated.
(ii) Development of an adsorption and re-evaporation model
The results of the Cs adsorption experiment based on themass transfer theory and thermodynamic data on Cs
adsorption products were analyzed and modeling toconsider the effects of chemical conditions such as theCsOH concentration in the gas phase was performed.
[Final Results]✓ The influence of chemical conditions wereexperimentally clarified, such as CsOH concentration in
the gas phase, atmosphere, Si concentration in steel
and temperature, on the Cs adsorption volume. A
chemisorption model based on the test values wasdeveloped.
✓ By using the model formula, the sensitivity (fluctuation
range) of Cs adsorption volume to the dryer andseparator according to changes in chemical conditions,such as CsOH concentration in the gas phase, wasexhibited.

14
12
10
8
6
4
2
0

A: Adsorption surface area
[cm2]
Md: Adsorption volume [μg]t: Adsorption time [s]

800℃
900℃
1000℃

0.01

0.1

1

CsOH concentration in the gas phase [μg/cc]
Figure 1. (Estimate) Cs adsorption volume to the dryer/separator in 1F2
where structural material temperature and CsOH concentration in the gas
phase were constant for 3 hours©International Research Institute for Nuclear Decommissioning

10

(2)-[2]-2 Evaluation of Particulate Cesium Compounds(1. Test on Reaction and Solidification in the Gas Phase)

44[Overview and Objectives]
To clarify the generation mechanism of insoluble cesium particles, a test on reaction and solidification in the gas phase in FY2017 was
conducted by using (1) Mo, (2) Zn coating, and (3) Ca as evaporation sources in addition to the standard simulation conditions of Cs-SiO that were examined in FY2016.
[FY2017 Research Details]
An additional test was performed to examine the following
possibilities:[1] Generating insoluble Cs particles in the environment wheremolybdate coexists.

[2] Entry of the Zn coating component used for the S/C inner wall
into insoluble Cs particles.

Quantitativearea[3] An insulation material or concrete are the original component of
insoluble Cs particles.
[Final Results][1] When the evaporation source, Mo, was added to the simulation
test conditions, soluble Mo-Cs-O particles were produced, but noinsoluble cesium particles were produced (Right figure (a)). In the
molybdate-bearing environment, it is believed that no insoluble
Cs particles are produced; presumably they were produced in theearly phase of core damage (where FP composition was
Cs>>Mo).

MoCs(a) Cross section of a Mo-Cs-O particle generated(Mo: 46%, Cs: 53%, Si: 1%-metallic element, (at% in the metallic element))[2] When the coating fragments of the S/C inner wall were added as
an evaporation source, insoluble cesium particles came tocontain Zn depending on the coating fragments’ loading volumes(Right figure (b)).
[3] When Ca-bearing material was added as an evaporation source,
insoluble cesium particles came to contain Ca depending on theevaporation property of the Ca-bearing material (Right figure (c)).
Presumably, neither concrete nor insulation was involved ingenerating insoluble Cs particles.

(b) Zn-bearing cesium particles(c) Ca-bearing cesium particles(Si: 56%, Cs: 17%, Zn: 22%-metallic element)(Si: 69%, Cs: 29%, Ca: 1%-metallic element)©International Research Institute for Nuclear Decommissioning

(2)-[2]-2 Evaluation of Particulate Cesium Compounds(2. Study on the Microstructure)[Overview and Objectives]

45

According to a report, a structure having nano-sized, network-like crystalline ZnFe2O4 was found in some insoluble Cs particles.
The generation mechanism of those particles was studied by simulating and evaluating the microstructure of insoluble Csparticles.
[FY2017 Research Details]A simulated sample was prepared with the following two techniques to observe the microstructure.
[1] The sample was adjusted so it had the same composition as the insoluble Cs particles. It was loaded in a crucible andmelted at 1,500oC in the air atmosphere, then quickly cooled it down with water.
[2] SiO2-Fe2O3-ZnO was adjusted so it had a composition similar to that of the insoluble Cs particles and hence had thepossibility of experiencing spinodal decomposition. Then, it was melted by gas levitation and laser heating and thensolidified.
Metrix: SiO2 glass

White zone: Zn–Fe-rich

Spherical particles: ZnFe2O4 crystalline

Black zone: SiO2-rich

Figure 1: TEM image of sample 1(Phase identification based onelectron diffraction results)

1 μm

Figure 2: SEM image of sample 2(Elementary analysis with EDS)

[Final Results]✓ In the SiO2 glass matrix of Sample [1], the generation of nano-sized crystalline ZnFe2O4 (Figure 1) was observed. The
particle observed by Yamaguchi et al.*1 (a structure formed through the dynamic dissolution of ZnFe 2O4) requires rapid
cooling after it is generated at a high temperature. It suggests that this type of structure was generated at a hightemperature in the RPV and then rapidly cooled in the S/C.
✓ Presumably, the network structure of the Zn-Fe-rich phase shown in Sample [2] (Figure 2) was generated by spinodal
decomposition. The microstructure observed by Furuki et al.,*2 (a network crystalline structure of ZnFe2O4) is nano-sizedand has a different dimension from the spinodal decomposition structure formed in the liquid phase mocked up under theconditions of Sample [2]. The origin of the nanostructure remains unknown.
*1: N. Yamaguchi et al., Sci. Rep. DOI: 10.1038/srep20548, 2016
*2: G. Furuki et al., Sci. Rep. DOI: 10.1038/srep42731, 2017

©International Research Institute for Nuclear Decommissioning

(2)-[2]-2 Evaluation of Particulate Cesium Compounds

46(3. Evaluation of Inorganic Zinc Coating)[Overview and Objectives]
A study on the possibility that inorganic Zn coating in the S/C inner wall was the material origin of the constituent elements(Zn and Si) of insoluble Cs particles, was conducted. A simulated coating sample was prepared by referring to construction
specifications and the coating components’ leaching behavior was examined under the test conditions simulating theaccident (with purified water at 140oC for 30 hours).
Table 1. Chemical analysis results of the coating before and after the test

[FY2017 Research Details]•
By using ICP and IC, the coating components
before and after the leaching test (Table 1)
was measured. Apart from Zn and Si, a small
amount of K and Cl was detected.

Before test
After test
Before test
After test

[Results]✓ Inorganic Zn coating components can only
be involved with Unit 2, where the RCIC
water source was switched from the CST to
the S/C. Hence, the RPV of Unit 2 is the sitewhere the insoluble Cs particles occurred in
the environment.
✓ From the amount of K2O supplied from thecoating (on the order of 100 g) and K2O
concentration in insoluble particles (approx.
1 wt%), the amount of insoluble Cs particles
generated is estimated to be on the order ofmore than 10 kg (Table 2).
✓ As the ratio of SiO2/K2O supplied from thecoating shows SiO2 was insufficient, theevaporation of silica from the steel materials
may have contributed to the amount.

Table 2. The estimated amount of insoluble Cs particles generated(based on the potassium amount)
Evaluation
values

Evaluation methods

1. Zinc coating
weight2,600 kg

Coating area (8,566 m2)  Ratio of the smootharea (0.8)  Coating thickness (100 μm)  Coating
density (3.8 g/cm2)

2. K content

3.12 kgK(3.76
kgK2O)

2,600 kg  0.0012 (Table 1 average)

3. Amount of K2O
dissolved

0.752
kgK2O

3.76 kgK2O  0.2 (dissolution rate; Table 1
average)

4. Brought into the
RPV

0.125
kgK2O

Dissolution amount  Amount of water
evaporated/Total water amount (500 m 3/3,000 m3)

12.5 kg

0.125 kgK2O/0.01 (K2O content by Kogure et al.)Items

5. Amount of Cs
particles generated©International Research Institute for Nuclear Decommissioning

(2)-[2]-2 Evaluation of Particulate Cesium Compounds(4. Oxidation and Evaporation Behavior of Silicon in Steel Material)

47[Overview and Objectives]
A study on the oxidation behavior of stainless steel and the existing state of silicon in such oxides under the temperature range
and atmosphere containing hydrogen and water vapor that presumably simulate the temperature range and atmosphere at thetime of the Fukushima Daiichi accident, was conducted. The thermodynamic properties of Cs-Si-(Fe)-O as a basic structure ofcesium compounds are clarified.
[FY2017 Research Details]•
As shown in Figures 1 and 2, the Si distribution
state in metallic phase near the metallic phaseobtained by heating up SUS304 grade
stainless steel was analyzed, in which Si
content changed under water vapor
atmosphere.
•
As shown in Figure 3, using a thermodynamic
database, the possibility that hydroxide andsuboxide are the silicon sources of the aforeFigure 2. Si distribution in stainless steel
Figure 1. Si distribution in stainless steel
mentioned particles was estimated andcontaining 1% Si, which was oxidized for 60
containing 1% Si, which was oxidized for 60
minutes at 1,200oC under atmosphere ofminutes at 1,200oC under atmosphere ofstudied.
H2/H2O/Ar
H2O/Ar•A Cs2O-SiO2 melt and copper, as basic
structures, in a graphite crucible undercoexistence of CO was equilibrated and the
SiO2 activity determined.
[Final Results]•
The existing state of Si oxides is divided mainly into Fe2SiO4 and SiO2,
largely depending on the state of the oxide layer of steel material.
•
When the oxidation rate is high, the oxide layer is separated into two,and Si exists in the inner oxide layer as Fe2SiO4; then, it is transformed
to SiO2 as the oxide layer peeled off due to long-time heating.
•Thus light was shed on the atmospheric dependence of silicon source ininsoluble cesium particles and the dependence of SiO2 activity in cesium
silicate melts on Raoult's law.
Figure 3. H2/H2O ratio dependence of vapor pressure of SiO
and Si(OH)4©International Research Institute for Nuclear Decommissioning

(2)-[2]-2 Evaluation of Particulate Cesium Compounds

48(5. Relation with the Accident Progression Sequence)
Time sequence

Generation process

During normal
operation••

The major portion of Cs stays in the pellets, but some portion precipitated at the grain boundary.
Mo, as intermetallic compound, precipitated mainly in crystal grains.

Earthquake (14:46, Mar 11), Tsunami, Loss of AC power (15:40–15:41, Mar 11), Switching of RCIC water source (CST to S/C)
While RCIC is
operating•

S/C water temperature rose → Exceeded the heat-resistant temperature of the coating → Coating components
leached → Coating components migrated due to the RCIC water flow → Evaporated to dryness (Si, Zn, K) on
the boiling surface in the RPVStop of RCIC (around 9 am, Mar 14), Forced depressurization of RPV (18:00, Mar 14), Decrease in water level, Start of core heatingCore heating

•
•
•
•
•
•
•

Fuel cladding tube failure
Evaporation of Fe crud and coating components (Zn, Si, and K) attached to the surface of fuel cladding
Evaporation of Zircaloy component (Sn)Meltdown of control rods (Meltdown forming Fe-B eutectic, without B’s involvement)
Fuel heating during the meltdown of the upper tie plate, oxidation and evaporation of SioDischarge of Cs due to grain boundary connectivity of pellets (>1,500 C)
Liquid phase of Cs2SiO4-SiO2 system → Densification and spheroidization of the condensation phase

Discharge fromRPV•

As a result of opening the safety valve relief (21:20, Mar 14), spherical cesium-bearing particles were rapidly
discharged from the RPV, along with a large amount of water vapor and non-condensable gas → Rapid cooldown→ Suppressed nucleation and spherical precipitation of ZnFe2O4 → Generation of insoluble Cs particles
Most of the particles were captured in the S/C water. Some of them were discharged to the environment.

•

Observed at environmental monitoring facilities in the Tokyo metropolitan area, including Meteorological Research Institute (7:00–16:00,
Mar 15)©International Research Institute for Nuclear Decommissioning

(2)-[2]-2 Evaluation of Particulate Cesium Compounds

49(6. Phenomenology of Material Mechanism)
Meltdown of control rods (CRs)⚫ Meltdown due to the eutectic reaction of
Fe-B (at about 1,200 oC) (the evaporationof silicon was suppressed)⚫ HBO2 evaporated, suppressed generationof CsBO2

Meltdown of the upper tie plate⚫ Cast steel contains Si of approx. 1.5wt%⚫ Presumably, the plate was heatedup with fuel during the fall andbecame the source of Si

Phase stability of insoluble Cs particles⚫ Two immiscible liquid phases anddissolution gap⚫ Rapid cooling that suppressed thenucleation and spherical precipitation of
ZnFe2O4⚫ Susceptibility for spinodal decomposition

Insoluble Csparticles
SiO2-Cs2Si4O9(ZnO-Fe3O4)

Si

Cs chemical form⚫
Under the environment where thegeneration of Cs2MoO4 and CsBO2 was
suppressed (CsOH environment),
amorphous insoluble Cs particles andcrystalline Cs2Si4O9, CsFeSiO4 were
produced

Si, Zn,K, Cl
CsOH

Involvement of S/C coating components (Unit 2)⚫
During the RCIC operation, the inorganic
zinc coating component leached andmigrated to the RPV⚫
Evaporated to dryness on the boilingsurface (of fuel)⚫
Evaporated due to heat generated from thecore

Fe, SnCs
Cs2MoO4

Cs

Mo

Heat generation of core fuel⚫ Evaporation of iron crud
⚫ Evaporation of cladding component(Sn)o⚫ Discharge of Cs and I > 1,500 C⚫ Discharge of Mo was limited < 2,300
o
C

CsI

Cesium silicate on the steel material
surface
Cs2Si4O9, CsFeSiO4

Vapor oxidation of the material of the structures at theupper portion⚫ Fe2SiO4 and SiO2 were formed on the SUS
surface⚫ Cesium silicate was generated by the reaction
with Cs⚫ Possibility of cesium silicate evaporating at hightemperature (>1,000 oC)©International Research Institute for Nuclear Decommissioning

(2)-[2]-3 Optimization of the Evaluation Model for Cesium Compounds

50[FY2017 Research Details](i) Optimization of the adsorption model
A correlation equation for the temperature-dependentadsorption rate was used in a reducing atmosphere basedon the existing data and it was determined that the effect ofCsOH concentration can be evaluated using the corrected
formula based on this project’s results (2)-[2]-1.
(ii) Development of a generation model of particulate cesium
It was determined that the Si discharge sources were theevaporated Si component in SUS core structure materials,
such as the control rod blade, core support plate, and theleached Si component in the S/C inner wall coating. Then,an Si discharge model and a particulate cesium generation
model was developed.
[Final Results]✓ The optimized adsorption model and the particulate Cs
generation model in SAMPSON was incorporated.
✓ A tool to evaluate the validity of the branching ratio “f” wasdeveloped based on the study results of this project (2)[2]-2.

Adsorption rate (cm/s)[Overview and Objectives]
Overview: Optimize the adsorption model regarding the products of reaction between CsOH and SUS304 steel under anoxidized/reducing atmosphere based on the JAEA experiment results, and develop a generation model ofparticulate cesium.
Objectives: Introduce the adsorption model and the silicon generation model into SAMPSON, and develop a chemical
characterization tool, based on the analysis results of the amount of silicon generated.

Correlation equation (oxidized atmosphere)
Correlation equation (reducing atmosphere)
Existing data (oxidized atmosphere)
Existing data (H2O atmosphere)
Existing data (reducing atmosphere)

Temperature (K)Figure 1. Temperature dependence of adsorption rate
Chemical equilibriumcomputation

Particulate cesium
Fuel

SUS materials(core structurematerials)

Coagulation anddepositsion

Coating(S/C Inner wall)

Figure 2. Generation model of particulate cesium©International Research Institute for Nuclear Decommissioning

(2)-[2]-4 Analysis and Evaluation of the Actual Units Using the Upgraded Model 51[Overview and Objectives]
Overview: Conduct analysis of Units 1 to 3 using SAMPSON, study the consistency with items such as actual
measurement data that can affect the atmosphere, and evaluate the reliability of analysis results
Objectives: Evaluate the cesium distribution in the RPV and PCV based on the SAMPSON analysis results[FY2017 Research Details](i) Analysis under the conditions of Fukushima Daiichi plant
The Units 1 to 3 using the SAMPSON were analyzed, forwhich the improved Cs behavior model wasincorporated.
(ii) Study on the factors that affect Cs behavior• Cs adsorption mainly occurs in an oxidized atmosphere,
and adsorption in a reducing atmosphere (P(H2) >P(H2O)) lasted for a short time.
• Generating particulate Cs needs a greater amount of Si
than that coming from the core structure materials,
requiring Si from the coatings inside the S/C. Therefore,
the study was conducted only on Fukushima Daiichi
Unit 2, where RCIC was run by using the S/C as thewater source.

Amount of Si produced fromthe coreAmount of Cs produced fromthe fuel

Figure 1. Time dependency of the amounts of Cs andSi generated in Fukushima Unit 2[Final Results]✓ Analysis was conducted and the Cs distribution in the RPV and PCV was calculated.
✓ It is believed that the D/W inside of Fukushima Daiichi Unit 2 has the highest contamination level when comparing
Units 1–3. Meanwhile, Unit 2 analysis results showed almost no cesium on the D/W inner wall, inside the pedestal,or on each Fukushima Unit. This result indicates that the stage of simulating the current cesium distribution in each
Fukushima Daiichi unit has not been reached.
✓ Current understanding and handling of the analysis codes are insufficient for analyzing the phenomenon.
©International Research Institute for Nuclear Decommissioning

(2)-[2]-5 Analysis of Samples Collected at Fukushima Daiichi NPS

52[Overview] The samples collected at Fukushima Daiichi NPS such as deposits in the contaminated water in the PCV of Unit 1,
contaminant attached onto the blockage inside the TIP piping in Unit 2, and a surface of a protective covering piece on
the Unit 2 operation floor were analyzed. A detailed observation by performing FE-TEM, with a focus on U-bearing
particles was also conducted, and the samples’ microscopic characteristics were sorted out.

[FY2017 Research Details and Results][2] Detailed analysis of contaminated samples
from Fukushima Daiichi NPS➢➢

The following was found in the deposits inthe contaminated water in the PCV of Unit 1:
rust containing Fe and Zn; C-bearing
material; structure material component;
glass material; Cu; Pb; Zr; and U-bearing
particles containing Cu (Figure “a” in theright).

(a) U-bearing area inside the deposits in the contaminatedwater of PCV in Unit 1(TEM-EDS mapping of the element U (right) and element Zr (left))

In obstacles collected from the TIP piping at
Unit 2, an area containing Zr apart from thestructure material components and Mo wasfound.
Uranium oxide➢

The following was observed on theprotective sheet on Unit 2’s operation floor:
uranium oxide particles that do not containZr; and vitreous particles containing Zn, Cs,
etc. (Figure “b” in the right).

Vitreous particles(b) Vitreous particles and uranium oxide on the protective sheet
on Unit 2’s operation floor©International Research Institute for Nuclear Decommissioning

(2)-[3] Knowledge Utilization through International Joint Research

53[Overview and Objectives]
The Benchmark Study of the Accident at the Fukushima Daiichi NPS (BSAF) Project Phase-2 was established by theOrganization for Economic Co-operation and Development (OECD)/the Nuclear Energy Agency (NEA). The BSAF project has
been hosted by the several operating agents including the Institute of Applied Energy (IAE) which utilizes the data obtained
through the project and understand the conditions inside the reactors.

[FY2017 Research Details]
In FY2017, the final year of the BSAF Project Phase-2, using major physical models, the analytical results provided byparticipating organizations on the accident progression behavior is compared and studied, within three weeks after the scram.
Their consistency is also studied with the actual measurements and on-site investigation results. The results were shared
among the BSAF participating organizations as the Phase-2 output.

[Final Report]⚫ In July 2017, the following meetings were held at the Kokukaikan hall in Tokyo: the PRG (Program Review Group) meeting;MB (Management Board) meeting; and the joint session with other projects for the Fukushima decommissioning. The pastand latest results of the internal investigations and Japan’s views regarding the accident scenario based on the knowledge
obtained so far, were reported on . Opinions with experts from participating organizations were exchanged. The following
were discussed: the results of the comparisons among the preliminary analyses, actual measurement data and on-site
investigations concerning the accident progression, fuel debris distribution and FP distribution within the three weeks after thescram; and the structure of the final report.
⚫ In January 2018, the PRG, MB and joint sessions at the OECD/NEA
headquarters in Paris were had. The evaluation results that theparticipating organizations provided on the accident progression within
three weeks after the accident were compared against the actual
measurement data and the on-site investigation results. It was
confirmed that our understanding of the accident progression
significantly deepened from Phase-1, as is demonstrated, for example,
by the basic consistency between the evaluation results concerning the
FP discharge amount and the amount discharged to the environment.
The knowledge obtained through such discussions with overseas
institutions was used, for our comprehensive analysis and evaluation,
which helped improve the estimation accuracy on the fuel debris
Comparative example of observation and analysis results of Cs-137
depositsion
distribution.
©International Research Institute for Nuclear Decommissioning

54

3. Research Details and Results
Summary for FY2017©International Research Institute for Nuclear Decommissioning55

3. Research Details and Results Summary for FY2017(1) Comprehensive analysis and evaluation of the conditions inside the reactors[1] Comprehensively analyze and evaluate based on the Units data and other R&D project
results
The conditions inside the RPVs and PCVs were comprehensively analyzed and evaluatedbased on the results of the following initiatives: “analysis and evaluation based on on-site
information;” “analysis and evaluation based on the data at the time of or after the accident
and inverse problem analysis;” and “analysis and evaluation utilizing analysis codes.” Then,the estimate diagrams of fuel debris distribution and FP distribution was developed.
[2] Establish a database for comprehensive analysis and evaluation
To share knowledge with overseas institutions more efficiently, the English search function
was added to the database and the graph display function for the measurement data, etc.
collected mainly within three weeks after the accident was improved. The adding of
documents and data with search tags to the database that was developed under this project,contributing to a more efficient, comprehensive analysis and evaluation was also continued.

©International Research Institute for Nuclear Decommissioning3. Research Details and Results Summary for FY2017

56(2) Estimation and evaluation of fuel debris behavior and fission product behavior and its characteristics thatwould be useful for the comprehensive analysis and evaluation[1] Reduction of uncertainties through analytical techniques
Sensitivity and other analyses were performed using the accident progression analysis codes, taking theboundary conditions and analysis models into account, with regard to the events assumed to have occurred
inside the reactors. Thus knowledge was obtained that will be useful for comprehensive analysis andevaluation.
[2] Evaluate FPs’ chemical characteristics
When evaluating the FPs’ chemical characteristics, light was shed on Cs in particular, as it was believed theinformation would be important for decommissioning. Cs distribution and its chemical characteristics werestudied, such as: identifying chemical species requiring special consideration, apart from the typical chemical
species of CsI and CsOH; the production amount of insoluble Cs particles observed in the environment; andthe possibility of uneven distribution of Cs due to a reaction with the structures at the upper portion of the RPV.
The samples collected on-site were also analyzed and the composition and spatial distribution of uranium andFPs to understand the conditions inside the reactors was studied.
[3] International joint research to use knowledge inside and outside Japan
By using a database that was developed for the project and through managing an international joint research
project (OECD/NEA BSAF Project Phase-2), the accident progression scenario and plant-related information
was shared with overseas institutions. The results provided by participating organizations concerning theaccident progression, fuel debris distribution, and FP distribution in the first three weeks after the accident werealso compared with the measurement data and the on-site investigation results. As a result, our sharedunderstanding about the accident progression and plant status deepened compared to Phase 1, leading to aremarkable reduction in the variety of analysis results from participating organizations and to the basic
consistency of the evaluated FP discharge amount with the measured discharge to the environment. This
deeper understanding of the accident progression contributed to improved accuracy of fuel debris distribution
estimates. The knowledge obtained through the discussions with overseas institutions was also used for ourcomprehensive analysis and evaluation.
©International Research Institute for Nuclear Decommissioning

57

Attachment➢ Attachment 1. Project Organization Chart (detailed)©International Research Institute for Nuclear DecommissioningAttachment 1. Project Organization Chart (detailed)(Consignment research)

International Research Institute for Nuclear Decommissioning (IRID)The University of Tokyo

IRID Head Office58○ Research management of implementation plans, their progress status, etc.
○ Project results summary and information dissemination(1)-[1]-1 Extraction of issues for further research to reduce uncertainties of the estimated conditions inside the reactors(1)-[1]-2 Comprehensive analysis and evaluation based on the data from the actual units and other researchdevelopment results(1)-[1]-1 Extraction of issues for further research to reduce uncertainties of the estimated conditions inside the reactors(1)-[1]-2 Comprehensive analysis and evaluation based on the data from the actual units and other researchdevelopment results(2)-[1]-3 (a) (4) Event sequence analysis at the time of core material slumping(2)-[1]-6-1 Simulated fuel assembly degradation test(2)-[2] Analysis and evaluation of chemical forms of FPs(outsourcing)
Hitachi Power
Solutions(Consignment research)
Waseda University(2)-[1]-2 Inverse problem evaluation using virtual reactors(2)-[1]- 2 Inverse problem evaluation using virtual reactors(Analysis and evaluation of the coolingcharacteristics of debris considering water injection)(2)-[1]-3-(a)-4 Event sequence analysis at the time of core material slumping[MELCOR analysis](Consignment research)
The University of Tokyo(outsourcing)
Sojitz Machinery
Corporation(2)-[1]-3-(a)-4 Event sequence analysis at the time of core material slumping(Analysis of pipe deformation)(2)-[1]-6-1 Simulated fuel assembly degradation test
(plasma heatup experiment)(outsourcing)
IRID/Hitachi GE(1)-[1]-1 Extraction of issues for further research to reduce uncertainties of the estimated conditions inside the reactors(1)-[1]-2 Comprehensive analysis and evaluation based on the data from actual units and other research developments results(2)-[1]-1 Evaluation of the concrete erosion state at the pedestal using sensitivity analysis(2)-[1]-3 Accident progression analysis using the analysis code

Inspection Development
Company Ltd.

(Consignment research)
Osaka University(2)-[1]-6-1 Simulated fuel assembly degradation test
(Service contract for CT imaging in the plasma melting test)(2)-[2]-2 Evaluation of particulate cesium compounds(Property evaluation test for cesium-bearing glass)(Consignment research)
Tokyo Institute of Technology

IRID/Toshiba Energy
Systems & Solutions(1)-[1]-1 Extraction of issues for further research to reduce uncertainties about the estimated conditions inside the reactors(1)- [1]-2 Comprehensive analysis and evaluation based on the data from the actual units and other research development results(2)-[1]-3 Accident progression analysis using the analysis code(outsourcing)Electric Power
Research Institute(EPRI), USA(2)-[1]-3 Accident progression analysis
using the analysis code

(Consignment research)
Tohoku University(outsourcing)
Art Kagaku
Co.,Ltd.
(outsourcing)
Nippon Nuclear Fuel
Development Co., Ltd.
(outsourcing)
Nippon Nuclear Fuel
Development Co., Ltd.

(2)-[2]-2 Evaluation of particulate cesium compounds(Study for evaluation of thermodynamic quantities of Cs-Si(Fe)-O system compound)(2)-[2]-2 Evaluation of particulate cesium compounds(Study of silicon behavior in steel materials)(2)-[2]-1 Response between cesium and steel materials, and re-evaporation(Evaluation test for adsorption and re-evaporation behavior)(2)-[2]-2 Evaluation of particulate cesium compounds(Study on the production conditions of particulatecesium)(2)-[2]-5 Analysis of samples collected at Fukushima Daiichi Nuclear
Power Station(Analysis of 1F contamination samples)

The Institute of Applied Energy (IAE)(1)-[1] Comprehensive analysis and evaluation based on the data from the actual units and otherresearch development results(1)-[2] Establishment of a database for comprehensive analysis and evaluation(2)-[1]-3 Accident progression analysis using the analysis code(2)-[2] Analysis and evaluation of chemical forms of FPs(2)-[3] International joint research to make use of knowledge inside and outside Japan(Affiliated organizations)
TEPCO(outsourcing)Mizuho Information & Research Institute, Inc.

(1)-[2] Management and maintenance of DB site and upgrade of display functions(2)-[2] Coding work associated with the FP model introduction(2)-[3] Management and maintenance of web portal for NEA BSAF project(outsourcing)
Advance Soft Corporation(2)-[2] Coding work associated with the FP model introduction©International Research Institute for Nuclear Decommissioning