Subsidy Project of Decommissioning and Contaminated Water Management
Started Since FY2020

Development of Technology for Further Increasing theRetrieval Scale of Fuel Debris and Reactor Internals(Development of Technology for Sorting and Distinction Between Fuel
Debris and Radioactive Waste)

Final ReportAugust 2022
International Research Institute for Nuclear Decommissioning(IRID)All rights reserved.
©International Research Institute for Nuclear Decommissioning (IRID)

No.1

Table of Contents
1. Introduction
Background and Purpose of Research, Goals, Implementation Items, Relation with
Other Research, Implementation Schedule and Project Organization
2. Implementation Details(1) Analytical evaluation of factors influencing measurement errors of prospective
measurement technologies① Selection of influencing factors② Setting the fluctuation range of the influencing factors③ Analytical simulation (④ Including study of the necessity of detailed
analysis and consolidation of technical issues)(2) Study of future research and development plans aiming for application of sorting
technology to actual equipment① Identification of technical issues for the purpose of practical application of
the sorting technology② Study of research and development plan (contents, duration, conditions)③ Consolidation of goals of the main processes
3. Summary©International Research Institute for Nuclear Decommissioning

No.2

About This Project
1. Introduction➢ Research Background and Purpose➢ Objective➢ Implementation Items, their Correlations, and Relations with Other
Research➢ Implementation Schedule➢ Project Organization©International Research Institute for Nuclear Decommissioning

No.3

Research Background and Purpose(Excerpts from solicitation information)[Purpose of developing technology for sorting and distinction between fuel debris and radioactive waste]⚫

During the decommissioning of the Fukushima Daiichi Nuclear Power Station, it is not logical to regard all objects retrieved
from inside the Primary Containment Vessel (PCV) as fuel debris.

⚫

The work from retrieving fuel debris to storing it can possibly be streamlined if a distinction can be made between fueldebris and radioactive waste using the results of measuring the amount of nuclear fuel material contained in the retrieved
objects as a guideline.

⚫

Prospective measurement technologies that can measure the amount of nuclear fuel material have been provided during
the Decommissioning and Contaminated Water Management Project (“Development of Technology for Further Increasing
the Scale of Retrieval of Fuel Debris and Reactor Internals” (FY2019-2020) Same shall apply hereinafter). However,
evaluation of the measurement errors, which is necessary for establishing the method of sorting and distinction based on
the amount of nuclear fuel material, has not been started.

⚫

Hence, it is essential to select factors influencing measurement errors arising from the amount of material other than
nuclear fuel material (water content, quantity of metallic components, quantity of control rods, quantity of concrete
resulting from MCCI (Molten Core Concrete Interaction), etc.) contained in the objects retrieved from inside the PCV, thestatus of filling inside the canisters, etc., and analyze and evaluate the impact that they have on measurement errors, inorder to select the required measurement technologies and equipment for establishing sorting technology in the future,with the measurement technologies*1 studied during the Decommissioning and Contaminated Water Management Projectin mind.
*1) Passive gamma rays measurement, passive neutron measurement, active neutron measurement, radioaparency
measurement, cosmic rays scattering measurement⚫

Technology for the following items will be developed.
① Analytical evaluation of factors influencing measurement errors of prospective measurement technologies② Study of future research and development plans aiming for application of sorting technology to actual equipment©International Research Institute for Nuclear Decommissioning

No.4

Objective[Project goal]
To create the research and development plan aiming for future application to actual equipment and identification of issues in
the measurement technologies, which is required for developing technology for sorting the objects retrieved from the PrimaryContainment Vessel during fuel debris retrieval, into fuel debris and radioactive waste for the purpose of further increasing thescale of fuel debris retrieval.

[Development steps based on the solicitation information and positioning of this year’s research](Step 0) Investigation of the technology concerning sorting and distinction (FY2019-2020)(Step 1) Feasibility study of measurement (Analyzing the possibility) ⇐ This year’s research (TRL2)① Setting the purpose and target (shape, density, etc.) of measurement② Nuclear materials → Evaluating the behavior of radiation incident upon the detector③ Identification, etc. of technical issues contributing to the concept of measurement(Step 2) Development of the measurement concept and re-evaluation of the assumed scenarios(Step 3) Basic design and software development(Step 4) Comprehensive verification tests using test manufacturing, simulated radiation source, etc.

(Step 5) Fabrication of actual equipment(Step 6) Actual operation* The results taking into consideration the positioning of this year’s research in the development plan up to practicalapplication at the start of the project have been referred to for the above-mentioned details. Investigation → Feasibility
Study → Concept → Basic design →Verification test →Fabricating actual equipment →Actual operation are defined as
the steps involved in general development.
©International Research Institute for Nuclear Decommissioning

No.5
Implementation Items, their Correlations, and Relations with Other Research (1/2)[Overview of implementation]

Factors influencing measurement errors arising from the amount of material other than nuclear fuel material containedin the objects retrieved from inside the Primary Containment Vessel, the status of filling inside the canisters, etc., wereselected, with the prospective measurement technologies*1) provided during the FY2019 Decommissioning andContaminated Water Management Project in mind, and their impact on measurement errors was analyzed and evaluatedby means of simulation.

(1) Analytical evaluation of factors influencing measurement errors of prospective measurement technologies
Factors that were expected to influence measurement errors of prospective measurement technologies*1) provided
during the Decommissioning and Contaminated Water Management Project were selected, the fluctuation range ofeach factor for each measurement technology was analytically simulated, and the extent to which each factor
influences the measurement errors was analyzed and evaluated. Also, the necessity to continue further analysis andevaluation in the future, and issues in technological development that will become necessary in the future for reducing
the measurement errors were studied.
(2) Study of future research and development plans aiming for application of sorting technology to actual equipment
Technical issues described above (1) that must be resolved in the future for the purpose of practical application of
the sorting technology were identified based on the evaluation results and technical issues studied in FY2019. Further,
organizing the conditions required according to the mid-and-long term research and development plan for resolving
those technical issues was considered.

*1） Passive gamma rays measurement, passive neutron measurement, active neutron measurement, active/passive neutrons +passive gamma rays measurement, radioaparency measurement, cosmic rays scattering measurement©International Research Institute for Nuclear DecommissioningImplementation Items, their Correlations, and Relations with Other Research (2/2) No.6
The following information was exchanged with other projects.
Input and output information related to the FY2020 Subsidy Project of Decommissioning and Contaminated Water Management- Development of technology for further increasing the retrieval scale of fuel debris and reactor internals(Development of technology for sorting and distinction between fuel debris and radioactive waste)
ID

Requesting project

Providing projectContent (overview)

Required by (time-line)

Use application of information

Remarks

Setting the analytical conditions for analytical
1

Debris Retrieval PJ

Characterization PJ

Fuel debris characterization

November 2020

evaluation of the parameters influencing

Information exchange asmeasurement errors in the measurement

neededtechnologies
Setting the analytical conditions for analytical2

Debris Retrieval PJ

Canister PJ

Shape of canister

November 2020

evaluation of the parameters influencing

Information exchange asmeasurement errors in the measurement

neededtechnologies
Setting the analytical conditions for analytical3

Debris Retrieval PJ

Treatment andShape of the waste storage

Disposal PJ

container

November 2020

evaluation of the parameters influencing

Information exchange asmeasurement errors in the measurement

neededtechnologies
Identification of technical issues for theProcess for containing, transfer
4

Debris Retrieval PJ

Canister PJ

and storage of fuel debris, andfuture development plans

purpose of practical application of the sorting
April 2021

concerning those systems

and distinction technology, and pre-conditions

Information exchange as

for consolidating the goals of the mainneededprocesses in developing the sorting anddistinction technology

Note) In this document, other related projects are expressed with the following abbreviations.
Debris Retrieval PJ: Development of Technology for Further Increasing the Retrieval Scale of Fuel Debris and Reactor Internals
Characterization PJ: Development of Technology for Fuel Debris Characterization and Analysis
Canister PJ: Development of Technology for Containing, Transfer and Storage of Fuel Debris
Treatment and Disposal PJ: Research and Development for Treatment and Disposal of Solid Wastes©International Research Institute for Nuclear Decommissioning

No.7

Implementation Schedule
FY2020
Nov
1

Dec

Jan

FY2021
Feb

Mar

Apr

May

Jun

Jul

Aug

Sept

Oct

Nov

Dec

Jan

Feb

Mar

Analytical evaluation of factors influencing measurement errors of prospective
measurement technologies① Selection of influencing factors

Final settings based on the investigation and fluctuation range settings
Selection of influencing factors
Fact-finding investigation pertaining to the influencing factors② Setting the fluctuation range of the influencing factors(To be reviewed as appropriate) *1
Setting the fluctuation range of the influencing factors*1: Additional study based on the
progress of analytical simulation③ Analytical simulation

Analysis and technical study of measurement technologies / Study of analytical techniques
Setting of base case conditions
Setting of sensitivity analysis conditions(To be reviewed as appropriate) *1

Setting of analysis model
Analytical simulation (Passive neutron measurement)(To be reviewed as appropriate)*1
Completion of analysis▼

Completion of analysis
Analytical simulation (Passive gamma rays measurement) ▼Completion of analysis▼

Analytical simulation (Active neutron measurement)

Completion of analysis▼
Analytical simulation (Passive/active neutron measurement + γ rays measurement)
Completion of analysis
Analytical simulation (X-ray transmission measurement (high energy X-ray CT method)) ▼Completion of analysis▼

Analytical simulation (Cosmic rays scattering measurement (Muon scattering method))④ Study of the necessity of detailed analysis and consolidation of technical issues
2

Study of future research and development plans aiming for application of sorting

Study of the necessity of detailed analysis and consolidation of technical issues

technology to actual equipment① Identification of technical issues for the purpose of practical application of the
Identification of technical issues

sorting technology② Study of research and development plan (contents, duration, conditions)

Study of research and development plan

③ Consolidation of goals of the main processes

Consolidation of goals of the main processes
Project Steering Committee Meeting

Major milestones (Debriefing session / presentation, etc.)▼

Interim Report▼

Project Steering
Final report
Committee Meeting▼▼©International Research Institute for Nuclear Decommissioning

No.8

Project Organization

Tokyo Electric Power Company Holdings, Inc.
○ Fukushima Daiichi Nuclear Power Station
Decommissioning Project Management

International Research Institute for Nuclear Decommissioning (IRID)○○

Coordination of overall planning and technology management
Coordination of technology administration including technology
development progress management

Mitsubishi Heavy
Industries, Ltd.

Hitachi-GE Nuclear
Energy, Ltd.

Toshiba Energy Systems
and Solutions Corporation

Japan Atomic Energy
Agency(1) Analytical evaluation of thesorting technology(Coordination, analyticalevaluation of factors
influencing measurement
errors, review of items
implemented by othercompanies)(2) Study of the future R&D planpertaining to the sorting
technology (Coordination,
study of research anddevelopment plan, review of
items implemented by othercompanies)(1) Analytical evaluation of thesorting technology (Analyticalevaluation of factors
influencing measurement
errors, review of items
implemented by othercompanies)(2) Study of the future R&D planpertaining to the sorting
technology (Study of
R&D plan, review of items
implemented by othercompanies)(1) Analytical evaluation of thesorting technology(Analytical evaluation offactors influencing
measurement errors, review
of items implemented byother companies)(2) Study of the future R&D planpertaining to the sorting
technology (Study of R&D
plan, review of items
implemented by othercompanies)(1) Analytical evaluation of thesorting technology (Setting
up the conditions foranalytical evaluation of factors
influencing measurement
errors, review of items
implemented by othercompanies)(2) Study of the future R&D planpertaining to the sorting
technology (Study of R&D
plan, review of items
implemented by othercompanies)

MHI-NS Engineering Co., Ltd.
・ Analysis of parameters influencing
measurement errors (Setting up theconditions / actual evaluation)
・ Provision of analysis results asfeedback to the equipment
・ Study on various technical issues andmethods for resolving them

Hitachi Industry & Control Solutions, Ltd.
• Analysis of parameters influencing
measurement errors (Development of
input/output conditions, actual evaluation,organization and compilation of analyticaldata)

Project teams to cooperate
for technologicaldevelopment
Development of Analysis andEstimation Technology for Fuel
Debris Characterization
Research and Development for
Treatment and Disposal of Solid
WastesDevelopment of Technology for
Containing, Transfer and Storageof Fuel Debris

Los Alamos National Laboratory• ”Study of applicability of cosmic rays
scattering measurement (Muon scattering
method) under high radiation environment”
related to technological issues for the purposeof practical application of the sorting
technology©International Research Institute for Nuclear Decommissioning

No.9

Items reported in the final report
2. Implementation Details(1) Analytical evaluation of factors influencing measurement errors of prospective measurement
technologies①

Selection of influencing factors②

Setting the fluctuation range of the influencing factors③

Analytical simulation
2.1Analysis conditions
2. 2 Analysis results
2.2.1 Passive neutrons
2.2.2 Passive gamma rays
2.2.3 Active neutrons
2.2.4 Passive/active neutrons + passive gamma rays
2.2.5 X-ray transmission measurement
2.2.6 Cosmic rays scattering measurement©International Research Institute for Nuclear Decommissioning2. Implementation Details – (1) Analytical evaluation of factors influencing
measurement errors of prospective measurement technologies

No.10① Selection of influencing factors
Identification of assumed variation
parameters

The assumed validation parameters for each measurement
technology were identified, including the reason for
selection.

② Setting the fluctuation range of the influencing factors
The variation parameters that were identified above were organized.
Organization of variation
The parameters that were likely to have comparable effects as perparameters
the analytical simulation model were organized.

Organization of variation parameters
required for each measurement
technology

Organization of theresulting influencing factors
Study of the maximum and minimum
values of the parameters

The identified parameters were examind against themeasurement principle for each measurement technology todetermine if they have an impact. At this time, the reason was
indicated as well. Also, the studies of the analyticalparameters was prioritized so that a significant conclusion
can be obtained.
The results, including the validity of the reason, wereorganized.
The maximum and minimum values of the parameters werestudied along with the reasons and basis, while stepping
further up to the measurement technology.

③ Analytical simulation
Determination of analytical conditions andimplementation of analytical simulation

The analytical conditions were determined and analyticalsimulation was implemented for each measurement
technology.
©International Research Institute for Nuclear Decommissioning2. Implementation Details - (1) ① Selection of influencing factors (1/2)
Table

Prospective measurement technologies

SymbolThe parameters that were likely to fluctuate in the case of eachmeasurement technology were identified, and from amongst them,influencing factors (① to ⑩) that were likely to have a significant
influence on the measurement errors were identified. Also, themeasurement techniques that the influencing factors influenced andtheir basis were consolidated.

No.11

Measurement technology

Pn

Passive neutron measurement technology

Pγ

Passive gamma rays measurement technology

An

Active neutron measurement technology

PAn+Pγ

Passive/active neutron measurement + γ rays
measurement technology

X

X-ray transmission measurement (High energy X-ray CT
method)M

Cosmic rays scattering measurement (Muon scattering
method)Table: Results of selecting the factors influencing measurement errors (1/2)
Influencing factor

Measurement
technique that thefactors influence

Basis of the measurement technique

Fuel debris properties
Fuel debris based① Fuel debris
compositionAll

Gd content(percentage)

Pn, An, PAn+Pγ, M

The absorption/scattering effect (energy distribution) of radiation (particularly neutrons) emitted by fuel debris
differs depending on the difference in amount contained (percentage).

B content(percentage)

Pn, An, PAn+Pγ, M

Same as above

MOX

All

②

Burn-upPn, Pγ, An,
PAn+Pγ, M

Since the composition of the fission products (FP) and the actinoids that form the fuel debris changes, thetype of radiation, the absorption/scattering effect of radiation in the fuel debris, etc. differs.

③

FP emission
rate

Pγ, X, M

The rate of emission of FP nuclides along with fuel meltdown.
The rate of emission of volatile FP nuclides at the time of fuel meltdown has an impact on the intensity of
gamma rays and distribution of energy emitted by the fuel debris.

④ Cooling period

Pn, Pγ, An,
PAn+Pγ, M

Same as “②The absorption/scattering effect of radiation differs depending on the difference in fuel debris composition(mixing ratio of fuel components and structures).

The radiation emitted by fuel debris is different than that emitted by uranium fuel. Also, fuel debris
composition changes and the absorption/scattering effect of radiation differs as well.

Burn-up”©International Research Institute for Nuclear Decommissioning2. Implementation Details - (1) ① Selection of influencing factors (2/2)
Table

No.12

Prospective measurement technologies

Symbol

Measurement technology

Pn

Passive neutron measurement technology

Pγ

Passive gamma rays measurement technology

An

Active neutron measurement technology

PAn+Pγ

Passive/active neutron measurement + γ rays
measurement technology

X

X-ray transmission measurement (High energy X-ray CT
method)M

Cosmic rays scattering measurement (Muon scattering
method)Table: Results of selecting the factors influencing measurement errors (2/2)
Influencing factor
Status of storage incontainers⑤

Moisturecontent

Measurement
technique that thefactors influence

Basis of the measurement technique

Pn, An, PAn+Pγ, M

The absorption/scattering effect (energy distribution) of radiation (particularly neutrons) emitted by fuel debris
differs depending on the difference in moisture content.

Measuringsystem⑥ Filling rate

All

Absorption/scattering effect of radiation differs inside the container.

⑦ Uneven
distribution

All

Same as above⑧

ContainerAll

The intensity (beam, etc.) of radiation reaching the detector differ depending on the container.

⑨

Irradiatingradiation
sourceAn, PAn+Pγ, X⑩ Detector

All

The mutual interactions within materials such as fuel debris, etc. differs depending on the energy of theirradiating radiation source (neutrons, X-ray).
The intensity (beam, etc.) of radiation reaching the detector differs depending on the location of the detector,
and thickness of shield and moderator, for each measurement technique.

©International Research Institute for Nuclear Decommissioning2. Implementation Details - (1) ② Setting the fluctuation range of the influencing factors (1/2)

No.13

On the basis of the properties of fuel debris and the principle of each measurement technology, the fluctuation range(minimum to maximum, etc.) of the identified influencing factors （① to ⑩） were set. (Table below)
Table : Results of setting the fluctuation range of the influencing factors (1/2)
Item
Fuel
debrisproperties

Influencing factor① Fuel debris
composition

Basis of the measurement technique

The fluctuation range was set based on
the mixing ratio of fuel components(UO2) and structures (ZrO2, SUS,
concrete).
(Details were organized based on
the analysis conditions.)

Since there can be infinite number of combinations of the mixture of fuel (UO 2) and structures (ZrO2,
SUS, concrete) in fuel debris, the following was assumed as the typical state of fuel debris.
・ Molten debris(Note 1)
Type
Definition
・ Uranium-rich debris
Molten debris
Fuel debris in which (U,Zr)O2 is the major component
・ Metallic debris(Note 1)
・ MCCI debris(Note 1)
Uranium-rich debris
Fuel debris in which UO2 is the major component
Metallic debris

Metallic layer accumulated at the bottom due todifference in density

MCCI debris

Layer of oxides consisting of concrete components

Gd content(percentage)

Minimum Gd content contained in
actual fuel debris to maximum Gd
content after more gets added at thetime of retrieval(1) Combustible Gd contained in fuel, (2) Gd2O3 added to maintain sub-criticality were assumed. (1)is considered to be likely to mix evenly in fuel debris, and hence was assumed to be equivalent to“actual conditions”. The largest amount of (2) was assumed to be likely to be in the container.

B content(percentage)

From 0% to maximum B content after
more gets added at the time of retrieval(1) B contained in control rods, (2) B4C added to maintain sub-criticality were assumed.
However, as control rods melt before fuel rods, considering the likelihood of control rods not mixing
evenly in fuel debris and fuel debris that does not contain control rods being retrieved, (1) was
assumed to be 0%. The largest amount of (2) was assumed to be likely to be in the container.

MOX present or MOX absent

MOX fuel is not loaded in Units 1 and 2. MOX fuel is loaded in Unit 3.

MOX
Fuel
debrisbased

Fluctuation range of the influencing
factor②

Burn-upApprox. 1.3 to approx. 51.3GWd/t

1.3GWd/t is the lowest burn-up per node in Unit 2 excluding the region with natural uranium. It is theleast in Units 1 to 3 and includes the minimum value of other units as well.
51.3GWd/t is the highest burn-up per node in Unit 2. It is the highest in Units 1 to 3 and includes the
maximum value of other units as well.

③

FP emission rate

Emission of volatile FP such as Cs, etc.
absent to emission present (99%)

The rate of emission of volatile FP nuclides at the time of fuel meltdown has an impact on theintensity of gamma rays and distribution of energy emitted by the fuel debris. Hence 0 emission tohigh emission was assumed.

20 to 40 years after the accident

From full-scale retrieval (FY2031) to Completion of decommissioning based on the Mid-and-LongTerm Roadmap.

④ Cooling period(Note 1) “Subsidy Project of Decommissioning and Contaminated Water Management (Development of technology for fuel debris
characterization and analysis)” in the FY2016 Supplementary Budget
Research Report (Interim report） March 2018

©International Research Institute for Nuclear DecommissioningNo.14

2. Implementation Details - (1) ② Setting the fluctuation range of the influencing factors (2/2)
Table : Results of setting the fluctuation range of the influencing factors (2/2)
Item
Status ofstorage incontainers

MeasuringsystemInfluencing factor

Fluctuation range of the influencing factor

Basis of the measurement technique⑤ Moisture content

0.1wt% to 70vol%0.1wt% was set based on the target drying value of 0.1wt% (Note 2) from theCanister PJ, on the premise that drying is performed before measurement.
70vol% was set by rounding up the maximum moisture content of 65.1vol%(residual amount of water considering the causes of error) from the draining testdata.

⑥ Filling rate

10 to 50vol%A fluctuation of ±20% from the filling rate of 30% (based on the information fromthe Subsidy Project of Development of Technology for Containing, Transfer andStorage of Fuel Debris) based on the specifications of the unit can was
assumed.

⑦ Uneven distribution

Following indicate extremely uneven distribution.
・ Radiation source unevenly located at thecenter of the container
・ Radiation source unevenly located near thesurface of the container

If fuel debris is collected in containers, there can be infinite possibilities of
uneven distribution. Hence, extremely uneven distribution was assumed andfirst the extent of impact was verified.

⑧ Container

Following are containers that are being
considered in the projects.
・ Unit can・ Canister
・ Waste storage container (or inner container)

From the perspective of verifying applicability, containers from the Debris
Retrieval PJ and Canister PJ that are currently being implemented, wereconsidered.
・ Unit can・ Canister (Note 2)・ Waste storage container (or inner container）

1.13MeV (average energy)
2.45MeV
14MeV

1.13MeV: Photonuclear reaction due to Bremsstrahlung X-rays
2.45MeV: D-D reaction
14MeV: D-T reaction

Maximum x-ray energy
6MeV
9MeV
15MeV

Selected based on the length up to which X-rays can penetrate through theobjects filled inside the container, and the X-ray energy of the linear electron
accelerator for non-destructive inspection that is available as a product in themarket.

Set for each measurement technology―⑨
Irradiatingradiation
sourceActive neutron
measurement
X-raytransmission
measurement⑩ Detector(Location of the detector,
shielding thickness, etc.)(Note 2) Subsidy Project of Decommissioning and Contaminated Water Management in the FY2018 Supplementary Budget - Development of Technology for Containing, Transfer and Storage of Fuel Debris(Note 3) Subsidy Project of Decommissioning and Contaminated Water Management in the FY2014 Supplementary Budget - Development of Technology for Containing, Transfer and Storage of Fuel Debris

Research Report (Final Report) March 2021
Research Report (Final Report) March 2017

©International Research Institute for Nuclear DecommissioningNo.15

2. Implementation Details(1) Analytical evaluation of factors influencing measurement errors of prospective
measurement technologies① Selection of influencing factors② Setting the fluctuation range of the influencing factors③ Analytical simulation
2.1Analysis conditions
2. 2 Analysis results
2.2.1 Passive neutrons
2.2.2 Passive gamma rays
2.2.3 Active neutrons
2.2.4 Passive/active neutrons + passive gamma rays
2.2.5 X-ray transmission measurement
2.2.6 Cosmic rays scattering measurement©International Research Institute for Nuclear Decommissioning2. Implementation Details – (1) ③ Analytical simulation (Division of implementation)

No.16

I.

Development of fuel debris radiation source model: Japan Atomic Energy Agency (JAEA)

II.

Analytical simulation of each measurement technique:

As per the table below
Applicable code (Cross-sectional
library) (*2)

No.

Measurement technology (*1)

1

Passive neutron measurement technology

MHI*
Hitachi-GE**MCNP5 (JENDL4.0)
PHITS 3.20 (JENDL4.0)

2

Passive gamma rays measurement
technology

MHI
Hitachi-GE

MCNP5 (JENDL4.0)
PHITS 3.20 (JENDL4.0)

3

Active neutron measurement technology

MHI
Hitachi-GE

MCNP5 (JENDL4.0)
PHITS 3.20 (JENDL4.0)

4

Passive/active neutron measurement +Gamma rays measurement technology

MHI

MCNP5 (JENDL4.0)

5

X-ray transmission measurement (Highenergy X-ray CT method)

Hitachi-GE

PHITS 3.20 (JENDL4.0)

6

Cosmic rays scattering measurement(Muon scattering method)

Toshiba ESS
Corporation***

MCNP6 (ENDF/B-VII)

Analysis led by*1: Measurement technology selected as a prospective technology that can be applied for sorting and distinction, during the FY2019 study.
*2: The applicable code and cross-sectional library is different for each company, but in the case of the measurement technologies for which the analysis work will be split between MHI and
Hitachi-GE, trial calculations were performed while keeping the calculation system and radiation source conditions the same, the results of those calculations were compared, and it wasverified that results meeting the goals of the feasibility study this time can be obtained.
The applicable code and cross-sectional library of Toshiba ESS is used extensively and has a proven track record.
*Mitsubishi Heavy Industries, Ltd.: MHI**Hitachi GE Nuclear Energy, Ltd.: Hitachi GE***Toshiba Energy Systems and Solution Corporation: Toshiba ESS©International Research Institute for Nuclear DecommissioningNo.17

2. Implementation Details – (1) ③ Analytical simulation（Overall flow）① Selection of influencing factors② Setting the fluctuation range of the influencing factors
Particle transport simulation(Neutron, photon, electron, muon)③ Analytical simulation
Development of the fuel debris
radiation source model (data on every material)
Development of model for the fuel debris collectedin containers considering the influencing factors(material mixing conditions, type of container, etc.)

Muon
Container

Hypothetical detector(Tally)Spontaneous fission
neutron source(Cm244, Pu240)(n,g)

Implementation of particle transport simulation
with the Monte Carlo method (MCNP, PHITS code)for every measurement technology

Neutrons
U235, Pu239
induced fission
FP (Cs137, Eu154, Co60)
PhotonEvaluation of the beam in front of the detector
for each sensitivity analysis case
(Including the detector response dependingon the method or case)Collected material©International Research Institute for Nuclear Decommissioning2.1Analysis conditions – Development of fuel debris radiation source model (1/2)

No.18

A model depicting the basic composition, photon source, and neutron source to be used for analyzing various types of fuel debris was developed.

Conditions for creating radiation source depending on the influencing factors(Conditions in blue are base cases)
Divided into 24 partsin the axial direction

Combustion compositionof fuelActivation composition of
the structures
Approx.
1,600 nuclides
Volatile FP emission simulation

3D nuclide inventory data (Note 1)
Example of Unit 2: 13,152 regions(Burnable poison and trace
impurities are considered)・ Region with minimum burn-up (1.3GWd/t)(Note 2)

Emission rate
Element(%)
元素
放出率(%)
Noble
gas
希ガス
99

・ Entire reactor core region (23GWd/t)

I
Cs
Te
Mo
Rb
Cd
Ba
Sb
Pd
Ag
Tc
Ru
Sr

・ Region with maximum burn-up (51GWd/t)・ High emission rate (99%) model・ Model based on the FP emission test (Note 3)
・ No emission model

Noble gas, Cs, etc.
19 elements

・ Duration of decay after accident: 40 years (Around FY2051)

Calculation of nuclidedecay

・ Duration of decay after accident: 30 years (Around FY2041)

・ Time period of starting further scale up
Approx.
1,600 nuclides (20 years： around FY2031)

Development of radiation model per unit volume element (Creation of photon
sourceComposition for calculating
particle transport

A line spectrum of approx. 33,500
units was created from the nuclear
data (ENDF/B-VIII.0 decay library),and radiation source with major
contribution of (ExI) and I, where I is
the strength and E is the energy, andradiation source of major nuclides(Eu154, etc.) that are targeted in thepassive gamma rays method were selected.
For the rest, a 47 groupstructure (10keV-20MeV)
was used to reduce thenumber of groups.

:voxel) for each radioactive material (UO2/MOX, ZrO2, SUS)

Nuclides that largely contribute to
the absorption reaction rate and theoverall reaction rate were selected(approx. 99.8% of the whole) fromnuclides that can be handled with
the nuclide transport library of the
MCNP, PHITS, GEANT codes.

Creation of neutron
source
The total number of spontaneous fission (SF) nuclides
was obtained considering SF nuclides (60 nuclides) from
Th232 to Fm257. The neutron energy spectrum was
expressed in terms of Watts for the effective massrepresented by the nuclides with maximum contribution(Cm224 or Pu240).

97
84
80
77
53
44
35
30
27
9.2
7.1
1.8
1.4

The percentage ofemission (100% - emission
rate = residual ratio) until itanchors as fuel debris after
the accident.

SF neutron source

Energy (MeV)

γ raysEnergy (MeV)(Note 1) Atomic Energy Society of Japan (AESJ) FY2021 Spring Annual Meeting (3B01, 3B02), FY2021 Fall Meeting (1I06).
(Note 2) Represented by Unit 2 as Unit 2 encompasses the burn-up range of Units 1 and 3. The region with natural uranium is notincluded while indicating the minimum burn-up.
（Note 3） PHEBUS-FPT4 test： Nucl. Eng. and Technol. 38(2), pp.163-174 (2006).

©International Research Institute for Nuclear Decommissioning2.1Analysis conditions – Development of fuel debris radiation source model (2/2)
Development of voxel-based mixed fuel debris model for simulation⚫ Handling of a great number of sensitivity analysis cases in an integrated manner⚫ Avoiding unrealistic models by using correlated parameters(Example of correlation: The average density of fuel debris with a higher
uranium concentration tends to be high.)

No.19

Fuel debris material (voxel)
UO2 (contains BP)

MOX (contains BP)

ZrO2

B4C
Control rods / criticality
prevention material

A variety of fuel debris was simulated by adjusting the number of voxels
for each material collected in the hypothetical container.
(The radiation source is proportionate to the number of voxels.)

Mixed uniformly and collected in various containers (simulation model)

SUS

Gd2O3
Criticality prevention
material

H2O

Concrete

For adjusting the
moisture content
Empty
For adjusting the filling rate
and apparent density

Unit canSize：1cm3
Weight: Material density： Radiation source ispresent.

18 types of voxel data were created including the difference in theradiation source model.
➡ A variety of fuel debris was simulated by using various combinations.

Canister
Waste storage
container

Molten debris: (U0.5, Zr0.5)O2
Uranium-rich debris
MCCI (Molten Core ConcreteInteraction) debris
Metallic debris©International Research Institute for Nuclear Decommissioning2.1Analysis conditions - Base case analysis conditions (1/2)
◆

No.20

Setting of base case analysis conditions(Sensitivity analysis: The sensitivity with respect to the minimum to maximum analysis results of individual influencing factors (parameters) was investigated on
the basis of the analysis results (beam, etc.) based on these conditions.)

Table: Base case analysis conditions
Item

Parameters

Values23GWd/tBurn-up
Radiation
sourceRadiationtransport
computation
container
model

Recent average burn-up of the Unit 2 reactor core planned to be investigated (Note 1)

FP emission rate

Standard emission model

Cooling period

20 yearsTime period of start of full-scale retrieval (FY2031)

Unit canContainer: Smallest single container
The mesh structure on the side and bottom was not modeled (Thickness wasconsidered as well)
Height: Area contributing to the dose rate, represented by 200mm wherein the amount ofradiation source becomes smaller
Other information about the container: According to the information from the Canister PJ and
Debris Retrieval PJ

Active neutron
measurement

14MeV

D-T reaction

X-ray transmission
measurement

9MeV

Has the ability to penetrate through the measurement target, shows intermediate values for the Xray energy of the linear electron accelerator for non-destructive inspection that is supplied as a
product.

Target container

Irradiatedradiation energy

Radiationtransport
computationfuel debris
composition
model

Approach towards setting the conditions

Emission rate based on the FP emission test (Phebus-FPT4)

Fuel debris composition(Volume ratio)

UO2: 50%
ZrO2: 50%
SUS: 0%
Conc: 0%
Gd2O3: 0%
B4C: 0%
MOX: 0%Empty (porosity): 0%・ Assumed to be molten debris. (mixture of UO2 and ZrO2)
The mixing ratio was according to the information from the Characterization PJ.
・ The combustible Gd contained in the fuel was considered in the UO 2 composition.
(Gd2O3（Note 2）was set to be absent.)
・ As control rods melt before fuel rods, they were assumed to not mix uniformly in the fueldebris. (B4C（Note 3）was set to be absent.)
・ MOX fuel was not considered.
・ The porosity was considered to be 0%

Moisture content (wt%)

1wt%

Value set based on the risk of hydrogen burning. (During the transport period (7 days), hydrogenconcentration inside the canister reduces the 1.5wt% moisture content that has reached 4vol%.)
According to the information from the Canister PJ

Filling rate (vol.%)

30vol%(Empty: 70vol%)According to the information from the Debris Retrieval PJ*

Uneven distribution-

The homogeneous model was used (UO2, ZrO2, H2O, Empty were uniformly distributed inside theunit can.)(Note 1) JAEA-Data-Code-2012-018, (Note 2) Criticality prevention material added for maintaining criticality was assumed.
(Note 3) Absorption material inside the 1F control rods was assumed.
*Subsidy Project of Development of Fuel Debris Retrieval©International Research Institute for Nuclear Decommissioning2.1 Analysis conditions - Base case analysis conditions (2/2) No.21
Table : Pre-conditions (Container and density)
Item

Container

Conditions

Values

Basis

Depth

200mm

Outer
diameter

210mm

Inner
diameter

206mm

Wall
thickness

Side: 2mm
Bottom: 5mm

Material

SUS316L

According to the information from the Canister PJ

Density

7.98g/cm3

According to JIS G 4304 (SUS316L)

UO2

10.525g/cm3

Actual conditions

ZrO2

5.56g/cm3

Actual conditions

H2O

1.0g/cm3

From the Chronological Scientific Tables FY2018 (Set by rounding up the density of
0.99820g/cm3 at 1atm and 20℃)

Air

0g/cm3

Handled as void

Area contributing to the dose rate, represented by 200mm wherein the amount ofradiation source becomes smaller
Planned value (According to the information from the Debris Retrieval PJ)

The mesh structure on the side and bottom was not considered.

Unit canActualdensity

FigureProposed structure of the unit can

From the Subsidy Project of Decommissioning and Contaminated Water Managementin the FY2018 Supplementary Budget - Development of Technology for Containing,
Transfer and Storage of Fuel Debris Research Report (Final Report) March 2021

Table : Policy for developing models

Image illustrating the assumed state of containing of fuel debris(example)
Air (Empty) or Water (H2O)

Fuel debris(UO2, ZrO2)

It was assumed that clumped
fuel debris is randomly filled in
the container.
Container (Unit can)Analysis model
Abundance of each component material(UO2, ZrO2, H2O, Empty) was mixed
throughout the container.
Further, the filling rate was modeled byhomologizing it with the apparent densityof fuel debris with respect to the total
capacity of the fuel debris canister.

The model of the detector system
conforms to each measurement
method.
©International Research Institute for Nuclear Decommissioning2.1 Analysis conditions - Procedure of sensitivity analysis
⚫

No.22

The variations in the beam, etc. that reaches the detection surface were verified based on the fluctuation range of the influencing
factor as per the procedures given below.
1.

With the results obtained using base case conditions as a guideline, individual influencing factors were changed to verify

variations.
2.

From the above-mentioned results, in order to understand physical phenomenon and enable systematic speculation
regarding variations that occur when multiple influencing factors are closely connected, cases wherein multiple influencing
factors were changed were analyzed as typical cases as well.
② Burn-up:
燃焼度： Unit2号炉心平均②
2 core average③
emission試験(Phebus-FPT4)ベース
rate: Test (Phebus-FPT4) base
③ FP
FP放出率：④
period:④ Cooling
冷却期間：20年20 years⑨ 照射放射線源：⑨ Irradiatingradiation source:
14MeV(※
中性子計測)
14MeV(*Activeneutronmeasurement)
９ＭｅＶ(*(※X-ray
X線透過計測:最大) measurement: Maximum
9MeV
transmission
energy)①
debris composition:①Fuel
燃料デブリ組成:UO2: 50 (vol%)ZrO2: 50 (vol%)SUS,
concrete: 0:0(vol%)SUS,(vol%)
Radiation
reaching the detector
検出器に到達する放射線
B4C, Gd2O3: 0 (vol%)
MOX: 0 (vol.%)
Empty(porosity):
Empty(気孔率) : 00(vol%)(vol%)⑤
content:
1 (wt%)⑤Moisture
含水率：1
(wt%)
⑥
rate: 30 (vol%)⑩ 検出器⑥Filling
充填率：30(vol%)⑩ Detector⑦
distribution:
None (=uniform)(検出器位置、遮蔽厚等)(Detector
position, shielding thickness, etc.)⑦Uneven
偏在：なし(=均質)⑧ Container:
容器：ユニット缶⑧
Unit can(φ210mm×H200mm)⚫

Parameter values in the figure indicate base case conditions.

Based on the above analytical evaluation, in order to reduce (predict) the variations caused by influencing factors, “④ Study of thenecessity of detailed analysis and consolidation of technical issues” will be implemented, and will be linked to future research anddevelopment planning.
©International Research Institute for Nuclear Decommissioning2.1Analysis conditions - Influencing factors to undergo sensitivity analysis for the measurement
techniques

No.23

Influencing factors to undergo sensitivity analysis (parameter study) for the measurement techniques are indicated by “○” inthe table. Its basis has been listed as well.
Measurement techniqueInfluencing factor

Passive
neutron

Passive γ

Active neutron〇〇

Gd content (percentage)〇

B content (percentage)

Combination

Basis, etc.

X-ray

Cosmic rays

(Note 1)

〇

〇

〇

〇

The absorption/scattering effect of radiation differs depending on the fuel debris
composition.

－

〇

〇

－

〇

The absorption/scattering effect (energy distribution) of radiation (particularly neutrons)
emitted by fuel debris differs depending on the difference in amount contained(percentage).

〇

－

〇

－

－

〇

Same as above

MOX

〇

〇

〇

－

〇

〇

The source of radiation emitted by fuel debris is different. Also, as the fuel debris
composition changes, the absorption/scattering effect of radiation differs as well.

② Burn-up

〇

〇

〇

〇

－

〇

Same as above③ FP (fission product)emission rate

－

〇

－

－

〇

〇

Same as above④ Cooling period

〇

〇

〇

－

－

〇

Same as above⑤ Moisture content

〇

－

〇

－

－

〇

The absorption/scattering effect (energy distribution) of radiation (particularly neutrons)
emitted by fuel debris differs depending on the difference in moisture content.

⑥ Filling rate

〇

〇

〇

〇

〇

〇

Absorption/scattering effect of radiation differs inside the container.

⑦ Uneven distribution

〇

〇

〇

〇

〇

〇

Same as above⑧ Container

〇

〇

〇

－

〇

〇

The intensity (beam, etc.) of radiation reaching the detector differs depending on thecontainer.

⑨ Irradiating radiation source

－

－

〇

－

〇

－

The mutual interactions within materials such as fuel debris, etc. differ depending on
the energy of the irradiating radiation source (neutrons, X-ray).
(*) Only active n measurement, radioparency measurement⑩ Detector(Location of the detector,
shielding thickness, etc.)〇(Note 3)〇(Note 3)〇(Note 3)－〇(Note 3)〇

The intensity (beam, etc.) of radiation reaching the detector differs depending on thelocation of the detector, and thickness of shield (moderator), for each measurement
technique.

① Fuel debris composition(Note2)(Note 1) Passive/active neutrons + passive gamma rays
(Note 2) Influence of all factors was investigated except ⑨ that is not related to cosmic rays scattering measurement (Muon scattering method). Only the influence of change in composition was handled in the 4cases of MOX, burn-up, FP emission rate and cooling period.
(Note 3) representative detector position, moderator, etc. were set.
©International Research Institute for Nuclear Decommissioning2.1Analysis conditions - Sensitivity analysis conditions (1/2)

No.24

The analysis conditions for the influencing factors (① to ⑩) to undergo sensitivity analysis (parameter study) were set.
Table: Sensitivity analysis conditions (1/2)
Item

Influencing factor

Fluctuation range of the influencing factor

Analysis conditions

Fuel debris
properties① Fuel debris
composition

The fluctuation range was set based on the mixing
ratio of fuel components (UO2) and structures (ZrO2,
SUS, concrete).

Since there can be infinite number of combinations of the mixture of fuel (UO2) andstructures (ZrO2, SUS, concrete) in fuel debris, the following was assumed as the
typical state of fuel debris.
・ Molten debris(Note 1)
・ Uranium-rich debris
・ Metallic debris(Note 1)
・ MCCI (Molten Core Concrete Interaction) debris (Note 1)
Total 8 cases. (Refer to No. 26 for details.)

Gd content(percentage)

Minimum Gd content contained in actual fuel debris
to maximum Gd content after more gets added at thetime of retrieval(1) Actual conditions ← Base
(2) Set assuming Gd2O3 is added to maintain sub-criticality.

B content(percentage)

From 0% to maximum B content after more gets
added at the time of retrieval (Note that, as control
rods melt before fuel rods, considering the likelihoodof control rods not mixing evenly in fuel debris andfuel debris that does not contain control rods being
retrieved, the minimum content was assumed to be0%.)(1) 0% ← Base
(2) B4C/UO2 volume ratio = 0.034(Equivalent to actual conditions: Core average of the amount of B mixed before theaccident)(3) Set assuming B4C is added to maintain criticality.

MOX

MOX present or MOX absent(1) MOX absent (Units 1 and 2) ← Base
(2) MOX present (Unit 3)② Burn-up

Approx. 1.3 to approx. 51.3GW/t(1) Approx. 1.3GWd/t (Lowest burn-up per node)(2) Approx. 23Gwd/t (Average) ← Base
(3) Approx. 51.3GWd/t (Highest burn-up per node)③ FP emission rate

Volatile FP emission absent to emission present(99%)(Refer to No. 14 for details.)(1) Zero emission(2) Emission rate based on the FP emission test (Phebus-FPT4) ← Base
(3) High emission (99%）④ Cooling period

20 to 40 years after the accident(1) 20 years after the accident ← Base(2) 30 years after the accident(3) 40 years after the accident

Fuel debris
based(Composition offuel components(UO2, ZrO2) andintensity ofradiation source)(Note 1) “Subsidy Project of Decommissioning and Contaminated Water Management (Development of technology for fuel debris characterization and analysis)”in the FY2016 Supplementary Budget
Research Report (Interim report） March 2018
©International Research Institute for Nuclear DecommissioningNo.252.1Analysis conditions - Sensitivity analysis conditions (2/2)
Table: Sensitivity analysis conditions (2/2)
Item

Influencing factor

Fluctuation range of the influencing factor

Analysis conditions

Status ofstorage incontainers⑤ Moisture content

0.1wt% to 70vol%(1) 0.1wt%(2) 1wt% ← Base
(3) Set upon seeing the results of (1), (2), and (4) and considering
the necessity.
(4) 70vol%0.1wt% was set based on the target drying value of 0.1wt% (Note 1) fromthe Canister PJ, on the premise that drying is performed before
measurement.
70vol% was set by rounding up the maximum moisture content of
65.1vol% (residual amount of water considering the causes of error) (Note2) from the draining test data.
⑥ Filling rate

10 to 50vol%A fluctuation of ±20% from the filling rate of 30% (based on theinformation from the subsidy project of Development of Fuel Debris
Retrieval) based on the specifications of the unit can was assumed.

⑦ Uneven distribution

Following indicate extremely uneven distribution.
・ Radiation source unevenly located at the center of the container
・ Radiation source unevenly located near the surface of the container

Same as on the left
(Refer to No. 27 for details)

Following are containers that are being considered in the projects.
・ Unit can・ Canister
・ Waste storage container (or inner container)(1) Unit can ← Base
(2) Canister(3) Waste storage container(Refer to No. 28 for details.)

Activeneutronmeasurement

1.13MeV (average energy)
2.45MeV
14MeV(1) 1.13MeV: Photonuclear reaction due to Bremsstrahlung Xrays(2) 2.45MeV (D-D reaction)(3) 14MeV (D-T reaction) ← Base

X-ray
transmission
measurement

Maximum x-ray energy
6MeV
9MeV
15MeV

⑧ Container

Measuringsystem⑨
Irradiatingradiation
source(1) 10vol%(2) 30vol% ← Base(3) 50vol%⑩ Detector(Location of the detector,
shielding thickness, etc.)

Set for each measurement technology(1) 6MeV(2) 9MeV ← Base(3) 15MeV
Same as on the left
(For details, refer to the models of the measurement techniques
in Section 2.2.)(Note 1): Subsidy Project of Decommissioning and Contaminated Water Management in the FY2018 Supplementary Budget - Development of Technology for Containing, Transfer and Storage ofFuel Debris
Research Report (Final Report) March 2021
(Note 2): Subsidy Project of Decommissioning and Contaminated Water Management in the FY2014 Supplementary Budget - Development of Technology for Containing, Transfer and Storage ofFuel Debris
Research Report (Final Report) March 2017

©International Research Institute for Nuclear Decommissioning2.1 Analysis conditions - Sensitivity analysis conditions - Fuel debris composition (Details) No.26➢ Fuel debris composition based on the FY2018 report of the Characterization PJ (Molten debris < Base >, metallic debris, MCCI
(Molten Core Concrete Interaction) debris) ＋ Uranium-rich debris was assumed.
➢ The sensitivity of nuclear fuel components, SUS and concrete to fluctuations in fuel debris components was investigated.
(Since components other than UO2 become shielding material, cases in which the detection efficiency gets worsened andcases in which measurement is affected were investigated.)

Chart enumerating the fuel debris composition (voxel) and volume ratio (vol%)Composition inside the container (Note 1)
Fuel debris components

Fuel debris

MoltendebrisUranium-richdebris

Other
Gd2O3

B4C

Reason (basis)

UO2

ZrO2

SUS

Concrete

15%
(50%)

15%
(50%)

0%

0%

Based on FY2018 report of the Characterization PJ*

7.5%
(25%)

22.5%
(75%)

0%

0%

Sensitivity investigation based on fluctuations in nuclear
fuel components

30%(100%)

0% (0%)

0%

0%
Base:
0%

Base:
0%

Fluctuations
consideredbased onsensitivity
analysisFluctuations
consideredbased onsensitivity
analysis

H2O

Base: 1wt%

Empty

Base:
Volumeexcluding fueldebris + H2O

Sensitivity investigation based on fluctuations in nuclear
fuel components

Metallic
debris

0.075%(0.25%)

0.075%(0.25%)

29.85%(99.5%)

0%

MCCI

1.05%
(3.5%)

1.05%(3.5%)

7.2%
(24%)

20.7%
(69%)

15%
(25%)

15%
(25%)

30%
(50%)

0%

15%
(25%)

15%
(25%)

0%

30%
(50%)

Sensitivity investigation based on fluctuations in
concrete components

15%(33.3%)

15%(33.3%)

0%15%(33.3%)

Sensitivity investigation based on fluctuations in
concrete components(Only neutron measurement)

Sensitivity
investigation
based onothermaterial
componentsthat form
fuel debris

Fluctuations
consideredbased onsensitivity
analysisFluctuations
consideredbased onsensitivity
analysisBased on FY2018 report of the Characterization PJ
Based on FY2018 report of the Characterization PJ
Sensitivity investigation based on fluctuations in SUS(Note 1) Percentage inside the container. (Percentage when fuel debris components are considered as 100% is mentioned inside parentheses.)*Subsidy Project of Development of Fuel Debris Characterization©International Research Institute for Nuclear Decommissioning2.1 Analysis conditions - Sensitivity analysis conditions - Uneven distribution model No.27■ System for analyzing cases of uneven distribution
Nuclear fuel material (UO2 and ZrO2) and other material was separated and distributed unevenly in the container.
Example: Cases of uneven distribution of MCCI (Molten Core Concrete Interaction) analyzed with the Passive γ, neutron, Active neutron
methods: At the center in the horizontal direction (#1-24) and on the exterior surface in the horizontal direction (#1-25)

Top view

MCCI at thecenter in thehorizontal
direction

Transverse cross-sectional view (A-A’)

Material other than
nuclear fuel (SUS +Concrete + H2O)

20.6
10.3

Unit can
4.36

9.75

A

A’
9.75

Nuclear fuel material(UO2 + ZrO2)

MCCI on theexterior
surface in thehorizontal
direction

Material other than
nuclear fuel (SUS +Concrete + H2O)

Nuclear fuel material(UO2 + ZrO2)

20.6
10.3

Unit can
4.36

A

Unit [cm]

2.18

9.75

A’
9.75

©International Research Institute for Nuclear Decommissioning2.1

Analysis conditions - Sensitivity analysis conditions – Shape of the container
Type of container

No.28

Evaluation model (Unit: mm)

Shape of container

φ210

1. Unit can
φ210

φ 210mm × 200mmH
Plate thickness 2mm

FueldebrisFigure: Proposed structureof the unit can[Reference: Subsidy project ofDevelopment of Fuel Debris
Retrieval]
Unit canFrom the Subsidy Projectof Decommissioning and Contaminated Water
Management in the FY2018 Supplementary
Budget - Development of Technology for
Containing, Transfer and Storage of Fuel Debris
Research Report(Final Report) March FY2021

200

5Unit can2
Vertical cross-section

Horizontal cross-section

φ240

25

2. Canister
Canister

φ 240mm × 928mmH
Plate thickness 10mm
10[Reference: Subsidy project ofDevelopment of Technology for
Containing, Transfer and Storageof Fuel Debris]Figure Proposed structure of canister(simple installation structure / inner diameter
220mm / without air supply mechanism)
From the “Subsidy Projectof Decommissioning and Contaminated Water Management in
the FY2018 Supplementary Budget (Development of Technology
for Containing, Transfer and Storage of Fuel Debris)”
FY2020
Final Report dated June FY2021

928
30
φ240
Fuel debris
Unit canVertical cross-section

Horizontal cross-section
500

3. Inner waste container

500

500mm□× 300mmH
Plate thickness 2mm

Inner waste container
500(Waste storage container)[Reference: Subsidy Project ofResearch and Development ofTreatment and Disposal of Solid
Waste]

300
Figure: Proposed structure of the waste storage
container
From the Subsidy Projectof Decommissioning and Contaminated Water
Management in the FY2018 Supplementary Budget(R&D for Treatment and Disposal of Solid Wastes)”
FY2019
Accomplishment Report dated December FY2020

2

Horizontal cross-section

2
Fuel debris
Inner waste container
Vertical cross-section©International Research Institute for Nuclear DecommissioningNo.29

2. Implementation Details(1) Analytical evaluation of factors influencing measurement errors of prospective
measurement technologies① Selection of influencing factors② Setting the fluctuation range of the influencing factors③ Analytical simulation①
Fuel debris
composition
2.2.1
Passive n

No. 37
No. 40No. 42
No. 45

2.1 Analysis conditions
2. 2 Analysis results
2.2.1 Passive neutrons
2.2.2 Passive gamma rays
2.2.3 Active neutrons
2.2.4 Passive/active neutrons + passive gamma rays
2.2.5 X-ray transmission measurement
②
③
④
⑤
⑥
⑦
⑧
2.2.6 BCosmic
rays
scattering
measurement
Gd
MOX
Burn-upFP
Cooling
Moisture
Filling
UnevenContainer
content

content

No. 52
to 54

No. 52
to 54

No. 42

No. 38
No. 43emission
rate

periodcontent

rate

distribution－

No. 41
No. 46

No. 39No. 44

No. 37
No. 40No. 42
No. 45

No. 58to 62

No. 50⑨
Irradiatingradiation
source－⑩Detector

No. 47
No. 48
No. 56
No. 72
to 75©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.30

2.2.1 Passive neutrons - Measurement concept⚫

Fuel debris contains nuclides (Cm-244, etc.) that originate from fuel and emit neutrons as a result of spontaneous fission.

⚫

In particular, since Cm-224 predominantly becomes a neutron generating nuclide as combustion progresses, it is assumed
that this nuclide will be measured.

⚫

However, the neutron beam becomes relatively small when the neutrons get absorbed due to the moisture contained in fueldebris or due to the neutron absorption material (Gd,B4C) for maintaining sub-criticality, or when fuel has a low burn-up.

⚫

Thus, the width of the neutron beam that reaches the measuring surface varies depending on the properties of fuel debris.

⚫

Hence, by determining the width to be measured based on the analytical evaluation, issues such as themeasurement range of the detector required for the measurement system, adjustment of the measurement

Neutron spectrum (n/cm2/s/△u)

distance or the neutron moderator system, etc. were identified.

Container

Neutron detector

Fuel debris

EnergyExample of analysis output (Moderator: Polyethylene)

Neutron©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.31

2.2.1 Passive neutrons - Analysis model

In order to make it possible to study the changes in the radiation flux depending on the distance from thecontainer, multiple layers of space were provided for cylindrical detectors at a fixed distance from the container(unit can, etc.).
→ The detectors were selected and placed based on the changes in the radiation flux.
The changes were used for studying shielding, etc.

Unit canFuel debris

200mm

A

Fuel debris

A’

10mm

150mm
1016mm

Φ210mmUnit canT-cross Tally
T-cross Tally

Analysis model top view

A-A’ cross sectional view

©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.32

2.2.1 Passive neutrons– Analysis model
Flow of study from beam to the detector installation location, detector response, etc. （PHITS analysis case)

The installation method, etc. of the prospective detector was studied for the base case as per theflow mentioned below.
The prospective detector was selected based on the detection sensitivity (cps/nv) and γ tolerance dose rate (Gy/h) → He-3, B-10, etc.
* Detection sensitivity: Sensitivity with respect to neutrons (cps/nv), γ tolerance dose rate: maximum γ dose rate (Gy/h) such that γ/n can be differentiatedAcquisition of beam data

⓪ Calculation cases for detector simulation
were selected from the energy spectrum of theneutron flux originating from fuel debris.

Following are 2 beam data. Output result of MCNP/PHITS, etc.
・Neutron flux originating from fuel debris (no moderator)
・ Gamma flux originating from fuel debris* Neutron / gamma flux = Energy flux Φ(E)[1/cm 2/s] from every neutron / gamma ray

Fuel debris① Distance (mm) from the container to thedetector or the shielding thickness (mm) was
set based on the dose rate (Gy/h) of the gamma
flux originating from fuel debris and the gamma
tolerance (Gy/h) of the prospective detector.

③ Cadmium thickness [mm] ② Moderator thickness [mm](For preventing back-scatter) (For thermalizing fast components)

Moderator(Polyethylene, etc.)

Cadmium

Detector(He-3, B-10, etc.)② Moderator thickness (mm) required for
decelerating the fast neutrons was set based on theenergy spectrum of the neutron flux originating fromfuel debris.
③ Cadmium thickness (mm) required forcontrolling back-scattering of thermal neutrons wasset based on the energy spectrum of the neutron
flux when a moderator is present.
Detector simulation was performed with the abovementioned conditions for determination (calculation
cases, detector position, etc.).
(Big picture image)Measurement time for each container was studied
based on the detector simulation results

Big picture (Image)Moderator (polyethylene)

Unit can① Distance [mm] from the container to the detector(For bringing it down up to the γ tolerance dose rate)

Cadmium
Unit canFuel debris

Detector
Applicability assessment items
1) Placement of detector … Parameters set for ① to ③ for each individual prospective detector
2) Measurement time … Calculated in terms of count rate (cps) from neutron flux (nv) and detector sensitivity (cps/nv)
for each individual prospective detector after the installation of ① to ③
Preliminary calculation of measurement time was performed based on the required statistical parameters.
3) Possibility of detection … Whether or not each individual prospective detector can be used, based on themeasurement time limitations©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.1 Passive neutrons - Table listing the analysis cases (1/3)No.33•

Considering the analysis conditions described in No. 2-1 as the base case, the analysis conditions that changed
to influencing factors (No. 2-2 to 29) were assumed as the sensitivity analysis conditions. All 29 cases
Influencing factors that changed from the base case under the sensitivity analysis conditions are highlighted
blue. Here, the parameters that changed in association with the above-mentioned changes are indicated byyellow hatching.

•Composition inside the container *1
Case No.

Type of fueldebris

2-1

2-2

2-3

2-4

2-5

2-6

2-7

2-8

2-9

2-10

Molten debris

Within the filling factor

FP emission

Cooling

Unevenrate

perioddistribution

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

Burn-up
Outside the filling factor

Composition
UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)UO2 : 7.5vol%(25vol%)ZrO2 : 22.5vol%(75vol%)MOX : 30vol%UO2 : 5vol%(50vol%)ZrO2 : 5vol%(50vol%)UO2 : 25vol%(50vol%)ZrO2 :25vol%(50vol%)UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)Container

Changed parameter

Total (Filling factor)
30vol%

30vol%

30vol%

10vol%

50vol%

H2 O (water content): 1wt%
Empty : Remainder
H2 O (water content): 1wt%
Empty : Remainder
H2 O (water content): 1wt%
Empty : Remainder
H2 O (water content): 1wt%
Empty : Remainder
H2 O (water content): 1wt%
Empty: Remainder

0GWd/tStandard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

Gd2O3 : 3vol%
30vol%

H2 O (water content): 1wt%
Empty: Remainder
Gd2O3 : 30vol%

30vol%

H2 O (water content): 1wt%
Empty: Remainder
B4C : 0.51vol%

30vol%

H2 O (water content): 1wt%
Empty: Remainder
B4C : 10vol%

30vol%

H2 O (water content): 1wt%
Empty: Remainder

30vol%H2O (water content): 0.1wt%
Empty : Remainder

Unit can(Φ210mm x H200mm)
Unit can(Φ210mm x H200mm)
Unit can(Φ210mm x H200mm)
Unit can(Φ210mm x H200mm)
Unit can(Φ210mm x H200mm)
Unit can(Φ210mm x H200mm)

Unit can(Φ210mm x H200mm)

Unit can(Φ210mm x H200mm)

Unit can(Φ210mm x H200mm)
Unit can(Φ210mm x H200mm)

Base case
Fuel debris
composition(UO2 , ZrO2 )MOX

filling factor

filling factor

Gd content

Gd content

B content

B content

Moisture content©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.34

2.2.1 Passive neutrons - Table listing the analysis cases (2/3)
Composition inside the container *1
Case No.

Type of fueldebris2-11

Molten debris

UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)UO2 : 15vol%(50vol%)2-14

ZrO2 : 15vol%(50vol%)UO2 : 15vol%(50vol%)2-15

ZrO2 : 15vol%(50vol%)UO2 : 15vol%(50vol%)2-16

ZrO2 : 15vol%(50vol%)UO2 : 15vol%(50vol%)2-17

ZrO2 : 15vol%(50vol%)UO2 : 15vol%(50vol%)2-18

ZrO2 : 15vol%(50vol%)UO2 : 0.48vol%[3.7kg]

2-19

2-20

Composition

ZrO2 : 15vol%(50vol%)2-13

SUS : 10vol%(95.4vol%)Uranium-rich

Cooling

rate

period

23.0GWd/t

Standard

20 yearsUniform

1.3GWd/t

Standard

20 yearsUniform

51GWd/tStandard

20 yearsUniform

23.0GWd/t

Standard

30 yearsUniform

23.0GWd/t

Standard

40 yearsUniform

23.0GWd/t

Standard

20 years

23.0GWd/t

Standard

20 years

23.0GWd/t

Standard

20 years

Outside the filling factor

UO2 : 15vol%(50vol%)2-12

FP emission

Burn-upWithin the filling factor

UO2 : 30vol%

Uneven distribution

Container

Changed parameter

Total (filling factor)
H2O (Moisture content):
30vol%

70vol%
Empty : 0vol%

30vol%

30vol%

30vol%

30vol%

30vol%

30vol%

30vol%

10.48vol%

30vol%H2O (Moisture content): 1wt%
Empty: Remainder
H2O (Moisture content): 1wt%
Empty: Remainder
H2O (Moisture content): 1wt%
Empty: Remainder
H2O (Moisture content): 1wt%
Empty: Remainder
H2O (Moisture content): 1wt%
Empty: Remainder
H2O (Moisture content): 1wt%
Empty: Remainder
H2O (Moisture content): 1wt%
Empty: Remainder

H2O (Moisture content): 1wt%
Empty: Remainder

H2O (Moisture content): 1wt%
Empty: Remainder

Unit can(Φ210mm x H200mm)
Unit can(Φ210mm x H200mm)
Unit can(Φ210mm x H200mm)
Unit can(Φ210mm x H200mm)
Unit can(Φ210mm x H200mm)

Horizontal direction Unit can(Center)(Φ210mm x H200mm)

Horizontal direction Unit can(Exterior surface) (Φ210mm x H200mm)

Moisture content

Burn-upBurn-up

Cooling period

Cooling period

Uneven distribution

Uneven distribution

Canister
Uniform(Φ220mm x H840mm,

Container

Thickness 10mm)
Waste storage container
23.0GWd/t

Standard

20 yearsUniform(Internal container

Container□ Φ500mm x H300mm)

23.0GWd/t

Standard

20 yearsUniform

Unit can(Φ210mm x H200mm)

Fuel debris
composition (type)©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.35

2.2.1 Passive neutrons - Table listing the analysis cases (3/3)
Composition inside the container *1
Case No.

Type of fuelBurn-upWithin the filling factor
Outside the filling factor

debrisComposition

Metallic debris

ZrO2 : 0.075vol%(0.25vol%)rate

Cooling period

Uneven distribution

Container

Changed parameter

Total (Filling factor)

UO2 : 0.075vol%(0.25vol%)2-21

FP emission

30vol%H2O (Moisture content): 1wt%
Empty: Remainder

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 years

23.0GWd/t

Standard

20 years

23.0GWd/t

Standard

20 years

23.0GWd/t

Standard

20 years

Unit can(Φ210mm x H200mm)

Fuel debris composition (type)

SUS : 29.85vol%(99.5vol%)UO2 : 15vol%(25vol%)2-22

ZrO2 : 15vol%(25vol%)

60vol%SUS : 30vol%(50vol%)H2O (Moisture content): 1wt%
Empty: Remainder

Unit can(Φ210mm x H200mm)

Fuel debris composition (SUS)UO2 : 1.05vol%(3.5vol%)ZrO2 : 1.05vol%(3.5vol%)
2-23

MCCI

SUS : 7.2vol%(24vol%)

30vol%

Concrete :H2O (Moisture content): 1wt%
Empty: Remainder

Unit can(Φ210mm x H200mm)

Fuel debris composition (type)20.7vol%(69vol%)UO2 : 15vol%(33.3vol%)2-24

ZrO2 : 15vol%(33.3vol%)45vol%

Concrete : 15vol%(33.3vol%)UO2 : 15vol%(25vol%)2-25

ZrO2 : 15vol%(25vol%)

60vol%

Concrete : 30vol%(50vol%)UO2 : 0.075vol%(0.25vol%)2-26

Metallic debris

ZrO2 : 0.075vol%(0.25vol%)

30vol%SUS : 29.85vol%(99.5vol%)UO2 : 0.075vol%(0.25vol%)2-27

ZrO2 : 0.075vol%(0.25vol%)

30vol%H2O (Moisture content): 1wt%
Empty: Remainder
H2O (Moisture content) 1wt%
Empty: Remainder

H2O (Moisture content): 1wt%
Empty: Remainder

H2O (Moisture content): 1wt%
Empty: Remainder

Horizontal direction(Center)

Unit can(Φ210mm x H200mm)

Unit can(Φ210mm x H200mm)

Unit can(Φ210mm x H200mm)

Horizontal direction

Unit can(Exterior surface)(Φ210mm x H200mm)

Horizontal direction

Unit canFuel debris composition(Concrete)

Fuel debris composition(Concrete)

Fuel debris composition (Type)
Uneven distribution

Fuel debris composition (Type)
Uneven distribution

SUS : 29.85vol%(99.5vol%)UO2 : 1.05vol%(3.5vol%)ZrO2 : 1.05vol%(3.5vol%)
2-28

MCCI

SUS : 7.2vol%(24vol%)

30vol%H2O (Moisture content): 1wt%
Empty: Remainder(Center)(Φ210mm x H200mm)

Fuel debris composition (Type)
Uneven distribution

Concrete :20.7vol%(69vol%)UO2 : 1.05vol%(3.5vol%)ZrO2 : 1.05vol%(3.5vol%)2-29

SUS : 7.2vol%(24vol%)

30vol%H2O (Moisture content): 1wt%
Empty: Remainder

Horizontal direction

Unit can(Exterior surface)(Φ210mm x H200mm)

Fuel debris composition (Type)
Uneven distribution

Concrete :20.7vol%(69vol%)*1: Percentage inside the container. Percentage, when 100% is considered to be within the filling rate, is mentioned inside parentheses©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.36

2.2.1 Passive neutrons - Analysis cases•

Sensitivity analysis related to the following analysis conditions was conducted, and the trend with respect to each
influencing factor has been consolidated.
Case

UO2

ZrO2

SUS

Concrete

Total(Filling rate)

Moisturecontent

Burn-up(GWd/t)Cooling
period(Years)Remarks
U content consideredas the analysis
condition

2-1

Base

15vol%

15vol%

0vol%-

30vol%

1wt%

23.0

20

8.79kg

2-3

MOX

MOX
30vol%

0vol%

0vol%

-

30vol%

1wt%

0.0

20

17.0kg2-4Filling rate (low)

5vol%

5vol%

0vol%

-

10vol%

1wt%

23.0

20

2.93kg2-5Filling rate (high)25vol%

25vol%

0vol%-

50vol%

1wt%

23.0

20

14.6kg

2-10

Moisture content(low)

15vol%

15vol%

0vol%-

30vol%

0.1wt%

23.0

20

8.79kg

2-11

Moisture content(high)

15vol%

15vol%

0vol%-

30vol%

70vol%

23.0

20

8.79kg2-12

Burn-up (low)15vol%

15vol%

0vol%-

30vol%

1wt%

1.3

20

8.98kg

2-13

Burn-up (high)15vol%

15vol%

0vol%-

30vol%

1wt%

51.0

20

8.58kg

2-14

Cooling period

15vol%

15vol%

0vol%-

30vol%

1wt%

23.0

30

8.79kg

2-15

Cooling period

15vol%

15vol%

0vol%-

30vol%

1wt%

23.0

40

8.79kg

2-21

Metallic debris

0.075vol％0.075vol%29.85vol%-

30vol%

1wt%

23.0

20

0.04kg

2-23

MCCI debris

1.05vol%

1.05vol%

7.2vol%20.7vol%

30vol%

1wt%

23.0

20

0.615kg

2-24

MCCI debris

15vol%

15vol%

0vol%

15vol%

45vol%

1wt%

23.0

20

8.79kg2-25

MCCI debris

15vol%

15vol%

0vol%

30vol%

60vol%

1wt%

23.0

20

8.79kg©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2-1 (Molten debris) @1cm

2-1 (Molten debris) @15cm

2-4 (Molten debris) @1cm

2-4 (Molten debris) @15cm

2-5 (Molten debris) @1cm

2-5 (Molten debris) @15cm

2-21 (Metallic debris) @1cm

2-21 (Metallic debris) @15cm

2-23 (MCCI) @1cm

2-23 (MCCI) @15cm

Neutron spectrum (n/cm2/s)

Neutron spectrum (n/cm2/s)

2.2.1 Passive neutrons - Analysis results [Fuel debris type]

Energy (MeV)

Energy (MeV)

Neutron spectrum at 1cm of the container surface

Neutron spectrum at 15cm of the container surface
Case
No.

Type offuel debris

2-1→ As compared to molten debris, the shape

No.37

2-4

Moltendebris

UO2

ZrO2

SUS

Concrete

Filling rate

15vol%(50vol%)15vol%(50vol%)

0vol%

0vol%

30vol%

5vol%(50vol%)5vol%(50vol%)

0vol%

0vol%

10vol%

25vol%(50vol%)25vol%(50vol%)

0vol%

0vol%

50vol%of the neutron spectrum of MCCI and

2-5

metallic debris indicated an increase in

2-21

Metallic
debris

0.075vol%(0.25vol%)0.075vol%(0.25vol%)29.85vol%(99.5vol%)

0vol%

30vol%

2-23

MCCI

1.05vol%(3.5vol%)

1.05vol%(3.5vol%)

7.2vol%(24vol%)20.7vol%(69vol%)30vol%

thermal neutron flux.

©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.38

2.2.1 Passive neutrons - Analysis results [Burn-up]
1E+03

1E+03

2-1燃焼度(ベース)@1cm
2-1
Burn-up (base) @1cm

1E+02

1E+01

2-12燃焼度(低)@1cm
2-12
Burn-up (low) @1cm

1E+01

1E+00

2-13
Burn-up (high) @1cm
2-13燃焼度(高)@1cm

2-1燃焼度(ベース)@15cm
2-1
Burn-up (base) @15cm
2-12
Burn-up (low) @15cm
2-12燃焼度(低)@15cm

1E+00

2-13
Burn-up (high) @15cm
2-13燃焼度(高)@15cm

1E-01

1E-01

Neutron spectrum (n/cm2/s)

Neutron spectrum (n/cm2/s)

1E+02

1E-02
1E-03
1E-04
1E-05
1E-06
1E-07
1E-08

1E-09
1E-10

1E-02
1E-03
1E-04
1E-05
1E-06
1E-07
1E-08
1E-09
1E-10
1E-11

1E-11
1E-11

1E-09

1E-07

1E-05

1E-03

1E-01

1E+01

1E-12
1E-11

1E-09

1E-07

1E-05

1E-03

1E-01

1E+01

Energy (MeV)

Energy (MeV)

Neutron spectrum at 1cm of the container surface

Neutron spectrum at 15cm of the container surface→ The intensity of neutron beams

Case
No.

changed depending on burn-up, but

2-1

the shape of the neutron spectrum did

2-12

not change.

2-13

Type offueldebris

Moltendebris

UO2

ZrO2

SUS

Concrete

Filling rate

Burn-up(GWd/t)15vol%(50vol%)15vol%(50vol%)

0vol%

0vol%

30vol%

23.0

15vol%(50vol%)15vol%(50vol%)

0vol%

0vol%

30vol%

1.3

15vol%(50vol%)15vol%(50vol%)

0vol%

0vol%

30vol%

51.0©International Research Institute for Nuclear Decommissioning2.2 Analysis results
1E+03

1E+03

1E+02

1E+02

1E+01

1E+01
1E+00

1E+00
1E-01
1E-02
1E-03
1E-04
1E-05

2-11
Moisture content (70vol)2-11含水率（70vol）

1E-06

2-1
Moisture content (1wt)
2-1含水率（1wt）

1E-07

2-10含水率（0.1wt）
2-10
Moisture content (0.1wt)

1E-08
1E-11

1E-09

1E-07

1E-05

1E-03

1E-01

1E+01

Neutron spectrum (n/cm2/s)

Neutron spectrum (n/cm2/s)

2.2.1 Passive neutrons - Analysis results [Moisture content] No.39

1E-01
1E-02
1E-03
1E-04
1E-05

2-11
Moisture content (70vol)2-11含水率（70vol）1E-06

2-12-1含水率（1wt）
Moisture content (1wt)

1E-07

2-10
Moisture content (0.1wt)
2-10含水率（0.1wt）

1E-08
1E-09
1E-11

1E-09

1E-07

Energy (MeV)

neutron flux increased.

1E-01

1E+01

Neutron spectrum at 15cm of the container surface

Case
No.

Type offuel debris

2-1

neutron flux decreased and thermal

1E-03

Energy (MeV)

Neutron spectrum at 1cm of the container surface→ As moisture content increased, fast

1E-05

2-10

2-11

Moltendebris

UO2

ZrO2

SUS

Concrete

Filling rate

Moisturecontent

15vol%(50vol%)15vol%(50vol%)

0vol%

0vol%

30vol%

1wt%

15vol%(50vol%)15vol%(50vol%)

0vol%

0vol%

30vol%

0.1wt%

15vol%(50vol%)15vol%(50vol%)

0vol%

0vol%

30vol%

70vol%©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.40

2.2.1 Passive neutrons - Analysis results [MCCI]
1E+03

1E+03

2-25MCCI conc. 30vol%

1E+02

2-24MCCI conc. 15vol%

1E+01

2-24MCCI conc. 15vol%1E+01

2-1溶融デブリconc.
0vol%
2-1 Molten debris 0vol%2-1溶融デブリconc.
0vol%
2-1 Molten debris 0vol%

1E+00

Neutron spectrum (n/cm2/s)

1E+00

Neutron spectrum (n/cm2/s)

2-25MCCI conc. 30vol%

1E+02

1E-01
1E-02
1E-03
1E-04
1E-05
1E-06
1E-07
1E-08

1E-01
1E-02
1E-03
1E-04
1E-05
1E-06
1E-07
1E-08

1E-09
1E-11

1E-09

1E-07

1E-05

1E-03

1E-01

1E+01

1E-09
1E-11

1E-09

1E-07

Energy (MeV)

Neutron spectrum at 1cm of the container surface→ As concrete increased, thermal
neutron flux increased.

1E-05

Energy (MeV)

1E-03

1E-01

1E+01

Neutron spectrum at 15cm of the container surface

Case
No.

Type offuel debris

UO2

ZrO2

SUS

Concrete

Filling rate

2-1

Moltendebris15vol%(50vol%)15vol%(50vol%)

0vol%

0vol%

30vol%

2-24

MCCI

15vol%(33vol%)15vol%(33vol%)0vol%

15vol%(33vol%)45vol%

2-25

MCCI

15vol%(25vol%)15vol%(25vol%)0vol%

30vol%(50vol%)60vol%©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.41

2.2.1 Passive neutrons - Analysis results [Cooling period]
1E+03

1E+03
1E+02

1E+02

2-14冷却期間30年
2-14
Cooling period 30 years
2-1
Cooling period 20 years
2-1冷却期間20年

1E+01

1E+01
1E+00

Neutron spectrum (n/cm2/s)

1E+00

Neutron spectrum (n/cm2/s)

2-15冷却期間40年
2-15 Cooling period 40 years
2-14冷却期間30年
2-14 Cooling period 30 years
2-1冷却期間20年
2-1 Cooling period 20 years

2-15冷却期間40年
2-15
Cooling period 40 years

1E-01
1E-02
1E-03
1E-04

1E-05
1E-06
1E-07

1E-01
1E-02
1E-03
1E-04
1E-05
1E-06
1E-07

1E-08

1E-08
1E-09
1E-11

1E-09

1E-07

1E-05

1E-03

1E-01

1E+01

1E-09
1E-11

1E-09

1E-07

1E-05

1E-03

1E-01

1E+01

Energy (MeV)

Energy (MeV)

Neutron spectrum at 1cm of the container surface

Neutron spectrum at 15cm of the container surface

Based on the dose rate obtained from the results of thisanalysis and the results of passive gamma rays analysis,
the study process mentioned in No. 32 ⓪ to ③ is planned
to be implemented.

Case
No.

Type of fueldebris2-1→ The intensity of neutron beams changed depending on
cooling period, but the shape of the neutron spectrum did
not change.

2-14

2-15

Moltendebris

UO2

ZrO2

SUS

Concrete

Filling rate

Cooling
period[Years]15vol%(50vol%)15vol%(50vol%)

0vol%

0vol%

30vol%

20

15vol%(50vol%)15vol%(50vol%)

0vol%

0vol%

30vol%

30

15vol%(50vol%)15vol%(50vol%)

0vol%

0vol%

30vol%

40→ Fast neutrons and thermal neutrons (integrated value) flux was evaluated based on the above results.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.42

2.2.1 Passive neutrons - Analysis results [U Mass]
Fast neutrons (0.5MeV to 5MeV)
Case
No.

Type of fueldebrisU Mass(kg)

1cm(n/cm2/s)

15cm (n/cm2/s)

Case
No.

U Mass(kg)

1cm(n/cm2/s)

15cm (n/cm2/s)

2-1

Molten debris

8.786

1.19E+02

3.83E+01

2-1

8.786

3.20E-04

1.03E-04

2-4

Molten debris

2.929

4.21E+01

1.35E+01

2-4

2.929

5.37E-06

1.66E-06

2-5

Molten debris

14.643

1.85E+02

5.95E+01

2-5

14.643

4.28E-03

1.38E-03

2-21

Metallic debris

0.044

3.00Eー02

6.37Eー02

2-21

0.044

3.94E-06

9.53E-07

2-23

MCCI

0.615

8.57E+00

2.75E+00

2-23

0.615

4.28E-05

1.37E-05

2-3

MOX

17.0

4.36E+01

1.40E+01

2-3

17.0

4.78E-05

1.54E-05

1.E-02

Molten debris @1cm
溶融デブリ＠1cm
Molten debris @15cm
溶融デブリ＠15cm

Base case (2-1)

MCCI@1cm
MCCI@1cm

1.5E+02y = 12.915x

MCC@15cm
MCCI@15cm
Metallic debris @1cm
金属デブリ@1cm
Metallic debris @15cm
金属デブリ@15cm
MOX@1cm
MOX@1cm

1.0E+02

MOX@15cm
MOX@15cm

MOX

y = 4.1565x
5.0E+01

Neutron beam (n/cm2/s)

2.0E+02

Neutron beam (n/cm2/s)

Thermal neutrons (0.4eV or less)Base case (2-1)

1.E-03

1.E-04
Moltendebris @1cm
溶融デブリ＠1cm
Moltendebris @15cm
溶融デブリ＠15cm

1.E-05MCCI@1cm
MCCI@1cm
MCC@15cm
MCCI@15cm
金属デブリ@1cm
Metallic
debris @1cm
金属デブリ@15cm
Metallic
debris @15cm

1.E-06

MOX@1cm
MOX@1cm

Molten debris

0.0E+00
0

2

4

6

8

10

12

14

MOX@15cm
MOX@15cm

16

18

1.E-07
0

2

4

6

8

10

12

14

16

U Mass (kg)U Mass (kg)Fast neutrons (Integrated value of 0.5MeV to 5MeV)

Thermal neutrons (Integrated value of 0.4eV or less)18→ Fast neutron flux was almost directly proportional to the U mass in the case of molten debris, MCCI and metallic debris.
As MOX was almost unburnt, less number of neutrons were generated. Hence it is not considered to be directly
proportional to U mass.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.43

2.2.1 Passive neutrons - Analysis results [Burn-up]
Fast neutrons (0.5MeV to 5MeV)

Thermal neutrons (0.4eV or less)

Case
No.

Type of fueldebrisBurn-up(GWd/t)1cm(n/cm2/s)

15cm (n/cm2/s)

Case
No.

Burn-up(GWd/t)1cm(n/cm2/s)

15cm (n/cm2/s)

2-12Molten debris

1.3

4.38E-02

1.41E-02

2-12

1.3

9.69E-08

3.09E-08

2-1

Molten debris

23.0

1.19E+02

3.83E+01

2-1

23.0

3.20E-04

1.03E-04

2-13

Molten debris

51.0

6.63E+02

2.13E+02

2-13

51.0

2.23E-03

7.29E-04

1.E+03

1.E-02

1.E+02

Moltendebris @1cm
溶融デブリ@1cm

Neutron flux (n/cm2/s)

Neutron flux (n/cm2/s)

溶融デブリ@1cm
Moltendebris @1cm
Moltendebris @15cm
溶融デブリ@15cm

1.E+01

Base case (2-1)
1.E+00

1.E-01

1.E-03

溶融デブリ@15cm
Moltendebris @15cm

1.E-04
1.E-05

Base case (2-1)
1.E-06
1.E-07

1.E-02

1.E-081

10100

Burn-up (GWd/t)

Fast neutrons (Integrated value of 0.5MeV to 5MeV)

1

10

Burn-up (GWd/t)

Thermal neutrons (Integrated value of 0.4eV or less)→ The neutron flux leaking from the container largely changed by orders of magnitude depending on the burn-up.
©International Research Institute for Nuclear Decommissioning100

2.2 Analysis results

2.2.1 Passive neutrons - Analysis results [Moisture content] No.44
Fast neutrons (0.5MeV to 5MeV)
Case
No.

Type of fueldebrisThermal neutrons (0.4eV or less)

Moisture content(vol%)

1cm(n/cm2/s)

15cm (n/cm2/s)

Case
No.

Moisture content(vol%)

1cm(n/cm2/s)

15cm (n/cm2/s)

2-10

Molten debris

0.24

1.22E+02

3.92E+01

2-10

0.24

2.37E-06

8.68E-07

2-1

Molten debris

2.44

1.19E+02

3.83E+01

2-1

2.44

3.20E-04

1.03E-04

2-11

Molten debris

70.0

3.01E+01

9.64E+00

2-11

70.0

1.22E+01

3.93E+00

1.E+03

1.E+02
Moltendebris @15cm
溶融デブリ@15cm

Neutron flux (n/cm2/s)

Neutron flux (n/cm2/s)

Moltendebris @1cm
溶融デブリ@1cm

1.E+02y = 123.85e-0.02x

Base case (2-1)

1.E+01y = 39.825e-0.02x

Moltendebris @1cm
溶融デブリ@1cm

1.E+01

Moltendebris @15cm
溶融デブリ@15cm

1.E+00
1.E-01
1.E-02
1.E-03

1.E-04
1.E-05

Base case (2-1)

1.E-06
1.E+00

1.E-07
0

10

20

30

40

50

60

70

80

0.1

1

10

100

Moisture content (vol%)Moisture content (vol%)Fast neutrons (Integrated value of 0.5MeV to 5MeV)

Thermal neutrons (Integrated value of 0.4eV or less)→ Along with the increase in moisture content, the fast neutron flux decreased,
and the thermal neutron flux increased largely by orders of magnitude.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.45

2.2.1 Passive neutrons - Analysis results [MCCI]Fast neutrons (0.5MeV to 5MeV)

Thermal neutrons (0.4eV or less)

Case
No.

Type of fueldebrisPercentage ofconcrete(vol%)

1cm(n/cm2/s)

15cm (n/cm2/s)

Case
No.

Percentage ofconcrete(vol%)

1cm(n/cm2/s)

15cm (n/cm2/s)

2-1

Molten debris

0

1.19E+02

3.83E+01

2-1

0

3.20E-04

1.03E-04

2-24

MCCI

15

1.15E+02

3.72E+01

2-24

15

1.98E-03

6.44E-04

2-25

MCCI

30

1.12E+02

3.59E+01

2-25

30

8.15E-03

2.65E-03

1.E-02

＠1㎝
＠15㎝

Neutron flux (n/cm2/s)

Neutron flux (n/cm2/s)

1.E+03y = 119.19e-0.002x

1.E+02y = 38.326e-0.002x
1.E+01

Base case (2-1)

＠1㎝
＠15㎝

1.E-03

Base case (2-1)
1.E-04

1.E+00
0

10

20

30

40

50

60

70

80

Percentage of concrete (vol%)Fast neutrons (Integrated value of 0.5MeV to 5MeV)

0

5

10

15

20

25

30

35

Percentage of concrete (vol%)Thermal neutrons (Integrated value of 0.4eV or less)→ Along with the increase in percentage of concrete, even though there was just a slightdecrease in the fast neutron flux, the thermal neutron flux increased by orders of magnitude.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.46

2.2.1 Passive neutrons - Analysis results [Cooling period]
Fast neutrons (0.5MeV to 5MeV)

Thermal neutrons (0.4eV or less)

Case
No.

Type of fueldebrisCooling period(Years)1cm(n/cm2/s)

15cm (n/cm2/s)

Case
No.

Cooling period(Years)1cm(n/cm2/s)

15cm (n/cm2/s)

2-1

Molten debris

20

1.19E+02

3.83E+01

2-1

20

3.20E-04

1.03E-04

2-14

Molten debris

30

8.26E+01

2.66E+01

2-1430

2.21E-04

7.15E-05

2-15

Molten debris

40

5.77E+01

1.86E+01

2-15

40

1.54E-04

4.98E-05

1.E+03

1.E-03

Moltendebris @1cm
溶融デブリ＠1cm

溶融デブリ＠15cm
Moltendebris @15cm

溶融デブリ＠15cm
Moltendebris @15cm

y = 245.7(1/2)^(x/19.1)
1.E+02

Base case (2-1)

y=79.0(1/2)^(x/19.1)

Neutron flux (n/cm2/s)

Neutron flux (n/cm2/s)

Moltendebris @1cm
溶融デブリ＠1cmy=0.0007(1/2)^(x/19.0)
1.E-04

Base case (2-1)

y=0.0002(1/2)^(x/19.0)

1.E-05

1.E+01
10

20

30

40

Cooling period (Years)
Fast neutrons (Integrated value of 0.5MeV to 5MeV)

50

10

20

30

40

50

Cooling period (Years)
Thermal neutrons (Integrated value of 0.4eV or less)→ The half-life of the neutron flux was 19.1 years which is slightly longer than the αdecay half-life of Cm-244 which is the main neutron radiation source.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.47

2.2.1 Passive neutrons - Analysis results
Study of a representative detector, measurement system (for identifying
issues) (1/2)

Approx. 5,000 barns

・representative detector: Helium 3 proportional counter・ Size and shape of detector : 1 inch cylindrical detector, length 40 inches
・ Moderator: High density polyethylene
・ Distance up to the detector : 150mm(From the outer surface of the unit canto the outer surface of the detector)

Approx.
barns

3He (n,p)<Analysis model>

0.025eV 0.1eV<Detector response>

195mm

Without moderator
With moderator
0.025eV

0.1eV

2MeV

Energy (MeV)Φ206mm

Φ210mm2MeV

He-3 cross sectional area (From JENDL4.0)

Neutron flux (n/cm2/s)

He-3 detector

1016mm(40in)Polyethylene
0.95g/cm3
Radiation source
UO2+ZrO2

200mm

Approx. 1 barns

3,000

100mm

Neutron flux (Radiation source: Trial calculation case, detector distance 15cm)
Φ25.4mm

50mm→ By installing polyethylene, measurement efficiency of He-3 detector increased several 1000 times.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.48

2.2.1 Passive neutrons - Analysis results
Study of a representative detector, measurement system (for identifying issues) (2/2)

Relation between U content and incident beams

All neutron beams (n/cm2/sec)

30vol%
No.2-20(Uranium-rich)

7.5vol%No.2-2

15vol%
No.2-1(Base case)All neutron beams*Burn-up constant at 23GWd/t

0.075vol%No.2-21(Metallic debris)

Weight of U (g)<Helium 3 proportional counter Incident
beam>
・As a result of the analysis in which only theamount of Uranium was changed, it was
found that the amount of Uranium and all
incident beams have a positive correlation.
・Even under conditions wherein polyethylene ispresent, Uranium content and incident beams
were proportionate.
→ However, the inclination is believed tolargely change with variations in otherparameters such as burn-up of fuel debris,
etc. (Refer to No. 43)

・The impact of the self-shielding effect causedby changes in the density (filling rate) of fueldebris did not appear significantly in any of theincident beams.

©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.1 Passive neutrons – Other sensitivity analyses (1/8)

No.49

The sensitivity of the neutron count rate was studied with respect to the following influencing factors:➢ Neutron absorption material➢ Configuration of container➢ Fuel debris composition (SUS)
Case
No.

UO2

ZrO2

SUS

Total
(Filling rate)

Other conditions

Remarks

2-1

15vol%

15vol%

0vol%

30vol%

-

Base

2-2

7.5vol%22.5vol%

0vol%

30vol%-

Molten debris

2-6

0vol%

15vol%

0vol%

30vol%

Gd2O3: Moderate

2-7

15vol%

15vol%

0vol%

30vol%

Gd2O3: High

2-8

15vol%

15vol%

0vol%

30vol%

B4C: Moderate

2-9

15vol%

15vol%

0vol%

30vol%

B4C: High

2-18

15vol%

15vol%

0vol%

30vol%

Canister (Φ220mm×H840mm)

Gd content

B content

Waste storage container

Shape ofcontainer

2-19

15vol%

15vol%

0vol%

30vol%

2-20

30vol%

0vol%

0vol%

30vol%-

Molten debris

2-21

0.075vol％0.075vol%29.85vol%

30vol%-

Metallic debris

2-22

15vol%

15vol%

30vol%

60vol%-

Metallic debris©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.1 Passive neutrons – Other sensitivity analyses (2/8)

No.50

Waste storage container (Inner container)（No. 2 - 19）➢

Impact due to difference in shape of container→ If the container is bigger than the unit can, the distance up to the detector increases.
Hence incident beams decrease.

Unit can

210mm
410mm
Vertical cross-section

Waste container(Inner container)

He-3
detector

Entire neutron flux (n/cm2/sec)

1016mm(40in)

15vol%×4
No.2-18 (Canister)
Other common conditions:

30vol%
No.2-20 (Uranium-rich)

15vol%
No.2-1 (Base case)All neutron beams

1016mm(40in)

7.5vol%No.2-2・ Burn-up 23GWd/t
・ FP emission rate:
Standard
・ Moisture content 1wt%
・ Cooling period: 20 years
・ Homogeneous model

0.075vol%No.2-21(Metallic debris)

500mm
800mm
Vertical cross-section

No.2-19(Waste container(Inner container))

Weight of U (g)He-3 detector©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.1 Passive neutrons – Other sensitivity analyses (3/8)

No.51➢ Impact depending on the amount of neutron absorption material (Gadolinium, Boron)
contained
UO2

ZrO2

SUS

Concrete

Filling rate

Neutronabsorption
material

Remarks

2-1

15vol%(50vol%)15vol%(50vol%)

0vol%

0vol%

30vol%

-

Base

2-6

15vol%(50vol%)15vol%(50vol%)

0vol%

0vol%

30vol%

Gd2O3: 3vol%

Gd content: Moderate

2-7

15vol%(50vol%)15vol%(50vol%)

0vol%

0vol%

30vol%

Gd2O3: 30vol%

Gd content: High

Case
No.

Type of fueldebris

Moltendebris2-8

15vol%(50vol%)15vol%(50vol%)

0vol%

0vol%

30vol%

B4C: 0.51vol%

B content: Moderate(Equivalent to actual conditions: Set
based on the core average of theamount of B mixed before theaccident)

2-9

15vol%(50vol%)15vol%(50vol%)

0vol%

0vol%

30vol%

B4C: 10vol%

B content: High

Other common conditions:・ Burn-up 23GWd/t
・ FP (fission product) emission
rate: Standard
・ Moisture content 1wt%
・ Homogeneous model・ Container: Unit can©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.1 Passive neutrons – Other sensitivity analyses (4/8)

No.52

Relation between the amount of neutron absorption material (Gadolinium, Boron) contained and the incident beams (1/3)[Without polyethylene]

He-3 detector

1016mm(40in)

195mm

200mm

Polyethylene
0.95g/cm3
Radiation source
UO2+ZrO2<Helium 3 proportional counter Incident beam>

Neutron spectrum (n/cm2/sec)<Analysis model>

Slowneutrons areabsorbed(Base case)

Gd2O3
B4C

Φ206mm
Φ210mm100mm
Φ25.4mm

Energy (MeV)
Neutron spectrum at the location of the detector

50mm→・ In cases with neutron absorption material, as slow neutrons (<1eV) got absorbed, they did not appear in the spectrum.
・ Since the cross-sectional area of Gd is large in the MeV region as well, the neutron flux between the fast and slow neutrons isgreater than the base case.
・ Since the absorption cross section of B in the KeV region is large, the neutron flux is lower than even the base case.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.1 Passive neutrons – Other sensitivity analyses (5/8)

No.53

He-3 detector

1016mm(40in)

195mm

200mm

Polyethylene
0.95g/cm3
Radiation source
UO2+ZrO2

Neutron spectrum (n/cm2/sec)

Relation between the amount of neutron absorption material (Gadolinium, Boron) contained
and the incident beams (2/3)<Helium 3 proportional counter<Analysis model>[With polyethylene]
Incident beam>(Base case)

Gd2O3
Gd2O3
B4C
B4C

Φ206mm
Φ210mmEnergy (MeV)
Neutron spectrum at the location of the detector

100mm
Φ25.4mm→ Impact of the neutron absorption material did not appear significantly in the spectrum.

50mm(The slow neutrons inside the unit can are absorbed by the neutron absorption material, but slow
neutrons on the unit can side of polyethylene do not reach up to the detector region due to thedeceleration/absorption effect of polyethylene regardless of the presence of the absorption
material, and are believed to not have any impact.)©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.1 Passive neutrons – Other sensitivity analyses (6/8)

No.54

Relation between the amount of neutron absorption material (Gadolinium, Boron) contained
and the incident beams (3/3)
No.2-1(Base)

No.2-1(Base)

No.2-7(Gd2O3: 30vol%)Other common conditions:・ Burn-up 23GWd/t
・ FP emission rate: Standard
・ Moisture content 1wt%
・ Homogeneous model・ Container: Unit can

Weight of Gd(kg)

No.2-8(B4C: 0.5vol%)No.2-9(B4C: 10vol%)Entire neutron flux (n/cm2/sec)

Entire neutron flux (n/cm2/sec)

No.2-6(Gd2O3: 3vol%)Weight of B
(kg)

→・Even if neutron absorption material was added, there was hardly any decrease in the entire neutron flux.
・ As the spectrum in the thermal neutron energy region that is mainly detected by the He-3 detector does not change
depending on the amount of Gd and B added, it is believed to have only a slight impact on the neutron count.

©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.1 Passive neutrons – Other sensitivity analyses (7/8)

No.55

Response spectrum for a representative detector➢ Correlation with Uranium content
Case

UO2

ZrO2

SUS

Concrete

Total(Filling rate)

U content setting

2-1

Base

15vol%

15vol%

0vol%-

30vol%

8.79kg2-2

Molten debris(Spot with lessU)

7.5vol%22.5vol%

0vol%-

30vol%

2.93kg

2-18

Canister

15vol%

15vol%

0vol%-

30vol%

35.1kg2-20

Molten debris(Uranium-rich)

30vol%

0vol%

0vol%

-

30vol%

17.6kg

2-21

Metallic debris

0.075vol％0.075vol%29.85vol%-

30vol%

0.04kg

2-22

Metallic debris(Lot of SUS)

15vol％

15vol%

30vol%-

60vol%

8.79kgOther common conditions:・ Burn-up 23GWd/t
・ FP emission rate: Standard
・ Moisture content 1wt%
・ Homogeneous model©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.1 Passive neutrons – Other sensitivity analyses (8/8)

No.56

Response spectrum for a representative detector
Correlation of U content and beams incident upon the Helium 3 proportional counter

Entire neutron flux (n/cm2/sec)

1016mm(40in)

Canister
15vol%×4
No.2-18 (Canister)

Unit can※ Burn-up constant at 23GWd/t

30vol%
No.2-20 (Uranium-rich)

7.5vol%No.2-20.075vol%No.2-21 (Metallic debris)

15vol%
No.2-1 (Base case)No.2-22 (30% addition of SUS)

Other common conditions:・ FP emission rate: Standard
・ Moisture content 1wt%
・ Homogeneous model

Entire neutron flux

Weight of U (g)→・Quantity of U and all incident beams had a positive correlation (when the burn-up was constant)・The impact of the self-shielding effect caused by changes in the density (filling rate) of fuel debris does not appear
significantly in all the incident beams.
・The impact is minor even when SUS is present (Case 2-22).
・In the case of canister (Φ220mm×H840mm), since neutrons emitted from the top and bottom ends of the container are
likely to leak outside the system, the neutron flux becomes lower than the direct line of proportion of the unit can (Case 218).
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.1 Passive neutrons – Sensitivity analysis

Uneven distribution (1/6)

No.57

Analysis by changing the method of filling fuel debris into the container■Tally settings of the uneven distribution cases#The tally was divided into 6 parts for the horizontal direction (outer surface) cases only.
Tally(Region divided into 6 parts)

Unit canFuel debris

(2)

(3)

Unit can
３０°

A
(4)

(1)

A’

9.75cm9.75cm10.3cm

(6)

(5)

Tally
(No vertical division)
Top view

Cross sectional view A-A’

The location and dimensions for installing the tally for the unit can are the same as described in slide No. 56.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.1 Passive neutrons – Sensitivity analysis

Uneven distribution (2/6)

No.58

Molten debris Uneven distribution case spectrum
1E+03

1E+01

Neutron spectrum (n/cm 2/s)

1E+00

1E+03

#2-1_ベースケース#2-1_Base case#2-16_tally_all#2-17_tally_all#2-17_tally_c1

1E+02
1E+01
1E+00

Neutron spectrum (n/cm 2/s)

1E+02

1E-01
1E-02
1E-03
1E-04
1E-05
1E-06
1E-07
1E-08
1E-09
1E-11

1E-09

1E-07

1E-05

1E-03

1E-01

1E+01

Energy (MeV)

Neutron spectrum at 1cm of the container surface#2-1_ベースケース#2-1_Basecase#2-16_tally_all#2-17_tally_all#2-17_tally_c1

1E-01
1E-02
1E-03
1E-04
1E-05
1E-06
1E-07
1E-08
1E-09
1E-11

1E-09

1E-07

1E-05

1E-03

1E-01

1E+01

Energy (MeV)

Neutron spectrum at 15cm of the container surface→ When fuel debris was thickened, thermal neutron flux increased.
Even when fuel debris was unevenly distributed, the shape of the neutron
spectrum did not change depending on the location of the sensor.

©International Research Institute for Nuclear Decommissioning2.2 Analysis results
2.2.1 Passive neutrons – Sensitivity analysis

Fast neutrons (0.5MeV to 5MeV)
CaseNo.
2-1 (Molten base)
2-16 (Center)

tally

15cm
(n/cm2/s)
3.83E+01
3.27E+01
3.28E+01
3.74E+01
3.47E+01
3.05E+01
2.88E+01

Thermal neutrons (0.4eV or less)Case
No.
2-1 (Molten base)
2-16 (Center)2-17 (Outersurface)

tally
all
all
all
1
2
3
4

1cm
(n/cm2/s)
3.20E-04
8.41E-03
8.71E-03
1.15E-02
9.61E-03
7.47E-03
6.71E-03

15cm(n/cm2/s)
1.03E-04
2.84E-03
2.90E-03
3.38E-03
3.08E-03
2.67E-03
2.49E-03(n/cm2/s)
Neutron flux (n/cm2/sec)

Neutron flux (n/cm2/sec)(n/cm2/s)

2-17 (Outersurface)

all
all
all
1
2
3
4

1cm
(n/cm2/s)
1.19E+02
9.44E+01
9.65E+01
1.28E+02
1.07E+02
8.18E+01
7.39E+01

No.59

Uneven distribution (3/6) Molten debris(Molten (Center)base)(Outer surface)
Case No.

Fast neutrons (Integrated value of 0.5MeV to 5MeV)(Molten (Center)base)(Outer surface)
Case No.

Thermal neutrons (Integrated value of 0.4eV or less)→ ・ When fuel debris was unevenly distributed, thermal neutron flux was 1 order of magnitude higher than the base case, but
the impact of uneven distribution on the measurement position was not more than 2 times.
・ The farther away the fast neutron flux was from the surface of the container, the smaller was the fluctuation.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results
2.2.1 Passive neutrons – Sensitivity analysis Uneven distribution (4/6) MCCIFast neutrons (0.5MeV to 5MeV)
Case No.

tally

2-23 (MCCI base)2-28 (Center)

all
all
all
1
2
3
4

15cm
(n/cm2/s)
2.75E+00
2.51E+00
2.65E+00
4.42E+00
3.16E+00
1.83E+00
1.49E+00

Thermal neutrons (0.4eV or less)Case No.

tally

2-23 (MCCI base)2-28 (Center)

all
all
all
1
2
3
4

2-29
(Outer surface)

1cm(n/cm2/s)
4.28E-05
6.57E-05
5.04E-05
4.87E-05
5.14E-05
4.96E-05
4.87E-05

15cm(n/cm2/s)
1.37E-05
2.12E-05
1.62E-05
1.59E-05
1.58E-05
1.63E-05
1.63E-05

Neutron flux (n/cm2/sec)(n/cm2/s)(n/cm2/s)
Neutron flux (n/cm2/sec)

2-29(Outer surface)

1cm(n/cm2/s)
8.57E+00
7.19E+00
8.16E+00
1.93E+01
8.68E+00
4.37E+00
3.57E+00

No.60(MCCI (Center)base)(Outer surface)
Case No.

Fast neutrons (Integrated value of 0.5MeV to 5MeV)(MCCI (Center)base)(Outer surface)
Case No.

Thermal neutrons (Integrated value of 0.4eV or less)→ The farther away the fast neutron flux was from the surface of the container, the smaller was the fluctuation
caused by uneven distribution of fuel debris.
The fluctuation in thermal neutron flux was by and large small.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results
2.2.1 Passive neutrons – Sensitivity analysis

Uneven distribution (5/6)

No.61

Comparison of the beams from molten debris and MCCI→Distance from the surface ofthe container (cm)Proportion of neutron flux withrespect to molten debris (base)Distance from the surface ofthe container (cm)Thermal neutrons (Integrated value of 0.4eV or
less)

Proportion of neutron flux withrespect to MCCI (base)

MCCI

Proportion of neutron flux withrespect to MCCI (base)

Moltendebris

Proportion of neutron flux withrespect to molten debris (base)

Fast neutrons (Integrated value of 0.5MeV to
5MeV)Distance from the surface ofthe container (cm)Distance from the surface ofthe container (cm)・ The fluctuations due to uneven distribution were larger in fast neutron flux in the case of MCCI as against
molten debris.
・ The impact of uneven distribution on thermal neutron flux was small as compared to the case of molten debris.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results
2.2.1 Passive neutrons – Sensitivity analysis➢

Uneven distribution of metallic debris
Impact was verified by means of the flux after penetrating polyethylene.

→・ Since the distance up to the tally changes due to uneven distribution, there isvariation in the incident beams.
・ Since self-shielding effect does not appear in passive neutron measurement, there
is little variation due to uneven distribution of the fuel debris locations as comparedto passive gamma measurement.
<Uniform>

No.62

Uneven distribution (6/6)

Case
No.

Type offuel debris

UO2

ZrO2

SUS

Filling rate

Unevendistribution

2-21

Metallic
debris

0.075vol%0.075vol%29.85vol%

30vol%

Uniform

2-26
2-27

Metallic
debris

0.075vol%0.075vol%29.85vol%

30vol%

Unevendistribution

Relation between the weight of Uranium and the entire neutron flux (log log graph)(No. 2-21)
Polyethylene

Other common conditions:

Tally

Horizontal cross-section<Uneven distribution>
90°(No. 2-26)

Molten debris (Unevenly distributed at
the center of the container)
Fuel debris radius: Approx. 1.3cm
Distance from outer surface of unit can to
the Tally: 15cm

180°

0°
Tally
SUS
270°

(No. 2-27)

Polyethylene

90°

Entire neutron flux (n/cm2/sec)(Distance from outer surface of unit can to
the Tally: 15cm)

No.2-18

・ Burn-up 23GWd/t
・ FP emission rate: Standard
・ Moisture content 1wt%
・ Homogeneous model・ Container: Unit can

No.2-20

No.2-1
No.2-2

Uneven distribution:
Near
Unevendistribution: Center

Entire neutron flux

No.2-21 (Metallic debris: Uniform)

Unevendistribution: Far
Molten debris (Unevenly distributed on
the inner surface of the container)

180°

0°

Tally

Weight of U (g)270°
Horizontal cross-section©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.1 Passive neutrons – Additional sensitivity analyses (1/8)

No.63■ Cases with a combination of fluctuations in 2 influencing factors
MoltenFuel debris composition
Molten Fuel debris

Metallic

MCCI

MOX

Filling factor

Gd content

B content

Moisture content

Burn-up

Cooling period

Uneven distribution

Container

-

x

x

x

x

x

x

x

x

-

-

Increase/decrease in flux

Can be evaluated by means of the Can be evaluated by means of the Can be evaluated by means of the Can be evaluated by means of the Can be evaluated by means of the Can be evaluated by means of thedepending on the filling factor

linear sum of each parameter

linear sum of each parameter

linear sum of each parameter

linear sum of each parameter

linear sum of each parameter

linear sum of each parameter

x

x

x

x

-

x

x

x

△

△

Increase/decrease in flux

Can be evaluated by means of the Can be evaluated by means of the Can be evaluated by means of thedepending on the filling factor

linear sum of each parameter

linear sum of each parameter

linear sum of each parameter

x

x

x

x

△

△

composition
MOX

Filling factor

Can be evaluated by means of the Can be evaluated by means of thexlinear sum of each parameter

linear sum of each parameterxx

Can be evaluated by means of the Can be evaluated by means of the Can be evaluated by means of the Can be evaluated by means of the Can be evaluated by means of the Can be evaluated by means of thelinear sum of each parameter
Gd content

linear sum of each parameter

linear sum of each parameter△〇〇

B content〇〇

linear sum of each parameter

linear sum of each parameter

linear sum of each parameter

x

△

x

〇〇

x

△

x

〇〇

1 case(High Gd - H20 50vol%)1 case(High B - H2 O 50vol%)MoistureBurn-upx

・In order to compare the impact of Gd and B respectively, analysis was performed by setting the same
volume ratio for both.
・If the filling rate is 30vol% and water is 70vol%, there is no room to add Gd or B.
・As against this, the proposal of keeping the water at 50vol%, adding 10vol% each of Gd and B andanalyzing, was studied.
・For further comparison, the molten debris base case was analyzed by keeping water at 50vol%.

periodUnevendistribution
Container
Metallic
MCCI

In order to check the extent of impact of Gd, B and moisture content, cases withcombinations of maximum setting values of the factors were additionally analyzed.
(The base cases of the above-mentioned factors have the minimum value or a smallvalue)〇〇

1 case(High B)〇〇△

content

Cooling〇〇

1 case(High Gd)〇〇

1 case(70vol%)x

1 case(High Gd)
1 case(High B)1 case(70vol%)xx

Attenuates depending on half-life

Attenuates depending on half-life

△

△

x

x
-

8 +1 (Water 50vol%) = 9 Case

■ Cases wherein influencing factors in which the neutron count rate is the least were combined: 1 case
Filling rate

Gd content

B content

Moisturecontent

Burn-up

Unevendistribution

10vol%(UO2:ZrO2=5vol%:5vol%)High(Gd2O3 30vol%)High(B4C 10vol%)

H2O
50vol%

Low(1.3GWd/t)

At the center inthe horizontal
direction

Total
10 cases・If filling rate is 30vol%, moisture content is 70vol%, Gd is 30vol%, B is 10vol%, then 100vol% is exceeded.

•

The molten debris base case was used for other factors. ・Proposed parameters: The following conditions wherein water is 50vol% were studied.

If filling rate is 10vol%, moisture content is 50vol%, Gd is 30vol%, B is 10vol%, then the total is 100vol%.
In this instance, even the case mentioned above with a combination of 2 influencing factors was analyzed while
keeping water at 50vol%.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.1 Passive neutrons – Additional sensitivity analyses (2/8)

Case

2-1

Base

Concrete

Total(Filling
rate)

Moisturecontent

UO2

ZrO2

SUS

15
vol%

15
vol%

0
vol%

-

30
vol%

1wt%

Gd2O3

B4C

Burn-up(GWd/t)No.64
Remarks
Cooling
U content considered
UnevenPeriodas the analysis
distribution(Years)condition

23.0

20

Uniform

8.79kgStep 2 analysis cases
2-30

Molten/ HighGd/ Water
50vol%

15
vol%

15
vol%

0
vol%

-

30
vol%

50
vol%

10
vol%

0
vol%

23.0

20

Uniform

8.79kg2-31

Molten/ High B/ Water
50vol%

15
vol%

15
vol%

0
vol%

-

30
vol%

50
vol%

0
vol%

10
vol%

23.0

20

Uniform

8.79kg2-32

Molten/ Water
50vol%

15
vol%

15
vol%

0
vol%

-

30
vol%

50
vol%

0
vol%

0
vol%

23.0

20

Uniform

8.79kg

2-33

Metal/ High Gd

0.075
vol%

0.075
vol%

29.85
vol%

-

30
vol%

1wt%

30
vol%

0
vol%

23.0

20

Uniform

0.0439kg

2-34

Metal
/ High B

0.075
vol%

0.075
vol%

29.85
vol%

-

30
vol%

1wt%

0
vol%

10
vol%

23.0

20

Uniform

0.0439kg

2-35

Metal
/ Water 70%

0.075
vol%

0.075
vol%

29.85
vol%

-

30
vol%

70
vol%

0
vol%

0
vol%

23.0

20

Uniform

0.0439kg

2-36

MCCI
/ High Gd

1.05
vol%

1.05
vol%

7.2
vol%

20.7
vol%

30
vol%

1wt%

30
vol%

0
vol%

23.0

20

Uniform

0.615kg

2-37

MCCI/ High B

1.05
vol%

1.05
vol%

7.2
vol%

20.7
vol%

30
vol%

1wt%

0
vol%

10
vol%

23.0

20

Uniform

0.615kg

2-38

MCCI/ High watercontent

1.05
vol%

1.05
vol%

7.2
vol%

20.7
vol%

30
vol%

70
vol%

0
vol%

0
vol%

23.0

20

Uniform

0.615kg

2-39

Moltencompoundfactors

5
vol%

5
vol%

0
vol%

-

10
vol%

50
vol%

30
vol%

10
vol%

1.3

20

Horizontal
direction(Center)

2.93kg©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.1 Passive neutrons – Additional sensitivity analyses (3/8)

No.65

2-1 (Base case)2-32 (Water 50 vol%)2-30 (Gd 10 vol%, water 50 vol%)2-31 (B 10 vol%, water 50 vol%)(n/cm2/s)
Neutron spectrum (n/cm2/s)

2/s)(n/cm2/s)
Neutron spectrum(n/cm

Molten debris

2-1 (Base case)2-32 (Water 50 vol%)2-30 (Gd 10 vol%, water 50 vol%)2-31 (B 10 vol%, water 50 vol%)2-39 (Case with compound factors)

2-39 (Case with compound factors)

Energy (MeV)

Energy (MeV)

Neutron spectrum at 1cm of the container surface

Neutron spectrum at 15cm of the container surface→ Neutron flux decreased on the whole in the cases with compound factors.

©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.1 Passive neutrons – Additional sensitivity analyses (4/8) Molten debris

No.66

Fast neutrons (0.5MeV to 5MeV)

Thermal neutrons (0.4eV or less)

Case
No.

Case
No.

1cm
(n/cm2/s)

15cm(n/cm2/s)

1.19E+02

3.83E+01

8.18E+01

2-30 (Gd 10vol%)1cm(n/cm2/s)

15cm(n/cm2/s)

3.20E-04

1.03E-04

2.63E+01

2-1(Molten debris base case)2- 32 (Water 50vol%)3.73E+00

1.21E+00

7.43E+01

2.39E+01

2-30 (Gd 10vol%)1.08E-01

3.48E-02

2-31 (B 10vol%)7.46E+01

2.40E+01

2-31 (B 10vol%)3.22E-02

1.04E-02

2-39 (Compound factors)

6.12E-03

2.07E-03

2-39 (Compound factors)

6.15E-07

2.01E-07

Neutron flux (n/cm2/s)(n/cm2/s)

Neutron flux (n/cm2/s)(n/cm2/s)

2-1(Molten debris base case)2- 32 (Water 50vol%)(Molten debris base (Water 50vol%) (High Gd, water 50vol%) (High B, water (Compound factors)
case)
50vol%)

Case no.

Fast neutrons (Integrated value of 0.5MeV to 5MeV)(Molten debris base (Water 50vol%) (High Gd, water 50vol%) (High B, water
50vol%)
case)(Compound factors)

Case no.

Thermal neutrons (Integrated value of 0.4eV or less)→ ・ Fast neutron flux in the case with compound factors decreased up to 5 digits of magnitude from the base case.
・ In the high Gd/B content cases and cases in which there was no Gd/B content，when the water was 50vol%, thedifference in fast neutron flux was about 9%.
・ Thermal neutron beams fluctuated exponentially.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.1 Passive neutrons – Additional sensitivity analyses (5/8)

No.67

2-1 (Base case)2-21 (Metallic debris)
2-33 (Metallic debris, Gd 30vol%)(n/cm2/s)
Neutron spectrum (n/cm2/s)

Neutron spectrum (n/cm2/s)(n/cm2/s)

Metallic debris

2-1 (Base case)2-21 (Metallic debris)
2-33 (Metallic debris, Gd 30vol%)2-34 (Metallic debris, B 10vol%)2-34 (Metallic debris, B 10vol%)2-35 (Metallic debris, water 70 vol%)2-35 (Metallic debris, water 70 vol%)Energy (MeV)

Neutron spectrum at 1cm of the container surface

Energy (MeV)

Neutron spectrum at 15cm of the container surface→ The spectrum of Gd 30vol% sharply decreased in the case of 1eV or less.

©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.1 Passive neutrons – Additional sensitivity analyses (6/8) Metallic debris
Fast neutrons (0.5MeV to 5MeV)

Thermal neutrons (0.4eV or less)

Case
No.

Case
No.

1cm
(n/cm2/s)

15cm(n/cm2/s)

6.00E-01

1.93E-01

5.27E-01

1.70E-01

2-21(Metallic debris base case)2-33 (Gd 30vol%)2-34 (B 10vol%)5.84E-01

1.88E-01

2-35 (Water 70vol%)3.26E-01

1.05E-01

15cm(n/cm2/s)

4.17E-06

1.36E-06

7.69E-08

2.46E-08

2-34 (B 10vol%)1.83E-08

6.00E-09

2-35 (Water 70vol%)4.58E-02

1.49E-02

Neutron flux (n/cm2/s)(n/cm2/s)

Neutron flux (n/cm2/s)(n/cm2/s)

2-21(Metallic debris base case)2-33 (Gd 30vol%)1cm(n/cm2/s)

No.68(Water 70vol%)(Metallic debris base case)
Case No.

Fast neutrons (Integrated value of 0.5MeV to 5MeV)(Metallic debris base case)(Water 70vol%)Case No.

Thermal neutrons (Integrated value of 0.4eV or less)→ ・ Fast neutron flux decreased to about half of that in the case of metallic debris base case with 70vol% of water.
・ Thermal neutron flux fluctuated exponentially.

©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.1 Passive neutrons – Additional sensitivity analyses (7/8)

No.69(n/cm2/s)
Neutron spectrum (n/cm2/s)

MCCI

2-1 (Base case)2-1 (Base case)2-38 (MCCI, water 70vol%)2-38 (MCCI, water 70vol%)Energy (MeV)

Neutron spectrum at 1cm of the container surface

Energy (MeV)

Neutron spectrum at 15cm of the container surface→ The spectrum of Gd 30vol% sharply decreased in the case of 1eV or less.

©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.1 Passive neutrons – Additional sensitivity analysis (8/8)
Fast neutrons (0.5MeV to 5MeV)

Thermal neutrons (0.4eV or less)

1cm(n/cm2/s)

15cm(n/cm2/s)

8.57E+00

2.75E+00

7.56E+00

2.43E+00

2-23(MCCI base case)2-36 (Gd 30vol%)2-37 (B 10vol%)8.36E+00

2.69E+00

2-38 (Water 70vol%)4.80E+00

1.54E+00

Case No.

1cm(n/cm2/s)

15cm(n/cm2/s)

4.28E-05

1.37E-05

1.09E-06

3.50E-07

2-37 (B 10vol%)3.18E-07

1.07E-07

2-38 (Water 70vol%)9.54E-01

3.11E-01

Case No.

(n/cm2/s)
Neutron flux (n/cm2/s)(n/cm2/s)
Neutron flux (n/cm2/s)

2-23(MCCI base case)2-36 (Gd 30vol%)No.70(MCCI base case)(Water 70vol%)Case No.

Fast neutrons (Integrated value of 0.5MeV to 5MeV)(MCCI base case)(Water 70vol%)Case No.

Thermal neutrons (Integrated value of 0.4eV or less)→ ・ Fast neutron flux decreased to about half of that in the case of metallic debris basecase with 70vol% water.
・ Thermal neutron flux fluctuated exponentially.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.71

2.2.1 Passive neutrons - Detector response
Study procedure◆ Simulation evaluation [Fuel debris conditions (Molten debris Base case No. 2-1)]① Optimization of polyethylene thickness➁ Cd impact, Pb impact◆ Detector response (measurement time) evaluation➁Cd①Polyethylene thickness➁Pb

Unit canFuel debris

Fuel debris
200mm
150mm

A

A’
1016mm

Φ210mmPolyethylene Cd

Pb

T-cross Tally

Analysis model top view

Unit canCross sectional view A-A’
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.1 Passive neutrons - Detector response ① Polyethylene thickness

10

1E+00

Thermal neutron flux (n/cm2/s)

Neutron flux (n/cm 2/s)

1E+01

No.72

1E-01

1E-02

2-1_DR_PE_40mm
2-1_DR_PE_50mm
2-1_DR_PE_60mm
2-1_DR_PE_70mm

1E-03

1E-04
1E-11

1E-09

1E-07

1E-05

1E-03

1E-01

1E+01

Energy (MeV)

Neutron spectrum at 15cm of the container surface

8

6

4

2

0
0

10

20

30

40

50

60

70

Polyethylene thickness (mm)Dependence of thermal neutron flux onpolyethylene thickness→ ・ Thermal neutron flux was the maximum when polyethylene thickness was 50mm to 60mm.
[Hereinafter, 50mm thick polyethylene will be used.]©International Research Institute for Nuclear Decommissioning80

2.2 Analysis results

2.2.1 Passive neutrons – Detector response ➁ Cd impact, Pb impact

No.73

Thermal neutrons (0.4eV or less)➢ Cd impact: Polyethylene 50mm + Cd 1mm➢ Pb impact: Polyethylene 50mm + Pb 20mm

Case
(2-1 Base case)

15cm(n/cm2/s)

PE50mm

8.17

PE50mm+Cd1mm

7.33

PE50mm+Pb20mm

8.52

1E+01

1E+03
1E+02

1E+00

1E+00
1E-01
1E-02
1E-03

2-1_DR_PE_50mm

1E-04

_Cd_
2-1_DR_Gd_1mm_PE_50mm

1E-05

2-1_DR_Pb_20mm_PE_50mm

1E-06
1E-11

1E-09

1E-07

1E-05

1E-03

1E-01

1E+01

Energy (MeV)

Neutron spectrum at 1cm of the container surface

Neutron flux (n/cm2/s)

Neutron flux (n/cm2/s)

1E+01

1E-01

1E-02

2-1_DR_PE_50mm
2-1_DR_Gd_1mm_PE_50mm
_Cd_

1E-03

2-1_DR_Pb_20mm_PE_50mm
1E-04
1E-11

1E-09

1E-07

1E-05

1E-03

1E-01

1E+01

Energy (MeV)

Neutron spectrum at 15cm of the container surface→ ・With Cd1mm, thermal neutron flux reduced by orders of magnitude at 1cm of the container surface ⇒ It controls the incidenceof thermal neutron flux upon fuel debris.
・There was a small variation in the shape of the neutron spectrum at the 15cm position of the container surface due to Cd1mm,
Pb20mm.
[Hereinafter＋ 50mm thick polyethylene and Pb20mm+Cd1mm will be used.]©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.74

2.2.1 Passive neutrons - Detector response (Evaluation of measurement time)[Preconditions]➢ Fuel debris to be evaluated: Molten debris Case 2-1 (Base case), Metallic debris
Case

UO2
(vol%)

ZrO2(vol%)

SUS(vol%)

Concrete

Case 2-21 (Small amount of Uranium)

Total(Filling
rate)

Moisturecontent

Gd2O3

B4C

Burn-up(GWd/t)Cooling
period(Year)

Unevendistribution

U content

2-1

Moltendebris

15

15

0

-

30vol%

1wt%

-

-

23.0

20

Consistent

8.79kg

2-21

Metallic
debris

0.075

0.075

29.85

-

30vol%

1wt%

-

-

23.0

20

Consistent

0.04kg➢Placement of detector, etc.
Polyethylene

Lead

Detector

Detector

The detector was placed 15 cm from the container surface, 50mm polyethylene was placed

50mm thickpolyethyleneon the inside and outside of the detector, 20mm lead was placed on the inner side of

20mm thicklead

the inner polyethylene.

➢

Detector Note 1)① He detector: Φ25.5mm, effective length 500mm,
Sensitivity 63.7 cps/(n/cm2/s)

150mm➁ B-10 detector: Φ25.5mm, effective length 1,000mm,
Sensitivity 12.3 cps/(n/cm2/s)➢

Detector Maximum gamma dose rate Note 2)① He detector: approx. 0.1 Gy/hFuel debris

Unit canFueldebrisUnit can➁ B-10 detector: approx. 10 Gy/h
The dose rate at a distance of 15cm from the container surface was approx. 1Gy/h(according to the 2.2.2 Passive Gamma Rays - Analysis Results (2/7 and 3/7)), and as a result of installing 20mm lead, the dose rate at the location of thedetector in this system was approx. 0.1Gy/h.

Note 1) https://etd.canon/ja/product/category/proportional/npc.html

・ Issue: When 20mm lead is installed, there is no margin in the gamma dose rate ofNote 2) Neutron Detectors T.W.Crane, M. Baker (1997)

the He detector. Optimization of the detector shielding or its installation location needs to be studied in the future.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.1 Passive neutrons - Detector response (Evaluation of measurement time)

No.75

1. Molten debris (Case 2-1) [Base case] (1 detector)
1) Thermal neutron flux: From the position of the detector in the system having 50mm polyethylene +20mm Pb (15cm from the container surface): 8.52 (n/cm2/s)
2) Count rate and measurement time (Measurement time required for 1σ error 1% (10,000 counts)):① He detector: 5.43E+02 cps (= 8.52 x 63.7) ⇒ 18 seconds (= 10,000/5.43E+02)➁ B-10 detector : 1.05E+02 cps (= 8.52 x 12.3) ⇒ 95 seconds (= 10,000/1.05E+02)
2. Metallic debris (Case 2-21) [Small amount of Uranium]
1) Thermal neutron flux: 3.88E-02 (= 8.52 x 0.04kg/8.79kg) (n/cm2/s)(Assumed to be proportionate to U mass based on 2.2.2 Passive Neutrons - Analysis Results [UMass] on Slide No. 66)
2) Count rate and measurement time (measurement time required for 1σ error 1% (10,000 counts))a) When there was 1 detector① He detector: 2.47 cps ( = 3.88E-02 x 63.7) ⇒ approx. 4,000 seconds (= 10,000/2.47)➁ B-10 detector: 0.477 cps (= 3.88E-02 x 12.3) ⇒ approx. 21,000 seconds (= 10,000/0.477)b）When there were 60 detectors (Maximum number of detectors lined up at a distance of 15cm
from the container surface)① He detector: approx. 70 seconds➁ B-10 detector: approx. 350 seconds
・ Issue: In the case of fuel debris that contains a small amount of Uranium (Example: metallic
debris), the measurement time was long particularly in the case of B-10 detectors.

©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.76

2.2.1 Passive neutrons – Summary➢

Major findings obtained from the analysis:
Impact
Major

Minor

Influencing factor② Burn-up⑤ Moisture content① Fuel debris composition:(Neutron absorption
material)(③ FP emission rate*1)④ Cooling period⑥ Filling rate:⑦ Uneven distribution⑧ Container

Findings
・The shape of the neutron spectrum that leaks from the container
does not change depending on the burn-up, however, the fluxchanges largely by the order of magnitude. [Refer to No. 43]
・When moisture content increases, the fast neutron flux
decreases and thermal neutron flux increases. [Refer to No. 44]

・The shape of the neutron spectrum and the neutron flux in thethermal energy area change due to the neutron absorption material,
but these changes do not appear significantly at the location of thedetector having a polyethylene moderator.
・There is a possibility of correcting the cooling period by means of
the retrieval time period and half-life.
・The filling rate has a minor impact.
・The farther away the fast neutron flux is from the surface ofthe container, the smaller is its impact.
・There is a possibility of being able to correct the shape of thecontainer.

*1 ③ The FP emission rate is the emission rate of gamma rays and does not have an
impact on the measurement of neutron radiation.

②
2 core average② Burn-up:
燃焼度： Unit2号炉心平均③
FPemissionrate:
Test (Pheus-FPT4) base
③ FP放出率： 試験(Phebus-FPT4)ベース④
Cooling
period:20④ 冷却期間： 20年 years① Fueldebris composition
燃料デブリ組成:UO2: 50 (vol%)ZrO2: 50 (vol%)SUS,:0 (vol%)
SUS, Concrete:
0 (v0l%)
B4C, Gd2O3: 0 (vol%)
MOX: 0 (vol.%)
Empty(porosity):
Empty(気孔率) : 00 (vol%)(vol%)⑤
content:⑤ Moisture
含水率：1(wt%) 1 (wt%)⑥
rate: 30(vol%)(vol%)⑥ Filling
充填率：30⑦ Uneven distribution: None (= Uniform)⑦ 偏在：なし (=均質)⑧ Container:
容器：ユニット缶⑧
Unit can(φ210mm×H200mm)

Parameter values in the figure indicate base case conditions.

➢

Study of the necessity of detailed analysis and consolidation of technical issues：•
Analysis of the extent of impact of heterogeneity on the flux on the measuring surface•
There is a possibility of correcting the rate of change in flux by correcting the impact of the burn-up
in fuel debris.
•
The flux reduces by the order of magnitude due to multiple influencing factors. And, even with thesame amount of nuclear material, neutron flux differs by the order of magnitude depending on the
difference in burn-up. Hence a measurement method that covers a broad range will be studied.
©International Research Institute for Nuclear Decommissioning

No.77

2. Implementation Details(1) Analytical evaluation of factors influencing measurement errors of prospective measurement
technologies①

Selection of influencing factors②

Setting the fluctuation range of the influencing factors③

Analytical simulation
2.1Analysis conditions
2. 2 Analysis results
2.2.1 Passive neutrons
2.2.2 Passive gamma
2.2.3 Active neutrons
2.2.4 Passive/active neutrons + passive gamma rays
2.2.5 X-ray transmission measurement
2.2.6 Cosmic rays scattering measurement①
Fuel debris
composition
2.2.2
Passive γ

No.83
No.85 to 87
No.93

Gd
content

B
content

－

－

MOX

No.85
No.86
No.88
No.91

②
Burn-up③
FP
emission
rate

No.85
No.86
No.88

No.85
No.86
No.88

④
Cooling
period⑤
Moisturecontent⑥
Fillingrate⑦
Unevendistribution⑧
Container⑨
Irradiatingradiation
source

No.85
No.86
No.88
No.92

－

No.85
to 87

No.97
No.98

No.94
No.95

－

⑩
Detector

No.84
No.89
No.100No.101©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.78

2.2.2 Passive gamma rays - Measurement concept⚫

Fuel debris contains gamma rays emitting nuclides (Am-241,Cs-137,Eu-154, etc.) originating from fuel and gamma rays emitting
nuclides (C-60, etc.) originating from radioactivation of traces of impurities in metals.

⚫

So nuclides originating from fuel need to be identified. Hence the energy spectrum of gamma rays is measured.

⚫

Further, since the gamma flux from the nuclides originating from fuel varies depending on the burn-up, etc. or the self-shielding

effect of the gamma rays differs depending on the state of enclosure, there is variation in the gamma flux reaching the measuring
surface.
⚫

Hence, by determining the width to be measured by means of analytical evaluation, issues such as selection of thenuclides to be measured, the measurement range of the detector required for the measurement system, changeability ofthe measurement distance or shielding, collimator, etc. were identified.

Container
Count rate (cps)Gamma rays detector
Fuel debris

Gamma rays
Energy (keV)

Example of analytical output (Ge detector)©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.79

2.2.2 Passive gamma rays - Analysis model

In order to make it possible to study the changes in the radiation flux depending on the distance from the container, multiple layers of space
were provided for cylindrical detectors at a fixed distance from the container (unit can, etc.).
→ The radiation dose rate calculated based on the changes in radiation flux will be used for selecting the detector, andstudying the shielding, collimator, etc.

Unit canFuel debris

200mm

Fuel debris

150mm

A’

A

10mm

1000mmUnit canT-cross Tally

Φ210mmT-cross Tally

Cross-sectional view A-A’
T-cross Tally

Analysis model top view

©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.80

2.2.2 Passive gamma rays - Table listing the analysis cases (1/2)•

Considering the analysis conditions described in No. 1-1 as the base case, the analysis conditions with changed
influencing factors (No. 1-2 to 24) were assumed as the sensitivity analysis conditions. All 24 cases
Influencing factors that changed from the base case under the sensitivity analysis conditions are highlighted
blue. Here, the parameters that changed in association with the above-mentioned changes are indicated byyellow hatching.

•Composition inside the container*1
Case No.

Type of fueldebris1-1

Molten debris

MOX : 30vol%UO2 : 5vol%(50vol%)1-4

ZrO2 : 5vol%(50vol%)UO2 : 25vol%(50vol%)1-5

ZrO2 :25vol%(50vol%)UO2 : 15vol%(50vol%)1-6

ZrO2 : 15vol%(50vol%)UO2 : 15vol%(50vol%)1-7

ZrO2 : 15vol%(50vol%)UO2 : 15vol%(50vol%)1-8

ZrO2 : 15vol%(50vol%)UO2 : 15vol%(50vol%)1-9

ZrO2 : 15vol%(50vol%)UO2 : 15vol%(50vol%)

1-10

1-12

1-13

1-14

1-15

UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)ZrO2 : 22.5vol%(75vol%)1-3

ZrO2 : 15vol%(50vol%)Molten debris

Burn-up

FP emission rate

Cooling period

Uneven distribution

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

0GWd/tStandard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

1.3GWd/t

Standard

20 yearsUniform

51GWd/tStandard

20 yearsUniform

23.0GWd/t

Zero emission

20 yearsUniform

23.0GWd/t

High emission

20 yearsUniform

23.0GWd/t

Standard

30 yearsUniform

23.0GWd/t

Standard

40 yearsUniform

23.0GWd/t

Standard

20 years

23.0GWd/t

Standard

20 years

23.0GWd/t

Standard

20 years

Container

Changed parameter

Outside the filling factor
Composition

UO2 : 7.5vol%(25vol%)

1-2

1-11

Within the filing factor

UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)UO2 : 0.48vol% [3.7kg]
SUS : 10vol%(95.4vol%)Total (Filling factor)

30vol%

30vol%

30vol%

10vol%

50vol%

30vol%

30vol%

30vol%

30vol%

30vol%

30vol%

30vol%

30vol%

30vol%

10.48vol%UO2 (Moisture content): 1wt%
Empty: Remainder
H2 O (Moisture content): 1wt%
Empty: Remainder
H2 O (Moisture content): 1wt%
Empty: Remainder
H2 O (Moisture content): 1wt%
Empty: Remainder
H2 O (Moisture content): 1wt%
Empty: Remainder
H2 O (Moisture content): 1wt%
Empty: Remainder
H2 O (Moisture content): 1wt%
Empty: Remainder
H2 O (Moisture content): 1wt%
Empty: Remainder
H2 O (Moisture content): 1wt%
Empty: Remainder
H2 O (Moisture content): 1wt%
Empty: Remainder
H2 O (Moisture content): 1wt%
Empty: Remainder
H2 O (Moisture content): 1wt%
Empty: Remainder
H2 O (Moisture content): 1wt%
Empty: Remainder

H2 O (Moisture content): 1wt%
Empty: Remainder

H2 O (Moisture content): 1wt%
Empty: Remainder

Unit can(Φ210mm x H200mm)
Unit can(Φ210mm x H200mm)
Unit can(Φ210mm x H200mm)
Unit can(Φ210mm x H200mm)
Unit can(Φ210mm x H200mm)
Unit can(Φ210mm x H200mm)
Unit can(Φ210mm x H200mm)
Unit can(Φ210mm x H200mm)
Unit can(Φ210mm x H200mm)
Unit can(Φ210mm x H200mm)
Unit can(Φ210mm x H200mm)

Horizontal direction(Center)

Unit can(Φ210mm x H200mm)

Horizontal direction(Outer surface)

Unit can(Φ210mm x H200mm)

Base case

Fuel debris composition(UO2, ZrO2)

MOX

Filling factor

Filling factor

Burn-upBurn-up

FP emission rate

FP emission rate

Cooling period

Cooling period

Uneven distribution

Uneven distribution

Canister
Uniform(Φ220mm x H840mm, Thickness

Container

10mm)
Waste storage container
23.0GWd/t

Standard

20 yearsUniform(Inner container □Φ500mm x

Container

H300mm)©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.81

2.2.2 Passive gamma rays - Table listing the analysis cases (2/2)
Composition inside the container*1
Case No.

1-16

Within the filing factor
Type of fuel debris

Uranium-richOutside the filling factor
Composition

Total (Filling factor)

UO2 : 30vol%

30vol%

UO2 : 0.075vol%(0.25vol%)1-17

Metallic debris

ZrO2 : 0.075vol%(0.25vol%)

30vol%

H2 O (Moisture content): 1wt%
Empty: Remainder
H2 O (Moisture content): 1wt%
Empty: Remainder

Burn-up

FP emission rate

Cooling period

Uneven distribution

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 years

Container

Unit can(Φ210mm x H200mm)
Unit can(Φ210mm x H200mm)

Changed parameter

Fuel debris composition(Type)
Fuel debris composition(Type)SUS : 29.85vol%(99.5vol%)UO2 : 15vol%(25vol%)1-18

ZrO2 : 15vol%(25vol%)

60vol%SUS : 30vol%(50vol%)H2 O (Moisture content): 1wt%
Empty: Remainder

Unit can(Φ210mm x H200mm)

Fuel debris composition（SUS）

UO2 : 1.05vol%(3.5vol%)ZrO2 : 1.05vol%(3.5vol%)
1-19

MCCI

SUS : 7.2vol%(24vol%)

30vol%

Concrete :H2 O (Moisture content): 1wt%
Empty: Remainder

Unit can(Φ210mm x H200mm)

Fuel debris composition(Type)

20.7vol%(69vol%)UO2 : 15vol%(25vol%)1-20

ZrO2 : 15vol%(25vol%)

60vol%

Concrete : 30vol%(50vol%)UO2 : 0.075vol%(0.25vol%)1-21

Metallic debris

ZrO2 : 0.075vol%(0.25vol%)

30vol%SUS : 29.85vol%(99.5vol%)UO2 : 0.075vol%(0.25vol%)1-22

ZrO2 : 0.075vol%(0.25vol%)

30vol%SUS : 29.85vol%(99.5vol%)H2 O (Moisture content): 1wt%
Empty: Remainder

H2 O (Moisture content): 1wt%
Empty: Remainder

H2 O (Moisture content): 1wt%
Empty: Remainder

23.0GWd/t

Standard

20 years

Horizontal direction(center)

Horizontal direction(outer surface)

Unit can(Φ210mm x H200mm)

Unit can(Φ210mm x H200mm)

Unit can(Φ210mm x H200mm)

Fuel debris composition(Concrete)

Fuel debris composition(Type)
Uneven distribution

Fuel debris composition(Type)
Uneven distribution

UO2 : 1.05vol%(3.5vol%)ZrO2 : 1.05vol%(3.5vol%)
1-23

MCCI

SUS : 7.2vol%(24vol%)

30vol%

H2 O (Moisture content): 1wt%
Empty: Remainder

23.0GWd/t

Standard

20 years

Horizontal direction(center)

Unit can(Φ210mm x H200mm)

Concrete :Fuel debris composition(Type)
Uneven distribution

20.7vol%(69vol%)UO2 : 1.05vol%(3.5vol%)ZrO2 : 1.05vol%(3.5vol%)1-24

SUS : 7.2vol%(24vol%)Concrete :30vol%

H2 O (Moisture content): 1wt%
Empty: Remainder

23.0GWd/t

Standard

20 years

Horizontal direction(outer surface)

Unit can(Φ210mm x H200mm)

Fuel debris composition(Type)
Uneven distribution

20.7vol%(69vol%)*1 : Percentage inside the container. Percentage, when 100% is considered to be within the filling rate, is mentioned inside parentheses.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.82

2.2.2 Passive gamma rays - Table listing the analysis cases•

Sensitivity analysis related to the following analysis conditions was conducted, and the trend with respect to each influencing factor
has been consolidated.
Case

UO2

ZrO2

SUS

Concrete

Total(Filling rate)

Burn-up(GWd/t)FP emission
rate

Remarks
Cooling period U content considered as(Years)the analysis condition

1-1

Base

15vol%

15vol%

0vol%-

30vol%

23.0

Standard

20

8.79kg

1-2

Molten debris(Small quantity of U)

7.5vol%22.5vol%

0vol%-

30vol%

23.0

Standard

20

4.39kg

1-3

MOX

MOX
30vol%

0vol%

0vol%

-

30vol%

0.0

Standard

20

17.0kg1-4

Filling rate (low)

5vol%

5vol%

0vol%

-

10vol%

23.0

Standard

20

2.93kg1-5Filling rate (high)25vol%

25vol%

0vol%-

50vol%

23.0

Standard

20

14.6kg

1-6

Burn-up (low)15vol%

15vol%

0vol%-

30vol%

1.3

Standard

20

8.98kg

1-7

Burn-up (high)15vol%

15vol%

0vol%-

30vol%

51.0

Standard

20

8.58kg

1-8

FP emission rate

15vol%

15vol%

0vol%-

30vol%

23.0

Zero emission

20

8.77kg

1-9

FP emission rate

15vol%

15vol%

0vol%-

30vol%

23.0

High emission

20

8.82kg

1-10

Cooling period

15vol%

15vol%

0vol%-

30vol%

23.0

Standard

30

8.79kg

1-11

Cooling period

15vol%

15vol%

0vol%-

30vol%

23.0

Standard

40

8.79kg

1-16

Molten debris(Uranium-rich)

30vol%

0vol%

0vol%

-

30vol%

23.0

Standard

20

17.6kg

1-17

Metallic debris

0.075vol％0.075vol%29.85vol%-

30vol%

23.0

Standard

20

0.04kg

1-19

MCCI debris

1.05vol%

1.05vol%

7.2vol%20.7vol%

30vol%

23.0

Standard

20

0.615kg©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.83

2.2.2 Passive gamma rays - Analysis results (1/7)
Flux at a position that is 100cm away from the container surface - Analysis of sensitivity to the type of fuel debris
1-1 (Molten debris)
Composed of UO2 and ZrO2

Flux (γ/cm2/s)

Flux (γ/cm2/s)

1-17 (Metallic debris)
SUS is 99.5%

Energy (MeV)

Energy (MeV)

Flux (γ/cm2/s)

1-19 (MCCI)
Concrete 69% and SUS 24%

Energy (MeV)

Case
No.

Type offuel debris

UO2

ZrO2

SUS

Concrete

Filling rate

1-1

Moltendebris15vol%(50vol%)15vol%(50vol%)

0vol%

0vol%

30vol%

1-17

Metallic
debris

0.075vol%(0.25vol%)0.075vol%(0.25vol%)29.85vol%(99.5vol%)

0vol%

30vol%

1-19

MCCI

1.05vol%(3.5vol%)

1.05vol%(3.5vol%)

7.2vol%(24vol%)20.7vol%(69vol%)30vol%

Flux spectrum largely differed in the case ofmolten debris, metallic debris and MCCI.

→ Evaluation was carried out using the above-mentioned flux, focusing on the dose rate and the peak of Eu-154 (1.27MeV).
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.84

2.2.2 Passive gamma rays - Analysis results (2/7)
Changes in flux depending on distance from the center of the container

In order to study the placement of the detector based on the dose rate, the changes in the flux from the container surface were evaluated.

Evaluation was carried out based on the flux from a distance of 1cm, 15cm and 100cm from the container surface by fitting the detector.
→ Changes in the flux are proportional to the square of the distance from the center of the container.

Changes in flux from the center of the container

Simulation system
20cm

15cm

100cm

Unit canTotal flux γ [106γ/s/cm2]

Fuel debris 1cm

21cm

Distance [cm] from thecontainer surface

Distance [cm] from thecenter of the container

Total flux
[γ/s/cm2]1

11.5

5.3×10815

25.5

1.0×108

100

115.5

5.2×106

Distance from the center of the container x[cm]©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.85

2.2.2 Passive gamma rays - Analysis results (3/7)
Conversion of flux to air kerma rate
Conversion of flux [γ/cm2/s] to air kerma rate [mGy/h]・The placement, shielding, etc. of the gamma rays detector wasused for evaluating the impact of gamma rays on the neutron
detector.
Metallic debris
Molten debris

JAERI-Data/Code 2000-044
Shielding Calculation Constants for Use in Effective
Dose Evaluation for Photons, Neutrons, andBremsstrahlung from Beta-rays

Table 3.3 Dose Conversion Coefficients for converting
photons into effective dose in AP orientation
Effective dose
per unit fluencein APorientation

Flux [γ/cm2/s]

MCCI

Linear interpolation of the conversion coefficient using JAEA data
as reference

Case No.

Flux [γ/cm2/s] at a position that is 100cm from the container surface
Metallic debris
MCCI

Air-kerma rate [mGy/h]

Molten debris

Case No.

*) Conversion coefficient for air kerma**) Effective dose in AP orientation per unit air kerma
in free air

Air kerma rate [mGy/h] at a position that is 100cm from the container surface©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.86

2.2.2 Passive gamma rays - Analysis results (4/7)
Evaluation of air kerma rate

Air kerma rate [mGy/h] at a position that is 100cm from the container surface (Repeated)

Molten debris

Metallic debris
MCCI (Molten Core Concrete Intraction)

Filling rate

Burn-upFP
emission rate

Cooling
Debrisperiod composition

Air kerma rate [mGy/h]Debriscomposition
MOX
Base case

Case No.
Air kerma rate fluctuations were larger when the fuel-based parameters were changed than when the fuel debris
composition was changed.
→ In particular, the dose rate largely changes due to the effect of MOX, burn-up and FP emission rate.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.87

2.2.2 Passive gamma rays - Analysis results (5/7)
Eu-154 (fuel debris composition) with respect to the amount of U

Passive gamma rays analysis results (Analysis performed by changing ① Fuel debris composition and ⑥ Filling rate)
Case
No.

Fuel debris
Type

Filling rate
(vol.%)

1-1
1-2
1-4
1-5
1-16
1-17
1-19

Molten debris
Molten debris
Molten debris
Molten debris
Molten debris
Metallic debris
MCCI

30
30
10
50
30
30
30

Entire fueldebris (kg)
15.8
13.4
5.28
26.4
20.5
15.6
7.71

Amount of U*(kg)

Total flux
[γ/s/cm2]

8.79
4.39
2.93
14.6
17.6
0.04390.615

5.23×106
3.63×106
3.40×106
5.63×106
6.93×106
7.63×106
1.08×106

1.27MeV peak(Eu-154)[γ/s/cm2]
1.77×105
1.01×105
1.03×105
1.94×105
2.77×105
1.50×103
7.15×103

1.27 MeV (Eu-154) [γ/s/cm2]*Amount of U = Total mass of Uranium isotopes
MCCI: Molten Core Concrete Interaction

5.0E+05← Flux of 1.27 MeV (Eu-154) with respect to amount of U

Molten debris
Metallic debris
MCCI

4.0E+05
3.0E+05

1-1

2.0E+05
1-4

1-16① Composition

If only fuel debris composition and filling rate were changed,
the flux of Eu-154 peak was proportionate to the amount of U.
→ There is a possibility of being able to perform quantitativeevaluation of the amount of U based on the gamma rays from
Eu-154

1-5⑥ Filling rate

1.0E+05
1-2

0.0E+00
0

5

10
15
Amount of U [kg]20
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.88

2.2.2 Passive gamma rays - Analysis results (6/7)
Eu-154 (fuel based) with respect to the amount of U

Passive gamma rays analysis results (Analysis performed by changing ② Burn-up, ③ FP emission rate and ④ Cooling period)

Case
No.

Burn-up[GWd/t]FP emission rate

Cooling period[Years]

1-1
1-3
1-6
1-7
1-8
1-9
1-10
1-11

23
0*2
1.3
51
23
23
23
23

Standard
Standard
Standard
Standard
Zero emission
High emission
Standard
Standard

20
20
20
20
20
20
30
40

Entire fuel Amount ofdebris (kg)
U*1
(kg)
15.8
8.79
20.3
17.0
15.8
8.98
15.8
8.58
15.8
8.77
15.8
8.82
15.8
8.79
15.8
8.79

Total flux
[γ/s/cm2]
5.23×106
2.80×105
3.06×105
1.16×106
2.81×107
1.15×106
3.81×106
2.89×106

1.27MeV peak(Eu-154)[γ/s/cm2]
1.75×105
1.02×103
5.92×102
4.45×105
1.76×105
1.75×105
7.84×104
3.50×104*1 Amount of U = Total mass of Uranium isotopes, *2 The MOX fuel in Case no. 1-3 has extremely low burn-up

Molten debris← Flux of 1.27 MeV (Eu-154) with respect to amount of U
1-7(51GWd/t)

Metallic debris

MCCI

1-1 (Base case),
1-8, 9 (FP emission rate)

1-10
1-11
(Cooling period)

1-6(1.3GWd/t)

1-3(0GWd/t)Even when fuel debris composition ratio was the same,
when burn-up changed the correlation between Eu-154 andamount of U deteriorated.
(Eu-154 affects the quantitative evaluation of the amount of U in
the order of Burn-up > Cooling period > FP emission rate.)

Amount of U (kg)©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.89

2.2.2 Passive gamma rays - Analysis results (7/7)
Response spectrum for a representative detector (Example of output for identifying issues)<Analysis model>

Lead collimator was required
because of a high count rate↓

<Detector response>

Counting rate (cps)・ representative detector: High purity germanium
semiconductor detector・ Size and shape of detector : 3 inch cylindrical detector
・ Distance up to the detector: 1000mm(From outer surface of the unit canto the center of the detector)

Base case (No.1-1)

Ge Detector(5.323g/cm3)

195mm

200mm

Radiation source
UO2+ZrO2

Φ206mm

Φ210mmSUS316L
7.98g/cm3

Counting rate (cps)Energy (keV)

Nuclear material may not be detected if the energy isequal to or lower than the Compton edge of Co-60.

Metallic debris (No.1-17）

1000mmEnergy (keV)

Further, if Eu-154 is not co-present with U, nuclear material cannot be detected.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.2 Passive gamma rays – Other sensitivity analyses (1/6)

No.90

The sensitivity to incident beams was studied with respect to the following influencing factors:
➢
➢
➢
➢

Case
No.

MOX
Cooling period
Shape of container
Implementation of analysis related to fluctuations in fuel debris composition (SUS)

UO2

ZrO2

SUS

Total
(Filling rate)

Other conditions

Remarks

1-1

15vol%

15vol%

0vol%

30vol%

-

Base

1-2

7.5vol%22.5vol%

0vol%

30vol%-

Molten debris

1-3

0vol%

15vol%

0vol%

30vol%

Pu02: 15vol%

1-10

15vol%

15vol%

0vol%

30vol%

Cooling period: 30 years

1-11

15vol%

15vol%

0vol%

30vol%

Cooling period: 40 years

1-14

15vol%

15vol%

0vol%

30vol%

Canister (Φ220mm×H840mm)

1-15

15vol%

15vol%

0vol%

30vol%

Waste storage container

1-16

30vol%

0vol%

0vol%

30vol%-

Molten debris

1-170.075vol％0.075vol%29.85vol%

30vol%-

Metallic debris

1-18

15vol%

15vol%

30vol%

60vol%

MOX

Cooling period

Shape ofcontainer

Metallic debris©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.2 Passive gamma rays – Other sensitivity analyses (2/6)

No.91
Other common conditions:・ FP emission rate: Standard
・ Moisture content 1wt%
・ Homogeneous model・ Container: Unit can➢ Impact due to difference in UO2 fuel and MOX fuel・ Am-241 peak generated by Pu contained in MOX was seen.
(However, from the perspective of detector response,
it may not be detected as it is hidden in the Compton zoneof Cs-137.)

Case
No.

Type offuel debris

1-6

UO2

ZrO2

PuO2

Fillingrate

Burn-up15vol%

15vol%

0vol%

30vol%

1.3GWd/t

0vol%

0vol%

30vol%

30vol%

0GWd/tMoltendebris1-3

Flux at 100cm from the container surface
1-3 (Extremely low burn-up of molten
debris (MOX))

Flux (γ/cm2/s)

Flux (γ/cm2/s)

1-6 (Low burn-up of molten debris (UO2)Energy (MeV)

Energy (MeV)©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.2 Passive gamma rays – Other sensitivity analyses (3/6)

No.92➢ Impact due to difference in cooling period
・ Incident flux attenuated basedon the half-life specific to thenuclide.

Case
No.

Type offuel debris

1-1

1-10

Moltendebris

1-11

UO2

ZrO2

SUS

Concrete

Filling rate

Cooling
period15vol%(50vol%)15vol%(50vol%)

0vol%

0vol%

30vol%

20 years

15vol%(50vol%)15vol%(50vol%)

0vol%

0vol%

30vol%

30 years

15vol%(50vol%)15vol%(50vol%)

0vol%

0vol%

30vol%

40 years

Other common conditions:・ Burn-up 23GWd/t
・ FP emission rate:
Standard
・ Moisture content 1wt%
・ Homogeneous model・ Container: Unit can

Flux at 100cm from the container surface

Co-60

Incident flux (γ/cm2/s)

Incident flux (γ/cm2/s)

Eu-154

Cooling period (years)

Half-life: approx. 8.6 years⇒ Attenuates approx. 1/5 times in 20 years

Cooling period (years)

Half-life: approx. 5.3 years⇒ Attenuates approx. 1/14 times in 20 years©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.2 Passive gamma rays – Other sensitivity analyses (4/6)➢ Impact due to difference in SUS content
Case
No.

・ As apparent density increased, incident flux
decreased due to the self-shielding effect.

Flux at 100cm from the container surface

Type offuel debris

No.93
Other common conditions:

UO2

ZrO2

SUS

Concrete

Filling rate

1-1

Moltendebris15vol%(50vol%)15vol%(50vol%)

0vol%

0vol%

30vol%

1-18

Metallic
debris15vol%(25vol%)15vol%(25vol%)30vol%(50vol%)0vol%

60vol%・ Burn-up 23GWd/t
・ FP emission rate: Standard
・ Moisture content 1wt%
・ Homogeneous model・ Container: Unit can

Incident flux per unit of Uranium (γ/cm2/s/gU)

Fuel debris apparent density and incident flux per unit of Uranium

F u e densitylApparentof fuel debris (g/cm3)©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.2 Passive gamma rays – Other sensitivity analyses (5/6)

No.94

Waste storage container (Inner container) （No. 1-15）➢ Impact due to difference in shape ofcontainer

Case
No.

・ Since the container was larger than the unit can as well,incident flux decreased due to self-shielding effect.

1-4

1-15

<Unit can>Type offuel debris

UO2

ZrO2

SUS

Filling rate

Container

Moltendebris

5vol%

5vol%

0vol%

10vol%

Unit canMetallic
Waste

0.5vol%（3.7kg)0vol%

10vol%

10.5vol%

Waste container(Inner container)

Other common conditions:・ Burn-up 23GWd/t
・ FP emission rate: Standard
・ Moisture content 1wt%
・ Cooling period: 20 years
・ Homogeneous model

Fuel debris apparent density and incident flux per unit of UraniumGe
Detector

Unit canVertical cross-section

Horizontal cross-section<Waste container (Inner container)>
500
500
Ge
Detector
500
300

Incident flux per unit of Uranium (γ/cm2/s/gU)

Fuel debris

Unit canWaste storage
container(Inner container)

2

Horizontal
cross-section

2
Inner waste container

Fuel debris

Vertical cross-section(Unit: mm)ApparentFueldensity of fuel debris (g/cm3)©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.2 Passive gamma rays – Other sensitivity analyses (6/6)

No.95➢ Impact due to difference in shape of container
Relation between U content and incident flux (at 100cm from the container surface)

Incident flux (γ/cm2/s/gU)

15vol%×4
No.1-14 (Canister)

30vol%
No.1-16 (U-rich)
15vol%
No.1-1 (Base case)
7.5vol%No.1-2Other common conditions:・ Burn-up 23GWd/t
・ FP emission rate: Standard
・ Moisture content 1wt%
・ Homogeneous model

CanisterGe Detector

Unit can0.075vol%No.1-17 (Metallic debris)Amount of U (g)・Amount of U and the entire incident flux had a positive correlation.
・The incident flux peaked due to the impact of the self-shielding effect caused by variations in fuel debris density (filling rate).
・In the case of the canister (Φ220mm×H840mm), since the detector was far away for the gamma rays coming from the top
and bottom ends of the container, and since the canister had thick walls of 10mm, the incident flux peaked further.

©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.2 Passive gamma rays - Sensitivity analyses Uneven distribution (1/3)

No.96

Analysis by changing the method of filling fuel debris in the container■Tally settings of the uneven distribution cases
surface) cases only.

#The tally was divided into 6 parts for the horizontal direction (outer

Tally
Unit can(Region divided equally into 6 parts)

Nuclear fuel

(2)

(3)

Unit can
30°

A
(4)

(1)

A’

9.75cm9.75cm10.3cm(6)(5)

Top view

Tally(No vertical division)

Tally(No vertical division)

Cross-sectional view A-A’

The location and dimensions for installing the tally for the unit can are the same as described in slide No. 79.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.97

2.2.2 Passive gamma rays - Sensitivity analyses Uneven distribution (2/3)a) Case of uneven distribution with 1.27MeV fluxb) Case of uneven distribution with 1.27MeV flux

1.E+08100%

1.E+07

1.E+06
1-1
1-12
1-13(1)1-13(2)
1-13(3)
1-13(4)

1.E+05

1.E+041

10100

Distance [cm] from the container surface

Ratio of 1.27 MeV (Eu-154) flux with the base case1.27 MeV (Eu-154) flux [γ/s/cm2]■Case of uneven distribution of molten debris:1.27MeV flux analysis results(For the uneven distribution cases, calculations were done by installing the tally at 1cm, 15cm and 100cm from thecontainer surface.)

90%
80%

70%
60%
50%

40%
1-12
1-13(1)1-13(2)
1-13(3)
1-13(4)

30%
20%

10%
0%
1

10

100

Distance [cm] from the container surfacea) Flux was higher when the molten debris was uniformly present (1-1) throughout the container, than when there wasuneven distribution (No. 1-12, 1-13).
b) The greater the distance from the surface of the container, the smaller was the impact of uneven distribution on the flux.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.2 Passive gamma rays - Sensitivity analyses Uneven distribution (3/3)

No.98

1.27MeV(Eu-154) flux (γ/s/cm2)■ Comparison of cases of uneven distribution of molten debris, metallic debris and MCCI
(Molten Core Concrete Interaction): 1.27MeV flux analysis results
Molten debris (No. 1-12)
Metallic debris (No. 1-21)
MCCI (No. 1-23)

Horizontal
Horizontal direction
direction(center)(outer surface)

Molten debris (No. 1-13)
Metallic debris (No. 1-22)
MCCI (No. 1-24)

Horizontal
direction(outer surface)

Horizontal
direction(outer surface)

Horizontal
direction(outer surface)

Since the proportion of material other than nuclear fuel (SUS, Concrete, etc.) was more in metallic debris and
MCCI than in molten debris, if nuclear fuel was unevenly distributed in metallic debris and MCCI, the fluxchanged largely.
* The molten debris base case is used for other factors.

©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.99

2.2.2 Passive gamma rays - Detector response (1/3)
Selection of prospective detectors and analysis conditions■List of prospective detectors
Prospective
detectors

Material and shape(Reference) Energy resolution

Remarks

Ge​

Cylindrical (Diameter and thickness
7.62cm)0.15%FWHM

Mirion Technologies (Canberra)

CZT

Cylindrical (Diameter 7.62cm andthickness 1cm)

0.8%FWHM

H3D (University of Michigan)

LaBr3

Cylindrical (Diameter and thickness
7.62cm)3.5%FWHM

ORTEC■Analysis system

Assuming that a collimator will be installed in the actual system, gamma rays were made to fall
perpendicularly on the surface of the detector.
(The impact of the thickness of the collimator, scattered rays, etc. is planned to be evaluated and studied
in the future.)

7.6cm

Example) Ge Detector
Flux in Gamma rays Case No. 1-1

Irradiatedperpendicularly on
the detector surface

10cm

Ge Detector

7.6cm©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1002.2.2 Passive gamma rays - Detector response (2/3)

Comparison of the response spectrum of prospective detectors■Response of each prospective detector in the base case
Energy spectrum for each detector assuming thedetector is 100cm from the container surface➢

Ge and CZT were able to distinguish between
1.27MeV(Eu-154) and 1.33MeV (Co-60), but this
distinction was difficult in the case of LaBr3.

108
Ge
CZT
LaBr3

Counting rate [cps]107

Total count rate and count rate at 1.27MeV peak

Indicator nuclide

106

Prospective
detector

Total count rate[cps]

1.27MeV count rate*[cps] (%)

Ge

2.2×1082.2×106(1.0%)

CZT

9.1×107

3.8×105(0.4%)

LaBr3

2.2×108

ー

105
104
103

* Counting rate at the base region resulting from scattered rays is
subtracted from that at the peak region resulting from gamma rays
with 1.27MeV.

102
0

0.5

1.0

1.5

2.0

➢

The measurement time was estimated from the countrate at 1.27MeV peak.

Detected Energy [MeV]©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1012.2.2 Passive gamma rays - Detector response (3/3)
Estimation of measurement time for the prospective detectors■Study of installation of every prospective detector (Assuming that the detector is installed at a
distance of 100cm from the container surface)
Collimator•

For a Ge detector with maximum count rate 25kcps*1, 3, a collimator ofdiameter approx. 0.1cm was required (2.2×108→2.5×104cps)•

For a CZT detector with maximum count rate 450kcps*2, 3, a collimator ofdiameter approx. 0.5cm was required (9.1×107→4.5×105cps)*1 Assuming that a gamma spectroscopy software of the same manufacturer

Moltendebris(No.1-1)

was used.

Detector

Unit can(Container)*2 Count rate limit of the H series detector of the same manufacturer.
*3 Dead time due to pile-up was not considered.

100cm■Estimation of the computation time of each prospective detector (Example: Assuming that the collimator is 20cm thick)Measurement time[seconds/round](Time required to obtain 104
counts at the 1.27MeV peak)

Measurement
frequency[Rounds /Container](Translationalscanning only)

Computational timefor each container

Prospective
detector

Collimator
diameter [cm]

Maximum countrate [cps]Proportion of the countrate of 1.27MeV

Ge

0.1

2.5×104 *1,3

1.0%

40

500

5.5 hours

CZT

0.5

4.5×105 *2,3

0.4%

6

20

2 minutes➢

In order to install a collimator for reducing the gamma rays incident upon the detector, measurement had to be performed while scanning thecontainer. Hence the measurement frequency increased due to which the process took time. The collimator installation and measurement
technique need to be improvised.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1022.2.2 Passive gamma rays - Additional sensitivity analyses (1/4)
Step 2 Analysis (1/4)

Five cases in which the peak count rate at 1.27MeV was less were analyzed taking multiple influencing factors into consideration.
■ Cases with a combination of fluctuations in 2 influencing factors
MoltenFuel debris composition
Fuel debris

Metallic

MCCI

MOX

Filling factor

Burn-up

FP emission rate

Cooling period

Uneven distribution

Container

-

△

△

x

x

△

x

-

-

Flux increases/decreases

Flux increases/decreases

No change in Eu-154 peak

Attenuation depending on

Flux increases/decreases with thehalf-life

increase/decrease in amount of U

x

△

△

x

composition

with the increase/decrease with the increase/decrease

MOX

in amount of U

in amount of U

(Self-shielding effect(Self-shielding effectpresent)

present)△-(Self-shielding effect present)

Flux increases/decreases

x

x

△

No change in Eu-154 peak

Attenuation depending on

Flux increases/decreases with thehalf-life

increase/decrease in amount ofwith the increase/decreasein amount of MOX

MOX(Self-shielding effect(Self-shielding effect present)

present)
Filling factor

△

x

x

△

△

△

Flux increases/decreases

No change in Eu-154 peak

Attenuation depending on

Flux increases/decreases with theFlux increases/decreases

Flux increases/decreases

half-life

increase/decrease in amount of U

with the increase/decrease with the(Self-shielding effect present)in amount of U

increase/decrease in(Self-shielding effect(Self-shielding effectamount of U

present)

present)(Self-shielding effectwith the increase/decreasein amount of U

Moltenpresent)
Burn-upxx△

No change in Eu-154 peak

Attenuation depending on

Self-shielding effect due to thehalf-life

form of the radiation source for

x

〇

〇

x

x

x

No change in Eu-154 peakNo change in Eu-154 peakxx

Attenuation depending on

Attenuation depending on

half-life

half-life

〇

〇

x

x

〇

〇

uneven distribution is present
FP emission

x

x

rate

Attenuation depending on

Expected to not be different thanhalf-life

the uneven distribution in Step 1

Cooling periodxx

Attenuation depending on half-life
Unevenx〇〇

distribution
Container
Metallic-

MCCI

Metallic debris and MCCI cases in which the burn-up and uneven distribution were changed were selected.
Metallic debris

MCCI

Burn-up

Small: 1.3GWd/t

Small: 1.3GWd/t

Uneven distribution

At the center in thehorizontal direction

At the center in thehorizontal direction

4 cases(Of these, 2 cases
are redundant)* The base case was used for other factors.

■Case in which the peak count rate at 1.27MeV was assumed to be the smallest
Filling rate

Burn-up

Uneven distribution

10vol% (UO2: 5vol% / ZrO2: 5vol%)Small: 1.3GWd/t

At the center in thehorizontal direction

Total
5 cases

1 case©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1032.2.2 Passive gamma rays - Additional sensitivity analyses (2/4)
Step 2 Analysis (2/4)
Five cases in which the peak count rate at 1.27MeV was less were analyzed taking multiple influencing factors into consideration.

■ Step 2 analysis cases
Remarks
U content
considered as
the analysis
condition

UO2

ZrO2

SUS

Concrete

Filling rate

Burn-up(GWd/t)FP
emissionrate

Cooling
periodUnevendistribution

1-21

Metallic debris
Unevendistribution(center)

25vol%

25vol%--

30vol%

23.0

Standard

20 years

Unevendistribution(Center)

0.0439kg

1-25

Metallic debris
Burn-up (low)0.075vol%0.075vol%29.85vol%-

30vol%

1.3

Standard

20 yearsUniform

0.0449kg

1-23

MCCI
Unevendistribution(center)

1.05vol%

1.05vol%

7.2vol%20.7vol%

30vol%

23.0

Standard

20 years

Unevendistribution(Center)

0.615kg

1-26

MCCI
Burn-up (low)1.05vol%

1.05vol%

7.2vol%20.7vol%

30vol%

1.3

Standard

20 yearsUniform

0.615kg

1-27

Molten debris
Worst case

5vol%

5vol%

-

-

10vol%

1.3

Standard

20 years

Unevendistribution(Center)

2.93kgCase(Repeated)(Repeated)

MCCI: Molten Core Concrete Interaction©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1042.2.2 Passive gamma rays - Additional sensitivity analyses (3/4)
Step 2 Analysis (3/4)
Five cases in which the peak count rate at 1.27MeV was less were analyzed taking multiple influencing factors into consideration.

■ MCCI (Molten Core Concrete Interaction) and metallic debris cases in which the burn-up and uneven distribution were changed(Red frame： Step 2 analysis cases)
Case
No.

Type of fueldebrisVariation
parameters

Amount of U*(kg)

Total flux
[γ/s/cm2]

1.27MeV peak[γ/s/cm2]

1-1

Molten debrisStandard

8.79

5.23×106

1.75×105―

1-6

Molten debris

8.97

5.37×100

0.0030%Comparison with 1-1

1-12

Molten debris

8.79

2.61×106

9.08×104

52%

Comparison with 1-1

1-19

MCCI

Burn-up (low)Uneven distribution(center)Standard

3.06×106

0.615

2.90×106

1.92×104

―

1-26

MCCI

Burn-up (low)0.629

2.11×106

2.01×102

1.0%

Comparison with 1-19

1-23

MCCI

Uneven distribution(center)

0.615

2.54×106

8.67×103

45%

Comparison with 1-19

1-17

Metallic debrisStandard

0.04397.53×106

1.52×103―

1-25

Metallic debris

Burn-up (low)0.04497.48×106

5.81×102

38%

Comparison with 1-17

Metallic debris

Uneven distribution(center)

0.0439

7.54×106

1.05×103

69%

Comparison with 1-17

1-21

Proportion of 1.27MeV peak when compared withthe standard case of each type of fuel debris (%)

Remarks*Amount of U = Total mass of Uranium isotopes

Standard

Molten debris
MCCI

Step 2 analysis case

Metallic debris•

The impact of burn-up was small in the case of MCCI and metallic debris
as compared to molten debris→This is because the amount of U contained in MCCI and metallic debris
is less.

•

The impact of uneven distribution was about the same regardless of thetype of fuel debris (around 40 to 70%)
→ Impact of disparity inside the container is higher than that of disparity incomposition

Low burn-up

Uneven distribution(center)©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1052.2.2 Passive gamma rays - Additional sensitivity analyses (4/4)
Step 2 Analysis (4/4)
Five cases in which the peak count rate at 1.27MeV was less were analyzed taking multiple influencing factors into consideration.
■Analysis of sensitivity to filling rate, burn-up, and uneven distribution in the case of molten debris(Red frame: Step 2 analysis case）
Entire debris
Amount of U*(kg)
Variation parameters
(kg)

Case
No.

Type of fuel debris

1-1

Molten debrisStandard

15.8

1-4

Molten debris

Filling rate (low)1-6

Molten debris

Burn-up (low)1-12

Molten debris

1-27

Molten debris

Uneven distribution(center)
Filling rate (low),
Burn-up (low),
Uneven distribution(center)

Total flux
[γ/s/cm2]

1.27MeV peak[γ/s/cm2]

8.79

5.23×106

1.75×105

5.28

2.93

3.40×106

1.03×105

5.90×10-1

15.8

8.97

3.06×105

5.92×102

3.38×10-3

15.8

8.79

2.61×106

9.08×104

5.18×10-1

5.28

2.99

7.28×104

1.42×102

Proportion of 1.27MeV peak when compared with 1-1(-)
ー1.03×10-3 **8.12×10-4*Amount of U = Total mass of Uranium isotopes

Base Filling rate
(low)**Product of 1-1, 1-4 and 1-6

Uneven distribution(center)Burn-up (low)Filling rate (low)Burn-up (low)Unevendistribution(center)
Step 2 analysis
cases

•
•

Count rate of 1.27MeV peak decreased about 3 digits of magnitude ascompared to the base case.
It was about the same as the product of the cases of filling rate, burn-up
and uneven distribution. Thus it is confirmed that all influencing factors
linearly contributed to it.

→ The measurement time is expected to be extremely long based on theanalysis of the detector response in the base case.

©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1062.2.2 Passive gamma rays – Summary➢

Major findings obtained from the analysis:
ImpactInfluencing factor

Findings

Major① Fuel debris
Composition② Burn-up⑥ Filling rate
⑦ Uneven distribution⑧ Container

・Even if only the fuel debris composition and filing factor are changed,
as long as the burn-up remains constant, there is a possibility of
being able to quantitatively evaluate the amount of U based on Eu154, but if the burn-up changes, the correlation between Eu-154
and amount of U deteriorates. [Refer to No. 88]
・As a result of the self-shielding effect due to the shape of thecontainer or the apparent density, the flux that is incident upon thedetector changes. [Refer to No. 93-95]
・If a lot of material other than nuclear fuel is contained, andfurthermore if nuclear fuel is unevenly distributed inside the container,
the flux incident upon the detector changes.

Minor③ FP emission rate
④ Cooling period(⑤ Moisture content*1)②
2 core average②Burn-up:
燃焼度：Unit2号炉心平均③
emission 試験(Phebus-FPT4)ベース
rate: Test (Phebus-FPT4) base
③FP
FP放出率：④
period:20 years④Cooling
冷却期間：20年

・There is minor impact of FP emission rate rate on Eu-154.
・There is a possibility of being able to correct the cooling period bymeans of the retrieval time period and half-life.

①
debris composition:① Fuel
燃料デブリ組成:UO2: 50 (vol%)ZrO2: 50 (vol%)SUS,concrete::0 0(vol%)SUS,(vol%)
B4C, Gd2O3: 0 (vol%)
MOX: 0 (vol.%)
Empty(vol%)
Empty(porosity):(気孔率) : 00(vol%)⑤
content:⑤Moisture
含水率：1(wt%) 1 (wt%)⑥
rate: 30(vol%)(vol%)⑥Filling
充填率：30⑦
Unevendistribution:
None (=uniform)⑦ 偏在：なし (=均質)⑧ Container:
容器：ユニット缶⑧
Unit can(φ210mm×H200mm)*1 Not analyzed as change in the moisture content has only a minor impact on selfshielding due to the penetration power of gamma rays➢

Parameter values in the figure indicatebase case conditions.

Study of the necessity of detailed analysis and consolidation of technical issues：•
Analysis of the extent of impact of heterogeneity on the flux on the measuring surface•
There is a possibility of correcting the rate of change in flux by correcting the impact of the burn-up in fuel debris.
•
If nuclear fuel is unevenly distributed in the container, the method of rotating the container and then carrying out
measurement, or the method of measuring the distribution of material mixed inside the container by means ofradioparency measurement, etc. and then correcting it, etc. are presumed to be effective. These methods are planned
to be studied from next year onwards.
©International Research Institute for Nuclear DecommissioningNo.107

2. Implementation Details(1) Analytical evaluation of factors influencing measurement errors of prospective measurement
technologies①

Selection of influencing factors②

Setting the fluctuation range of the influencing factors③

Analytical simulation
2.1Analysis conditions
2. 2 Analysis results
2.2.1 Passive neutrons
2.2.2 Passive gamma rays
2.2.3 Active neutrons
2.2.4 Passive/active neutrons + passive gamma rays
2.2.5 X-ray transmission measurement
2.2.6 Cosmic rays scattering measurement①
Fuel debris
composition
2.2.3
ActiveneutronNo. 114
No. 128 to 130
No. 143
No. 144Gdcontent

B
content

MOX

No. 145

No. 145

No. 115②
Burn-up③
FP
emissionrate④
Cooling
period⑤
Moisturecontent⑥
Fillingrate⑦
Unevendistribution

No. 122
to 124－

No. 125
to 127

No. 119
to 121

No. 116
to 118
No. 144

No. 133
to 135⑧
Container

No. 146⑨
Irradiatingradiation
source⑩Detector

No. 138
No. 147

No. 140
to 142©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1082.2.3 Active neutrons - Measurement concept⚫

Pulsed neutrons are irradiated from the neutron generator to the fuel debris, and neutrons produced by fissile nuclides (U-235, Pu239, etc.) contained in fuel debris are measured.

⚫

However, as the produced neutron flux is affected by the amount of nuclear material in fuel debris, the abundance of fissile
nuclides attributable to burn-up, neutron absorption material (B, Gd, etc.),etc., there is variation in the neutron flux reaching themeasurement surface.
⚫

Based on the above, the range to be measured was determined by means of analytical evaluation, and issues such as themeasurement range of the detector required for the measurement system,the layout for measurement, etc. wereidentified.

Fast neutrons from the neutron generator
Fast neutrons from the neutron source + neutrons produced by fast fission(Since 2 signals are mixed, the information is not of much use.)

Count

Part to be measured:
Prompt neutrons produced by the fission of U, Pu
caused by thermal neutrons

Fission neutrons

Neutron detector

Neutron tube●: Neutrons
Spontaneous fission neutrons●: Nuclear material(UO2)

Nuclear fuel present

Nuclear fuel absent
Signal from Cm244 + signal from delayed neutrons
Neutron source ON

Neutron source OFF

Neutron counter
Time©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1092.2.3 Active neutrons - Analysis model

Radiation source for pulsed neutrons (Neutron tube ING031 manufactured by Russia's VNIIA is assumed)
Neutron energy: 14.1MeV, Neutron generation rate: 2x108n/s, Neutron generation direction: isotropic,
Pulse width: 1.2µs, Repetitive frequency: 100Hz
In order to scatter / moderate pulsed neutrons, polyethylene was installed around the unit can and the radiation source for pulsed
neutrons.
→ Used for selecting the detector and studying the layout based on the changes in neutron flux (Total energy and 0.4eV or more(assuming that thermal neutrons are absorbed by Cd)) time.
50mm
Polyethylene
Radiation source forpulsed neutrons(Point source)

Radiation source forpulsed neutrons(Point source)

Unit canFuel debris

Fuel debris

125mm

Polyethylene

10mm
200mm
1016mmA’A

Φ210mmUnit canT-cross Tally

Φ250mmAnalysis model top viewΦ350mmCross-sectional view A-A’
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1102.2.3 Active neutrons - Table listing the analysis cases (1/3)••

Considering the analysis conditions described in No. 3-1 as the base case, the analysis conditions that changed
to influencing factors (No. 3-2 to 27) were assumed as the sensitivity analysis conditions. All 27 cases
Influencing factors that changed from the base case under the sensitivity analysis conditions are highlighted
blue. Here, the parameters that changed in association with the above-mentioned changes are indicated byyellow hatching.
Composition inside the container *1

Case No.

Within the filling factor

Burn-upType of fuel debris
Composition
3-1

3-2

Molten debris

UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)UO2 : 7.5vol%(25vol%)

3-4

3-5

3-6

3-7

3-8

3-9

3-10

MOX : 30vol%UO2 : 5vol%(50vol%)ZrO2 : 5vol%(50vol%)UO2 : 25vol%(50vol%)ZrO2 :25vol%(50vol%)UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)UnevenContainer

distribution

Irradiating neutron
sourceChanged parameter

Total (filling factor)
30vol%

30vol%

ZrO2 : 22.5vol%(75vol%)3-3

FP emission rate Cooling period

Outside the filling factor

30vol%

10vol%

50vol%H2O (Moisture content): 1wt%
Empty: Remainder
H2O (Moisture content): 1wt%
Empty: Remainder
H2O (Moisture content): 1wt%
Empty: Remainder
H2O (Moisture content): 1wt%
Empty: Remainder
H2O (Moisture content): 1wt%
Empty: Remainder

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

0GWd/tStandard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

Gd2O3 : 3vol%
30vol%H2O (Moisture content): 1wt%
Empty: Remainder
Gd2O3 : 30vol%

30vol%H2O (Moisture content): 1wt%
Empty: Remainder
B4C : 0.51vol%

30vol%H2O (Moisture content): 1wt%

Unit canD-T reaction(Φ210mm x H200mm)(14.1MeV)

Unit canD-T reaction

Fuel debris composition(Φ210mm x H200mm)(14.1MeV)(UO2, ZrO2)

Unit canD-T reaction(Φ210mm x H200mm)(14.1MeV)

Unit canD-T reaction(Φ210mm x H200mm)(14.1MeV)

Unit canD-T reaction(Φ210mm x H200mm)(14.1MeV)

Unit canD-T reaction(Φ210mm x H200mm)(14.1MeV)

Unit canD-T reaction(Φ210mm x H200mm)(14.1MeV)

Unit canD-T reaction(Φ210mm x H200mm)(14.1MeV)

Unit canD-T reaction(Φ210mm x H200mm)(14.1MeV)

Unit canD-T reaction(Φ210mm x H200mm)(14.1MeV)

Base case

MOX

Filling factor

Filling factor

Gd content

Gd content

B content

Empty: Remainder
B4C : 10vol%
30vol%H2O (Moisture content): 1wt%
Empty: Remainder
H2O (Moisture content):

30vol%

0.1wt%
Empty: Remainder

B content

Moisture content©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1112.2.3 Active neutrons - Table listing the analysis cases (2/3)

Composition inside the container *1
Case No.

Burn-upWithin the filling factor
Type of fuel debris

Outside the filling factor
Composition

3-11

3-12

3-13

3-14

3-15

3-16

3-17

3-18
3-19

3-20

Molten debris

UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)UO2 : 0.48vol%[3.7kg]
SUS : 10vol%(95.4vol%)UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)FP emission
rate

Cooling period

Uneven distribution

Container

Irradiating neutron
sourceChanged parameter

Total (filling factor)
H2O (Moisture content):
30vol%

70vol%

23.0GWd/t

Standard

20 yearsUniform

1.3GWd/t

Standard

20 yearsUniform

51GWd/tStandard

20 yearsUniform

23.0GWd/t

Standard

30 yearsUniform

23.0GWd/t

Standard

40 yearsUniform

23.0GWd/t

Standard

20 years

23.0GWd/t

Standard

20 years

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

Empty : 0vol%
30vol%

30vol%

30vol%

30vol%

30vol%

30vol%

30vol%

10.48vol%
30vol%H2O (Moisture content): 1wt%
Empty: Remainder
H2O (Moisture content): 1wt%
Empty: Remainder
H2O (Moisture content): 1wt%
Empty: Remainder
H2O (Moisture content): 1wt%
Empty: Remainder
H2O (Moisture content): 1wt%
Empty: Remainder
H2O (Moisture content): 1wt%
Empty: Remainder
H2O (Moisture content): 1wt%
Empty: Remainder
H2O (Moisture content): 1wt%
Empty: Remainder
H2O (Moisture content): 1wt%
Empty: Remainder

Unit canD-T reaction(Φ210mm x H200mm)(14.1MeV)

Unit canD-T reaction(Φ210mm x H200mm)(14.1MeV)

Unit canD-T reaction(Φ210mm x H200mm)(14.1MeV)

Unit canD-T reaction(Φ210mm x H200mm)(14.1MeV)

Unit canD-T reaction(Φ210mm x H200mm)(14.1MeV)

Horizontal Direction Unit can(Center)(Φ210mm x H200mm)

Horizontal Direction Unit can(Outer surface)

D-T reaction(14.1MeV)
D-T reaction

Moisture content

Burn-upBurn-up

Cooling period

Cooling period

Uneven distribution

Uneven distribution(Φ210mm x H200mm)(14.1MeV)

Unit canD-T reaction(Φ220mm x H840mm, Thickness 10mm)(14.1MeV)

Waste storage container

D-T reaction(Inner container□Φ500mm x H300mm)(14.1MeV)

Unit canD-D reaction

Irradiating neutron(Φ210mm x H200mm)(2.45MeV)

sourceContainer
Container©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1122.2.3 Active neutrons - Table listing the analysis cases (3/3)
Composition inside the container *1
Case No.

Within the filling factor

Burn-upType of fuel debris

Outside the filling factor
Composition
UO2 : 15vol%(50vol%)3-21

Molten debris

3-22

Uranium-rich

UO2 : 30vol%3-23

Metallic debris

ZrO2 : 0.075vol%(0.25vol%)ZrO2 : 15vol%(50vol%)

30vol%

30vol%

30vol%SUS : 29.85vol%(99.5vol%)UO2 : 15vol%(25vol%)ZrO2 : 15vol%(25vol%)rate

Cooling period

UnevenContainer

distribution

Irradiating neutron source

Changed parameter

Accelerator based

Irradiating radiation

neutron source

sourceTotal (filling factor)

UO2 : 0.075vol%(0.25vol%)3-24

FP emission

60vol%SUS : 30vol%(50vol%)H2O (Moisture content): 1wt%
Empty: Remainder
H2O (Moisture content): 1wt%
Empty: Remainder

H2O (Moisture content): 1wt%
Empty: Remainder

H2O (Moisture content): 1wt%
Empty: Remainder

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

23.0GWd/t

Standard

20 yearsUniform

Unit can(Φ210mm x H200mm)
Unit can(Φ210mm x H200mm)

Unit can(Φ210mm x H200mm)

Unit can(Φ210mm x H200mm)

D-T reaction (14.1MeV)

D-T reaction (14.1MeV)

D-T reaction (14.1MeV)

Fuel debris composition(type)

Fuel debris composition(type)

Fuel debris composition(type)

UO2 : 1.05vol%(3.5vol%)
3-25

MCCI

ZrO2 : 1.05vol%(3.5vol%)

30vol%SUS : 7.2vol%(24vol%)H2O (Moisture content): 1wt%
Empty: Remainder

Unit can(Φ210mm x H200mm)

D-T reaction (14.1MeV)

Fuel debris composition(type)

Concrete : 20.7vol%(69vol%)UO2 : 15vol%(33.3vol%)3-26

ZrO2 : 15vol%(33.3vol%)45vol%

Concrete : 15vol%(33.3vol%)UO2 : 15vol%(25vol%)3-27

ZrO2 : 15vol%(25vol%)Concrete : 30vol%(50vol%)

60vol%H2O (Moisture content): 1wt%
Empty: Remainder

H2O (Moisture content): 1wt%
Empty: Remainder

Unit can(Φ210mm x H200mm)

Unit can(Φ210mm x H200mm)

D-T reaction (14.1MeV)

D-T reaction (14.1MeV)

Fuel debris composition(concrete)

Fuel debris composition(concrete)*1: Percentage inside the container. Percentage, when 100% is considered to be within the filling rate, is mentioned inside parentheses©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1132.2.3 Active neutrons - Analysis results [Base case]➢ Neutrons were irradiated to fuel debris, and it was verified that neutron flux time change due to the nuclear fission caused byfissile nuclides (U-235, etc.) can be obtained through analysis. Here, the neutron flux (after penetration of polyethylene) at themeasurement surface was evaluated.
➢ Further, the flux from spontaneous fissile nuclides (Cm-244, etc.) that cause measurement noise was evaluated as well.

Neutron source:
・ Energy: 14.1MeV・ Rate of incidence: 2×108 n/s
・ Pulse width: 1.2μs
・ Repetitive frequency: 100Hｚ
Fuel debris:
・UO2：15vol%・ZrO2：15vol%
・ Cooling period: 20 years
・ Burn-up: 23GWd/t
・ FP emission rate: Standard
・ Moisture content: 1wt%
・ Container: Unit can*

Neutron flux (n/cm2/sec)

Analysis conditions(Base case)

Fissile nuclides (U-235, etc.) present

steady components (Cm-244, etc.)

Fissile nuclides (U-235, etc.) absent

Time (μs)* Neutron flux energy: >0.4eV©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1142.2.3 Active neutrons – Analysis details [Base case]
50µs to 2,000µs

Evaluated using integrated value

10 5

of ②, integrated value of ③, and

0.4eV or moreTime integrated neutron flux*1) [n/cm2]

10 4① Pulsed fast neutron component

10 3

integrated value of ②+③ at 50µs

to 2,000µs② ‘Prompt + delayed neutron components(Fissile material absent)→ Only neutrons from the radiation source for pulsed
neutrons

10 2② Prompt + delayed neutron components(Fissile material present)
②+③

10 1

Component

Time integrated
neutron flux*2)[n/cm2]10 0②

1.45E+02③

9.47E+00②＋③

1.55E+02

10 -1
10 -2
③ steady component（Cm-244）
10 -3
0

500

1000
Time [µs]

1500

2000

*1) A value 100 times the integrated value between the time bins (10μs) for 1 pulse, considering a repetitive frequency of 100Hz.
*2) A value that is 100 times the integrated value between 50μs and 2000μs for 1 pulse, considering a repetitive frequency of 100Hz.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1152.2.3 Active neutrons – Analysis results [MOX]

In the case of MOX fuel, the components of ② largely varied at an early time (100µs or less), but had almost the

10
10
10

10
10
10
10
10

10

5

4

Component

Time integrated
neutron flux
[n/cm2]②

1.68E+02③

9.59E+00②＋③

1.78E+02②

1.45E+02③

9.47E+00②＋③

1.55E+02

3

MOX
2

MOX

10

1

0

Base case

-1

-2

-3

Base caseTime integrated neutron flux [n/cm2]

same value from 500µs onwards.

The steady component (Cm-244) in the base case, andthe steady component in MOX were almost the same.

-4

0

500

1000

1500

2000

Time [µs]©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1162.2.3 Active neutrons – Analysis results [Filling rate]
The steady component (Cm-244) of ③ almost did not change even if the filling rate (U mass) was changed.

10

5

10

4

10

3

10

2

10

1

10

0

10

-1

10

-2

10

-3

10

-4

Component

Time integrated
neutron flux
[n/cm2]

10%

Base case (30%)
10%Base caseTime integrated neutron flux [n/cm2]

Components of ② largely varied at an early time (100µs or less), but from 500µs onwards, the variation became almost constant.

The steady component (Cm-244) in the base case 30% andthe steady component in 10% were almost the same.

0

500

1000

1500

2000

②

7.59E+01③

9.72E+00②＋③

8.56E+01②

1.45E+02③

9.47E+00②＋③

1.55E+02

Time [µs]©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1172.2.3 Active neutrons – Analysis results [Filling rate]
The steady component (Cm-244) of ③ almost did not change even if the filling rate (U mass) was changed.

10

5

10

4

10

3

10

2

Component

The steady component (Cm-244) in base case 30%

1

10

0

10

-1

10

-2

②

1.90E+02③

9.09E+00②＋③

1.99E+02②

1.45E+02③

9.47E+00②＋③

1.55E+02

50%
50
%

10

Base case (30%)

The steady component (Cm-244) in 50%

10Time integrated
neutron flux
[n/cm2]

Base caseTime integrated neutron flux [n/cm2]

Components of ② largely varied at an early time (100µs or less), but from 500µs onwards, the variation became almost constant.

-3

0

500

1000

1500

2000

Time [µs]©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1182.2.3 Active neutrons – Analysis results [Filling rate]

The steady component (Cm-244) of ③ almost did not change even if the filling rate (U mass) was changed.

Time integrated neutron flux [x 102 n/cm2]

The integrated value of ② was not linear with respect to filling rate (= U mas). It monotonically increased.

3

②
③
②＋③
②
③
②＋③
②
③
②＋③

2

10%

30%
(Base case)
50%

1

0
0

10

20

30

40

50

60

Filling rate [%]©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1192.2.3 Active neutrons – Analysis results [Moisture content]
The components of ② and ③ had almost the same value in the base case (1wt%) and in the case of 0.1wt%.

Time integrated neutron flux [n/cm2]

10
10

5
Component

4

10

3

10

2

10

1

10

0

10

-1

10

-2

10

-3

The steady component (Cm-244) in base case 1wt% and thesteady component in the case of 0.1wt% were almost the
same.

0.1
wt%

Time
integrated
neutron flux
[n/cm2](Reference)
Integrationinterval
30 to 2,000µs②

1.48E+02

1.73E+02③

9.63E+00

9.73E+00②＋③

1.57E+02

1.83E+02②

1.45E+02

1.70E+02③

9.47E+00

9.57E+00②＋③

1.55E+02

1.80E+02

Almost the same in base case and in the case of 0.1wt%

Base
case

0

500

1000

1500

2000

Time [µs]©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1202.2.3 Active neutrons – Analysis results [Moisture content]

When the moisture content was 70vol%, at 80µs or less, the components of ② had a higher value than the base case, but 80µs
onwards the value of the base case increased.

When compared with the integrated value in the interval of 30 to 2000µs, the value for the components of ② was almost the same for

10

5

10

4

10

3

ComponentTimeintegrated
neutron flux
[n/cm2](Reference)
Integrationinterval
30 to 2,000µs②

1.23E+02

1.75E+02③

6.19E+00

6.26E+00②＋③

1.29E+02

1.81E+02②

1.45E+02

1.70E+02③

9.47E+00

9.57E+00②＋③

1.55E+02

1.80E+02

The steady component (Cm-244) in the base case

10

2

10

1

10

0

10

-1

10

-2

10

-3

70vol%

Base case (1.0wt%)

70
vol
%

Base caseTime integrated neutron flux [n/cm2]

the base case and the case with moisture content 70vol%.

The steady component (Cm-244) in the caseof moisture content 70vol%

0

500

1000

1500

2000

Time [µs]©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.3 Active neutrons – Analysis results [Moisture content]

No.121In the case of integration interval 50 to 2,000µs, when the moisture content became 70vol%, the neutron flux reduced,
but if the integration interval was changed to 30 to 2,000µs, there was a possibility of being able to perform measurement

Time integrated neutron flux [x 102 n/cm2]

without the influence of moisture content.

3

0.1wt%

2

1.0wt%

70vol%

②
③
②＋③
②
③
②＋③
②
③
②＋③

0.1wt%

1.0wt%(Base case)70vol%(Reference) Integration interval 30 to 2,000µs

1
②
②＋③

0.1wt%②

1.0wt%(Base case)

②＋③

0
0.01

0.1

1

10

100

②
②＋③

70vol%

Moisture content [vol%]©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1222.2.3 Active neutrons – Analysis results [Burn-up]

Time integrated neutron flux [n/cm2]

The value of steady component (Cm-244) of ③ in the case of 1.3GWd/t was about 4 digits of magnitude smaller than ②.

10

5

10

4

10

3

10

2

10

1

10

0

10

-1

10

-2

10

-3

10

-4

10

-5

Base case (23.0GWd/t)
1.3
GWd/t

1.3GWd/t

Component

Time integrated
neutron flux
[n/cm2]②

5.45E+01③

3.41E-03②＋③

5.45E+01②

1.45E+02③

9.47E+00②＋③

1.55E+02

The steady component (Cm-244) in the base case
Base
case

The steady component (Cm-244) in the case of 1.3GWd/t

0

500

1000

1500

2000

Time [µs]©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1232.2.3 Active neutrons – Analysis results [Burn-up]
The value of steady component (Cm-244) of ③ in the case of 51GWd/t was almost 5 times or more than theTime integrated neutron flux [n/cm2]

base case, but it was about 1/4 when compared with the steady components of ②.

10

5

10

4

10

3

10

2

10
10

Component

The steady component (Cm-244) in the case of 51GWd/t

Base case (23.0GWd/t)
51
GWd/t

51GWd/t

1

Time integrated
neutron flux
[n/cm2]②

1.91E+02③

5.29E+01②＋③

2.44E+02②

1.45E+02③

9.47E+00②＋③

1.55E+02

0

10

-1

10

-2

Base
case

The steady component (Cm-244) in the base case

10

-3

0

500

1000

1500

2000

Time [µs]©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1242.2.3 Active neutrons – Analysis results [Burn-up]

When burn-up increased, the integrated value of the components of ② of the neutron flux monotonically increased, but as the burn-up increased, theproportion of increase of the integrated value reduced.

Time integrated neutron flux [n/cm2]

The integrated value of the components of ③ sharply increased when the burn-up increased, but even in the case of 51GWd/t, it was about 1/4 of ②.

10

10

3

②
③
②＋③
②
③
②＋③
②
③
②＋③

1

10

-1

10

-3

0

10

20

30

40

50

1.3GWd/t

23.1GWd/t(Base case)
51GWd/t60

Burn-up [GWd/t]©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1252.2.3 Active neutrons – Analysis results [Cooling period]

10

5

10

4

10

3

10

2

10

1

10

0

10

-1

10

-2

10

-3

10

-4

The steady component (Cm-244) in the base case of 20
years, and the steady component in the case of 30 year
were almost the same.

Almost the same in the base case and in the case of 30 years

30
years

Base caseTime integrated neutron flux [n/cm2]

The cooling period had almost no impact on the time change pertaining to the components of ② and ③.

0

500

1000

1500

Component

Time integrated
neutron flux
[n/cm2]②

1.43E+02③

9.47E+00②＋③

1.53E+02②

1.45E+02③

9.47E+00②＋③

1.55E+02

2000

Time [µs]©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1262.2.3 Active neutrons – Analysis results [Cooling period]

10

5

10

4

10

3

10

2

10

1

10

0

10

-1

10

-2

10

-3

10

-4

The steady component (Cm-244) in the base case of 20 years,
and the steady component in the case of 40 year were almost
the same.

Almost the same in the base case and in the case of 40 years

40
years

Base caseTime integrated neutron flux [n/cm2]

The cooling period had almost no impact on the time change pertaining to the components of ② and ③.

0

500

1000

1500

Component

Time integrated
neutron flux
[n/cm2]②

1.43E+02③

9.46E+00②＋③

1.53E+02②

1.45E+02③

9.47E+00②＋③

1.55E+02

2000

Time [µs]©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1272.2.3 Active neutrons – Analysis results [Cooling period]

Time integrated neutron flux [x 102 n/cm2]

The cooling period had almost no impact on the integrated value pertaining to the components of ② and ③.

3
②
③
②＋③
②
③
②＋③
②
③
②＋③

2

20 years(Base case)
30 years

40 years

1

0
0

10

20

30

40

50

Cooling period [years]©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1282.2.3 Active neutrons – Analysis results [MCCI]
There was almost no difference in the time change in the components of ② and ③ in the case of MCCI (Molten Core ConcreteTime integrated neutron flux [n/cm2]

Interaction) and molten debris.

10

5

10

4

10

3

10

2

10

1

10

0

There was almost no change in the case of Concrete 15% and
30% and the base case.

Concrete
15%

Concrete
30%

Component

Time integrated
neutron flux
[n/cm2]②

1.42E+02③

9.24E+00②＋③

1.51E+02②

1.39E+02③

9.00E+00

10

-1

②＋③

1.48E+02

10

-2

②

1.45E+02

10

-3

③

9.47E+00②＋③

1.55E+02

Base case

0

500

1000

Time [µs]

1500

2000

©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.3 Active neutrons – Analysis results [MCCI]

No.129When the amount of UO2 and ZrO2 was the same, there was almost no difference in the time change pertaining to the components of ② and ③,
even if the amount of concrete changed.
When the amount of UO2 and ZrO2 was different, there was almost no difference in the time change pertaining to the components of ③, but in theTime integrated neutron flux [n/cm2]

case of components of ②, when the amount of UO2 and ZrO2 was approx. 1/15, the integrated value of the neutron flux was approx. 1/7.

Time integrated
neutron flux
[n/cm2]10

5Component

10

4

②

2.11E+01103③

9.18E+00

10

2

②＋③

3.02E+0110

1②

1.42E+02③

9.24E+00②＋③

1.51E+02②

1.39E+02③

9.00E+00②＋③

1.48E+02

10

Concrete
20.7%

Concrete 20.7%(UO2:1.05wt%,ZrO2:1.05wt%)
Concrete
15
%

0

10

-1

10

-2

10

Concrete 15% and 30%(UO2:15wt%,ZrO2:15wt%)

Concrete
30%

-3

0

500

1000

1500

2000

Time [µs]©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.3 Active neutrons – Analysis results [MCCI]

No.130

There was almost no difference in the integrated value of the components of ② and ③ in the case of MCCI (Molten CoreTime integrated neutron flux [x 102 n/cm2]

Concrete Interaction) and molten debris.

3
②
③
②＋③
②
③
②＋③
②
③
②＋③

2

Base case
Concrete: 15%

Concrete: 30%

1

0
0

10

20

30

40

50

Concrete [%]©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1312.2.3 Active neutrons – Sensitivity analyses [Uneven distribution model (A)]

Tally
(Region divided equally into 6 parts)

50mm
Radiation source forpulsed neutrons(Point source)

Radiation source for pulsed neutrons(Point source)

Polyethylene

Unit can(3)
Fuel debris(2)

Polyethylene

Fuel debris(Diameter 155mm)

125mm
10mm

200mm

A

(1)

(4)

(5)

(6)

A’

1016mmUnit canT-cross Tally

Analysis model top view

Φ210mm
Φ250mm

Φ350mmCross-sectional view A-A’

©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1322.2.3 Active neutrons – Sensitivity analyses [Uneven distribution model (B)]

Tally
(Region divided equally into 6 parts)
Radiation source for pulsed neutrons(Point source)(3)

Fuel debris

50mm

Radiation sourcefor pulsed neutrons(Point source)

Polyethylene

Unit can(2)

Polyethylene

Fuel debris(Diameter 155mm)

125mm
10mm

200mm

A

(1)

(4)

(5)

(6)

A’

1016mmUnit canT-cross Tally

Analysis model top view

Φ210mm
Φ250mm

Φ350mmCross-sectional view A-A’

©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.3 Active neutrons – Sensitivity analyses [Uneven distribution model (A)]No.133The change in the component (Cm-244) of ③ resulting from uneven distribution was 10% or less. However, the components of ②

Time integrated neutron flux [n/cm2]

largely changed early on (100µs or less) but the change was small 500µs onwards.

10

5

10

4

10

3

10

2

10

1

10

0

10

-1

10

-2

10

-3

10

-4

Component

Time integrated
neutron flux
[n/cm2]②

8.07E+01③

8.86E+00②＋③

8.95E+01②

1.45E+02③

9.47E+00②＋③

1.55E+02

The steady component (Cm-244) in the base case
Unevendistribution
model

Base case

Uneven distribution model ①

Base case

The steady component (Cm-244) in uneven distribution model ①

0

500

1000

Time [µs]

1500

2000
In the uneven distribution model (A), thevalues of (1) to (6) were all the same.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.3 Active neutrons – Sensitivity analyses [Uneven distribution model (B)]The change in the steady component (Cm-244) of ③ due to uneven distribution was

10% or less. However, the components of ② largely changed at an early time (100µs orless) but the change was small 500µs onwards.
5

10

4

10

3

10

2

10

1

10

(1)

Uneven distribution model (B)Time integrated neutron flux [n/cm2]

10

Base case ②+③(1) ②+③(4) ②+③

0

10

-1

10

-2

(2)

(3)

(4)

(1) ③
Mean

Base case (4) ③
10

-3

0

500

1000

Time [µs]

1500

2000

Base case

No.134Component

Time integrated
neutron flux
[n/cm2]②

8.49E+01③

1.03E+01②＋③

9.52E+01②

7.97E+01③

9.33E+00②＋③

8.90E+01②

7.28E+01③

8.14E+00②＋③

8.10E+01②

7.04E+01③

7.77E+00②＋③

7.82E+01②

7.67E+01③

8.84E+00②＋③

8.56E+01②

1.45E+02③

9.47E+00②＋③

1.55E+02©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.3 Active neutrons – Sensitivity analyses [Uneven distribution models]No.135

There was a difference of less than 5% in the average value of the integrated values of the components of ② in uneven distribution

Time integrated neutron flux [x 102 n/cm2]

models (A) and (B), but in the base case it was approx. double the values of the uneven distribution models (A) and (B).

2
②
②＋③

1

0
Base
case

(A)

(B)
Average

(B)
(1)

(B)
(2)

(B)
(3)

(B)
(4)
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1362.2.3 Active neutrons – Other sensitivity analyses results

Accelerator based neutron source

Repetition

Pulsed bremsstrahlung
Pulsed bremsstrahlung
Target for generating neutrons(Be : 5mm thick)Intensity

Pulsed electron beam
Energy: 6MeV
Average current value: 100µA
Pulse width: 4µs
Repetitive frequency: 100Hz

Intensity

Pulsed neutrons

Pulsed neutrons

Repetition

Target for generating bremsstrahlung (W: 4mm thick)

10ms

Pulsed electron beam

4µs
Average neutron energy: 1.13MeV

Repetition

Intensity

Rate of neutron generation: 4.5×108 n/s

Time©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1372.2.3 Active neutrons – Other sensitivity analyses results

Accelerator based neutron source
Neutron generation pulse width: 4µs
Rate of neutron generation: 4.5×108 n/s
Average neutron energy: 1.13MeV

Radiation sourcefor pulsed neutrons(Point source)

50mm

Radiation sourcefor pulsed neutrons(Point source)

Unit canFueldebrisFueldebrisPolyethylene

125mm

Polyethylene

10mm
200mm

A

A’

1016mm

Φ210mmUnit canT-cross Tally

Φ250mmAnalysis model top viewΦ350mmCross-sectional view A-A’
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1382.2.3 Active neutrons – Other sensitivity analyses results

Rate of pulsed neutron generation: 4.5×108 n/s: 2.0×108 n/s(When it is the same as the base case)
The integrated value of the components of ② per unit of the rate of neutron

Accelerator based neutron source

(1)
(2)

10

5

10

4

10

3

10

2

10

1

10

0

10

Base caseAccelerator based neutron source

Time integrated neutron flux [n/cm2]

generation in the case of accelerator based neutron source was approx. 3 times
that of the base case.

Accelerator based neutron source

(1)

(2)

Component

Time integrated
neutron flux
[n/cm2]②

9.15E+02③

9.47E+00②＋③

9.25E+02②

4.07E+02③

9.47E+00②＋③

4.16E+02②

1.45E+02③

9.47E+00②＋③

1.55E+02

-1

10

-2

10

-3

Base case

0

500

1000

1500

2000

Time [µs]©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1392.2.3 Active neutrons – Other sensitivity analyses results

Response spectrum for a representative detector (System)
50mm

Unit canFuel debris

Polyethylene

Fuel debris

Polyethylene
200mm

A

Neutron radiation source
Analysis model top view

1016mmA’

B-10 detector
Unit canOR
Cd(2mm
He-3 detector
thick)

Cd(2mmthick)

Φ210mm
Φ250mm

Φ350mmCross-sectional view A-A’

B-10 detectorOR
He-3 detector©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1402.2.3 Active neutrons – Other sensitivity analyses results
Response spectrum for a representative detector (Energy spectrum of the neutron radiation source)
Energy spectrum of the neutrons on the side of the outer cylinder of Cd in the base case(Integrated value at the interval of 50µs to 2,000µs)

Cd cut neutrons with energy lower than
0.4eV.

©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1412.2.3 Active neutrons – Other sensitivity analyses results
Response spectrum for a representative detector (B-10 detector)
B-10 detector: Diameter (inner diameter) 25.4mm, effective length: 1000mm

Housing: 0.5mm thick SUS304

B-10 thickness 0.8547µm (=0.2mg/cm2)
Ar gas pressure 0.3atm

Count rate in the base case

Alpha rays
Alpha rays and Li-7 that impart 80keV or

Li-7

more energy were measured.
→ 26.2cps
Time required for getting a 10,000 count: 382 seconds©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1422.2.3 Active neutrons – Other sensitivity analyses results

Response spectrum for a representative detector (He-3 detector)
He-3 detector: Diameter (inner diameter) 25.4mm, effective length: 1000mm

Housing: 0.5mm thick SUS304

He-3 pressure 4atm

Count rate in the base caseProtons and tritium that impart 80keV ormore energy were measured.

→ 326.7cps
Time required for getting a 10,000 count: 31 seconds

Protons

Tritium©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1432.2.3 Active neutrons- Analysis results [Sensitivity Analysis]
Sensitivity to the amount of Uranium (Influencing factor ①)
・ The amount of Uranium and neutron flux has a proportional relationship.
・ When the composition largely differed from molten debris such as in the case of metallic debris, etc.,
since the deceleration and level of thermalization of neutrons generated by the source for pulsed
neutrons differed, the trend differed.

Time integrated neutron flux [n/cm 2]✓(Uranium rich)(Base case)(Metallic debris)

Weight of Uranium (g)©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1442.2.3 Active neutrons- Analysis results [Sensitivity Analysis]
Sensitivity to apparent density (Influencing factors ① and ⑥)
・ The difference in apparent density of fuel debris had a small impact
on neutron flux.

(There was slight difference since the deceleration and level ofthermalization of neutrons differs depending on the fuel debris
composition.)
Time integrated neutron flux [n/cm 2] / Weight of U [g]

✓

Case
No.

Type offuel debris

UO2

ZrO2

SUS

Concrete

Filling rate

3-1

Moltendebris15vol%(50vol%)15vol%(50vol%)

0vol%

0vol%

30vol%

3-22

Uraniumrich

30vol%

0vol%

0vol%

0vol%

30vol%

3-24

Metallic
debris15vol%(25vol%)15vol%(25vol%)30vol%(50vol%)0vol%

60vol%(Base case)(Uranium rich)(Metallic debris)

Apparent density (g/cc)©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1452.2.3 Active neutrons- Analysis results [Sensitivity Analysis]✓

Sensitivity to the neutron absorption material (Influencing factor ①)
・ The neutron flux reaching the surface of the detector reduced due to neutron absorption material such as Gd, B, etc.
・ The larger the amount of neutron absorption material introduced, the more the reduction in neutron flux.
(Almost all the neutron flux (③) was from the steady components (Cm-244, etc.) Particularly in the case of Gd.)

<Boron>Time integrated neutron flux [n/cm 2]

Time integrated neutron flux [n/cm 2]<Gadolinium>

Gd2O3 percentage (vol%)(Base case)

B4C percentage (vol%)©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1462.2.3 Active neutrons- Analysis results [Sensitivity Analysis]
Sensitivity to the shape of container (Influencing factor ⑧)

・ When the shape of the container was different, the amount of Uranium and neutron flux were not proportionate due to the impact of themeasurement system.
(In the case of containers other than unit cans, a large number of neutrons from the source for pulsed neutrons leaked outside the system.)Time integrated neutron flux [n/cm2] / Weight of U [g]✓(Unit can)(Canister)(Waste container)

Weight of U (g)Source for pulsed
neutrons

Source for pulsed
neutrons

Source forpulsedneutrons

Unit canCanister
Waste container(Inner container)©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1472.2.3 Active neutrons- Analysis results [Sensitivity Analysis]✓

Sensitivity to the energy of neutrons from the accelerator based neutron source (Influencing factor ⑨)

Time integrated neutron flux [n/cm 2] / Weight of U [g]・ The lower the energy of the irradiated neutrons, higher was the neutron flux.
・ This was because, when the neutron energy was low, together with the deceleration
effect of polythylene, the neutrons decelerated and thermalized thereby inducing thermal
neutron fission.
The cross-sectional area of H which forms polyethylene showed a difference of approx. 4times between the case of 1MeV and 10MeV.
→ There is a possibility of increasing the neutron flux by improvising the deceleration
system.

1MeV 10MeV(Average 1.13MeV)

Neutron energy (MeV)©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1482.2.3 Active neutrons – Summary➢

Major findings obtained from the analysis:
ImpactInfluencing factor

Major① Fuel debris
composition:(Neutron absorption
material)② Burn-up⑦ Uneven distribution⑨ Irradiating radiation
sourceMinor(③ FP emission rate*1)④ Cooling period⑤ Moisture content⑥ Filling rate:⑧ Container

Findings
・When the amount of neutron absorbing agent
increases, the amount of thermal neutrons in the unit candecreases and the nuclear fission is suppressed. [Refer
to No. 145]
・If the burn-up increases, the amount of U-235 in fueldebris decreases, and since Cm-244, etc. that becomes asource of noise increases, sensitivity reduces.[Refer to
No. 122 - 124]
・If the size of debris is around the same, the impact ofuneven distribution is small, but if the debris is small,sensitivity reduces.
・There is a possibility of being able to correct the impactof the energy of irradiated neutrons on sensitivity by
optimizing the system.
・ Cooling period has a minor impact.
・ There is a possibility of being able to minimize the impact
of moisture content.
・ The filling rate (apparent density) has a minor impact.
・ There is a possibility of being able to correct the shape ofthe container.

*1 ③ FP emission rate is the emission rate of gamma rays and does nothave an impact on the measurement of neutron radiation.

② Burn-up:
燃焼度： Unit
2号炉心平均2 core average③ FP
FP放出率：
試験(Phebus-FPT4)ベース
emission
rate: Test (Phebus-FPT4) base
④ Cooling
冷却期間：20年20 years
period:⑨ Irradiatingradiation source
Neutron 14MeV① Fueldebris composition:
燃料デブリ組成:UO2: 50 (vol%)ZrO2: 50 (vol%)SUS,(vol%)
SUS,concrete::00(vol%)B4C, Gd2O3: 0 (vol%)
MOX: 0 (vol.%)
Empty(porosity):(vol%)
Empty(気孔率)0: 0(vol%)⑤ Moisture content: 1 (wt%)⑤ 含水率：1 (wt%)⑥ Filling rate: 30 (vol%)⑥Uneven
充填率：30(vol%)None⑦
distribution:(=uniform)⑧ Container: Unit can⑦ 偏在：なし (=均質)⑧ 容器：ユニット缶(φ210mm×H200mm)

Parameter values in the figure indicate base case conditions.

➢

Study of the necessity of detailed analysis and consolidation of technological issues•
Combination of different measurement techniques (prompt gamma rays analysis, etc.) needs to be studied forcorrecting the influence of neutron absorption material contained in fuel debris.
•
There is a possibility of correcting the rate of change in flux by correcting the impact of the burn-up in fuel debris.

©International Research Institute for Nuclear DecommissioningNo.149

2. Implementation Details(1) Analytical evaluation of factors influencing measurement errors of prospective
measurement technologies① Selection of influencing factors② Setting the fluctuation range of the influencing factors③ Analytical simulation
2.1Analysis conditions
2. 2 Analysis results
2.2.1 Passive neutrons
2.2.2 Passive gamma rays
2.2.3 Active neutrons
2.2.4 Passive/active neutrons + passive gamma rays
2.2.5 X-ray transmission measurement
①
②
③
④
⑤
⑥
⑦
⑧
2.2.6
Cosmicrays
scattering
measurement
Fuel debris
Gd
B
MOX
Burn-upFP
Cooling
Moisture
Filling
UnevenContainer
2.2.4
Combination

composition

content

content

No. 156
No. 157

No. 157－－

No. 153
No. 157

emission
rate

Periodcontent

rate

distribution

－

－

－

No. 157

No. 156
No. 157－⑨
Irradiatingradiation
source⑩Detector－－©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1502.2.4 Passive/active neutrons + passive gamma rays - Measurement concept⚫

Fuel debris contains gamma ray emitting nuclides originating from fuel (Eu-154, etc.), and nuclides originating from fuel that emit
neutrons (Cm-244, etc.) due to spontaneous fission. Hence neutron beams and gamma rays that are passive and originate fromfuel can be measured. Also, by irrdiating neutrons, since neutrons from the fissile nuclides (U-235, etc,) contained in fuel debris are
emitted, neutron beams that are active and originate from fuel can be measured.

⚫

However, it was confirmed through analytical evaluation (2.2.1, 2.2.2, 2.2.3) of each measurement that measured values varied
due to influencing factors, and such conditions (influencing factors and their fluctuation range) made measurement difficult.

⚫

Based on the above, analysis was conducted to find out whether or not measurement would be possible by combining
measurement techniques to compensate for these conditions that render measurement difficult.

Variations in measured values due to influencing factors(Uncertainty in the estimation of the amount of nuclear material)

Passive neutrons
Fission neutrons＋

Neutron detectors

Passive gamma rays

Neutron tube
γ rays detector

Spontaneous fission neutrons

Active neutrons

Neutron detector●: Neutrons: Gamma rays
Passive neutron measurement +Gamma rays measurement●: Neutrons●: Nuclear material (UO2)
Active neutron measurement

Figure Basic concept of the measurement principle for the
Passive/active neutron beam measurement + γ rays measurement technology

Combination
measurement

Less

More

Quantity of nuclear material (estimated value)
Figure Image illustrating approach towards sorting out the
combination of measurement techniques©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1512.2.4 Passive/active neutrons + passive gamma rays – Concept of analysis<Purpose>⚫

Based on past results of analyzing individual measurement techniques, ”Burn-up”, Self-shielding”, and“Neutron absorption material” were focused on as influencing factors due to which the variation inmeasured value increases. The possibility of compensating for the variation in measured values due tothese influencing factors by combining measurement techniques was verified.
Findings based on results of analyzing stand-alone measurement technologies:
Factors with a significant impact
Measurement
technique

Passive γ

Findings

・When burn-up changes the correlation between Eu-154
and amount of U deteriorates.
・As a result of the self-shielding effect , the flux that isincident upon the detector changes.

Passive n・When burn-up changes the correlation between Cm-244
and amount of U deteriorates. (Flux changes by the orderof magnitude)(・Moisture content has a minor impact(Note 1))

Active n・When burn-up increases, sensitivity reduces.
・Fission gets suppressed due to the neutron absorption
material.

Item

Influencing factor

Burn-up② Burn-up

State of fuel debris in which variations inmeasured values are presumed to be large
Measured value: Low

Measured value:
High① Fuel debris
composition⑥ Filling rate
⑦ Uneven
distribution⑧ Container

Low burn-up (Eu154: small
amount) + major selfshielding (Fuel debris
composition: metallic debris +uneven distribution: center)

Burn-up② Burn-up

Low burn-up( Cm244: Small quantity)

High burn-up(Cm244: Large
quantity)

Burn-up② Burn-up

Neutronabsorption
material① Fuel debris
composition(Gd, B content)High burn-up + major neutron
absorption (Gd, B: Large
quantities)―

Self-shielding of
gamma rays

High burn-up(Eu154: large
amount)(Note 1) Impact is considered to be minor as based on the analysis of sensitivity to moisture content in active neutrons, the variation in steadycomponents (Cm-224 of the passive neutrons) is small in systems wherein polyethylene has been provided©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.4 Passive/active neutrons + passive gamma rays – Findings based on the
results of past analyses⚫

No.152

Based on results of past analyses, the possibility of compensating for the variation caused by individual influencing
factors by combining 3 measurement techniques, was inferred. (Table below)
Rough comparison of the extent of variation of the influencing factors (Note 2) and thepossibility of compensating for it

Conditions considered as factors having a significant impact (Note 1)(Note 3)

Neutron absorption
material

Self-shielding of gamma rays

Passive γ

Analyzed cases

Burn-up
Fuel debris
composition

Filling rate

Unevendistribution

Gd, B content

MediumMoltenMedium--Low

MoltenMedium--Base case

MoltenMedium--Medium

Metallic

Medium--MediumMCCI

Medium--MediumMoltenLeast-

Possibility ofcompensating

Variation

Possibility ofcompensating

Variation

Possibility ofcompensating

-

〇

-

〇

-

〇

Medium〇(Since BG
component
is small)

Major

x
Major

☆
Sensitivity analysis (Fuel debris
composition)(Note 4)

Active n

Variationx
Sensitivity analysis (Burn-up)(Note 4)

High

Passive n☆

△

Medium

△

Minor

〇

Minor

〇

Medium

△

Minor

〇

Minor

〇

-

Sensitivity analysis (Filling rate)(Note 5)

Minor

〇

Minor

〇

Minor

〇

Sensitivity analysis (Uneven distribution)(Note 5)

Minor

〇

Minor

〇

Minor

〇

Minor

〇

Major

x

MediumMoltenMediumCenter-MediumMoltenMedium-Maximum

Minor
Sensitivity analysis (Gd, B content)(Note 5)〇(Not evaluated but isdetermined qualitatively)(Note 1) Base case conditions are indicated in black. Conditions that have changed from the base case are indicated in blue. Note that, conditions other than the factors mentioned here are all base caseconditions.
(Note 2) When changes in flux per unit of the amount of U are about double or half: Small variation, about less than 1 order of magnitude: Moderate variation, 1 order of magnitude or more: Large variation(Note 3) 〇： Sufficient flux reaches the measurement surface, there is little variation and measurement is possible, △： Flux changes about 1 order of magnitude or less but can be corrected, etc., ×： Flux reduces
extremely and hence measurement is difficult,☆： Sufficient flux reaches the measurement surface, but based on the information obtained using the passive γ・n measurement technique, a mutually complementary relationship is required for estimating
nuclear material.
(Note 4) Since conditions that lead to an increase in variations differ depending on the measurement technique, multiple conditions are set.
(Fuel debris composition: Metallic debris, MCCI debris, Burn-up: Low, high)(Note 5) Only the conditions in which variations increase and flux decreases are consolidated. (Filling rate: least, Uneven distribution: center, Gd, B content: greatest)©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.4 Passive/active neutrons + passive gamma rays – Findings based on the
results of past analyses

No.153➢ Measures in response to variations due to burn-up
The fact that the tendency of Cm-244 (passive neutrons) and Eu-154 (passive gamma rays) to increase differs
was focused on.
→ Based on the ratio of Cm-244 / Eu-154, measurement error due to burn-up can be corrected

Passive neutrons (Cm-244)

Passive gamma rays (Eu-154)

Ratio

*

* Ratio when flux in the case of
average core burn-up (23Gwd/t) is
assumed to be 1.

Burn-up (GWd/t)©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.4 Passive/active neutrons + passive gamma rays – Analyzed cases⚫

No.154The combination of factors having a large impact according to the results of past analyses, was studied andit was decided to verify the possibility of compensating by analyzing the following 3 cases (in the blue frame)

as passive/active neutrons + passive gamma rays.

Conditions considered as factors having a significant impact (Note 1)
Neutronabsorption
material

Self-shielding of gamma rays
Burn-up
Fuel debris
composition

Filling
rate

Unevendistribution

Gd, B
content

Policy for setting up the analysis conditions

Medium

Metallic

MediumCenter-Verification of the possibility of compensating for “self-shielding” (large variation and low flux when the passive γ
measurement technique is used) by means of the passive n measurement technique .

Low

Metallic

MediumCenter-Verification of the possibility of compensating for “burn-up” and “self-shielding” (large variation and low flux when
the passive γ・n measurement technique is used) by means of the active n measurement technique.

High

MoltenMedium-Maximum

Verification of the possibility of compensating for “neutron absorption material” and “burn-up” (large variation andlow flux when the active n measurement technique is used) by means of the passive γ and passive n measurement
technique .

Low

Metallic

Least

CenterMaximum

Verification of the possibility of compensating for “burn-up”, “self-shielding”, and “neutron absorption material”(large variation and low flux when 3 measurement techniques are used) by means of a combination of the passive γ・n and active n measurement technique.

(Note 1) Base case conditions are indicated in black. Conditions that have changed from the base case are indicated in blue. Note that, conditions other than thefactors mentioned here are all base case conditions.

Since the condition in this case is high burn-up, it is presumed that the factor can be compensated by means of thepassive γ, n measurement method. Hence this case has been excluded from the analysis.

©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1552.2.4 Passive/active neutrons + passive gamma rays - Analyzed cases••

Analytical simulation models using individual measurement methods were used.
Considering the analysis conditions described in No. 4-1 as the base case, the analysis conditions withchanged influencing factors (No. 4-2 to 4) based on the policy for setting up analysis conditions mentioned on
the previous page, were assumed as the sensitivity analysis conditions. All 3 cases(Base case has been analyzed using analytical simulation for individual measurement techniques)Composition inside the container*1
Case No.

Measurement
technique *1

Within filling factor

Passive γ
Passive η

Molten debris

Active η

4-2

UO2 : 15vol%(50vol%)ZrO2 : 15vol%(50vol%)ZrO2 : 0.075vol%(0.25vol%)Passive η

30vol%

30vol%SUS : 29.85vol%(99.5vol%)Passive γ
4-3

Passive η
Active η

UO2 : 0.075vol%(0.25vol%)
Metallic debris

ZrO2 : 0.075vol%(0.25vol%)

30vol%SUS : 29.85vol%(99.5vol%)Active η

ZrO2 : 0.025vol%(0.25vol%)SUS : 9.95vol%(99.5vol%)H2O (Moisture content): 1wt%
Empty: Remainder

H2O (Moisture content): 1wt%
Empty: Remainder

23.0GWd/t

Standard

20 years

23.0GWd/t

Standard

20 yearsStandard

20 years

Uneven distribution

Container

10vol%

1.3GWd/t*3

Empty: Remainder

23.0GWd/t

H2O (Moisture content): 1wt%
Empty: Remainder

Uniform

Horizontal direction

H2O (Moisture content): 1wt%

Gd2O3 : 30vol%UO2 : 0.025vol%(0.25vol%)4-4

periodIrradiating neutron
sourceChanged parameter

Total (Filling factor)

UO2 : 0.075vol%(0.25vol%)Passive γ

Coolingrate

Outside filling factor
Composition

4-1

FP emission

Burn-upType of fuel debris

1.3GWd/t*3
23.0GWd/t(center)(center)

Horizontal direction
20 years(center)

D-T reaction(Φ210mm x H200mm)(14.1MeV)

Unit can(Φ210mm x H200mm)

Horizontal direction

Standard

Unit can-Base case

Fuel debris composition (type)
Uneven distribution

Unit canD-T reaction

Fuel debris composition (type)(Φ210mm x H200mm)(14.1MeV)

Burn-up, Uneven distribution

Unit canD-T reaction

Filling factor, Gd content,(Φ210mm x H200mm)(14.1MeV)

Moisture content, Burn-up,

Fuel debris composition (type)

Uneven distribution*1 : Not analyzed here, if there are prospects of being able to conduct measurement using other measurement techniques, or if it is believed that qualitative measurement isdifficult, based on sensitivity analysis of measurement techniques.
*2 : Percentage inside the container. Percentage, when 100% is considered to be within the filling rate, is mentioned inside parentheses.
*3 : UO2 and ZrO2 are assumed to have low burn-up (1.3GWd/t), and SUS is assumed to have base burn-up (23.0GWd/t).

©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1562.2.4 Passive/active neutrons + passive gamma rays - Analysis results<Metallic debris unevenly distributed at the center (Cases in which flux measured using the passive γ
measurement technique decreased) [Analysis No. 4-2]><Passive γ><Passive n>
Relation between the weight of Uranium and the entire flux (log-log graph)
The peak of Eu-154 was barely
visible when metallic debris isunevenly distributed.

Flux (γ/cm2/s)

Entire flux (n/cm2/sec)

Other common conditions• Burn-up 23GWd/t• FP emission rate: standard• Moisture content 1wt%
• Homogeneous model• Container: Unit can

Unevendistribution: near
Unevendistribution: center

No. 2-19 (Metallic debris: homogeneous)

Entire flux

Uneven distribution: far

There is a possibility of compensating

Energy (MeV)

Weight of U (g)(Note) As it was confirmed that measurement is possible with passive n, the analysis of active n was omitted.

For cases when measurement of nuclear material is expected to be difficult with passive γ (self-shielding
effect: large), the possibility of compensating by measuring nuclear material originating from fuel using
the passive n measurement method will be verified.

©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.4 Passive/active neutrons + passive gamma rays - Analysis results

No.157<Metallic debris unevenly distributed at the center and having a low burn-up (Cases in which flux measured using the passive γ・nmeasurement technique decreased) [Analysis No. 4-3]><Metallic debris unevenly distributed at the center and having a low burn-up (Large quantity of Gd)(Cases in which flux measured using the passive γ, passive n and active n measurement technique decreased) [Analysis No. 4-4]><Active n>* Due to the low burn-up, it is difficult to measure nuclear material originating fromfuel by measuring passive γ and passive n. (Based on past analysis results)

Time integrated neutron flux [n/cm 2]

No. 3-22 (Uranium-rich)
No. 3-1 (Base case)No. 4-3 (Low burn-up, uneven distribution at the center, metallic debris)

Flux is extremely low.
No. 4-4 (Low burn-up, uneven distribution at the center, metallic debris, large quantity of Gd)

Weight of U (g)The measured value reduced substantially because of suppression of fission due to small
quantities of nuclear material and due to the presence of neutron absorption material.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1582.2.4 Passive/active neutrons + passive gamma rays – Summary➢

Major findings obtained from the analysis:•

When individual measurement techniques are used individually, “burn-up”, “self-shielding” and“neutron absorption material” are the influencing factors causing measurement errors.

•

When 3 measurement techniques are combined, it was confirmed that there is a possibility of reducing
measurement errors (reduction in variation in the measured values).
Conditions under which measurement becomes difficult(Factors with a significant impact)

Measurement technique
Passive n

Passive γ

Active n

Burn-up (low)

x

x

〇

Burn-up (high)

☆

☆

△

Self-shielding of gamma rays

Fuel debris
composition,
filling rate,
uneven distribution,
etc.

〇

△

〇

〇

Neutron absorption material

Gd, B content

〇

〇

x

〇

Item

Influencing factor

Burn-up

Combination〇〇： Sufficient flux reaches the measurement surface, there is little variation and measurement is possible, △： Flux changes about 1 order ofmagnitude or less but can be corrected, etc., ×： flux reduces extremely and hence measurement is difficult.
☆： Sufficient flux reaches the measurement surface, but based on the information obtained using the passive γ・n measurement technique, amutually complementary relationship is required for estimating nuclear material.

©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.4 Passive/active neutrons + passive gamma rays – Summary

No.159➢ Study of the necessity of detailed analysis and consolidation of technical issues•

The technique for correcting the influence of factors that have a significant
impact needs to be studied (Following are typical examples).
Burn-up: Correction based on the difference in the tendency of Eu=154(passive γ) and Cm-244 (passive n) to increase.
Neutron absorption material: Study of combination of different measurement
techniques (prompt gamma rays analysis, etc.)•

Based on the results of analyses conducted this year, analytical simulation
including the detector model will be performed with the purpose of developing
the concept of the equipment.

©International Research Institute for Nuclear DecommissioningNo.160

2. Implementation Details(1) Analytical evaluation of factors influencing measurement errors of prospective measurement
technologies①

Selection of influencing factors②

Setting the fluctuation range of the influencing factors③

Analytical simulation
2.1Analysis conditions
2. 2 Analysis results
2.2.1 Passive neutrons
2.2.2 Passive gamma rays
2.2.3 Active neutrons
2.2.4 Passive/active neutrons + passive gamma rays
2.2.5 X-ray transmission measurement
2.2.6 Cosmic rays scattering measurement①
Fuel debris
composition

2.2.5
X-ray

No. 166
No. 167

Gd
content

B
content

MOX－－

No. 166
No. 167②
Burn-up③
FP
emissionrate④
Cooling
Period⑤
Moisturecontent⑥
Filling
rate

Unevendistribution⑦－

No. 166
No. 167－－

No. 166
No. 167

No. 166
No. 167⑧
Container

No. 166
No. 167⑨
Irradiatingradiation
sourceNo. 166
No. 167⑩
Detector

No. 169
to 172©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.5 X-ray transmission measurement - Measurement concept•

No.161

Measurement principle: From amongst the X-rays that are irradiated while rotating the debris, the x-rays thatpenetrate the debris are measured, and a tomographic image is produced by means of computational image
reconstruction using the data measured in all directions.
•

Measurement system: Debris is placed between the accelerator based X-ray source and the detector tally•

Radiation source conditions: X-ray spectrum calculated from the EGS*•

Detector response: Total amount of X-ray energy for every pulse irradiated by the accelerator based X-ray source•

Measurement time: 10 to 15 seconds per image

Object to be tested

Detector

X-ray beam
X-ray source(accelerator)
Rotating table

In the radioparency measurement equipment,
the object to be measured placed on the rotatingtable is placed in such a way that the accelerator
based x-ray source and the x-ray detector are on
either sides, and by irradiating X-rays whileturning around the rotating table once,
radioparency data in all directions with respect tothe measured object is obtained. A tomographic
image including the inside of the object being
measuredis
producedbyperforming
computational image reconstruction using thisradioparency data.

Structure of the radioparency measurement (Highenergy X-ray CT method) equipment*Particle Transport Code： Electron Gamma Shower©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1622.2.5 X-ray transmission measurement – Analysis conditions
Analysis conditions (1/3)

Analysis model

Flux evaluation system for radioparency measurement (High energy X-ray CT method)

Unit [mm]
L: Distance from the rotation center up to
the front of the detector

PHITS simulation space

X-ray source
Rotating tableyz

Unit canx

Φ210
Detector* Assuming that concrete is placed in front of the detector, the spatial width in the direction of the Z axis is set inaccordance with the scanning pitch (1mm).
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.5 X-ray transmission measurement – Analysis conditions

No.163

Analysis conditions (2/3) Flow of analysis

PHITS simulation system
Bird’s eye view

Flow of analysis
A: Flux evaluation for all cases

L
210

Flux for each unit of energy of the γ rays and
X-rays incident upon the detecting element
1ch is calculated.

206

10
X-ray
sourceyz

γ rays contribution rate (energy flux ratio ofthe γ rays and X-rays) with respect to thedetecting element 1ch is evaluated.

Air
Tally
(Detector space)

X

Molten debris Unit can
Lateral view
Molten debris

CT image is calculated by means of the CT
simulator incorporative of the detector
parameters, etc., and the CT value and error
with respect to fuel debris composition (UO2,
SUS, etc.) is evaluated.

L
210
206

z

X-ray
source
y

X

Air

Tally
10
Unit canUnit: mm

The relation between the γ ray contribution ratio
and the CT value of fuel debris composition(UO2, SUS, etc.) and its error is evaluated.
B: Evaluation of CT image for the typical cases©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.5 X-ray transmission measurement – Analysis conditions

No.164

Analysis conditions (3/3) Analysis cases
Of the 20 passive gamma rays cases, the cases pertaining to burn-up and cooling period were excluded, and 18 cases in which
X-ray energy was added were analyzed.
Case

UO2

ZrO2

SUS

Concrete

Filling rate

FP emission rate

Unevendistribution

Container

X-ray energy

5-1

Base

15vol%

15vol%--

30vol%

Standard

Uniform

Unit can

9MeV

5-2

Small amount of
UO2

7.5vol%22.5vol%--

30vol%

Standard

Uniform

Unit can

9MeV

5-3

MOX

30vol%---

30vol%

Standard

Uniform

Unit can

9MeV

5vol%

5vol%

-

-

10vol%

Standard

Uniform

Unit can

9MeV

25vol%

25vol%--

50vol%

Standard

Uniform

Unit can

9MeV

15vol%

15vol%--

30vol%

Zero emission

Uniform

Unit can

9MeV

5-9

15vol%

15vol%--

30vol%High emission

Uniform

Unit can

9MeV

5-12

15vol%

15vol%--

30vol%

Standard

At the center inthe horizontal
direction

Unit can

9MeV

15vol%

15vol%--

30vol%

Standard

Outer surface inthe horizontal
direction

Unit can

9MeV

15vol%

15vol%--

30vol%

Standard

Uniform

Canister

9MeV

0.5vol%10vol%--

10.5vol%Standard

Uniform

Inner waste container

9MeV

30vol%---

30vol%

Standard

Uniform

Unit can

9MeV

0.075vol%0.075vol%29.85vol%-

30vol%

Standard

Uniform

Unit can

9MeV

15vol%

15vol%

30vol%-

60vol%

Standard

Uniform

Unit can

9MeV

1.05vol%

1.05vol%

7.2vol%20.7vol%

30vol%

Standard

Uniform

Unit can

9MeV

15vol%

15vol%

0vol%

30vol%

60vol%

Standard

Uniform

Unit can

9MeV

15vol%

15vol%--

30vol%

Standard

Uniform

Unit can

6MeV

15vol%

15vol%--

30vol%

Standard

Uniform

Unit can

15MeV

5-4
Filling rate
5-5
5-8
FP emission rate

Uneven distribution
5-13
5-14
Container
5-15

5-16

Uranium-rich

5-17
Metallic debris
5-18
5-19
MCCI debris
5-20
5-21
5-22

X-ray irradiationenergy©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.5 X-ray transmission measurement – Analysis results

No.165

Analysis results (1/3) Flux evaluation of the base caseBase case analysis conditions in the radioparency measurement system
Case No.

Fuel debris
Type

UO2

ZrO2

H2O

Filling rate*

Entire fuel debris

Amount of U**

5-1

Molten debris

15vol%

15vol%

1wt%

30vol%

15.8kg8.79kg*Filling rate does not include H O, **Amount of U = Total mass of Uranium isotopes
2

Flux of energy from x-rays and energy from the γ rays in theradioparency measurement system

X-ray

γ raysTotal flux
[photon/cm2/s]

3.5×104

8.5×102

Average energy
[MeV]

3.3

0.63

X raysFlux [photons/cm2/s]

γ raysResults of base case analysis in the radioparency
measurement system•

Since as against x-rays, the total flux of γ rays
was low (<3%), and the average energy waslow, it is assumed that the impact of noise fromγ rays on the CT image is low.

Energy [MeV]©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1662.2.5 X-ray transmission measurement – Analysis results
Analysis results (2/3) Flux evaluation of all cases

The γ ray and x-ray flux was calculated for all 18 cases （Since the analysis of γ rays is the same as passive gamma rays, it was omitted).
It was found that the x-ray flux reduces following an exponential function with respect to the apparent density of fuel debris.

■Analysis results: X-ray flux with respect to the apparent density of fuel debris
1.5×105

9×10108×10107×1010CT image with lot ofγ ray noise

5-8(Zero FP emission)

6×1010Flux evaluation

5×10104×10103×10102×10105-3
5-1,14,
21,22

5-16
5-5

5-2
5-20

5-15
5-4

1×10105-12.13
5-18(Metallic debris(Unevenand high filling distribution)
rate)

X-ray flux [photon/cm2/s]

γ ray intensity of fuel debris [photon/s]

Small

γ ray flux

Large

■Analysis conditions： Apparent density of fuel debris and γ ray intensity

5-4

1.2×105

5-19
Fitting:
3.3×105 e-x

9.0×104

6.0×104

3.0×104

5-15(Inner waste
container)

5-21(6MeV)

5-17
5-9

0

0

5-14
(Canister)5-22(9MeV)
5-2
5-17

5-1,8,9
5-20
5-3
5-16
5-5
5-18

5-12.13

5-19

Apparent density of fuel debris [g/cm3]
Large

X-ray flux

Apparent density of fuel debris [g/cm3]

Small

・When the apparent density of fuel debris was large and intensity of γ rays washigh, since the proportion of γ ray flux was larger than the x-ray flux, it is inferredthat the γ ray noise in the CT image becomes bigger.

・X-ray flux reduced following an exponential function with respect to the apparent density offuel debris.
・ The values that deviated from the exponential function were values pertaining to the caseswith containers that have no correlation with apparent density (5-14,15) or X-ray irradiationenergy (5-21,22).
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.5 X-ray transmission measurement – Analysis results

No.167

Analysis results (3/3) Flux evaluation of all cases
In order to study the impact of γ ray noise on the CT image, the γ ray and X-ray energy flux （the value obtained byadding flux of each energy and the energy product) respectively was obtained, and the proportion of γ rays as against Xrays was calculated.
γ ray and X-ray energy flux (left vertical axis) and γ ray / X-ray energy flux ratio (right vertical axis)
109

X raysUneven distribution(Center and outer surface)

Metallic debris

107γ/X energy flux ratio

Energy flux [MeV/photon/cm2/s]

108

γ rays

106
105
104

103
102
101
100
5-1 5-2 5-3 5-4 5-5 5-8 5-9 5-12 5-13 5-14 5-15 5-16 5-17 5-18 5-19 5-20 5-21 5-22

Case No.
••

The maximum γ/X ray energy flux ratio was approx. 6%. The proportion of γ rays was large in uneven distribution (5-12,13) and metallic debris (5-18).
In the CT image evaluation, in addition to the base case, uneven distribution and metallic debris will be added to the typical cases and analyzed.

©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1682.2.5 X-ray transmission measurement – Analysis conditions
Analysis conditions

Selection of typical cases for evaluating the CT image

Four cases were selected in all, including the base case (5-1) and the cases in which γ ray noise is assumed to have a major impact on the CT image.
Case

UO2

ZrO2

SUS

Concrete

Filling rate

FP emission rate

Unevendistribution

Container

X-ray energy

5-1

Molten debris (base case)

15vol%

15vol%--

30vol%

Standard

Uniform

Unit can

9MeV

5-2

Small amount of UO2

7.5vol%22.5vol%--

30vol%

Standard

Uniform

Unit can

9MeV

5-3

MOX

30vol%---

30vol%

Standard

Uniform

Unit can

9MeV

5vol%

5vol%

-

-

10vol%

Standard

Uniform

Unit can

9MeV

25vol%

25vol%--

50vol%

Standard

Uniform

Unit can

9MeV

15vol%

15vol%--

30vol%

Zero emission

Uniform

Unit can

9MeV

5-9

15vol%

15vol%--

30vol%High emission

Uniform

Unit can

9MeV

5-12

15vol%

15vol%--

30vol%

Standard

At the center in thehorizontal
direction

Unit can

9MeV

15vol%

15vol%--

30vol%

Standard

Outer surface inthe horizontal
direction

Unit can

9MeV

15vol%

15vol%--

30vol%

Standard

Uniform

Canister

9MeV

0.5vol%10vol%--

10.5vol%Standard

Uniform

Inner waste container

9MeV

5-4
Filling rate
5-5
5-8
FP emission rate

Uneven distribution

5-13

5-14
Container
5-15
5-16

Uranium-rich

30vol%---

30vol%

Standard

Uniform

Unit can

9MeV

5-17

Metallic debris

0.075vol%0.075vol%29.85vol%-

30vol%

Standard

Uniform

Unit can

9MeV

5-18

Metallic debris (High filling rate)

15vol%

15vol%

30vol%-

60vol%

Standard

Uniform

Unit can

9MeV

5-19

MCCI (Molten Core Concrete Interaction)

1.05vol%

1.05vol%

7.2vol%20.7vol%

30vol%

Standard

Uniform

Unit can

9MeV

5-20

MCCI (High filling rate)

15vol%

15vol%

0vol%

30vol%

60vol%

Standard

Uniform

Unit can

9MeV

15vol%

15vol%--

30vol%

Standard

Uniform

Unit can

6MeV

15vol%

15vol%--

30vol%

Standard

Uniform

Unit can

15MeV

5-21
5-22

X-ray irradiation energy©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.5 X-ray transmission measurement – Analysis results

No.169Analysis results (1/5) CT image evaluation of typical cases
Based on the computational results of the CT simulator [1] incorporative of detector parameters, etc. equivalent to the actual
equipment, the CT image was reconstructed.
Case 5-1: Molten debris (Base case) … UO2: 15vol%, ZrO2: 15vol%, H2O: 1wt%

Heterogeneous model
The inside of the container
depicted in terms of voxels(1cm3)

15
10

Unit can
H2O

5

ZrO2
UO2

Random cross-section

Output image

0

2000

10

Y [cm]

Y [cm]…

Input image

CT Value

15

5

1500

0

1000

-5

-5

-10

-10

500
0

Air
Calculated assuming
a cross section of anarbitrary location➢-15-15 -10-50
5
X [cm]

10

15

It was possible to recognize UO2, ZrO2, and H2O which are fuel constituent materials and the container-15-15 -10-50
5
X [cm]

10

15* CT value is computed considering average CT value of container (unit can) as 1000.
* CT value outside the container is set to 0

[1] So Kitazawa, et al., Simulations of MeV energy computed tomography, NDT & E International, Volume .38 Issue 4 (2005)©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1702.2.5 X-ray transmission measurement – Analysis results
Analysis results (2/5) CT image evaluation of typical cases

Case 5-18： Metallic debris (High filling rate) … UO2: 15vol%, ZrO2: 15vol%, SUS: 30vol%, H2O: 1wt%

Heterogeneous model
The inside of thecontainer depicted interms of voxels(1cm3)

15
Input image

CT Value

15

Unit canOutput image

10

2000

10
ZrO2

Random cross-section

UO2

Y [cm]

Y [cm]

…

5

0
-5

SUS
H2O

-10

5

1500

0

1000

-5

500

-10
0

Air
Calculated assuming across section of anarbitrary location-15-15 -10-50
5
X [cm]

1015➢ Even when there was lot of metal and the filling rate was high, it
was passible to recognize each material.

-15-15 -10-50
5
X [cm]

10

15* CT value is computed considering average CT value ofcontainer (unit can) as 1000.
* CT value outside the container is set to 0
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1712.2.5 X-ray transmission measurement – Analysis results
Analysis results (3/5) CT image evaluation of typical cases

Case 5-12： Molten debris unevenly distributed (at the center in the horizontal direction) … UO2: 15vol%, ZrO2: 15vol%, H2O: 1wt%

Uneven distribution model
Nuclear fuel (UO2 + ZrO2)
unevenly distributed
inside the container

15
Input image

Unit canOutput image

10
ZrO2

5

Cross section of the center

➢
➢

UO2

Y [cm]

Y [cm]

ZrO2

Calculated assuming across section of the centerof the container.

2000

10

Air + H2O

UO2

CT Value

15

0

5

1500

0

1000

-5

-5

-10

-10

-15
-15 -10

Air + H2O

-5

0
5
X [cm]

10

15

It was possible to recognize UO2 and ZrO2 in the uneven distribution case as well.
However, since the apparent density decreased as H2O got mixed with water, H2O could
not be distinguished from air.

500
0

-15
-15 -10-50
5
X [cm]

10

15* CT value is computed considering average CT value ofcontainer (unit can) as 1000.
* CT value outside the container is set to 0
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1722.2.5 X-ray transmission measurement – Analysis results
Analysis results (4/5) CT image evaluation of typical cases

Case 5-13: Molten debris unevenly distributed (outer surface in the horizontal direction) … UO2: 15vol%, ZrO2: 15vol%,H2O: 1wt%

Uneven distribution model
Nuclear fuel (UO2 and ZrO2) isunevenly distributed inside the 15
container

Input image

Unit canOutput image

10
ZrO2

5

Cross section of the center➢

UO2

Y [cm]

Y [cm]

ZrO2

Calculated assuming across section of thecenter of the container.

2000

10

Air + H2O

UO2

CT Value

15

0

5

1500

0

1000

-5

-5

-10

-10

-15
-15 -10

Air + H2O

-5

0
5
X [cm]

10

Results were the same as case 5-12 with uneven distribution (at thecenter in the horizontal direction).

15

500
0

-15
-15 -10-50
5
X [cm]

10

15* CT value is computed considering average CT value of container (unit can) as 1000.
* CT value outside the container is set to 0

©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1732.2.5 X-ray transmission measurement – Analysis results
Analysis results (5/5) CT image evaluation of typical cases

In order to evaluate the ability to distinguish material based on the CT value, the average of the CT values of material voxels and the error
were evaluated.
(The CT values of material voxels excluding those that were at the border between materials were identified and error was calculated
based on 3σ method.)

Average CT values of materials and the error

2500

CT Value

2000

UO2

1500
1000

SUS
ZrO2

500

Case No.
Analysis conditions

➢
➢

0

H2O
Air

5-1
Molten debris(Base case)

5-18
Metallic debris(High filling rate)* Uneven distribution cases
were excluded as H2O and Air
were mixed in those cases.
5-12
Uneven distribution ofmolten debris(At the center in thehorizontal direction)

5-13
Uneven distribution ofmolten debris(Outer surface in thehorizontal direction)

As the CT value of UO2 largely differed from other materials, it could be identified.
Under the conditions this time, even SUS, ZrO2 could be identified, however, if multiple influencing factors overlap, it is likely thatthese cannot be identified.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1742.2.5 X-ray transmission measurement
Summary■Major findings obtained from the analysis➢➢

Flux evaluation for all cases：•
γ/X ray energy flux ratio that serves as the indicator for γ ray noise in CT images was maximum approx. 6%. It was
found that the proportion of γ rays was higher in the uneven distribution cases and metallic debris (high filling rate)cases.
Evaluation of CT image from the typical cases：•
From the evaluation of the CT image that was reconstructed based on the results of calculations using the CT
simulator incorporative of detector parameters, etc. equivalent to the actual equipment, and assuming that theheterogeneous model is made up of 1 cm square voxels, it was confirmed that nuclear fuel (UO2) and othermaterials can be identified.

■Study of the necessity of detailed analysis and consolidation of technical issues➢

Flux evaluation for all cases：•
The γ/X ray energy flux ratio that underwent sensitivity analysis in this project was maximum approx. 6%, which is low.
→ Cases in which multiple influencing factors are changed will be additionally analyzed, and the necessity of shielding γ rays
will be studied.

➢

Evaluation of CT image from the typical cases：•
If the apparent density of material changes (air mixes at a size that is lower than spatial resolution) it becomes difficult to
identify material.
→ Assuming cases wherein the apparent density changes, the applicability of CT measurement using a different X-ray
energy will be studied.
•If multiple materials mix at a size that is lower than spatial resolution, it becomes difficult to identify the material.
→ Methods for evaluating the amount of fuel debris, etc. mixed at a size that is lower than spatial resolution will be studied.

©International Research Institute for Nuclear DecommissioningNo.175

2. Implementation Details(1) Analytical evaluation of factors influencing measurement errors of prospective measurement
technologies①

Selection of influencing factors②

Setting the fluctuation range of the influencing factors③

Analytical simulation
2.1Analysis conditions
2. 2 Analysis results
2.2.1 Passive neutrons
2.2.2 Passive gamma rays
2.2.3 Active neutrons
2.2.4 Passive/active neutrons + passive gamma rays
2.2.5 X-ray transmission measurement
2.2.6 Cosmic rays scattering measurement①
Fuel debris
composition
2.2.6
Cosmicrays

No. 179
No. 184

Gdcontent

B
content

MOX②
Burn-up③
FP
emissionrate④
Cooling
Period⑤
Moisturecontent⑥
Filling
rate

Unevendistribution⑦

No. 180

No. 180

No. 179

No. 181No. 181No. 181No. 180

No. 179

No. 182⑧
Container⑨
Irradiatingradiation
source⑩Detector

No. 183－

All©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.6 Cosmic rays scattering measurement - Measurement concept

No.176•

Measurement principle: Measurement of changes in the trajectory of muons penetrating fuel debris•

Measurement system: Placement of muon trajectory detector above and below the sample•

Radiation source conditions: Setting of the muon energy and angle spectrum through EXPACS*•

Detector response: Muon scattering angle distribution (calculated from the changes in trajectory above and below)•

Measurement time: 1 hour (Conditions under which it is assumed that 10,000 or more valid events can be obtained)
Incident muons

Muon trajectory detector (1)Frequency

Muon scattering angle
𝜃0 =13.6 𝑥𝑥
1 + 0.038 ln
𝛽𝑐𝑝 𝑋0
𝑋0

X0: Radiation length(substance dependent)
x: Penetration distanceβcp: Energy

SampleFrequency

Frequency

Muon energy [MeV]Scattered muons

Muon trajectory
detector (2)Cosmic rays scattering measurement system

Muon incidence angle [rad]Conditions forincident muons

Muon scattering angle distribution※EXPACS (https://phits.jaea.go.jp/expacs/jpn.html)©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1772.2.6 Cosmic rays scattering measurement - Analysis model•••

Scope of simulation: The muon trajectory between the upper and lower detectors is reproduced by means ofsimulation
Scope of data analysis: Identification of data on muon scattering around the sample
Data analysis technique: As a primary evaluation, the muon scattering average angles (Total scattering in thearea being analyzed / Total number of muons penetrating the area being analyzed) are compared.
Area
simulatedMuon trajectory detector

Area analyzed

1m

Unit can2m

Muon scattering
distribution

2m

Simulation system

Simulation results(Muon scattering distribution)©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1782.2.6 Cosmic rays scattering measurement - Analyzed cases
Case

UO2

ZrO2

SUS

Concrete

Total(Filling rate)

Burn-up(GWd/t)FP emission rate

Cooling period(Years)6-1

Molten debris

15vol%

15vol%

0vol%-

30vol%

23

Standard

20

6-2

Molten debris (small quantityof U)

7.5vol%22.5vol%

0vol%-

30vol%

23

Standard

20

6-3

MOX

15vol% (MOX)

15vol%

0vol%-

30vol%

0Standard

20

6-4

Filling rate (low)

5vol%

5vol%

0vol%

-

10vol%

23

Standard

20

6-5Filling rate (high)25vol%

25vol%

0vol%-

50vol%

23

Standard

20

6-6

Gd (low)

15vol%

15vol%

0vol%-

30vol%

23

Standard

20

6-7

Gd (high)

15vol%

15vol%

0vol%-

30vol%

23

Standard

20

6-8

B (low)

15vol%

15vol%

0vol%-

30vol%

23

Standard

20

6-9

B (high)

15vol%

15vol%

0vol%-

30vol%

23

Standard

20

6-10

Water content (low)15vol%

15vol%

0vol%-

30vol%

23

Standard

20

6-11

Water content (high)15vol%

15vol%

0vol%-

30vol%

23

Standard

20

6-12

Burn-up (low)15vol%

15vol%

0vol%-

30vol%

1.3

Standard

20

Burn-up (high)15vol%

15vol%

0vol%-

30vol%

51

Standard

20

15vol%

15vol%

0vol%-

30vol%

23

Zero emission

20

15vol%

15vol%

0vol%-

30vol%

23

High emission

20

6-13
6-14
6-15

FP emission rate (Zeroemission)
FP emission rate (highemission)

6-16

Cooling period

15vol%

15vol%

0vol%-

30vol%

23

Standard30

6-17

Cooling period

15vol%

15vol%

0vol%-

30vol%

23

Standard40

6-18

Uneven distribution 1

15vol%

15vol%

0vol%-

30vol%

23

Standard20

6-19

Uneven distribution 2

15vol%

15vol%

0vol%-

30vol%

23

Standard20

6-20

Container 1

15vol%

15vol%

0vol%-

30vol%

23

Standard20

6-21

Container 2

15vol%

15vol%

0vol%-

30vol%

23

Standard

20

6-22

Uranium-rich

30vol%

0vol%

0vol%

-

30vol%

23

Standard

20

6-23

Metallic debris

0.075vol%0.075vol%29.85vol%-

30vol%

23

Standard

20

6-24

Metallic debris

15vol%

15vol%

30vol%-

60vol%

23

Standard

20

6-25

MCCI debris

1.05vol%

1.05vol%

7.2vol%20.7vol%

30vol%

23

Standard

20©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.1792.2.6 Cosmic rays scattering measurement - Analysis results: (1) Amount of U, filling rate

Muon scattering angle [rad]➢Evaluation of the correlation between the amount of U +TRU contained in fuel debris and muon
scattering
Parameters: U content, MOX fuel, filling rate

Muon scattering angle [rad]•Weight of U + TRU [kg]Molten debris with a different U content and MOX fuel(6-1, 6-2, 6-3)•••

Filling rate [vol%]Changes in the filling rate of molten debris(6-1, 6-4, 6-5)

The muon scattering value increased as the weight of U +TRU in fuel debris increased.
When the composition was the same, scattering value increased with the filling rate.
The difference between UO2 fuel and MOX was small.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.6 Cosmic rays scattering measurement - Analysis results: (2) Gd, B content No.180

Muon scattering angle [rad]Muon scattering angle [rad]Muon scattering angle [rad]• Evaluation of the impact of the proportion of each component contained in fuel debris➢ Parameters: Gd2O3, B4C, Moisture content

Gd2O
Gd2O3
proportion[vol%]
3

Gd2O3 (Density: 7.41g/cm3)(6-1, 6-6, 6-7)••

C
B4CB4proportion[vol%]

B4C (Density: 1.76g/cm3)(6-1, 6-8, 6-9)

Weight of U + TRU [kg]H2Oproportion [wt%]
H2O

H2O (Density: 1.0g/cm3)(6-1, 6-10, 6-11)

The impact of the proportion of Gd2O3 which has comparatively higher density was large.
The impact of B4C and moisture content which have a lower density was small.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.181

2.2.6 Cosmic rays scattering measurement - Analysis results (3): Burn-up, FP, cooling period
Evaluation of impact of operating conditions and cooling conditions, etc.
Parameters: Burn-up, FP emission rate, cooling period

Muon scattering angle [rad]Muon scattering angle [rad]Muon scattering angle [rad]

•
➢

Zero
emission

Burn-up(6-1, 6-12, 6-13)••

NormalHighemission

FP emission rate
(6-1, 6-14, 6-15)

Cooling period [y]Cooling period(6-1, 6-16, 6-17)

Impact of burn-up, FP emission rate, and cooling period was small.
Impact of difference in isotopes of the same element on muon scattering was small.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.6 Cosmic rays scattering measurement - Analysis results: (4) Volume of fuel No.182
debris and uneven distribution➢

Evaluation of impact of the volume of fuel debris and uneven distribution of its location
The impact of different volumes, density and location in the base case and in cases with the same
components and weight was evaluated.
Parameters: Volume, density and location① Uniform distribution

Volume: 6,500cm3
Density: 2.413g/cm3② Uneven distribution(center)③ Uneven distribution(on the sides)

Muon scattering angle [rad]••Uneven distribution conditions

Volume: 1,950cm3
Density: 8.0425g/cm3① Uniform distribution② Uneven③ Uneven distribution
distribution (center)(on the sides)

Uneven distribution conditions(6-1, 6-18, 6-19)••

When the weight remained the same but the volume changed, the muon scattering value changed.
Impact of difference in location was small when the volume remained the same.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.6 Cosmic rays scattering measurement - Analysis results: (5) Canister••➢No.183

Evaluation of the impact of difference in the container in which fuel debris is collected
The fuel debris composition was set to the same conditions as the base case, and the filling rate was set at
30% of the capacity of the container.
Parameters: Shape of the container

容器種類
Type
of container

Evaluation
model
評価モデル
(Unit: mm)（単位：mm）

Shape容器形状
of the container

1.
can
1. Unit
ユニット缶

φ210

φ210

Figure Proposed structure of
the unit can[Reference:
Fuel[Ref. デブリ取り出しPJ]
Debris Retrieval PJ*]

Unit canFuel
デブリdebrisFrom
the Subsidy Project
図 ユニット缶の構造案
of Decommissioning and Contaminated
Water Management in the FY2018
平成３０年度補正予算
廃炉・汚染水対
Supplementary
Budget - Developmentof
策事業費補助金
燃料デブリ収納・移送Technology
for Containing,
Transfer andStorage of Fuel
Debris・保管技術の開発
研究報告書（最終報
Research告）令和３年３月よりReport (Final Report) March
2021

Inner waste container

5
ユニット缶
Unit can2

Horizontal
cross-section
水平断面2.
2.Canister
収納缶

Verticalcross section
鉛直断面

φ240

25
Canister
収納缶

φ 240mm
240mmx×928mmHΦ
928mmH
板厚10mm10mm
Thickness

10

Figure Proposed structure of canister (Simple
installation structure / Inner diameter 220mm /without air feeder)
From the Subsidy Projectof ”Decommissioning andContaminated Water
平成３０年度補正予算「廃炉・汚染水対策事業費
Management in the FY2018 Supplementary Budget
補助金（燃料デブリ収納・移送・保管技術の開発）」(Development of Technology for Containing, Transfer
2020年度最終報告2021年6月より
and Storageof Fuel Debris PJ” FY2020 Final Report
dated June 2021

[Ref. 収納缶PJ][Reference:
Canister
PJ**]

928

30
φ240
Horizontal
cross-section
水平断面

Fuel debris
デブリ

Unit can
ユニット缶

Verticalcross section
鉛直断面

500

Waste
inner container
廃棄物内容器(Waste storage(廃棄物保管容器)
container)[Reference:
Treatment[Ref. 処理・処分PJ]
and Disposal PJ***]

Figure Proposed structure of thewaste storage container
From the Subsidy Project
図 廃棄物保管容器の構造案
of ”Decommissioning
and Contaminat
平成３０年度補正予算
ed Water Management in the FY2018「廃炉・汚染水対策事業費補助金
Supplementary Budget (Research and Development for（固体廃棄物の処理・処分に関する研究開発）」
Treatment
and Disposal of Solid Wastes)” FY2019
Accomplishment
Report dated2020年12月より
December 20202019年度成果

500
300

Unit can2
2

Horizontal
cross-section
水平断面

Fuel
デブリ
debrisVertical鉛直断面
cross section

Waste innercontainer
廃棄物内容器*Subsidy Project of Development of Technology for Fuel Debris Retrieval**Subsidy Project of Development of Technology for Containing, Transfer and Storage of Fuel Debris***Subsidy Project of R&D for Treatment and Disposal of Solid Waste•

Canister

500

3. Waste inner container
3. 廃棄物内容器
500mm□x
300mmH
500mm □×300mmHThickness
板厚2mm 2mm

200

Muon scattering angle [rad]φφ 210mm
210mmx×200mmH200mmHThickness
板厚2mm 2mm

Weight of U+TRU [kg]Unit can, canister, inner waste container(6-1, 6-20, 6-21)

Even if the type of container was different, the scattering value increased in accordance with the muon scattering angle and theweight of U+TRU.

•

Parameter survey by container needs to be conducted for detailed evaluation.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.6 Cosmic rays scattering measurement - Analysis results: (6) Type of fuel debris No.184•

Muon scattering angle [rad]➢Evaluation of the impact of the type of fuel debris (molten debris, Uranium-rich, metallic debris, MCCI (Molten
Core Concrete Interaction))
Parameters： Fuel debris composition

Molten debris
Uranium-rich

Metallic debris

MCCI

Weight of U+TRU [kg]Molten debris, Uranium-rich, metallic debris, MCCI
(6-1, 6-22, 6-23, 6-25)•

The muon scattering value increased in accordance with the weight of U +TRU even when the fuel debris composition was
different.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.6 Cosmic rays scattering measurement - Analysis results: (7) Creation ofreference data••

Reference data•

Evaluation of muon scattering angle with respect to the amount of U as reference data•

Simulation was conducted with only UO2 as the uniform component.

① Constant volume (Adjustment of the amount of U by changing the density):••

No.185Constant volume (unit can capacity), Density 0 to 10.525[g/cm3]② Constant density (Adjustment of the amount of U by changing the volume)：•

Constant density (10.525[g/cm3]), spherical (radius 1 to 10[cm])Reference data (Constant volume)

Muon scattering angle [rad]Reference data (Constant density)

Weight of U+TRU [kg]•The muon scattering angle differed in accordance with the weight of U which varies when the volume and density of fuel debris changed.
©International Research Institute for Nuclear Decommissioning2.2 Analysis results

No.186
2.2.6 Cosmic rays scattering measurement - Analysis results: (8) Comparison withreference data•

Evaluation of the overall trend by comparing the reference data and the analyzed cases

Proportion of Gd content (high)Muon scattering angle [rad]Proportion of SUS content (high)Molten debris(small quantity of U)① Variation in measured values•
Variation of ±30% when the weight
was the same as the base case•
Largest when there was Gd content•
Smallest under high density
conditions

Filling rate (high)
Uranium-rich

Metallic debris
MCCI

Uneven distribution
Filling rate (low)Analyzed cases
Reference data (constant volume)
Reference data (constant density)② Lower limit for identification•
Even in the case of metallic debris
and MCCI conditions where U
content was less, scattering occurred
above a certain level.
•
If it is converted to the weight of U in
the reference data it corresponds toapprox. 2.5kg to 5.0kg

Weight of U+TRU [kg]•The reference data evaluated based independently on UO2 and the trend of the analyzed cases matched by and large.

•

The scattering value increased due to the Gd and SUS content.

•

The scattering value decreased in the case of uneven distribution (small volume, high density)©International Research Institute for Nuclear Decommissioning2.2 Analysis results

2.2.6 Cosmic rays scattering measurement – Summary

No.187⚫

Major findings obtained from the analysis:•
A mutual correlation can be seen between the muon scattering value and the weight of heavy elements(U +TRU) present in fuel debris.
•
Scattering value increases due to SUS, Gd2O3 that have a high density.
•
Scattering value fluctuates depending on the volume of fuel.
•Burn-up, FP emission rate, and cooling period have little impact.

⚫

Approximation of measurement accuracy (Primary evaluation based on the average scattering angle value)•
Variation in measured value: ±30% （Fluctuates depending on the conditions in the case of fuel withthe same weight)•
Lower limit for identification: 2.5kg to 5.0kg (Scattering occurs above a certain level even when there is
no U or only minute quantity of U)⚫

Study of the necessity of detailed analysis and consolidation of technical issues•
Detailed analysis of muon scattering angle➢ Study of the method of distinguishing between scattering caused by U, Pu, etc., and scattering
caused by other material➢ Study of the techniques for analyzing muon scattering angle distribution•
Evaluation of the impact of the volume and density of fuel➢ Study of the combination of data analysis and volume evaluation based on image analysis•
Impact of radiation from fuel debris on the sensor➢ Verification of the radiation resistance of the sensor and study of improvement measures©International Research Institute for Nuclear DecommissioningNo.188

2. Implementation Details(2) Study of future research and development plans aiming for application ofsorting technology to actual equipment① Identification of technical issues for the purpose of practical applicationof the sorting technology② Study of research and development plan (Contents, duration, conditions)③ Goals of the main processes©International Research Institute for Nuclear Decommissioning

2.(2) ① Identification of technical issues for the purpose of practical application of the sorting
technology (1/9)

No.189[Work procedures]

A common form was created for identifying issues by measurement technique.

Development of technology for sorting fuel debris and radioactive waste
燃料デブリと放射性廃棄物の仕分け技術の開発
Identification of technical issues for the purpose of実用化に向けた技術課題抽出
practical application
Classification
分類
Items
Accuracy,
Device,Details
No. Principle,
項目
内容
原理，
精度，
装置，
conditions,efficiency,
equipment,
etc.
条件等

etc.
効率等

✓ The step mentioned in section
2.(2)② at which action needs tobe taken is listed here.
Note)

対象計測技術：1.
Target measurement
technology: 1

etc.
設備等

Response
policy
対処方針

Corresponding
対応開発
development
step
ステップ

備考
Remarks

1-1

1-2

1-3

✓ Principle, conditions, etc.: Issues arising from the measurement principle or
the conditions applied, etc.
✓ Accuracy, efficiency, etc.: Issues pertaining to measurement accuracy
identified through analytical simulation✓ Device, equipment, etc.: Issues related to equipment structure, placement,
handling, etc.

✓ Development policy for resolving theissues is mentioned here
✓ Refined as much as possible to ensure itis captured in the research anddevelopment planNote) Indicates that the process of resolving the issue starts at the corresponding development step to get an idea of the prospects.
©International Research Institute for Nuclear DecommissioningNo.190
2.(2) ① Identification of technical issues for the purpose of practical application of
the sorting technology (2/9)[Division of work]
・ The companies responsible for each measurement technique are listed below (following the divisionof analysis work).

・ Overall review is conducted by all organizations including JAEA.
No.

Prospective measurement technology

Division of work

1

Passive neutron measurement technology

Hitachi-GE, MHI

2

Passive gamma rays measurement technology

Hitachi-GE, MHI

3

Active neutron measurement technology

Hitachi-GE, MHI

4

Passive/active neutrons and passive gamma rays measurement
MHI
technology

5

X-ray transmission measurement (High energy X-ray CT method)

Hitachi-GE

6

Cosmic rays scattering measurement (Muon scattering method)

Toshiba ESS©International Research Institute for Nuclear DecommissioningNo.191
2.(2) ① Identification of technical issues for the purpose of practical application of
the sorting technology (3/9)[Consolidation of identification results]
・ From among the issues identified, those considered to be “Key issues” from the following 3 viewpoints were marked.
(In the [Overview of identification results] described hereinafter, the identified key issues are indicated by a red frame.

Classification

View points on key issues

Example of identified issue

A

Issue concerning the principle- Unprecedented

Development of an algorithm for estimating theamount of nuclear material based on the measured
valuesB

Issues that will require time to beresolved

Issues that cannot be resolved unless implemented
before the testC

The properties of fuel debris are unknown, andIssues beyond the scope of
there are major uncertainties in the preconditions
development of the sorting technology for development. Issues that could arise based on
new findings when these unknowns are ascertained.

©International Research Institute for Nuclear Decommissioning

2.(2) ① Identification of technical issues for the purpose of practical application of thesorting technology (4/9)[Target measurement technique]： 1. to 4. Passive/active neutrons and passive gamma rays

[Overview of identification results]•

No.192

Based on the analytical evaluation in 2.(1), it was confirmed that sorting and measuring is possible by combining
measurement techniques depending on the influencing factors.

•

Assuming the concept of measurement by combining the measurement technologies described in 1 to 3, in addition to theissues pertaining to individual measurement techniques (1, 2, 3) that need to be resolved, the technical issues concerning

combination measurement technique (4) were consolidated.
1.

Passive neutron measurement technology

2.

Passive gamma rays measurement technology

3.

Active neutron measurement technology

4.

Passive/active neutrons and passive γ rays measurement technology

・ Gamma rays produced by the spontaneous
fission of fissile nuclides are measured.
Pulse height
波高分析装置analyzer

同時計数回路Coincidence
circuit

Enumerated
計数値
value
Neutrons
中性子

Nuclearmaterial
核物質

Spontaneous
自発核分裂
fission

Neutrons
中性子

中
性
子
検
出
器

Neutron detector

Neutron detector

中
性
子
検
出
器

Conversion
換算アルゴリズムalgorithm

換算アルゴリズム
Conversion
algorithm

γ rays detector
γ線検出器

Amountof
核物質量nuclear material
Container
容器・ Nuclear fission of fissile nuclides is induced byirradiating neutrons from the outside.
・ The prompt neutrons and delayed neutrons
produced as a result, are measured.
同時計数回路Coincidence
circuit

Enumerated
計数値
value

γ rays
γ線
Nuclearmaterial
核物質

3. Active neutron measurement technology

Amountof
核物質量nuclear material

照射
中 Irradiated中性子
neutrons
性
子
発
生
装
置

Nuclearmaterial
核物質

Nuclear
核分裂
fission

Prompt /即発/遅発
delayed中性子中neutrons
性子
検
出器

Neutron detector

・ Neutrons produced by the spontaneous
fission of fissile nuclides are measured.

2. Passive gamma rays measurement
technology

Neutron generator

1. Passive neutron measurement technology

Enumerated
計数値
value
換算アルゴリズム
Conversion
algorithm

Amountof
核物質量nuclear material

Container
容器
容器
Container©International Research Institute for Nuclear Decommissioning

2.(2) ① Identification of technical issues for the purpose of practical application of the sorting
No.193technology (5/9)[Target measurement technique]： 1. to 4. Passive/active neutrons and passive gamma rays
[Overview of identification results]
Note)

The key issues from among the technical issues identified with respect to each individual measurement technique (1, 2, 3) are consolidated.
No.

Items

Details

Method of converting the measured
1-3

The analyst will be requested to verify the state of co-presence byState of co-presence of Cm244 and U is unknown

2-①

value to evaluation value

1-10

Corresponding
development step

Response policy

conducting a sampling analysis of fuel debris.

C

Algorithm

The method of converting the effective mass of Cm244 into

The correlation between the number of spontaneous fission neutrons(Estimation of the amount ofamount of U/Pu, when fuel with different burn-ups is mixed, is

produced and the amount of U/Pu will be studied using a mixed simulation

nuclear material)

yet to be determined.

of fuel debris composition from the 3D nuclide inventory data.

Issues concerning simulation and the prediction accuracy havePrediction accuracy will be verified and calibrated by testing simulated

not been ascertained.

debris.

It is difficult to conduct a calibration test in which the actual

A mock calibration test will be conducted and the validity of the calibration

system, contents and radiation source intensity are simulated.

curve obtained from the simulation will be evaluated.

1-13
Verification of simulation accuracy
3-11

2-③

A, B
2-⑤

B

1-14
2-15

Calibration test

2-⑤

B

3-14
Method of converting the measured
2-2

The analyst will be requested to verify the state of co-presence by
The state of co-presence of Eu154 and U is unknown.

2-①

value to evaluation value

conducting a sampling analysis of fuel debris.

C

The method of correcting self-shielding by means of simulation using theCorrection of the self-shielding

The self-shielding effect of gamma rays depends on fuel debris

effect(density, amount of U, uneven distribution, etc.) and is diverse.

full energy peak of the energy of multiple γ rays emitted from Eu154, etc.,

2-12

2-③

A, B

will be studied.
In the DDA (FNDI) method, if more than a certain level of

The scope of application will be verified by means of simulated tests using

neutron absorption material is present, the fission components

Gd and B, or the introduction of prompt gamma rays method, etc. will becannot be observed.

studied.

Error reduction
3-1

2-①

(Neutron absorption material)

Step 2 (Measurement concept / scenario development)① Study of target performance value② Development of the concept of the equipment by means ofanalysis③ Study of evaluation methods for nuclear fuel material, etc.

2-⑤

④ Re-examination of sorting scenario⑤ Elemental technology verification test

Step 3 (Basic design and software development)① Basic design of the equipment② Software development③ Elemental technology verification test (ongoing)

A, B

Step 4 (Test manufacturing and comprehensive verification test)① Test manufacturing of measuring equipment② Demonstration test (Hot, etc.), improvementNote) Indicates that the process of resolving the issue starts at the corresponding development step to get an idea of the prospects.
©International Research Institute for Nuclear Decommissioning

2.(2) ① Identification of technical issues for the purpose of practical application of the sorting
No.194technology (6/9)[Target measurement technique]： 1. to 4. Passive/active neutrons and passive gamma rays
[Overview of identification results]
No.

4-1

4-2

4-3

Items

Note)

Details

Corresponding
development step

Response policy

Splitting or merging of cells during neutron measurement and passive γ

A comparative evaluation of the streamlining of layout due to merging / reduction inrays measurement

processing time due to splitting will be conducted.

Streamlined layout of the neutron generator, detector, shield, moderator,

The equipment will be configured for the process in which the side to be measured will

2- ②

etc.

be moved and measurement will be carried out sequentially.

(2-⑤)

・Radiation resistant components and equipment will be used.

2- ②

2- ②

Measurement cell

Optimization of the placement of detector

Radiation resistance

Irradiation degradation of the detector, electronic equipment, cables, etc.

・Shield will be installed

4-4

4-5

The scope of measurement and the performance (accuracy, measurement time, etc.)

signals / disturbing signals

will be evaluated by means of performance assessment analysis.

Algorithm

Difficulty in developing a generic and versatile algorithm

Algorithms exclusive to the category of the target of measurement will be used.

2-③

Study of measurement technique otherPossibility of enhancing the measurement accuracy by a measurement
Introduction of the PGA method will be considered.

2-③

than the DDA method

technique other than the DDA method studied this year

γ rays generated due to the nuclear

Need for a design that considers the γ rays generated due to the nuclear

reaction of neutrons

reaction of neutrons

4-7

4-9

2- ②

Performance assessment under the most stringent conditions of effective
Performance assessment analysis

4-6

4-8

(2-①, 2-⑤)

The phenomenon will be understood by comparing the simulation and the test.

Reduction in detection sensitivity due to accumulation of leaked fuel・Measurement using sealed containers

debris・A design that enables decontamination inside the equipment

Enhancement of the maintainability of the numerous detectors

A structure that enables direct maintenance from outside the cell

Accumulation of contamination

Maintainability

Step 2 (Measurement concept / scenario development)① Study of target performance value② Development of the concept of the equipment by means ofanalysis③ Study of evaluation methods for nuclear fuel material, etc.

(2-①, 2-⑤)

A, B

A, B

2-②、2-③、2-⑤B

3-① onwards④ Re-examination of sorting scenario⑤ Elemental technology verification test

Step 3 (Basic design and software development)① Basic design of the equipment② Software development③ Elemental technology verification test (ongoing)

3-① onwards

Step 4 (Test manufacturing and comprehensive verification test)① Test manufacturing of measuring equipment② Demonstration test (Hot, etc.), improvementNote) Indicates that the process of resolving the issue starts at the corresponding development step to get an idea of the prospects.

©International Research Institute for Nuclear Decommissioning

2.(2) ① Identification of technical issues for the purpose of practical application of the sorting
technology (7/9)[Overview of identification results]
No.

5-1

5-2

5-3

5-4

No.195[Target measurement technique]: 5. X-ray transmission measurement (High energy X-ray CT method) 1/2
Note)
Corresponding
development
step

Items

Details

Response policy

Measurement principle

・Since only information on density can beobtained from past radioparency measurements,a measurement technique that enables
evaluation of the amount of nuclear fuel material
by making a distinction between nuclear fuel andstructures (iron, concrete), etc. needs to bestudied.

・As a distinction can be made between nuclear fueland structures (iron, concrete), etc. by means ofradioparency measurement using different X-ray
energies, that measurement technique will be
studied.

2-①
2-②
2-③
2-⑤

・Fuel debris that is smaller than the spatial
resolution is difficult to distinguish using the CT
value.

・The method of calculating the amount of nuclear
fuel that is mixed, based on the increase from theCT value when only iron or concrete are present,
will be studied.

2-①
2-②
2-③
2-⑤ A, B

・Since the dose rate of fuel debris changes
extensively, the S/N of the image could
deteriorate and the measurement accuracy islikely to worsen.

・The necessity of a shield installed in front of thedetector for reducing the dose rate of gamma rays
will be studied.

2-①
2-②

・An algorithm for estimating the amount of
nuclear material based on the values measured
by radioparency measurement using different xray energies, is required.
・An estimation algorithm that combines theradioparency measurement technique with othermeasurement methods is required.

・The relation between each material and the CT
values will be evaluated beforehand, and thenuclear material will be identified based on the CT
value.
・The method of providing information on
distribution of iron, concrete, neutron absorbing
agent, etc. that are factors inhibiting estimation of
the amount of nuclear material using othermeasurement techniques, using the CT measured
values, and performing evaluation will be studied.

2－③
3- ②

Impact of spatial
resolution, etc.
High dose rate gamma
rays
Noise impact

Estimation of the amountof nuclear material
Algorithm

BA, B

Note) Indicates that the process of resolving the issue starts at the corresponding development step to get an idea of the prospects.
©International Research Institute for Nuclear Decommissioning

2.(2) ① Identification of technical issues for the purpose of practical application of the sorting
technology (8/9)[Overview of identification results]

No.196[Target measurement technique]: 5. X-ray transmission measurement (High energy X-ray CT method) 2/2
Note)
Corresponding
development
step

No.

Items

Details

Response policy

5-5Reduction of backgroundradiation

・The larger the background, longer it takes forthe target nuclides to be detected, and the lower
detectable limit is likely to worsen.

・Since the dose rate of accelerator based X-rays ishigher than the background dose rate, a combined
use of the shield in existing equipment for
preventing leakage of X-rays to the outside will be
studied.

2－①
3－①
4－①

5-6

Accumulation ofcontamination

・Fuel debris that has leaked from the container
when it was measured is likely to accumulate
inside the equipment.

・A sealed container will be used.
・A design that enables decontamination inside theequipment will be used.

4－①

5-7

Radiation resistance (γrays)

・The detectors, electronic equipment, cables,
etc. in the vicinity of the measurement cell are
expected to undergo irradiation degradation.

・Since the dose rate of accelerator based x-rays ishigher than the fuel debris dose rate, design ofexisting equipment can be used for the shield, etc.

3－①
4－①

5-8

Radiation resistance(neutrons)

・The detector is expected to degrade faster due
to the spontaneous fission neutrons.

・The need to install a shield will be studied.

3－①
4－①

Maintainability

・It is assumed that the shields around theequipment need to be handled during
maintenance.

・The possibility of being able to carry out direct
maintenance from outside the cell will be studied.
・Measures need to be taken to ensure
maintainability of the radiation source, detector, etc.

3－①
4－①

Daily inspection

・Calibration needs to be performed on a daily
basis in order to monitor ageing degradation or
random failures, and for accurate measurement.

・There is a proven track record of carrying out
measurement for a period of 10 to 15 seconds
without loading fuel debris and monitoring thestatus of the equipment, during normal operation.

4－①

5-9

5-10

Note) Indicates that the process of resolving the issue starts at the corresponding development step to get an idea of the prospects.
©International Research Institute for Nuclear Decommissioning

2.(2) ① Identification of technical issues for the purpose of practical application of thesorting technology (9/9)[Overview of identification results]
No.

No.197[Target measurement technique]: 6. Cosmic rays scattering measurement (Muon scattering method)
Corresponding
development
step

Items

Details

Response policy

Estimation of the amount of U

Development of the technique of evaluating the amount of U
from the muon scattering distribution

・Statistics, functions, etc. that have a strong correlation withamount of U will be selected from the muon scattering
distribution.
・Methods for enhancing the accuracy of estimating the amountof U by combining with other measured values will be studied.

2-③
2-⑤
3-①

The radiation generated by fuel debris (mainly gamma rays)
is incident upon the sensor in the detector which results indead time with respect to muons.

・Background reduction effect will be evaluated by evaluating the
gamma rays flux and neutron flux under the measurement
conditions.

2-①
2- ②
2-⑤

Placement of detector

The distance between the sample and the detector needs to
be increased in order to reduce the background radiation
originating from fuel debris, but the larger the distance
between detectors the lower is the muon count rate.

・The correlation of the detector placement with the count rate of
background radiation and the muon count rate will be evaluated.
・A measurement system that can sufficiently reduce thebackground radiation will be studied.

2- ②
2-④
3-①

Shielding

The sample needs to be covered with a shield in order toreduce the background radiation originating from fuel debris,but the larger the shield lower is the muon scattering angle
measurement accuracy.

・The correlation of the amount of shielding with the count rate of
background radiation and the muon scattering angle
measurement accuracy will be evaluated.
・The shielding thickness that can sufficiently reduce thebackground radiation will be studied.

2-②
2-④
3-①

6-5

Detector specifications

The detector is made up of numerous sensors, and eachindividual sensor has dead time for a certain period after theincidence of gamma rays. The gamma rays count rate canbe reduced by downsizing each individual sensor, but thecost increases.

6-6

Fluctuation in the measured
value depending on the shapeof fuel debris

The measured value fluctuates depending on the shape offuel debris even though its composition and weight are the
same.

6-1

6-2

6-3

6-4

Background radiation
originating from fuel debris

・Sensor size and performance corresponding to the backgroundconditions will be studied.

・The method of correcting the measured value by estimating theshape of fuel debris based on the spatial distribution of muon
scattering angle, will be studied.

A

B

2-②
2-⑤
3-①
3-②

2-③
3-②

A

Note) Indicates that the process of resolving the issue starts at the corresponding development step to get an idea of the prospects.
©International Research Institute for Nuclear Decommissioning

Note)No.198

2. Implementation Details(2) Study of future research and development plans aiming for application ofsorting technology to actual equipment① Identification of technical issues for the purpose of practical applicationof the sorting technology② Study of research and development plan (Contents, duration, conditions)③ Goals of the main processes©International Research Institute for Nuclear Decommissioning

2.(2)② Research and development plan (Contents, duration, conditions) (1/11)[Development steps and R&D items]

No.199[Target measurement technique]： 1. to 4. Passive/active neutrons and passive gamma rays

(Step 1) Feasibility study of measurement (Analyzing the possibility) ⇐ This year’s research① Setting the purpose and target (shape, density, etc.) of measurement② Nuclear material → Evaluating the behavior of radiation incident upon the detector③ Identification, etc. of technical issues contributing to the concept of measurement(Step 2) Development of the measurement concept and re-evaluation of the assumed scenarios

Response to the identified issues① Study of target performance value required for sorting (Temporarily set for about 1 year ＋ updated after ② to ④)
Target performance value of the measuring equipment (measurement time, equipment size, lower detectable limit, error,
radiation resistance, etc.) will be set.

1-3, 2-2, 3-1, 4-3, 4-4② Development of the concept of the equipment by means of detector response analysis (about 2 years)
Analytical simulation including the detector model will be performed with the purpose of developing the concept of the equipment
and evaluating performance.

4-1, 4-2, 4-3, 4-4, 4-7③ Study of method for evaluating nuclear fuel material, etc. based on the measured quantity (about 2 years)
Evaluation method for estimating the amount of nuclear fuel material based on measured values will be developed (Including
combination of measurement technologies).

1-10, 2-12, 4-5, 4-6, 4-7④ Re-examination of sorting scenarios (Temporarily set for about 2 year ＋ updated after ① to ③)
The applicable sorting/segregation scenarios, the location where the measuring equipment will be used, etc. will be studied.
⑤ Elemental technology verification test using existing equipment, etc. (About 3 years: excluding permission and authorization)
Applicability will be verified by means of elemental technology verification test using existing non-destructive measuring
equipment, etc.

1-13, 1-14, 2-15, 3-1, 3-11, 3-14, 4-2,
4-3, 4-4, 4-7(Step 3) Basic design and software development① Basic design of the equipment② Software development for estimating the amount of nuclear material

4-8, 4-9③ Elemental technology verification test (ongoing)(Step 4) Comprehensive verification test using prototypes, simulated radiation source, etc. (Demonstration test using the hotlaboratory, etc.)① Test manufacturing of measuring equipment② Demonstration test using spent fuel with a known composition or actual fuel debris⇒ Improvement in software for estimating quantity of nuclear material⇒ Improvement in placement of detector, error evaluation(Step 5) Fabrication of actual equipment ⇒ actual operation©International Research Institute for Nuclear Decommissioning

2.(2)② Research and development plan (Contents, duration, conditions) (2/11)[Development steps and R&D items]

No.200[Target measurement technique]: 5. X-ray transmission measurement (High energy X-ray CT method)(Step 1) Feasibility study of measurement (Analyzing the possibility) ⇐ This year’s research① Setting the purpose and target (shape, density, etc.) of measurement② Nuclear material → Evaluating the behavior of radiation incident upon the detector③ Identification, etc. of technological issues contributing to the concept of measurement(Step 2) Development of the measurement concept and re-evaluation of the assumed scenarios

Response to the identified issues① Study of target performance value required for sorting (Temporarily set for about 1 year ＋ updated after ② to ④)
Target performance value of the measuring equipment (measurement time, equipment size, lower detectable limit, error,
radiation resistance, etc.) will be set.

5-1, 5-2, 5-3, 5-5② Development of the concept of the equipment by means of detector response analysis (about 2 years)
Analytical simulation including the detector model will be performed with the purpose of developing the concept of the equipment
and evaluating performance.

5-1, 5-2, 5-3③ Study of method for evaluating nuclear fuel material, etc. based on the measured quantity (about 2 years)
Evaluation method for estimating the amount of nuclear fuel material based on measured values will be developed (Including
combination of measurement technologies).

5-1, 5-3, 5-4④ Re-examination of sorting scenarios (Temporarily set for about 2 year ＋ updated after ① to ③)
The applicable sorting/segregation scenarios, the location where the measuring equipment will be used, etc. will be studied.
⑤ Elemental technology verification test using existing equipment, etc. (About 3 years: excluding permission and authorization)
Applicability will be verified by means of elemental technology verification test using existing non-destructive measuring
equipment, etc.

5-1, 5-3(Step 3) Basic design and software development① Basic design of the equipment

5-5, 5-7, 5-8, 5-9② Software development for estimating the amount of nuclear material

5-4③ Elemental technology verification test (ongoing)(Step 4) Comprehensive verification test using prototypes, simulated radiation source, etc. (Demonstration test using the hotlaboratory, etc.)① Test manufacturing of measuring equipment

5-5, 5-6, 5-7, 5-8, 5-9, 5-10

② Demonstration test using spent fuel with a known composition or actual fuel debris⇒ Improvement in software for estimating quantity of nuclear material⇒ Improvement in placement of detector, error evaluation(Step 5) Fabrication of actual equipment ⇒ actual operation©International Research Institute for Nuclear Decommissioning

2.(2)② Research and development plan (Contents, duration, conditions) (3/11)[Development steps and R&D items]

No.201[Target measurement technique]: 6. Cosmic rays scattering measurement (Muon scattering method)(Step 1) Feasibility study of measurement (Analyzing the possibility) ⇐ This year’s research① Setting the purpose and target (shape, density, etc.) of measurement② Nuclear material → Evaluating the behavior of radiation incident upon the detector③ Identification, etc. of technological issues contributing to the concept of measurement(Step 2) Development of the measurement concept and re-evaluation of the assumed scenarios

Response to the identified issues① Study of target performance value required for sorting (Temporarily set for about 1 year ＋ updated after ② to ④)
Target performance value of the measuring equipment (measurement time, equipment size, lower detectable limit, error,
radiation resistance, etc.) will be set.

6-2

② Development of the concept of the equipment by means of detector response analysis (about 2 years)
Analytical simulation including the detector model will be performed with the purpose of developing the concept of the equipment
and evaluating performance.

6-2, 6-3, 6-5③ Study of method for evaluating nuclear fuel material, etc. based on the measured quantity (about 2 years)
Evaluation method for estimating the amount of nuclear fuel material based on measured values will be developed (Including
combination of measurement technologies).

6-1, 6-6④ Re-examination of sorting scenarios (Temporarily set for about 2 year ＋ updated after ① to ③)
The applicable sorting/segregation scenarios, the location where the measuring equipment will be used, etc. will be studied.

6-3, 6-4⑤ Elemental technology verification test using existing equipment, etc. (About 3 years: excluding permission and authorization)
Applicability will be verified by means of elemental technology verification test using existing non-destructive measuring
equipment, etc.

6-1, 6-2, 6-5(Step 3) Basic design and software development① Basic design of the equipment

6-1, 6-3, 6-4, 6-5② Software development for estimating the amount of nuclear material

6-5, 6-6(Step 4) Comprehensive verification test using prototypes, simulated radiation source, etc. (Demonstration test using the hotlaboratory, etc.)① Test manufacturing of measuring equipment② Demonstration test using spent fuel with a known composition or actual fuel debris⇒ Improvement in software for estimating quantity of nuclear material⇒ Improvement in placement of detector, error evaluation(Step 5) Fabrication of actual equipment ⇒ actual operation©International Research Institute for Nuclear DecommissioningNo.202

2.(2)② Research and development plan (Contents, duration, conditions) (4/11)[Approach towards the criteria for sorting]● Sorting category and criteria (Tentative plan in accordance with the current approach)
Criteria (tentative)

Sorting category

Main purpose of sortinga: Canister(Inner diameter 220mm, heightapprox. 1m)*1*2(None of theconditions
mentioned on
the right are
met)Concentration of U-235
in fuel debris: Lower
than reactivity equivalent
to 1.5wt%

Quantity of
nuclear material
is equal to or
lesser than
3.7kg/container

Quantity of nuclear
material is equal to or
lesser than theamount (to be
determined) in thecase of which physicalprotection andsafeguards end

〇

－

－

－

b: Impact mitigation type canister(Inner diameter 400mm)⚫ Enhancing retrieval throughput⚫ Reducing the scale of fueldebris storage

－

〇

－

－

c: Waste storage container(Control level)

Ensuring criticality safety

－

－

〇

－

d: Waste storage container(Contamination level)

Streamlining of storage andmanagement of waste

－

－

〇

〇

*1: According to results of past evaluations conducted under the Subsidy Project ofDevelopment of Technology for Containing, Transfer and Storage of Fuel Debris, if it isassumed that all fuel debris is composed of U-235 and U-238, as long as weight percent of
U-235 in fuel debris is about 1.7wt% or lower, canisters with inner diameter 400mm can beused. Hence, expecting some margin in this 1.7wt%, 1.5wt% will be set.

Will be updated at Step 2.

*2: Considering the placement and stacking while storing, the least critical mass (approx.
30Kg) will be set on the condition that it would be divided equally in 8 waste storage
containers.

©International Research Institute for Nuclear Decommissioning

2.(2)② Research and development plan (Contents, duration, conditions) (5/11)

No.203(Step 2) Development of the measurement concept and re-evaluation of the assumed scenarios① Study of the values of performance objectives required for sorting (Temporarily set
for about 1 year ＋ updated after ② to ④)

The values of performance objectives of the measuring equipment will betentatively set so that the measurement concept for sorting and the sorting
scenarios can be studied.
➢The values of performance objectives of the measuring equipment will be set,
depending on the measurement location (pretreatment building, storage facility,etc.) during the fuel debris retrieval process, and the target container, so that themeasurement concept for sorting and the sorting scenarios can be studied.
➢Following are examples of the performance objectives that will be set.
- Measured quantity, evaluated quantity- Lower detectable limit, error, measurement time
- Radiation resistance, equipment size, etc.
➢These performance objectives (tentative) will be tentatively set as the preliminary
proposal for about 1 year, and will be updated along with ④ after studying ② to ④
and incorporating the results.

©International Research Institute for Nuclear Decommissioning

2.(2)② Research and development plan (Contents, duration, conditions) (6/11)

No.204(Step 2) Development of the measurement concept and re-evaluation of the assumed scenarios② Development of the concept of the equipment by means of detector response
analysis (for about 2 years)
Analytical simulation including the detector model will be performed with the purposeof developing the concept of the equipment and evaluating performance.
➢A Monte Carlo simulation will be conducted for the stand-alone measurement
technology studied in FY2021 based on the value of the performance objectives(tentative) in ① and the issues in the technologies identified in FY2021, and atentative concept for the equipment will be proposed.
➢The detector model here will take into consideration the placement of a reasonabledetector, moderator, γ rays shielding material, etc., their radiation resistance, etc.
as a system that assumes actual equipment.
➢The performance of this equipment concept will be evaluated focusing on the fueldebris conditions based on the issues identified in FY2021 pertaining to eachmeasurement technology, besides the common cases of typical fuel debris.
➢For this performance evaluation, the amount directly measured using each method(for example, effective mass of Cm244 in the case of passive neutron method, etc.)
will be used, rather than the evaluated value.

©International Research Institute for Nuclear Decommissioning

2.(2)② Research and development plan (Contents, duration, conditions) (7/11)

No.205(Step 2) Development of the measurement concept and re-evaluation of the assumed scenarios③ Study of the technique for evaluating the amount of nuclear fuel material, etc. based on the measured
quantity (about 2 years)
The method for evaluating the amount to be evaluated (quantity of nuclear fuel material, etc.) based on
the measured quantity will be studied.
➢ The method for deriving the amount or properties to be evaluated from the measured quantity,
in the case of the equipment concept studied in ② will be studied and a tentative plan will becreated.
➢ The amount or properties to be evaluated are basically pertaining to the amount of U, but inorder to discover a broad range of possibilities pertaining to each measurement technology(broad range of possibilities of the sorting scenarios), they will be considered as indicators,
etc. related to the Pu quantity, fissile quantity, isotopic composition, moisture content,
criticality risk, and the evaluated quantity that is believed to be applicable to sorting or nuclear
fuel material control, including the relative values, will be studied.
➢ The method for derivation refers to setting dedicated algorithms or parameters for estimating
the evaluation quantity, based on the property of the measured quantity to vary depending on
the conditions of the objects to be measured, or on the correlation between measured quantityand evaluation quantity, and wherein multiple measurement methods are combined if required.
➢ In this study, test analysis and investigations will be conducted as required.

©International Research Institute for Nuclear Decommissioning

2.(2)② Research and development plan (Contents, duration, conditions) (8/11)

No.206(Step 2) Development of the measurement concept and re-evaluation of the assumed scenarios④ Re-examination of sorting scenarios(Temporarily set for about 2 years ＋ updated after ① to ③)
The applicable sorting/segregation scenarios will be studied, and prospective options will be
proposed.
➢ During the process from retrieval to storage of fuel debris, possible sorting/segregation
scenarios will be studied based on the results of ① to ③, and prospective options will
be proposed.
➢ During this study, the practicality/ rationality of managing fuel debris and waste at the
Fukushima Daiichi Nuclear Power Station as well as the feasibility of the proposed
technology will be taken into consideration, and the location of measurement,
measurement technology (multiple combinations are also possible), measured quantity,
management method, etc. will be consolidated.
➢ A preliminary proposal will be tentatively set for about 2 years, and will be updated
along with ① after studying ① to ③ and incorporating the study results.

©International Research Institute for Nuclear Decommissioning

2.(2)② Research and development plan (Contents, duration, conditions) (9/11)

No.207(Step 2) Development of the measurement concept and re-evaluation of the assumed scenarios⑤ Elemental technology verification test using existing equipment, etc.
(For about 3 years: excluding permission and authorization)
Applicability will be verified by means of elemental technology verification test using
existing non-destructive measuring equipment, etc.
➢Assuming that existing non-destructive measurement equipment or newly
installed small equipment will be used, in anticipation of measurement of actual
nuclear fuel material in the future, by implementing simulation and preliminary
tests (Refer to the next page for prospective existing equipment), verification tests
for element technology will be conducted.
➢The above-mentioned test plan will be created in FY2022 (including study of TMI-2
debris measurement). In addition, the permission and authorization required for
implementing the elemental technology verification tests will be consolidated.

©International Research Institute for Nuclear Decommissioning

2.(2)② Research and development plan (Contents, duration, conditions) (10/11)

No.208[Target measurement technique]: 5. X-ray transmission measurement (High energy X-ray CT method)(Step 2) Development of the measurement concept and re-evaluation of the assumed scenarios① Study of target performance value required for sorting (Temporarily set for about 1 year ＋ updated after ② to ④)
Values of the performance objectives of the measuring equipment- Measurement time
- Equipment size
- Setting of lower detectable limit, error, radiation resistance, etc.

② Development of the concept of the equipment by means of detector response analysis (about 2 years)
Analytical simulation including the detector model will be performed with the purpose of developing the concept of the equipment andevaluating performance.
- The radioparency measurement method using different x-ray energies will be evaluated by means of analysis, and the applicability
will be examined.
- The shielding, etc. for reducing gamma rays originating from fuel debris will be evaluated by means of analysis, and its need will be
examined.
- The method of evaluating the amount of fuel debris, etc. mixed at a size that is lower than spatial resolution will be studied by meansof analysis.
③ Study of method for evaluating nuclear fuel material, etc. based on the measured quantity (about 2 years)
Evaluation technique for estimating the amount of nuclear fuel material based on measured values will be developed.
- The stand-alone radioparency measurement method and its combination with other measurement methods, etc. will be studied bymeans of analysis.
- The algorithm for estimating and evaluating the amount of nuclear fuel material will be studied based on the results of studying ②
and combinations of measurement methods, etc.

④ Re-examination of sorting scenarios (Temporarily set for about 2 years ＋ updated after ① to ③)
The applicable sorting/segregation scenarios, the location where the measuring equipment will be used, etc. will be studied.
- Considering the practicality/ rationality of managing fuel debris and waste at the Fukushima Daiichi Nuclear Power Station as well as
the feasibility of the proposed technology, the location of measurement, measurement technology (multiple combinations are also
possible), measured quantity, management method, etc. will be consolidated, and the sorting scenarios will be developed once again.

⑤ Elemental technology verification test using existing equipment, etc. (About 3 years: excluding permission and authorization)
Applicability will be verified by means of elemental technology verification test using existing non-destructive measuring equipment,etc.
- Tests will be conducted using simulated fuel debris (cold) test pieces, by means of existing radioparency measurement equipmentthat can irradiate different x-ray energies, and applicability to fuel debris will be verified.
©International Research Institute for Nuclear Decommissioning

2.(2)② Research and development plan (Contents, duration, conditions) (11/11)

No.209[Target measurement technique]: 6. Cosmic rays scattering measurement (Muon scattering method)(Step 2) Development of the measurement concept and re-evaluation of the assumed scenarios① Study of target performance value required for sorting (Temporarily set for about 1 year ＋ updated after ② to ④)
Values of the performance objectives of the measuring equipment- Measurement time
- Equipment size
- Setting of lower detectable limit, error, radiation resistance, etc.

② Development of the concept of the equipment by means of detector response analysis (about 2 years)
Analytical simulation including the detector model will be performed with the purpose of developing the concept of theequipment and evaluating performance.
- Creation of simulation model including the detector model- Evaluation of the correlation of the background radiation count rate resulting from detector placement, shielding, etc. andthe muon count rate
- Study of required specifications of the detector corresponding to the measurement conditions③ Study of method for evaluating nuclear fuel material, etc. based on the measured quantity (about 2 years)
Evaluation method for estimating the amount of nuclear fuel material based on measured values will be developed.
- Study of the method of evaluating the amount of U by combining muon scattering distribution and other measured values,etc.
- Imaging by analyzing the spatial distribution of muon scattering angles and development of a correction method dependingon the shape of fuel debris④ Re-examination of sorting scenarios (Temporarily set for about 2 years ＋ updated after ① to ③)
Applicable sorting/segregation scenarios, the location where the measuring equipment will be used, etc. will be studied.
- Study of combination with other techniques, division of work, etc.
- Study of optimization and positioning of this technique in the fuel debris sorting scenarios as a whole⑤ Elemental technology verification test using existing equipment, etc. (About 3 years: excluding permission andauthorization)
Applicability will be verified by means of elemental technology verification test using existing non-destructive measuring
equipment, etc.
- Sample measurement test using existing muon measurement facilities- Development of elemental technologies such as circuit, etc. and combination tests with existing facilities©International Research Institute for Nuclear DecommissioningNo.210

2. Implementation Details(2) Study of future research and development plans aiming for application ofsorting technology to actual equipment① Identification of technical issues for the purpose of practical applicationof the sorting technology② Study of research and development plan (Contents, duration, conditions)③ Goals of the main processes©International Research Institute for Nuclear DecommissioningNo.2112.(2)③ Goals of the main processes (1/3)[Target measurement technique]： 1. to 4. Passive/active neutrons and passive gamma rays

[Proposed development schedule]
Item

2020

2021

Short term (until FY2024)

Mid- and long-term (FY2025 onwards)

Key dates
Step 1 (Measurement FS)

Increase in scale of retrieval▽

Setting up of scenarios and updatingof the mid- and long-term plan▽▽

Step 2 (Measurement concept / scenario development)① Study of target performance value② Development of the concept of the equipment by meansof analysis③ Study of evaluation methods for nuclear fuel material, etc.
④ Re-examination of sorting scenario⑤ Elemental technology verification test
Step 3 (Basic design and software development)① Basic design of the equipment② Software development③ Elemental technology verification test (ongoing)

Step 4 (Test manufacturing and comprehensive verificationtest)① Test manufacturing of measuring equipment② Demonstration test (Hot, etc.), improvement

Start of Step 3 when future
prospects can be seen in Step 2

Starting of operation

Step 5 (Fabrication of actual equipment)① Designing of actual equipment② Fabrication, installation and testing of actual equipment

Reflection into
evaluation
methodimprovements

Coordination of retrieval
conditions, etc.

Refer to 2.(2)② for goals of Step 2.

Subsidy Project of Development ofTechnology for Containing, Transfer
and Storage of Fuel Debris

Actual fuel debris
Analysis results

Operation(improved)

Analysis facility©International Research Institute for Nuclear DecommissioningNo.2122.(2)③ Goals of the main processes (2/3)[Proposed development schedule]
Item
Key dates
Step 1 (Measurement FS)

2020

2021

[Target measurement technique]: 5. X-ray transmission measurement (High energy X-ray CT
method)Short term (until
FY2024)Mid- and long-term (FY2025 onwards)
Increase in scale of retrieval▽

Setting up of scenarios and updating of
the mid- and long-term plan▽▽

Step 2 (Measurement concept / scenario
development)① Study of target performance value② Development of the concept of the equipment
by means of analysis③ Study of evaluation techniques for nuclear
fuel material, etc.
④ Re-examination of sorting scenario⑤ Elemental technology verification test
Step 3 (Basic design and software development)① Basic design of the equipment② Software development③ Elemental technology verification test
(ongoing)
Step 4 (Test manufacturing and comprehensive
verification test)① Test manufacturing of measuring equipment② Demonstration test (Hot, etc.), improvement

Start of Step 3 when future
prospects can be seen in
Step 2

Step 5 (Fabrication of actual equipment)① Designing of actual equipment② Fabrication, installation and testing of actual
equipment

Refer to 2.(2)② for goals of Step 2.

Starting of operation
Reflection into
evaluation
methodimprovements
Coordination of retrieval
conditions, etc.

Subsidy Project of Development ofTechnology for Containing, Transfer andStorage of Fuel Debris

Operation(improved)

Actual fuel debris
Analysis results

Analysis facility©International Research Institute for Nuclear DecommissioningNo.213

2.(2)③ Goals of the main processes (3/3)[Proposed development schedule]
Item
Key dates
Step 1 (Measurement FS)

2020

2021

[Target measurement technique]: 6. Cosmic rays scattering measurement (Muon scattering method)
Short term (until
FY2024)Mid- and long-term (FY2025 onwards)
Increase in scale of retrieval▽

Setting up of scenarios and updatingof the mid- and long-term plan▽▽

Step 2 (Measurement concept / scenario
development)① Study of target performance value② Development of the concept of the equipment
by means of analysis③ Study of evaluation techniques for nuclear
fuel material, etc.
④ Re-examination of sorting scenario⑤ Elemental technology verification test
Step 3 (Basic design and software development)① Basic design of the equipment② Software development
Step 4 (Prototype manufacturing andcomprehensive verification test)① Prototype manufacturing of measuring
equipment② Demonstration test (Hot, etc.), improvement
Step 5 (Fabrication of actual equipment)① Designing of actual equipment② Fabrication, installation and testing of actual
equipment

Refer to 2.(2)② for goals of Step 2.

Start of Step 3 when
future prospects canbe seen in Step 2

Coordination of retrieval
conditions, etc.

Subsidy Project of Development ofTechnology for Containing, Transfer
and Storage of Fuel Debris

Reflection into
evaluation
methodimprovements

Starting of operation

Operation(improved)

Actual fuel debris
Analysis results

Analysis facility©International Research Institute for Nuclear Decommissioning3. Summary

No.214

Level of achievement of this project
Issues were identified for setting the required specifications contributing to the future research and development
plan by conducting analytical simulation for the regions where existing technology cannot be applied.

1) Analytical evaluation of parameters influencing measurement errors of prospective measurement technologies① Analytical simulation

・The intensity, etc. of the impact of factors on measurement errors was evaluated for eachmeasurement technology based on the correlation between the range of fluctuation of thefactors and the changes in the measured flux, etc. in the direction of the measuring surface
and information on the extent of impact was consolidated. (TRL: Level 2)② Study of the need for detailed
analysis and consolidation of technicalissues*1

The need for further detailed analysis and the issues in reducing measurement errors wereidentified based on the results of analysis.

2) Study of future research and development plans aiming for application to actual equipment① Identification of technical issues
for the purpose of practicalapplication of the sorting
technology*1

・The challenges in studying the measurement technologies required for sorting were identified
based on the technical issues identified through analytical evaluation of parameters having an
impact on the measurement error and the investigation of measurement technologies for sortingwhich was conducted in FY2019.

② Goals of the main processes
involved in developing the sorting
technology*1

・In addition to studying the contents of research and development, the development procedures,and development period for resolving the technical issues identified in the previous section, theapproach towards the pre-conditions and judgment criteria required while examining the
research and development plan were consolidated.
・The technologies to be adopted were narrowed down, and based on that the goals of the mainprocesses involved in developing the sorting technology were organized.

*1： As identification and consolidation of issues in developing technologies, and organization of the goals, etc. are different than development items, the Technology Readiness Level (TRL)
has not been set.
©International Research Institute for Nuclear Decommissioning