ISSN 2188-1456

JAEA R&D Review

2014

Radiation measurement from the sky using theUnmanned Airplane for Radiation Monitoring
System（a）燃料集合体の溶融初期段階（b）炉心支持板への溶融物の移行（c）下部プレナムへの溶融物の侵入と蓄積

Results of simplified simulations on the relocation
損傷していない燃料
溶融した構造物
behavior of simulated molten materials
溶融した燃料

溶融後固化した燃料

固体状の構造物

Message from the President

President Shojiro MATSUURA

We sincerely thank you for your understanding and continued support of our research and development activities.
Since the foundation of the Japan Atomic Energy Agency (JAEA), this publication has been brought out every yearwith the intention of informing you of the day-to-day accomplishments of the JAEA.
As the only research institute in Japan dedicated to comprehensive nuclear research and development, JAEA
has focused on investing efforts toward the recovery from the accident at the Tokyo Electric Power Company,
Incorporated (TEPCO) Fukushima Daiichi Nuclear Power Station (NPS) by our scientific and technical expertise.
Furthermore, JAEA could not prevent the occurrence of a series of incidents, such as inadequate maintenancemanagement at the “Monju” and leakages from the Hadron Experimental Facility of Japan Proton Accelerator
Research Complex (J-PARC). JAEA is endeavoring to recover the public’s trust by improving ourselves. JAEA
defined the year beginning October 2013 as an intensive organizational reform period, all JAEA united in aneffort to achieve the necessary reform. With strong resolution to recreate ourselves anew, we have made efforts toenhance corporate planning functions, safety management functions, and internal control mechanisms, including
reorganizing our research and development framework into six departments. For “Monju,” we established an
independent management system for the power plant and enhanced the quality assurance system. As part of the
J-PARC reforms, we equipped the facilities with safety management systems to guarantee safety.
In April 2014, the “Strategic Energy Plan” was approved by the Cabinet, and the position of nuclear power was
designed in the Plan as “a crucial base-load power source, contributing to stability of the energy supply and demand
structure” in Japan. In addition, it was expressed in the “Basic Policies for the Economic and Fiscal Management
and Reform 2014,” approved by the Cabinet in June 2014.
For research and development related to the decommissioning of the TEPCO’s Fukushima Daiichi NPS, wehave promoted the test on the characterization of the fuel debris, remote decontamination technology, the long-term
integrity of the fuel assembly in the spent fuel storage pool for the preparation of its removal, and the treatment
and disposal of radioactive waste caused by the accident. We have performed research and development activities
towards the quick return of the residents such as radiation monitoring in forests and rivers, migration of radioactive
cesium, and calculation of decontamination effects. In addition, we are making efforts to contribute to the recovery
from the accident by developing improved decontamination materials, etc.
In the research and development of the fast breeder reactor cycle technology, including “Monju” we aim tocollect the research results stipulated in the “Monju Research Plan,” and are working towards lifting the security
measures put in place by the Nuclear Regulation Authority and complying with the new regulation standards. We
are also conducting a research and development activities about reduction and stabilization of radioactive wastes, aswell as transmutation technology using accelerators.

Moreover, for the activities on the backend of nuclear energy utilization, which are necessary for advancing our
project, we have performed research and development on decommissioning of our own facilities, and the treatment
and disposal of radioactive waste. We are also moving forward with the development of technologies related to theprocessing and disposal of low-level radioactive waste, underground disposal and the light-water reactor fuel cycle.
Regarding nuclear hydrogen and heat application, we are making steady progress in research and development
efforts to improve the safety of high-temperature gas-cooled reactors. Furthermore, we aim to conduct trials thatconnect a High-Temperature engineering Test Reactor (HTTR) with heat utilization facilities (gas turbine power
generation, hydrogen production, etc.) as a step toward the practical application of high-temperature gas-cooled
reactors. In the research and development aimed at the practical utilization of nuclear fusion energy, we, as adomestic agency/implementing agency of Japan under roles based on the international framework (ITER Project/
BA activity) have been promoting our operations using the world’s most advanced technologies along the plan that
the JAEA prescribed. As part of the research and development on quantum beam science, we have been producing
a variety of results in areas ranging from the fundamentals of science and technology to industrial applications,through various quantum beam facility groups owned by the JAEA. Moreover, we have recognized that ensuring
nuclear safety is an important step in the use of nuclear power, and we have diligently conducted nuclear safety
research for the improvement of safety in the utilization of nuclear energy. In addition, we have promoted a wide
variety of engineering research activities, which will serve as the basis in the development of nuclear energy
utilization in Japan; these activities have led to many accomplishments.
Hereafter, we are steadily proceeding with structural reforms in accordance with the plan that the JAEA has set
forth. With the increased expectations toward the JAEA’s role in nuclear technology development after the accident
at the TEPCO’s Fukushima Daiichi NPS, we are now actively involved in medium-to-long-term human resource
development for nuclear technology, industry-academia-government collaborations, promotion of international
cooperation with overseas research institutes, and the dissemination of the achievements made in research anddevelopment, without failing to prioritize operations.
We are pleased that you will gain further understanding of the accomplishments of the JAEA through thispublication, and we thank you for your continued encouragement and guidance in our research and development
activities.

JAEA R&D Review 2014
About This Publication and the Outline of Organization of JAEA

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and Development Relating to the Accident
1 Researchat the TEPCO’s Fukushima Daiichi NPS
Environmental Remediation and Plant Decommissioning

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1. The Migration of Radioactive Iodine in Seawater
Variation of Iodine-129 Concentration before and after the Accident at the TEPCOʼs Fukushima Daiichi NPS
2. Ground Deposition Maps of Iodine-131 Obtained Using Airborne Surveys at an Early Stage of the Accident at 1F
A Joint Study with JAEA and DOE/NNSA
3. Visualization of the Distribution of Radiocesium Concentration at the Bottom of Irrigation Ponds
Measurement of Irrigation Ponds Using the Plastic Scintillation Fiber
4. Why Have Air Dose Rates Decreased with Time?
Depth Distributions of Radioactive Cesium in Soils
5. Provision of Radionuclide Distribution Data
The Collection and Quick Provision of Environmental Monitoring Data
6. Simulating the Migration Behavior of Radioactive Cesium in the Environment
A Case Study of the Ogaki Dam Reservoir
7. Chemical Bonding between Radioactive Cesium and Clay Minerals Revealed by Supercomputer
First-Principles Calculation Studies of the Chemical Adsorption Form of Cesium on Clay Minerals
8. Translocation of Radioactive Fallout Cesium in Trees
Analysis of the Distribution of Radioactive Cesium in Plants Using Autoradiography
9. Easy Monitoring of the Radioactive Cesium Concentration in Water
A New Assay System Using a Cesium Adsorption Disk and the GM Survey Meter
10. Assessment of Radiation Dose Reduction of Radioactive Cesium in Buildings
Numerical Simulation Analysis Using Building Models
11. Detection of Defects in the Fuel Cladding Tube
Development of a Defect Detection Technique Using High-Resolution X-ray Computed Tomography
12. Laser Spectroscopic Analysis of Underwater Debris in a Nuclear Reactor CoreA Contribution to the Decommissioning of the TEPCO’s Fukushima Daiichi NPS
13. Melting and Crushing of Fuel Debris with Laser Light as a Heat Source
Countermeasures to Fuel Debris Characterized by Indefinite Shapes, Higher Hardness, Multi-Ingredients, and Porous Bodies
14. Maintaining the Integrity of the Reactor Vessel until the Completion of Fuel Removal
The Influence of Radiation on the Corrosion of Vessel Steels and Corrosion Measures
15. Estimation of the Chemical Form of Fuel Debris in a Reactor Pressure Vessel
Estimation with the Help of Thermodynamic Calculation
16. What Has Formed from the Melted Fuel and Control Blade?
Investigating the Characteristics of Solidified Core Melt
17. Management of Removed Fuel Debris
A Study of Treatment Scenarios for Fuel Debris after Removal
18. An Exploration of the Reason for Iodine-131 Release after Core Cool Down
Release of Iodine-131 from Contaminated Water in the Basement of Reactor Buildings
19. Safe Storage of Spent Cesium Adsorbent
Characterization and Evaluation for Hydrogen Production and Corrosion
20. Development of an Analysis Method for a Hard-To-Measure Nuclide in Wastewater
Cadmium-113m Analysis Method Using β-ray Measurement
21. Toward the Treatment and Subsequent Disposal of the Accident Waste
An Analysis of Rubble and Trees Collected from the Site of the TEPCOʼs Fukushima Daiichi NPS

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2 Nuclear Safety Research

Implementing Continuous Improvements for Safety33

1. Fuel Failure Processes Revealed by Numerical Approach
Simulation of Fracture Behavior Using Damage Mechanics Model
2. Improving Structural Integrity Assessment of Reactor Pressure Vessels
Effect of Neutron Irradiation on Fracture Properties of Stainless Overlay Cladding for Reactor Pressure Vessel
3. Structural Integrity Assessment of Nuclear Components Operated over a Long Period
Probabilistic Fracture Mechanics Analysis for Ni-Based Alloy Welds
4. Investigating Gas Inflow Influences in Accidents
Concentration Analysis with Gas Measurement Device
5. Dissolution of Compacted Clays in High-Level Radioactive Waste Repositories
Alkaline Dissolution Inhibited by Accessory Minerals in Bentonite
6. Acquisition of Evaluation Data for Severe Accident in Reprocessing Plant
Acquisition and Evaluation of Airborne-Release Data from Boiling Accident

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3 Advanced Science Research

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Advanced Science Pioneers the Future40

1. Flow of Spins Resulting from Acoustic Wave
Towards Rare-Metal-Free Magnetic Devices
2. Elucidation of the Spin State of Conduction Electrons in Graphene
Significant Progress for Spintronic Applications of Two Dimensional Materials
3. A Precursor to Exotic Superconductivity
Nematicity and Superconductivity in URu2Si2
4. Hydrogen-Induced Insulation Degradation of Ceramic Capacitors
Microscopic Evidence Provided by Muons
5. A Study to Reveal the Mechanism of Radiobiological Effects
DNA Damage Induces Aberrations in Normal Chromosomes in a Living Cell
6. Discovery of Uranium Migration in Groundwater with Silica Colloid
Devisal of an Effective Method for Elucidating the Chemical Forms of Trace Actinides in Environmental Water
7. Search forHypernuclei
Can We Make Neutron-Rich Hypernuclei?

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4 Nuclear Science and Engineering Research

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Promoting Basic R&D on Nuclear Energy and Creation of Innovative Technology to Meet Social Needs

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1. Improving the Prediction Accuracy of the Decay Heat from Nuclear Force
Clarify Effect of Tensor Force on β Decay
2. Understanding the Behavior of Molten Material in Severe Accidents
Developing Numerical Simulation Method of Melt Relocation
3. Improving Uranium Accountancy
Demonstration of a Novel Nondestructive Assay for Nuclear Waste Drums at Ningyo-toge
4. Improving the Evaluation Accuracy of the Long-Term Integrity of Nuclear Reactors
Developing a Model of the Microstructural Evolution under Neutron Irradiation
5. Developing a Database of Fundamental Properties for Partitioning and Transmutation
Evaluating Thermodynamic Data on Intermetallic Compounds with Electrochemical Techniques
6. Extraction of Strontium Using a Macrocyclic Compound
Extraction Enhanced by Applying Ionic Liquid Solvents
7. Controlling Valences by Electrocatalysis of Pt Particles
Rapid and Selective Valence-Control Method for Actinide

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8. Selective Detection of Higher Enriched Uranium Particles
Secondary Ion Mass Spectrometry Combined with Solid-State Nuclear-Track Detection
9. Mechanism of Greenhouse Gas Emission from Rice Paddy
Simulation of Atmospheric Ammonia Exchange Using the Land Surface Model for Transport of Radioactive Materials
10. Influence of Physical Characteristics on Internal Dosimetry
Use of Human Models to Compare Caucasians with Japanese
11. Ensuring the Reliability of Measurements of Exposure of Workers and Public to High-Energy γ-rays
Establishment of the High Energy γ-ray Calibration Field
12. Measuring Radiation Intensity in Nuclear Reactors
Developing γ-ray Detectors for Remote Inspection in Primary Containment Vessel
13. Technical Development for In-Pile Irradiation-Assisted Stress Corrosion Cracking Test Capsules
Research on Safety Evaluation of Light Water Reactors
14. Developing a Real-Time Management System for Workers in Radiation-Controlled Areas
Worker-Safety Management in Radiation Controlled Areas

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5 Quantum Beam Science Research

Quantum Beam Science and Technology Researches and Research Sites
R&D Using Quantum Beam Facilities and Fundamental Technologies

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1. Biocompatible Plastic Fabrication by Ion Beam
Development of Micro/nanofabrication and Surface Modification Technique Using Focused Ion Beam for Medical Applications
2. Application of Composite-Type Optical Fiber Technique for Medical Instrument Development
Industrialization of Composite-Type Optical Fiber Technique and Startup of JAEA-Supported Venture Business3. Doping Effect of Ferroelectrics
Improvement of Ferromagnetic Mechanism of Bismuth Ferrite
4. Distinction between Polarizations Induced by Magnetic Cycloids and Exchange Striction
Importance of Magnetic Cycloids for Multiferroic Materials
5. Development of an Aluminum-Based Hydrogen Storage Alloy
To Realize Light-Weight Hydrogen Storage Alloys
6. Synchrotron Radiation Mössbauer Absorption Spectroscopy Improved by Electron Detection
Highly Efficient Mössbauer Spectroscopy of Many Periodic Table Elements Is Now Achievable
7. Toward the Development of Protein Absorbents for Rare and Harmful Metals
Structural Characteristics of Halophilic Proteins Clarified by a Quantum Beam
8. Observation of Living Cells under a Soft X-ray Microscope
Laser Plasma Soft X-ray Microscope with a Combination of Intense Laser Plasma Soft X-ray Source and Contact Microscopy
9. Inhibition of Cadmium Movement in Crops by a Peptide
Visualization of Elements in Roots Using a Positron-Emitting Tracer Imaging System
10. Challenge for Nanoscale Analysis by X-ray Spectroscopy in Combination with Electron Microscopy
Development of a Multilayer Grating Spectrometer for Use with Electron Microscopes
11. Discovery of X-ray Coherent Mirage
Indication of the Potential of New X-ray Optics Using Plasma
12. Transparency of Sodium in the Ultra-Violet Spectral Range
Imaging by Light Illumination through an 8-mm-Thick Sodium Sample
13. Innovative Technique for Elucidating Diffusion through Materials
Development of an Advanced Analysis Method for Quasielastic Neutron Scattering
14. Achievement of High-Quality Pulsed Cold Neutron Beam
Performance Evaluation of the J-PARC Cryogenic Hydrogen System
15. Dose Rate after Beam Operation of J-PARC 3 GeV Synchrotron
Investigation and Measurement of Dose Rate Distribution of Radioactive Nuclides Produced in the Accelerator for Higher Output Power
16. Magnetic Shielding via Vacuum Chambers for Beam Orbit Stabilization
Development of Vacuum Chambers of Magnetic Materials at J-PARC
17. Observation of the Beam Injected into the Acceleration Region of the Cyclotron
Development of an Emittance and Acceptance Measurement System
JAEA R&D Review 2014

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6 Nuclear Hydrogen and Heat Application Research

Research and Development of High Temperature Gas-Cooled Reactor, Hydrogen Production, and Heat Application Technologies

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1. Demonstration Plan for the World’s First Nuclear Heat Application with the High Temperature Engineering Test Reactor
Toward the Industrial Use of Nuclear Heat
2. Compatibility between Effective Plutonium Incineration and Nuclear Proliferation ResistanceConceptual Study of a Plutonium Burner High Temperature Gas-Cooled Reactor System
3. Economic Evaluation of a High Temperature Gas-Cooled Reactor Iodine-Sulfur Process Hydrogen Production System
Investigation of Hydrogen Production Cost and Competitiveness4. Toward a Demonstration that Reactor Safety Can Be Secured Using Only Natural Phenomena
To Establish a Test Procedure to be Employed Soon after Restart83

7 Research and Development of Advanced Nuclear Systems

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R&D of Fast Reactor Cycle Technology

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1. Investigation of Molten Core Material Behavior
An Experimental Study on the Containment of Core Disruptive Accidents
2. Aiming for Robustness for the Reactor Building to Withstand Severe Natural Disasters
Evaluation of External Hazards to the Reactor Building
3. Numerical Simulation for Safety Evaluation of the Steam Generators in Fast Reactors
A Multiphysics Evaluation System for Tube Failure Accidents
4. Challenges Facing the Improvement of Inspection Technology for Steam Generator Tubes
Large Scale 3D Simulations of ISI of FBR “Monju” SG Tubes Using ECT
5. Development of Long Life Fuel Cladding Tube for the Fast Reactor
11Cr-ODS Steel with Sufficient Strength, Toughness, and Corrosion Resistance
6. Toward a New Reprocessing Process for the Transitional Period to the Fast Reactor Cycle
The Development of a Uranium and Plutonium Co-recovery Process (Co-processing Process)

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Research and Development Relating to Nuclear Fuel Cycle
8 Backend
and Reprocessing of Spent Nuclear Fuel

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Research and Development Relating to Nuclear Fuel Cycle Backend and Reprocessing of Spent Nuclear Fuel94

1. Toward an Optimized Decommissioning Plan
Development of Formula to Evaluate Manpower Requirements for Dismantling Nuclear Facility
2. Efficient Separation of Uranium from Sludge-Like Uranium-Bearing Waste
Developing Process to Selectively Separate Uranium by Using Hydrochloric Acid
3. Developing Rapid Method to Analyze Radioactive Waste
Developing Method to Analyze 242mAm in Low-Level Radioactive Waste
4. Study to Improve the Safety of Trench Facility
Suppression of Infiltrated Water to Reduce Radioactive Substances from the Facility
5. Visualization of Geosynthesis Procedure
Geosynthesis Data Flow Diagram (Construction Phase)
6. Research in Onshore Mud Volcanism
Origin of Muddy Sand from Kamihoronobe Mud Volcano
7. Investigation of Damage to Rock Mass in Gallery Excavation
Based on Seismic Velocity and Fracture Distribution around Gallery Wall
8. Detection of Unrecognized Volcanic Activity by Sand Grain
Developing Method based on Mineral and Chemical Analysis of Sand Grain in Sedimentary Rock9. Large Inland Earthquake Triggered by Latent Magmatism
Helium Isotopes as a Tool for Detecting Concealed Active Faults
10. Prediction of Radionuclide Speciation in Groundwater
Determination of Standard Thermodynamic Data of Se(VI)/(IV) Couple

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11. Prediction of Long-Term Alteration Characteristics of HFSC
Clarification of Decreasing pH and Development of Alteration Model
12. Evaluation of Long-Term Geological Evolution
Study of the Safety of Geological Disposal Considering Uplift and Erosion
13. Improvement of Ruthenium Removal Rate by Electrochemical Oxidation
Improvement of Stable Operation for Vitrification Process

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9 Nuclear Fusion Research and Development
Toward the Practical Use of Fusion Energy

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1. Lifetime Extension of Beam Dump by Dispersing the Deposition of Laser Energy
A High Laser Energy-Resistant Beam Dump for Thomson Scattering in ITER
2. Investigation of Strain Using Neutron Diffraction
A Foundation for Improvement of ITER Conductor Performance
3. Failure Behavior of the Hot Isostatic Pressed Joint in the Fusion Blanket
The Latest Highlight of Research and Development of Reduced Activation Ferritic/Martensitic Steel for Use in the Fusion DEMO Reactor
4. Toward the Construction of an Intense Neutron Source that Produces Fusion Reactor Conditions
Engineering Validation Research and Engineering Design of the Lithium Facility
5. Disassembly of the Japan Torus-60 Tokamak Device for the Japan Torus-60 Super Advanced Project
Disassembly of Large Radioactivated Structures
6. Accurate Winding of a Nb3Sn Superconducting Magnet
Development of JT-60SA Central Solenoid for Precise Plasma Control
7. World Record-Setting Dual-frequency Gyrotron
The Development for Electron Cyclotron Resonance Heating in JT-60SA
8. Improvement of the Stability Analysis Model for Rotating Plasmas
Introducing Rotation Effects into the Kinetic-Magnetohydrodynamic Model
9. Safe Shutdown Scenarios for Fusion Reactors
The Development of a High-Energy Runaway Electron Simulation Code
10. Removing Fuel from the Nuclear Fusion Blanket
Investigation of Tritium Recovery from a Future Fusion Blanket
11. Stably Supplying Fuel Tritium to Fusion Reactors
Establishment of a New Pebble Fabrication Technology for an Advanced Tritium Breeder

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10 Computational Science and E-Systems Research

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Advanced Computing and Simulation Technology to Support Nuclear Research and Development

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1. Helium Induces Grain Boundary Fracture of Iron
Finding a Universal Law of Grain Boundary Strength Loss Using a Supercomputer
2. Toward Reliability Enhancement of Seismic Risk Assessment of NPP
Proposal for a Ground Motion Generation Method that Combines a Fault Model and Frequency Prediction
3. Simulations for Superconductors with High-Performance Eigenvalue Solver
Toward Real-Size Scale Simulations
4. Massively Parallel Techniques for Realizing 105 Times Higher Spatio-Temporal Resolutions
The K Computer Sheds Light on Multi-Scale Turbulence in Fusion Plasma

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11 Development of Science & Technology for Nuclear Nonproliferation
Development of Technology and Human Resources in the Area of Nuclear Nonproliferationand Security to Support the Peaceful Use of Nuclear Energy

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1. A Probabilistic Expression for Physical Protection Performance in a Nuclear Facility
A Quantification Method for Physical Protection Performance

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Promotion of Collaboration Intellectual Property Held by JAEA

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JAEA R&D Review 20147

About This Publication and the Outline of Organization of JAEA
This publication introduces our latest research and development (R&D) results in each field. Each chapter presents the activities of
one R&D Directorate. As shown in the organization chart, the various R&D Directorates perform their activities through R&D centers
or institutes. Some of these centers or institutes consist of only one site, while others are located at two or more sites, depending on
the R&D activities. The R&D centers and institutes are located throughout Japan, as shown in the map below. The following brief
introduction outlines the research undertaken by each R&D Directorate at various R&D centers and institutes.
1. The Sector of Fukushima Research and Development is engaged in R&D to recover from the accident at the Tokyo Electric Power
Company, Incorporated Fukushima Daiichi Nuclear Power Station. The Fukushima Decommissioning Technology Center is conducting
R&D to restore the nuclear power plant. The Fukushima Environmental Safety Center in the Fukushima prefecture is conducting R&D
concerning decontamination technologies to restore the environment, outreach activities concerning radiation, and investigations ofinternal exposure of the human body. Additionally, the Nuclear Plant Decommissioning Safety Research Establishment is conducting
research and analysis of radioactive materials, development of remote-controlled robots, and establishment of research base facilities.
2. The Nuclear Safety Research Center, Sector of Nuclear Safety Research and Emergency Preparedness, is in charge of safety
research that supports the national nuclear safety regulatory bodies that regulate nuclear power plants, nuclear fuel cycle facilities, andradioactive waste disposal facilities. This work is conducted at the Nuclear Science Research Institute and at the Tsuruga Head Office.
3. The Advanced Science Research Center, Sector of Nuclear Science Research, conducts pioneering research in basic fields of nuclear
power science mainly through the Nuclear Science Research Institute and the Takasaki Advanced Radiation Research Institute.
4. The Nuclear Science and Engineering Center, Sector of Nuclear Science Research, is engaged in key and basic research on various
elemental technologies that support nuclear power use. These efforts are carried out at the Nuclear Science Research Institute and the
Oarai Research and Development Center.
5. The Quantum Beam Science Center, Sector of Nuclear Science Research, is engaged in research using neutrons at the Nuclear
Science Research Institute and J-PARC. Work using electron beams, gamma rays, and ion beams is conducted at the Takasaki Advanced
Radiation Research Institute. Research using lasers and synchrotron radiation is performed at the Kansai Photon Science Institute.

Japan Atomic Energy Agency -Outline of OrganizationManagement Sector
Monju Reorganization Headquarters

AuditorResearch and Development Sector

As of December, 2014

Sector of Fukushima Research and Development

Sector of Fast Reactor Research and Development

Office of Strategic Planning

Fukushima Administrative Department

Prototype Fast Breeder Reactor Monju

Safety and Nuclear Security
Administration Department

Nuclear Plant Decommissioning
Safety Research Establishment

Monju Project Management and Engineering Center

Legal and Audit Department

Fukushima Decommissioning Technology Center

Advanced Fast Reactor Cycle System
Research and Development Center

President
R&D Program Management Department

Nuclear Science Research Institute

Nuclear Fuel Cycle Engineering Laboratories

Executive
Vice
President

General Affairs Department

Nuclear Fuel Cycle Engineering Laboratories

Oarai Research and Development Center

Personnel Department

Oarai Research and Development Center

Executive
Director

Financial Affairs Department

Fukushima Environmental Safety Center

Contract Department
Public Relations Department

Sector of Nuclear Safety Researchand Emergency Preparedness

Office of JAEA Reorganization
Nuclear Safety Research Center

Preparatory Office for a New Research Institute

Nuclear Emergency Assistance and Training Center

Common Mission Sector

Sector of Decommissioning andRadioactive Waste Management
Radioactive Waste Management
and Disposal Project Department
Geological Disposal Researchand Development Department

Nuclear Fuel Cycle Engineering Laboratories
Fugen Decommissioning Engineering Center

Sector of Nuclear Science ResearchIntellectual Resources Management
and R&D Collaboration Department
Office of International Affairs
Construction Department

Center for Computational Science & e-Systems

Tono Geoscience Center
Advanced Science Research Center
Horonobe Underground Research Center
Nuclear Science and Engineering Center
Aomori Research and Development Center

Nuclear Engineering Research Collaboration Center
Nuclear Science Research Institute

Nuclear Human Resource Development Center

Quantum Beam Science Center

Integrated Support Center for Nuclear
Nonproliferation and Nuclear Security

Takasaki Advanced Radiation Research Institute

Tsuruga Head Office8

Ningyo-toge Environmental Engineering Center
Nuclear Science Research Institute

Oarai Research and Development Center

Kansai Photon Science InstituteSector of Fusion Research and Development

Oarai Research and Development Center

Naka Fusion InstituteTokai Administration Center

Nuclear Hydrogen and Heat Application Research Center

Rokkasho Fusion Institute

Oarai Research and Development Center

J-PARC Center

Aomori Research and Development Center

Fuel Removal Project TeamJAEA R&D Review 2014

6. The Nuclear Hydrogen and Heat Application Research Center, Sector of Nuclear Science Research, conducts R&D on technologies forhigh temperature gas-cooled reactors (HTGRs) and thermochemical hydrogen production at the Oarai Research and Development Center.
7. The Sector of Fast Reactor Research and Development is conducting R&D toward the establishment of fast reactor (FR) cycles toaddress long-term energy security and global environmental issues. In Tsuruga, staff at the Prototype Fast Breeder Reactor Monju andMonju Project Management and Engineering Center are conducting R&D on “MONJU.” R&D activities are also aimed at enhancing thesafety of the FR system at the Oarai Research and Development Center, and manufacturing plutonium fuel and reprocessing spent FBR
fuel at the Nuclear Fuel Cycle Engineering Laboratories.
8. The Sector of Decommissioning and Radioactive Waste Management develops technologies for the safe and rational decommissioningof nuclear power facilities as well as measures for processing and disposing radioactive waste in their R&D centers or institutes. This
sector also conducts multidisciplinary R&D aimed at improving the reliability of geological isolation of high-level radioactive waste
in Japan. A particular focus involves establishing techniques for investigating the deep geological environment through R&D at the
Tono Geoscience Center and the Horonobe Underground Research Center. At the Nuclear Fuel Cycle Engineering Laboratories, thefocus is on improving technologies for designing disposal facilities and safety assessment. Additionally, ongoing work is focused on thedevelopment of a next-generation knowledge management system based on the above R&D activities. Furthermore, the development of
nuclear fuel cycle technology for LWRs is in progress at the Nuclear Fuel Cycle Engineering Laboratories.
9. The Sector of Fusion Research and Development is performing fusion R&D as a domestic agency of the International Thermonuclear
Experimental Reactor (ITER) project and an implementing agency of Broader Approach (BA) activities. The procurement activity of the
ITER project, the upgrade of JT-60 to a superconducting machine as a BA activity, fusion plasma research, and R&D on various element
technologies are conducted at the Naka Fusion Institute. Moreover, the International Fusion Energy Research Center project and theEngineering Validation and Engineering Design Activities of the International Fusion Material Irradiation Facility as a BA activity are
performed mainly at the Rokkasho Fusion Institute.
10. The Center for Computational Science & e-Systems conducts research work on advanced simulation technology and on basic
technology in computational science, as well as operation and maintenance of computer systems. These efforts are carried out mainly at
the Nuclear Science Research Institute and the Kashiwa Office.
11. The Integrated Support Center for Nuclear Nonproliferation and Nuclear Security develops technologies for nuclear
nonproliferation and safeguards to ensure peaceful nuclear energy use. These developments are conducted at the Nuclear Science
Research Institute and the Techno Community Square Ricotti.

R&D Institutes/Centers of JAEA
Horonobe Underground Research Center
Tono Geoscience Center
R&D of Geoscientific Researchin order to build a firm scientificand technological basis for
Geological Disposal of HLW

R&D of Geoscientific
Research andGeological Disposalof HLW

Aomori Research andDevelopment Center
Broader Approach Activities, Support
for Nuclear Fuel Cycle Business,
Decommissioning of Nuclear Ship

Fukushima Area
Tsuruga Head Office
Prototype Fast Breeder Reactor MONJU,
and Monju Project Management and Engineering Center,
FUGEN Decommissioning Engineering Center
MONJU
FUGEN

Response to the accident
at the TEPCO’s Fukushima
Daiichi Nuclear Power Station

Ningyo-toge Environmental
Engineering Center

Tokai Area
Nuclear Safety Research, Nuclear Science andTechnology, Neutron Science, Nuclear Fuel-Cycle
Technology Development, Radioactive Waste
Treatment and Disposal Technology Development,
Human Resource Development, etc.
Nuclear Science
Nuclear Fuel Cycle
Research InstituteEngineering Laboratories

J-PARC Center

Decommissioning of
uranium enrichment
plants andenvironmental
remediation activities
at Ningyo-toge mine

Materials and Life Science using
Neutrons and Muons, Particle
and Nuclear Physics using Kaonsand Neutrinos at Japan Proton
Accelerator Research Complex

Kansai Photon Science Institute
Oarai Research and Development Center

Photon & Synchrotron
Radiation Science

Tokyo • Kashiwa Area
c RIKEN/JASRI

Computational Science & e-Systems
Research

Takasaki Advanced Radiation
Research InstituteR&D of Radiation Applications

Research and Test Reactors;
JMTR, JOYO and HTTR,
R&D of FBR Cycle System Technology,
Irradiation Test Technology,
High Temperature Gas-cooled Reactor
Technology and its Hydrogen Production
Technology

Naka Fusion InstituteFusion R&D, ITER Support,
Broader Approach Activities

JAEA R&D Review 20149

１ Research and Development Relating to the Accident at the TEPCO s Fukushima Daiichi NPS
Environmental Remediation and Plant Decommissioning
Environmental
Remediation
Radiation monitoringof environment

Human resource
development

Environmental restoration
activities

Communications
and public relations
Meeting answering
questions about radiation(by the end of March, 2014)
Aizu
28 groups,1693 people

Wooden house

Iwaki
Kennan40 groups,
16 groups,1552 people 3154 people

Calculation result

1st floor

RF
0.8

BathroomKenhoku
82 groups,
6063 people

Sousou20 groups,
1021 people
Kenchu43 groups,4833 people

Minami Aizu
3 groups,431 people

Measurement andestimation of exposure doses
WC

LDK
0.5

Room1 Room2

0.3

Window

1st floor

HospitalCalculation result RF
0.3

X-ray
room

Conference
roomConsulting
room
Treatment
room

Officeroom

0.1

Window

Fig.1-1 JAEA’s post-accident efforts at environmental remediation for the recovery of Fukushima
These activities are introduced on the web page of JAEA Sector of Fukushima Research andDevelopment (http://fukushima.jaea.go.jp).

Environmental Remediation

As a designated public institution under the Disaster
Countermeasures Basic Act, we started taking action
immediately following the Great East Japan Earthquake ofMarch 11, 2011. We have undertaken diverse actions, such asmeasurement of radiation. A sample of our efforts (described
below) are shown in Fig.1-1, and are currently being continued
for the recovery of Fukushima.
Radiation Monitoring of the Environment
It is necessary to exactly evaluate the contamination status
and radiation dose rates caused by the release of radioactive
materials during the accident at the Tokyo Electric Power
Company, Incorporated (TEPCO) Fukushima Daiichi Nuclear
Power Station (NPS) (1F accident). Thus, the variation of theradioactive iodine distribution in the sea before and after the
1F accident (Topic 1-1) has been investigated. In addition,
we have made a map of the contamination by radioactive
iodine, based on airborne monitoring data taken just after the
1F accident (Topic 1-2), resulting in an estimate of the spreadof radioactive materials in the early stage. Development ofa visualization technique for the radiation distribution at thebottom of the storage reservoir (Topic 1-3) and measurementof the soil radioactivity distribution (Topic 1-4) werealso performed. Moreover, monitoring data from various
measurements were collected and disclosed (Topic 1-5),
and information such as monitoring data of the air dose ratemeasured by local buses in the Fukushima prefecture was
uploaded to the web site (http://info-fukushima.jaea.go.jp/
joho_en/) on the activities for the recovery of Fukushima.
Environmental Restoration Activities
For the recovery of the environment, performing quick
decontamination is important. The portal site “Cleanup navi,”
which provides the technical information obtained by thedecontamination model verification project sponsored by the
Cabinet Office, has been established online (http://c-navi.jaea.
go.jp/en/index.html). For understanding long-term changes tothe distribution of radioactive cesium (Cs) in the environment,
10

JAEA R&D Review 2014

a project for researching long-term environmental change inFukushima is being promoted. The behavior of radioactive
Cs in the dam lake (Topic 1-6), the Cs absorption mechanismof clay minerals (Topic 1-7), and the behavior of radioactive
Cs in plants (Topic 1-8) were studied. In addition, a simple
method for measuring the concentration of radioactive Cs
in aqueous solutions (Topic 1-9) was developed. Based on
the above activities, studies toward optimization and highefficiency of decontamination for environmental recovery will
be continued in the future.
Human Resources Development
We are working to spread knowledge of decontamination
by dispatching experts to municipalities to conductdecontamination technology consultation and instruction, field
investigation, and briefing session support for local inhabitants.
In addition, we are creating guidelines on decontamination,
giving advice through one-stop contacts, and cooperating in
the development of a municipal decontamination plan basedon Japan Atomic Energy Agency’s (JAEA’s) decontamination
support system, “RESET.”
Communications and Public Relations
We conduct “meetings answering questions about radiation”
for various groups, such as students at nursery schools,
kindergartens, elementary schools, and junior high schools
in the Fukushima prefecture, to promote greater scientificunderstanding of radiation. By the end of March, 2014, we
had offered such meetings at 232 places, to approximately
18750 people.
Measurement and Estimation of Exposure Doses
We monitored internal radiation exposure using theTokai Research and Development Center’s whole bodycounter (WBC) and the mobile WBC car for inhabitants of
the Fukushima prefecture. By the end of March, 2014, we
measured exposure from about 61800 people. In addition,
evaluation of the radiation dose reduction in different
buildings (Topic 1-10) was performed.

(a)

(b)
Estimation of Status inside Vessel andInvestigation of Cause of Accident

Removal of Spent Fuel and Fuel Debris

Plant
Decommissioning

Mg

viewing
position

Component of fuel assembly

(c)

(d)
Radioactive Waste Management

Research base facility

Fig.1-2 Overview of JAEA’s Activities toward the Decommissioning of the TEPCO’s Fukushima Daiichi NPS(a) A detailed inspection of samples taken from the unirradiated fuel assembly at 1F4, (b) an example of simulation results for
failure of the lower head of the reactor pressure vessel, (c) JAEA staff picking rubble around 1F4 for analysis, and (d) an imageof the Remote-controlled Equipment and Device Development Facility.

Plant Decommissioning
Immediately after the accident, we began providing technicaladvice to the government and TEPCO, and conducting R&D
for removal of the molten core materials (the fuel debris),radioactive waste management, and so forth (Fig.1-2).
Removal of Spent Fuel and Fuel Debris
Some of the fuels could have been damaged by fallingconcrete debris and the like. Hence, we are currently conducting
R&D for inspection of the fuel assembly (Topic 1-11).
For defueling the fuel debris in the reactor core, we areresearching methods of inspecting the inside of the reactor
core (Topic 1-12), as well as developing a cutting and crushingmethod for extracting the fuel debris (Topic 1-13). Since sea
water was injected into reactor vessel for emergency cooling,
the possibility of corrosion problems with the structural
materials was also considered. We conducted an examinationfor evaluating this effect (Topic 1-14).
For safe handling, storage, processing, and disposal, it
is important to grasp the characteristics of the fuel debris
before defueling. We estimated the chemical forms of thefuel debris by a combination of melt-progression analysis
and thermodynamic equilibration (Topic 1-15) to investigate
the mechanical properties using simulated debris made of U,
Zr, and B4C control blades (Topic 1-16) for the purposes ofmanagement of the fuel debris after removal (Topic 1-17).
Estimation of Status inside the Vessel and Investigationof the Cause of the Accident
The fuel in the reactors reached a high temperature andbecame molten when the core cooling system ceased operation

during the station blackout caused by the tsunami. To elucidate
the progress of the melting phenomena in the reactor core, weestimated the status in the core by means of severe accident
analytical codes, simulated tests, and so forth, contributing toa detailed analysis of the progress of the severe accident.
Radioactive Waste Management
A massive volume of contaminated water has been
generated due to the tsunami and the inflow of groundwater in the reactor buildings; this contaminated water is
temporarily stored at the plant site. We evaluated whether
the iodine in the contaminated water could be releasedin the atmosphere (Topic 1-18), and developed analyticaltechniques for “difficult-to-measure” radioactive nuclides in
the contaminated water (Topic 1-19).
Some water treatment systems are being applied to theremoval of radioactive nuclides, e.g., cesium and strontium,from the contaminated water. We are addressing the long-term
storage, processing, and disposal of secondary waste producedby treatment of this contaminated water (Topic 1-20).
We collected rubble and trees from 1F, and analyzed theirradioactivity concentrations (Topic 1-21).
Research Base Facility
We will construct a Research Base Facility in the Fukushima
Prefecture for the purpose of establishing basic technologies
and methods (such as remote-controlling systems and analysisof radioactive materials) that are indispensable to thedecommissioning of 1F. A Remote-controlled Equipment and
Device Development Facility started construction in August, 2014.

JAEA R&D Review 201411

Research and Development Relating to the Accident at the TEPCO’s Fukushima Daiichi NPS

1-1 The Migration of Radioactive Iodine in Seawater− Variation of Iodine-129 Concentration before and after the Accident at the TEPCO’s Fukushima Daiichi NPS −(b) Two months(c) Four months

40˚
b1
b2
39˚

b3

1F

1F

b8

c1
b9

37˚
b10

Latitude

a1

7.0
6.0

35˚

a3

1F

8.0

c4

b12

a2
40˚

9.0

c3

b11

36˚

45˚

10.0

c2

(d) Six months

5.0(e) Eight months

4.0

40˚

a4

3.0

35˚

135˚

1.0
140˚

145˚

Longitude

150˚

155˚

160˚

Latitude

30˚
130˚

2.0

39˚

a5

38˚

0.0

1F

1F
d1
d2

37˚

7

50˚

38˚

I concentration (×10 atoms/ℓ)(a) Before the accident

b5 b6
b7

129

Latitude

b4

e1
e2

d3
36˚

35˚
140˚

d4

141˚

142˚

143˚

Longitude144˚

145˚ 140˚

141˚

142˚

143˚

144˚

145˚

Longitude

Fig.1-3 Oceanographic stations and concentration of 129I measured in this study
Oceanographic stations and concentrations of 129I before and after the accident are shown. Concentrations over
108 atoms/ℓ are shown in red.

If radioactive iodine is accidentally released into theenvironment from a nuclear facility, it is necessary to survey
the migration of this iodine through the environment. We
have developed a technique for measuring iodine-129 (129I) in
environmental samples using accelerator mass spectrometry(AMS), and we have investigated the background level of
129
I to prepare for unexpected accidents. In March 2011,
the accident at the TEPCO’s Fukushima Daiichi NPS (1F)
caused radioactive iodine to be released into the ocean. We
have studied the migration of the accident-derived 129I in thesea area around Fukushima using our developed method andbackground data.
Before the accident, seawater samples had been collected atfive stations in the western North Pacific Ocean (Fig.1-3(a)).
After the accident, samples were collected at 12 stations
after two months (Fig.1-3(b)), four stations after four months(Fig.1-3(c)) and six months (Fig.1-3(d)), and two stations
after eight months (Fig.1-3(e)). Concentrations of 129I in theseawater samples were measured using AMS.
The concentrations of 129I in surface seawater aroundthe western North Pacific Ocean before the accident were(0.94−1.83) × 107 atoms/ℓ. After the accident, the concentrationsof 129I were (1.08−89.8) × 107 atoms/ℓ. The maximum value
was observed at station c3 four months after the accident, andthe minimum was observed at stations b10 and c4 two andfour months after the accident, respectively. The maximum
was 80 times higher than the background value before theaccident. The minimum concentrations were at the same level
as the background value before the accident. A wide range of
129
I concentrations in surface seawater were observed aroundthe 1F site for several months following the accident. It isbelieved that this result was induced by the complex flow
field with the Oyashio and Kuroshio currents flowing into thisregion, and the corresponding mesoscale eddies. Since theconcentration of 129I in surface seawater after eight months
from the accident decreased to the background level, it has
been shown that the concentration of accident-derived 129I has
been reduced by time.
The findings of this study can be applied to simulation
models that predict the distribution of radionuclides inseawater.

Reference
Suzuki, T. et al., Iodine-129 Concentration in Seawater near Fukushima before and after the Accident at the Fukushima Daiichi Nuclear Power Plant,
Biogeosciences, vol.10, no.6, 2013, p.3839-3847.

12

JAEA R&D Review 2014

Research and Development Relating to the Accident at the TEPCO’s Fukushima Daiichi NPS

1-2 Ground Deposition Maps of Iodine-131 Obtained Using Airborne Surveys
at an Early Stage of the Accident at 1F− A Joint Study with JAEA and DOE/NNSA −(a)(c)
I deposition
2
concentration (Bq/m )(as of April 3, 2011)

Counts / Channel

131

I (365 keV)

3

103000 K <1000 – 3000 K
600 – 1000 K
300 – 600 K
100 – 300 K
60 – 100 K
30 – 60 K
10 – 30 K

134

Cs (796 keV)

2

10

1

10

0

10

Unoo Zi.

FUKUSHIMA≦ 10 K-1

10

Mano G.

500

0

1000

(b)

1500

2000

2500

3
3000
Niida G.

Energy (keV)

(e)

(d)
4

4

10

10

Counts / Channel

131

134

I

3

Cs

1F

3

10

10

2

2

10

10

1

1

10

20 km

10

0

0

10

Fig.1-4 Airplane used for the airborne
monitoring (a), and the measurement
instrument comprising three embedded
NaI scintillation detectors (b)The aerial measuring system installedin the air plane is composed of anarray of three 5 cm × 10 cm × 40 cm NaI
detectors. Each detector system produces
a 1024-channel energy spectrum, once
per second with the readout of all the
spectrometers synchronized with thetime from a GPS receiver.

ABUKUMA G.

131

4

10

10

250 300 350 400 450

700 750 800 850 900

Energy (keV)

Energy (keV)

30 km

Shioya Zi.

Fig.1-5 Spectral data
Spectral data obtained by airborne
monitoring (c), and the extracted
energy peak of iodine-131 (131I) (d)and cesium-134 (134Cs) (e) obtainedusing the Gaussian fitting method.

Radioactive substances were spread across a wide areain the aftermath of the accident at the TEPCO’s Fukushima
Daiichi NPS (1F).
We obtained the airborne monitoring data that was collectedby the U.S. Department of Energy National Nuclear Security
Administration (DOE/NNSA) at an early stage, March 17 toApril 5, 2011 (Fig.1-4), and analyzed the spectral information
in a joint research project with DOE/NNSA (Fig.1-5). Within
this data, the energy peak corresponding to iodine-131 (131I)
was observed. DOE/NNSA and we have developed a new
method for analyzing 131I to determine the ground deposition
concentrations and create distribution maps.
Our main focus was on three airborne monitoring flights
on April 2 and 3, 2011, and the energy peak (365 keV) of131
I (Fig.1-5(d)). The deposition concentrations of 131I wereanalyzed by developing a method for calculating the response
to ground deposited activity at varying aircraft heights. Thiswas conducted using Monte Carlo simulations, which extract
the iodine peak and simulate the attenuation of γ-rays in the air.
From this analysis, the 131I deposition distribution was

Same G.

0

5

10

15

20 km

Fig.1-6 131I deposition distribution (as of April
3, 2011)(map drawn using map data provided by the Geospatial
Information Authority of Japan)

High concentrations of 131I have been deposited
to the northwest and south of 1F.

determined, and its maps were created. Fig.1-6 shows the
distribution of 131I depositions at varying concentrations,corrected for radioactive decay up to April 3 when monitoring
was completed.
Analysis of cesium-134 (134Cs), which has a longer halflife, was conducted using the same method as that for131
I. Consequently, high concentrations of 131I, 134Cs, andcesium-137 (137Cs) were shown to be present to the northwestof 1F. In the area south of 1F, high concentrations were also
detected.
To verify the validity of the method, a comparison was
performed with the Japanese Ministry of Education, Culture,
Sports, Science and Technology (MEXT) Emergency
Operation Center’s (EOC’s) soil sampling data (131I, 134Cs)
taken on June 14, 2011. The results from the present analysis
were corrected for radioactive decay for the purposes ofthis comparison. The deposition concentrations of 131I and134
Cs measured on the ground agreed well with the airborne
monitoring results. This newly developed method showed thedetailed distribution of 131I.

Reference
Torii, T. et al., Enhanced Analysis Methods to Derive the Spatial Distribution of 131I Deposition on the Ground by Airborne Surveys at an Early Stage after
the Fukushima Daiichi Nuclear Power Plant Accident, Health Physics, vol.105, no.2, 2013, p.192-200.
JAEA R&D Review 201413

Research and Development Relating to the Accident at the TEPCO’s Fukushima Daiichi NPS

1-3 Visualization of the Distribution of Radiocesium Concentration
at the Bottom of Irrigation Ponds− Measurement of Irrigation Ponds Using the Plastic Scintillation Fiber −

N
0

5

10

15

20

25
m

Data
processing
unit

Measurement
positon of PSF

PSF

Concentration of radiocesium
Sediment sampling(kBq / kg-wet)

Scintillation fiber

Photomultiplier tube

Fig.1-7 A radiation measuring system forthe bottoms of water bodies using PSF
15 bundles of PSF with a diameter of 1 mm.

There are approximately 3700 irrigation ponds in the
Fukushima prefecture. Contamination by radiocesium at thebottom of these ponds has been a concern since the accident
at the TEPCOʼs Fukushima Daiichi NPS (1F). The irrigation
ponds collect the water that falls in the catchment area as rain,
which is then used for agriculture. In general, the radiocesium
concentration in sediment is evaluated by measuring collected
samples in a laboratory. The drawback of this method arethat the samples become radioactive waste. In addition, theevaluation of the distribution of over the entire irrigation pond
is difficult. In Fukushima, a quick and direct technique formeasuring a wide area of the bottom of a pond is needed.
We developed the plastic scintillation fiber (PSF) following
the accident at 1F. PSF has the following characteristics,(1) a detector is long (max.20 m). (2) the water resistanceof the detector is high, and (3) PSF can be put along themeasuring object. Therefore, PSF was applied to themeasurement of irrigation ponds. A photo of the radiation
measurement system based on PSF is shown in Fig.1-7. PSF
is an optical fiber with a plastic scintillator at its core. Opticalsensors (photomultiplier tubes) were located at both ends ofthe optical fiber, and the amount of radiation was measured. In
addition, the position at which radioactive emission took place
was pinpointed based on the difference in the times at which(kBq / kg-wet)

> 40

> 40

30-40

30-40

20-30

20-30

10-20

10-20

< 10

< 10

Fig.1-8 Example of a radiocesium distribution map of anirrigation pond in Fukushima Prefecture
PSF measurements were performed at a line spacing of 5 m. The
map was created using Kriging, which is an interpolation method.
This map accorded with the result of sediment sampling.

the two sensors counted it. The count rate collected by PSF
was used to extrapolate the radiocesium concentration at thebottom of the pond by the results to those taken by waterproof
γ-ray spectrometer that was calibrated by Monte Carlocalculation code. The assumption under which the radiocesium
concentration was extrapolated was that radiocesium wasdistributed uniformly over 10 cm from the surface layer of thesediment. There is good agreement between the concentration
extrapolated from PSF data and that obtained from sedimentcore samples. Using this method, measurement of an irrigation
pond 1000 m2 in area took about four days (with five people
to each team). An example of a radiocesium distribution map
obtained from an irrigation pond in Fukushima is shown in
Fig.1-8. The PSF measurements were performed at a line
spacing of 5 m. The map was created using the interpolation
method known as Kriging. This map accorded with the result
obtained by sediment sampling.
We are transferring these techniques to Midori Net
Fukushima (http://www.midorinet-fukushima.jp) by making
a technical support contract with them. JAEA is going todevelop the technology and provide the support that isnecessary to utilize these techniques for the measurement ofradiocesium in irrigation ponds in Fukushima.

Reference
Sanada, Y. et al., Research and Development of In-Situ Measurement for Radiation Distribution of Waterbed, JAEA-Research 2014-005, 2014, 67p.
(in Japanese).

14

JAEA R&D Review 2014

Research and Development Relating to the Accident at the TEPCO’s Fukushima Daiichi NPS

1-4 Why Have Air Dose Rates Decreased with Time?
− Depth Distributions of Radioactive Cesium in Soils −Concentration (Bq/kg)
0

1000

2000

Depth (cm)

2

3000
134

Cs

137

Cs

4
6
8
10

Nishigo village0

Depth (cm)

2
4
6

1.0

Relative ratio of averaged air dose rates

0

1.00
0.82
0.65

0.51

Approximate curve (Least-square method)

0.45

y = 1.05e−0.00084x
R2 = 0.995
Decreasing rate by physical decay of radioactive Cs
Averaged air dose rates measured using survey meters
Averaged air dose rates estimated under consideration ofalteration in the depth distributions of radioactive Cs
0.1

8

0
10

0.61

400

600

800

1000

Number of days elapsed from March 11, 2011 (d)Koriyama city

Fig.1-9 Depth distributions of radioactive Cs in
the soils(Sampling method: scraper plate; Implementation
period: October 28 to November 29, 2013) A typical
depth distribution wherein concentrations of radioactive
Cs exponentially decrease with depth.

200

Fig.1-10 Temporal changes in averaged air dose rates above theground measured using survey meters
The decreasing rate of the physical decay of radioactive Cs, and thetemporal changes in the air dose rate estimated under consideration ofvarious depth distributions of radioactive Cs.

With funding from the Ministry of Education, Culture,
Sports, Science and Technology of Japan (MEXT) and the
Nuclear Regulation Authority (NRA), we have conductedcomprehensive large-scale environmental monitoring asa national project since June, 2011, and provided reliable
information on the distributions of the air dose rate and theradioactive nuclides deposited in soil on the ground. This has
been done to assess the consequences of the accident at theTEPCO’s Fukushima Daiichi NPS (1F), and to give basic
information for judgment on substantial measures such as
decontamination, rearrangement of evacuation zones and soon. As a part of environmental monitoring, depth distributionsof radioactive cesium (Cs) in soils have been obtained usinga scraper plate method, which can collect soil samples ofdifferent depths with minimal cross-contamination.
During the environmental monitoring performed in
November, 2013, radioactive Cs deposited in the soils remained
within 5 cm of the surface on the whole. Exponential decreasesof the concentrations of radioactive Cs with depth (Fig.1-9)
were observed in many situations. The depth distribution was
found with repeated environmental monitoring to graduallyprogress deeper into the soil.
The temporal change of the averaged air dose rates was alsomeasured above the ground at various locations around the
1F using survey meters, and the results relative to a first result
are shown in Fig.1-10. The reduction rate of the averaged air
dose rates was steeper than that caused by the physical decayof radioactive Cs alone. In addition, the results obtained in a
migration study performed as a part of the project showed thatthe horizontal advection of radioactive Cs was negligibly small.
To find the reason for this difference, simulations of the air
dose rate as a function of the depth distribution of cesium-137(137Cs) were conducted with the Particle and Heavy Ion
Transport code System (PHITS). The calculation results in theapproximated curve (Fig.1-10) were in good agreement withthe plots of the averaged air dose rates. The results indicatethat the averaged air dose rates have decreased because slightchanges in the depth distribution of 137Cs with time haveincreased the shielding effect of the soil against γ-rays.
The present study was sponsored by the Secretariat of the
Nuclear Regulation Authority (NRA).

Reference
Saito, K., Mapping and Modelling of Radionuclide Distribution on the Ground due to the Fukushima Accident, Radiation Protection Dosimetry, vol.160,
no.4, 2014, p.283-287.
JAEA R&D Review 201415

Research and Development Relating to the Accident at the TEPCO’s Fukushima Daiichi NPS

1-5 Provision of Radionuclide Distribution Data
− The Collection and Quick Provision of Environmental Monitoring Data −Environmental monitoring database

Data list

Maps, Charts

Documents

Analysis tools

Data aggregation, integration
Fukushima
Pref.
NSR

METI

MEXT

MHLW
ENV

CAO

・・・

MAFF

Intuitive image,Comparable format

MLIT

Fig.1-11 An environmental monitoring database system
Integrating many kinds of monitoring data and publishing them in a uniform fashion improves data reusability andcomparability.

30 cm
20 cm

Fukushima
KURAMA-Ⅱ

Aiduwakamatsu
Kooriyama

Real-timel ivisualization
i
li

Iwaki

Installing a compact γ-ray survey
meter on public buses
Surveys the air dose rates
continuously

Providing a distribution map

Collecting the data
through a network
Automatic correction

Automatic visualization Providing a time variation graphTransmit the result
from a website

Fig.1-12 A real-time visualization system for the distribution of air dose rates in the Fukushima Prefecture
By installing a compact γ-ray survey system on public buses, survey data will be collected, compensated, visualized
automatically, and finally made available to the public.

To estimate the impact of radionuclides emitted by theaccident at the TEPCOʼs Fukushima Daiichi NPS, it wasnecessary to identify the current distribution of radionuclides
and the tendency of this distribution to change in time.
We conducted research projects to provide environmental
monitoring data, which were essential for these identifications.
One of the research projects was the development ofan environmental monitoring database. Many kinds ofenvironmental monitoring data collected by ministries,
agencies, and local governments have previously been
published in different formats. We republished the data
in a uniform fashion by developing a tool to integratethem automatically, thereby improving data usability andcompatibility (Fig.1-11). Over 300 million records of data
published by 21 organizations were republished. The average
number of accessing the new database was 2.5 times larger(about 166 thousand accesses) than the former database, which

was developed under a previous MEXT project.
In cooperation with the Fukushima Prefecture and Kyoto
University, we conducted another research project that provided
real-time visualization of the air dose rate data monitored by thecompact γ-ray survey system (KURAMA-II). Because of its
limited budget, air dose surveys with KURAMA-II have only
been conducted a few times before. The use of public buses
that run in residential areas on a daily basis and of servers thatcollect, compensate, and visualize the data automatically madereal-time data provision possible (Fig.1-12). Furthermore,just in case a similar accident occurs, it is expected that thissystem can provide data in real time. Real-time visualized air
dose rate data is now available on a large monitor in the NBF
Unix Building near Fukushima Station.
This present study was partly sponsored by the Nuclear
Regulation Authority (NRA), and by the Fukushima Prefecture.

Reference
Seki, A. et al., Development of Radionuclide Distribution Database and Map System on the Fukushima Nuclear Accident, Progress in Nuclear Science andTechnology, vol.4, 2014, p.47-50.

16

JAEA R&D Review 2014

Research and Development Relating to the Accident at the TEPCO’s Fukushima Daiichi NPS

1-6 Simulating the Migration Behavior of Radioactive Cesium in the Environment− A Case Study of the Ogaki Dam Reservoir −(a)(b)
Sediment transport (Ogaki Dam Basin)(ty−1)

Ogaki Dam
Reservoir

0.0E+00

4.0E+03

8.0E+03

1.2E+04

To rivers
To ocean
Deposit in dam

Ogaki Dam Basin

Fig.1-13 (a) The Ogaki Dam Reservoir (map drawn using DEM data provided by the Geophysical Survey Institute)and (b) the simulation results for the amount of sediment movement in the basin
A large amount of soil erodes and moves as sediment during a flood, entering into water systems, such as rivers andreservoirs. Most of the sediment is then deposited on the reservoir bed. The results indicate the importance of the reservoir
to Cs migration.

(c)

(d)

Amount ofdeposition (m)
1.00

(e)

Amount ofdeposition (m)0.250

Amount ofdeposition (m)0.0750

0.800

0.200

0.0602

0.600

0.150

0.0454

0.401

0.101

0.03060.201

0.05080.01580.00100

0.00100

0.00100

Fig.1-14 Examples of sediment deposition in terms of (c) sand, (d) silt, and (e) clay on the reservoir bed (mapdrawn using DEM data provided by the Geophysical Survey Institute)
Almost all of the sand and silt are deposited on the reservoir bed, but some clay leaves the reservoir and travels downriver.

Radioactive cesium (Cs), which was deposited on land after
the accident at the TEPCO’s Fukushima Daiichi NPS (1F),
adheres strongly to soils, especially if they are rich in clay.
Its migration is mainly caused by soil erosion and sedimenttransport within surface water during times of widespreadflooding. Although the alleviation of all sediment movement is
impossible, adequate operation of dam reservoirs may mitigate
the dispersion of sediment-absorbed radioactive Cs.
We simulated the movements of sediment and radioactive
Cs in the Ogaki Dam Reservoir in Fukushima Prefecture using
several simulation models to understand sediment migration
behavior during floods. We chose this reservoir because it is locatedin the middle of the Ukedo river, the river which was mostaffected in terms of radioactive Cs by the accident at 1F.
Fig.1-13 shows the Ogaki Dam Basin and the simulation
results for the amount of sediment that entered into the river
system, exported to the ocean, and deposited on the reservoir
bed. This simulation was performed using our watershed model,
the Soil and Cs Transport model. As can be seen from thefigure, almost all of the sediment that entered the upstream partof the river was eventually deposited on the reservoir bed.
Fig.1-14 shows the results for the amounts of sand, silt, andclay that were deposited on the reservoir bed 3 days after atypical flood event. This simulation was performed with the
Nays 2D code. Almost all the sand and silt was deposited on
the reservoir bed. However, because of the size difference,
the locations of the largest deposits of each sediment grade
differed. The sand was mainly deposited near the entrance ofthe reservoir and the silt was deposited almost everywhere
under consideration. Clay, on the other hand, remained
suspended in the reservoir for longer. More than half of it was
eventually deposited on the reservoir bed, while the rest left
the reservoir and flowed downriver in our simulation.
In addition, we have simulated the same problem using
different water heights within the reservoir. When the waterlevel was lowered, the sand and silt generally behaved in
the same manner as above. However, more than half of theclay passed through the reservoir and flowed downstream ina shorter time than it took in the previous case. The results
indicate that adjusting the water height in the reservoir may
provide a countermeasure against radioactive Cs migration
from, or accumulation in, the reservoir. At present, intensive
field investigations are ongoing with the goal of validating oursimulation results and refining the model as necessary.

Reference
Kitamura, A. et al., Predicting Sediment and Cesium-137 Discharge from Catchments in Eastern Fukushima, Anthropocene, vol.5, 2014, p.22-31.

JAEA R&D Review 201417

Research and Development Relating to the Accident at the TEPCO’s Fukushima Daiichi NPS

1-7 Chemical Bonding between Radioactive Cesium and Clay Minerals Revealed by Supercomputer− First-Principles Calculation Studies of the Chemical Adsorption Form of Cesium on Clay Minerals −(a)

KElectron Density: Oxygen
K : Potassium: Electron

K

No Covalent Bonding(Ionic Bonding Only)

Muscovite(b)

Cs

From lower position

Electron Density: OxygenCs : Cesium: Electron

Cs
Muscovite

Electron Shearing(Ionic and Covalent Bonding)

From higher position

Fig.1-15 Model of clay minerals and electron density(a) A typical muscovite model and the electron density of K. (b) Cs-adsorbed muscovite and the electron density of Cs.

Large amounts of radioactive nuclides were released into
the environment by the accident at the TEPCO’s Fukushima
Daiichi NPS in 2011. The relatively short-life nuclides likeradioactive iodine have almost entirely decayed, but theradioactive cesium (Cs) still remains in the topsoil, and it willremain a primary γ-emitter source for a long time.
It is well-known that Cs is strongly adsorbed onto soil
and is difficult to extract from it. At present, the governmentis operating large-scale decontamination by removal of thecontaminated soil and storing it. However, storage sites
are extra environmental burden. Therefore, there is a great
demand for the development of efficient separation methodsof Cs from soil.
To achieve the above purpose, it is important to clarify detailsof the form of chemical adsorption of the Cs onto the soil,e.g., which ingredient in the soil is essential for adsorption
and chemical bonding. If we obtain such information, we
can select one of the most efficient decontamination methods
based on scientific ideas and approach the final goal at areasonable speed.
In general, the adsorption process of Cs on topsoil isregarded as follows. First, Cs takes the form of a dissolved
ion just before the fallout. Afterwards, the dissolved Cs is
selectively and irreversibly adsorbed by micaceous clay

minerals through ion exchange with the potassium (K)preoccupied by these minerals. However, the chemical form
produced by the adsorption as yet remains unknown. Thus, westudied the form using a numerical technique called a “firstprinciples calculation method” using a JAEA supercomputer.
We selected muscovite as a target model system in clay
minerals and evaluated the electronic states of muscovite
together with cations (K and Cs) as shown in Fig.1-15.
Consequently, we found that the chemical bond between
the oxygen in muscovite and Cs is a combination of ionicand covariant although one between the oxygen and K is
only ionic (Fig.1-15). The covalent bonding is caused by
the sharing of electrons between bonded atoms. Due to thiscounterintuitive character, the bonding strength is relatively
enhanced. The condition necessary for making a covalent bondis the closeness of energies of the electron orbitals between
the bonding atoms. Indeed, we found that the electron orbitalsof oxygen and Cs are very close in terms of energy, whichis in marked contrast to the clear orbital energy difference
between oxygen and K. We are now trying to reveal methods
for cutting the special bond between Cs and clay minerals
to reduce the volume of waste soils produced by large-scaledecontamination efforts in Fukushima.

Reference
Okumura, M. et al., First-Principles Studies on Cesium Adsorption of Clay Minerals; Mechanism and Chemical Bonding, Nippon Genshiryoku Gakkai-Shi,
vol.56, no.6, 2014, p.20-25 (in Japanese).

18

JAEA R&D Review 2014

Research and Development Relating to the Accident at the TEPCO’s Fukushima Daiichi NPS

1-8 Translocation of Radioactive Fallout Cesium in Trees− Analysis of the Distribution of Radioactive Cesium in Plants Using Autoradiography −Leaves grown after the accident

Leaves, male
flowers and braｎchesgrown after theaccident

Leaves grown
before the accident

Branches andfemale flowers
grown before theaccident

Fig.1-16 Autoradiograph (AR) image of Torreya nucifera
sampled after 2 months of the accident(left) Photograph of T. nucifera, (right) AR image. Black spots in
the AR image show radiation sources in branches and leaves,indicating the presence of radioactive Cs. Red circles indicate
the leaves (light green colored leaves in the photo) that grew
after the accident. Black spots are practically zero within the redcircles, indicating that the radioactive Cs in the old leaves (yellow
arrow) was rarely transported into the young leaves.

Fig.1-17 AR image of Cryptomeria japonica sampled after
about 2 years of the accident(left) Photo of C. japonica, (right) AR image. Blue circles indicate
the branches, female flowers, and old leaves in the photo (yellow
arrows) that grew before the accident. Red circles indicatethat male flowers and leaves that emerged after the accident
contain some amount of radioactive Cs inside them.

How were radionuclides attached to and transported inforest plants and soil after the fallout from the accident at theTEPCO’s Fukushima Daiichi NPS? This question must be
resolved to quantify present and future forest contamination.
Distributions of radionuclides in plants and soil samples
are usually measured by cutting small fractions from thesamples. Unfortunately, this technique requires considerable
time and effort. Thus, we have applied an autoradiography
technique to analyze the distribution of radionuclides within
the plants and soil.
We collected Torreya nucifera and Cryptomeria japonica
samples in Iitate-mura, Soma-gun, Fukushima in May 2011
and Okuma-machi, Futaba-gun, Fukushima in December 2012,
respectively. We placed the samples on an imaging plate (IP)coated with fluorescent paint that reacts with radiation andphotographed the spatial distribution of radionuclides. The
distribution of radionuclides in the branches and leaves ofT. nucifera (Fig.1-16(left)) was analyzed by autoradiograph (AR)images, where many black spots were present (Fig.1-16(right)).
These black spots appeared on IP in response to radiation
originating from radioactive Cesium (Cs). Note that 131I was

decayed out in May. Fig.1-16(left) illustrates dark and lightgreen leaves that grew before (old) and after (young) the
accident, respectively. Black spots were present on the darkgreen leaves but not on the light green leaves, indicatingthat radionuclides attached to the old leaves were rarelytransported to the young leaves. The same result was obtained
from analysis of Cryptomeria japonica.
The AR image (Fig.1-17(right)) in C. japonica (Fig.1-17(left))shows many black spots on the branches and the female
flowers that emerged before the accident. On the contrary,broadened black areas were observed on the leaves andfemale flowers that emerged after the accident, indicating thatradioactive Cs was translocated from the tree surface to thenew branches and leaves. Since the top of the leaves and male
flowers are actively growing, and since K ions are known tobe accumulated in the active areas, radioactive Cs may havebeen transported from the old branches to the newer area.
These results clearly show that radioactive Cs deposited on
the tree surface is slowly translocated to new growing areas
over timescales of the order of years.

Reference
Sakamoto, F., Ohnuki, T. et al., Determination of Local-Area Distribution and Relocation of Radioactive Cesium in Trees from Fukushima Daiichi Nuclear
Power Plant by Autoradiography Analysis, Nippon Genshiryoku Gakkai Wabun Ronbunshi, vol.12, no.4, 2013, p.257-266 (in Japanese).
JAEA R&D Review 201419

Research and Development Relating to the Accident at the TEPCO’s Fukushima Daiichi NPS

1-9 Easy Monitoring of the Radioactive Cesium Concentration in Water− A New Assay System Using a Cesium Adsorption Disk and the GM Survey Meter −waterproof bag
plastic support stickwater-filled tank
GM survey meterCs diskFig.1-18 A new assay system for monitoring the radioactive
Cs concentration in water
Radioactive Cs in water is adsorbed onto a Cs disk, and its
radioactivity is measured with a GM survey meter. The Cs disk
is then fixed onto the probe of the GM survey meter and sunk ina water-filled tank to decrease the background count rate.

Table 1-1 Application to outdoor school swimming poolwater decontamination
Our assay system was applied to the decontamination of an
outdoor school swimming pool in Fukushima in July 2011. As aresult, it was quickly confirmed at the site of the decontaminationthat the pool water treated had a radioactive Cs concentrationof under ~60 Bq/ℓ, which was below the effluent standard valueof 200 Bq/ℓ at that time. This demonstrated the viability of ourassay system.

Date(2011)

Cs concentration(Bq/ℓ)

Throughput ofpool water(m3)

July 2<37

0-65

July 948±965-13156±10

131-18362±13

183-235<30

235-236

July 10

July 11In areas contaminated by the discharge of radioactive cesium

sensitivity to β-rays, an easy method of use involving only(Cs) from the accident at the TEPCOʼs Fukushima Daiichi

portable apparatuses, and applicability to water samples

NPS, continual monitoring of water and decontamination

contaminated only with 134Cs and 137Cs. In addition, because

activities have been conducted. These field operations

the calibration factor of 137Cs is larger than that of 134Cs, thedemand an easy method for monitoring the radioactive Cs

use of the former factor led to conservative evaluations.

concentration in water.

In the decontamination of an outdoor school swimming

We developed a new assay system using a cesium adsorption

pool in Fukushima in July 2011, our assay system was applied

disk (Cs disk) and a Geiger-Mueller (GM) survey meter forto the determination of the radioactive Cs concentration in thefield operations (Fig.1-18). This assay system comprisedpool water. About 100 mℓ of the samples collected from the(1) the adsorption of radioactive Cs on a Cs disk by filtering

pool water had a radioactive Cs concentration below ~60 Bq/ℓ.

of water, (2) the determination of the β count rate of the Cs

Since this value was below the effluent standard value at thatdisk using the GM survey meter, and (3) the calculation of Cs

time (200 Bq/ℓ), the decontaminated pool water was pumped

radioactivity on the basis of a calibration factor. The Cs diskout into a nearby river (Table 1-1).

(Sumitomo 3M Co. Ltd., Empore

TM

Cs Rad Disk) contains Cs

adsorbent in its matrix.
Some features of our assay system included: high detection

Our assay system is expected to be applied to the monitoringof tap water, because its detection limit reaches 1 Bq after
about 10 min of the measuring time.

Reference
Nagano, T. et al., Method for Monitoring Radioactive Cesium Concentration in Water using Cesium Adsorption Disk and GM Survey Meter, Nippon
Hoshasen Anzen Kanri Gakkai-Shi (Japanese Journal of Radiation Safety Management), vol.11, no.2, 2012, p.139-145 (in Japanese).

20

JAEA R&D Review 2014

Research and Development Relating to the Accident at the TEPCO’s Fukushima Daiichi NPS

1-10 Assessment of Radiation Dose Reduction of Radioactive Cesium in Buildings− Numerical Simulation Analysis Using Building Models −(c) Wooden house(a) Wooden house
Tiled roof(e) Wooden houseRF
0.8

Bathroom

Wooden board

WC
0.6

LDK
Room1

Room2
0.4

Glass window(b) Hospital

Window (f) Hospital(d) Hospital
Glass window

Kitchen

RF
0.3

X-ray
room

Conference

room

0.2

Consulting
room

Officeroom

0.1

Concrete

Treatmentroom

Fig.1-19 Outside appearance of building
modelsBuilding models were constructed in three
dimensions.

Fig.1-20 1st floor layouts in building
modelsBuilding models were constructedaccounting for room compartments.

Radioactive nuclides were discharged by the accident at
the TEPCO’s Fukushima Daiichi NPS (1F) and radioactive
cesium (Cs) still remains in the environment. Precise radiationdose assessment is important for public health and for plans
for the return of evacuees. Information about the dose ratemeasured in the open air has been made available for eacharea by radiation monitoring, which has been conducted since
the accident at 1F. On the other hand, residents spend large
amounts of their daily lives inside buildings, where dose rates
are reduced compared with their values in open air. Thus,
evaluation of dose reduction in various buildings is required
for the assessment of external radiation dose exposure
accounting for individual lifestyles.
We selected 27 types of buildings based on surveys ofbuildings in the Fukushima area, and investigated the dose
reduction in these buildings by means of numerical simulation
analysis. These selected buildings were constructed as threedimensional models (Fig.1-19) with consideration of theirinner structures and compartmentalization of their rooms
based on their usages (Fig.1-20). By incorporating the building
models into the radiation transport code, PHITS, the behaviors

Fig.1-21 Distribution of the radiationdose RF
While RF gradually decreases fromthe outer walls to the center of wooden
houses, γ-ray contributions from
windows are essential in concretebuildings.

of γ-rays were simulated from radioactive Cs homogenously
distributed over the ground surface to inside the buildings.
We computed the dose ratios of the inside of the buildings tothe outside, the so-called reduction factor (RF), with thesesimulations. In addition, the distribution inside the buildings
was illustrated on top of the layout (Fig.1-21) to analyze theinfluence of building’s configuration on RF.
In wooden houses, the radiation dose decreases towards thecenter (Fig.1-21(e)) and RF becomes lower for a house with alarger size. This tendency is due to the absence of radioactive
Cs beneath the house. In contrast, the dose rates insideconcrete buildings are lower than those inside wooden houses
because of the high shielding effect of concrete walls. The RF
distribution in concrete buildings is largely influenced by thelocation and size of the windows, which offer little shielding
effect (Fig.1-21(f)).
We developed a technique for analyzing the influence ofbuilding configuration on dose reduction in various buildings.
The obtained data are expected to be useful for dose estimation
and reduction countermeasures for returning residents.

Reference
Furuta, T. et al., Analyses of Radiation Shielding and Dose Reduction in Buildings for Gamma-rays Emitted from Radioactive Cesium in Environment
Discharged by a Nuclear Accident, JAEA-Research 2014-003, 2014, 100p. (in Japanese).
JAEA R&D Review 201421

Research and Development Relating to the Accident at the TEPCO’s Fukushima Daiichi NPS

1-11 Detection of Defects in the Fuel Cladding Tube
− Development of a Defect Detection Technique Using High-Resolution X-ray Computed Tomography −Zry-2 cladding tube

Slit of 0.1 mm width
Without corrosion layer
Corrosion layer

9000

Region of basic material (SUS)7000

Zry-2 cladding tube

1 mm

Slit

Fig.1-22 X-ray CT image of a zircalloy-2 cladding tube
with a slit
A 0.1-mm-wide slit can be observed on the X-ray CT image.

CT value

Insidethe SUS tube

Outsidethe SUS tube

5000

3000

Decreasing
the CT value

1000

−1000

Basic material (SUS)0

Basic material (SUS)Corrosion layer

2

3

Measurement position of CT value (mm)Fig.1-24 CT value distribution of a SUS tube
The CT value decreases with the density. The CT value
distribution of the SUS tube with a corrosion layer was lower
than that of the SUS tube without the corrosion layer.

Measurement
position of CT
value

Corrosion layer

1

1 mm

Fig.1-23 X-ray CT image of a SUS tube with a corrosion layer
The corrosion layer could be discriminated from the basic
material of a SUS tube.

The fuel assemblies of the TEPCO’s Fukushima Daiichi
NPS (1F) will be transported from the spent fuel pool to theshared pool and stored for a long time. In the severe accidentof 1F, concrete fragments fell into the spent fuel pool dueto the hydrogen explosion, and sea water was temporarily
injected to cool the fuel assembly. There is a possibility ofdefects (the corrosion and the crack) forming on the fuelassembly component materials (cladding tube, tie plate,spacer, support grid, etc.) due to the effects of the concrete
fragments and sea water. Thus, it is necessary to evaluate theintegrity of the fuel assemblies to safely transport from thespent fuel pool and store in the shared pool.
In this study, we developed a technique for detecting cracks
and the corrosion damage on the fuel cladding tube using an
X-ray computed tomography (CT) technique, which is thenon-destructive way to examine the irradiated fuel assembly.
Zircalloy-2 (Zry-2) cladding tube specimens were
manufactured with slits designed to simulate cracks, and an
X-ray CT examination was performed on them to obtain data
about crack-detection performance with this apparatus. In

addition, SUS tube specimens having corrosion layers were
manufactured and also subjected to X-ray CT examination
to obtain basic data about detection performance on the
corrosion layer. The 0.1-mm-wide slit on the Zry-2 cladding
tube could be observed on the X-ray CT image in Fig.1-22,
and a corrosion layer (gray region) with a thickness of morethan 0.35 mm could be discriminated from the basic material(white region) in the X-ray CT image in Fig.1-23. An X-ray
CT examination of the Zry-2 cladding tube based on this result
has been conducted.
The X-ray CT image is composed of CT values, which
depended upon the density. This image analysis technique
was developed to evaluate the slit width and the corrosion
thickness using the CT value distribution (Fig.1-24). The
defect on the fuel cladding tube can be evaluated without
dismantling of the fuel assembly by using this image analysis
technique. At present, a high performance image analysis
technique is being developed to increase the evaluation
accuracy.

Reference
Ishimi, A. et al., X-ray CT Basic Data about Inspection of Irradiated Fuel Assembly, JAEA-Data/Code 2014-012, 2014, 72p. (in Japanese).

22

JAEA R&D Review 2014

Research and Development Relating to the Accident at the TEPCO’s Fukushima Daiichi NPS

1-12 Laser Spectroscopic Analysis of Underwater Debris in a Nuclear Reactor Core
− A Contribution to the Decommissioning of the TEPCO’s Fukushima Daiichi NPS −

(a)

N2

Optical fiber
Gas tube

Laser

LIBS Probe
Probe
head
SampleWater

Spectrometer

Emission intensity(a.u.)(b)in waterwith gas flowin airin water
800

820

840

860

880

800

820

840

860

880

Wavelength (nm)Fig.1-25 Prototype of the transportable LIBS apparatus
This instrument, whose size is 1.2 m (W) × 1.5 m (D) × 0.5 m (H),is composed of lasers, optical fibers, a LIBS probe, and aspectrometer.

Fig.1-26 Typical experimental set-up and results(a) Image of the experimental set-up in the LIBS measurement.
(b) LIBS emission spectra of zircaloy in air and water (left) andin water with gas flow (right).

To decommission the TEPCO’s Fukushima Daiichi NPS(1F), environmental information is needed about the insideof the post-accident reactor core. In 1F, molten fuel debris(a mixture of melted fuel core material, fuel cladding, andconstruction material) might be submerged in water inside thereactor core. There is not enough space for the direct inspectionof the reactor core, because many pipes for water andelectricity occupy the inside. Moreover, a high radiation fieldhinders people’s access. Thus, a remote sensing technique thatis applicable in narrow spaces is crucial for the inspection of 1F.
For this purpose, we have proposed a combination of afiber optic transmission system and laser-induced breakdown
spectroscopy (LIBS), which is an analysis technique thatcomprises ablating a material with a focused laser beam anddetecting emission from the ablation plasma. A prototypeof the transportable LIBS apparatus is shown in Fig.1-25.
Among the most pivotal parts was the fiber optics, which
needed to have high resistivity against radiation. To test theradiation resistivity, the fiber optics employed in our apparatus

were exposed to 60Co γ-rays with a total dose of 1.6 MGyin the Takasaki Institute. Comparing the optical conditions
before and after the radiation exposure, we confirmed thatthe transmissivity of the fiber optics in the infrared region(730 –1100 nm) was not degraded by the irradiation, andthus, infrared can be used for the laser and the emission
transmission.
In addition to radiation resistivity, the LIBS analysis ofunderwater debris demands a special technique, since water
quickly quenches the plasma emission that is induced by laser
ablation. Hence, we attached a gas-flow system to the headof the LIBS probe and performed the LIBS analysis undera quasi-atmospheric environment in water (Fig.1-26(a)).
Fig.1-26(b) is the emission spectrum of zircaloy in water using
the gas flow system.
Towards in situ observation and analysis of the debris, weare improving the system that comprises our LIBS apparatus
combined with radiation measurement and image detection
through optical fibers.

Reference
Saeki, M. et al., Development of a Fiber-Coupled Laser-Induced Breakdown Spectroscopy Instrument for Analysis of Underwater Debris in a Nuclear
Reactor Core, Journal of Nuclear Science and Technology, vol.51, issues 7-8, 2014, p.930-938.
JAEA R&D Review 201423

Research and Development Relating to the Accident at the TEPCO’s Fukushima Daiichi NPS

1-13 Melting and Crushing of Fuel Debris with Laser Light as a Heat Source− Countermeasures to Fuel Debris Characterized by Indefinite Shapes,
Higher Hardness, Multi-Ingredients, and Porous Bodies −(a) The establishment of a laser optical
path by an assist gas flow
Alumina Pellet

m
m
82

20 mm
12 mm

45
mm

Laser
Irradiation
Condition

Accuracy: ±0.01 mm
Accuracy: ±10 ℃

Two-color
Thermometers

Laser
Scanner

Gas Flow Rate

Resolution: ±0.01 nm

LIBS

••••

Environmental
Sensors

Assist Gas Flow(b) A snapshot of the melting andcrushing process in water
Laser Light

Laser Processing Head

Adaptive Controller (PC)

ee

p

Ve
l

oc

it y

Gap Length

SUS 304 Plate

Sw

Reflection
Lights

10 mm

m
m

100 mmtG
L a asA s s e r F l ow
sis Lig st G ht
as
Fl
ow
s

sis
As

Surface TemperatureEnvir
Se onm
ns en
or ta
s
l

Surface Shape,
Material Properties,

Laser
Processing
Head

89

120
mm

mm
230

Adaptive Control System

Laser Processing Head

Robot
Controller

..

z

q
y

Heat Conduction

Liquid Metal

Laser
Controller

Assist Gas
Controller

SUS 304 Plate••••

Simultaneous
Controller
Robot Arm with Adaptive Control Systemx
Fig.1-27 An adaptive control system using
laser light for melting and crushing fuel debris
The system is adaptively controlled toward
suitable laser irradiation conditions through theuse of various environmental sensors.

Liquid Metal

Fig.1-28 Block diagram of the adaptive
control system
The system prevents decreases in themelting and crushing performances bysimultaneous control of the robot’s motion,
the laser irradiation, and the assist gas flow.

Laser light is characterized by higher power density, local
processability, remote controllability, and so forth, and can beused to melt and crush various materials, such as fuel debris,which do not depend upon fracture toughness.
An adaptive control system for yielding suitable laser
irradiation conditions, shown in Fig.1-27, is being developed
using various environmental sensors, such as a laser scanner
to catch indefinite shapes, a LIBS sensor to identify multiingredient materials, and a two-color thermometer to monitor
melting conditions in laser irradiated positions for the removalof fuel debris and in-vessel structures from the TEPCO’s
Fukushima Daiichi NPS (1F).
Fig.1-28 indicates a block diagram of the adaptive control
system. The system comprises environmental sensors; apersonal computer containing the adaptive control algorithm;
and a simultaneous controller for robot motion, laser
irradiation, and assist gas flow. While control procedures are
employed in the system to obtain suitable laser irradiation
conditions, behaviors in the course of the laser melting process
are monitored by checking the two-color thermometer signals
every time step. Thus, suitable laser irradiation conditions are

Fig.1-29 A melting and crushingexperiment in water
It was confirmed that the system
has the potential to melt and crush
fuel debris in water.

determined from control rules based on the response surfacefunction.
The adaptive control system was applied to a melting
and crushing water experiment using a combined test piececomprising a SUS304 plate (20 mmt) and an alumina pellet(8 mmO.D. × 10 mmh) simulating fuel debris. Fig.1-29 shows
experimental photos. A steady-state assist gas flow field of
350 ℓ/min was established as a laser optical path before being
used to melt and crush transients (Fig.1-29(a)). In the transient
experiments, the SUS304 plate and the alumina pellet of the
combined test piece were irradiated by continuous laser lightat 6 kW and pulsate laser light over 100 ms, respectively. The
sweep velocity of the laser processing head was kept constantat 30 mm/min. As shown in Fig.1-29(b), the combined test
piece was melted and crushed well by the use of the adaptive
control system.
The results obtained from the water experiment are very
encouraging in the sense that the adaptive control system may
be used as an efficient tool for removing fuel debris and invessel structures from the 1F.

References
Muramatsu, T. et al., Research on Removal Technologies of Fuel Debris and In-Vessel Structures using Laser Light (I) ‒Research Plan and Research
Activities on FY2012‒, JAEA-Research 2013-024, 2013, 49p. (in Japanese).
Muramatsu, T. et al., Research on Removal Technologies of Fuel Debris and In-Vessel Structures using Laser Light (II) ‒Research Activities on FY2013‒,
JAEA-Research 2014-018, 2014, 41p. (in Japanese).

24

JAEA R&D Review 2014

Research and Development Relating to the Accident at the TEPCO’s Fukushima Daiichi NPS

1-14 Maintaining the Integrity of the Reactor Vessel until the Completion of Fuel Removal− The Influence of Radiation on the Corrosion of Vessel Steels and Corrosion Measures −Cooling water

60

Co sources cover

Condenser
Thermocouple

Flask
4.4 kGy/h
0.2 kGy/hSpecimens

60

Co sourcesrays
γγ-rays
-rays
1.6 m

Diluted seawater (50 ℃)
HeaterCorrosion weight loss (mg/cm2)

20

Test temp.: 50 ℃
Test time: 500 h
PCV steel: SGV480

15
Decrease of
corrosion weight loss
10
Further decrease
50
No N 2H4
In air

No N 2H4
In N 2

100 mg/ℓ N 2H4
In N 2

Seawater was poured into reactors 1F1-1F3 after theaccident at the TEPCO’s Fukushima Daiichi NPS (1F).
Therefore, the primary containment vessel (PCV) composedof carbon steel will be corroded by contact with seawaterover the long term, and there is a possibility of the loss ofstructural strength and/or the formation of holes in the vessel.
Furthermore, the decomposition of water molecules due tohigh doses of radiation (i.e., water radiolysis) in the PCV
causes the production of chemically active species, such ashydrogen peroxide, which could accelerate the corrosionof the vessel steel. It is, therefore, necessary to predict andcontrol the corrosion of the steel until the completion of theremoval of nuclear fuel and debris.
As shown in Fig.1-30, specimens of SGV480 steel–identicalto the actual steel used in the PCV–were immersed in diluted
artificial seawater and irradiated by γ-rays for 500 h. Fig.1-31
shows the results of the corrosion test under irradiation.
“Corrosion weight loss” on the vertical axis is the weight lost by

Fig.1-30 Corrosion testing of steel specimens
under γ-ray irradiation
To simulate the conditions inside PCVs, corrosion
testing of steel specimens in diluted seawater
under γ-ray irradiation was conducted.

4.4 kGy/h irradiation
0.2 kGy/h irradiation
No irradiation

Fig.1-31 Result of the corrosion test of PCV steel
Measures to suppress the corrosion of steel by nitrogen
substitution of the gas phase of PCV and the addition ofchemicals were examined. The nitrogen substitution in theflask together with hydrazine addition could suppress corrosion
to a high degree.

the specimen due to corrosion, and was obtained by removing
the rust from the specimen, weighing it, and comparing theresult to that obtained before the specimen’s immersion. Afterthe accident at 1F, nitrogen gas was enclosed in PCVs for the
purposes of preventing hydrogen explosions. Corrosion tests
under irradiation in the test flask in which nitrogen gas was
enclosed showed a decrease of the weight loss, as shown in
Fig.1-31. Since the oxygen partial pressure of the gas phasein the flask decreased through nitrogen substitution, theconcentration of dissolved oxygen in the diluted seawater also
decreased, and consequently the corrosion rates of specimens
decreased as well. Furthermore, the case that used nitrogen
substitution together with the addition of hydrazine with theproperty of removing oxygen from water is shown on theright-hand side of Fig.1-31. From this result, it is believedthat the corrosion of PCVs, wherein nitrogen is enclosed andhydrazine has been injected, is still under control at present.

Reference
Nakano, J. et al., Effects of Hydrazine Addition and N2 Atmosphere on the Corrosion of Reactor Vessel Steels in Diluted Seawater under γ-rays Irradiation,
Journal of Nuclear Science and Technology, vol.51, issues 7-8, 2014, p.977-986.
JAEA R&D Review 201425

Research and Development Relating to the Accident at the TEPCO’s Fukushima Daiichi NPS

1-15 Estimation of the Chemical Form of Fuel Debris in a Reactor Pressure Vessel− Estimation with the Help of Thermodynamic Calculation −Increase in P(O2)• Oxidation of Zr, Fe, andUO2 in this order

Increase in temperature• Dissolution of Zr and Fe in UO2
matrixCommon logarithm of oxygen partial pressure
log10P (O2)(atm)(U,Zr,Fe)O2,(Zr,U)O2, Liq (O-rich)(U,Zr,Fe)O2,
ZrO2, FeOx.

−10・(U,Zr)O2-cubic・(Zr,U)O2-tetragonal

steam dryer

0

U4O9, U3O8,
ZrO2, FeOx.

core region (middle part)

RPV

Liq (O-rich)steam separator(U,Zr,Fe)O2, (Zr,U)O2,
Liq (Fe-rich)

top guide・Zr(O)
metaloxidemass fraction of
metal and oxide(U,Zr)O2,(Zr,U)O2, Fe−20

core region (lower part)UO2, ZrO2, Fe

Zr (O),Liq (O-poor)−30

Liq (O-poor)(U,Zr)O2, (Zr,U)O2,
Fe2 (Zr,U), Zr(O)−40

Zr-U alloy,
FeaUbZrc,
Liq (O-poor)

lower plenum(almost intact)

Low P(O2)• Reduction of U
by Zr
• Formation of Fe-ZrU alloy・Zr(O)
・Fe2(Zr,U)・Fe3UZr2

・(U,Zr)O2-cubiccore plate

oxide
metal

Color index for Zr amount
large

small

mass fraction of
metal and oxide

−50
800

1000

1200

1400

1600

1800

2000

2200

2400

2600

2800

3000

3200

Temperature (K)–Details for calculation–• Composition: 65wt% UO2; 27wt% Zr; 8wt% Fe• Software: FactSage6.2 (http://www.factsage.com/)• Database: TDnucl (http://www.crct.polymtl.ca/fact/documentation/)

Estimated damage status in
RPV after core degradation(The larger the amount of Zr
becomes, the darker the cellis indicated)–Result of thermodynamic calculation–
The mass fraction of metal/oxide andthe chemical forms of fuel debris

Fig.1-32 Calculated phase diagram of temperature vs.
oxygen partial pressure (P(O2)) for the UO2-Zr-Fe system
The chemical form of the fuel debris is expected to change
depending upon the temperature and P(O2) during the melt
progression of the BWR core. Oxygen partial pressure aswell as temperature is expected to increase with the melt
progression.

Fig.1-33 Chemical form estimation for in-vessel fuel debris
after core degradation
For 1F2, the spatial distribution of core materials andtemperature were evaluated based on the results of melt
progression analysis (left side of figure). Based on thisinformation, the phase mass fraction under equilibrium wasestimated (right side of figure).

To prepare for the removal of fuel debris at the TEPCO’s
Fukushima Daiichi NPS (1F), data on the physical propertiesof fuel debris are necessary. However, the actual situation
inside the damaged core of 1F is hardly known. In this study,
the chemical forms of fuel debris were estimated throughthermodynamic calculation.
As a preliminary evaluation, the effects of temperature andoxygen partial pressure (P(O2)) on the chemical forms of thefuel debris were investigated under constant composition of
the core materials (UO2, Zr, and Fe). As shown in Fig.1-32,
at low P(O2), the metal Zr tends to react with UO2 and Fe.
With an increase in P(O2), the core materials are oxidized in
the order Zr, Fe, and UO2. Metallic phases, such as Zr(O) andFe2(Zr,U), are expected to be formed as long as Zr is not fully
oxidized. In addition, an increase in temperature is expected
to cause the formation of mixed oxides, such as (U,Zr)O2, and(U,Zr,Fe)O2.
The composition of the core materials in the RPV after

core degradation was evaluated from the results of the melt
progression analysis conducted by Ishikawa et al (2012 Fall
Meeting of the Atomic Energy Society of Japan). Based on
this information, a thermodynamic calculation was conducted.
Fig.1-33 shows the results of chemical form estimation for
1F Unit 2 (1F2). The middle part of the core region was
occupied by mixed oxides, such as (U,Zr)O2, while the lower
part near the core plate was rich in metal phases, such as
Zr(O) and Fe2(Zr,U). This trend is consistent with the resultof the preliminary evaluation described above. However, if
the RPV is severely damaged, the concrete from the pressure
containment vessel floor also needs to be considered.
We are accumulating data on mechanical properties suchas the hardnesses and fracture toughnesses of the typical
phases of fuel debris, including metal phases. Those data willcontribute to the appropriate selection of defueling methods
and tools.

Reference
Ikeuchi, H. et al., Suggestion of Typical Phases of In-Vessel Fuel-Debris by Thermodynamic Calculation for Decommissioning Technology of FukushimaDaiichi Nuclear Power Station, Proceedings of International Nuclear Fuel Cycle Conference (GLOBAL 2013), Salt Lake City, Utah, USA, 2013, paper 8174,
p.1349-1356., in CD-ROM.

26

JAEA R&D Review 2014

Research and Development Relating to the Accident at the TEPCO’s Fukushima Daiichi NPS

1-16 What Has Formed from the Melted Fuel and Control Blade?
− Investigating the Characteristics of Solidified Core Melt −

(a)

(b)

(c)

(f)

(g)

25 μm

25 μm

Metallic part (alloys and borides)

(d)
ZrB2

(e)

Borides(Fe,Cr,Ni)2B(U,Zr)O2 − as-melted

MagnifiedDioxidessintered

1 mm

UO2

(U,Zr)O2(Fe,Cr,Ni)2 (U,Zr)
Fe-Cr-Ni

0

50 μm

5

Alloys1015
Microhardness (GPa)20

Fig.1-34 The appearance and cross-sectional images of asolidified melt specimens
Photos (a)–(c) show the starting mixture, melted state,and solidified button, respectively. (d) and (e) are themicrophotographs of the cross section. The needle-like
precipitation of crystalline ZrB2 is found in the metallic part.

Fig.1-35 Comparison of the microhardnesses of phasesin the solidified melt specimens
Photos (f) and (g) show examples of indentation on (U0.5Zr0.5)O2
and ZrB2, respectively. Hardness increases in the order ofalloys, dioxides, and borides.

For the decommissioning of the accident at the TEPCO’s
Fukushima Daiichi NPS (1F), the removal of fuel debris
and its appropriate management are imperative missions.
Therefore, we have been studying the characteristics of fueldebris using simulated specimens.
The fuel debris formed in the reactor cores mainlycomprises materials from the fuel (UO2 and Zr) and control
blade (B4C and stainless steel). However, the chemical form ofboron in the fuel debris has not been well identified.
To simulate the solidified fuel debris in the reactor cores, somepelletized mixtures of B4C, stainless steel, Zr, and (U,Zr)O2(a mixed oxide of UO2 and ZrO2) were arc-melted under an
Ar atmosphere, as shown in photos (a)–(c) of Fig.1-34. The
chemical forms in the solidified specimens were analyzed
along the cross section (photos (d) and (e)). The specimens
likely consist of (U,Zr)O2 ceramic part and metallic part. The
latter is found to include the precipitation of borides expressed
as ZrB2 and (Fe,Cr,Ni)2B in the alloy matrix of Fe-Cr-Ni and(Fe,Cr,Ni)2(Zr,U) intermetallic.
To examine the influence of oxygen in the atmosphere,
the solidified specimens were annealed at 1500 °C in an Ar0.1%O2 mixed gas flow for 10 h. Consequently, the Zr and U
in the alloy, and the Zr in ZrB2 were oxidized to form a Zr-rich(Zr,U)O2 layer on the surface. The boron isolated from ZrB2
alternatively formed (Fe,Cr,Ni)2B inside the oxide layer.
As a basic mechanical property of fuel debris, themicrohardnesses of the phases in the specimen were measured
by a Vickers tester (Fig.1-35). It is found that the borides,
especially ZrB2, are extremely hard compounds in all thephases. The dense precipitation of borides in the fuel debris
may potentially be a barrier against machining tools forremoval operations.
The present study was sponsored by the Agency for Natural
Resources and Energy, Ministry of Economy, Trade andIndustry of Japan (METI).

Reference
Takano, M. et al., Characterization of Solidified Melt among Materials of UO2 Fuel and B4C Control Blade, Journal of Nuclear Science and Technology,
vol.51, issues 7-8, 2014, p.859-875.
JAEA R&D Review 201427

Research and Development Relating to the Accident at the TEPCO’s Fukushima Daiichi NPS

1-17 Management of Removed Fuel Debris− A Study of Treatment Scenarios for Fuel Debris after Removal −RemovalCollection

Transport

Pretreatment

Transport

Short term
StorageTransport

Cask
DebrisFuels debris

Canister

MCCI productFor large
piecesOthermaterials
・Tool
・Core
internals
・Others

Cask

Conventional
cask

Facility
Conventional
facility(Commonpool)

For finesForsuspendedparticulate

New caskNew
facility

Cask

Conventional
cask
Type
・Transport only・Transport/Storage
Condition in cask
・Dry/Wet
New cask
Type・Transport only・Transport/Storage
Condition in cask
・Dry/Wet

StoragefacilityDrystorageWet
storage

Conventional
cask
Type
・Transport only・Transport/Storage
Condition in cask
・Dry/Wet
New cask
Type・Transport only・Transport/Storage
Condition in cask
・Dry/Wet

Treatment
Long term
storage
DirectdisposalConventional
waste (TRU)

Stabilizing
treatment

New waste(HAW)

Aqueousprocessing
PUREX

Transport

Canisterstorage fordirect disposal

Conventional
waste (HAW/TRU)

New
methodNew waste(TRU)

Pyrochemical
processing

New waste
(HAW)

Cask

Final
disposalType 1waste
disposal(Isolation type)

Conventionalcask

GeologicaldisposalNew
cask

Type 2waste
disposal(Control type)
Near surface
disposal
Subsurface
disposal

Fig.1-36 Complete scenario ideas from the removal of fuel debris to final disposal
Various options exist from removal to final disposal.

Table 1-2 Characteristics of each treatment scenario idea
The treatment scenario concepts based on several assumed methods are prioritized based on the extent to which they use
existing technology.
Aqueous processing

Long term storage

Direct disposal

Stabilizing treatment

Definition

Storage until disposal
strategies are defined

Simple treatment for final
disposal without U/Pu recovery

Stabilizing treatment for final
disposal without U/Pu recovery

Present treatment for final disposal with positive U/Pu recovery

Basetechnology

Intermediate storage ofspent fuelDirect disposal of spent fuelSolidification of high-level waste

Aqueous reprocessing (PUREX)

OutlineUse simple treatment
Store it after
selecting a suitable
method from among
TMI-2 experiences
and short-term
storage methods
Basically postpone
the final disposal

Use simple treatment
Geological disposal after
storage until the decisionon the final disposal site is
madeDispose as just spent fuelwas removed or by adding
some treatment to it

Use simple treatment
Stabilize the waste to reduce
its leaching rate assuming it isvitrified
Avoid dissolution to reduce thegeneration of secondary waste

The removal of fuel debris from the TEPCO’s Fukushima
Daiichi NPS is scheduled to begin around 2020. Treatment(processing and disposal) methods for the removed fuel debris
are expected to be finalized during the 20–25 year period
following December 2011, when the emission of radioactive
material and its radiation dose were confirmed to have beensignificantly inhibited. Therefore, it is necessary to consider
an approach of comparative scenario evaluation and obtainfundamental information for selecting treatment methods by
the beginning of the removal. In this study, the characteristics
and technological problems of each treatment scenario wereextracted and put in order, in preparation for comparative
scenario evaluation in the future.
Several scenarios for the removal and disposal of fuel debris
are shown in Fig.1-36. Among these scenarios, “Treatment” isregarded as a pre-processing step toward the manufacturing ofa stable waste form for final disposal.
The selection of treatment methods was implemented forthe following policies (Table 1-2).
(1) Storage until disposal strategies are defined

PUREX

New method

Pyrochemical processing

Pyrochemical reprocessing(Metal electro-refining)

Convert into a manageable form of waste based on existing disposal classification
when possible
Use simple processing
with technical feasibility
Recover only nuclear
material that can berecovered easily

Give priority to the recovery of nuclear material and to
the reduction of radioactivity in waste(2) Simple treatment for final disposal without U/Pu recovery(3) Stabilizing treatment for final disposal without U/Pu
recovery(4) Present treatment for final disposal with positive U/Pu
recovery
Elemental technologies for each treatment scenario were
evaluated in terms of economy, amount of nuclear waste, andtechnological difficulties, on the premise that they functionin practice. Consequently, it has been found that long-term
storage has more advantageous aspects than the other methods,
both economically and technologically. Direct disposal, on
the other hand, has been found to produce the least amount ofwaste.
In summary, long term storage and direct disposal are
the most advantageous scenarios, though they have sometechnological problems. The stabilizing treatment, aqueous
processing, and pyrochemical processing options are inferior
to the rest in terms of economy and the amount of waste.
The present study was sponsored by the Ministry of
Economy, Trade and Industry of Japan (METI).

Reference
Fuel Debris Conditioning Technology Development Group, Fukushima Project Team et al., Fuel Debris Characterization and Treatment Technologies
Development for TEPCOʼs Fukushima Daiichi Nuclear Power Station ‒2012 Annual Research and Development Report‒, JAEA-Review 2013-066, 2014,
p.83-92 (in Japanese).

28

JAEA R&D Review 2014

Research and Development Relating to the Accident at the TEPCO’s Fukushima Daiichi NPS

1-18 An Exploration of the Reason for Iodine-131 Release after Core Cool Down
− Release of Iodine-131 from Contaminated Water in the Basement of Reactor Buildings −Table 1-3 Estimated fractions of 131I
in accumulated water to each core
inventory in late March

All the decay heat is assumed to
be given to accumulated water

Analytical areaof existing SA
codes, such as
MELCOR

ReactorPressure Vessel

pH=7.1; Ignored radiolytic
conversion from I– to I2

deposition–

Condensor
TrenchGas-liquid
partition
+

26

17

Csl

–

Cs , I , I2

3SPEEDIMELCOR
Simplified model

10

Steam
generation

Leak from somewhere of S/Cor Reactor Core Isolation
Cooling System74

18

Simplified model

I2

0.51

2

Accumulated water–

2I + hv ⇔ I 2 + 2e ← Move right with high dose rate (radiolysis)–+
I 2 + H2O ⇔ HOI + I + H ← Move right with high pH (hydrolysis)

Decrease in iodine conc.
at surface of accumulated
water due to fresh water
injection on March 26

10

16

10

15

10

14

10

13

10

⎧

I2

Release Rate of 131I (Bq/h)

Suppression
Chamber (S/C)

Turbine Building

1

12

10

Release by
gas-Iiquid partition

11

10

⎩

Containment

I (%)

⎨

Reactor Building

131

Unit

Ventilation
rate

I2

10

10

Fig.1-37 Schematic of contaminated water accumulationin the basements of reactor building and turbine building
and the release from the accumulated water considered inpresent simplified model
During the 1F accident, somewhere of the reactor coolant
system failed and contaminated water accumulated in thebasements of the reactor building. Since the release of 131I
from the contaminated water due to gas-liquid partition is notconsidered to be negligible, a simplified model of the releasefrom the contaminated water was prepared, as indicated in red.

Fig.1-38 A comparison of the 131I release rate between oursimplified model and SPEEDI reverse estimation
Although our simplified model did not predict an increase in therelease rate on March 23 and 30 due to shortages of cooling waterand re-melt, the model mostly reproduced the continuous releasefrom the accumulated water after March 17 and the decrease in
the release rate due to fresh water injection on March 26.

During the process of core cooling in the accident at theTEPCO’s Fukushima Daiichi NPS (1F), large amounts ofcontaminated water accumulated in the basements of thereactor buildings (Fig.1-37). The present study estimated thequantity of I-131 (131I) in the water as of late March in 2011
based on the data made available to the press. The estimation
showed that a significant fraction of the 131I inventory was
dissolved in water (Table 1-3).
Most of 131I released from the fuel would be dissolved as I−
in water, while some of the I− could become I2 and be released
as a gas phase due to the gas-liquid partition in the case ofacidity. This and the ventilation of the reactor buildings resultin an increase in the release of 131I into the environment.
Many analyses of 131I release have been attempted, mainly
using the two methods. One is reverse estimation using theatmospheric dispersion code SPEEDI combined with theenvironmental monitoring data. The other is estimation using
severe accident analysis codes, such as MELCOR. No large
difference was found in the cumulative release to environment
between the two methods. However, SPEEDI predicted
continuous release for longer than one month, whereas
MELCOR predicted that it would stop after March 17 due tothe decrease in release from the cooled core (Fig.1-38).

In the present study, we prepared a simplified model of 131I
release from the accumulated water, and tried to simulate thecontinuous 131I release predicted by SPEEDI after March 17.
The pH in the water after March 17 was assumed to beconstant at the value of 7.1 measured on March 27. The
ventilation rate of the reactor building was determined fromthe steam generation rate, assuming that all the heat from thedecay is given to the water. The decrease in the I2 release at thetime of fresh water injection for core cooling on March 26 wassimulated by the stratification of water, due to difference in
specific gravity between fresh and sea waters. The simplified
model well-reproduced the release baseline predicted by
SPEEDI (Fig.1-38).
The present study showed that the 131I release from theaccumulated water may explain the release between March
17 and 26. This strongly suggests a need for improvement of
the current MELCOR approach, which treats the release only
from containment breaks.
We plan to further improve the present simplified modelthrough detailed investigation of the effect of revaporization
or resuspension of 131I once-deposited on the wall surface, as
well as the stratification of accumulated water at the time of
fresh water injection.

3/8

3/13

3/18

3/23

3/28

4/2

4/7

Date in 2011

Reference
Hidaka, A. et al., Quantities of I-131 and Cs-137 in Accumulated Water in the Basements of Reactor Buildings in Process of Core Cooling at Fukushima
Daiichi Nuclear Power Plants Accident and Its Influence on Late Phase Source Terms, Journal of Nuclear Science and Technology, vol.51, issue 4, 2014,
p.413-424.
JAEA R&D Review 201429

Research and Development Relating to the Accident at the TEPCO’s Fukushima Daiichi NPS

1-19 Safe Storage of Spent Cesium Adsorbent− Characterization and Evaluation for Hydrogen Production and Corrosion −(a) Vessel(b) Temperature

Air-H2(℃)
211

Vent tube

Plug
Shield192

Water inlet

Air
Water outlet
Radioactive
adsorbent
Residualwater

Max 211 ℃(c) Hydrogen concentration
(%)
1.59

Max 1.6%

1.44

174

1.28

156

1.12

137

0.96

119

0.80

101

0.64

82

0.48

64

0.32

45

0.16

27

0.00

Fig.1-39 An example of the thermal hydraulic analysis of a Cs adsorption vessel with residual water
A reference vessel (decay heat 504 W, water level 24 cm, hydrogen production 20.5 ℓ/d, absorbed dose rate at thebottom 755 Gy/h) was evaluated. Air is introduced by the water inlet and released through the vent plug. The maximum
temperature and hydrogen concentration are 211 °C and 1.6%, respectively. The temperature of the residual water isbelow 47 °C.

Potential (V vs. SHE)

0.8
SUS316L contacted with herschelite
60 ℃
diluted artificial sea water

0.6

Fig.1-40 Electrochemical corrosion tests of stainless steel SUS316L
The electrochemical potentials of SUS316L in contact with herschelite
were measured in diluted sea water to determine its immunity to localized
corrosion. The critical potential for localized corrosion decreases as the−
Cl ion concentration increases. Conditions for localized corrosion aresatisfied when the critical potential for localized corrosion is lower than
the steady-state corrosion potential. Under the conditions of the reference
vessel shown in Fig.1-39, the localized corrosion of SUS316L in contact−
with herschelite would not immediately occur at Cl ion concentrations of20000 ppm.

Satisfiedcorrosion
conditions

Steady-state
Critical potential for
corrosion potential localized corrosion
5 kGy/h

0.4
750 Gy/h

0.2

Unsatisfied
corrosion conditions

400 Gy/h
0 Gy/h

0
2
10

10

3

4

10

5

10

–

Cl ion concentration (ppm)

Stainless steel vessels containing mineral adsorbents(herschelite, etc.), which possess high selectivity towardcesium (Cs), are used for the treatment of radioactive saline
water at the accident at the TEPCO’s Fukushima Daiichi
NPS. The spent vessel is highly radioactive and temporarily
stored with the shield (Fig.1-39(a)). For long-term storage,
we evaluated the hydrogen production and corrosion ofherschelite adsorption vessels.
As fundamental properties of herschelite, we investigated
its thermal conductivity, hydrogen production yield, and Cs
adsorption coefficients, as well as determined the Cs activity,
decay heat, absorbed dose rate, and hydrogen production rate.
The hydrogen concentration in a reference vessel of uniform
Cs distribution was evaluated based on these fundamental
properties. Fig.1-39(b) shows that the maximum temperature
is 211 °C, below the self-ignition temperature of hydrogen(500–571 °C). This thermal hydraulic behavior maintains the

maximum hydrogen concentration below 1.6% (Fig.1-39(c))
and well below the lower explosive limit of 4%.
An electrochemical method was employed to determine
the immunity to localized corrosion. Fig.1-40 shows thepotentials of SUS316L in a mixture of herschelite and sea
water. The critical potential for localized corrosion decreased−
as the Cl ion concentration increased. The conditions for
localized corrosion are satisfied when the critical potential
is lower than the steady-state corrosion potential. Therefore,
under the conditions (temperature < 60 °C, adsorbed dose rateof 755 Gy/h) shown in Fig.1-39, the localized corrosion of
SUS316L in contact with herschelite would not immediately
occur at Cl− ion concentrations of 20000 ppm.
Part of this study was the results of research entrusted to JAEA
by the Agency for Natural Resources and Energy, Ministry of
Economy, Trade and Industry of Japan (METI).

Reference
Yamagishi, I. et al., Characterization and Storage of Radioactive Zeolite Waste, Journal of Nuclear Science and Technology, vol.51, issues 7-8, 2014,
p.1044-1053.

30

JAEA R&D Review 2014

Research and Development Relating to the Accident at the TEPCO’s Fukushima Daiichi NPS

1-20 Development of an Analysis Method for a Hard-To-Measure Nuclide in Wastewater− Cadmium-113m Analysis Method Using β-ray Measurement −30 mℓ

SampleCd 1 mg
evaporation

100
collected effluent

Sr

2 M HCｌ10 mℓ

50
Waste

0
100
Fe
Recovery (%)

1.2 M HCｌ10 mℓ
1 M HNO3 15 mℓ
1 M HNO3 30 mℓ

Anion
Exchangeresin

evaporation
0.1 M HNO 3
Sample solution

50

10 mℓ

(9 mℓ)(0.5 mℓ)scintillator 11 mℓ

LSC
measurement

0
100

Zn
50

0.5 M HNO 3
ICP-AESmeasurement

Fig.1-42 Analytical procedure for 113mCd
Eluate (1 M HNO3) discards the first 15 mℓ and then collects
the next 30 mℓ.

0
100
Cd

100

36

Cl

50

10 20 25 30 35 40 45 50 55 60 65 70 75(a) (b)(c)
Amount of effluent (mℓ)(a) Load sample (b) Column washing with 1.2 M HCl(c) Elution with 1 M HNO3

Fig.1-41 Recovery of Cd, Zn, Fe, and Sr as functions of
the amount of effluent
Separation condition: The sample is prepared in 10 mℓ of
a 2 M HCl solution. (a) The solution is passed through ananion exchange resin. Cd is retained on the resin. (b) The
Resin is washed with 10 mℓ of 1.2 M HCl. (c) Cd is eluted
from the resin with 55 mℓ of 1 M HNO3.

In the TEPCOʼs Fukushima Daiichi NPS, the wastewatercontaining radioactive nuclides is treated by water treatment
systems. To check the removal performance of radioactive
nuclides by the Advanced Liquid Processing System (ALPS)(which is one of the water treatment systems), it is necessary toevaluate the activity concentration of nuclides in water treatedby ALPS. Since the γ-ray emission ratio of Cadmium-113m( 113mCd) (which is one of nuclides to be evaluated) issmall (0.023%), we cannot apply non-destructive γ-ray
measurement. Therefore, a new analysis method that measures
β-rays (emission ratio: 99.9%) was examined.
To measure the β-rays of 113mCd with sufficient accuracy, it
was necessary to separate the other nuclides and elements. Ananion-exchange procedure was applied to this separation. The
examination result of the separation condition of Cd and theother elements is shown in Fig.1-41. Most of the Cd eluted

Counting efficiency (％)

0

113m

Cd

80

99

Tc

60
14

C

40
20
0
150

250

350

450

550

Quench indicating parameter (tSIE)

Fig.1-43 Counting efficiency as a function of the quench
indicating parameter (tSIE)
The β-ray counting efficiency for 113mCd was estimated using
14
C, 99Tc, and 36Cl standard solutions.

to 45–50 mℓ of effluent. When 35–65 mℓ of effluent wascollected, Cd could be efficiently separated and collected fromthe other elements. The 113mCd analysis procedure establishedbased on this result is shown in Fig.1-42. β-rays of 113mCd were
measured using a liquid scintillation counter (LSC). Since
the counting efficiency of LSC is decreased by quenching
phenomena, it is necessary to calibrate its efficiency. However,since a standard solution of 113mCd was difficult to purchasefrom vendors, the β-ray counting efficiency of 113mCd wasestimated using other β-ray-emitting nuclides (Fig.1-43).
When the sample, which was a simulation of water treatedby ALPS, was analyzed using this established analysis
method, more than 97% of Cd was recovered, and a minimum
detectable activity of 0.004 Bq/mℓ was obtained. Throughthe above examination, we contributed to the performanceassessment of ALPS.

Reference
Yasuda, M. et al., Determination Method using Liquid Scintillation Counter for 113mCd in Wastewater in Fukushima-1 Nuclear Power Plant, Bunseki Kagaku,
vol.63, no.4, 2014, p.345-350 (in Japanese).
JAEA R&D Review 201431

Research and Development Relating to the Accident at the TEPCO’s Fukushima Daiichi NPS

1-21 Toward the Treatment and Subsequent Disposal of the Accident Waste− An Analysis of Rubble and Trees Collected from the Site of the TEPCO’s Fukushima Daiichi NPS −Stock yard
7/26

Fig.1-44 Sampling locations of rubble and trees(Tanaka, K. et al., JNST, vol.51, issues 7-8, 2014, p.1032-1043.)

7/26

7/27

1F1
1F6
1F5

Rubble around 1F1, 1F3, and 1F4 was collected on June 25 andJuly 27, 2012. Tree samples were collected on July 26, 2012.

1F2
6/25

1F1
1F2
1F3

1F3
6/25

7/26

1F4

1F4

6/25

rubblecut treesliving tree

(b)
1F1
3

10

1F3

10%

1F4

1%

cut tree

0.1%

1F3 living tree

0.01%

10 1

10−1
ND

10−3 0
10

102

104

Radioactivity concentration of

106
137

Radioactivity concentration of 3H (Bq/g)

Radioactivity concentration of 90Sr (Bq/g)

(a)

1F1

10

3

1F3

10%

1F4

1%

cut tree

0.1%

1F3 living tree

0.01%

10 1

10−1

10−3 010

Cs (Bq/g)

102

104

Radioactivity concentration of

106
137

Cs (Bq/g)

Fig.1-45 Concentrations of 3H and 90Sr as functions of those of 137Cs(Tanaka, K. et al., JNST, vol.51, issues 7-8, 2014, p.1032-1043.)

These figures show the radioactivity concentrations of (a) 90Sr and (b) 3H as functions of the concentration of 137Cs. It is clarified
that the concentrations of 90Sr did depend on those of 137Cs, but those of 3H did not.

After the accident at the TEPCO’s Fukushima Daiichi NPS(1F, The “Unit 1-6 at 1F” is abbreviated to “1F1-1F6”), hugequantities of rubble and many trees were contaminated byradionuclides, including 137Cs and 131I, which were generated
at the site of 1F. To determine a strategy for the treatment andsubsequent disposal of the rubble and trees, it was essential
to clarify their radionuclide and radioactivity concentrations.
To determine the radionuclides in and on the rubble and trees,
samples were collected at the location shown Fig.1-44.
Important radionuclides (including α-ray and β-ray emitting
radionuclides, which need chemical analysis) were selectedbased on a safety assessment of the waste disposal. Analyticalmethods for the important radionuclides have alreadybeen developed by JAEA, and applied to various types ofradioactive waste generated by research and industrial andmedical facilities. We modified the existing analytical methods
and analyzed the rubble and trees.
The ratios of the radioactivity concentrations 3H and 90Sr to
those of 137Cs determined in this study are plotted in Fig.1-45.
As shown in Fig.1-45, the concentration of 3H did not dependon the sampling location. In contrast, the radioactivity ratiosof 90Sr to 137Cs clearly depended on the sampling location. The
observed results for 90Sr also implied that the 90Sr/137Cs ratio
was different for each Unit of 1F. We are establishing methods
for the treatment and disposal of radioactive waste on the basisof this data.
This study was sponsored by the Agency for Ministry of
Economy, Trade and Industry of Japan (METI) in the FY 2013
and 2014.

Reference
Tanaka, K. et al., Radiochemical Analysis of Rubble and Trees Collected from Fukushima Daiichi Nuclear Power Station, Journal of Nuclear Science andTechnology, vol.51, issues 7-8, 2014, p.1032-1043.

32

JAEA R&D Review 2014

２ Nuclear Safety Research
Implementing Continuous Improvements for Safety

Nuclear Safety Research Center
Fuel Manufacturing Plant

Research Planning and
Coordination Office
Office for Analysis of Eventand Regulatory Information

Fuel
Cycle Reprocessing

Thermohydraulic Safety Research Gr.

Reactor Safety Research Unit

Plant

NSRR

Fuel Safety Research Gr.

Materials and Structural
Integrity Research Unit

Structural Integrity Research Gr.

Risk Analysis andApplications Research Unit

Severe Accident Analysis Research Gr.

Fuel Cycle Facility
Nuclear Power Plant
・Fuel safety
・Safety evaluation of・Thermohydraulics safety fuel cycle facilities
・Evaluation of material aging
・Degradation mechanism offuel and material under
neutron irradiation

Materials and Water Chemistry Research Gr.

RFEF

Waste Disposal Facility
・Safety evaluation ofradioactive waste
disposalRadiation Risk Analysis Research Gr.
Fuel Cycle Safety Research Gr.

Fuel Cycle Safety Research Unit
Criticality Safety Research Gr.

Environmental Safety
Research Unit

Waste Safety Research Gr.
Environmental Safety Research Gr.

THYNC

LSTF

JMTR

JMTR-HL

NUCEF

Fig.2-1 Organization of the nuclear safety research
centerStarting April 2014, we have been strengthening theorganization to promote studies of SAs, including
the initiating event, progress of the accident, and theenvironmental impact, as well as the safety assessment
due to the design basis accidents for several nuclear
facilities.

Fig.2-2 Nuclear safety research areas and related JAEA
facilities
NSRC acquires experimental data from various Japan Atomic
Energy Agency facilities. These data for risk assessment andsafety evaluation of nuclear facilities contribute to international
efforts to develop evaluation methods and criteria.

To ensure the safe use of current nuclear facilities,
the Nuclear Safety Research Center (NSRC) studies thepossibility of serious accidents at nuclear installations andtheir consequences. A massive and severe accident (SA)occurred at the Tokyo Electric Power Company, Incorporated
Fukushima Daiichi Nuclear Power Station (1F) as a resultof the Great East Japan earthquake and tsunami. NSRC has
been collaborating with the Japanese government to instituteappropriate emergency response procedures based on researchfindings and developments. The specialists at NSRC regretthat they were unable to prevent the accident.
Continuous efforts should be made to avoid serious damage
while we utilize nuclear energy. A proper understandingof the situation and continuous improvement of current
technologies are important for making the best and safest
use of these technologies. In fact, the International Atomic
Energy Agency, which investigated the accident at 1F,
highlighted the importance of “continuous improvement.” To
contribute to “continuous improvement aimed at the highest
level of safety,” we strengthened the organization to promotestudies of SA prevention and evaluation to reduce the risk ofaccidents at nuclear facilities and to improve nuclear disasterprevention using analyses such as environmental impact
assessments (Fig.2-1).
We are committed to safety research and have devoted
to this cause the various facilities shown in Fig.2-2. These
special facilities allow the handling of radionuclides and thesimulation of accident conditions.
This chapter presents the results of recent safety researchon the following topics: simulation of fuel-cladding failure(Topic 2-1), effect of neutron irradiation on fracture propertiesof stainless overlay cladding for reactor pressure vessel(Topic 2-2), development of probabilistic structural assessment
code for nuclear components with long-term operation(Topic 2-3), analytical techniques of gas inflow during loss of
coolant accidents at pressurized water reactors (Topic 2-4),
study on applicability of compacted clays to waste-disposal
facility (Topic 2-5), and SA evaluation regarding loss ofcooling function of highly active liquid waste in reprocessing
plant (Topic 2-6).
The following Topics solicited by NSRC and related to theaccident at 1F are described in Chapter 1: release of I-131
from contaminated water in the basement of the reactor
building (Topic 1-18).
JAEA R&D Review 201433

Nuclear Safety Research

2-1 Fuel Failure Processes Revealed by Numerical Approach− Simulation of Fracture Behavior Using Damage Mechanics Model −

(a)

(d)

Hoop direction

Interpretations:
Peripheral
crack
generation

Radialdirection

Through-wall
penetration(e)Damage parameter
1.0
0.0

Elapsed time
(ms)

0.0

Cladding
thickness~600 μm

0.07(b)Many dimples
in fractography

Deep initial crack(100 μm)
50 μm

(c)

0.37

Shallow initial crack(50 μm)~10 μm

0.38

0.39

Fig.2-3 Failure of long-term used fuel cladding andmodelling(a) We anticipated that peripheral cracks are at the origin of
the observed cladding failure. (b) Many dimples indicate thatfracture occurred in a ductile manner. (c) The cladding wasmodeled as an assembly of elements that break upon finite
deformation.

Fig.2-4 Simulation results(d) Fracture appearance depends on initial crack length.
(e) Penetration through the wall with very small increment intime (and strain).

Reactivity initiated accident (RIA), a power excursion
caused by control rod ejection, is one of the accidents assumed
to confirm safety design of nuclear reactors. In Japan, thefailure limit of long-term used fuels for RIA is postulated to
be lower than that of fresh fuels, based on RIA-simulated
tests conducted at the Nuclear Safety Research Reactor.
Although the lower failure limit was considered to be causedby through-wall penetration of cracks which were generated
in the cladding peripheral region with hydrogen absorbed andaccumulated during reactor operation (Fig.2-3(a)), directly
observing a phenomenon as rapid as fuel failure inside thereactor is not possible. Thus, the anticipated mechanism was
not confirmed and the detailed processes remain unknown.
To overcome this difficulty, we developed a numericalmodel of fuel cladding to simulate a RIA. The cladding wasmodeled as an assembly of elements (Fig.2-3(c)), each about
10 μm in size. This element was designed to break upon finite
deformation, as determined independently of the RIA tests.

We then simulated the failure of the virtual cladding under
RIA conditions, which is quite a new approach in this field,
realized on the basis of accumulated test data and simulation
techniques.
The simulation well reproduced the observations: fracturescharacterized by a predominant ductile shear in the shallowcrack case and straight-crack propagation in the deep-crack
case (Fig.2-4(d)), and penetration through the wall that
occurs in a very short time (<1 ms) (Fig.2-4(e)). This result
strongly supports the aforementioned hypothesis and reveals
that predominant ductile shear (Fig.2-3(a)) occurs when asoftened region extends along the strain field around the crack
tip and collapses simultaneously at a certain deformation.
A sensitivity analysis using the model provides information on
the influence of the loading state and temperature distribution
inside the cladding on the failure limit, which supports theongoing development of methods for safety evaluation that
can accurately predict the failure limit of long-term used fuels.

Reference
Udagawa, Y. et al., Simulation of the Fracture Behavior of Zircaloy-4 Cladding under Reactivity-Initiated Accident Conditions with a Damage Mechanics
Model Combined with Fuel Performance Codes FEMAXI-7 and RANNS, Journal of Nuclear Science and Technology, vol.51, issue 2, 2014, p.208-219.

34

JAEA R&D Review 2014

Nuclear Safety Research

2-2 Improving Structural Integrity Assessment of Reactor Pressure Vessels− Effect of Neutron Irradiation on Fracture Properties of Stainless Overlay Cladding
for Reactor Pressure Vessel −Stainless Overlay Cladding
≈5 mm

500

■Before Irradiation■After Irradiation (1.2 × 1024 n/m2)

Base Metal

Reactor Pressure
VesselTensile StressCore
Distance from
Inner Surface

Fracture Resistance (kJ/m2)

Test Temperature: 120 ℃

Emergency Core Cooling Water

400
300
200
100
0
0.0

1.0

2.0

3.0

Crack Extension (mm)Fig.2-5 RPV and PTS event
When the emergency core cooling water is injected into
RPV under high temperature and pressure condition, tensilestress occurs at the inner surface of RPV because of thethermal stress. This stress is superimposed on the stresscreated by the internal pressure.

Fig.2-6 Fracture resistance curves of stainless overlay
cladding
The fracture resistance of cladding materials increases as
the crack extension proceeds. The fracture resistance curvesindicate no obvious degradation of the cladding materials
after neutron irradiation.

The reactor pressure vessel (RPV) is the most importantcomponent for safe operation of a nuclear power plant.
Pressurized thermal shock (PTS) is one of the most severe
events that may lead to the failure of the pressurized water
reactor (PWR). PTS is a transient event caused by rapid
cooling of the inner surface of the RPV, which is consequently
due to the injection of emergency core-cooling water. PTS
causes high tensile stress on the inner surface of the RPV(Fig.2-5). Assessing the structural integrity of the RPV during
a PTS event postulates the formation of an axial semi-ellipticalcrack on the inner surface of the RPV base metal or weld
metal. To prevent the corrosive effects of the reactor coolant,
the inner surface of a RPV is overlaid with cladding made of
5-mm-thick austenitic stainless steel. A PTS event may leadto higher tensile stress in the cladding region. Understanding
the properties of fractures in cladding materials that have been
irradiated by neutrons is necessary to improve the assessmentof the structural integrity of a RPV.

We tested the elastic-plastic fracture toughness of cladding
materials subjected to a neutron fluence of 1.2 × 1024 n/m2.
Typical fracture resistance curves for cladding materials
before and after irradiation are shown in Fig.2-6. The test
temperature of 120 °C is the temperature range in which
tensile stress increases during a PTS event. As shown in the
figure, the fracture resistance of cladding materials increases
as the crack extension proceeds. No obvious degradation
appears in the fracture resistance curves of the cladding
materials after neutron irradiation. These properties are similar
to what is found at other temperatures, and no brittle fractureappears in the low-temperature region. These results suggestthat the fracture properties of the cladding region change little
over the long-term operation of a RPV.
This study provides data on the materials’ properties
that will improve the assessment of structural integrity andcontribute to the preparation by academic societies of therequisite standards.

Reference
Tobita, T. et al., Effect of Neutron Irradiation on the Mechanical Properties of Weld Overlay Cladding for Reactor Pressure Vessel, Journal of Nuclear
Materials, vol.452, issues 1-3, 2014, p.61-68.
JAEA R&D Review 201435

Nuclear Safety Research

2-3 Structural Integrity Assessment of Nuclear Components Operated over a Long Period− Probabilistic Fracture Mechanics Analysis for Ni-Based Alloy Welds −Wastage of reactor pressure
vessel due to leakage

Reactor pressure vessel

Ni-based alloy

Low alloy steel

Postulated crack

Total number ofnozzles: 69

PWSCC
Dissimilar weld

Weld overlay cladding

Ni-based alloy

Stainless steel

Cumulative leak probability (%)

NozzlePASCAL-NP
Leak detection

100

10

1

3/69 nozzles(about 4.3%)
0.1

0.01

0.001
0

Approximately 16 years
1020

Operating time (year)

Fig.2-7 Schematic of PWSCC at a vessel head penetration
PWSCC occurs at dissimilar welds at the penetration nozzles
of reactor vessel heads of pressurized water reactors due tomaterial (Ni-based alloy), environmental (water condition),and stress (weld residual stress, etc.) factors.

Fig.2-8 Case study for leakage due to PWSCC at the DavisBesse nuclear power plant
A case study was performed using the PASCAL-NP code todescribe vessel head penetrations caused by PWSCC at theDavis-Besse nuclear power plant (United States). The study
considers the scatter in time for crack initiation, the crack growthrate, the residual stress distribution, etc. Analytical results forcumulative leak probability are consistent with the leak detection
data.

During long-term operation, primary water stress corrosion

complicated dissimilar metal welds, including NiSCC in a

cracking (PWSCC) was detected at dissimilar metal welds of

BWR. Using a Monte Carlo approach and considering the

Ni-based alloys at vessel head penetration of the pressurized

scatter described above, this code evaluates the probabilities

water reactors (PWRs) (Fig.2-7).

of failure for events such as leakage and break of components.

Some features of PWSCC differ from those of SCC reported

Fig.2-8 shows the results of a case study on leakage nearat piping joints of primary loop recirculation system in boilingthe vessel head penetrations at the Davis-Besse nuclear power

water reactors (BWRs), e.g., PWSCC initiates and grows inplant, which suffered from PWSCC. The analytical cumulative

Ni-based alloys at geometrically complicated welds.

leak probability exceeds 5% after 16 years. These PFM

Some factors related to PWSCC, such as crack initiation

analysis results are consistent with the leak detection data,time, crack growth rate, and residual stress distribution, have

which provides about 4.3% probability (i.e., 3 of 69 leaks

large scatter.

were detected).

To account for this scatter, we developed a probabilistic

PFM is expected to be a rational and useful method tofracture mechanics (PFM) analysis code PASCAL-NP toestimate failure probabilities due to unexperienced events,

assess the structural integrity of dissimilar metal welds

to assess risks considering aging degradation of nuclear

subjected to PWSCC.

components, and to develop a plan of risk-informed in-service

PASCAL-NP contains functions that describe crack

inspection. To advance the assessment of structural integrity,

initiation and crack growth and evaluate failures for various

we continuously improve PASCAL-NP by incorporating thepatterns of crack locations and orientations at geometricallylatest knowledge.

Reference
Udagawa, M. et al., Userʼs Manuals of Probabilistic Fracture Mechanics Analysis Code for Ni-Based Alloy Welds, PASCAL-NP, JAEA-Data/Code 2013-013,
2013, 145p. (in Japanese).

36

JAEA R&D Review 2014

Nuclear Safety Research

2-4 Investigating Gas Inflow Influences in Accidents− Concentration Analysis with Gas Measurement Device −

14

10

6

Fully opening of
SG safety valves
Inflow of non-condensable gas

2

Small-break
LOCA

Fig.2-9 PWR small-break LOCA andaccident-management measure
SG secondary depressurization isan effective accident-management
measure for core cooling because
steam condenses in the U-tubes during
the PWR small-break LOCA.

Fully opening ofSG relief valves

8

4

Pressure
vessel100

120

Air concentration (%)

Outletplenum16

Pressure (MPa)Upper head

U-tubesAccumulator
systemAir conc. by Zirconia sensor
Air conc. by steam temp. and press.

Primary pressure
Secondary pressure

SG secondary
depressurization
Steam
generator(SG)

Pressurizer

80

60

40

1st measure.

01000 2000 3000 4000 5000 6000 7000 8000

Time after break (s)Fig.2-10 Primary and secondarypressures under non-condensable
gas inflow
The decreasing rate of the primarypressure became smaller than thatof the secondary pressure after noncondensable gas inflow during the SG
secondary depressurization period,
suggesting a delay in the coolant
injection into the core.

Previous research at the large scale test facility (LSTF)
simulated mainly small-break loss-of-coolant accidents(LOCAs) in PWR to clarify specific thermal-hydraulicphenomena and to confirm the effectiveness of accident
management (AM) measure.
When the high-pressure injection system of an emergency
core cooling system (ECCS) completely fails during a PWR
small-break LOCA, core damage would occur due to the loss ofthe primary coolant and the delay of water injection into the coreby the actuation of an ECCS, such as the accumulator (ACC)system. Steam generator (SG) secondary depressurization is aneffective AM measure for core cooling because of the steam
condensation in the SG U-tubes (Fig.2-9). When the isolationof ACC system is failed, ingress of non-condensable gas used
for the pressurization of ACC tanks to the primary system maycause degradation in the primary depressurization after thecompletion of the ACC coolant injection. It is thus necessary
to clarify how gas inflow influences the small-break LOCA
progression with the SG secondary depressurization.
A Zirconia sensor has been employed for measuring oxygen
concentration in engineering applications, such as exhaust gas
for chimney. A gas-measurement device to directly measureoxygen concentration in steam-gas mixtures was developed
using a Zirconia sensor under high-pressure condition (<2 MPa),
and a pressure control valve for the Zirconia sensor was

2nd measure. 3rd measure.

20
9000 9500 10000 10500 11000 11500 12000 12500

Time after break (s)Fig.2-11 Air concentration at SG
outlet plenum
Non-condensable gas from ACC tanks
accumulated in SG outlet plenum
through air concentration, as confirmed
by a Zirconia sensor and the estimatedvalue based on steam temperature andpressure at SG outlet plenum.

installed up-stream of the sensor. The gas concentration could
thus be measured at the SG outlet plenum and the upper headof the vessel, where non-condensable gas may have passed
through.
We conducted a simulation test on a PWR horizontal leg
small-break LOCA for the OECD/NEA ROSA Project at the
LSTF. SG secondary depressurization as an AM action was
initiated by fully opening the SG relief valves 10 min after
the safety injection signal, considering a grace period to startan operator action. We further assumed that enhanced SG
depressurization was conducted by fully opening the SG safetyvalves when the primary pressure decreased to 2 MPa. The
primary pressure decreased following the secondary pressure
after the start of the SG secondary depressurization, while thedecreasing rate of the primary pressure became smaller thanthat of the secondary pressure after the non-condensable gas
inflow (Fig.2-10). The gas concentration was analyzed throughthe measurements at the SG outlet plenum (Fig.2-11) and theupper head of the vessel. The results of the analysis confirmedthat non-condensable gas from the ACC tanks accumulates in
the U-tubes and in the outlet plenum of the SG.
Detailed experimental data obtained from the LSTF
simulations were useful for the validation and verification ofthermal-hydraulic best-estimate computer codes of the ROSA
Project organizations, such as nuclear regulatory agencies.

Reference
Takeda, T. et al., Measurement of Non-Condensable Gas in a PWR Small-Break LOCA Simulation Test with LSTF for OECD/NEA ROSA Project and
RELAP5 Post-Test Analysis, Experimental Thermal and Fluid Science, vol.51, 2013, p.112-121.
JAEA R&D Review 201437

Nuclear Safety Research

2-5 Dissolution of Compacted Clays in High-Level Radioactive Waste Repositories− Alkaline Dissolution Inhibited by Accessory Minerals in Bentonite −10−2

CementBentonite
Highly alkaline
component

Overpack
High-level radioactive waste(Vitrified waste)
Bedrock

Radionuclide

Dissolution rate of montmorillonite (kg m−3 s−1)

Vitrified waste

Overpack

CementBedrock

Bentonite

Highly alkaline
component

Compacted montmorillonite (this study)
Compacted bentonite

Calculated values using a previous dissolution
model of pulverized montmorillonite
10−3

10−4

10−5
Dissolution rate of montmorillonite
in compacted bentonite
10−6
10−2

10−1

100

Activity of OH− (mol dm−3)

Fig.2-12 Conceptual view of disposal system for HLW
Vitrified HLW is encapsulated in an iron container (overpack),surrounded by an engineered buffer material (bentonite), andplaced in stable bedrock. Cement used for mechanical supportof shafts and drifts increases the alkalinity of surrounding water,and over a long period, it likely deteriorates the properties ofthe bentonite buffer.

Fig.2-13 Dissolution rate of montmorillonite in highly
alkaline solutions
The dissolution rate of montmorillonite was higher in compacted
montmorillonite than in compacted bentonite. Dissolution ofmontmorillonite was inhibited in the compacted bentonite. The
difference is due to decreased activity of OH− stemming fromthe dissolution of accessory minerals.

In geological disposal of high-level radioactive waste(HLW), vitrified waste containing long-lived radionuclides
is encapsulated in an iron container called an overpack,surrounded by an engineered bentonite buffer material(compacted bentonite), and placed in a repository in stablebedrock, as shown in Fig.2-12. The high sorptivity andlow permeability of bentonite retards outward radionuclide
movement. Because the radionuclide-confining property ofthe bentonite depends on the montmorillonite content, thedissolution of montmorillonite induced by the highly alkaline
component supplied from the cementitious materials in thesystem must be revealed to evaluate the long-term barrier
function of this system.
Previous studies on the dissolution rates of clays in alkaline
solutions used pure pulverized montmorillonite; hence, therates were obtained for a much lower solid/solution ratio
than that warranted by the disposal conditions. Whether the
rates may be applied to predict the dissolution of compacted
bentonite under current disposal conditions remains a question.
Therefore, by experiment, we determined the dissolution rateof montmorillonite in compacted bentonite. The dissolution

rate thus obtained for compacted bentonite was lower anddepended more strongly on the activity of hydroxide ions(OH−) than the rates obtained in previous studies. However,
the mechanism behind such differences has not yet beenclarified.
The present study experimentally investigated the dissolution
rate of pure compacted montmorillonite at 130 °C. The
dissolution rate of montmorillonite is found to be higher anddepends slightly less on the activity of OH− in pure compacted
montmorillonite than on its activity in compacted bentonite, as
shown in Fig.2-13. These results suggest that the dissolutionof montmorillonite in compacted bentonite is inhibitedby decreasing the activity of OH− by dissolving accessory
minerals, such as silica, in the bentonite. The logarithmic plotof the dissolution rate shows that the effect of the inhibition
would thus be enhanced at lower activity of OH−.
The present study was sponsored by the Secretariat of the
Nuclear Regulation Authority (NRA) (formerly the Nuclear
and Industrial Safety Agency (NISA), Ministry of Economy,
Trade and Industry (METI), Japan).

Reference
Sawaguchi, T. et al., Alkaline Dissolution Behavior of Montmorillonite under Compacted Condition, Genshiryoku Bakkuendo Kenkyu (Journal of Nuclear
Fuel Cycle and Environment), vol.20, no.2, 2013, p.71-78 (in Japanese).

38

JAEA R&D Review 2014

Nuclear Safety Research

2-6 Acquisition of Evaluation Data for Severe Accident in Reprocessing Plant− Acquisition and Evaluation of Airborne-Release Data from Boiling Accident −Condenser
Furnace

Air filter

Gas absorber
(5 ℃)

Air
Pump
Coolant(5 ℃)
MFC
Silica gel

Mass flow
controller
1 ℓ/min

Simulated HALW
Condensate

Fig.2-14 Apparatus to obtain release ratio of each element
Simulated HALW was heated to 300 °C, and condensates were collected at regular timeintervals. The release ratios of elements were obtained by analyzing the respective condensates.
10−1

350
NO2– Added

10−2

No addedRelease ratio (–)

Temperature (℃)

300
250
200
150
100

NO2– Added

No added

Ru
Cs

Ru
Cs

10−3
10−4
10−5
10−6

50

10−7

0
0

10

20

30

40

50

60

70

80

90

Time (min)

0

10

20

30

40

50

60

70

80

90

Time (min)Fig.2-15 Time dependence of sample temperature
Simulated HALW continued boiling for about 70 min anddried out after boiling.

Fig.2-16 Time dependence of release ratios
The nitrite ion (NO2−) is thought to prevent nitric acid fromoxidizing nitrosyl Ru early in the boiling step.

When the high active liquid waste (HALW) is no longer
cooled, it may boil and dry out, and radioactive materials,
which are dissolved in HALW, may be released in the air.
Therefore, understanding the release and transport of thesematerials is necessary to evaluate safety issues in the caseof an accident. This study simulates the release ratio of eachelement by virtually heating HALW, which was prepared using
a stable isotope selected with reference to the composition
of HALW. Fig.2-14 shows the apparatus used to obtain therelease ratio. In the simulation, HALW was heated to 300 °C
with the vapor flow rate being set equal to that estimated foran accident. Figs.2-15 and 2-16 show the time dependence ofthe temperature and the release ratio, respectively. The releaseratio of ruthenium (Ru) rapidly increases beginning at thelatter part of boiling step and continuing to the drying step.
At 300 °C, it increases to about 10−2. The release ratio of Ru
is three orders of magnitude greater than that of nonvolatileelements, such as cesium (Cs). The release ratio of Ru appears
to be greater than that of Cs because nitric acid is concentrated
in the dried out HALW and concentrated nitric acid oxidizes

nitrosyl Ru to create a volatile Ru species. However, in theboiling step, both Ru and Cs are considered to be present
in the mist, which is generated by a bubble breaking in theboiling HALW. These elements are then released in the air.
Fig.2-16 shows that the release ratio of Ru is greater than thatof Cs at the boiling step without adding nitrite ions (NO2−) as
the radiolysis product of nitric acid in HALW. Upon adding
NO2−, the release ratio of Ru becomes the same as that of Cs.
Consequently, NO2− is thought to prevent nitric acid fromoxidizing nitrosyl Ru to create a volatile Ru species in HALW
early in the boiling step.
The release data of each element were obtained in thissimulation. In the next step of the simulation, we will obtaintransportation data for each element released in the air anddevelop a model to analyze the release and transport due to aboiling accident.
The present study was accomplished in collaboration withJapan Nuclear Fuel Limited (JNFL) and the Japan Nuclear
Energy Safety Organization (JNES).

Reference
Amano, Y. et al., Study on Release and Transport of Aerial Radioactive Materials in Reprocessing Plant, Proceedings of International Nuclear Fuel Cycle
Conference (GLOBAL 2013), Salt Lake City, Utah, USA, 2013, paper 7568, p.1411-1417., in CD-ROM.
JAEA R&D Review 201439

３ Advanced Science Research
Advanced Science Pioneers the Future
Cultivation of Unexploded Areas• Expansion of the area of basic nuclear energy science• Internationalization, cultivation of human resources

to open a new frontier of nuclear
energy science

Advanced Materials
Research

Frontier Researchon Heavy Element Systems

Spintronics,
Condensed matter theory,
Simulation in materials science

Actinide materials science,Superheavy elements,
Physics of heavy nuclei

Conventional
Nuclear Energy
Research

Advanced Atomic-Energy
ScienceResearch on Radiation
Field Effects
Hadron physics,
Radionuclide transport,
DNA radiation damage, Positron beamInternational Collaborations

Research Institutes
in/outside JAEA

Trains PhD students, PostDocs, and other scholars, and provides human resources in/outside JAEA

Fig.3-1 Advanced research to open a new frontier of nuclear science
Nuclear research trains PhD students, postdocs and other scholars, and provides human resources both in and outside JAEA.

Basic science supports the greater part of atomic energyrelated science and technology. In the development of atomic
energy research, it is particularly crucial to respond to the newphase of atomic energy that we will be entering in the next
few decades, as well as constantly seeking solutions to today’s
energy problems.
As for the Advanced Science Research Center (ASRC), ourrole is to explore the unlimited possibility of atomic energy
from the perspective of basic science. In this process, we aim
to open up new research fields and strive to further developscience and technology. Beginning in FY 2010, we haveconducted our mid-term plan, focusing on three main areas:advanced materials research, frontier research on heavy element
systems, and research on radiation field effects (Fig.3-1).
Advanced materials research has been conducted withthe aim of creating spintronic materials and advancing
the theory of solid state physics. In the research of heavy
element systems, we have performed diverse studies on
actinide compounds and development of new materials, aswell as studying the nuclear science of superheavy elements.
Our research on radiation field effects has included hadron
physics, the study of radiation effects on bio-molecules, andthe technical development and application of spin polarized
positron beams. Through interactions between the three areasand collaboration with other divisions, we have been working
on cultivating new fields of atomic energy science (Fig.3-1).
We have made a number of highlighted achievements
during FY 2013. Through advanced materials science, a
40

JAEA R&D Review 2014

new principle of the generation of a magnetic flow withcopper and aluminum was proposed (Topic 3-1). A methodof determining the spin status of a conduction electronin graphene was revealed (Topic 3-2). Substantial results
have been obtained by frontier research on heavy element
systems; the study of the nematicity and superconductivityof a uranium (U) compound (Topic 3-3), and the elucidationof the dielectric degradation of a ceramic condenser due tohydrogen impurities in ferroelectric materials (Topic 3-4) aremarked examples. Prominent results toward an explanation of
the ecological effects of radiation were also acquired throughresearch on radiation field effects. The discovery of the factthat DNA damage has an effect on normal chromosomes(Topic 3-5), and the development of an effective method foranalyzing the behavior of U elements in various environments(Topic 3-6) were cases in point. In hadron physics research at
the Japan Proton Accelerator Research Complex (J-PARC),
challenging experiments to extend the limit of the neutronnumber fraction were conducted. The experiments included asearch for neutron-rich Λ hyper nuclei, which have a higher
fraction of neutrons than usual (Topic 3-7).
In an effort to respond to the accident at the Tokyo ElectricPower Company, Incorporated Fukushima Daiichi Nuclear
Power Station, we have successfully visualized the movementsof radiocesium deposited into trees and foliage (Chapter 1,
Topic 1-8). As a part of the research into the environmental
fate of radioactive materials, we have demonstrated a basic
study of bio-decontamination methods using fungi.

Advanced Science Research

3-1 Flow of Spins Resulting from Acoustic Wave
− Towards Rare-Metal-Free Magnetic Devices −zMagnetic field due to rotational deformation
Spin current
Surface acoustic wave inducedby mechanical resonator
x

2π/k
y

Electron spins

Schematic picture of a metal
with a surface acoustic wave

Fig.3-2 Spin-current generation by acoustic wave
We have proposed a new method for spin-current generation
using the deformational motion of metal induced by acoustic
waves.

Fig.3-3 Magnetic pattern induced on the surface
Spin current due to the acoustic wave induces a magnetic
stripe pattern on the metal surface, whose spatial period is thesame as the wavelength of the acoustic wave (2π/k, where k is
the wave number of the acoustic wave).

Electrons have two fundamental properties, electricity andinto metals using piezoelectric devices, which convert electric

magnetism. In particular, the origin of magnetism is known as

signals into acoustic waves, the so-called surface acoustic“spin,” which is related to the spinning motion of electrons.

waves (SAWs) are created (Fig.3-2). SAWs propagate along

Recent progress of nanotechnology enables us to control

the x-axis and the excited local rotational deformation of thethe “spin current,” a flow of spins. Researchers have been

metal. The rotational deformation induces magnetic fields

studying the relation between spin current and magnetism,

whose intensity increases toward the metal surface. In general,and developing next-generation electronic devices using spin

spin has the following properties:

current.

(1) Spin aligns parallel to the magnetic field

Conventionally, the generation of spin current has required(2) Spin moves to where a stronger magnetic field is appliedmagnets or rare metals since such materials have special

Thus, the magnetic field induced by the SAW aligns thequalities that align electron spins. In other words, common

third component of spins to the z-axis and flows the spins

metals, such as aluminum (Al) or copper (Cu), cannot bealong the y-axis. As a result, a stripe pattern of spins appears

utilized for spin-current generation because they have weakon the metal surface, as shown in Fig.3-3.

spin-control properties.

We have also found that sufficient spin current can be

We have found that the deformation of crystals can aligncreated for magnetic devices if a SAW with several GHz range

spins and have proposed a new method for spin-current

is injected into Al or Cu. This means that our method does not

generation using the deformation of common materials, suchrequire magnets or rare metals for spin-current generation.

as Al or Cu.

Hence, our findings are expected to offer more options for

We formulated a theory of spin current in metals deformedby acoustic wave injection. When acoustic waves are injectedspin-current generation than ever before and a new route todevelop rare-metal-free magnetic devices.

Reference
Matsuo, M., Ieda, J. et al., Mechanical Generation of Spin Current by Spin-Rotation Coupling, Physical Review B, vol.87, issue 18, 2013,
p.180402-1-180402-4.
JAEA R&D Review 201441

Advanced Science Research

3-2 Elucidation of the Spin State of Conduction Electrons in Graphene− Significant Progress for Spintronic Applications of Two Dimensional Materials −Spin detector with ultimate
surface sensitivity+2eSingle-layer
graphene

Nucleus1s
2s

Allowed9Spin asymmetry (%)

Spin-dependent
Spin-polarized
emissionmetastable helium

Nickel

6
3
0
−3
4
2

6

8

10

12

14

Graphene

0−2

Conduction electronsof graphene−4Graphene / ferromagnetic
metal interface

6

8

10

12

14

16

Kinetic energy of emitted electrons (eV)

NickelFig.3-4 Schematic of the experimental method
When a low velocity spin-polarized metastable helium beam
is incident on the sample surface, the helium atoms reboundabove the surface without penetrating into the interior. During
the collision, a surface electron of graphene moves toward
the helium atom, and the helium atom ejects a substitutional
electron. Since the transfer from the sample surface to thehelium atom is allowed only for electrons with a specific spin
direction, the ejected electron carries the spin information of thegraphene surface electrons.

Fig.3-5 Spin asymmetries of graphene and nickel
The horizontal axes indicate the electron energy in graphene
and nickel, respectively. The energies surrounded by a brokenline square correspond to the energies of the conduction
electrons. The spin asymmetry (vertical axis) reflects electronic
spin state. If there is no spin polarization, the asymmetry
becomes 0%. At the graphene-nickel junction, the induction ofspin polarization in graphene is observed. The spin polarizationof graphene is found to be in the opposite direction to that ofnickel.

Since the discovery of a convenient fabrication method
using micromechanical exfoliation from bulk graphite(2010 Nobel Prize in physics), graphene has attracted
world-wide attention as an innovative nanoelectronic material.
In the spintronics field, graphene is also expected to be an ideal
spin-transport material. Techniques for the manipulation ofelectron spins in graphene are indispensable for realizing theapplications of graphene to spintronic devices; in particular,
the technique of spin injection from a magnetic electrode isa key issue. For the purposes of spintronic application, it isnecessary to understand the spin state of graphene in contact
with magnetic metal. However, it has been difficult to use
conventional techniques to analyze the electronic spin statesof graphene, since the weak signals from graphene of a single
atomic layer are easily buried in the strong signals from themagnetic metal substrate.
In this study, selective detection of the electron spinsin graphene at a junction of graphene and magnetic metal
was successfully demonstrated by employing a spinpolarized metastable helium beam with an extremely highsurface sensitivity (Fig.3-4). The graphene-magnetic metaljunction was prepared by chemically synthesizing graphene
on a nickel thin film. Fig.3-5 shows the spin state of thegraphene electrons obtained by the irradiation of the spinpolarized metastable helium beam. The conduction electronsof graphene contacted with nickel were found to be spinpolarized in the opposite direction from the conduction
electrons of nickel. The opposite spin polarization directionsof graphene and nickel could explain why the spin injection
efficiency is very low in graphene spin devices with graphenemagnetic metal contacts.
The present study provides direct information on how theconduction electrons of graphene were affected by the contact
with magnetic metals. Our research achievement is also
expected to greatly contribute to the design of spin-related
functions based on various two-dimensional materials that are
attractive for future spintronic application.
The present study was accomplished in collaboration with
National Institute for Materials Science (NIMS) and the
University of Science and Technology of China.

Reference
Entani, S. et al., Spin Polarization of Single-Layer Graphene Epitaxially Grown on Ni(111) Thin Film, Carbon, vol.61, 2013, p.134-139.

42

JAEA R&D Review 2014

Advanced Science Research

3-3 A Precursor to Exotic Superconductivity− Nematicity and Superconductivity in URu2Si2 −H=5.19 T
0.5 GaussL(θ) (Gauss)magnetic field

2 rotation axes

5K

10 K

12.5 K

16 K

20 K

0

45

90

θ (degree)
Fig.3-6 Two-axis rotating stage
The sample can be rotated precisely along two perpendicularaxes in an applied field H.

Fig.3-7 Field-angle (θ) dependence of Si-NMR line
width, L(θ)
Below 17.5 K, the amplitude of the θ-dependence
increases. The pointed minimum at 45° indicates 2-foldanisotropy.

In the superconducting state, two electrons can forma superconducting pair. To form such pairs, an attractive
interaction between the electrons is necessary. In conventional
superconductors, the origin of the attractive interaction isa lattice vibration. In contrast, a magnetic fluctuation isconsidered to be the origin in actinide superconducting
compounds. Since a high superconducting transition
temperature Tc can be expected for magnetic fluctuationinduced superconductivity, it is important to clarify its
mechanism.
In URu2Si2, a so-called “hidden order state” appears below
17.5 K (−256 °C). As the superconducting state appears at
lower a temperature of 2 K (−271 °C), magnetic fluctuationin the hidden order precursor state is believed to induce thesuperconductivity. However, the hidden order has not been
identified in 25 years, which is one of the important issues
facing condensed matter physics.
URu2Si2 has 4-fold symmetry in the paramagnetic state
above 17.5 K. Recently, however, it has been revealed thatthe electronic state shows a symmetry breaking from a 4-foldstate to a 2-fold state in the hidden ordering regime. In thisstudy, the magnitude of the 2-fold anisotropy was determined
precisely using nuclear magnetic resonance (NMR). At
first, a two-axis rotating stage was developed to rotate asample precisely (Fig.3-6). Using this stage, the field angle(θ) dependence of the Si-NMR line width was precisely
determined (here θ is defined against the [110] crystal
direction in the basal plane). Based on this, the magnitudeof the 2-fold anisotropy was found to be smaller than thepreviously reported value (Fig.3-7).
A future superconductor with a higher Tc is considered to berealizable in a magnetic fluctuation-mediated superconducting
system, since the Tc of phonon mediated superconductors
has already reached its peak. If a new room-temperature
superconductor were to be found, it would be quite useful for
improving energy efficiency in electronic devices.
We are going to further investigate actinide compounds toclarify the mechanism behind superconductivity in detail.

Reference
Kambe, S. et al., NMR Study of In-Plane Twofold Ordering in URu2Si2, Physical Review Letters, vol.110, issue 24, 2013, p.246406-1-246406-5.

JAEA R&D Review 201443

Advanced Science Research

3-4 Hydrogen-Induced Insulation Degradation of Ceramic Capacitors− Microscopic Evidence Provided by Muons −BariumTitanium
Oxygen1.0

Positive muon that
released an electron

Electron weakly boundto a positive muon

Fraction

0.8

0.6

0.4

Positive muon thatbound an electron

0.2

Positive muon
0.0

−250

−200

−150

−100

−50

0

Temperature (°C)

Fig.3-8 A schematic of the muon-electron bound state in
BaTiO3
The widely spread electron orbital suggests that the electron isvery weakly bound to the positive muon in BaTiO3.

Fig.3-9 Electron release from the hydrogen-like muonelectron bound state in BaTiO3
The black squares and red circles represent the fractions ofpositive muons with and without a bound electron, respectively.
The solid curves are the best fits to an ionization model.

Multilayer ceramic capacitors (MLCCs) are fundamental
components of electronic circuits and are indispensable forstate-of-the-art electronic devices. Since the performanceof MLCCs heavily depends on that of dielectric materials,
controlling the quality of dielectric materials is vitally
important. As the internal electrodes of typical MLCCs
contain an easily-oxidizable element, the annealing process tounify the dielectrics and electrodes is performed in a reducing
atmosphere containing hydrogen (H). On the other hand,
this process can cause insulation degradation, depending on
the annealing conditions. Intensive studies are under way toidentify the origin of this degradation.
In this work, we focused on the risk of H incorporation
into dielectrics in the annealing process, and studied the
behavior of H in a typical dielectric material, barium titanate(BaTiO3). The electronic structure of H impurities in BaTiO3
is not obvious, since H can assume various charge states inmaterials. To tackle this issue, we used positive muons inplace of H. It is well established that the electronic structureof a muon-electron bound state is identical to that of H

except for small isotope corrections. Thus, we can imitate
H impurities in BaTiO3 with positive muons and selectively
probe their influences. Another advantage of using a muon is
its high sensitivity; one can obtain microscopic information
corresponding to H impurities in the dilute limit, which isdifficult to access by other experimental techniques.
A positive muon beam was irradiated to a BaTiO3 single
crystal at the J-PARC muon facility, and the local electronic
structure around the implanted muons was investigated bymeans of the muon spin rotation technique. We observed
a signal from muons that weakly bound an electron below−190 °C, which suggested that the electron orbital was widely
spread, as shown in Fig.3-8. The weakly bound electrons were
gradually released with increasing temperature, as shownin Fig.3-9. The released electrons could move freely aroundthe crystal and led to electric conductivity, thus decreasing
the insulating performance of BaTiO3. Hydrogen impurities
in BaTiO3 are also thought to release electrons according toa similar mechanism, resulting in insulation degradation at
device operating temperatures.

Reference
Ito, T. U. et al., Shallow Donor Level Associated with Hydrogen Impurities in Undoped BaTiO3, Applied Physics Letters, vol.103, issue 4, 2013,
p.042905-1-042905-4.

44

JAEA R&D Review 2014

Advanced Science Research

3-5 A Study to Reveal the Mechanism of Radiobiological Effects− DNA Damage Induces Aberrations in Normal Chromosomes in a Living Cell −UV-A（365 nm）
Microcells with ahuman chromosome
Microcell fusion Method(a)Unirradiatedmouse cells

Cell division
Cell division

(c)

(b)

(d)

(e)

(f)

Cell division

Clone cells with
exposed humanchromosome
Analysis of chromosome
aberrations

Fig.3-10 Transfer of damaged DNA by a microcell
fusion method
Microcells containing a human chromosome were
exposed to UV-A, and then transferred to unirradiatedmouse cells. The cells were cultured, and many clones
with a stable cell division function were selected. The
resulting chromosome aberrations were microscopicallyobserved in the cells and analyzed.

As one of the genetic effects through multiple repetitionsof cell division, chromosome aberrations are occasionallyinduced in the progeny cells after ionizing irradiation.
This is known to be “genetic instability,” which is strongly
involved in the induction of late radiation effects, such as
carcinogenesis. The detailed mechanism behind it, however,has not yet been fully understood.
In this study, we focused on DNA damage as a trigger of theinstability, particularly DNA base damage, which might persist
even after repeated cell divisions. The bases normally work as
elementary units of genetic code in DNA, and their alterationis thought to raise mutation frequencies. In addition to DNA
damage, irradiation could also induce damage to cytoplasmic
organelles. To exclude the cytoplasmic effects, we applieda microcell fusion method to observe genetic instabilities
induced only by DNA base damage.
Sub-cellular small structures called microcells, including
one human chromosome, were irradiated and then transferred
into unirradiated normal mouse cells (Fig.3-10). The humanchromosome is considerably different from those of the mouse
cell, such that the exposed human chromosome was easily
distinguished from unirradiated mouse chromosomes by meansof microscopic observations. It was confirmed beforehandthat normal chromosome transfer did not induce any genetic

Fig.3-11 Typical chromosome aberrations microscopicallyobserved
Human and mouse chromosomes are indicated in red and blue,
respectively. Yellow arrows show pieces of apparatus, the so-call“centromere,” used to pull the chromosomes into both daughter
cells during cell division. Normal human (a) and mouse (b)chromosomes are shown in the upper panels. The abnormal
chromosomes comprise a mouse chromosome and a humanchromosome (c), two human chromosomes (d), and two (e) andthree (f) mouse chromosomes, which are shown in the lower
panels.

instabilities. Ultraviolet light (wavelength: 365 nm), known
as UV-A, was used as the source of the irradiation, because it
has been known to preferentially induce the same sort of DNA
base lesions that are induced by ionizing radiation. After thechromosome transfer, the cells were cultured for 20 days or amonth to obtain various clonal cell lines that stably undergo
cell division.
The analysis of the chromosomes in the clones revealed thatcertain types of aberrations, such as fusions between normalmouse chromosomes, were frequently induced, in addition tofusions between the human chromosomes, or between humanand mouse chromosomes (Fig.3-11). Furthermore, it wasobserved that the number of chromosomes in the clone cells
abnormally increased from the normal number of 43 to twicethat number or more.
These evidences indicate that a novel mechanism of signaltransfer from damaged to normal chromosomes could resultin the genetic instability in the unirradiated cells. Detailed
investigation is expected to reveal the induction process ofradiation carcinogenesis in the low dose region in the near future.
The present study was partly supported by the Ministry of
Education, Culture, Sports, Science and Technology of Japan(MEXT) KAKENHI Grant-in-Aid for Young Scientists (B)(No.20710045).

Reference
Urushibara, A., Yokoya, A. et al., Induction of Genetic Instability by Transfer of a UV-A-Irradiated Chromosome, Mutation Research, vol.766, 2014, p.29-34.

JAEA R&D Review 201445

Advanced Science Research

3-6 Discovery of Uranium Migration in Groundwater with Silica Colloid− Devisal of an Effective Method for Elucidating the Chemical Forms of Trace Actinides
in Environmental Water −EluentUltraviolet-visible absorption
spectrometer(UV-Vis)Negatively charged column

Positively charged column
(c)

(a)

U

U

Sampleinjection
Pre-collected
colloid only
(Conventional
method)Whole water
sample(This study)

(d)

(b)
Iodide (I−)
Na+

I−
Na+

ColumnMass spectrometer(ICP-MS)

Fig.3-12 SEC-UV-Vis-ICP-MS analysis
Sample components are separated by their size as
they pass through a column of eluent. The eluent was
introduced on-line to an ultraviolet-visible (UV-Vis)
spectrometer and then to an inductively coupled plasma
mass spectrometer (ICP-MS) to analyze its components.
The conventional method analyzes colloids only, using
a single column. This study used two columns with
opposite charges to analyze both the colloid and theions by directly injecting environmental water into thecolumns. The results obtained by the two columns werecompared.

10

20
30
Retention time / min

40 10

20
30
40
Retention time / min50

Fig.3-13 SEC-UV-Vis-ICP-MS analysis of Horonobe groundwater(b), (d) Ultraviolet absorption. (a), (c) 238U. The sizes of components
can be estimated based on the retention times of colloids withknown sizes. Larger colloids are eluted faster. For a colloid, the sizes
estimated by two columns with opposite column charges are similar.
For an ion, very different sizes are estimated, because the retention
times differ greatly between the two columns, due to the interaction
between the ion and the charge of the column (b, d).

It is thought that actinides do not migrate in environmental
water (EW) in ionic form, but rather, migrate together withcolloids by being adsorbed onto them. However, this theory
has not yet been experimentally confirmed because there
have not been any methods for effectively investigating
trace actinides in EW and thereby determining what colloids
transport them and whether actinide ions can migrate bythemselves.
The SEC-UV-Vis-ICP-MS method has been used toanalyze colloids in EW. This method separates colloids by
their sizes and detects trace elements in them (Fig.3-12). In
the conventional method, relatively large colloids in EW arecollected in advance, and are then injected into a separation
column for analysis. This method has the disadvantage ofbeing incapable of analyzing ions and small colloids. To solvethese problems, we took note of the charge of the separation
columns. By injecting EW as it is into a separation column,
actinides in all states are detected. Since colloids and ions are
separated by sizes and charges, respectively, it is predictedthat colloids and ions can be distinguished by comparing thetwo results obtained by columns with opposite charges.
We analyzed trace uranium (U) in the groundwater collectedat Horonobe, Hokkaido (Fig.3-13). A single peak of U was
detected by each column. The sizes of the uranium-bearing
substances, estimated from the times when the peaks appeared,
were very similar to each other. This result shows that U waspresent as a colloid.
Trace U does not form an intrinsic (true) colloid. All
the silica contained abundantly in the groundwater was
eluted with U. This result revealed that U was present asa pseudocolloid, which comprised U adsorbed onto silica
colloid. To our knowledge, we are the first to have discovered
this type of pseudocolloid of U. The results also revealed that
this groundwater did not contain U ions, because no other U
peaks were detected.
This study devised a powerful analytical method for
elucidating the forms of trace actinides in EW. This method
is broadly applicable to various water samples, including
experimental samples, and future application in various
research fields is expected.

Reference
Kozai, N. et al., Characterization of Saline Groundwater at Horonobe, Hokkaido, Japan by SEC-UV-ICP-MS: Speciation of Uranium and Iodine, Water
Research, vol.47, issue 4, 2013, p.1570-1584.

46

JAEA R&D Review 2014

Advanced Science Research

3-7 Search for Λ6 H Hypernuclei− Can We Make Neutron-Rich Hypernuclei? −Natural stable nuclei(nb / sr / (MeV / c2))60
4Expand by

Unstable nuclei“glue-like effect”

Neutron-rich hypernuclei

B

10

Be

9

7

6

3

p

Li

He 4He

2

H

3

H

n

7

Li

6

B

11

B

12

B

13
Λ

B

Differential cross section

proton number

8

50

14
Λ

B

13
Be 10Be 11Be 12ΛBe ΛBe

8

Li

9

Li

10
Λ

Li

11
Λ

Li

8

He

He Λ9He

6
Λ

H

3
2

40

4
Λ

H+2n

1
30
0

5800

5810

5820

5830

20
10

7
Λ

H

0
5800

neutron number

5850

5900

−

Mass of X at π +6Li→K ++X reaction (MeV / c2)

Fig.3-14 Nuclear chart of ordinary nuclei and hypernuclei
The horizontal axis shows the number of neutrons, andthe vertical axis shows the number of protons. The greyboxes correspond to existing natural nuclei. The blue boxes
correspond to unstable nuclei measured by experiments. The
production of hypernuclei, indicated by yellow boxes, is possibleusing glue-like effects at J-PARC.

Fig.3-15 Mass of X in the reaction: π + 6Li K+ + X
A magnified view of the Λ bound region is shown in the inset.
The arrow labeled Λ4 H + 2n shows the particle decay threshold(5801.7 MeV / c2).

Ordinary nuclei comprise nucleons, i.e., protons and

The FINUDA (FIsica NUcleare a DAfne) group has reported–

neutrons. On the other hand, Λ hypernuclei comprise one Λ

three candidate events forparticle together with nucleons. The Λ particle, which is of thewas as large as about 1 MeV/c2. We should confirm the mass

species “hyperon,” has a new degree of freedom: strangeness.

with a larger amount of data.

This introduces an additional attractive force between the Λ. However, the error in the massThis study was conducted with the K1.8 beam line ofparticle and the nucleons, and causes a shrinkage of the nuclei.

J-PARC’s Hadron Experimental Facility.

We call this a “glue-like effect.” Using this, the nuclear chart

the reaction π + Li→K +X(X=expands in the strangeness sector. Fig.3-14 shows the nuclear

to one neutron and one Λ particle. Fig.3-15 shows the mass

chart of ordinary and hypernuclei. The region of hypernuclei

spectrum of X. The arrow in the inset indicates the sum of(represented by yellow boxes) can be expanded into theand two neutrons, and thus, the maximum-mass threshold of- 6

neutron-rich region.

is produced via), which changes two protons+. Unfortunately, no significant peak structure was observed

Thus, we attempted to produce a

hypernucleus, whicharound this arrow. An upper limit of the cross section wascomprises one proton, four neutrons, and one Λ. The coreestimated to be 1.2 nb/sr. From this result, we determined that

nucleus H is unbound. Success in making

the production cross section of5

by adding a Λmust be much lower if it

could allow the production of very neutron-rich nuclei, withexists and indicated that there is room for the discussion of theneutron-proton ratios as large as four.

theoretical model of the production mechanism.

Reference
Sugimura, H. et al., Search forHypernucleus by the 6Li(π‒, K +) Reaction at pπ = 1.2 GeV/c, Physics Letters B, vol.729, 2014, p.39-44.
‒

JAEA R&D Review 201447

４ Nuclear Science and Engineering Research
Promoting Basic R&D on Nuclear Energy and Creation of Innovative Technology to Meet Social Needs
Nuclear Science and Engineering Research
Supporting various activities ofgovernment and projects both insideand outside JAEA

Providing technological platform
to industries and universities

Tackling the recovery
from the accident at the TEPCO’s
Fukushima Daiichi NPS

Conservation and improvement of basis
for nuclear science and engineering

Preparing basic codes anddatabases

Creating innovative technology(Partitioning and transmutation, etc.)

Cultivating human resource andbasic technology

Fig.4-1 Roles of nuclear science and engineering research
We serve various roles to conserve and improve the foundation of nuclear science and engineering.

Nuclear Engineering Research Collaboration Center
Special Gr. for
Advanced
Reprocessing
Materials
Development

Special Gr. for
Advanced
Technology on
LWR Thermal
Hydraulics

Special Gr. forGraphite &Carbon Materials
Characterization

Special Gr. for
Vitrification
Technology

Special Gr. for
HTGRTechnology
Development

Special Gr. forSpecial Gr. forSpecial Gr. for
Generation
Advancing
Simulation ofTechnology using
Nuclear
Molten Materials
Accelerator
Hydrogen Safety

Nuclear Science and Engineering Center
Nuclear Data andReactorEngineering Unit

Fuels andMaterials
Engineering Unit

NuclearChemistry Unit

Environment andRadiation
Sciences Unit

Fundamental
Decommissioning
Technology
Development Unit

Partitioning andTransmutation
Technology Unit

Fig.4-2 Nuclear Engineering Research Collaboration CenterTo create innovative and practical products, joint research with industry and academia is being promoted.

Nuclear science and engineering research aims to promote
activities, as shown in Fig.4-1, including recovering from theaccident at the Tokyo Electric Power Company, Incorporated
Fukushima Daiichi Nuclear Power Station (1F) anddeveloping transmutation technologies to reduce the amountof long-lived radioactive waste. In the field of nuclear data
and reactor engineering, we are working on the Japanese
Evaluated Nuclear Data Library (JENDL), measuring nuclear
data, and developing methods to design nuclear reactors using
advanced theoretical, experimental, and simulation approaches(Topics 4-1, 4-2, and 4-3). In the fuels and materials
engineering field, we are promoting research on nuclear fuels
and materials used in reactors and on fuel-cycle facilities(Topics 4-4 and 4-5). In the nuclear chemistry field, we arepromoting the analysis of basic data for reprocessing anddetecting extremely small amounts of nuclear materials(Topics 4-6, 4-7, and 4-8). In the environmental and radiation
science field, we are promoting studies on radionuclides
in the environment and the development of a database for
48

JAEA R&D Review 2014

radiation protection (Topics 4-9 and 4-10). Joint research withindustry and academia is also being promoted via the Nuclear
Engineering Research Collaboration Center (Fig.4-2).
Developing the fundamental technology to decommission 1F
involves enhancing our capabilities to characterize fuel debris,to analyze the durability of materials, to develop criticality
management strategies, and to estimate radioisotope production(Chapter 1, Topics 1-1, 1-9, 1-10, 1-14, 1-16, and 1-19).
Using accelerator-driven subcritical systems (ADS), we are
studying technologies to partition minor actinide (MA) nuclides(e.g., neptunium and americium) and fission products fromspent nuclear fuel, and to perform MA transmutation.
Basic technology development is conducted at the JAEA
research facilities. These technologies target radiation
measurement (Topics 4-11 and 4-12) methods to assess
durability of capsule materials in the reactor exposed to longterm irradiation (Topic 4-13) and safety management systems
for the radiation-controlled area (Topic 4-14).

Nuclear Science and Engineering Research

4-1 Improving the Prediction Accuracy of the Decay Heat from Nuclear Force− Clarify Effect of Tensor Force on β Decay −

132

Sn

3

β decay

p

n

p

2

132

p

1

Excitation Energy ofTheory (Without Tensor)

Theory (With Tensor)

Repulsive

Half-Lifenn

Attractive Repulsive

n

p

132

Attractive

Fig.4-3 Interaction due to tensor force between
proton and neutron
The tensor force is either attractive or repulsive
depending on the direction of the “spin” (spin axis isrepresented by arrows in the figure) and the relative
positions of the proton and neutron. The attractive
tensor force plays an important role in β decay.

Experiment

Experiment
39 s
Theory (With Tensor)
72 sTheory (Without Tensor) 838 s Ground StateSb (MeV)p : proton n: neutron0

Sb

Fig.4-4 Half-life of 132Sn and excitation energy of daughter
nucleus 132Sb with and without tensor force
The horizontal lines in the figure indicate the excitationenergy of 132Sb. By considering the tensor force, the half-lifeof 132Sn and the excitation energy of 132Sb obtained by theory
approach the experimental results.

Fission products (FPs) generated in nuclear power plants

is not accounted for in the conventional models. The tensor

and fragments produced by nuclear disintegration reactions inforce is one of the fundamental nuclear forces that govern theaccelerators contain unstable nuclei. Unstable nuclei typicallyinteraction between protons and neutrons, the constituentschange into more stable nuclei by β decay. β decay involvesof the nucleus. Fig.4-3 shows the characteristic featureemission of an electron (positron), an antineutrino (neutrino),of the tensor force. The tensor force is either attractive or

and a photon, and delivers energy to the environment. The

repulsive depending on the “spin” of the particles involved.

energy released in β decay of a single nucleus is very small.

Spin is analogous to the rotation of the particle and the

However, spent nuclear fuel contains an enormous quantity

relative position of proton and neutron. We developed aof FPs, which produces a large amount of heat (decay heat).

new theoretical model that includes the tensor force and

The decay heat is one of the problems that complicate

numerically analyzed it to study how the tensor force affects

the reprocessing of spent nuclear fuel and the disposal of

β decay. We found that the attractive tensor force plays anradioactive waste. In addition, decay heat is the main reason

important role in β decay. As shown in Fig.4-4, the half-life

that lots of heat is generated in the damaged nuclear reactorsof the magic-number nucleus, tin-132 (132Sn), approaches theof the TEPCO’s Fukushima Daiichi NPS. However, someexperimental result and the excitation energy of its daughter

aspects of β decay remain unexplained by current theoretical

nucleus, antimony-132 (132Sb), is accurately reproduced.

models. In particular, theoretical models cannot reproduce

This result contributes to improving the accuracy ofenergy by β decay or the half-life of nuclei with magic

predicting decay heat and should be useful in various fields

numbers. This fact reduces the accuracy predicting decay heat.

of research, such as cosmic nucleosynthesis and fundamental

To solve this problem, we focus on the tensor force, which

research in nuclear physics.

Reference
Minato, F. et al., Impact of Tensor Force on β Decay of Magic and Semimagic Nuclei, Physical Review Letters, vol.110, issue 12, 2013, p.122501-1-122501-5.

JAEA R&D Review 201449

Nuclear Science and Engineering Research

4-2 Understanding the Behavior of Molten Material in Severe Accidents− Developing Numerical Simulation Method of Melt Relocation −(a) The early stage of the core melting(b) Melt relocates to the core plate

Fig.4-5 Transition of molten-material relocation(a) Fuels were melted by heating from above, and then themelted fuel relocated downward. (b) Molten fuels accumulate
on the core plate and some flows into the control-rod guidetubes (CRGTs). (c) Molten fuel enters the lower plenum
through the CRGTs and accumulates in the lower region.

(c) Melt intrudes into and accumulates
at the lower plenum

Intact fuelMolten structure

Molten fuelSolid structure

Solidified fuel after melt

In the accident at the TEPCO’s Fukushima Daiichi NPS

the remaining structure (nonheating element), we improved(1F), the meltdown is believed to have occurred in the reactor

the code so that it can distinguish between these two. We alsopressure vessel (RPV), following which the molten fueldesigned a numerical method that can efficiently calculate

reached the lower head. Although understanding in detailthe changing shape of the molten materials, which enables

how molten fuels distribute in the RPV is very important,the code to treat thermal hydraulics, including complicatedthe distribution is quite difficult to measure because of theinteractions between molten fuels and the other structureshigh radiation. In this case, a numerical simulation is useful.

shown in Fig.4-5. Thus, by providing a detailed treatment

Although some severe accident (SA) codes have beenof molten materials and a simulation of the relocation of theused to numerically simulate SAs, detailed simulations ofmulticomponent melt, we implemented a unified simulation ofmelt relocation remain very limited. In addition, because

the system, including the heating material (uranium fuel) andthe boiling water reactor (BWR) accident at 1F is without

the nonheating materials (Zry, SUS, B4C, etc.). Therefore, theprecedent in the world; most simulations to date deal withresults from this code are expected to provide information thatpressurized water reactors (PWRs). Therefore, we arewill complement the results obtained from existing SA codes.

developing a new simulation code to elucidate the details of

In the future, we will implement some physical models andmelt relocation in SAs. Because at least two components arealso validate the code to ensure an accurate understanding of

involved inside the RPV, uranium fuel (heating element) andhow melt behaves inside a RPV.

Reference
Yamashita, S. et al., Development of Numerical Simulation Method for Relocation Behavior of Molten Materials in Nuclear Reactors: Analysis of Relocation
Behavior for Molten Materials with a Simulated Decay Heat Model, Proceedings of the 22nd International Conference on Nuclear Engineering (ICONE 22),
Prague, Czech Republic, 2014, ICONE22-30972, 9p., in CD-ROM.

50

JAEA R&D Review 2014

Nuclear Science and Engineering Research

4-3 Improving Uranium Accountancy− Demonstration of a Novel Nondestructive Assay for Nuclear Waste Drums at Ningyo-toge −3

He detector tube

Shielding 0.5 m
concrete

Detector
bank

Drum
U

1.2 m

Neutrons
Neutron generator
Polyethylene moderator

Fig.4-6 Photograph of NDA device (JAWAS-N) built at
Ningyo-toge
This NDA device was constructed using 0.5-m-thick concrete
blocks, wherein a neutron generator, 3He detector bank, andturntable were installed (Fig.4-7). On the outside, the device is
2 m long, 2 m wide, and 2.2 m high. On the inside it is 1 m long,
1 m wide, and 1.2 m high.

When nuclear-facility operators place U-contaminateddismantling wastes in waste drums for strict storage, they arelegally required to implement nuclear material accountancy.
Therefore, the Ningyo-toge Environmental Engineering
Center (NEEC) of Japan Atomic Energy Agency (JAEA)
uses the passive γ-ray measurement method to determine theamount of uranium (U) contained in the drums.
However, with this method, the results of the measurements
are strongly perturbed when the matrix and U arenonuniformly distributed in the drum. Occasionally, themeasurement error is large and becomes a matter of great
concern for U accountancy.
To overcome this difficulty, the Nuclear Science andEngineering Center (NSEC) of JAEA developed a nondestructive
assay (NDA) for 235U called the fast-neutron direct-interrogation(FNDI) method. To promote the practical application of thismethod, an NDA device (JAWAS-N) that uses the FNDI
method was newly constructed at the NEEC (Fig.4-6).

Neutron counting (counts)

60000

Fig.4-7 Schematic innerview of JAWAS-N
The FNDI method directly
irradiates fast neutron to adrum and detects only thefast neutrons resulting fromthe nuclear-fission between
the moderated neutronsand 235U.

● Test data using metallic waste
in a drum

40000

20000

0
0

500

1000
235

1500

U mass (mg)

Fig.4-8 Linear relationship between 235U mass and theneutron counting measured in JAWAS-N
The mass of 235U contained in a waste drum is specified byneutron counting, whose results are independent of spatial(nonuniform) distribution of 235U.

In this method, fast neutrons (14.4 MeV) from a neutron
generator irradiate a waste drum, and the fast neutrons that
result from nuclear fission between the moderated neutronsand 235U in the drum are detected (Fig.4-7). Since fast neutrons
are not strongly affected by the nonuniform distribution of
matrix and U, the 235U mass can be determined by counting
the neutrons that correspond to 235U.
To date, fundamental research has shown that the resultof neutron counting increases in proportion with the 235U
mass in the drum, as shown in Fig.4-8; hence, the 235U mass
could be obtained with small error regardless of the chemicalcomposition and location of nuclear materials in the drum.
Since the FNDI method can precisely determine the total
mass of 235U, the other isotopes can be identified if theenrichment of U is known. Combining this method with theconventional methods enables the proper determination of themass of U (238U+235U). Therefore, more R&D is planned tofurther improve this method.

Reference
Japan Atomic Energy Agency, Measurement Method for Mass of Fissile Materials, and the Measurement Deivce, Patent Application Publication,
pub.no.JP2014-174123, 2014-09-22, (in Japanese).
JAEA R&D Review 201451

Nuclear Science and Engineering Research

4-4 Improving the Evaluation Accuracy of the Long-Term Integrity of Nuclear Reactors− Developing a Model of the Microstructural Evolution under Neutron Irradiation −(a)Prediction model
Prediction model ofmicrostructural evolution

Example ofmicrostructure
100 nm

Optimization

Surveillance test data

Prediction curve for future
neutron irradiation embrittlement

Defect cluster

Increase in the index
temperature for
embrittlement (℃)(b)
Prediction of future values

Impurity conc.: ~ 10−4
Impurity conc.: ~ 4 × 10−5
Impurity conc.: < 5 × 10−8(Carbon)

1025

Impurity conc.: 10−4
Impurity conc.: 10−5
Impurity conc.: 10−61024Previous model

1023
1022
1021
1020
10−5

10−4

10−3

10−2

10−1

100

Irradiation dose (number of displacements per atom)＊

Prediction curve
Surveillance test data

Number density of defect clusters (m−3)

Empirical
correlation

Tuning parameters

Present model

Experimental data

A model developed in this study
reducing the tuning parameters

Neutron irradiation dose (operation term)＊

Corresponds to neutron irradiation embrittlement

Fig.4-9 Schematic of the model used to predict embrittlement
due to neutron irradiation(a) Flow chart showing steps involved in creating curve to
predict irradiation embrittlement, and (b) schematic of theprediction curve. We reduced the tuning parameters bydeveloping a model to predict the microstructural evolution.

Fig.4-10 Density of defect clusters as a function of neutronirradiation dose
Compared with the prediction of a previous model (dashed line),
which did not include impurity effects, the present model (solidlines) with impurity effects better reproduces the experimentallyobserved defect density as a function of neutron-irradiation
dose.

Metallic materials subjected to long-term neutron irradiation
due to fission in nuclear reactors suffer microstructuralmodifications, which result in neutron-irradiation
embrittlement of these materials. To understand and predictthe rate at which irradiation embrittlement proceeds is thus
a vital issue for ensuring the long-term integrity of nuclear
reactors.
Fig.4-9 shows a schematic of the current model used to
predict neutron-irradiation embrittlement. The prediction curveis estimated by solving empirical equations that correlate thecalculated microstructural change to the increase in irradiation
embrittlement. By optimizing the parameters involved in
reproducing data on irradiation embrittlement obtained fromsurveillance pieces, the future integrity of nuclear reactors
can be predicted. However, these parameters must sometimes
be significantly modified, especially as the operating life of
nuclear reactors increases; hence, improving the accuracy ofthe model used to make these predictions is a crucial issue.
To address this issue, we modeled how the migration of
defect clusters (produced by irradiation) is affected by impurity
atoms as fundamental physical processes that were notconsidered in previous models. With this approach, we aim todevelop a versatile model capable of treating various metallic
materials and irradiation conditions in a unified fashion.
Based on the latest results of research on how impurities
affect the formation and the growth of defect clusters (which
is one of the primary causes of irradiation embrittlement), we
developed a model to predict the microstructural evolution ofmaterials and demonstrated its validity. Fig.4-10 shows how
defect-cluster density depends on the neutron-irradiation dosein pure iron, which is the basic material of nuclear reactors.
This model reduces the range over which parameters mustbe tuned and significantly improves the consistency between
model and experimental data (e.g., it reproduces the saturationin defect density upon increasing neutron-irradiation dose andthe defect density as a function of impurity concentration).
This result shows the importance of incorporating more
fundamental physical processes into prediction models. Byfurther developing the present model, we expect to improve
the accuracy with which the increase in neutron-irradiation
embrittlement is predicted, thus better securing long-term
operation of nuclear reactors.

Reference
Abe, Y. et al., Effect of Carbon Impurity Content on Microstructural Evolution in Neutron-Irradiated Alpha Iron: Cluster Dynamics Modeling, Materials
Research Society Symposium Proceedings, vol.1535, 2013, 7p.

52

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Nuclear Science and Engineering Research

4-5 Developing a Database of Fundamental Properties for Partitioning and Transmutation− Evaluating Thermodynamic Data on Intermetallic Compounds with Electrochemical Techniques −−0.8

Potentiostat

Digital thermometer

Reference electrode

Counter electrode

Working electrode

Glovebox

Argon
gas

Co-depositonof Ce and Cd
on the working
electrode

HeaterLithium chloride – Potassium chloride +Cerium chloride + Cadmium chloride

Potential relative to reference electrode (V)PCmeasured−1.2

Equilibrium potential of Ce
measured in advance,
Eeq (reference potential)

−1.6

E5
E3

E6

E4

E2

−2.0
E1

0

100

200

300

Time (s)Fig.4-11 Concept of the electrochemical method applied in
the present study
The Gibbs free energy of formation of Ce-Cd intermetallic
compounds is evaluated based on the plateau potentials ofintermetallic compounds formed by co-deposition of Ce and Cd
on the working electrode in molten salts of lithium chloride andpotassium chloride.

Fig.4-12 Chronopotentiogram of equilibrium potentialof Ce-Cd intermetallic compounds relative to referenceelectrode (673 K)
The red solid line indicates the variation in potential. Bluedashed lines indicate the equilibrium potential for Ce and the six
plateau potentials for Ce-Cd intermetallic compounds (E1–E6).

Partitioning and transmutation is a promising concept
for advanced nuclear fuel cycles. In this approach, minor
actinides (MAs) are recycled and transmuted to reduce theenvironmental impact of nuclear waste. Since MAs are highly
radioactive and give off significant decay heat, a pyrochemical
method is a proper candidate for reprocessing. In the
pyrochemical reprocessing method proposed by JAEA, MAs
are selectively collected in a liquid cadmium (Cd) cathodein the molten salt bath, following which they are recoveredas alloys by Cd distillation and separation. Developing sucha pyrochemical reprocessing requires a database of various
fundamental properties. Cd distillation and separation requires
thermodynamic data on various Cd intermetallic compounds
because the decomposition of the intermetallic compounds
may be dominant in the last stage of Cd distillation.
Fig.4-11 presents the electrochemical method used in thepresent study, in which the three-electrode technique wasadopted for electrochemical measurements, from which
the Gibbs free energy of formation of various intermetallic
compounds was determined.
Cerium-Cadmium (Ce-Cd) intermetallic compounds form byco-deposition of Ce and Cd metals on the working electrode.
The Gibbs free energy of formation of Ce-Cd intermetallic
compounds was determined from the difference between two
potentials: the redox reaction potential for Ce3+/Ce and theplateau potential for the Ce-Cd intermetallic compound (E1–E6)
in the molten salt (Fig.4-12).
These results represent the first determination of the Gibbs
free energy of formation of these six Ce-Cd intermetallic
compounds.
Based on the similarity between lanthanide (Ln) andactinide (An), we plan to use An with this electrochemical
method and evaluate the Gibbs free energy of formation of
An-Cd intermetallic compounds.

Reference
Shibata, H. et al., Evaluation of Gibbs Free Energies of Formation of Ce-Cd Intermetallic Compounds using Electrochemical Techniques, Journal of Physics
and Chemistry of Solids, vol.75, issue 8, 2014, p.972-976.
JAEA R&D Review 201453

Nuclear Science and Engineering Research

4-6 Extraction of Strontium Using a Macrocyclic Compound− Extraction Enhanced by Applying Ionic Liquid Solvents −

O

N
N

O

N
OH

N
N

N

O

O
O

β-Diketone

OH

O

β-Diketone

Crown ether

Fig.4-13 Chemical structure of the macrocyclic compoundz
100
Extractability (%)

Ionic liquid

y
x

80
60
40
20

Chloroform

0
4

5

6

7

8

9

Aqueous phase pH

Fig.4-14 Extraction of strontium in the ionic liquid andchloroform
Strontium can be extracted in a lower-pH region when using
an ionic liquid, resulting in significant improvement in theextraction.

Strontium-90 (90Sr), which is a fission product, is a β-ray
emitting nuclide with a half-life of 28.8 years. In the human
body, it accumulates in the bones. Thus, an efficient technique
is required to remove 90Sr from contaminated soil andseawater.
One effective technique to separate and remove metal
ions including radionuclides is solvent extraction, which is atechnique for extracting a target substance from an aqueous
solution into an organic solvent (containing extractants)
immiscible with water. This technique is widely used inindustry to separate and recover rare metals.
As an extractant for strontium, we synthesized a macrocyclic
compound (Fig.4-13) bearing two β-diketone moieties andused it to investigate solvent extraction with an ionic liquid.
Ionic liquids have received significant attention recently as areplacement for organic solvents in the extracting medium.
Ionic liquids are room-temperature molten salts composed
only of ions and are expected to serve as functional solventswith unprecedented advantages due to their properties, which
differ from those of conventional solvents.
To evaluate the advantages of using an ionic liquid as an
extracting solvent to extract strontium ions, we compared thecharacteristics of the ionic liquid for extraction with those of

Fig.4-15 Crystal structure of the strontium complex
extracted into chloroform
Strontium ( ) is coordinated by six oxygen ( ) and two
nitrogen ( ) atoms from the macrocyclic compound and one
oxygen ( ) atom from the solvent.
andare carbon andsulfur atoms, respectively.

chloroform, which is known as a general extracting solvent(organic solvent). We thus clarified that the extraction ofstrontium using the ionic liquid was remarkably enhanced
compared with that using chloroform (Fig.4-14). A subsequent
study comparing the extraction into the ionic liquid with that
into chloroform shows that the 1:1 complex of strontium
forms with the macrocyclic compound and is extracted
in both systems. Analysis of the crystal structure of thecomplex extracted into chloroform reveals that strontium is
encapsulated in the cavity of the ring and is coordinated by thetwo β-diketone moieties (Fig.4-15).
Furthermore, to understand why extraction into the ionicliquid is enhanced compared with chloroform, the structuresof the complexes extracted into both solvents were analyzed
by X-ray absorption spectroscopy. The strontium–oxygen
distance in the complex in the ionic liquid is 2.51 Å, whichis less than the distance found for the complex in chloroform(2.55 Å). Therefore, we conclude that strontium is morestrongly held in the ionic liquid and the shorter strontium–oxygen distance leads to the enhanced extraction.
The present study was accomplished in collaboration with
Kanazawa University.

Reference
Okamura, H. et al., Specific Cooperative Effect of a Macrocyclic Receptor for Metal Ion Transfer into an Ionic Liquid, Analytical Chemistry, vol.84,
issue 21, 2012, p.9332-9339.

54

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4-7 Controlling Valences by Electrocatalysis of Pt Particles− Rapid and Selective Valence-Control Method for Actinide −+0.5 V
U(VI)
Np(V)
Pu(III)

GC fiber(a) GC+0.15 V
U(VI)
Np(IV)Pu(III)(b) Pt/GC

0.8
0.4
0
Potential vs Ag/AgCl (V)Fig.4-16 Flow electrolysis with column
electrode
The column electrode is a working electrodeof GC fiber packed tightly in a porous-glasstube diaphragm. Electrolysis while letting
the sample solution flow through the GC
fibers enables a rapid and effective valence
control of ions.

IV)→
Np(

III)

−0.4

0.8

V)
U(VI)→U
(I

Np(

V)→
N

p(IV

)

Np(

(III)
Pu(VI)→Pu
Np(V
I)→N
p

V)

(V)

Pu(V
)→

VI

U(VI)→U(I

Sample solution

Pu(V
I)→P
u(V)

V

Np(V
I)→N
p(V)

Valence states (–)IV

Np(V)→
Np(III)

Reduction

glass diaphragm

Pu(II
I)

III

0.4
0
Potential vs Ag/AgCl (V)−0.4

Fig.4-17 Valence changes of U, Np, and Pu ions by flow electrolysis
Valence changes of U, Np, and Pu as a function of potential applied byelectrolysis with (a) GC fiber column electrode and (b) platinized GC (Pt/GC)
fiber column electrode. For the Pt/GC electrode, Np can be controlled to be
tetravalent. Hence, different oxidation states can be prepared for the U, Np,
and Pu ions.

Actinides (An) such as uranium (U), neptunium (Np), andplutonium (Pu), in aqueous solutions exist as ionic species
ranging from trivalent to hexavalent configuration. The
chemical properties of these solutions depend on the valenceof the ions; hence, a proper understanding of how to precisely
adjust the valence of An ions in solution is required for
efficient chemical separation and precise analysis.
We are developing a flow electrolysis method that uses asworking electrode the column packed with glassy carbon (GC)fiber (Fig.4-16). Flow electrolysis is conducted by passing
the solution through the column electrode, which can rapidly
reach exhaustive electrolysis. The redox of An ions between
tetravalent (An4+) and pentavalent (AnO2+) accompanies thetransformation of ions. Consequently, the reaction is slow,
and large overpotentials are necessary. We find that the
overpotential decreases as a result of electrocatalysis when
platinum (Pt) is used as the working electrode material.
Reaction rates at normal Pt surfaces are very small, however,
and a long time period is needed to reach exhaustive
electrolysis.
To address this problem, we tried to develop a rapid
electrolysis wherein the column electrode exploitedelectrocatalytic effect. A column electrode with the platinized

GC (Pt/GC) fiber serving as electrode was prepared byelectrodeposition of platinum black onto GC fiber from achloroplatinic acid solution.
The relationships between the applied potential and thevalence of U, Np, and Pu upon electrolytic reduction with
GC and Pt/GC electrodes are shown in Fig.4-17. At theconventional GC electrode (a), Np(V) is immediately reduced
to Np(III), whereas Pu(V) is slowly reduced to Pu(III) becauseof the large overpotential for Np(V) and Pu(V) reduction.
These results indicate that the selective control of valence isdifficult. However, at the Pt/GC electrode (b), Np(IV) can be
precisely prepared. Thus, U, Np, and Pu ions can be preparedin various oxidation states (e.g., U(VI), Np(IV), and Pu(III)when +0.15 V is applied).
The results of this study may be applied to obtain precise
coulometric speciation of individual valence ions bymeasuring electric charge during electrolysis. They may also
be applied to controlling the valence for the separation of Anions.
The present study was partly supported by the Ministry of
Education, Culture, Sports, Science and Technology of Japan(MEXT) KAKENHI Grant-in-Aid for Scientific Research (C)(No.22560832).

Reference
Kitatsuji, Y. et al., Flow Electrolysis of U, Np and Pu Ions Utilizing Electrocatalysis at a Column Electrode with Platinized Glassy Carbon Fiber Working
Electrode, Electrochimica Acta, vol.74, 2012, p.215-221.
JAEA R&D Review 201455

Nuclear Science and Engineering Research

4-8 Selective Detection of Higher Enriched Uranium Particles− Secondary Ion Mass Spectrometry Combined with Solid-State Nuclear-Track Detection −(a)Film containing particles

Nuclear track detector

U particles

Tracks10 μm

Fig.4-19 Identification of U particles
A film containing particles was fabricated and placed incontact with a nuclear-track detector. U particles wereidentified by observing the tracks.

(b)

0.20
0.15
0.10
0.05

234

10 μm

U abundance ratio (%)

0.25

0.00

Fig.4-18 Images of (a) fission tracks and (b) alpha tracks
from U particles
Fission or alpha tracks were created in a nuclear-track
detector by U particles and were observed with an optical
microscope.

Undeclared nuclear activities related to nuclear-weapon
production should be monitored to ensure peaceful use of
atomic energy. We routinely analyze environmental samples
taken at nuclear facilities by inspectors of International Atomic
Energy Agency (IAEA). Based on the isotope ratio results,
IAEA verifies nuclear materials used and nuclear activities in
the facilities.
Secondary ion mass spectrometry (SIMS) is a techniqueused to analyze isotope ratios. Here, individual particles areirradiated with focused ion beams and sputtered ions of eachisotope are detected. To measure individual particles with
SIMS, the locations of such particles in the sample should be
exactly identified before analysis. In this study, we developed
a technique to identify uranium (U) particles by detecting
nuclear-tracks. The procedure is as follows. A film containing
particles is fabricated and placed in contact with a nucleartrack detector. The sample is then irradiated with thermal
neutrons in a nuclear reactor. When U in a particle absorbs
neutrons, the U undergoes fission, which creates tracks inthe detector. Since the tracks can be observed with an optical

0

10

5

15

20

235

U abundance ratio (%)

Fig.4-20 U isotope abundances in individual particles
taken at a nuclear facility and measured with SIMS
Two particles with 235U abundances over 9% were detected.
Such particles were detected only when using nuclear-track
detection.

microscope (Fig.4-18(a)), U particles can be identified byobserving the tracks (Fig.4-19). As another method, alphatracks created by the alpha decay of U can be used to identify
U particles (Fig.4-18(b)). In this case, no neutron source isnecessary.
Since the number of tracks relates to the 235U isotope
abundance in the particle, we can identify more-enriched U
particles by selecting the particles with a larger number oftracks. The identification of U particles abundant in 235U isespecially important to detect nuclear activities related tonuclear-weapon production. Fig.4-20 shows a SIMS result
for a sample taken from a nuclear facility. Here, we identified
particles with a larger number of tracks and analyzed the
U isotope abundances with SIMS. Consequently, twoparticles with higher 235U abundances can be detected, which
is impossible to do by SIMS without prior nuclear-track
detection.
The present study was sponsored by the Secretariat of the
Nuclear Regulation Authority (NRA).

Reference
Esaka, F. et al., Secondary Ion Mass Spectrometry Combined with Alpha Track Detection for Isotope Abundance Ratio Analysis of Individual UraniumBearing Particles, Talanta, vol.120, 2014, p.349-354.

56

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4-9 Mechanism of Greenhouse Gas Emission from Rice Paddy− Simulation of Atmospheric Ammonia Exchange Using the Land Surface Modelfor Transport of Radioactive Materials −(a)Exchange(deposition/
emission) Chemicalfertilizer
Compost
NH3
Brown rice

0
−2
−4
−6
−8

N2

−10
12/5

N2O

Straw

NO

12/7

12/8

12/9

12/10

12/11

12/12

12/13

6
4

Soil process

Irrigation

Absorption
Soil

NH3 exchange(nmol m−2 s−1)

Paddy
water

Run-off(b)

Nitrogenfixation

Rice

12/6

Winter 2009 (fallow period: bare soil condition)

Denitrification
Seedling

NH3 exchange(nmol m−2 s−1)

Output from paddy field

Calc.

2

Input to paddy field
N2O

Obs.
4

2
0
−2
−4

Leaching

−6
7/28

7/29

7/30

7/31

8/1

8/2

8/3

8/4

8/5

Summer 2010 (cropping period: flooded water condition)

Fig.4-21 Nitrogen cycling at rice paddy
Greenhouse gases, such as nitrous oxide, relate to various
processes. Bidirectional NH 3 exchange (deposition andemission) between the atmosphere and the rice paddy is hardto accurately be quantified.

Fig.4-22 SOLVEG simulation of NH3 exchange over rice
paddy (a) during fallow period and (b) cropping period
Upon incorporating NH3 exchange process into SOLVEG, themodel reproduced the dynamics observed in NH3 exchange(positive: emission, negative: deposition) bet ween theatmosphere and the rice paddy.

To develop the model for various environmental materialtransport studies, we applied the SOLVEG (Multi-layer
Atmosphere-SOiL-VEGetation Model) to those studies,
which is the land surface model for prediction of radioactive
materials transfer between atmosphere and land surface. In
the present study and in cooperation with National Institutefor Agro-Environmental Sciences, Japan, we newly modeled
ammonia (NH3) exchange process above rice paddy fields.
Rice paddies are not only the major arable ecosystems providing
the staple food in Asia, but also substantial sources of greenhousegases such as nitrous oxide (N2O). The emission of greenhousedepends on complicated processes of nitrogen cycling between
the atmosphere, rice paddies and soil (Fig.4-21). NH3 in theatmosphere above paddy fields is absorbed by the stomata
of rice foliage and/or is dissolved by paddy water (NH3deposition). However, NH3 is often emitted from the paddy
when nitrogen fertilizer is provided (NH3 emission). The
accurate estimate of this bidirectional NH3 exchange (thedifference between NH3 emission and deposition) is notstraightforward and remains a key issue in the prediction of
the greenhouse gas emission from rice paddies.
Hence, we developed new modules for SOLVEG to predictthe paddy water temperature and the NH3 concentrations in thefoliage and at the surface of paddy water or soil. The modified
SOLVEG model reproduced the observed NH3 exchange rate
at the Mase rice paddy site in Ibaraki, Japan (Fig.4-22). We
also investigated how amount of NH3 is exchanged during rice
growth. The result revealed that the recapture by the foliageof NH3 from the surface of the paddy water (NH3 recapture)tends to increase with rice growth. Once maturity is reached,most volatilized NH3 is recaptured in rice canopies and nolonger transfers to the atmosphere. By allowing the reductionof the vaporization loss of NH3 to the atmosphere, this result isuseful for managing supplemental fertilization of rice paddies.
The present study was partly supported by Japan Society for
the Promotion of Science (JSPS) KAKENHI Grant-in-Aid for
Scientific Research (A) (No.22248026).

Reference
Katata, G. et al., Coupling Atmospheric Ammonia Exchange Process over a Rice Paddy Field with a Multi-Layer Atmosphere-Soil-Vegetation Model,
Agricultural and Forest Meteorology, vol.180, 2013, p.1-21.
JAEA R&D Review 201457

Nuclear Science and Engineering Research

4-10 Influence of Physical Characteristics on Internal Dosimetry− Use of Human Models to Compare Caucasians with Japanese −Physique

Average of
ICRP reference value ofJapanese adult male Caucasian adult male

Height (cm)

170

176

Weight (kg)64

73

Average of
ICRP reference value ofOrgan mass
Japanese adult male Caucasian adult male
(kg)
Liver

1.600

1.800

Colon

0.330

0.370

Adipose tissues

13.90018.200

Lungs

1.200

1.200

Thyroid

0.019

0.020

Brain

1.470

1.450

Kidneys

0.320

0.310

Assessing the internal dose due to the intake ofradioisotopes (RIs) requires the knowledge of the ratio ofenergy deposited in the target organ to energy emitted byradiation from the source regions. These ratios are called thespecific absorbed fractions (SAFs) and are calculated withhuman models that reproduce the shapes of organs, tissues,
and body. For dose assessments, the International Commission
on Radiological Protection (ICRP) has defined reference
human models (ICRP models) that have the heights, weights,
and organ masses appropriate for Caucasians. In the future,
SAF data and effective doses per unit intake of RI (i.e., dosecoefficients) will be published based on the calculations using
the ICRP models.
The typical Japanese body size, however, is generally
smaller than that of a Caucasian, and organ masses forJapanese also differ from those of Caucasians, as shownin Table 4-1. We analyzed discrepancies in SAFs anddose coefficients between Japanese and Caucasian due todifferences in body size and organ mass (these are the physicalcharacteristics), because dose coefficients given by the ICRP
are used to establish radiation-protection standards in Japan.
We developed a human model named JM-103, whose

1600
1500
1400

Number of cases

Table 4-1 Averages of Japanese and reference valuesof Caucasians for physiques and organ masses inadult male
Body size for Japanese is usually smaller than that for
Caucasian. However, the relative mass of each organ fora Japanese with respect to that for a Caucasian differs.

1000
900
800
300
200
100
0
-20 -15 -10 -50

5 10 15 20 25 30 35 40

Difference in dose coefficients (%)
Fig.4-23 Histogram showing difference in dosecoefficients between JM-103 and ICRP models
The percent difference was calculated as follows:(Dose of JM-103 / Dose of ICRP model − 1) × 100

physical characteristics are those of an average Japanese adultmale. In this study, calculations of SAFs using JM-103 were
performed for photons and electrons at 15 energies from 10 keV
to 5 MeV, and for various combinations of 41 source regions
and 33 target organs. Comparing the results of this calculation
with SAFs based on the ICRP model reveals differences
between the SAFs derived from organ masses and the weightsof the two models. Next, we calculated the dose coefficients(De) using the SAFs of JM-103 and the ICRP model for
2894 cases, considering intake pathways and the chemical
forms of 923 RIs.
Fig.4-23 shows the distribution of the difference in De
between the JM-103 and ICRP models. The maximum De
was about 40%, which was due to the differences in the SAFs
between the models. However, the results show that, in 97% of
all cases calculated, the De ranged from −10% to 10%. Thus,
only in a limited number of cases the De were significantly
affected by differences between the representative physicalcharacteristics of Japanese and Caucasians. Thus, we conclude
that the dose coefficients used by ICRP are applicable to theJapanese population for radiation protection.

Reference
Manabe, K. et al., Comparison of Internal Doses Calculated using the Specific Absorbed Fractions of the Average Adult Japanese Male Phantom with Thoseof the Reference Computational Phantom-Adult Male of ICRP Publication 110, Physics in Medicine and Biology, vol.59, no.5, 2014, p.1255-1270.

58

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Nuclear Science and Engineering Research

4-11 Ensuring the Reliability of Measurements of Exposure of Workers and Public to High-Energy γ-rays− Establishment of the High Energy γ-ray Calibration Field −Beam line for the high-energy
γ-ray calibration field

60°

Dose Monitor
2”φ × 2” NaI (Tl)
600 cmPMMA Build up Plate (BUP)(30 × 30 × 3 cm3)
BUP was positioned 15 cm in front ofa point of test
Point of test0°

Proton beam
CaF2
Target (Pt backing)10−17

Calculated exposure (C kg−1 photon−1)

4 MV Van de Graaff accelerator

100 cm

Fig.4-24 Schematic of the high-energy γ-ray calibration
field (at Facility of Radiation Standards)

The 6-7 MeV high energy γ-ray calibration field wasestablished at the Facility of Radiation Standards and serves toensure the reliability of readings for dosimeters used in highenergy γ-ray fields.
The reliability of dosimeter readings is ensured by the so-called“calibration,” which donates the periodic checks of whether
dosimeter readings are correct. Well-defined radiation fields,
called “calibration fields,” are required for proper calibration.
The FRS has established various types of radiation calibration
fields, including X-ray, γ-ray, β-ray and neutrons, which itoffers to companies and institutes working in the field ofradiation protection.
For exposure to γ-rays, fixed γ-ray calibration fields are only
available up to 2 MeV, and dosimeters have been calibrated inthese fields. In contrast, the emission of 6 MeV γ-rays resulting
from the 16O(n, p)16N reaction in the coolant water is widely
recognized as a substantial source of radiation exposure,which is particularly relevant for BWRs. Furthermore, fortherapeutic and industrial purposes, γ-rays with energies
greater than several tenths of MeVs are produced by electron
accelerators and high-energy X-ray generators. The calibration
field for 6-7 MeV γ-rays should tell us how widely available
dosimeters respond to γ-rays with these energies or higher.

Ideal exposure
Build-Up Cap
Build-Up Plate (30 cm × 30 cm) + thin BUC

10−18

10−19
10−2

10−1

100

101

Total thickness of a build-up material (cm)Fig.4-25 Exposures calculated at test point as a functionof thickness and shape of build-up materials
Ideal exposures were calculated based on the γ-ray fluence at
100 cm and the mass-energy absorption coefficient for dry air,which reflects the attenuation of γ-ray fluence with increasing
thicknesses of build-up materials. The ideal exposure
represents the exposure predicted when electrons due to
γ-rays fully contribute to charges collected at a point of test as
well.

However, no high-energy γ-ray calibration field is available inJapan for radiation protection. We thus developed a 6-7 MeV
γ-ray field produced by the nuclear reaction 19F(p, αγ)16O using
the 4 MV Van de Graaff accelerator at the FRS (Fig.4-24) for
calibration purposes.
Because of the high energy of incident γ-rays, introducing
an ionization chamber (IC) with a thicker build-up cap isrequired to precisely determine air kerma rates at a point of test
(100 cm). This means that measurement conditions significantly
differ from those of routine calibrations. We investigated themethod of using a conventional IC coupled with the samebuild-up plate (BUP) introduced for routine calibrations.
Throughout a series of Monte Carlo calculations, as shownin Fig.4-25, we assessed the measurement arrangement forthe conventional IC and for the BUP, and found the optimalarrangement to be a BUP made of polymethylmethacrylatewith the dimension of 30×30×3 cm3 and positioned 15 cm infront of a point of test. The method would allow us to properly
calibrate the dosimeters that are most widely available.
In addition, introducing a cylindrical 2”ϕ × 2” NaI(Tl)scintillation detector as monitor allows us to quite precisely
determine reference air kerma rates at a point of test whenirradiating with γ-rays for regular calibrations.

Reference
Kowatari, M. et al., Measurement of Air Kerma Rates for 6- to 7-MeV High-Energy γ-ray Field by Ionisation Chamber and Build-Up
Plate, Radiation Protection Dosimetry, vol.162, no.4, 2014, p.446-458.
JAEA R&D Review 201459

Nuclear Science and Engineering Research

4-12 Measuring Radiation Intensity in Nuclear Reactors− Developing γ-ray Detectors for Remote Inspection in Primary Containment Vessel −several tens to several hundreds mm

(a)

(d)

Core wire

EmitterCollector

Insulator

Current (A)a few mm
MI cable(b)( × 10−10)
3
φ 3 × 50 mm
φ 3 × 100 mm
φ 3 × 150 mm

2

1

Controlling andData Collecting PC

Irradiation Room

0
0

SPGD

1

2

3

4

5

Dose rate (kGy/h)

Co Source
GPIB Cable(e)

MI Cable( × 10−12)6
φ 2 × 100 mm
φ 3 × 150 mm

Switching Picoammeter
Unit

(c)
SPGD

Current (A)Coaxial
Cable

4

2

0
0
Dosimeter

Thermocouple

50

100

Dose rate (Gy/h)

Fig.4-26 Schematic of SPGD and irradiation experiment(a) SPGD is a γ-ray detector with a diameter of only a few
millimeters and comprises an emitter, collector, and insulator.
(b) Output currents from SPGDs when irradiated undersame conditions that were measured by a switching unit
and picoammeter. (c) Irradiation conditions were precisely
determined by thermocouple and dosimeters positioned near
the SPGDs.

Fig.4-27 Output currents of SPGDs with γ-ray-dose rate(d) Output currents of SPGDs used in this study were
proportional to the dose rates up to and over 4000 Gy/h.
(e) Lower-detection limit was about 10 Gy/h.

The fuel debris must be removed to decommission the
TEPCOʼs Fukushima Daiichi NPS (1F). However, the highradiation environment precludes a visual confirmation of the
primary containment vessel (PCV) at 1F.
Therefore, we are developing a method to measure radiationthat uses a self-powered gamma detector (SPGD) as a means to
obtain the distribution of the debris in PCVs. The SPGD does
not require a complex structure and an external electric powersource because the output current is obtained from electrons
generated by γ-rays impinging on the emitter. This means that
the SPGD can be only a few millimeters in diameter and cantherefore access intricate piping systems (Fig.4-26).
To determine the applicability of the SPGD be to 1F, wefabricated thin SPGDs with lead (Pb) emitters of various
lengths and diameters, and investigated their γ-ray detection
limits. The relationship between the γ-ray intensity and theoutput current was obtained in irradiation rooms where the of
γ-ray intensity distribution was known. The results indicatethat the length of emitter influences the output current morethan the diameter.
Until now, the dose rate in the PCV was considered to beup to tens of gray per hour, with the higher-intensity radiation
doubtless occurring closer to the core. In a high-dose-rateenvironment, the output current was linear in dose rate to over
4000 Gy/h (Fig.4-27(d)). In a low-dose-rate environment,
above 10 Gy/h, the average discrepancy with respect to theapproximate curve (dashed lines) was estimated to be ~14%,
as indicated by dash-dotted lines (Fig.4-27(e)). Thus, the
SPGDs used in this study can measure the γ-ray-dose rate withfair accuracy. These results show that the measuring rangeof the SPGDs corresponds to the intensity of radiation in the
PCV of 1F and that the SPGDs may serve as a detector in suchan environment.
By leveraging the results of the present study, we plan tofurther improve the structure of the SPGDs to increase their
sensitivity and their ability to interrogate intricate pipe systems.

Reference
Takeuchi, T. et al., Development of a Self-Powered γ Detector, Journal of Nuclear Science and Technology, vol.51, issues 7-8, 2014, p.939-943.

60

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Nuclear Science and Engineering Research

4-13 Technical Development for In-Pile Irradiation-Assisted Stress Corrosion Cracking Test Capsules− Research on Safety Evaluation of Light Water Reactors −
(a)

(b)

50

100

800

40

80

600

●
■
▲
●

30

400

20

●
■
●
●
■
▲

200
0
200

26

2

▲ 1.0 × 10 n/m
26
2■ 2.5 × 10 n/m
26
2● 3.9 × 10 n/m

300
400
500
Test temperature (K)

10
0
600

Fraction of SCC (%)

1000

Total elongation (%)

Yield stress (MPa)Fig.4-28 Assembling works in hot cell(a) Pre-irradiated specimens are installed in an in-pile test capsule and (b) welding is
conducted in a hot cell by the remote-welding apparatus.

60

SSRT in water (DO=8 – 32 ppm)
Neutron fluence: 2.0 –3.9 × 1026 n/m2
Test temp.: 423 – 533 K

40This study
Kodama et al.
Nakano et al.

20
0
300

400
500
Test temperature (K)600

Fig.4-29 Result of tensile test in air
Irradiated specimens are sufficiently ductile at the estimatedtemperature of the outer tube of the pre-irradiation capsule.

Fig.4-30 SCC fracture rate measured by SSRTThe SCC fracture is not observed in irradiated specimensat the estimated temperature of the outer-tube of the
pre-irradiation capsule.

Irradiation-assisted stress corrosion cracking (IASCC) is
recognized as one of the key safety issues for managing thecore components of light water reactors (LWRs). We plan to
do in situ crack-propagation experiments to simulate IASCC
in the Japan materials testing reactor (JMTR).
To perform these experiments, specimens must be irradiatedup to the threshold of IASCC. Next, the specimens are to be
relocated to in-pile test capsules and the capsules are to beassembled. Two technical hurdles remain to be solved: one is thedevelopment of welding and assembling techniques appropriate
for capsules in a hot cell, and the other is the development ofmethods to evaluate the integrity of pre-irradiation capsules thatwill be irradiated in the JMTR.
We therefore developed a remote-welding apparatus thatcan rotate a capsule to perform circumferential welding.
We tested the resulting welds many times by varying theparameters of rotation speed and welding current. However,obtaining a suitable weld in a single welding pass proved
impossible. Thus, we used two-pass circumferential welding.
This method efficiently preheats the first weld, which is thencompletely by the second weld, making it possible to weld andassemble capsules in the hot cell. Fig.4-28 shows photographsof welding and assembling.
We also evaluated the integrity of the structural material of
the capsules. When specimens are irradiated, the structural
materials of the pre-irradiation capsule also receive irradiation
up to the threshold fluence of IASCC. Thus, we must evaluate
the integrity of the heavily irradiated structural materials of
the capsules.
Specimens were fabricated from stainless steel irradiatedin the JMTR for over 20 years (neutron irradiation at
1.0–3.9 × 1026 n/m2). Tensile tests in air and slow-strain-rate
testing (SSRT) in water were performed.
Fig.4-29 shows the result of the tensile tests, and Fig.4-30
shows the SCC fracture rate determined after SSRT.
Under irradiation, the temperature of the outer tube of thecapsule is estimated to be 423 K. These results show that thestructural material of the capsule is sufficiently ductile andwill not undergo SCC fracturing at the operating temperature.
By overcoming these two hurdles, we can perform the in-pile
IASCC tests required to investigate life-time management,
which is important for evaluating the safety of LWRs.

Reference
Shibata, A. et al., Development of Remote Welding Techniques for In-Pile IASCC Capsules and Evaluation of Material Integrity on Capsules for Long
Irradiation Period, Journal of Nuclear Materials, vol.422, issues 1-3, 2012, p.14-19.
JAEA R&D Review 201461

Nuclear Science and Engineering Research

4-14 Developing a Real-Time Management System for Workers in Radiation-Controlled Areas− Worker-Safety Management in Radiation Controlled Areas −JMTR
Location
SensorWireless
Dosimeter with a
Postural Sensor

Items carriedInstalled wirelessreceiversin the reactor
building

Real-time multifunctional access‐control system- Collect worker’s information- Data management by server- Transmission of data to related people

・Location information (where they are in the building?)
・Time information (how much time has possessed since entry?)・Dose information (how much exposure?)
・Posture information (standing or lying?)

Sharing worker’s information
inside and outside
the reactor building in real time

Management by
work situation
outsidethe reactor
building

Check of work’s
status inside thereactor building

Fig.4-31 Schematic of a real-time multifunctional access-control system
Developed using the patent of Hitachi Aloka Medical, Ltd. and the joint patent of JAEA. Access control andindividual dose are managed in real time.

In a radiation-controlled area, such as a reactor building, thesafety management of access control and dose for workers,
including prevention of radiation hazards, is required. Advances
in management techniques are always required, as is the use of
the latest technology.
In the Japan Materials Testing Reactor (JMTR), no system,
such as an administrator, that was able to manage the access
control and individual dose for workers in the reactor building
in real time, existed. Therefore, the administrator had the
difficult job of managing, and this management system
was difficult for the workers in the reactor to understand.
Furthermore, information sharing was sometimes insufficienton both sides.
To confront this situation, the JAEA and Hitachi Aloka
Medical, Ltd. developed a “real-time multifunctional accesscontrol system” (Fig.4-31), which comprises items that workers
carried (location sensors and wireless dosimeters with posturesensors), wireless receivers, and data acquisition servers.
Each worker’s location, time of entering, dose, and posture
can be displayed on a PC or tablet in real time. Information
may now be shared between the administrator and workers,

between workers, and with people inside and outside the reactor
building. The status of workers (where they are in the building,
how much time has elapsed since their entry, their exposure,and whether they are standing or lying) is monitored by the PC.
The management is improved due to the visibility offered by agraphical display of the worker’s dose and location on a map.
Furthermore, safety management is facilitated by an alarm thatis triggered as a function of dose, posture, and entering time.
In particular, combining the wireless dosimeter and posturesensor is useful for safety management because the posture
information can quickly save a worker who has an accident,
even when he/she is working alone.
Safety management of workers in a radiation area has
been improved by the installation of this system, and by
the information sharing it promotes between administrators
and workers, and with people inside and outside the reactor
building. In addition, since this system is wireless and based on
notebook PCs, it can contribute to the management of workers’
conditions in outdoor decontamination projects, such as thedecontamination of the Fukushima area, which does not havesufficient radiation control.

Reference
Hiyama, K. et al., Development of Safety Management System for Works in Radiation Controlled Area (Joint Research), JAEA-Technology 2013-045, 2014,
32p. (in Japanese).

62

JAEA R&D Review 2014

５ Quantum Beam Science Research
Quantum Beam Science and Technology Researches and Research Sites− R&D Using Quantum Beam Facilities and Fundamental Technologies −“Quantum beam” is a generic term for neutron beams, ion beams, electron beams, high-intensity lasers, and synchrotron X-rays,
which are generated from accelerators, high-intensity laser facilities, and research reactors. Recently, “quantum beam technology”
has been extensively developed, with the most advanced manufacturing and observations being conducted using highly controlled
quantum beams.
There are several large quantum beam facilities such as J-PARC, TIARA, J-KAREN, SPring-8 beamlines, and JRR-3 in JAEA.
We have generated numerous results from these facilities. In this chapter, we introduce representative research results on “quantum
beam” science and technology.

Development of Beam Technologies for Quantum Beam Facilities
Technological Development at J-PARC
J-PARC comprises a series of three proton accelerators—Linac, 3 GeV synchrotron (RCS), and 50 GeV synchrotron—and three experimental facilities. The facilities include theMaterials and Life Science Experimental Facility (MLF),
which provides neutron and/or muon beams for a wide rangeof studies, the Hadron Experimental Facility for nuclear
and particle physics experiments using particles such as
K-mesons, and the Neutrino Experimental Facility for theT2K experiment of particle physics using neutrinos. All theseexperimental facilities are open to users from across the globe.
In the fiscal year (FY) 2013, there were significant advancesat Linac and RCS toward the upgrade of the beam power
to 1 MW. Annular-ring coupled structure (ACS) cavities(Fig.5-1), developed in J-PARC, were installed to upgrade
the acceleration energy from 181 MeV to 400 MeV at thedownstream of Linac. The newly installed ACS section isshown to have the following features. The accelerating beams
are very efficient and suppress the transverse accelerating
field caused by the axial symmetry of an ACS. At the RCS,
by virtue of the newly developed beam tuning techniques, thebeam losses can be reduced to an acceptable level not only at
the injection point but also in the acceleration section.
Consequently, we successfully delivered a 532 kW beamTechnological Development at TIARAThe Takasaki Ion Accelerators for Advanced Radiation
Application (TIARA) consists of four ion accelerators. Together
with an electron accelerator and gamma-ray irradiation facilities,
the TIARA accelerators are available to researchers in JAEA
and other organizations for R&D activities on new functionaland environmentally friendly materials, biotechnology, theradiation effects of materials, and quantum beam analysis.
Practical technological developments currently in progress
involve microbeam formation, single ion hits, techniques forlarge-area uniform irradiation, efficient beam transmission andacceleration (Topic 5-17) at the cyclotron, three-dimensional
in-air PIXE analysis, and a three-dimensional microfabrication
technique at the electrostatic accelerators.
In FY 2013, a technique for controlling the irradiation field shapeof the large-area uniform irradiation of a several-hundred-MeV
heavy ion beam was developed using a multi-pole magnetic
field. The beam intensity distribution is made uniform by a pairof octupole magnets, and the shape of the uniform irradiation
field is controlled by quadrupole magnets (Fig.5-2). According
to both beam optical calculations and experimental results,
the two quadrupole magnets closest to the irradiation chamber
are most effective for shape control. A quantitative method

to the neutron source of MLF, the highest power thus far
reported. A test operation with a 1 MW beam was scheduled
for October 2014.
Users at MLF conducted experiments at 18 neutron and2 muon beamlines with a proton beam power of 300 kW over4 run cycles. The number of submitted proposals was 533,
approximately the number submitted in FY 2012. This chapter
presents the outcomes from MLF (Topics 5-13, 5-14) and RCS(Topics 5-15, 5-16).

Fig.5-1 Annular-ring coupled structure (ACS) cavities
installed at the downstream of Linac

for estimating the shape and uniformity of the field was alsoestablished using a radiochromic film. Irradiation fields ofvarious sizes and aspect ratios, such as rectangles and ribbons,
were formed using these techniques.
(a)

Film

Octupole-Quadrupole Magnets

(b)
13 cm
9 cm

(c)

Fig.5-2 Large-area uniform
irradiation field measured
with a radiochromic film
Using (a) the multi-polemagnet system, Ar beams
of 520 MeV and 385 MeV are
formed into (b) a rectangle
and (c) a ribbon. The shaded
areas show beam intensity
distribution. The uniformity of(b) and (c) is ±7% and 10%,respectively.
2 cm

17 cm

JAEA R&D Review 201463

under the Nanotechnology Platform Project entrusted to us by
MEXT. This project was granted a supplementary budget for
FY 2012, which was invested in new equipment for nitride
semiconductor film formation, and renewing the kappa-type
X-ray diffractometer of two JAEA beamlines in FY 2013.

Kansai Photon Science InstituteIn the Kizu District, we eagerly engage in R&D activities
concerning lasers. For instance, we develop high quality lasers
and study the use of electron beams, ion beams, and X-rays
generated by high-intensity short-pulse lasers. Currently, we
are improving the quality of our high-intensity short-pulse
lasers (Fig.5-3). Since FY 2013, we have been installinghigh repetition rate pump lasers and renewing the pulse
compression gratings to achieve a more frequent and higher
power beam supply.
In the Harima District, we have been developing andimproving a state-of-the-art analysis technique for expressing
the functional and reaction mechanisms of materials. For
this purpose, we employ four JAEA synchrotron radiation
beamlines at SPring-8. These beamlines are also applied
to nanotechnology, energy, and environmental studies; forexample, a decontamination technique for revitalization at
Fukushima.
Furthermore, external researchers are now being supported

Fig.5-3 JAEA Kansai Advanced Relativistic ENgineering(J-KAREN) laser system

Advanced R&D on Quantum Beam Science and Technology
JAEA’s large quantum beam facilities have been employed
in diverse science and technology fields, mainly forsustainable living and green innovations (Fig.5-4).
By controlling the beam parameters, quantum beams canbe used to probe atomic- or molecular-level information.
Quantum beams can further investigate materials at nanometerscales (i.e., at the atomic or molecular level), because theyinteract with the material’s constituent atoms to change theirconfiguration, composition, and electronic state. In medical
applications, these beams are used for radiotherapy, which
focuses a beam onto cancer cells (Fig.5-5).
We have been developing new beam sources and enhancing
the beam intensities. These improvements will assist
the clarification of unknown phenomena and extend theapplicability of quantum beams. Additionally, quantum beam“probes” have yielded good results in materials science,
environment and energy, medicine, and biotechnology, which
are related to life sustainability and green innovations.
In this chapter, we introduce our recent research (Topics 5-1,

5-2, 5-3, 5-4, 5-5, 5-6, 5-7, 5-8, 5-9, 5-10, 5-11, and 5-12) on
advanced beam technology and the application of quantum
beams to the above mentioned fields.
We are also contributing to the recovery from the accident at
the Tokyo Electric Power Company, Incorporated Fukushima
Daiichi Nuclear Power Station. For example, we have beendeveloping improved decontamination materials (Chapter 1,
Topic 1-12).
Functions of Quantum Beam
Observe

CreateObservation at
atomic resolution

Atomic level
fabrication andcreation

Cure

Dr. Ｑ

Cure andtreatment
at cell level

Fig.5-5 Characteristics of quantum beamsLife Innovation

Green Innovation
Material Science

Medicine andBiotechnology

Environment
and Energy

Restoration and Revival
after the accident
at the TEPCO’s
Fukushima Daiichi NPS

Advanced BeamTechnology
NeutronProtonSynchrotron
X-ray

JRR-3

Research Reactors

TIARA

Radioisotopes

Electron

Ion

Photonγ-ray

J-PARC

Accelerators

SPring-8

High-Intensity Laser Facilities

LasersFig.5-4 Quantum beam facilities and research system for quantum beam science and technology in JAEA

64

JAEA R&D Review 2014

Quantum Beam Science Research

5-1 Biocompatible Plastic Fabrication by Ion Beam
− Development of Micro/nanofabrication and Surface Modification TechniqueUsing Focused Ion Beam for Medical Applications −(a)(b)
200 nm1 μm(c)

Intensity (a.u.)

Non-irradiated
O-C=O

(d)

C-O

C-C/C-H

1E+15 ions/cm2
C=C

1 μm

z
292

y
x, y: 5 μm

x

290

288

286

284

282

Binding Energy (eV)z: 50 nmFig.5-6 Microfabricated poly (L-lactic acid) using focused ion
beam
Poly (L-lactic acid) was precisely fabricated by a 30 kV Ga focusedion beam; (a) 80-nm-diameter holes, (b) 60-240-nm-wide grooves,(c) micro concave-convex structure, (d) 100-nm-wide alphabets.

Micro/nanofabrication techniques for fabricating biocompatible
materials are essential for the development of devices usedin advanced medical treatment and biological research.
Biocompatible materials have low heat and chemical stabilities;
therefore, they cannot be precisely fabricated by any standard
method. Here we developed a new technique for the micro/
nanofabrication of biocompatible plastic, which exhibits locallyhigh cell adhesion, using a focused ion beam (FIB).
As the target material, we selected poly (L-lactic acid)(PLLA), a highly biocompatible and biodegradable polymer.
PLLA is typically used in medical applications such as
implants and sutures that are degraded and absorbed by
the body after healing. The irradiation effect on the surface
modification of PLLA was investigated using 30 kV Ga FIB.
The heat induced by FIB irradiation enhanced the diffusionand desorption of the decomposed PLLA fragments. On
the other hand, fabrication accuracy was reduced by thethermal deformation of the heat-unstable PLLA. To increasefabrication precision, we reduced the thermal deformation
by optimizing the sample preparation and irradiation
conditions, such as the dose, dose rate, PLLA thickness,

Fig.5-7 Chemical bonds on the fabricated bottom ofpoly (L-lactic acid) measured by X-ray photoelectron
spectroscopy
Under irradiation by a focused ion beam, the peaks
attributed to the C-O bond and to the C-C and C-Hbonds at 284.8 eV decreased, while that attributed to
the C=C bond at 284.1 eV increased.

and beam diameter. Micrographs of successfully fabricatedmicrostructures, such as fine holes, grooves, and alphabets, arepresented in Fig.5-6.
Fig.5-7 shows the X-ray photoelectron spectra of the C1s
in PLLA before and after irradiation. It appears that the PLLA
surface was directly carbonized and became C=C-bond rich.
Because decomposed fragments such as H2, CO2, and CH4
were physically sputtered and desorbed to/from the PLLA
surface, the surface gradually became similar to a diamondlike carbon (DLC). The number of C=C bonds is known tostrongly affect cellular adhesion to DLC surfaces. Thus, thesurface modification technique can be used to control cellularadhesion onto PLLA.
This micro/nanofabrication technique was achieved bymerging fine processing and material modification techniques.
It is potentially applicable to the manufacture of biocompatible
devices, such as micromachines and lab-on-a-chip, used intissue engineering and bio-research to enhance the longevity
and well-being of society.
The present study was accomplished in collaboration with
Osaka University and Waseda University.

Reference
Oyama, T. G., Nagasawa, N. et al., Micro/Nanofabrication of Poly(L-lactic acid) using Focused Ion Beam Direct Etching, Applied Physics Letters, vol.103,
issue 16, 2013, p.163105-1-163105-4.
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5-2 Application of Composite-Type Optical Fiber Technique for Medical Instrument Development− Industrialization of Composite-Type Optical Fiber Technique and Startup
of JAEA-Supported Venture Business −Optical fiber
for image transmission(~φ4 μm, ~9000 pixels)

Cautery laser fiber(~φ0.1 mm, 1 unit)

Fiberscope

Light source~φ1 mm

Optical fiber
for illumination light transmission(~φ30 μm)

Lens

Front view

Sectional viewFig.5-8 Schematic of the composite-type optical fiberscope system
This configuration enables the fabrication of a composite-type opticalfiberscope suited to each medical field.

Composite-type
optical fiberscope
Image
processing
devicePC

Laser source

Image device

Outer dia.: φ1.02 mm
φ0.2 mm

Same axle type
Observation & Treatment
Diameter: φ1.1 mm
LASER and IMAGE istransmitted on the same axis.
Parallel axle type
Observation & Treatment
Diameter: φ0.83 mm, φ1.02 mm
LASER and IMAGE istransmitted on the parallel axis.

Coupling
deviceBlood flow
measurement
deviceLength: 3 m
Tip
φ0.4 mm

Extra-fine type
Observation
Diameter: φ0.83 mm
The finest diameter in the lineup

PD LaserThe PDT device for a peripheral lung-cancer
treatment (unapproved medical device)

Outer dia.: φ0.83 mm

Fig.5-10 Example of industrial development of the device
This basic system was developed by the JAEA-authorized
venture business for the treatment of peripheral lung cancer.
Four medical institutions are promoting a clinical study of the
new treatment.

Fig.5-9 Minimally invasive laser treatment system and the tipsof various fiberscopes
This system enables various tests and laser surgeries.

Minimally invasive treatments have been promoted byrecent advances in medical engineering. Such treatments may
reduce the burden of care and abridge hospitalization time,enhancing patients’ quality of life. The composite-type opticalfiber can transmit an image and a laser beam in parallel. With
this technology, a specific tool has been developed that allows
remote manipulation of the thermo-nuclear experimental
reactor and the large-sized nuclear installation at JAEA.
Because the developed technique is applicable to various
regions in the body, it may greatly benefit the burgeoning
minimally invasive surgery market. We are applying thistechnique to the development of various medical instruments.
Fig.5-8 shows the configuration of the composite-type
optical fiberscope (same axle type). The center core andsurrounding optical fibers transmit the laser beam and theimage, respectively. The left panel of Fig.5-9 is a photographof the minimally invasive laser treatment system. The rightpanels show the tip of the fiberscope adapted to various tests
and treatments. Currently, we are trialing this system on lungcancer therapy.
Although central lung cancer can be treated byphotodynamic therapy (PDT) at the transbronchial site,
PDT is unsuitable for early peripheral lung cancer because
existing endoscopes cannot reach the peripheral lung field.
On the other hand, a composite-type optical fiberscopewith an outside diameter of 1 mm can be inserted into theperipheral lung field, allowing detailed observation andprecise irradiation of peripheral lung cancer. The developed
fiberscope is also expected to provide a new PDT laser surgery
device. Toward these goals, we have patented the compositetype optical fiberscope, and accumulated knowledge of the
instrumental assembly. Furthermore, we have networked withrelated organizations and several medical scientific societies.
The composite-type optical fiber technique has been evaluatedas compatible with medical needs and eminently suited to a
new business enterprise.
Therefore, it was adopted in the FY 2012 “Program for
Creating STart-ups from Advanced Research and Technology”(START Program), which competes for government grants
from Ministry of Education, Culture, Sports, Science andTechnology (MEXT).
These efforts resulted in the successful launch of the new
venture business “OK Fiber Technology Co., Ltd.” of JAEA
in September, 2013. The business aims to distribute the device
based on the composite-type optical fiber technique acrossthe globe. In this venture, the PDT device for peripheral lungcancer treatment is being continuously developed (Fig.5-10).
The present study was partly supported by the Program for
Creating STart-ups from Advanced Research and Technology(START Program) “The industry and medical tool development
project based on the composite-type optical fiber” of the
Ministry of Education, Culture, Sports, Science and Technologyof Japan (MEXT).

Reference
Oka, K. et al., Medical Application of Composite-Type Optical Fiberscope, Oyo Butsuri, vol.80, no.12, 2011, p.1069-1072 (in Japanese).

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Quantum Beam Science Research

5-3 Doping Effect of Ferroelectrics− Improvement of Ferromagnetic Mechanism of Bismuth Ferrite −
P
1.5

P
E
0

μ (E)

1.0

Mn-K edgeZn- and Mn-codoped

Mn-doped

0.5
0.0
+2

Fig.5-11 Hysteresis loop of ferroelectric material
An electric field applied to a ferroelectric material redistributes
the electric charge in the material. This phenomenon iscalled polarization. The direction of polarization in a crystal isvisualized by the D-E hysteresis loop.

6520 6540 6560
Photon Energy (eV)

6580

6600

Fig.5-13 XAFS spectra of doped Mn
The valence of Mn differs between samples doped with both
Zn and Mn, and those doped with Mn only.

Zn, Mn codoped annealed

100
Polarization (μC/cm2)

6500

+3

50
0
−50

pure BiFeO3 × 5

P

E

−100
−100

−50
0
50
Electric Field (kV/cm)

100

Fig.5-12 D-E hysteresis loop of bismuth ferrite
Although bismuth ferrite is slightly affected by an external
electric field, its ferroelectric properties are dramatically
improved by doping with small amounts of Zn and Mn, and ahysteresis loop develops.

When a ferroelectric thin film is implanted with a small
number of ions from the ion beam irradiation equipment, its
ferroelectric property is often improved. However, as few
ions can become embedded in a thin film, a detailed structural
analysis is precluded, and the mechanism by which the ionsimprove the ferromagnetic properties cannot be clarified.
However, ion implantation into bulk medium induces the same
effect, allowing investigation of the doping effect.
Although bismuth ferrite (BiFeO 3 ) exhibits small
polarization, its ferroelectric property is dramatically improvedby codoping with small quantities of zinc (Zn) and manganese(Mn). Figs.5-11 and 5-12 illustrate the polarization principle
and the D-E hysteresis loop of BiFeO3, respectively. The
coercive electric field of BiFeO3 is very large and spontaneouspolarization is rare. However, spontaneous polarization isincited by codoping with Zn and Mn.
Here we investigated ferromagnetic improvement in BiFeO3
codoped with small quantities of Mn and Zn codoping. As
the valence of Zn is stable, the investigation relies on theFig.5-14 Defect-induced polarization reversal
Since the valence and local structure of the host phase differsfrom that of the dopants, the doped Zn and Mn generate
defect-induced polarization, which nucleates domain reversal.
Consequently, the ferroelectricity of bismuth ferrite improves
and the hysteresis loop opens.

fluctuating valence of Mn. Moreover, local structure analysis
is more suitable than conventional crystal structure analysis
because the surrounding doped ions break the translational
symmetry of the BiFeO3 structure. The X-ray absorption finestructure (XAFS) was then measured using beamline BL14B1
in SPring-8, which is dedicated to JAEA.
Fig.5-13 shows the results of the XAFS measurements.
The Mn valence state in codoped BiFeO3 (+2) differs fromthat in Mn-doped BiFeO3. Because hetero-valence ion doping
alters the structure of the bismuth sites, it induces polarization
defects. These defects initiate domain reversal in the presenceof an electric field. The polarization reversal induced by
the defect, which improves the ferroelectric property of thecodoped BiFeO3, is shown in Fig.5-14.
The present study was partly sponsored by the Ministry of
Education, Culture, Sports, Science and Technology of Japan(MEXT) KAKENHI Grant-in-Aid for Scientific Research (C)(No.21560877).

Reference
Yoneda, Y. et al., Electronic and Local Structures of Mn-Doped BiFeO3 Crystals, Physical Review B, vol.86, no.18, 2012, p.184112-1-184112-11.

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5-4 Distinction between Polarizations Induced by Magnetic Cycloids and Exchange Striction− Importance of Magnetic Cycloids for Multiferroic Materials −

(a)
Mn4+

P by magnetic cycloidal structure

P by magnetic exchange striction

Polarization (−)

Polarization (+)

OYTotal P
P by cycloid

YMn2O5100

Mn3+
P (nC/cm2)

75

MnO6

octahedron

50
25

(b)
V

Unpolarized neutrons
incident on the sample

Diffracted neutrons
are polarized

0
−25
−273

−263

−243
−253
Temperature (℃)−233

Fig.5-15 Cycloidal structure of YMn2O5 and experimental
set-up(a) Mn4+ spins () form a cycloidal structure.
(b) The cycloidal structure () aligns the diffracted neutron
spins.

Fig.5-16 Polarization (P) in YMn2O5
At temperatures between −238 °C and −253 °C, magnetic
exchange striction induces positive P, whereas, at temperatures
below −253 °C, cycloidal structures induce negative P. Data
points indicate P induced by cycloidal structures estimated frompolarized neutron diffraction.

We successfully distinguished the electric polarizationinduced by magnetic chiral structures and magnetic exchange
striction in multiferroic compounds by simultaneously
measuring the spontaneous polarization and polarized neutron
diffraction.
The magnetoelectric (ME) effects of multiferroic
compounds arise from coupling between magnetism andferroelectricity. This characteristic of multiferroic compounds
is expected to be exploited in future devices. Below 240 °C,
RMn2O5 compounds (where R denotes the rare-earth element
Bi or Y) establish ferroelectricity and magnetic order as
their Mn4+ spins spontaneously form cycloidal arrangements(Fig.5-15(a)). Below −250 °C, this system displays a richvariety of ME effects depending on the R atoms. The originof the ME effect in this compound has been explained by twomodels, one driven by the cycloidal structure and the other bymagnetic exchange striction. However, the correct model has
yet to be clarified.

We found that both models can be distinguished by thecharacter of polarized neutrons which can detect cycloidal
structures. To this end, we performed polarized neutron
diffraction measurements on a single YMn2O5 crystal. The
electric polarization was controlled by electric fields provided
by electrodes attached to the crystal surface. Fig.5-15(b) showsa schematic of the experiment. By simultaneously measuring thechirality of the cycloidal structure and the electric polarizationin the presence of an electric field at different temperatures,
we found that exchange striction occurs below −238 °C (asevidenced by the positive polarization in Fig.5-16), whereas amagnetic cycloidal structure develops below −253 °C (giving
rise to negative polarization). Given the rich ME effects below−253 °C, we infer that polarization induced by cycloidal
structures is an important property of multiferroic materials,
and can be exploited in future applications.
The present study was accomplished in collaboration with
Tohoku University and is summarized in the Reference.

Reference
Wakimoto, S. et al., Role of Magnetic Chirality in Polarization Flip upon a Commensurate-Incommensurate Magnetic Phase Transition in YMn2O5,
Physical Review B, vol.88, issue 14, 2013, p.140403-1-140403-5.

68

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5-5 Development of an Aluminum-Based Hydrogen Storage Alloy− To Realize Light-Weight Hydrogen Storage Alloys −180Time (s)Hydrogenation

Intensity

10 GPa, ca. 800 ˚C

120
60
0

0.26

0.24

0.22

0.20

0.18

AluminumHydrogen

CopperLattice spacing (nm)Fig.5-17 In situ synchrotron radiation X-ray diffraction
profiles of hydrogenated Al2Cu
Closed circles ( ) indicate the Bragg peaks from Al2CuH.
The hydrogenation reaction began 60 s after incubation of thesample at ca. 800 °C.

Fig.5-18 Crystal structure of Al2CuH
The interstitial nature of the synthesized aluminum-based
hydride, Al2CuH, was confirmed both experimentally and by
theoretical calculations.

Developing safe and efficient hydrogen storage is among
the primary technological challenges to realizing a hydrogenbased economy. Lightweight hydrogen storage materials are
required for automotive applications. Aluminum is a suitablematerial because it is lightweight, harmless, and abundant innature. However, although complex aluminum hydrides havebeen extensively investigated, they have not been considered
for hydrogen storage applications. Apart from complex
aluminum hydrides, few aluminum-based interstitial hydrides
have been synthesized to date, although interstitial hydride—referred to as a “hydrogen-storage alloy”—exhibits
excellent hydrogen reversibility.
We synthesized an aluminum-based interstitial hydride byhydrogenating a powdered aluminum–copper alloy (Al2Cu)
at high pressure and temperature. Under these conditions,hydrogen becomes extremely reactive. High-pressure andhigh-temperature conditions were generated by a cubic-type
multi-anvil apparatus. The hydrogenation conditions were
explored by an in situ synchrotron radiation X-ray diffraction
measurement system installed on the BL14B1 at SPring-8.
Fig.5-17 shows the X-ray diffraction profiles of Al2Cu

hydrogenated at 10 GPa. After 60 s incubation at ca.
800 °C, new Bragg peaks (● ) began to appear in the
profile. The appearance of the peaks indicated that Al2Cu
was hydrogenated to Al2CuH. The formed hydride wasrecovered at ambient conditions, and its crystal structure wascharacterized by a powder X-ray diffractometer. The obtained
crystal structure of Al2CuH (Fig.5-18) is consistent with theformation of an aluminum-based interstitial hydride. This
structure is further supported by first-principles calculations.
We conclude that an aluminum-based interstitial hydride,Al2CuH, was successfully synthesized. The experimental
and theoretical results of this study will assist the explorationof other aluminum-based interstitial hydrides, and thedevelopment of practical lightweight hydrogen storage
materials.
The present study was partly sponsored by New Energy
and Industrial Technology Development Organization(NEDO), and by Japan Society for the Promotion of Science(JSPS) KAKENHI Grant-in-Aid for Scientific Research(Nos. 25220911, 25420725).

Reference
Saitoh, H. et al., Synthesis and Formation Process of Al2CuHx: A New Class of Interstitial Aluminum-Based Alloy Hydride, APL Materials, vol.1, no.3, 2013,
p.032113-1-032113-7.
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5-6 Synchrotron Radiation Mössbauer Absorption Spectroscopy Improved by Electron Detection− Highly Efficient Mössbauer Spectroscopy of Many Periodic Table Elements Is Now Achievable −Vacuum chamber(He-flow cryostat)
SR

Sample(YbB12)Detector
Electrons
X-raysVelocitytransducer
Oscillations in Scatterer
Doppler velocity (YbB12)

Intensity

Scattering
Signals

Absorption
ProfileCounts (/10 h)

2000

Synchrotron Radiation Mössbauer spectrometer

1.3 mm/s

1900
1800
1700
1600
YbB12 T=20 K

1500
1400
−15

−10

−5
0
5
Velocity (mm /s)

10

15

Fig.5-20 174Yb Mössbauer spectrum
The transmitter is YbB12 at 20 K, and the scatterer is YbB12 at
26 K.

H

Scan

He

Li Be

Unsuitable

B

C

N

O

F Ne

Na Mg

Mössbauer-active probe

Al

Si

P

S

Cl Ar

K Ca Sc Ti

Velocity

Rb Sr

Y

Cs Ba ＊

Fig.5-19 Schematic of the SRMS spectrometer (upper
panel), an outside view of the APD detector placed in a
vacuum chamber (lower left panel), and scheme of theenergy scan process for obtaining the Mössbauer spectrum(lower right panel)

Synchrotron radiation Mössbauer spectroscopy (SRMS)
is a powerful tool used in diverse research areas. SRMS
provides information on magnetism, valences, crystal fields,
and electron densities by exploiting the hyperfine interactions
between the nucleus and its surrounding electrons. So far,the spectrum has been measured by detecting the fluorescent
X-rays and γ-rays released after nuclear resonant absorption
while largely disregarding the concurrent internal conversion
electrons (ICEs). To effectively utilize the ICEs, we developed
an advanced SRMS spectrometer that detects the ICEs alongwith the X-rays and γ-rays. The device is constructed froma windowless avalanche photodiode detector and a vacuum
chamber with a helium (He) gas-flow cryostat (Fig.5-19). In
this system, the synchrotron radiation (SR) X-rays transmitted
through the target sample (transmitter) are resonantly scatteredby the reference sample (scatterer). The resulting spectrum isacquired by counting the delayed scattering signals from theresonantly excited nuclei in the scatterer, and plotting them as
a function of the Doppler velocity.
As a feasibility study, we measured the Mössbauer spectrumof the 76.5 keV excited state of 174Yb, which exhibits a short
lifetime (τ = 2.58 ns). The experiment was conducted at theV

Cr Mn Fe Co Ni Cu Zn Ga Ge As Se Br Kr

Zr Nb Mo Tc Ru Rh Pd Ag Cd In Sn Sb Te
Hf Ta W Re Os IrI

Xe

Pt Au Hg Tl Pb Bi Po At Rn

Fr Ra ＊＊ 104~
＊Lanthanide

La Ca Pr Nd Pm Sm Eu Gd Tb Dy Ho Er Tm Yb Lu＊＊Actinide

Ac Th Pa U Np Pu Am Cm Bk Cf Es Fm Md No Lr

Fig.5-21 Elements highlighted in light blue are suitablefor Mössbauer spectroscopy

undulator beamlines BL11XU and BL09XU of SPring-8.
YbB12 powder samples were used as both transmitter andscatterer. The counting rate (measurement efficiency) of theconstructed device was up to 500% higher than that of theprevious detection system, and the spectrum was obtained
within 10 h. A clear absorption profile is shown in Fig.5-20. Theline width is approximately 1.3 mm/s, sufficient to determine
the 174Yb-hyperfine parameters.
In summary, a novel system for acquiring SRMS
was successfully developed at SPring-8. The improved
measurement efficiency enabled the first observation of
the 174Yb SR-Mössbauer spectrum at 76.5 keV. Clearly, thedeveloped device will enable SRMS studies of many other
Mössbauer-active nuclides (Fig.5-21). Moreover, a highbrilliance SR source offers new research possibilities intransmission and scattering Mössbauer experiments. We
believe that SR-based Mössbauer spectroscopy has entered a
new stage in both fundamental and practical studies.
The present study was partly supported by Japan Society for
the Promotion of Science (JSPS) KAKENHI Grant-in-Aid for
Scientific Research (S) (No.24221005) and a Grant-in-Aid for
Research Activity Start-up (No.24810014).

Reference
Masuda, R., Mitsui, T. et al., Synchrotron Radiation-Based Mössbauer Spectra of 174Yb Measured with Internal Conversion Electrons, Applied Physics
Letters, vol.104, no.8, 2014, p.082411-1-082411-5.

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5-7 Toward the Development of Protein Absorbents for Rare and Harmful Metals− Structural Characteristics of Halophilic Proteins Clarified by a Quantum Beam −

(a)

HaAP

HaBLA

(2.8)

(2.1)

HaNDK
(1.6)
Wild-type dimer(b)

Mutant tetramer

Hydrophobic cluster

Zn2+
Zn2+
Mg2+

Fig.5-22 X-ray crystal structures of the halophilic proteins(a) Molecular surfaces of HaNDK, HaBLA, and HaAP. The numbers
in parentheses are the ratios of positively charged to negatively
charged residues, indicating the acidic residue contents of proteins.
(b) Hydrophobic cluster (magenta sticks) formed by the internalhydrophobic residues of HaAP. This cluster may attract a substrate(an aromatic or aliphatic phosphoester) to the catalytic site (white
sticks). Mg2+ and Zn2+ are known to catalyze the enzymatic reaction ofalkaline phosphate.

Catalytic site

Halophilic bacteria living in salt-rich environments
such as salt lakes, halite, and salted foods produce special
proteins called “halophilic proteins.” The surfaces of these
proteins exhibit abundant quantities of acidic amino acids;
consequently, they become negatively charged. We have
exploited this property to develop protein absorbents ofrare and harmful metals. As the molecular design of such
absorbents requires knowledge of the protein’s tertiary
structure, we conducted X-ray crystallographic analysis ofseveral halophilic proteins; namely, nucleoside diphosphatekinase (HaNDK), β-lactamase (HaBLA), and alkaline
phosphatase (HaAP) (Fig.5-22(a)). Here we describe thestructural study of HaAP, which possesses the highest numberof acidic residues among the three candidate proteins.
A detailed structure of HaAP (Fig.5-22(a)) was acquired
using the X-ray beams at SPring-8 and the Photon Factory.
The negative charge density at the molecular surface of
HaAP (0.0028 e /A2) is strikingly higher than those of alkaline
phosphatases from other halophiles and a non-halophilic
Escherichia coli strain (0.0004–0.0019 e/A2), whose X-ray
structures are known. This property suggests that HaAP caninteract with various metal ions, including rare and harmful
metal ions.

Next, we clarified the relationship between the structure andenzymatic function of HaAP. Halophilic proteins usually lose
their enzymatic activity at low salt concentrations, as theirtertiary structures are destabilized by electrostatic repulsion
between the side chains of the acidic residues. However,
HaAP is exceptional because it maintains enzyme activity
over a wide salt concentration range (1–4 mol/ℓ NaCl). Our
structural study revealed more internal hydrophobic residues
in HaAP than in alkaline phosphatases from other halophiles
and non-halophilic E. coli (24–27). These abundant internalhydrophobic residues may help to stabilize the HaAP, thereby
preserving its enzymatic activity at low salt concentrations(Fig.5-22(b)). Therefore, protein absorbents modeled on the
HaAP structure might be functional in a wide range of saline
environments, from freshwater regions to salt lakes.
Moreover, using the designed protein absorbents, rare andharmful metals can be collected by unique methods. For
example, if the protein absorbents are expressed in specificparts of plants, such as their seeds, we can envision an
efficient, cost-economical means of sequestering rare andharmful metals from different environments.
The present study was accomplished in collaboration with
Kagoshima University.

Reference
Arai, S. et al., Structural Characteristics of Alkaline Phosphatase from the Moderately Halophilic Bacterium Halomonas sp. 593, Acta Crystallographica
Section D, vol.70, no.3, 2014, p.811-820.
JAEA R&D Review 201471

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5-8 Observation of Living Cells under a Soft X-ray Microscope− Laser Plasma Soft X-ray Microscope with a Combination of Intense Laser Plasma
Soft X-ray Source and Contact Microscopy −(a)silicon nitride window(b)

X-ray photoresist

laser plasmahigh power laser

soft X-ray

cultivation medium

cells cultivated directly
onto the photoresist

10 μm

Fig.5-23 Conceptual design of the laser plasma soft X-ray
microscope
Soft X-rays generated from the laser plasma were irradiatedonto cells directly cultivated on an X-ray photoresist, andthe soft X-ray images of the cells were recorded on thephotoresist.

(c)

10 μm

Fig.5-24 (a) Soft X-ray image and (b) fluorescent imageof living cells
The red, blue, and green regions in the fluorescent image
are the mitochondria, nuclei, and cytoskeleton, respectively.

(d)

(e)
mitochondrion

nucleusmitochondria

mitochondria
5 μm

1 μm

100 nm

Fig.5-25 Soft X-ray images of the inner structures of living cells at different scales
Nuclei and mitochondria are clearly recognized.

We have developed a laser plasma soft X-ray microscopethat combines a laser plasma soft X-ray source with contact
X-ray microscopy. The laser plasma soft X-ray source offers
high brightness and short pulse duration, whereas contact
X-ray microscopy allows observation of cells cultivated in situ.
By combining these two techniques, we can observe living
cells at a high spatial resolution. Intense soft X-rays weregenerated by irradiating a thin-foiled gold target with anintense laser beam. Fig.5-23 shows the conceptual design ofthe laser plasma soft X-ray microscope. Soft X-rays generated
from the laser plasma were irradiated onto cells directly
cultivated on an X-ray photoresist, and the soft X-ray imagesof the cells were recorded on the photoresist. The cells werethen labeled with several fluorescent dyes and observed undera fluorescence microscope. The resulting fluorescent images
were directly compared with the soft X-ray images.
Fig.5-24 shows the soft X-ray image (a) and the fluorescentimage (b) of the living cells. The red, blue, and green regionsin the fluorescent image are the mitochondria, nuclei, andcytoskeleton, respectively. Comparing the fluorescent images,
we find that all organelles are clearly identified in the soft
X-ray image.
Soft X-ray images of the inner structures of living cells
are presented in Fig.5-25. Panel (c) clearly shows the
mitochondria surrounding the nucleus. The mesh-likestructures of the mitochondria are visible in panel (d), whilepanel (e) reveals a single mitochondrion, the first image ofthis organelle in living cells. The laser plasma soft X-ray
microscope is expected to contribute further to life sciencestudies.
The present study was partly supported by Japan Society for
the Promotion of Science (JSPS) KAKENHI Grant-in-Aid for
Scientific Research (C) (No.25390134).

Reference
Kado, M. et al., In Situ Observation of Cellular Organelles with a Contact X-ray Microscope, Journal of Physics: Conference Series, vol.463, 2013,
p.012056-1-012056-4.

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5-9 Inhibition of Cadmium Movement in Crops by a Peptide− Visualization of Elements in Roots Using a Positron-Emitting Tracer Imaging System −Without
GSH
treatment treatment

Cd in hydroponic solution (cps)

36 h

107

24 h

Without treatment
GSH treatment

4h

under
ground

12 h

2h

above
ground20 min

600
500
400
300
200

Cd exclusion from the root
100
0
0

6

12

18

24

30

36

Time (h)Fig.5-26 Root-zone imaging of Cd in oilseed rape plants
The plants were inserted into an acrylic vessel (left) andradioactive Cd was fed to the root. Subsequently, we could
view the uptake of Cd from the hydroponic solution by the root
and the movement to the aboveground part of the plant (right).
The bright regions in the right panel show the Cd distribution.

Fig.5-27 Time course of 107Cd radioactivity in solution
In the untreated plants, Cd was continuously taken up and the107
Cd content in solution correspondingly decreased. On theother hand, in the plant treated with GSH, the 107Cd content
in solution increased 12 h after 107Cd feeding. This result
indicates that Cd was excluded into solution from the root.

Reducing cadmium (Cd) accumulation in crops is a globally
important food safety requirement. To reach this goal, studies
have proposed numerous approaches such as cleansing of Cdcontaminated soil by crops that readily accumulate the metal,
and exploring the crop cultivation conditions that minimize
Cd uptake. In promoting these studies, it is very important toobserve directly the process of Cd uptake from the roots. We
have been developing a non-invasive imaging method called
the positron-emitting tracer imaging system (PETIS) thatcan visualize the movement of various radioactive elements
in intact plants. However, PETIS observations have been
limited to the aboveground parts of the plant, because it wasconsidered that, if the roots were immersed in a hydroponicsolution containing a radiotracer, the visualization of theelement of interest through the roots would be hindered by theintense radiation.
In this study, we developed a specialized vessel thatminimizes the volume of the hydroponic solution in
the field of view. The aim was to improve the counting
efficiency of radiation signals from the roots. Within thevessel, the movement of the elements of interest into andthrough the roots can be directly observed. Using this newly
developed “root-zone imaging,” we observed the movementof a radioisotope of Cd ( 107Cd) in oilseed rape plants.
The movement of the radioisotope from the roots to theaboveground parts was suppressed by feeding a tripeptide,
glutathione (GSH), to the roots (Fig.5-26). Moreover, adetailed analysis of the dynamic image revealed that Cd waspartially excluded from the root to the solution after feedingwith GSH (Fig.5-27). From these results, we deduced that
GSH activates the exclusion of Cd from the roots, suppressing
its movement to the aboveground parts.
This study suggests that the application of GSH is very
promising in the development of a technique for reducing Cd
in crops. Furthermore, as the root-zone imaging is applicableto other elements, it will facilitate the understanding of generalmechanisms by which plants take up various nutrients andenvironmental pollutants.
The present study was partly supported by Japan Society for
the Promotion of Science (JSPS) KAKENHI Grant-in-Aid for
Scientific Research (B) (Nos.17380194, 19380185, 23380155,
23380194).

Reference
Nakamura, S., Suzui, N. et al., Application of Glutathione to Roots Selectively Inhibits Cadmium Transport from Roots to Shoots in Oilseed Rape, Journal of
Experimental Botany, vol.64, issue 4, 2013, p.1073-1081.
JAEA R&D Review 201473

Quantum Beam Science Research

5-10 Challenge for Nanoscale Analysis by X-ray Spectroscopy in Combination
with Electron Microscopy− Development of a Multilayer Grating Spectrometer for Use with Electron Microscopes −(a)3 nm groove depth(c)Diffraction efficiency (%)(b)
twice thicker thanother B4C layers

W/B4C multilayer

B4C

Gold single layer(50 nm thickenss)W

5.6 nmGrating(2400 grooves/mm)

10
W/B4C multilayer grating
1

10 times
200 times

0.1
Gold-coated grating
0.01

Grating

2000

2500

3000
3500
Energy (eV)4000

Fig.5-28 Schematics of (a) a gold-coated grating and (b) an aperiodic W/B4C multilayer grating. (c) The measured
diffraction efficiencies of both gratings are plotted as functions of photon energy at a constant angle of incidence
The W/B4C multilayer grating shows uniformly high diffraction efficiency across the measured energy range.

(d)

(e)
MG spectrometer

Electron beam

In-Lα

1

Intensity (a.u.)

Detector

Source point(Specimen)
Multilayer grating (MG)
X-raysMG spectrometer
Conventional X-ray
spectrometer (EDS)

ITO

TEM

In-Lβ1In-Lβ4In-Lβ3In-Lβ2,15
0
3000

Sn-Lα
3200

3400

3600

(Terauchi, M. et al., JEM,
vol.62, issue 3, 2013,
p.391-395.)

Sn-Lβ1
3800

4000

Energy (eV)Fig.5-29 MG spectrometer installed in (d) a TEM and (e) emission spectra from ITO
The MG spectrometer better resolves the fine structure of the spectra, such as the Sn-Lα and In-Lβ1 peaks, thanconventional EDS.

Electron microscopes (EMs) are useful tools for the
structural analysis of materials at nanometer scales. X-rays are
generated when a material is irradiated by an electron beam.
If the X-ray intensity distribution is plotted as a functionof the photon energy by using an X-ray spectrometer witha diffraction grating featuring many narrow grooves on its
surface, we can analyze the valence electronic structure
underlying the material properties. As conventional soft X-ray
gratings are coated with a thin gold film, their use becomes
impractical at energies around the gold absorption edges(approximately 2.2 keV). To overcome this problem, we
designed an aperiodic W/B4C multilayer structure and applied
it to a wideband multilayer grating covering 2.0–4.0 keV at aconstant angle of incidence.
Fig.5-28 presents schematics of an Au-coated grating (AG)(a) and the designed aperiodic W/B4C multilayer grating(MG) (b). The measured diffraction efficiency curves of both
gratings are plotted in panel (c). The diffraction efficienciesof the MG at 2.1 keV and 4.0 keV are 10 and 200 times
higher than those of the AG, respectively. The MG also shows

uniformly high efficiency over the entire energy range. This
result is attributable to the reflective characteristic of the B4C
layer just below the topmost W layer, whose thickness is twicethat of the other layers in the structure, as shown in Fig.5-28(b).
Fig.5-29 is a photograph of the MG spectrometer installed
in a transmission EM. Also shown are the emission spectrafrom indium-tin-oxide measured by the MG spectrometer anda conventional energy-dispersive spectrometer (EDS). The
L emission spectra of Sn and In (e.g., Sn-Lα and In-Lβ1) are
more clearly resolved by the high-resolution MG spectrometer
than by EDS.
By combining X-ray spectroscopy with EM, we haveestablished a unique technique for extracting morphologicaland electronic structure information. This technique is
expected to assist the development of electronic devices andfunctional materials.
The present study was partly supported by the project ofCollaborative Development of Innovative SEEDs from theJapan Science and Technology Agency (JST).

Reference
Imazono, T., Development of a Soft X-ray Flat-Field Spectrograph in the 50-4000 eV Range and Its Application to Electron Microscopes, Oyo Butsuri,
vol.83, no.4, 2014, p.288-292 (in Japanese).

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5-11 Discovery of X-ray Coherent Mirage− Indication of the Potential of New X-ray Optics Using Plasma −(a)Plasma X-ray laser(wavelength 13.9 nm)(b)Plasma lens(concave)

Imaginary
source of X-ray(mirage)Real sourceof X-ray

X

(c)

Z

Experiment

Detector
X-ray laser amplifier(plasma)
X-ray laser oscillator

Calculation

Y
(d)

−33

(e)
Z, mm

−27

60
40
20

X,

−60

0

−21
−15
−9

−40

μm

−20

−20

0

−40
−60

20
40
60

m
Y, μ

−3
3
−50 −40 −30 −20 −10

0

10

20

30

40

50

Y, μm

Fig.5-30(a) Schematic of the formation of an X-ray mirage. The plasma acts like a convex lens, refracting a portion of the X-ray laser
beam. An imaginary source, displaced from the position of the original source, is formed.
(b) Interference fringes of the X-ray laser obtained at the detector position.
(c) Comparison of the experimental fringes with the calculated result.
(d), (e) We reproduced the imaginary source by theoretical modeling based on the experimental result.
(d) Spatial profile of the imaginary source in the X–Y plane.
(e) Spatial profile in the Y–Z plane.

Mirages are well-known phenomena aroused in the Earth’s
atmosphere. They result from the uncontrolled deflection oflight rays from rectilinear propagation in inhomogeneous
media. Mirages in the X-ray range have been consideredextremely difficult to form because the refractive indices of
conventional materials to X-rays are close to unity, which
essentially excludes refraction.
JAEA has developed plasma X-ray lasers with excellent
narrow beam divergence. We injected these X-ray laser beams
into a plasma that amplifies X-rays of the same wavelength(Fig.5-30(a)), and obtained a coaxial interference pattern(Fig.5-30(b)). Coaxial interference patterns are generated if
two coherent X-ray sources are positioned on the same straightline from the observation point. To explain the observed
phenomena, we considered that the plasma behaves as aninhomogeneous atmosphere, i.e., a portion of the X-ray laser
beam is refracted and expanded by a concave lens effect in theplasma, forming an imaginary source that appears to exist at
the lens focal point. Indeed, if the imaginary source is assumed

to lie inside the plasma, the calculated and experimental fringe
patterns are completely consistent (Fig.5-30(c)). Considering
its similarity to ordinary mirage formation, we conclude that
we have generated the first example of an X-ray mirage.
To visualize the imaginary source (the X-ray mirage) on
the basis of the experimental result, we have developed atheoretical calculation code, accounting for X-ray refraction
and amplification in the plasma. The simulated spatial profilesof the imaginary source are presented in Figs.5-30(d) and(e). The formation of an X-ray mirage was found to strongly
depend on the amplification effects of the plasma.
The X-ray mirage discovered in the present study may be
adopted in novel plasma-based X-ray optics and methods
for diagnosing plasmas, providing useful information of thedensity profiles, amplification effects, and other properties ofplasmas.
The present study was partly supported by Japan Society for
the Promotion of Science (JSPS) KAKENHI Grant-in-Aid for
Scientific Research (B) (No.25289244).

Reference
Magnitskiy, S., Pikuz, T. et al., Observation and Theory of X-ray Mirages, Nature Communications, vol.4, 2013, p.1936-1-1936-7.

JAEA R&D Review 201475

Quantum Beam Science Research

5-12 Transparency of Sodium in the Ultra-Violet Spectral Range− Imaging by Light Illumination through an 8-mm-Thick Sodium Sample −MgF2 lensSodium sample
Thickness=8 mm

Phosphor+ CCD camera

MgF2 lens
MgF2 lensMetal mesh

Deuterium lamp
Ultra-violet light

Sodium sample

Mesh

MgF2 windows
5 cm

Fig.5-31 Set-up of the imaging experiment
The ultra violet light illuminates the sodium sample throughoptical elements to form an image on the camera.

Ultra-violet lightFig.5-32 Photograph of an 8-mm-thick sodium sample
and the imaging system
At the center of the photo, the fresh sodium sample is visible
with metallic reflection.
1.0

mm

Glass plate

Fig.5-33 Image of a metal mesh composed of 100-μm-thick wires spacedby 1 mm, illuminated by ultra-violet light that shone through an 8-mm-thicksodium sample
The blackened area corresponds to the area covered by a 1-mm-thick glass
plate, which is opaque to this range of wavelengths.

The surface of fresh sodium reflects visible light. As iswell known, surface electrons oscillate when exposed toincident light and emit light at the same wavelength. At
shorter wavelengths of incident light, the electrons cannot
follow the electric field of the light, and the light penetrates
the sodium interior. However, the light is significantly dumpedby phenomena such as electron-phonon scattering. To date,transparent sodium has not been experimentally investigated.
On the other hand, the optical properties of alkali metals(including sodium) have been intensively studied since thepioneering work of Wood in the1930s. In the1960s, the Oak
Ridge National Laboratory team obtained a plasma wavelengthof 218 nm. At wavelengths below 218 nm, sodium is partiallytransparent down to some threshold; namely, the core electron
excitation wavelength (40.5 nm). The Oak Ridge team tried
to measure the transmittance of a sub-μm-thick solid sodium
layer coated onto a quartz and lithium fluoride (LiF) substrate.
However, the thickness restricts the dynamic range of thetransmittance. Therefore, we require a considerably thicker
sodium region to clarify the transmittance.

In this study, we fabricate relatively thick sodium samples(>1 mm) in the spectral range beyond 115 nm, whichcorresponds to the shortest transmission wavelength ofmagnesium fluoride (MgF2) windows, and evaluate theirtransmittance. First, we determined the spectral transmittance
between 115 nm and ~200 nm. At ~120 nm, the transmittanceof 3-mm-thick solid sodium samples admitting the MgF2
windows was several tens of percent. Furthermore, thetransmittance weakly depended on temperature up to 150 °C(note that the melting point of solid sodium is 97 °C).
To confirm the transmittance, we also performed a simple
imaging experiment with an 8-mm-thick sodium sample.
Figs.5-31 and 5-32 show a schematic of the experiment and aphotograph of the setup, respectively. We recorded a clear imageof a metal mesh with 1 mm separation onto a two-dimensionalcharge coupled device detector (Fig.5-33). This result
encourages us to design an optical imaging device for viewing
objects inside or through solid or liquid sodium media, and to
determine the physical origin of high transmittance, which has
not previously been demonstrated.

Reference
Daido, H. et al., Demonstration of Partially Transparent Thick Metallic Sodium in the Vacuum Ultraviolet Spectral Range, Optics Express, vol.21, issue 23,
2013, p.28182-28188.

76

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5-13 Innovative Technique for Elucidating Diffusion through Materials− Development of an Advanced Analysis Method for Quasielastic Neutron Scattering −Relaxation time (ps)
100
6

10

1

0.1

0.01

3.0

0

10

−2
1

−4
−6
0 0.5

1.0

2.0

3.0

4.0

2.5

2.0

2.0
1.5

1.5

Slow

2.5

1.0

Intensity of MDF

2Momentum transfer (Å−1)

100Scattering intensity

Energy transfer (meV)

3.0

4

Fast

0.5

1.0
0.0

0.5

Intermediate−1

Momentum transfer (Å )
0.001

0.01

0.1

1

10

100

Width (meV)Fig.5-34 Quasielastic neutron scattering of water
Quasielastic scattering spectra obtained by a neutron
scattering experiment. The signal is spread like a moundaround the elastic scattering of zero energy transfer. Because
more than one diffusion mode exists, the routes, speeds,
and types of the modes are difficult to identify because theirsignals overlap. The red dot indicates the signal generated by
the sample cell.

Fig.5-35 Calculated mode distribution function of water
Using the newly developed analysis, we computed the mode
distribution function (MDF) of water. Each diffusion mode isindependently represented. Consequently, we can directly
evaluate the number of modes, and their types and speeds.

Diffusion is an important process in materials and living
tissue. For instance, it is involved in the functional expressionof proteins, and plays a role in the capabilities of ion andproton conductors, which are now being investigated as battery
materials. Therefore, by elucidating diffusion mechanisms,
we can control the functional expressions and capabilities
of materials. Quasielastic neutron scattering (QENS) is
among the few powerful techniques that can investigate therelationships between time and space in the diffusion motionsof atoms and molecules. Therefore, QENS allows an in-depth
study of diffusion. However, the correctness of diffusion
models is difficult to demonstrate by conventional QENS
analysis, as the model must be assumed prior to analysis, andtherefore cannot naturally emerge from the results.
We have developed a versatile analysis method that requires
no model assumptions. By this method, we can understand themodel from the obtained results. We represent all diffusion
processes of a material as a set of “simple diffusions.” Ina “simple diffusion,” on average, the particles gradually
spread out with a constant relaxation time. In this scenario,
the diffusion process can be converted into the intensity

distribution of the relaxation time. We call this distribution
the mode distribution function (MDF). The MDF directly
provides the number of diffusion motions, and their types andspeeds. The conversion to MDF adopts the maximum entropy
method based on information theory.
We applied this new method to a diffusion study of water.
QENS experiments were performed on AMATERAS at
J-PARC. AMATERAS can measure over a wide area athigh resolution and is ideally suited for this method. A
QENS spectrum of liquid water is shown in Fig.5-34. The
characteristics of the diffusion are not directly interpretablefrom this figure. However, the MDF obtained from the QENS
spectrum (Fig.5-35) clearly shows three types of motions. In
particular, an unsuspected intermediate motion is discovered.
Therefore, the MDF reveals the detailed characters ofdiffusion motions. This result is important for understanding
the overall picture of diffusion in water.
We hope that such diffusion studies will advance a wide
range of research fields and uncover new knowledge that will
benefit science and technology.

Reference
Kikuchi, T. et al., Mode-Distribution Analysis of Quasielastic Neutron Scattering and Application to Liquid Water, Physical Review E, vol.87, no.6, 2013,
p.062314-1-062314-8.
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Quantum Beam Science Research

5-14 Achievement of High-Quality Pulsed Cold Neutron Beam
− Performance Evaluation of the J-PARC Cryogenic Hydrogen System −Helium blanket chamber
Pump

Accumulator

Orifice typeheaterHelium refrigerator
Ortho-parahydrogenconvertor

Expansion turbine
Compressor

Fig.5-36 Cryogenic hydrogen system for the J-PARC
spallation neutron source
Supercritical hydrogen at 20 K and 1.5 MPa is supplied tothe moderators, which convert the generated fast neutrons
to cold neutrons.

Hydrogen loop
HeaterModerators
Helium flow
Hydrogen flow

Cold evaporator

Allowable value(100 kPa)

100
Theoretical value

80

Measured value

60
40
20

Without accumulator
Pressure rise (kPa)Para-hydrogenconcentration (%)

100

80
60
40
Predicted value

20

Measured value
0

0

50

100

150

200

250

300

Hydrogen temperature (K)

0

0

200

400

600

800

1000

Proton beam power (kW)Fig.5-37 Measured para-hydrogen concentration during acool-down process
The measured para-hydrogen concentration, which agrees
with the theoretical values, exceeds 99% at 20 K.

Fig.5-38 Effect of proton beam power on pressure rise in
the hydrogen loop
At each power of the proton beam, the developed system
reduces the pressure rise to below the maximum allowed value.

At J-PARC, 3 GeV protons with a power of 1 MW are
injected onto a mercury target at a repetition rate of 25 Hz,
producing fast neutrons via the spallation reaction. The highenergy neutrons are slowed to thermal and/or cold neutrons inhydrogen moderators, which receive supercritical hydrogenat 1.5 MPa and below 20 K. We aimed to maintain the parahydrogen concentration in the hydrogen moderator above
99%. Cold neutrons are effectively slowed by the excitationof para-hydrogen to ortho-hydrogen. Moreover, the neutron
scattering cross section of para-hydrogen is lower than thatof ortho-hydrogen by two orders of magnitude. Therefore,
we obtained a pulsed cold neutron profile with a narrow fullwidth-half-maximum (approximately 100 μs) and a short tail.
These pulsed neutrons are suitable for crystal and magnetic
structural analyses.
The presence of ortho-hydrogen and large changes inhydrogen density degrade the pulse shape. Therefore, wedeveloped a cryogenic hydrogen system in which supercriticalpara-hydrogen circulates at the largest flow rate thus far
achieved (9.5 m3/h). This large flow rate reduces the fluctuationsin the moderator temperature to below 3 K (Fig.5-36).
We installed an ortho-para hydrogen convertor with acatalyst of hydrous ferric oxide. Furthermore, we developed
a hydrogen gas sampling system, which safely fills a glass

cell with hydrogen gas obtained from a high-pressure, lowtemperature environment. The para-hydrogen concentration
was measured during the cool-down process by a laser
Raman spectrometric method. The measured and theoreticalpara-hydrogen concentrations closely match and satisfy therequirements at 20 K (Fig.5-37).
One major challenge is to mitigate the pressure fluctuations
below 0.1 MPa. These fluctuations, caused by large stepwise
heat loads when the beam is on, are problematic because
supercritical hydrogen behaves as an incompressible fluid
and lacks a vapor-liquid interface. To overcome this problem,
we developed a novel equipment with a bellows structure,
called an “accumulator.” Pressure fluctuations are mitigatedby spontaneous expansion or contraction of the bellows,which enclose helium gas (boiling point = 4.2 K). At beam
powers below 532 kW, the measured pressure rises agree withthe predicted values. We expect that the pressure rise can be
maintained below the upper allowable limit at a beam powerof 1 MW (Fig.5-38).
This technology will enable the steady production of highquality cold neutron beams at the rated 1 MW operation.
Potentially, it will advance neutron beam research, thereby
contributing to science, technological developments, andindustrial promotion.

Reference
Tatsumoto, H. et al., Operational Characteristics of the J-PARC Cryogenic Hydrogen System for a Spallation Neutron Source, AIP Conference Proceedings
1573, 2014, p.66-73.

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5-15 Dose Rate after Beam Operation of J-PARC 3 GeV Synchrotron− Investigation and Measurement of Dose Rate Distribution of Radioactive Nuclides Producedin the Accelerator for Higher Output Power −Arc section[beam loss point (1)]: 0.055 mSv/h
Downstream of the injection point [beam loss point (2)]:0.97 mSv/h (Before measurement: 6.2 mSv/h)
Arc section
Downstream of the injection point [beam loss point (1)]:0.012 mSv/h[beam loss point (2)]:
1 mSv/h (Before measurement: 4.5 mSv/h)
Injection branch [beam loss point (3)]:
1 mSv/h (Before measurement: 1.8 mSv/h)

Plastic scintillation counter signals (a.u.)

Arc section [beam loss point (1)]: 0.08 mSv/h

1500

1200

900

600

0

10

15

20

25

Average number of foil hits (count)

Arc section [beam loss point (1)]: 0.01 mSv/h

Arc section [beam loss point (1)]: 0.025 mSv/h

Fig.5-39 Residual dose distribution of J-PARC 3 GeV synchrotronThis figure shows the dose rates at the surfaces of the magnets
and vacuum chambers after accelerator operation. These data
were acquired after continuous operation at 200 kW for one month,and 300 kW for 3 days. Data were acquired on July 2012 except at
point (3), which was measured on March 2014 after reducing thebeam loss at that point.

The 3 GeV rapid cycling synchrotron (RCS) at J-PARC has
been operational since 2007. The RCS ring generates a highpower proton beam (1 MW) at a repetition rate of 25 Hz,
which is supplied to the main ring and the neutron production
target. In the RCS, negative hydrogen ions are accelerated
to 400 MeV by a linac and passed through a thin carbon
foil, which converts them into protons by removing the twoelectrons. The protons are then injected into the synchrotron
and accelerated to 3 GeV in 20 ms.
In a high-intensity accelerator such as J-PARC, minimizing
the beam loss is vital to retaining the activation of the accelerator
components within an allowable level. Based on experience
with the RCS and other accelerators, provided that the dose rate
immediately after beam stop is below 1 mSv/h at the accelerator
component surfaces, the exposure dose during maintenancework can be suppressed to less than several hundred μSv.
Therefore, to ensure that the dose rate remains below 1 mSv/h,
we must understand the distribution of the dose rate generated
by the beam loss. To this end, we measured the dose rate on
the equipment surfaces during each maintenance period, andelucidated the dose rate conditions produced by the various

Fig.5-40 Beam loss as a function of average numberof foil hits
This graph shows the output signals of the plastic
scintillation counter as a function of the average numberof times that the circulating beam hits the carbon foil. The
plastic scintillation counter was positioned downstream
from the carbon foil. The average number of foil hits wascontrolled by varying the injection orbit or the carbon foil
position. The radiation generated by interaction between
the carbon foil and the circulating beam increased with theaverage number of foil hits, thereby increasing the signal
intensity.

operation modes. Additionally, we considered the cause of theactivation and appropriate countermeasures.
Fig.5-39 shows the measured dose rate around the RCS.
Most of the radioactivity generated by the accelerator
components is concentrated at the injection section (the areaof the straight line on the left side of Fig.5-40). Away fromthe injection section, the dose rate remains below 0.1 mSv/h[beam loss point (1)]. In the injection section, peaks of local
activation appear downstream of the injection point [beamloss point (2)] and at the branch between the injection andthe circulating beam [see beam loss point (3), Fig.5-39]. The
results indicate that the activation at point 2 is sourced frominteraction between the beam and the thin carbon film (seebeam-study results in Fig.5-40), while the beam loss at point 3
is caused by increased pressure of the injection beam transport
line. After decreasing the beam loss by inserting anothersource, we increased the beam power while retaining the dose
rate below 1 mSv/h at both loss points.
In summary, we increased the beam power by reducing thebeam loss.

Reference
Yamamoto, K. et al., Beam Power and Residual Dose History of J-PARC RCS, Progress in Nuclear Science and Technology, vol.4, 2014, p.238-242.

JAEA R&D Review 201479

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5-16 Magnetic Shielding via Vacuum Chambers for Beam Orbit Stabilization− Development of Vacuum Chambers of Magnetic Materials at J-PARC −Fig.5-41 Location of the vacuum chambersof magnetic materials in the 3 GeV
synchrotron at J-PARC
At the extraction section in the synchrotron,
stray magnetic fields from a nearby
magnet leak into the synchrotron beamline, distorting the beam orbit. The vacuum
chambers of magnetic materials weredeveloped to shield the beam from theseunwanted magnetic fields.

Unwanted magnetic field from a nearby magnet (~10−3 T)
Bellows

Bellows

Beam pipes

Beam pipes

Beam pipes

1m

Beam pipes

Beam
pipe

Beam pipe(a)Bellows

Bellows

Outer magnetic field 2.0 × 10−3 T

4.5e−3Magnetic material:
Pipe parts, bellows parts, flange parts
Pipe parts, bellows parts
Pipe parts

1.5e−32e−4

3e−3

Beam pipeof magnetic materials

2e−3

Outer magnetic field 4.0 × 10−3 T

1e−30−600 −400

−200

0

200

400

600

800

Position (mm)

(b)

1e−4

0

0

1000

2000

3000

Position (mm)Pressure (Pa)

Magnetic field inside
the vacuum chambers (T)Beam axis

Beam pipe

Magnetic field inside
the vacuum chamber (T)Turbo molecular pump
Quadrupole magnet
Turbo molecular pump
Modeled region to calculate the magnetic shielding effect (Fig.5-42)

10−5

Ultra-high vacuum region (under 10−5 Pa)

10−6
10−7
10−8
0

10

20

30

40

Fig.5-42 Calculated magnetic field inside the vacuum
chambers, assuming that each component of the vacuum
chambers is constructed from magnetic materials
The area surrounded by the green rectangle in Fig.5-41 wasmodeled in the calculation. The most effective magnetic
shielding was achieved when all components were constructedfrom magnetic materials.

Fig.5-43 Measured magnetic shielding performance andvacuum performance of the vacuum chamber(a) The vacuum chamber of magnetic materials effectively
shielded the external magnetic field.
(b) The vacuum chamber achieved ultra-high vacuum conditions.

A main source of beam loss in synchrotrons is beam orbit
distortion, caused by the entry of unwanted stray magnetic
fields from the magnets at a nearby beam line. In the J-PARC
3 GeV synchrotron, a magnet along the beam extraction
line injected a ~10−3 T stray magnetic field into the vacuum
chambers, distorting the beam orbit by approximately 10 mmfrom the chamber center. Stray magnetic fields are mosteffectively shielded by completely surrounding the beam
region with magnetic material. Therefore, we decided to
manufacture the vacuum chambers of magnetic materials. The
developed vacuum chambers demonstrate superior magnetic
shielding and vacuum performances.
Fig.5-41 shows the layout of the extraction section of
the synchrotron, where the vacuum chambers of magnetic
materials are installed. According to the beam orbit
calculation, the external magnetic field should be suppressed
to less than one-tenth in this beam region. Fig.5-42 shows thecalculated magnetic field, assuming that each component of
the vacuum chambers is constructed from magnetic materials.

To ensure effective shielding, all the components, including
the bellows, flange, and pipe parts, should be made frommagnetic materials. To this end, permalloy and ferritic
stainless steel were selected for the thin (pipes and bellows)
and thick (flanges) parts, respectively.
Machining generates internal stress, which generally reduces
the magnetic permeability of a material. However, magnetic
permeability can be recovered by annealing. On the other
hand, to reduce outgassing, vacuum materials are baked athigh temperature under high-vacuum conditions. To achievemagnetic annealing while reducing the outgassing, we heattreated the vacuum chambers under high vacuum. Fig.5-43
shows the magnetic shielding and vacuum performances.
The external magnetic field was suppressed to less than onetenth inside the chamber, and the magnetic materials ensured
ultra-high vacuum. By constructing the vacuum chambers ofmagnetic materials, we can realize superior magnetic shielding
and vacuum performances and a stable beam orbit within thesynchrotron.

Time (h)Reference
Kamiya, J. et al., Vacuum Chamber Made of Soft Magnetic Material with High Permeability, Vacuum, vol.98, 2013, p.12-17.

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5-17 Observation of the Beam Injected into the Acceleration Region of the Cyclotron− Development of an Emittance and Acceptance Measurement System −Beam intensity monitor 2(a)Phase space collimator

Ion source
SolenoidmagnetSteering
magnetPosition
slit

Divergence
slit

Beam
intensity
monitor 1

Cyclotron
(b)

angle

angle

positionangle

position

position

Emittance① Effectively expanding
emittance② Dividing into
small regions③ Detecting accelerated
regionsFig.5-44 (a) Emittance and acceptance measurement system and(b) acceptance measurement procedure
The system is composed of electromagnets that effectively expand
beam emittance, a phase space collimator that limits the position andangle range of the beam, and two beam intensity monitors.

20

Angle (mrad)

15

emittance

105
0

Fig.5-45 Results of emittance and acceptance measurements−5(modified from Kashiwagi, H. et al., Rev. Sci. Instrum., vol.85, no.2, 2014,
p.02A735-1-02A735-5.)

−10
−15

acceptance−20−50 −40 −30 −20 −10

0

10

20

30

40

50

The relationship between the acceleration region and injected beam
region is visualized by displaying the measured emittance andacceptance in one graph.

Position (mm)

The TIARA cyclotron facility produces various types
of ion beams for different researches (such as bioscienceand material sciences) by changing the ion species in eachexperiment. Beam injection tuning that maximizes the ratioof the accelerated beam current to the beam current injected
into the cyclotron is required to ensure sufficient current foreach experiment. The tuning corresponds to the overlappingof emittance (a region occupied by the injected beam) andacceptance (an acceleration region) in phase space. Because
any emittance regions outside the acceptance are notaccelerated, the emittance must be matched to the acceptance.
Although various methods are available for measuring theemittance, no method has been developed for measuring
the acceptance of a cyclotron. To fill this gap, we deviseda method for measuring the acceptance, and developed atransverse emittance and acceptance measurement system forvisualizing the emittance–acceptance relationship.
Figs.5-44(a) and (b) are schematics of the measurement
system and procedure, respectively. The acceptance isdetermined by the following three steps: (1) The beamemittance from the ion source is effectively expanded bysolenoid magnets and a steering magnet to ensure sufficientmeasurement range (Fig.5-44(b)①). (2) The expanded
emittance is divided into many small regions by two slits calledphase space collimator (Fig.5-44(b)②), and the beamlets withthe regions are sequentially injected into the cyclotron. (3) The
acceptance is obtained by aggregating the accelerated beamlets
detected by beam intensity monitor 2 (Fig.5-44(b)③). The
emittance is obtained by measuring the beam intensity ofeach region divided by the phase space collimator using beam
intensity monitor 1.
Fig.5-45 shows the measured emittance and acceptanceon a single plot. We succeeded to visualize the relationshipbetween the injected beam emittance and the acceptance of thecyclotron using the system.
On the basis of the results of the emittance and acceptancemeasurements, we will develop a beam injection tuning
system that overlaps the emittance and acceptance tomaximize the ratio of the accelerated beam.

Reference
Kashiwagi, H. et al., A Transverse Emittance and Acceptance Measurement System in a Low-Energy Beam Transport Line, Review of Scientific Instruments,
vol.85, no.2, 2014, p.02A735-1-02A735-5.
JAEA R&D Review 201481

６ Nuclear Hydrogen and Heat Application Research
Research and Development of High Temperature Gas-Cooled Reactor, Hydrogen Production,and Heat Application Technologies
Major features

HTGR Commercial Plant

Low-temperature
heat application• Excellent inherent safety• Excellent economy even smallsized reactor• Multi-purpose heat application• Seawater desalination• District heating• Agriculture, fishery

High Temperature engineering
Test Reactor (HTTR): Oarai,Ibaraki

Particle fuel coatedwith ceramics

1998
2001
2004
2010

Criticality
Full power operation
950 ˚C operation
950 ˚C operationfor 50 days

UO2

• Hydrogen production• Steam supply
Electric generation • High efficiency
power generation
by gas turbine

World’s most advanced homegrown technology• Ceramics coating (TRISO) for Uranium fuelTemperature limit: 1600 ℃

Diameter ~1 mm
Isotropic graphite

High strength, high heat conduction,
radiation resistance• Heat-resistant, oxidation-resistant Ni-base superalloy• Heat insulation, anti-thermal expansion, anti-vibration
structure
Service temperature: ~950 ℃

Thermal power: 30 MW
Outlet temperature : 950 °C
Coolant pressure : 4 MPa

Long term (~3 times of LWR),
complete retainment of fission
products (100~1000 times of HTGRs in
USA and Germany)• 3-dimensional isotropic, radiation-resistant graphite
Temperature limit: 2400 ℃

Heat-resistant alloy

High-temperature
heat application

High-temperature (950 ℃) heatcan be extracted from the reactor.

• High-temperature, high-pressure He handling
Service temperature: ~950 ℃

Chemically, nuclear, and mechanicallystable coolingFig.6-1 Outline of HTGR; Features, Heat utilization, Major specifications, and HTTR technologies
HTGRs are helium gas-cooled, graphite-moderated thermal neutron reactors capable of meeting various heat application requirements,
such as hydrogen production and power generation. In particular, they are strongly expected to regain the public’s trust in nuclear
power because of their inherent safety.

To reduce greenhouse gas emissions and remedy the fragilityof Japanʼs energy supply-demand structure arising fromexcessive dependence on fossil fuel resources from abroad,
nuclear energy should be used not only for power generation
but also various heat applications. The Strategic Energy Plan,
Large-boned Policy, and Japan Revitalization Strategy, decided
by the cabinet in 2014, indicated the promotion of research anddevelopment of a high temperature gas-cooled reactor (HTGR) as
a national policy because of the reactor’s excellent inherent safety
and ability to meet energy demands from various industries.
An HTGR can supply temperatures of 950 °C using inert
helium (He) gas as a coolant. Three other cutting-edge
Japanese technologies that make it feasible to obtain such hightemperatures in an HTGR have been developed by JAEA. The
first is a technology for fabricating ceramic-coated fuel particles
with a diameter of approximately 1 mm. Ceramics remain stableeven at 2500 °C and provide superior heat-resistant coating
layers for containing radioactive fission products within fuelparticles. The second is a technology for fabricating isotropic
and radiation-resistant graphite blocks. The third is a technology
for manufacturing heat- and oxidation-resistant super-alloys
with careful composition of alloy contents. With these major
technologies as well as further knowledge of high temperature
82

JAEA R&D Review 2014

structural design and the like, we generated a temperature of
950 °C in the high temperature engineering test reactor (HTTR)
for the first time in the world in 2004, and maintained thistemperature over 50 days of operation in 2010 (Fig.6-1).
It is feasible to prevent explosive gas generation by physicalphenomena, as well as accidental overheating and oxidation of
the fuel coating layers, without any engineered safety features.
The inherent safety features assure no harmful release ofradionuclides to general public and environment in any accident.
The heat from HTGR is useful for not only power generation
but also hydrogen production for fuel cell cars and direct
reduction iron-making, and for supplying steam to industries.
Waste heat can be utilized for desalination, among otherapplications. Research and development of the thermochemical
IS process for producing hydrogen by splitting water molecules
has been conducted. A continuous hydrogen production test isunderway to establish the applicability of this technology toindustrial plants, with the test facility consisting of components
made of existing steel and ceramic materials.
For practical applications of HTGR, we intend to perform
integration tests of HTTR and heat application systems for
power generation and hydrogen production, and research such
Topics as advanced safety (Topics 6-1, 6-2, 6-3, 6-4).

Nuclear Hydrogen and Heat Application Research

6-1 Demonstration Plan for the World’s First Nuclear Heat Application
with the High Temperature Engineering Test Reactor− Toward the Industrial Use of Nuclear Heat −Secondary helium
Tertiary helium
Pressurized water

Isolation valve

H2 plant

Reactor building
Second IHX

Offset distance

Co-axial duct
Co-axial duct
Compressor

TurbineHelium gas
turbineGenerator

Precooler

Recuperator

H2 plant

Cooler
IHX
Water cooling system
Water cooling system

Fig.6-2 Flow diagram of the candidate heat application
system coupled to the HTTR
The nuclear heat transferred to the secondary helium (He)system at the intermediate heat exchanger (IHX) is used by the
He gas turbine system and the H2 plant for power generationand hydrogen production.

Fig.6-3 Candidate heat application system layout coupled
to the HTTR
An appropriate offset distance is set between the reactor
building and H2 plant to ensure the integrity of the reactor
building (including safety-related systems, structures, andcomponents) and the habitability of the reactor control room
against leakages of combustible and hazardous chemicals.

For the utilization of nuclear heat in industrial applications
such as hydrogen production and high-efficiency power
generation, the following technical issues should be resolved.
First, a formulation of the safety requirements and a validationof the design considerations for coupling a hydrogenproduction plant (H2 plant) to a nuclear facility must becompleted. Second, the overall performance of the He gas
turbine must be confirmed. The present study identifies thetests that must be conducted using the HTTR. Additionally,
plant concepts are proposed for the heat application system tobe coupled to the HTTR.
We aim to formulate safety requirements through a licensing
review of the Nuclear Regulation Authority for coupling an
H2 plant to the HTTR. The design considerations will also bevalidated by a test simulating H2 plant accident. Furthermore,
continuous operation at the rated power, in addition to start-up
and shut-down test operations of the He gas turbine, will be
performed to demonstrate safe and reliable operation at steadyand transient conditions assumed in the actual environment forthe performance confirmation of the He gas turbine.
System configurations are investigated to minimize theamount of modification to the existing HTTR facility, as wellas to simulate the configuration of a commercial system.
A cascade configuration with a second intermediate heatexchanger (IHX), which transfers heat to the H2 plant, anda He gas turbine, which consists of a turbine, compressor,
generator, precooler, and recuperator, is selected as one of
the candidates (Fig.6-2). Additionally, the plant layout ispreliminarily determined with sufficient offset distance toensure the integrity of the reactor building (including safetyrelated systems, structures, and components) and habitabilityof the reactor control room against leakages of combustible
and hazardous chemicals (Fig.6-3).
Further study will be conducted for system design,
component design, and layout design to finalize the plant
concept. Furthermore, a safety evaluation will be performed
to expedite the first-of-a-kind demonstration of nuclear heatapplication with the HTTR.

Reference
Sato, H. et al., HTTR Demonstration Test Plan for Industrial Utilization of Nuclear Heat, JAEA-Technology 2014-031, 2014, 30p. (in Japanese).

JAEA R&D Review 201483

Nuclear Hydrogen and Heat Application Research

6-2 Compatibility between Effective Plutonium Incineration and Nuclear Proliferation Resistance− Conceptual Study of a Plutonium Burner High Temperature Gas-Cooled Reactor System −U/Pu separation

U/Pu separation

Pu

U

0.35

Solution of spent fuelPu

U

U purification

Pu purification

U purification

Pu purification

U denitration

Mixing with U

U denitration

Mixing with YSZ

U product storage

Mixed denitration

U product storage

MOX product storage

To MOX fuel fabrication process

Current reprocessing facility

To PuO2-YSZ coated fuel particle
fabrication process

Proposed reprocessing facility

Nuclear proliferation resistance

Solution of spent fuelCurrent facility
Proposed facility

U/Pu separation

0.30

To fuel fabrication
process

0.25
0.20

Pu purification

In the case of connecting with the fuel fabrication
process and the storage facility whose nuclear
proliferation resistances are 0.22

0.15
Denitration and storage
0.10
0

200

400

600

800

6775

Residence time of Pu (h)Fig.6-4 Pu flow in the reprocessing facilities
Connecting the Pu nitrate solution line directly to the fuelfabrication process allows the elimination of the mixed
denitration process and MOX product storage.

Fig.6-5 Nuclear proliferation resistance based on the
AFCI method
The nuclear proliferation resistance of the proposed
reprocessing facility is superior to that of the current systembecause it has an AFCI nuclear proliferation resistancevalue of 0.28 and a shorter residence time of Pu from theseparation process.

We have been conducting a conceptual study on using aplutonium (Pu) burner HTGR system to achieve effective
incineration of 239Pu, which is significant for nuclear nonproliferation in the long term.
In Japanʼs reprocessing facility, Pu recovered from lightwater reactor spent fuel is mixed with uranium (U), as shownin Fig.6-4, to prevent an increase in nuclear proliferation risk
by avoiding the existence of pure Pu. The Pu/U mixed solution
line is connected to the denitration process, and then Pu/U
mixed dioxide fuel (MOX fuel) is fabricated.
When used in a reactor, the MOX fuel generates additional
Pu through a neutron capture reaction with 238U. In otherwords, more efficient incineration of Pu would require Pu fuelwithout U. One challenge to fabricating Pu fuel without U is
how to do it without reducing nuclear proliferation resistancein the reprocessing facility. In the proposed reprocessing
facility shown in Fig.6-4, Pu is mixed with yttria-stabilized
zirconia (YSZ) instead of U. YSZ is essentially chemicallyinactive, and the Pu/YSZ mixed solution line is connected to
the fabrication process for PuO2-YSZ coated fuel particles.
The nuclear proliferation resistances of the proposed andcurrent reprocessing facilities are evaluated quantitatively
using the US Advanced Fuel Cycle Initiative (AFCI) method.
The results in Fig.6-5 show that the proposed reprocessing

facility is superior to the current system in terms of nuclear
proliferation resistance. Additionally, the fabrication of
PuO2-YSZ coated fuel particles is possible without major
research or development, because a similar technology has
already been developed and used for the fabrication of UO2coated fuel particles for the HTTR.
PuO2-YSZ coated fuel particles are composed of 0.5-mmdiameter PuO2-YSZ kernels and four coating layers made ofpyrolytic carbon and SiC. Owing to the use of YSZ as fuelmatrix, it is very difficult to recover Pu from a PuO2-YSZ
coated fuel particle. As a result, the PuO2-YSZ coated fuelparticle is considered to have excellent nuclear proliferation
resistance in many aspects including fuel fabrication, reactor
operation, and disposal of the spent fuel. Additionally, thePuO2-YSZ coated fuel particle is excellent for safe geologicaldisposal because the leach rate of the fission product in YSZ is
two orders of magnitude smaller than that in glass.
A preliminary analysis has been performed for a Pu burner
HTGR rated at 600 MWt to investigate the incineration rateof 239Pu. The result shows that 95% of 239Pu is incinerated
by optimizing a fuel shuffling scheme. In future work, wewill prepare the nuclear design of a Pu burner HTGR and afuel design that would allow a very high burnup of around
500 GWd/t.

Reference
Fukaya, Y., Goto, M. et al., Proposal of a Plutonium Burner System Based on HTGR with High Proliferation Resistance, Journal of Nuclear Science andTechnology, vol.51, issue 6, 2014, p.818-831.

84

JAEA R&D Review 2014

Nuclear Hydrogen and Heat Application Research

6-3 Economic Evaluation of a High Temperature Gas-Cooled Reactor Iodine-Sulfur
Process Hydrogen Production System− Investigation of Hydrogen Production Cost and Competitiveness −Capital Cost
Energy Cost

Operating Cost
CCS＊

Table 6-1 Competitiveness of the HTGR-IS hydrogen production
cost in the industry
HTGR-IS hydrogen production cost satisfies the requirement
for hydrogen reduction steelmaking and fuel cell vehicles (FCV)in the case when the hydrogen production thermal efficiency isover 40% and HTGR availability is over 80%. A goal of our R&D
is to reduce the cost.

＊＊

HyPr-RING

Coal
Gasification

Economical
Economical
HTGR-IS Hydrogen
Competitive
Competitive
Hydrogen Production
HTGR
HydrogenHydrogen
Production Thermal Availability Production Cost for
Production
Cost
Hydrogen Reduction
Efficiency
Cost(%)
for FCV
Steelmaking(%)(JPY/Nm3)(JPY/Nm3)(JPY/Nm3)

Steam
MethaneReforming

HTGR-IS

0

10
20
30
Hydrogen Production Cost (JPY/m 3)

45.2

30

70

41.4

30

80

＊

CCS: Carbon dioxide Capture and Storage＊＊
HyPr-RING: A hydrogen production process wherein CaO is fed to a coalgasifier to absorb CO2 as CaCO3

Fig.6-6 Hydrogen production costs under processes using
fossil fuels and HTGR-IS
The competitiveness of a future HTGR-IS hydrogen production
system with existing hydrogen production processes is shown.
The competitiveness of HTGR-IS is partly due to there being
no need for CCS.

Stable, large-scale hydrogen production technologies
are needed for the expected low-carbon society. The
thermochemical iodine-sulfur (IS) process is one promisingtechnology that harnesses heat energy from HTGRs.
Aiming at the formulation of a safety standard forconnecting the HTTR to a hydrogen production facility, andat the establishment of a design applicable to the standard, weare verifying the integrity of process components and a facility
made of industrial materials as a technological base for aconnection test. Although an economic evaluation of HTGR-IS
hydrogen production by a practical plant is necessary, accurate
cost estimation using an optimized large-scale commercial
plant is difficult at present.
An economic evaluation of hydrogen production by a futurecommercial HTGR-IS system was performed on the basis of
economic evaluation data from an existing commercial plantusing fossil fuel as a raw material. The following operational
data were assumed: hydrogen production thermal efficiencyof 50%, HTGR availability of 80%, and renewal of hydrogenproduction facilities every 10 years. The construction costof a hydrogen production facility was set at two times that

33
31.4

40

80

25.4

50

80

39-48

of a naphtha reforming plant of similar scale, assuming
component rationalization and material cost reduction bytechnological development. Heat and electricity prices wereset at 0.7 JPY/MJ and 5.8 JPY/kWh. The hydrogen production
cost was estimated at 25.4 JPY/Nm3. Capital and energy costs
account for 13% and 78% of the total hydrogen production
cost, respectively. Decreasing the HTGR construction cost andincreasing HTGR availability are important for cost reduction.
In Fig.6-6, this estimation is compared with evaluationsof existing hydrogen production processes based on fossil
fuels. The proposed HTGR-IS hydrogen production system
is competitive with the existing systems, without any carbon
dioxide capture and storage (CCS) being required. Considering
the competitiveness of HTGR-IS hydrogen production costsin the industry, as in Table 6-1, the process satisfies the costrequirements for hydrogen reduction steelmaking and fuel cellvehicles in the case when the hydrogen production thermal
efficiency is over 40% and the HTGR availability is over
80%. We will continue research and development of HTGR-IS
hydrogen production, setting low cost as a goal.

Reference
Iwatsuki, J. et al., Economic Evaluation of HTGR IS Process Hydrogen Production System, JAEA-Review 2014-037, 2014, 14p. (in Japanese).

JAEA R&D Review 201485

Nuclear Hydrogen and Heat Application Research

6-4 Toward a Demonstration that Reactor Safety Can Be Secured Using Only Natural Phenomena− To Establish a Test Procedure to be Employed Soon after Restart −Cooling tower

Water coolingtube

Cooler

Permanent
reflector
block
Naturalconvection

180

200
Temperature of permanent reﬂector block

180

160

160

Heat

Circulation pump

Heat
Hot plenum
block

Thermalradiation

Water coolingtrains of VCS
inactive

Operating GCs
of primarycooling system(Heat input)

Temperature (℃)

140
Water
Primary
biologicalshielding140

Temperature of hot plenum block120

Temperature of reactor pressure vessel

100

120

Water ﬂow rate in VCS

80

100

Temperature of water cooling tube

80

60

60

40

40

20

Water flow rate in VCS (t/h)

Reactorpressure vessel

20

Temperature of primary biological shielding

0

0
0

1

2

3

4

5

6

7

Elapsed time (h)Fig.6-7 Vessel cooling system (VCS)
The VCS, which is set around the reactor pressure vesselon the surface of the primary biological shielding made ofconcrete, passively removes the retained residual and decay
heat from the reactor core via the reactor pressure vessel by
natural convection and thermal radiation.

Fig.6-8 Effectiveness of the method for restricting local
temperature rise of the water cooling tube
The water circulation pump was restarted twice for several
minutes after 3 and 5 h following the start of the test. The
temperature changes in the permanent reflector block, hot
plenum block, reactor pressure vessel, and primary biologicalshielding were negligibly small, in contrast to that in the watercooling tube.

One of the safety demonstration tests for the HTTR is theloss of forced core cooling test, in which the forced cooling of
the reactor core is lost without inserting control rods into thecore and cooling by the vessel cooling system (VCS), as shownin Fig.6-7. It will be demonstrated that the HTGR is inherently
safe because of natural phenomena within the actual reactor,
even if those functions are not available to shut it down orcool it. This makes it possible to design a reactor that is free ofsevere accidents.
On the other hand, the local temperature of the uncovered
cooling tube by the thermal reflector exceeds the limit fromviewpoint for long-term use during the test, although thereactor remains safe. This temperature rise can be restrictedby restarting the circulation pump for a sufficiently shortperiod of time. However, it is difficult to predict numericallywhether restarting the circulation pump would influence thetemperatures of reactor internal structures such as a permanent
reflector block, because the configuration of the water coolingtubes is complicated.
The reactor cannot be restarted unless the conformity to the
new regulation standards is checked. Thus, in order to establish
a safe and promising test procedure, a test was conducted using
non-nuclear heat input from gas circulators while stopping
water flow in the VCS. Multiple thermocouples were set tomeasure the temperatures of water cooling tubes to specify the

position at which the temperature locally increased.
The test results showed the locally increased temperature
position, although the temperature was sufficiently lower thanthe maximum allowable working temperature. On comparing
the measured and numerical results, it was found that the local
temperature rise of the cooling tube was restricted by starting up
the circulation pump for a sufficiently short time (5 min) against
the duration period of the test (7 h). On the other hand, themeasured temperatures of the permanent reflector blocks, hot
plenum blocks, reactor pressure vessel, and primary biologicalshielding (which were focused on as aspects of safety), were notinfluenced by the short-term pump restart, as shown in Fig.6-8.
Furthermore, from the results of the flow rate measurement of
natural circulation of water by a high-accuracy electromagnetic
flow meter, it was found that the natural circulation flow did nothave any cooling effect on the temperature of the water coolingtube, because the flow rate increased and then decreased soonafter the VCS was stopped.
Through this investigation, a procedure for the loss of forced
core cooling test was established, which may be used soonafter the restart of HTTR. The safety demonstration test, whichshows that the inherent safety of the HTGR can be maintainedby natural phenomena, has attracted attention from around theworld, and has been conducted as an international joint research
project.

Reference
Takada, S. et al., Proposal of Safety Demonstration Test Plan of HTTR by Cold Test of Loss of Forced Cooling with Vessel Cooling System Inactive,
JAEA-Technology 2014-001, 2014, 34p. (in Japanese).

86

JAEA R&D Review 2014

７ Research and Development of Advanced Nuclear Systems
R&D of Fast Reactor Cycle Technology
R&D of FR Cycle Technology
 Volume Reduction and Mitigation of the Degree of Harmfulness of Radioactive Waste Safety enhancement of a Fast Breeder Reactor / Fast Reactor

R&D in “MONJU”

 Compile Research Results of Fast
Breeder Reactor
Experimental Fast Reactor “JOYO”

ChemicalProcessing Facility

Plutonium Fuel
Production Facility

International
Cooperation

Advanced
Sodium-cooled
Fast Reactor
Concept(JSFR)※1※2

Establishment of SDC,SDG

Planning of International
Joint Experiment on Severe
Accident Countermeasures

Prototype Fast Breeder Reactor “MONJU”

International Workshop on
International Collaborative
Research using “MONJU”※1 JSFR: Japan sodium-cooled Fast Reactor※2 SDC: Safety Design Criteria, SDG: Safety Design Guidelines

Fig.7-1 Overview of the research and development of fast reactor cycle technology
In line with Japan’s new national energy policies, we are planning to promote research and development on “volume reduction andmitigation of the degree of harmfulness of radioactive waste” and “safety enhancement of a fast breeder reactor/fast reactor,” utilizing
international cooperation from FY 2014 onwards.

Prior to FY 2011, research and development activities had
been conducted toward the commercialization of fast reactor(FR) cycle technology with the cooperation of electric utilities;
the main concept was a combination of a sodium-cooled FR,
advanced aqueous reprocessing, and simplified pelletizing fuelfabrication. After the accident at the Tokyo Electric Power
Company, Incorporated Fukushima Daiichi Nuclear Power
Station, we had focused our activities on efforts required toretain the technical capacity and to enhance safety of FR. Then,in April 2014, the Cabinet of Japan decided to approve the
new Strategic Energy Plan as the basis for the orientation ofJapanʼs new energy policies. In accordance with this plan, we
have implemented R&D activities on “volume reduction andmitigation of the degree of harmfulness of radioactive waste”
and “safety enhancement of a fast breeder reactor (FBR)/FR,”utilizing international cooperation from FY 2014 onwards.
On the safety enhancement issue, “Safety Design Criteria(SDC)” that we proposed were approved at the GenerationIV International Forum (GIF) in May 2013. “Safety Design
Guidelines (SDG),” which are guidance documents for
SDC, are currently being developed (Fig.7-1). This chapter
describes the subsequent enhancement of safety and reliability,

considering the features of a sodium-cooled FR and basic
research on fuel cycle technology.
Experimental data showed a possibility of retaining degradedcore material in a reactor vessel during a core destructive
accident in a FBR/FR (Topic 7-1).
The reactor building of the Japan sodium-cooled FR (JSFR) is
designed to have high resistance against extreme wind, snowfall,
and fire, as well as earthquakes and tsunamis (Topic 7-2).
Regarding the sodium/water chemical reaction, which is an
important issue for ensuring the safety and protection of thesteam generator (SG), an analysis system has been developedfor evaluating heat transfer tube failure propagation byreacting jet (Topic 7-3).
For the inspection of SG tubes, a technique for a large-scale
simulation of a magnetic field around the structure has beendeveloped (Topic 7-4).
A promising candidate material for long-life fuel cladding
tubes has been developed through efforts at improving
chemical compositions and fabrication methods (Topic 7-5).
With respect to the FBR fuel cycle, the fundamental data havebeen acquired to establish future reprocessing during a period of
co-existence of light water reactors and FBR/FR (Topic 7-6).

JAEA R&D Review 201487

Research and Development of Advanced Nuclear Systems

7-1 Investigation of Molten Core Material Behavior− An Experimental Study on the Containment of Core Disruptive Accidents −Injection nozzle(inner dia.: 40 mm)

Lower sodium
plenumWater(30 ℃)
0.0 ms

Molten corematerial
Lower structureof reactor vessel
Distance forfragmentation

Vapor bubble

150 mmWater
surfaceNozzleexit

Simulated molten corematerial (350 ℃)

4.0 msVapor bubble

8.0 ms

12.0 ms

Debris particles

SodiumDebris(Particles ofcore material)
14.0 ms

16.0 ms

18.0 ms

20.0 ms

Fig.7-2 In-vessel retention (IVR) of molten core material
Although a column of molten core material may impose alarge thermal load on the lower structure and thus degrade theintegrity of a reactor vessel, if the column is fragmented into
smaller particles, it will be easier to achieve IVR by the efficientquenching of the molten core material.

Fig.7-3 Images of fragmentation behavior
We conducted a fundamental experiment under visualobservation using a high density melt and water to simulate
molten core material and sodium, respectively. Our observations
clarified the fragmentation behavior induced by local coolantboiling and vapor expansion.

To contain the impact of core disruptive accidents (CDAs)
of sodium-cooled fast reactors in the reactor vessels, it isimportant to retain molten core material within the vessel
during CDAs. This concept is called in-vessel retention (IVR).
Molten core material discharged into the lower sodium
plenum through control rod guide tubes has the potential toimpose a significant thermal load on the lower structures of
the reactor vessels, and thus may cause reactor vessel failure,compromising IVR. However, as shown in Fig.7-2, if themolten core material is fragmented into smaller particles
well before it reaches the lower structures, the thermal load
should be significantly reduced by enhanced quenching of thismaterial. Hence, the development of a method for evaluating
the distance for fragmentation is crucial for achieving IVR.
In this study, to obtain experimental data on the distancefor fragmentation, we conducted a fundamental experimentthat simulated the discharge behavior of molten core material
under a sodium-cooled condition using simulants whichenabled visual observation, as well as the establishment of awide range experimental condition (e.g., discharge diameters).
In the sodium-cooled condition, it is believed that a heattransfer mode with a liquid-liquid direct contact should bemaintained, since even though sodium boiling would occur,
the formation of a stable vapor film should be hindered at
the contact interface between the molten core material andsodium. Fig.7-3 shows the fragmentation behavior of thesimulated molten core material in such a heat transfer mode,
which was established by a combination of the simulated
molten core material and water. The experimental results
revealed that fragmentation in the sodium-cooled condition
was induced by thermal factors, including local coolant
boiling and vapor expansion, in contrast to light water reactor
conditions where the fragmentation would be induced byhydrodynamic instabilities. Moreover, the experimental results
showed that thermal interactions facilitated the fragmentation
and reduced the distance for fragmentation down toapproximately 10 percent of the value predicted using theexisting representative correlation.
The present study, which contributes to the developmentof a method for evaluating the distance for fragmentation,suggests that the high coolability of sodium is effective for
IVR. This study was sponsored by the Ministry of Education,
Culture, Sports, Science and Technology of Japan (MEXT).

Reference
Matsuba, K. et al., Fundamental Experiment on the Distance for Fragmentation of Molten Core Material during Core Disruptive Accidents in SodiumCooled Fast Reactors, International Electronic Journal of Nuclear Safety and Simulation, vol.4, no.4, 2013, p.272-277.

88

JAEA R&D Review 2014

Research and Development of Advanced Nuclear Systems

7-2 Aiming for Robustness for the Reactor Building to Withstand Severe Natural Disasters− Evaluation of External Hazards to the Reactor Building −Partition
Steel Plate

Stud
Penetration
SleeveTable 7-1 Results of structural integrity evaluation of a reactor
building and seismic isolation system against an earthquake
The results of the ear thquake evaluation show that thestructural integrity and stability of the reactor building and theadvanced seismic isolation system were maintained, and themargin (evaluated value/requirement) was greater than two.

Tie Bar

Evaluated items

Concrete

SurfaceSteel Plate

Structural integrity

Steel plate reinforced Concrete structure

2.8 / 2.9

Shear force (NS / EW)

2.2 / 3.6

Overturning

7.7

Sliding

6.5

Vertical bearing capacity

7.8

Shear strain

2.1

Tensile strain

9.7

Compressed strain

2.4

Displacement

2.3

Reactor building

Stability

Laminated rubber bearing

Oil damper

Laminated
Seismic isolation rubber bearing
systemAdvanced seismic isolation system

Fig.7-4 Steel plate reinforced concrete (SC) structures andthe advanced seismic isolation system
SC structure comprising a pair of steel plates with a concrete
filling in between, and an advanced seismic isolation systemwith laminated rubber bearings and dampers have been
investigated.

Strong Wind・Structural Integrity
・Stability

Snow Pack
・Load

Reactor building

Tsunami・Structural lntegrity
・Stability

Oil damper

Table 7-2 Results of structural integrity evaluation of a reactor
building and seismic isolation system against a tsunami
The results of the tsunami evaluation show that the structural
integrity and stability of the reactor building and advanced
seismic isolation system were maintained if an appropriate wall
thickness was set.
Evaluated items

Stack

SC Wall

Advanced seismic
isolation system

Fire Disaster・Deflagration
Temperature
Load
・Fire
Temperature

MarginBending Moment (NS / EW)

Wall Member(wall thickness
1250 mm)
MemberMarginBending Moment

1.2

Shear force

1.3

Bending Moment(NS / EW)

471 / 286

Shear force(NS / EW)

32 / 16

BuildingSCstructure

Base mat
Earthquake
・Structural Integrity
・Stability

Fig.7-5 Evaluated external hazards
The evaluated external hazards were earthquake, tsunami,snowpack, strong wind, and fire disaster.

A steel plate reinforced concrete (SC) structure has been
considered for the building wherein the sodium-cooled fast
reactor (SFR) is housed. The SC structure comprises a pair ofsteel plates with a concrete filling in between, held together bystuds, etc. In addition, an advanced seismic isolation systemwith laminated rubber bearings and oil dampers has been
investigated for adoption against earthquakes (Fig.7-4).
In response to the Great East Japan Earthquake and theaccident at the TEPCO’s Fukushima Daiichi NPS in March
2011, the structural integrity and stability of reactor buildings
against external hazards have been re-evaluated. Fig.7-5 showsthe evaluations for earthquake, tsunami, snowpack, strong wind,and fire. These external hazards are threats to the structural
integrity and stability of the reactor building; hence, to protect
the safety of a reactor building, it must be constructed in such a

Stability

Seismic
isolationsystemLaminatedrubberbearing
Oil damper

Overturning

20

Sliding

22

Vertical bearing capacity

11

Shear strain

3.2

Tensile strain

402

Compressed strain– (Not occur)

Displacement

1.8

way as to withstand them. In fact, it is necessary to ensure thatloads due to earthquakes and tsunamis do not damage an outerwall nor affect the reactor building inside.
The results of the evaluation against earthquake and tsunami
are shown in Tables 7-1 and 7-2, respectively. They show thatthe structural integrity of a reactor building can presumably bemaintained with appropriate wall thickness, and that seismic
or tsunami loads do not lead to the reactor building being
damaged. Evaluation results were also obtained showingthat structural integrity could be maintained against otherevaluated events, such as snowpack, strong wind, and fire.
This study includes some of the results of the “technicaldevelopment program on a commercialized FBR plant”
entrusted to Japan Atomic Energy Agency (JAEA) by the
Ministry of Economy, Trade and Industry of Japan (METI).

Reference
Yamamoto, T. et al., Evaluation of External Hazard on JSFR Reactor Building, Proceedings of 2013 International Congress on Advances in Nuclear Power
Plants (ICAPP 2013), Jeju Island, Korea, 2013, paper FD211, 9p., in USB Flash Drive.
JAEA R&D Review 201489

Research and Development of Advanced Nuclear Systems

7-3 Numerical Simulation for Safety Evaluation of the Steam Generators in Fast Reactors− A Multiphysics Evaluation System for Tube Failure Accidents −(a) Related experiments (blue) and their applications (red)Water/
vapor

Exp. on underexpanded jet
Gas discharging in water(SERAPHIM)
Exp. on self wastage
Self wastage of microscopic
hole (SEWAP)

Rea
cting
jet

LiquidsodiumWater/
vapor

Exp. on over-heating rupture
Water-side heat transfer (RELAP5),
sodium-side heat transfer, and rapidlyheated tube rupture (TACT)Exp. on target wastage (LDI+FAC)
Liquid droplet impingement erosion,
corrosion by high-temperature NaOH,
and water vapor discharging in liquidsodium (SERAPHIM, TACT)

Vessel(Simulated SG)

Water vapor
line

Cover gas
Reaction productrelease line

Exp. on chemical reaction
Sodium-water counter-flow
diffusive reaction and differential
thermal analysis (SERAPHIM)(b) An example: an experiment on water vapor discharging in sodium

Failed tube

Exp. under actual
conditions of SG

Liquid sodium level
Target tube

20 mmNozzleWastage markReacting jet
Discharging tube

Integrated validation exp. (Mechanistic evaluation
system and long-time process analysis method)

Knowledge database

Sodiumline

Experimental result(wastage of target tube)

Fig.7-6 Phenomena appearing under tube failure accidents in a steam generator and their related experiments
Some experiments related to tube failure accidents were performed to develop and validate the numerical methods.
(c) An overview of the multiphysics evaluation system
Mechanistic evaluation system(1)(4)
Failed tube

(2)(1) SERAPHIM code: Evaluate
behaviors of reacting jet & wastage
environment(3)
Adjacent tube

(4) SEWAP code: Evaluate
enlargement of initial crack

B.C.

(2) TACT code: Evaluate heat transferfrom reacting jet to tube wall andpossibility of failure propagation(d) An example: analysis of water vapor discharging in sodiumusing SERAPHIM code01300 ℃

01600 m/s

Agreement withmeasured data
High-temperature
region0500 m/s

90° rotatedAgreement with wastage
position in exp. (Fig.7-6(b))
Underexpansion(supersonic)High
velocity

B.C.
(5) Long-time process analysis method(6) Pressure wave propagation
analysis method(3) RELAP5 code: Evaluate thermal
hydraulic behavior of water inside tube
Discharging tube

Gas-phase temperature Gas-phase velocity

Liquid droplet velocity

Fig.7-7 A multiphysics evaluation system and an example of numerical simulation
To evaluate tube failure accidents, a multiphysics evaluation system comprising mechanistic numerical methods was developed.
The applicability of the numerical methods was confirmed through the analyses of the related experiments.

When pressurized water or vapor leaks from a failed
heat transfer tube in a steam generator (SG) of sodiumcooled fast reactors, a high-velocity, high-temperature jet
reacts chemically with the surrounding sodium and causes
wastage and degradation of the mechanical strength on theadjacent tubes (Fig.7-6). Significant progress of such material
damage may lead to failure propagation. Prevention of
failure propagation is a major concern in designing the SG.
In this study, a multiphysics evaluation system comprisingmechanistic numerical methods was newly developed toevaluate the possibility of the occurrence of the failure
propagation. The developed system is applicable to a wider
variety of operating conditions and design options of the SG
than the conventional evaluation method based on exhaustive
demonstration experiments.
The systematic experiments shown in Fig.7-6 wereconducted for the elucidation of phenomena and theacquisition of data to validate the numerical methods. In
these experiments, we succeeded in identifying a dominant
process in the sodium-water chemical reaction, measuring
the sodium-side and water-side heat transfer behaviors,
measuring the rupture behaviors of the rapidly-heated tube,
deriving a correlation between the wastage rate and someparameters relevant to liquid droplet impingement erosion
and flow accelerated corrosion, and so forth. As an example,
Fig.7-6 shows the experiment on water vapor discharging
in liquid sodium to obtain data for target wastage under
actual plant conditions. A knowledge database including
the quantitative uncertainty of the experimental data wasconstructed. As shown in Fig.7-7, the mechanistic numericalmethods were developed for the three regions: (1) the reacting
jet, (2) the heat transfer tube, and (3) the inside of the tube.
About region (1), a SERAPHIM code with an incorporated
wastage environment evaluation model was developed. The
applicability of the SERAPHIM code was confirmed throughthe analysis of the abovementioned water vapor discharging inliquid sodium (Fig.7-7). A TACT code involving the numericalmodels for heat transfer from the reacting jet to the tube
wall and failure judgment (region (2)), and a RELAP5 code
involving the water-side heat transfer correlation applicable to
the rapid heating condition (region (3)) were also constructed.
Our evaluation system can make a large contribution toward
designing the SG with an appropriate safety margin.
The present study was partly supported by the Ministry of
Education, Culture, Sports, Science and Technology of Japan(MEXT).

Reference
Uchibori, A. et al., Development of Numerical Evaluation Methods for Multi-Physics Phenomena under Tube Failure Accident in Steam Generator of
Sodium-Cooled Fast Reactor, Nippon Kikai Gakkai Ronbunshu, B Hen, vol.79, no.808, 2013, p.2635-2639 (in Japanese).

90

JAEA R&D Review 2014

Research and Development of Advanced Nuclear Systems

7-4 Challenges Facing the Improvement of Inspection Technology for Steam Generator Tubes− Large Scale 3D Simulations of ISI of FBR “Monju” SG Tubes Using ECT −(c)
OD 20%
grooveHigh(b)

Support plate
Exciter coil position

OD 30%
partialMagnetic vector potential(a)

SG helical tube

Low

Large SP with 3
SG tubes as in Na
tank experiment(2011)SG helical tube

Fig.7-8 (a) View of the mock-up test model; (b) FEM model; (c) Magnetic vector eddy current field distribution(a) A large support plate (SP) model with three SG helical tubes. (b) FEM model using 2nd order tetrahedrons. (c) The magnetic
vector eddy current field distribution obtained by scanning the upper SG helical tube with ECT sensors.
SP
SP+OD20％ groove

Simulation (250 Hz)

SP
SP+OD20％ groove

Experiment (250 Hz)

SP
SP+OD30％ partial

Simulation (250 Hz)

8

8

6

6

6

4

4

4

4

2

2

2

2

0
−2

0
−2

Im (V)

8

6

Im (V)8

Im (V)Im (V)

Experiment (250 Hz)

0
−2

0
−2

−4

−4

−4

−4

−6

−6

−6

−6

−6

−4

−2

0

Re (V)

2

4

6

8

−6

−4

−2

0

2

4

6

Re (V)

8

−6

−4

−2

0

Re (V)

2

4

6

8

SP
SP+OD30％ partial

−6

−4

−2

0

2

4

6

8

Re (V)

Detection coil signal complex representation (Voltage vector)

Fig.7-9 Comparison between ECT measurements and FEM simulations of SP and defects
The OD20% groove is a fully circumferential defect and the OD30% partial groove is a quarter-circumferential defect. The excitation
frequency in both experiment and simulation is 250 Hz.

The eddy current technique (ECT) is a well-establishedmethod for detecting defects in metallic structures, whichis easier and more repeatable than alternatives; it works bymonitoring changes in a coilʼs impedance. When applying
ECT to the assessment of the soundness of the SG tubes of“Monju” FBR, the complexity of the large metallic support
plate (SP) structure surrounding the long SG helical tubes
adds additional electromagnetic interference or noise in the
primary electromagnetic field created by the ECT coils. To
improve the inspection technology for FBR SG tubes, precise
evaluation of these effects requires reliable and verified
numerical electromagnetic tools operating in conditions as
close as possible to those in “Monju.”
A three dimensional (3D) electromagnetic code, using finite
element method (FEM), researched and developed in-house at
JAEA, analyzed the electromagnetic field of an ECT systemcomposed of excitation-detection coils. The code was validatedby comparing numerical simulations with experimental
measurements using three SG helical tubes with a large SP
mock-up (Fig.7-8(a)). Large scale numerical simulations of

FEM models with up to 5 million second-order tetrahedrons(Fig.7-8(b)) were run on JAEA supercomputers with
1024 parallel CPU cores, distributed in a mixed Open MP-MPI
environment (Fig.7-8(c)). Variation in the defect signals due tohelical structures and defect proximity to the SP was comparedagainst measurements for outer tube defect (OD) grooves
spanned full circumferential with depth 20%tw (from tube
wall thickness), and around defects that spanned only a quarterpart of the circumference and had a depth of 30%tw. A fullcomparison with the experimental results for defect and SP
signals validated the FEM computation of ECT signals for SG
helical tubes (Fig.7-9).
Large scale numerical simulations using reliably tested
and verified FEM codes are aimed at providing improved
understanding of the ECT signal/noise ratio, where
electromagnetic interference appears due to the close proximityof multiple tubes or intricate 3D structure geometries tofacilitate further development of inspection technology for“Monju” SG tubes.

Reference
Mihalache, O. et al., Large Scale 3D FEM Electromagnetic Simulations and Validations for FBR Steam Generator Tubes, Studies in Applied
Electromagnetics and Mechanics, vol.38, 2014, p.94-102.
JAEA R&D Review 201491

Research and Development of Advanced Nuclear Systems

7-5 Development of Long Life Fuel Cladding Tube for the Fast Reactor− 11Cr-ODS Steel with Sufficient Strength, Toughness, and Corrosion Resistance −9Cr-ODS steel (Pre-mix/Pre-alloy)
11Cr-ODS steel (Pre-mix)
11Cr-ODS steel (Full pre-alloy)

High

10

8

Strength increase

100
90
80
70

Low

50
10

100

6
Approximately
4 times42

Conventional 11Cr steel(PNC-FMS)

60

Absorbed energy (J)200
Toughness

Creep rupture stress (MPa)300

Conventional 11Cr steel (PNC-FMS)
11Cr-ODS steel (Pre-mix)
11Cr-ODS steel (Full pre-alloy)

1000

10000

Time to rupture (h)

0
−200

−100

0

100

200

300

Temperature (℃)

Fig.7-10 Creep strength of 9 and 11Cr-ODS steels at 700 °C
The 9 and 11Cr-ODS steels exhibit the same level of creep
strength, showing that the effect of Cr increase on creep
strength is small. The suspected reason is that 11Cr-ODS
steel is sufficiently strengthened by the application of oxide
dispersion.

Fig.7-11 Toughness of full pre-alloy and premix ODS steels
We have succeeded in producing 11Cr-ODS steel with good
toughness. The toughness is equal to that of conventional
11Cr steel (PNC-FMS) with good toughness, which is notstrengthened by oxide dispersion.

To make the advanced fast reactor (FR) safe and economical,
it is necessary to develop fuel cladding tube with superior
strength and dimensional stability, since the cladding tube will
be used under severe conditions (e.g., high temperature andhigh-dose neutron irradiation). Hence, we have researched anddeveloped oxide dispersion-strengthened (ODS) steel cladding
tubes. 9Cr-ODS steel cladding tubes have been produced asa primary candidate, and engineering requirements for massscale production of these tubes have been studied.
If the dissolution of cladding tubes in nitric acid increased
during the spent fuel reprocessing process, the amount ofhigh-level radioactive waste (vitrified waste) and the costof the fuel-cycle system would both increase. Therefore, wehave also developed ODS steels containing over 9wt% Cr to
improve corrosion resistance in nitric acid.
High-Cr ODS steel should be granted properties equivalent
to those of 9Cr-ODS steel to be satisfactory for use as afuel cladding tube. Therefore, the matrix must be tempered
martensite, as it is for 9Cr-ODS steel. We selected 11Cr forthe high-Cr martensitic ODS steel, and modified the otherelement concentrations based on our research into the structure
control (i.e., crystal structure and grain size) and strengthening
mechanism. The high-temperature and long-term strength ofthe optimized 11Cr-ODS steels were evaluated by creep tests
at 700 °C which is assumed to be the highest temperaturethat the cladding will attain during normal operation of anadvanced FR. The creep strength of 11Cr-ODS steels wasmuch better than that of conventional 11Cr heat-resistant steel(PNC-FMS), and nearly equal to that of the 9Cr-ODS steels,as shown in Fig.7-10.
In general, high strength steels tend to have low toughness,i.e., they are weak against cracks and impacts. In addition, ODS
steel is subject to the influence of defects, such as inclusions(i.e., coarse oxide and nitride), that degrade its toughness. We
applied a full pre-alloy metallurgy method to the productionof 11Cr-ODS steels, instead of a simple pre-mix metallurgy
method. The pre-alloy method can greatly decrease inclusions,leading to fracture. The toughness of the full pre-alloy ODS
steel is much greater than that of pre-mix ODS steel, as shownin Fig.7-11.
We have confirmed the potential of 11Cr-ODS steel withgood corrosion resistance for use as the material for the longlife fuel cladding tube. In the next step, we will proceed to
conduct several kinds of tests on the cladding tubes composedof full pre-alloy 11Cr-ODS steels.

ReferenceTanno, T. et al., Evaluation of Mechanical Properties and Nano-Meso Structures of 9-11%Cr ODS Steels, Journal of Nuclear Materials, vol.440, issues 1-3,
2013, p.568-574.

92

JAEA R&D Review 2014

Research and Development of Advanced Nuclear Systems

7-6 Toward a New Reprocessing Process for the Transitional Period to the Fast Reactor Cycle− The Development of a Uranium and Plutonium Co-recovery Process (Co-processing Process) −Co-decontamination and partitioning cycle
Pu, U loaded solvent

TBP/nDD(solvent)Partitioning(Pu: Back extraction
U: Back extraction in part)U washing(removed)

Co-decontamination(U, Pu: Extraction)

U, Pu mixed solution

Pu, U concentration tuning

Pu/U purification cycle
Oxidation

Solvent reflux
stream(removed)

TBP/nDD(solvent)Extraction/Washing

U washing(removed)

Solvent washing

Pu–U strip
Total Pu, U recovery

U, Pu mixed solution

Fig.7-12 Schematic of the co-recovery process
The elimination of the U washing stages and the solvent reflux stream, which are both partof the conventional PUREX process, prevents Pu isolation in the process and improves its
proliferation resistance.

Pu/U Ratios in Feed Solution
1%

3%

20%

Reductant(U(IV))

Diluted nitric acidand reductant (HAN)

Pu,Uloaded solvent
Partitioning

Pu Concentrationin Organic Phase (g/ℓ)

1E+01
1E+00
1E−01
1E−02
1E−03
1E−04

Analytical Limit

1E−05

Number of Mixer-Settlers→Solvent
Exit

Fig.7-13 Reduction of Pu in the partitioning stage
Pu is stripped into an aqueous phase in this stage by reducing
it to trivalency using U(IV) and HAN as reductants.

Fig.7-14 Pu concentration profiles in the partitioning stage
Pu is completely stripped in this stage because Pu concentrations
in solvent effluents are less than the analytical limit.

The transition from the current light-water reactor (LWR)
cycle to the FR cycle will span several decades. In this period,
LWRs and FRs will coexist, and not only UO2 fuel but also
mixed oxide (MOX) fuels will be discharged from thesereactors. For the purpose of reprocessing these spent fuelswith varying plutonium contents, a U-Pu co-recovery process(co-processing process) is currently being developed. In contrast
to the conventional reprocessing process (PUREX), which
recovers Pu separately from U, the co-processing process will
recover the two elements together to prevent the isolation of Pu
to improve the proliferation resistance (Fig.7-12).
Mixer-settler tests of the co-processing process have been
conducted using U/Pu nitric acid solutions at the Operational
Testing Laboratory (OTL) of the Tokai Reprocessing Plant(TRP). To assure complete stripping of Pu in a mixture with
U in the partitioning stage, the appropriate usage of reductants(U(IV) and HAN)—which reduce Pu to a non-extractable
trivalent state—has been evaluated in the tests (Fig.7-13).
The adopted Pu/U ratios of the feed solutions used in thetests were 1%, 3%, and 20%, considering the composition of
the future spent fuels from a LWR, a LWR-MOX hybrid, anda FR-MOX reactor. In the tests, complete stripping of Pu in a
mixture with U has been observed at all Pu/U ratios (Fig.7-14),
and the U/Pu ratios in the recovered U/Pu mixed solutions
could be controlled within the 0.5–2.0 U/Pu ratio range, which
was suitable for the fabrication of FR-MOX fuel.
Further tests are to be conducted to improve the process,and will use centrifugal contactors.
The present study was sponsored by the Agency for Natural
Resources and Energy, Ministry of Economy, Trade andIndustry, of Japan (METI), in the FY 2010 and 2011.

Reference
Yamamoto, K., Ohbu, T. et al., Development of U and Pu Co-Recovery Process (Co-Processing) for Future Reprocessing, Proceedings of International
Nuclear Fuel Cycle Conference (GLOBAL 2013), Salt Lake City, USA, 2013, paper 7797, 4p., in CD-ROM.
JAEA R&D Review 201493

８ Research and Development Relating to Nuclear Fuel Cycle Backend and Reprocessing of Spent Nuclear Fuel
Progress in Decommissioning of Nuclear Facilities and Treatment and Disposalof Radioactive Waste
Safe and efficient decommissioning of our nuclear facilities and treatment and disposal of radioactive wastes are important issues
to consider in our research and development activities. We are currently setting up systems to decommission nuclear facilities andmanage radioactive waste and are also developing the related technologies (Fig.8-1).
Furthermore, we will be responsible for disposing radioactive waste generated not only from our research facilities but also fromuniversities, industrial facilities, and other sites.
R&D for the decommissioning of nuclear facilities
We have developed an engineering system for decommissioning
and a verification or evaluation system for clearance. It is
important to assess the applicability of these systems to thedecommissioning of several nuclear facilities of JAEA. We
already confirmed that the system is applicable to a research
reactor. This time, we applied the system to decommissioningof the Uranium Refining and Conversion Plant, which is oneof the nuclear fuel cycle facilities, and showed that the system
is applicable to it. (Topic 8-1).
R&D for waste treatment and disposal
There are various chemical forms of sludge-like uraniumbearing waste, and its uranium concentration is very high.
We must select which uranium separation process to apply
to various uranium-bearing wastes. Therefore, we have beendeveloping a process for selective separation of uranium byusing mineral acid. As part of this study, we confirmed that theprocess that uses hydrochloric acid is applicable to the CaF2
sludge and spent filter aid (Topic 8-2).
R&D to characterize radioactive waste
For the safe disposal of radioactive waste packages, theirradioactive content must be evaluated. Therefore, we developed
related techniques to enable reasonable and effective analysis

Planning of disposal
facilitiesof the radioactive waste. However, because some radionuclides
contained in the radioactive waste are hard to analyze, wedeveloped a new method to analyze 242mAm (Topic 8-3).
Moreover, to deal with the accident at the Tokyo Electric Power
Company, Incorporated (TEPCO) Fukushima Daiichi Nuclear
Power Station (NPS), we are developing a new analyticalmethod to measure the concentration of radionuclides in thewastewater. Our systematic method to analyze radioactive
waste is also being applied to determine the concentrationof radionuclides in or on the rubble and trees (Chapter 1,
Topics 1-20, 1-21).
Disposal of low-level radioactive waste
Trench-type disposal is a method for disposing low-level
radioactive waste in a facility without engineered barrier locatedat a few meters below the surface. To improve the safety of thetrench facility, it is important to reduce the rate at which rainfall
infiltrates into the waste layer (due to the low permeability ofcover soil). Therefore, we evaluated the rate of water infiltration
when a geomembrane and a low-permeability soil layer are
installed in the upper cover soil, considering the permeabilityof these layers. We found out that using a low-permeability soil
layer and a geomembrane is effective to reduce water infiltration
rate into the waste layer of the trench facility (Topic 8-4).

Design for shallow land disposal facilities

Treatment andstorage ofradioactive waste

Waste
Operation and maintenance packageof existing waste treatment
facilities (Advanced volume
reduction facilities, etc.)
Construction ofwaste treatment
facilities at each site

Decommissioning of
nuclear facilities

Decommissioning of aging nuclear facilities or nuclear
facilities that have completed their missions
Consolidation/Reuse of facilities and equipment

Waste package
inspection

Clearance

Waste from dismantling

Developmentof relevant
technology

Fig.8-1 Outline of measures for decommissioningof nuclear facilities and treatment and disposalof radioactive waste
We are setting up systems to decommission
nuclear facilities and manage radioactive waste. In
addition, we are developing related technologies(e.g., decommissioning, treatment, and disposal)and constructing radioactive waste treatment anddisposal facilities.

R&D to Improve Technology and Reliability of Geological Disposal in Japan
Geological disposal is one option for long-term isolationfrom human environments of high-level radioactive waste(HLW) produced during nuclear power generation. Thisis a critical issue that the present generation must sensibly
deal and that remains crucial despite any revision of thenational nuclear energy policy. In Japan, spent fuel frompower reactors is reprocessed to extract reusable uranium and
94

JAEA R&D Review 2014

plutonium for power generation. The liquids separated fromthe spent fuel during chemical reprocessing are solidified
into a stable glass form. In the Japanese disposal concept,
vitrified wastes are encapsulated in a thick steel overpack
surrounded by highly compacted bentonite and then placedin a stable geological environment at more than 300 m below
the surface (Fig.8-2). Implementing geological disposal ofOverpack(e.g., carbon steel)

Stable Geological
Environment(Natural Barrier)(greater than 300 m
below surface)Fig.8-2 Schematic view of basic concept for the geologicaldisposal of high-level radioactive waste (HLW) in Japan

performance of the disposal system, of engineered barriersystems, and of the long-term chemistry and migration ofradionuclides at Tokai to expand our knowledge base forgeological disposal (Topics 8-10, 8-11, and 8-12). These studies
exploit data and information about geological environmentsthat were obtained through geoscience research at both
URLs. The prototype knowledge management system thatwas developed in 2010 is being improved to systematicallyprovide and transfer multiple associated R&D results to both
implementer and regulator and to ensure their safety.
Our efforts also focus on promoting public understanding by
disseminating relevant information and by opening our R&D
facilities to the public.

●Horonobe URL＊ (Sedimentary Rock)＊●Mizunami URL (Crystalline Rock)Cooperation and
Collaboration

BufferMaterial(e.g., bentonite)

Horonobe UndergroundResearch Center(Horonobe, Hokkaido)

Tono Geoscience Center(Toki & Mizunami, Gifu)

Other Organizations

Vitrified
Waste

Engineered Barrier System

HLW is a long-term project that will last over 100 years. The
project begins with site selection and continues to repository
construction and operation, which will be followed by backfill
for repository closure. It is thus of great importance to proceed
efficiently with the project, as a national responsibility, bycontinuously improving its sound technical basis and applying
these attitudes to implementation, regulatory activities, and,most importantly, to enhance public confidence. To thisend, we made and will continue to make steady progress inresearch and development (R&D) in various fields, such as
in geoscience, repository engineering, and safety assessment,
to improve the technologies used for and reliable geologicaldisposal in Japan.
At present, our R&D focuses specifically on projects at two
underground research laboratories (URLs)—one at Mizunami,
which researches crystalline rocks, and the other at Horonobe,
which researches sedimentary formations (Fig.8-3)—with
the main aim being to develop a sound technical basis forformulating and implementing safety regulations. At the endof 2013, some research galleries were made available for use
at a depth of 500 m at Mizunami and at 350 m at Horonobe.
Multidisciplinary investigations are ongoing, because thereliability of various investigative techniques should be tested
and verified before the site characterization program begins(Topics 8-5, 8-6, and 8-7). In addition, studies on tectonics,
volcanic and faulting activities, and so on are in progress toevaluate the long-term stability of geological environments inJapan (Topics 8-8 and 8-9). In parallel with such geoscience
efforts, we are conducting an extensive study to assess the

JAEA Knowledge
Management System

Knowledge Base(Concept image)

Nuclear Fuel Cycle
Engineering Laboratories(Tokai, Ibaraki)(Concept image)・Data・Documents
・Software・Experience and Expertise・Simulations
・Evidence・Expert Systems●Engineering-scale Test & Research Facility (ENTRY)●Quantitative Assessment Radionuclide
Migration Experiment Facility (QUALITY)
ENTRY

＊

URL: UndergroundResearch Laboratory

QUALITY

Implementer

Regulator

Public Understanding
and Confidence

Fig.8-3 System for implementing JAEA R&D activities

Implementing Safety Measures at Tokai Reprocessing Plant and Enhancing Reprocessing Technologies
After the Great East Japan Earthquake and the accident at
the TEPCO’s Fukushima Daiichi NPS, various emergency
safety measures were implemented at the Tokai Reprocessing
Plant (TRP). To reduce the probability of hazards at the TRP,
efforts focused on solidifying and stabilizing as soon as
possible the highly active liquid waste (HALW) at the Tokai
Vitrification Facility (TVF) and the plutonium solution at
the Plutonium Conversion Development Facility (PCDF).
Vitrifying the all stored HALW is estimated to require two
decades. Thus, an advanced glass melter is under development

to make steady progress on the solidification and stabilizationof the HALW.
For low-level radioactive effluents, a cement-based solidification
method accompanying a nitrate-ion decomposition process has
been developed to reduce the environmental impact.
The following studies on advanced reprocessing technologies
are ongoing:- Separating Ruthenium from HALW (Topic 8-13).
- Developing Uranium and Plutonium co-recovery process.

JAEA R&D Review 201495

Research and Development Relating to Nuclear Fuel Cycle Backend and Reprocessing of Spent Nuclear Fuel

8-1 Toward an Optimized Decommissioning Plan
− Development of Formula to Evaluate Manpower Requirements for Dismantling Nuclear Facility −ClassifierC
as err＊Tower & Vessel

Hydrator

Separator

DeviceTrap

Crusher

Compressor

Dehydrator

Manpower needs y (man-hour)

1200

Tower
T
To
w r&
we
Ve
V
Vesselesse

Simplified Formula y = 105.6 w(Determination Coefficient R2 = 0.95)

1000

800

600

400

200

Sep
Separato
Separator
paratoror
0

Cradle

0

2

4

6

8

10

Weight of DP equipment w (ton)Fig.8-4 Typical DP equipment and its state before
dismantling
All DP equipment was contaminated with U compounds
and installed in “Cradle” or on “Concrete base.”

Fig.8-5 Correlation between the weight of DP equipment
and manpower requirements for dismantling DP equipment
Almost all types of DP equipment were dismantled by thesimilar procedure. Because of this, the manpower requirements
for dismantling these types of DP equipment could be
estimated by one simplified formula with high accuracy,
although the number of data or their range for each type of DP
equipment was limited.

Nuclear facilities are dismantled when their mission iscomplete. To implement effective dismantling, we are setting
up a system to estimate a priori project management datasuch as cost, manpower requirements, exposure dose, andwaste generation. This system uses formulas established byanalyzing data collected during previous dismantling projects.
The formulas to determine the manpower requirements for
dismantling the equipment, for example, were represented bylinear approximations with the equipment weight according
to the equipment type. We have already confirmed that theseformulas are applicable to a research reactor, but not to othertypes of nuclear facilities. Therefore, we verified the formulas
for use with one of the uranium (U) handling facility: the U
Refining and Conversion Plant (URCP).
At the URCP, many types of equipment were dismantled,such as “Tower & Vessel,” “Trap,” “Crusher,” “Separator,”and “Compressor.” The correlation between equipment weightand manpower required for dismantling were analyzed,
revealing that the formulas for the type of equipment could

be also applied to the URCP. However, few actual data werecollected for some types of equipment that making formulas
for these was impossible. Moreover, some formulas had low
accuracy because every equipment was very light and the
distribution of data was localized.
Therefore, we considered grouping the actual data together
for equipment that had similar installation and dismantlingprocedures. In this case, we determined that most types ofdry conversion process-related equipment (DP equipment)
used to run U powder inside were supported by “Cradle”
and “Concrete base” (Fig.8-4) and were dismantled by asimilar procedure. Because of this, a simplified formula was
fit to these types of DP equipment (Fig.8-5). The simplified
formula depends on the total weight of the DP equipment
and accurately provides the manpower requirements for
dismantling all DP equipment.
We are currently developing an appropriate formula as per
the type of nuclear facility, and improve the system to theuseful one to optimize the decommissioning plan.

Concretebase
CCon rette bncase＊Here, “Classifier” is categorized as “Crusher.”＊

Reference
Izumo, S. et al., Development of Evaluation Models of Manpower Needs for Dismantling the Dry Conversion Process-Related Equipment in Uranium
Refining and Conversion Plant (URCP), Proceedings of the ASME 2013 15th International Conference on Environmental Remediation and Radioactive
Waste Management (ICEM 2013), Brussels, Belgium, 2013, ICEM2013-96097, 9p., in CD-ROM.

96

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Research and Development Relating to Nuclear Fuel Cycle Backend and Reprocessing of Spent Nuclear Fuel

8-2 Efficient Separation of Uranium from Sludge-Like Uranium-Bearing Waste− Developing Process to Selectively Separate Uranium by Using Hydrochloric Acid −Sludge-like uranium,
spent adsorbent
Dissolution process
Hydrochloric acid,
Aluminum chloride

Insoluble
residue

Precipitation process
Hydrogen peroxide

Uranium
peroxide

Removal of trace
uranium process

Spent resin

Chelating resin

Fig.8-6 Drums for storing sludge-like uraniumbearing waste
Neutralized
precipitate

Neutralization process
Sodium hydroxide

Neutralized
precipitation

Aluminum recovery
process

Sodium hydroxide

Removal of tracefluorine process

Chelating resin

Disposal

Discharge

Fig.8-8 Process using hydrochloric acid to separate uranium
from sludge-like uranium-bearing waste
CaF2 sludge
spent filter aid
Sludge-like uranium

spent adsorbent

Fig.8-7 Sludge-like uranium and the spent adsorbent

Technologies for producing uranium (U) hexafluoride weredeveloped from 1964 to 2001 at the Ningyo-toge Environmental
Engineering Center. Consequently, approximately 1500 t ofuranium-bearing waste was generated and safely stored indrums (Fig.8-6).
Some waste is in a pellet form, and other wastes have a highfraction of insoluble component, U, or hazardous elements(Fig.8-7). To remove U and hazardous elements from various
uranium-bearing wastes, we proposed the hydrochloric acid(HCl) process (Fig.8-8). U is dissolved in HCl (dissolution
process) and separated as a precipitate (precipitation process)by using hydrogen peroxide. Trace U and hazardous materials
remaining in the solution are removed by using a chelating
resin that absorbs the ions that need to be removed. Aluminum
chloride is added in the solution during the dissolution stages
to decrease the concentration of impurities mixed in withthe separated U. The aluminum is then selectively recovered

and reused. The insoluble residue, the spent resin, and theneutralized precipitate should be disposed of as a radioactive
waste. As part of the present study, we investigated whether
the process can be used for CaF2 sludge and the spent filter aid.
Approximately 90% of the U was separated at the precipitation
stage. Furthermore, highly concentrated U (60–70wt%),
which might be useful as a resource, was separated. The study
confirmed that the concentration of U in solution dropped
below the discharge standard concentration and that, byremoving trace U, the concentration of U in the neutralized
precipitate became sufficiently low.
The investigation results confirm that the majority of the
U in CaF2 and the spent filter aid is separated and that U
concentration becomes sufficiently low to be disposed of.
We will investigate whether the process can be used for othersludge and spent adsorbents.

Reference
Ohashi, Y. et al., Technique for Recovering Uranium from Sludge-Like Uranium-Bearing Wastes using Hydrochloric Acid, Journal of Nuclear Science andTechnology, vol.51, issue 2, 2014, p.251-265.
JAEA R&D Review 201497

Research and Development Relating to Nuclear Fuel Cycle Backend and Reprocessing of Spent Nuclear Fuel

8-3 Developing Rapid Method to Analyze Radioactive Waste− Developing Method to Analyze 242mAm in Low-Level Radioactive Waste −Reservoir

250

Samplesolution

An3+ fraction200

241+242m+243

Am+Cm

Extraction
chromatographic
resin

Count

243

150

Am (239Np)(242Am+239Np)

242

100

Am

50
0
100

150

200

250

Channel number
Pump

Fig.8-9 System of extraction chromatography
A cartridge-type of extraction chromatographic
resin was used. Solution in a reservoir is pumped
from downstream into the cartridge.

Fig.8-10 β-ray spectrum of 242Am and so on
An3+ fraction contains 241+242m+243Am and Cm, and β-ray spectrum of242
Am and 239Np, which are progeny nuclides of 242mAm and 243Am,
respectively is obtained. 243Am source was prepared to estimate β-ray
spectrum derived from 239Np, which was subtracted from the spectrumof the An3+ fraction to obtain the β-ray spectrum of 242Am.

The radioactive inventory of low-level radioactive waste(LLW) must be evaluated to dispose of LLW. However, somenuclides that are important for safety are difficult to analyze,
such as americium-242m (242mAm). Thermal ionization mass
spectrometry (TIMS) is a conventional method to determine
the content of 242mAm. 242mAm can also be analyzed by α-ray of
curium-242 (242Cm), which is the progeny nuclide of 242mAm(242Cm method). Although the TIMS measuring time is short,more sample is required than for 242Cm method. However,less sample is available for the 242Cm method, it takes several
months to determine the 242mAm content. In addition, bothmethods require separation of Am from Cm.
In the present study, we developed a new method todetermine the 242mAm content that involves measuring β-ray
emission from 242Am, which is the progeny nuclide of 242mAm.
To measure β-rays from 242Am, Am must be separated
from β-ray emitting nuclides in LLW. Therefore, we used
a separation method based on extraction chromatography(Fig.8-9). First, trivalent actinides (An3+) and lanthanides(Ln3+) were separated from the major elements in LLW by

using transuranic resin (Eichrom Technologies), which has
selectivity for transuranic elements. Next, An3+ was separated
from Ln3+, which has similar chemical properties and contains
some β-ray emitting nuclides, by using a tetravalent actinide
resin (Eichrom Technologies). This An3+ fraction contains
Cm, but the influence of Cm on the β-ray measurement canbe subtracted because there is no β-ray emitting Cm in LLW.
Conversely, β-rays from neptunium-239 (239Np), which is aprogeny nuclide of 243Am, have to be detected and considered.
Therefore, a 243Am source was prepared to measure β-ray
spectra of 239Np, which were subtracted from those of An3+ toobtain the radioactivity of 242Am (Fig.8-10). As a verification,
the value obtained was compared with that determined by the242
Cm method.
The separation step in the new method to determine theradioactivity of 242Am is simpler than that for the TIMS and242
Cm methods. Less sample is required by this new methodthan by the TIMS method and is comparable to that requiredby the 242Cm method. The measuring time for this new method
is considerably shorter than that for the 242Cm method.

Reference
Shimada, A. et al., A New Method to Analyze 242mAm in Low-Level Radioactive Waste Based on Extraction Chromatography and β-ray Spectrometry,
Analytical Chemistry, vol.85, no.16, 2013, p.7726-7731.

98

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Research and Development Relating to Nuclear Fuel Cycle Backend and Reprocessing of Spent Nuclear Fuel

8-4 Study to Improve the Safety of Trench Facility− Suppression of Infiltrated Water to Reduce Radioactive Substances from the Facility −Table 8-1 Resulting runoff, evapotranspiration, andwater infiltration into upper cover soil
44% of the rainfall infiltrates into upper cover soil.
(unit: mm/year)

Precipitation

Precipitation

Runoff

1264

85

Evapotranspiration Infiltration in theupper cover soil

Upper cover soil

Evapotranspiration
Runoff

627

552

Infiltration to theupper cover soil＊1

Model 1-good
Model 1-poor
Model 2

Drainage

Groundwater level＊1＊2

Drainagelayer

Waste layer

Low-permeability
soil

Infiltration to the upper cover soil
= Precipitation – (Evapotranspiration + Runoff)
Infiltration to waste layer= Infiltration to the upper cover soil – Drainage

10000

Annual average amount of water
infiltration to the waste layer(mm/year)Geomembrane

Infiltration towaste layer＊2

Annual average amount ofwater infiltration to theupper cover soil
(552 mm/year)

1000

Reduced to
one-tenth

Reduced to
one-tenth

100

10

1
1E−9

Fig.8-11 Model of water flow through upper cover soilat the facility
Upper cover soil conditions:
Model 1:
Geomembrane over 25-cm-thick low-permeability soil layer
Model 2:50-cm-thick low-per meabilit y soil layers with nogeomembrane

Fig.8-12 Result of water infiltration in the waste layer
The water infiltration in the waste layer is adequately
reduced when the permeability of low-permeability soil
layer is low.

We are considering to dispose of very low-level radioactive
waste in the trench facility. The waste layer is placed above
the groundwater level; therefore, to evaluate the safety of the
facility, we must determine if the upper cover soil prevents
rainfall from infiltrating into the waste layer.
As shown in Fig.8-11, two models were evaluated, i.e.,
Model 1, in which drainage, geomembrane, and a lowpermeability soil layer were installed in the upper coversoil, and Model 2, in which a low-permeability soil layer
thicker than that of Model 1 was installed, but without ageomembrane. Specifically, after obtaining the rainfall
infiltration in the upper cover soil based on the weather
conditions of Japan, we determined for both the models theextent to which water infiltrates in the waste layer as a functionof the permeability coefficients of the low-permeability soil
layer.
A set of pseudo weather data was used in this determination.
Using a Fourier series and a gamma distribution, the pseudo
weather data were created from real weather data (i.e.,
precipitation, temperature, and solar radiation) observed
daily at a single measurement point in the Kanto area, Japan,
which has an annual rainfall of approximately 1200 mm.

For the geomembrane, we used the following parameters:a permeability coefficient of 1 × 10−11 m/sec (standard value),a constant deflection (poor) or slight deflection (good), andthe number of flaws per hectare. The permeability coefficientsof the low-permeability soil layer were set within the range
between that of sandy soil and clay. Based on these conditions,
we calculated the water infiltration in the upper cover soil andthe waste layer of the facility.
The results indicate an annual average water infiltration in theupper cover soil of approximately 550 mm for an average annual
rainfall of approximately 1200 mm (Table 8-1). Fig.8-12 showsthat the annual average water infiltration in the waste layer was
reduced to less than one-tenth of the water infiltration in theupper cover soil when the low-permeability soil layer has apermeability coefficient below 1 × 10−7 m/sec in Model 1 andbelow 1 × 10−9 m/sec in Model 2. In addition, we confirmed thatdrainage increases as the permeability coefficient of the lowpermeability soil layer decreases.
The study concludes that using a low-permeability soil
layer is effective for suppressing water infiltration in the waste
layer of the trench facility. Moreover, placing a geomembrane
suppresses water infiltration.

1E−8

1E−7

1E−6

1E−5

Permeability coefficients of
the low-permeability soil layer (m/sec)

Reference
Kurosawa, R., Sakai, A. et al., Evaluation of Infiltration Water through the Upper Cover Soil in Trench Type Disposal Facility for Low Level Radioactive
Wastes Generated from Research, Industrial and Medical Facilities, JAEA-Technology 2014-013, 2014, 89p. (in Japanese).
JAEA R&D Review 201499

Research and Development Relating to Nuclear Fuel Cycle Backend and Reprocessing of Spent Nuclear Fuel

8-5 Visualization of Geosynthesis Procedure− Geosynthesis Data Flow Diagram (Construction Phase) −Sampling and chemical
analysis of inflow water

Petrological and mineralogical data
(Rock texture / Mineralogicalcomposition, etc.)
Geochemical data
(Chemical / Isotopes composition, etc.)

Laboratory tests and investigations(Petrological and mineralogy
characterization / Age dating test,
Hydraulic test / PetrophysicalCharacterization /Rock mechanical characterization)

Petrophysical data
(Porosity / Density, etc.)

Sedimentary cover- Horizontal distribution- Stratigraphy and lithology- Mineralogical / Petrologicalcharacteristics

Spatial distribution and geometry of host rockSpatial distribution and geometry of transport
pathways(groundwater flowpaths)
Geological model(Regional scale)

Basement granite- Depth and geometry of
unconformity- Lithological and structural domain- Petrological / Mineralogicalcharacteristics

Long-term evolution of geology
Groundwater flow field

Hydrogeological model(Regional scale)

Groundwater flow simulation(Regional scale)

Groundwater flux
Long-term evolution of groundwater flow

Discontinuities (Fault / Joint)- 3D distribution and geometry- Petrological / Mineralogicalcharacteristics

Spatial distribution of groundwater salinity

Geological model(Local scale)

Spatial distribution of pH, redox condition

Change in groundwater chemistry

Groundwater table distribution
Core logging

Recharge rate (Infiltration)

Investigations of borehole wall
image

Surface hydrological data
(River flux / Precipitation /Evapotranspiration /Groundwater level, etc.)

Single-hole hydraulic test
Sampling and chemical analysis ofgroundwater

Deep hydraulic data
(Inflow points / Hydraulic conductivity /Transmissivity / Storage coefficient, etc.)

Long-term groundwater level /Pressure changes

Hydrochemical model(Local scale)

Basement granite- Hydraulic characteristics(Hydraulic property)- Flow porosity

Effect of colloid / Organics / Microbes on
nuclide transport / Retardation
Spatial distribution of EDZ

Geological model(Site scale)Hydrogeological / Mechanical property of EDZ
Geochemical property of EDZ

Discontinuities (Fault / Joint)- Hydraulic characteristics(Hydraulic property)- Flow porosity

Hydrogeological model(Site scale)

Groundwater flow simulation(Site scale)Hydrochemical simulation(Site scale)Hydrochemical model(Site scale)Rock mechanical model(Site scale)Stress / Excavation analysis(Site scale)Stress condition in EDZ
Spatial distribution of higher-permeability rocks,
aquifers and surface waters

Hydraulic head distribution
Cross-hole tomography survey
Amount of inflow water

Spatial variability of water fluxes in higherpermeability rocks, aquifers and surface waters

Hydrochemical data of groundwater(Chemical / Isotopic composition /Colloids, organic matters andmicrobes, etc.)

Surface tilt measurement

Geophysical survey
Surface hydrological investigations,
Meteorological observation
Measurement of water inflow

Pressure measurement

Hydrochemical measurement

Groundwater in sedimentary cover- Distribution of groundwater chemistry- Origin and residence time ofgroundwater

Geological model(Block scale)
Volume of inflow into underground tunnels

Sedimentary cover- Sorption / Diffusion characteristics- Geometrical structure of solutetransport field

Hydrogeological model(Block scale)

Groundwater flow simulation(Block scale)

Basement granite- Sorption / Diffusion characteristics- Geometry of pathways / Porosities- Pore structure characteristics

Hydrochemical simulation(Block scale)Hydrochemical model(Block scale)

Solute transport data
(Sorption / Diffusion coefficient /Pore structure, etc.)
Rock mechanical and thermal data
(Unconfined compression strength /Young's modulus / Thermal
conductivity / Geothermal gradient)

Sedimentary cover / Granite- Rock mechanical / Thermal
characteristics

Initial stress data

Initial stress state

Spatial variability of petrophysical /Geomechanical properties of rocks
Distribution of discontinuities intersecting
underground tunnels

Groundwater in granite- Distribution of groundwater chemistry- Origin and residence time ofgroundwater- Chemical characteristics of colloids /Organic matters / Microbes

Discontinuities (Fault / Joint)- Sorption characteristics- Depth of matrix diffusion- Pore structure characteristics

Local stress regime

Chemistry of inflow into underground tunnels

Spatial variability of geothermal gradient
Thermal rock properties

Solute transport
conceptual model
Effect on groundwater pressure
Rock mechanical model(Block scale)Stress / Excavation analysis(Block scale)Effect on groundwater chemistry
Effect on river water chemistryVibration and noise

Stress measurement

Items related to thecharacterization

Items not related tothe characterization

Link between items related
to the characterization

Confirmation of consistency

Links connected

Groundwater flow
condition
into the tunnel

Laboratory tests and investigations(Petrological and mineralogy
characterization / Age dating test,
Hydraulic test / Petrophysicalcharacterization /Rock mechanical characterization)

Hydrochemical data of surface water(Chemical / Isotopic composition, etc.)

Geomechanical /Hydraulic propertiesof tunnel near-fieldenvironment

Amount of inflow water
Cross-hole hydraulic test

In-situ rock mechanical tests

Monitoring

Hydrochemical simulation(Local scale)

Sorption capacity and diffusivity of rock matrix
and of transport pathways

Subsurface
thermalconditions

Borehole investigation

Fluid logging

Groundwater flow simulation(Local scale)

Geometry of transport pathways depth ofdiffusion-accessible rock matrix

Sedimentary cover- Hydraulic characteristics(Hydraulic property)- Flow porosity

Geophysical logging

Hydrogeological model(Local scale)

Environmental effect byshaft construction

Investigation in the gallery

Shaft wall observation

Data of geological structure(Lineament / Fault / Joint distribution)

Groundwaterflow
characteristics

Stratigraphical and lithological data
(Geological map and columns, etc.)

Geophysical survey

Geological structure

Specific aims / Issues to be understoodSize and geometry of host rock;heterogeneity within host rockGeochemical
characteristics
of groundwater

Conceptualization / Modeling / Simulation

Transport /Retardationof nuclides

Interpretation / Dataset

Geologicalenvironment
in EDZ

Raw / Conditioned data

Dilution of
nuclidesField Investigations

Links not connected

Fig.8-13 Example of geosynthesis data flow diagram (groundwater geochemistry: site scale)

We are pursuing a geoscience research and development
project called the Mizunami Underground ResearchLaboratory (MIU) project. The project studies a crystalline
rock environment to construct a scientific and technologicalbasis for the geological disposal of high-level radioactive
waste.
The MIU project is planned in three overlapping phases, i.e.,the surface-based investigation phase (phase I), the construction
phase (phase II), and the operation phase (phase III). Currently,
the project is in phases II and III.
In phase II, the adequacy of models of the geologicalenvironment established in phase I is evaluated by using
data accumulated during phase II. Based on the results ofthis evaluation, we determine whether various elemental
technologies adopted to characterize the geologicalenvironment in phase I are applicable and feasible. These

technologies include the planning, investigation, and modelingmethods. Furthermore, a series of procedures are organized toevaluate design, construction, and safety, and a geosynthesis
data flow diagram is established (Fig.8-13). This data flow
diagram integrates the data flow from investigation tomodeling and analysis. It combines in a rational way the itemsof the investigation that make the results reflect the safetyassessment and design.
In this sense, the geosynthesis data flow diagram provides
a rational framework, from “investigation” to “modeling”
and “analysis,” for achieving individual goals and tasks.
The experience and knowledge obtained by performing theinvestigation, modeling, and evaluation should be compiled
by using the geosynthesis data flow diagram.
This approach should prove useful for both implementing
and regulating high-level radioactive waste disposal.

Reference
Hama, K. et al., Mizunami Underground Research Laboratory Project-Preparation of Geosynthesis Data Flow Diagram (Construction Phase),
JAEA-Data/Code 2013-010, 2013, 58p. (in Japanese).

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Research and Development Relating to Nuclear Fuel Cycle Backend and Reprocessing of Spent Nuclear Fuel

8-6 Research in Onshore Mud Volcanism− Origin of Muddy Sand from Kamihoronobe Mud Volcano −4.0

1.0
Wakkanai Fm.
Masuporo Fm.
Haboro Fm.
Hakobuchi Fm.
Kamihoronobe
Mud Volcano

K2O (%)

3.0

2.0

TiO2 (%)(a)

0.5

Muddy sand sediment
1.0

0.0

0.0

↑
Groundwater and bubbling gas

0

(b)

5

10
Al2O3 (%)

15

20

20

40

60
SiO2 (%)

80

100

Fig.8-15 Bivariate plots of K2O–Al2O3 and SiO2–TiO2 relationships formuddy sand and core samples from the area of Fig.8-14(modified from Miyakawa, K. et al., G3, vol.14, no.12, 2013, p.4980-4988.)

Tempoku Drillhole

Fault
nobe
Horo

(c) Depth (GL. -m) Heulandite
Koetoi Fm.

N(d)QuartzOpal-CT

Wakkanai Fm.

Kamihoronobe
Mud Volcano (KMV)(N45˚2’3” E141˚56’50”)

Horonobe UndergroundResearch Laboratory0

Plagioclase

1000

Heulandite

Masuporo Fm.

Smectite

2000

Illite
ult
ri Fa
aga
Om

Magaribuchi Fm.

Pyrite3000

GypsumHaboro Fm.
Matsuo Anticline →Hakobuchi Fm.

55km
km

Yezo Fm.

Alluvium

Sarabetsu Fm.

Yuchi Fm.

Wakkanai Fm.

Masuporo Fm.

Soya coal-bearing Fm.

Koetoi Fm.

Anticline

Syncline

Fault

Calcite4000

Siderite
Chlorite

5000
0

10

20

0

2

4

6

8

10

XRD relative intensity

Concealed fault

Fault (or buried Fault) inferred from subsurface data

Fig.8-14 (a) Photograph of vent from which
muddy sand and gas was sampled. (b) Geologicalmap of the area of KMV(modified from Miyakawa, K. et al., G3, vol.14, no.12, 2013,
p.4980-4988.)

Fig.8-16 Abundances of minerals as a function of depth: (c) heulandite
from a report on Tempoku Drilling (JNOC, 1995) and (d) data from thepresent study(modified from Miyakawa, K. et al., G3, vol.14, no.12, 2013, p.4980-4988.)

XRD relative intensity, which is defined as the XRD intensity from a mineral in
the sample divided by the intensity of a quartz standard and multiplied by 100.

Recently, applied earth science research into onshore mud
volcanism has focused the safety of underground facilities,
disaster mitigation, and the greenhouse effect on the climate.
A topographical heightening created by a mud irruption alongwith high-pressurized groundwater and/or gas is called a mud
volcano (MV), and the genetic process of creating a MV iscalled mud volcanism or the MV phenomenon.
Our R&D program must assess the geosphere stability toisolate radioactive waste over the long term. The goals of thisR&D activity may be summarized in the following sequential
steps: (1) determine the relationship between anomalously
high formation pressure and mud volcanism, (2) narrow therange of geological settings where MV can occur in the future,and (3) clarify the impact of MV to the waste disposal system.
Because only two onshore MVs are known in Japan (i.e., theNiikappu MV in Hokkaido and the Matsudai MV in Niigata),
information from the case studies of other onshore MVs isneeded to completely characterize onshore MVs. In this study,
we investigated the origin of muddy sand sediments from the
Kamihoronobe MV (KMV) located in Hokkaido (Fig.8-14).
Fig.8-15 compares the SiO2, TiO2, K2O, and Al2O3 contentsof KMV samples with those of core samples from eachformation in this area, as described in the Tempoku Drilling(TD) (JNOC, 1995; Fig.8-14(b)). Fig.8-15 suggests that thechemical compositions of the KMV samples indicate a mixof materials from the Masuporo, Haboro, and Hakobuchi
formations. By analyzing with X-ray diffraction (XRD), we
compared the mineral composition of muddy sand samples
with that of core samples (see published TD report) (Fig.8-16).
Fig.8-16(c) shows that stratigraphically, heulandite, which
is a type of clay mineral, is detected only in the Hakobuchi
Formation, and Fig.8-16(d) shows that heulandite occurs in
the muddy sand sediments. Considering that the Masuporo
Formation is the youngest sedimentary formation below the
KMV, the depth of the Hakobuchi Formation was estimatedfrom the thickness of the formation to be 2200–2400 m(Fig.8-16(c)). Based on this evidence, we concluded thatmuddy sand from the KMV migrated from depths of over
2200–2400 m and was incorporated in the rock fragmentsfrom overlying formations.
Although we concluded that the materials came from great
depths, we still need to study how MVs are generated andtheir lifetime or cycle time.

Reference
Miyakawa, K., Tokiwa, T. et al., The Origin of Muddy Sand Sediments Associated with Mud Volcanism in the Horonobe Area of Northern Hokkaido, Japan,
Geochemistry Geophysics Geosystems, vol.14, no.12, 2013, p.4980-4988.
JAEA R&D Review 2014101

Research and Development Relating to Nuclear Fuel Cycle Backend and Reprocessing of Spent Nuclear Fuel

8-7 Investigation of Damage to Rock Mass in Gallery Excavation− Based on Seismic Velocity and Fracture Distribution around Gallery Wall −(a) Picture of gallery wall(b) Fracture mapping of gallery wall
Water
Nodule (Hard)

N-017-14 (N38E68W)
Interval of the fracture
3–5 cmNodule (Hard)
Nodule (Hard)

N-017-13 (N26E72W)

Nodule (Hard)

N-017-12 (N18E68W)

Nodule (Hard)

Interval of the fracture
1–3 cm

N-017-07 (N2E80W)

Water

Interval of the fracture
1–3 cmInterval of the fracture
3–5 cmVelocity of rock mass measured
before excavation (1.67 km/s)

Water

Nodule (Hard)

N-017-11 (N22E78W)
Interval of the fracture
2–5 cmNodule (Hard)

1.8

Water

N-017-06 (N2W86W)
Water

Interval of the fracture
1–5 cmCheck boring

Nodule (Hard)

N-017-10 (N2W82W)
Water
Gas

Nodule (Hard)

1.7

Water
Gas

Water

Water
Interval of the fracture
3–10 cm

Water
Nodule (Hard)

N-017-08 (N10E22E)

Water

N-017-05 (N4W26E)
Interval of the fracture
3–10 cm

Interval of the fracture
3–10 cm

Fig.8-17 Example of fractures induced by gallery excavation
Gray lines in (b) indicate fractures induced by gallery excavation.

(c) Pre-excavation(d) After the excavationof the survey area

(Rock) (Gallery)

(m) 4

3

2

1

1.1

0

(m) 4

1.3

1.5

3

2

1.7

1

1.9

0

2.1

(m) 4

(Rock) (Gallery)

3

2

1

1.6
1.5
1.4
1.3
1.2
1.1

(e) 64 days after the
pre-excavation

Driving direction(Rock) (Gallery)

Seismic velocity (km/s)

N-017-09 (N38E40E)
Nodule (Hard)0

Seismic source
Receiver

1.0
0.0

0.5

1.0

1.5

2.0

Density of fractures (1/m)

Fig.8-19 Relationship between seismic velocity anddensity of fracture
The circles in the graph represent the relationshipbetween seismic velocity of each ray path and fracture
density calculated based on the number of fracturesintersected by each ray path.

Seismic velocity (km/s)

Fig.8-18 Results of seismic tomography survey
Velocity contour map of rock mass around gallery wall.

In the construction of a deep underground gallery, fractures
are induced around the gallery wall because of stress redistribution
in response to excavation, as shown in Fig.8-17. In addition,
because of these fractures, the hydraulic conductivity of
the rock mass increases. Such a zone is called an excavation
damaged zone (EDZ). To dispose of high-level radioactive
waste, evaluation of an EDZ is important for safety assessment.
The Horonobe Underground Research Laboratory is
conducting a seismic tomography survey at the depth of eachgallery (140 m, 250 m, and 350 m) to investigate the extent
of an EDZ. Fig.8-18 shows how seismic velocity changes in
response to excavation. These results were obtained from aseismic tomography survey performed in the 250 m gallery.
From the results shown in Fig.8-18(c), the seismic velocityof the pre-excavation rock mass ranges from 1.6 to 1.8 km/s.
From the results shown in Figs.8-18(d) and (e), the layer inwhich the seismic velocity decreased remarkably (<1.4 km/s)
extended approximately 1.0 m in the gallery wall. To investigate

the mechanism producing the variations in seismic velocity,
we obtained the number of fractures intersected by the raypath of a seismic wave based on the result of the fracture
mapping of the gallery wall. Next, the fracture density foreach ray path was calculated by dividing the number offractures by the length of each ray path. Fig.8-19 shows
the relationship between seismic velocity and fracture
density. The figure shows that seismic velocity decreases
almost linearly as fracture density increases. Thus, we
expect that fracture density around the gallery wall could be
estimated from the seismic velocity around the gallery wall.
Furthermore, hydraulic tests reveal that enhanced hydraulicconductivity extends approximately 1.2 m into the gallery
wall. Consequently, the EDZ in the 250 m gallery is estimatedto extend approximately 1.2 m in the gallery wall. In the near
future, we will construct a conceptual model of an EDZ basedon the results of in situ tests and observations made in the
350 m gallery and in vertical shafts.

Reference
Aoyagi, K. et al., Quantitative Assessment of an Excavation Damaged Zone from Variations in Seismic Velocity and Fracture Distribution around a Gallery
in the Horonobe Underground Research Laboratory, Rock Engineering and Rock Mechanics; Structures in and on Rock Masses, p.487-492.

102

JAEA R&D Review 2014

Research and Development Relating to Nuclear Fuel Cycle Backend and Reprocessing of Spent Nuclear Fuel

8-8 Detection of Unrecognized Volcanic Activity by Sand Grain− Developing Method based on Mineral and Chemical Analysis of Sand Grain in Sedimentary Rock −Mizunami Group

Columnar section
depth

0m

Bulk mineralcomposition(all heavy minerals)

35°N

134°E

135°E

Sedimentary rock contemporaneouswith Mizunami Group

136°E

137°E

Volcanic rock after depositing of
Mizunami Group

Fig.8-20 Distribution of sedimentary rock contemporaneouswith Mizunami Group and volcanic rock after depositing
Mizunami Group in Southwest Japan
Mizunami Group was deposited contemporaneously with the
opening of the Japan Sea. Contemporaneous sedimentaryrock, distributed around the Seto Inland Sea and from theSan'in district to the Chubu District, does not contain volcanic
rock. Large-scale volcanic activity is not inferred in the area.

Togari Member

pyroxene only)

20 m
30 m
40 m

Tsukiyoshi
MemberAkeyo Formation
Hongo Fm.

133°E

10 m

50 m
60 m
70 m
80 m
90 m

Toki Lignite-bearing Formation

34°N

Histograms ofrefractive index(biotite-amphibolite- of plagioclase

Heavy mineralcomposition

100 m
110 m
120 m
130 m
140 m
150 m

The Mizunami Group was deposited around the Mizunami
Underground Research Laboratory between 20 and 15 million
years ago. The group was deposited contemporaneously withthe back-arc spreading event that formed the present-day
Japan Sea and that involved extensive volcanic activity along
the coast of the Japan Sea. However, in Southwest Japan,
no volcanic activity seemed to have occurred during thedeposition of the Mizunami Group (Fig.8-20).
When we studied the depositional age of the Mizunami Group,
we found volcanic ash dispersed in the group. Such information,
however, was obtained within a limited horizon, so we collectedsandstone samples over a wider stratigraphic horizon andexamined them to see if they contained volcanic ash.
In general, volcanic ash is composed of minerals
and volcanic glass, but in older strata, volcanic ash isdiagenetically altered to diagenetic minerals such as zeolite.
We tried to find minerals that originated from volcanic ash.
The Mizunami Group normally consists of sand and clay
derived from basement granite and older sedimentary rock, so
we tried to identify heavy minerals and plagioclase derived
from other rock types.

1.550

1.560

1.570

labradorite

bytwoniteanorthite

0Lithology

20 40 60 80 100 0

Minerals

20 40 60 80 100 0

20 40 60 80 100(%)

Heavy minerals

conglomerate

light mineralpyroxenesandstone

heavy mineralamphibolemudstone

volcanic glass

biotitetuff

rock fragment

othersligniteothers

1.580

1.540

1.530

andesine

170 m

Fig.8-21 Columnar sections of the studied drill core of the
Mizunami Group with bulk and heavy mineral compositions
and refractive indices of plagioclase
Subspecies of plagioclase can be identified based on refractive
index. Large amounts of volcanic glass are contained in theupper part of the group, and zeolite, which is an alteration
product of volcanic glass, is contained in the lower part. Thus,
the Mizunami Group contains large amounts of volcanic glass.

albiteoligoclase

160 m

fault

The investigation revealed that sandstone contains volcanic
glass and/or alteration products and that the sandstonemineralogically divides into three types: (1) characterized
by the presence of biotite and variations in plagioclase fromalbite to oligoclase, (2) characterized by the dominanceof amphibole and labradorite, and (3) characterized by thedominance of pyroxene and andesine (Fig.8-21). Because
basement granite contains biotite and albite, type 1 sandstone
derived from granite and types 2 and 3 derived fromvolcanic ash. Volcanic activity, which supplied volcanic ash
to the Mizunami area, is divided into two phases based on
variations in mineralogical and chemical compositions. Thesemineralogical and geochemical compositions suggest that the
volcanic activity evolved over time.
The volcanic ash does not derive from volcanic activity
along the coast of the Japan Sea. This study indicates that
unrecognized volcanic activity occurred in and aroundthe Mizunami area, even though the origin of the ash is
unclear. This study reveals that unknown volcanic activity
contemporaneous with Japan Sea formation occurred from
central to western part of the Japanese Island.

Reference

Sasao, E., Petrographic Study of the Miocene Mizunami Group, Central Japan; Detection of Unrecognized Volcanic Activity in the Setouchi Province,
Island Arc, vol.22, issue 2, 2013, p.170‑184.
JAEA R&D Review 2014103

Research and Development Relating to Nuclear Fuel Cycle Backend and Reprocessing of Spent Nuclear Fuel

8-9 Large Inland Earthquake Triggered by Latent Magmatism− Helium Isotopes as a Tool for Detecting Concealed Active Faults −35.6°N

Helium isotope ratio
R/RA＜0.5
0.5≦R/RA＜1.0
1.0≦R/RA＜2.0
2.0≦R/RA＜4.0
4.0≦R/RA

37.5°N

Helium isotope ratio
R/RA≦1.0
1.0＜R/RA≦2.0
2.0＜R/RA≦3.8
3.8＜R/RA≦5.4
5.4＜R/RA

37.0°N

Quaternary volcanoes
Active faults

A’
35.4°N

Mainshock
Aftershocks
Inferred fault
segments
35.2°N

36.5°N

36.0°N

10 km

A

35.0°N132.8°E

133.0°E

133.2°E

133.4°E

133.6°E

Fig.8-23 Geographical distribution of helium isotope ratios
in aftershock area of 2000 western Tottori earthquake
The helium isotope ratios in the aftershock area decrease with
distance from the inferred faults.

35.0°N

34.5°N
137.0°E

137.5°E

138.0°E

138.5°E

139.0°E

Fig.8-22 Geographical distribution of helium isotope
ratios around the Itoigawa-Shizuoka tectonic line
The helium isotope ratios (R = 3He/4He) from hot springand drinking wells along the Itoigawa-Shizuoka tectonic
line are larger than the atmospheric value (RA).

Depth (km)

0

A

A’
Upwelling
mantle helium

10

1000
Source fault of mainshock20
Latent magma and fluids(Conductor)

30
40

10000

0

20

40
60
Distance (km)

100

80

Resistivity (Ωm)

35.5°N10

Fig.8-24 Electrical resistivity structure beneath aftershockarea of 2000 western Tottori earthquake
A prominent conductor is visible on the southwestern side of
the source fault. The conductor is attributed to latent magma
and related fluids.

Earthquake faulting is one of the most important factors
affecting the long-term stability of a geological disposal
system, because it could cause a dynamic destruction of
basement rocks. Recently, some large earthquakes with nonoticeable surface expression of earthquake faulting haveoccurred, implying the existence of concealed active faults.
Therefore, we developed a geochemical approach based onhelium isotopes to detect concealed active faults.
The helium isotope ratios (3He/4He) observed in volcanic
regions are usually similar to those of mantle-derived helium,
which is several times higher than atmospheric value. These
ratios should allow us to determine migration pathways, suchas active faults, from the subcrustal mantle. We mapped the
distribution of the helium isotope ratio around the active
faults of the Itoigawa-Shizuoka tectonic line (ISTL) and of
the seismic source region of the western Tottori earthquake byusing helium isotope data from groundwater samples from hot
springs and drinking water wells.
The spatial variation of the helium isotope ratio along

the ISTL revealed that there is a high helium isotope ratiothat occurs along the active faults (Fig.8-22), which can beinterpreted as the leakage of mantle helium within the crust
migrating upward through active faults.
An inland earthquake (M 7.3) occurred in the western Tottoriin 2000 where no active fault was related to the earthquake.
Groundwater samples collected in the aftershock area arecharacterized by helium isotope ratios several times higher than
the atmospheric value. The helium isotope ratios decrease with
distance from the inferred fault segments (Fig.8-23). In addition,
magnetotelluric soundings were taken in the aftershock area
in order to image subsurface electrical resistivity structureand detected an anomalous conductive body, which may
be attributed to latent magma and related fluids, on thesouthwestern side of the source fault (Fig.8-24). These findings
indicate mantle helium leaking from latent magma through theconcealed faults. Moreover, the upwelling of fluids to the crust
could contribute to triggering a large earthquake by weakening
the shear strength of the source fault.

Reference
Umeda, K., Asamori, K. et al., Release of Mantle and Crustal Helium from a Fault Following an Inland Earthquake, Applied Geochemistry, vol.37, 2013,
p.134-141.

104

JAEA R&D Review 2014

Research and Development Relating to Nuclear Fuel Cycle Backend and Reprocessing of Spent Nuclear Fuel

8-10 Prediction of Radionuclide Speciation in Groundwater− Determination of Standard Thermodynamic Data of Se(VI)/(IV) Couple −50

30

Current (μA)

20
10
0
−10
−20

−40
−50
0.4

half-wave
potential (E1/2)
0.6
0.8
1.0
1.2
1.4
Potential reference to thestandard hydrogen electrode (V)Standard potential reference to thestandard hydrogen electrode (mV)

1060

40

−30

slope=ΔS/2F=−0.3±0.1 mV/K
ΔS: Molar entropy (J K−1 mol−1)
F: Faraday constant (C mol−1)

1050
1040
1030
1020
1010
1000
990
980
273

283

293
303
313
Temperature (K)323

Fig.8-25 Cyclic voltammogram of Se(VI)/(IV) couple
The growth of the redox wave indicates the increase in theactive electrode area where electron transfer is possible dueto the removal of the passivating layer by repetitive potential
cycling. The anodic and cathodic peaks correspond to theoxidation of Se(IV) and the reduction of Se(VI), respectively.

Fig.8-26 Molar entropy of Se(VI)/(IV) couple−
HSeO 4 /H 2SeO 3 standard redox potential as a function oftemperature. The solid line is a weighted linear regression of
the data. ΔS is derived from the slope of this line.

We assess the safety of the geological disposal based on theassumption that the radionuclides contained in vitrified waste
are transported to the biosphere by flowing groundwater. We
obtained the necessary data for this assessment.
Selenium-79(79Se) is a long-lived fission product. Se species
vary, depending on the oxidation/reduction potential E.
Radionuclides are transported through the buffer material andhost rocks and are retarded by sorption onto them. Because
each species has unique sorption characteristics, the migrationof Se depends strongly on E. Therefore, we accounted forthe oxidation of the original reducing groundwater due to adecrease in the depth of repositories and/or the presence of anoxidizing agent, such as H2O2 that radiolysis of groundwater
generates. The oxidation state of Se is Se(IV) or Se(VI) in
oxidized groundwater. Because the sorption characteristics
of Se(IV) differ from those of Se(VI), the accurate predictionof the ratio Se(VI) to Se(IV) is crucial for understanding
the migration of Se. The standard redox potential E 0 of theSe(VI)/(IV) couple determines Se(VI) to Se(IV) ratios at
298.15 K in calculating the Se speciation. This potential isdifficult to determine because the equilibrium of the Se(VI)/(IV)couple, whose reaction rate is slow, is not rapidly established,

and the experimentally determined E is not affected by theSe(VI) to Se(IV) ratio in solution. We therefore chose to use
cyclic voltammetry CV, which is a common method used tomeasure the half-wave potential E1/2. The potential E1/2 is where
the concentration of an oxidant equals that of a reductant.
Even for the redox reaction with a slow rate, CV allows us
to measure E1/2. Fig.8-25 shows the cyclic voltammogramof the Se(VI)/(IV) couple, where E1/2 is midway between
the anodic and cathodic peak potentials. The value of E 0
is derived from the concentration dependence of E1/2.
In the geological environment, temperatures exceed 298 K.
Because the molar entropy of the reaction ΔS yields thetemperature derivatives of E 0, ΔS of the Se(VI)/(IV) couple
is also required to predict the Se(VI) to Se(IV) ratio at atemperature above 298 K. We therefore obtained ΔS for thiscouple from the temperature dependence of E 0 (Fig.8-26).
The results for E 0 and ΔS, which agree with publised values ofthe molar enthalpy of the Se(VI)/(IV) couple, will contributeto a more reliable prediction of Se speciation under various
environments and will be input in the JAEA thermodynamic
database.

References
Doi, R., Determination of the Selenium (VI)/(IV) Standard Redox Potential by Cyclic Voltammetry, Journal of Nuclear Science and Technology, vol.51, issue 1,
2014, p.56-63.
Doi, R., Molar Entropy of the Selenium (VI)/(IV) Couple Obtained by Cyclic Voltammetry, Journal of Nuclear Science and Technology, vol.51, issue 3,
2014, p.359-368.
JAEA R&D Review 2014105

Research and Development Relating to Nuclear Fuel Cycle Backend and Reprocessing of Spent Nuclear Fuel

8-11 Prediction of Long-Term Alteration Characteristics of HFSC
− Clarification of Decreasing pH and Development of Alteration Model −pH

13

OPC (Provided in literature)
HFSC with low SO3 (Experiment)
HFSC with high SO3 (Experiment)

12

11

10

1

100

Fig.8-27 Experimental pH value in dissolution experiments
Hydrates made from HFSC containing different amount of SO3
were immersed in ion-exchanged water with differing liquid-tosolid ratios. These pH values reflect a possible scenario for pH
evolution in an underground repository.

1000

liquid/solid ratio (dm3/kg)(b) Same model as in panel (a), but includes alkali
adsorption phenomena(a) Model considers only dissolution and/or
precipitation phenomena

13

pH

pH

13

12

Calc. (HFSC with low SO3)
Calc. (HFSC with high SO3)
Exp. (HFSC with low SO3)
Exp. (HFSC with high SO3)

12

11

11
1

10

100

1000

1

liquid/solid ratio (dm3/kg)

10

100

1000

liquid/solid ratio (dm3/kg)

Fig.8-28 Comparison between experimental and calculated pH
Differences between experimental and calculated values (a) can be improved by including alkali adsorption in the model (b).

Cementitious materials are expected to be used to construct
an underground repository for the geological disposal ofradioactive wastes. Degradation of ordinary Portland cement(OPC) by the ingress of groundwater can generate a high-pH
plume (≥ 12.5), which would probably affect the performanceof the bentonite buffer used to contain the radioactive wastes
and the transport properties of the host rock. Therefore, a new
type of high-volume fly ash silica fume cement (HFSC) is
being developed that will lead to a lower plume than for OPC.
The long-term alteration characteristics of HFSC need to bepredicted to assist in demonstrating the safety of a geologicaldisposal system. The prediction should be made based on anunderstanding of the mechanisms responsible for alteration
phenomena. In this study, dissolution experiments wereused with geochemical equilibrium calculations to make apreliminary assessment of the pH-lowering mechanism in
HFSC and to predict the long-term alteration characteristics.
These calculations, based on the understanding that pH
decreases only when Ca(OH)2 is absent and the molar Ca/Si
ratio in the calcium silicate hydrate (C–S–H) gel is lowered,
did not satisfactorily reproduce the experimental results.
Dissolution experiments were performed with HFSC
containing different amount of SO3, with the focus being
on SO42− as a counter ion and on the response of the alkali

elements (i.e., Na and K). HFSC containing either highor low SO3 wt% was immersed in ion-exchanged water atvarious liquid-to-solid ratios, and the resulting pH and liquidcomposition were measured. At lower liquid-to-solid ratios(≤ 10), the pH was higher for HFSC with low SO3 than for
HFSC with high SO3 (Fig.8-27). However, the pH calculatedwith models that only consider dissolution and/or precipitation
phenomena differs significantly from the experimental valuesfor HFSC with low SO3 (Fig.8-28(a)) and so could not be
explained by the SO3 content alone.
An improved chemical equilibrium model was developed
by including Na and K adsorption on the C–S–H gel alongwith the dissolution and/or precipitation phenomena.
The pH calculated with the improved model agrees wellwith the experimental results (Fig.8-28(b)). The model thus
developed is expected to help improve our understanding of
the mechanisms responsible for lowering pH in HFSC andcan be used to predict possible scenarios of pH evolution in ageological disposal system.
Part of this study was sponsored by the Ministry of
Economy, Trade and Industry of Japan (METI) through theprogram “Advanced Assessment Technology Development for
Impact of Cement on Geological Disposal System for Longlived Radioactive Waste” in FY 2011 and 2012.

Reference
Hoshino, S., Honda, A. et al., Mechanism of Alkalinity Lowering and Chemical Equilibrium Model of High Fly Ash Silica Fume Cement, Konkurito Kogaku
Ronbunshu (Concrete Research and Technology), vol.25, 2014, p.97-107 (in Japanese).

106

JAEA R&D Review 2014

Research and Development Relating to Nuclear Fuel Cycle Backend and Reprocessing of Spent Nuclear Fuel

8-12 Evaluation of Long-Term Geological Evolution− Study of the Safety of Geological Disposal Considering Uplift and Erosion −Weatheredzone

Time after closure■ Change in mean altitude■ Topographic change■ Decrease in repository depth

AltitudeTime of closure

Topography(Dispersion of altitude)

Downstream(Uplift / erosion rate: 0.1 mm/year)
Middle course(Uplift / erosion rate: 0.5 mm/year)
Upstream(Uplift / erosion rate: 1.0 mm/year)

Mean altitude

ErosionRepository
1.0E+04

Frequency

UpliftRepository

Fig.8-29 Conceptual uplift erosion model for repositoryand evolution of topography over time
Uplift, erosion, and changes in topography are evaluated using
a landscape evolution model, which helps in estimating the timerequired for a repository to reach the surface of the ground andthe number of waste packages eroded as a function of time.

Dose (μSv/year)

IntactRock

300 μSv/year

1.0E+03■ Number of affected canisters■ Geological conditions and their temporalchanges (e.g., hydrological andgeochemical features)＊1

1.0E+02

1.0E+01

1.0E+00

1.0E−01

1.0E−02

−11.0

−10.5

−10.0

−9.5

−9.0

−8.5

−8.0

−7.5

−7.0

Logarithmic mean of transmissivity distribution＊2 (log(m2/s))【Middle course】
Average uplift and erosion rate
Unequal erosion
Sl＊

1 The maximum permitted dose has not yet been decided in Japan and so thedose = 300 μSv/year is adopted from Publication 81 by the ICRP.
＊
2 The transmissivity (hydraulic conductivity × aperture) distribution expresses acharacteristic of the hydrogeologic environment in view of its hydraulic heterogeneity.

ion

op

er

ss

e
Width of verticallyeceroded areaern 50-100 m Slop

ec

es

sio

LateralerosionDeepening

Lateralerosion【Downstream】
Relatively low uplift anderosion rate
Uniform erosion
Width of horizontallyeroded area
≈2000 m【Upstream】
Relatively high uplift anderosion rate
Linear erosion (V-shaped)
Slope

Slope

Valley depth
100-700 m

Lateral erosion
Lateral erosion
Deepening

Fig.8-31 Results of safety assessment assuming that
repository reaches ground surface by uplift and erosion
The effect on humans due to repository is evaluated byconsidering the uncertainty of transmissivity to examine
the various conditions of the geological environment. The
result shows that the effect is less than the targeted criterion
suggested by the international organization even if therepository reaches the ground surface.

Valley width200-1200 m

Fig.8-30 Conceptual erosion model

Japan is tectonically active, so a site for the geologicaldisposal of high-level radioactive waste (HLW) must be
demonstrated to be stable over the long term. Areas identified
by geological investigation that are more likely to experiencecatastrophic natural events should be avoided (e.g., areas withmajor fault movements or volcanic activity). More widespread
phenomena identified on regional and global scales, however,are more difficult to avoid (e.g., uplift or erosion and changes
in sea level). For uplift and erosion, the distance between a
300-m-deep repository and the biosphere can be reduced, andin the worst case, the repository could become exposed at
the ground surface. The present study provides a method toevaluate the effects of uplift and erosion on the safety of thegeological disposal of HLW.
In previous safety assessments, the rate of erosion was
assumed equal to the rate of uplift, so the repository would
approach the ground surface at a constant velocity. Afterreaching the ground surface, the repository was then assumed
to erode uniformly. In reality, there are many areas withdifferent rates of uplift and erosion. A systematic method to
identify and express the parameters for safety assessments
has been developed by using more realistic rates of uplift anderosion (Fig.8-29).
Fluvial erosion is the main erosion process in Japan; so toevaluate the subaerial exposure of a repository, the present
study models erosion depending on the precise locations
along a river’s course (Fig.8-30). The effect on human healthis evaluated based on the assumption that radionuclides are
released in the biosphere by erosion. In all cases considered,the dose rates are below the target value suggested by theInternational Commission on Radiation Protection (ICRP)(Fig.8-31). These results can be explained by the low amountof erosion because of the localized upstream deepening, thelonger downstream decay time due to the slow rate of upliftand erosion, even considering the large amount of erosion due
to uniform deepening, and the average of these two conditionsin the middle course.

Reference
Wakasugi, K. et al., Bounding Analysis of Uplift and Erosion Scenario for an HLW Repository, Proceedings of 21st International Conference on Nuclear
Engineering (ICONE 21), Chengdu, China, 2013, ICONE21-16724, 9p., in DVD-ROM.
JAEA R&D Review 2014107

Research and Development Relating to Nuclear Fuel Cycle Backend and Reprocessing of Spent Nuclear Fuel

8-13 Improvement of Ruthenium Removal Rate by Electrochemical Oxidation− Improvement of Stable Operation for Vitrification Process −Potentiostat

A

B

D

C

E

1.2
Initial Ru Concentration; 3000 mg/ℓ

Water (7 ℃)
Inlet
Integrating Flowmeter

Reference
Electrode(Ag/AgCl)

Thermocouple
Diaphragm

Absorbing Liquid
6M-HCl
(C2H5OH: 1%)

1.5 V(vs Ag/AgCl)
Counter
Electrode(Pt)

Working
Electrode(Pt)

Heater(Room Temperature to 95 ℃)

Ratio of Ru concentration

Salt Bridge

1.0

0.8

0.6

0.4

0.2

Fig.8-32 Schematic diagram of oxidation cell

6 h electrolysis for conditions A and C

0

Table 8-2 Experimental conditions for electrochemical
oxidation

Temperature Ce concentration Reflux/
Evaporation(℃)(mg/ℓ)

Diaphragm

A
B
C
D

95
50
95
95

0
0
3.0×1030

Reflux
Reflux
RefluxEvaporation

○
○
○
○

E

95

0

Reflux×

One of the platinum group elements, i.e., ruthenium (Ru), is amajor fission product in high-level liquid waste (HLLW), whichis to be disposed as vitrified waste. Platinum group elements
precipitate in a glass fusing furnace. These precipitates influence
the flow characteristics and electrical conductivity of a melting
glass, which complicates the vitrification process. To stabilize
the process, these platinum group elements should be removed
from the solution before feeding it in the glass fusing furnace.
Ru tetraoxide (RuO4) has a high vapor pressure, which is
sufficiently high to run off from the solution even at room
temperature. If Ru ions in the solution could be oxidized totetraoxide, Ru could be removed from the liquid phase.
We studied an electrochemical oxidation method to oxidize
Ru ions in solution to separate Ru as tetraoxide (Fig.8-32). The
advantage of this method is that it does not require additional
oxidizing or extraction reagents. Electrical oxidation,
however, is necessary for long oxidation times. To increase theoxidation rate, we conducted experiments to clarify how thefollowing four fundamental conditions (Table 8-2) influenceelectrochemical oxidation. Results of experiments involving adecreasing rate of Ru concentration are shown in Fig.8-33.
(1) Electrolyte temperature
When the electrolyte reaches a high temperature, Ru ions
easily migrate in the solution, so the oxidation rate can beincreased.
(2) Presence of promoter elements
Cerium ions are reported to promote the Ru oxidation rate.

0

1

2

3

4

5

6

7

Time (h)Fig.8-33 Decreasing ratio of Ru concentration, [Ru]t/[Ru]0,
with time
Experimental results showing decreasing rate of Ru concentration
ratio, i.e., [Ru]t/[Ru]0. Electrical oxidation under the conditions A
and C are appropriate for fast oxidation.

We confirmed that Ru migrates to the gas phase effectively withcerium. Cerium ions are electrically oxidized to tetravalent ions,
i.e., Ce(IV), which stabilize oxidized octavalent Ru, preventing
a back reaction, in the solution. Therefore, the removal of RuO4
to gas phase becomes more efficient.
(3) Evaporation or reflux of condensed phase
We were concerned about the possible absorption of gaseous
tetraoxide Ru into the condensed water in the condenser
and its return in the water to the electrolyte cell as reduced
Ru dioxide. But the evaporation or reflux conditions do notdirectly affect the electrochemical oxidation efficiency.
(4) Diaphragm for counter electrode
Simple electrical oxidation system that is easy to maintainshould be designed considering the HLLW treatment. If thediaphragm at the counter electrode is not necessary, it can beeliminated to simplify maintenance. However, the diaphragmis necessary for efficient oxidation.
We obtained the following conditions for fast electrochemical
oxidation: (1) higher temperature (95 °C), (2) Ce coexistence
at 3000 mg/ℓ, and (3) use of a diaphragm for counterelectrode. The oxidation rate achieved should be three times
faster under these optimized conditions.
The present study was partly supported by “Developmentof High-level Liquid Waste Conditioning Technology for
Advanced Nuclear Fuel Cycle” from the Ministry of Education,
Culture, Sports, Science and Technology of Japan (MEXT).

Reference
Sato, S. et al., Optimization for Removal of Ruthenium from Nitric Acid Solution by Volatilizing with Electrochemical Oxidation, Journal of Nuclear Scienceand Technology, vol.49, issue 2, 2012, p.182-188.

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９ Nuclear Fusion Research and Development
Toward the Practical Use of Fusion Energy
Fusion Plasma Research
Toward ITER and DEMO
○Develop Plasma
Performance for DEMO
○Support ITER
○Cultivate Human
Resources
Large Tokamak Device JT-60
・Exp. was completed・Upgrade to JT-60SA started

DEMO
ReactorFusion Frontier Research

ITER Project
Domestic Agency for ITER Project
Demonstration of Fusion Energy
・Cadarache
in France
・Construction
Started in
December 2007

JT-60

Develop Technological
Basis for DEMO
○R&D of Breeding Blanket○DEMO Design, R&D○Simulation of Burning
Plasma, Theoretical
Model○Remote Experiment○Engineering Validationand Engineering Design
Activities for IFMIF
International Fusion Energy
Research Center

BlanketJT-60SA
JT-60SA

Broader Approach (BA) Activities
Implementing Agency for BA Activities

Satellite Tokamak

Naka Fusion InstituteSupport ITER, Accelerate Early Realization of DEMO
・International Fusion Energy Research Center・Design Activity for Material Irradiation Facility
・Satellite Tokamak
Activities Started In June 2007

Design Activity for IFMIF

Rokkasho
Fusion InstituteFig.9-1 Steps involved in the development of the fusion DEMO reactor
Fusion plasma research and fusion frontier research are being pursued for the development of a DEMO reactor, aimed at an early
realization of fusion energy.

Crucial developments toward the practical use of fusionenergy in the fields of fusion plasma research and fusion frontierresearch are being pursued through intensive international
cooperation. The International Thermonuclear Experimental
Reactor (ITER) project, Broader Approach (BA) activities, andother collaborations (Fig.9-1) are aimed at the early realizationof a fusion demonstration power plant (DEMO) reactor. Fusion
plasma research is mainly being advanced by the Naka Fusion
Institute, and fusion frontier research is mainly being advancedby the Rokkasho Fusion Institute.
ITER Project
The ITER project is an international cooperative project
to demonstrate the scientific and technological feasibility offusion energy through the construction and operation of anexperimental reactor. The ITER agreement came into force in
October 2007, and the Japan Atomic Energy Agency (JAEA)
was designated as the domestic agency fronting the ITER
project in Japan. JAEA has proceeded with the preparation of
the equipment that Japan agreed to provide, and has achieved
various results in terms of technological development(Topics 9-1 and 9-2). In particular, JAEA began producing
a superconducting coil for the actual machine ahead of any
other country participating in ITER.
BA Activities
The Broader Approach (BA) activities are joint projects by
Japan and the European Union to conduct supporting research
for ITER and research and development (R&D) for a DEMO
reactor, which is the next step of ITER, aimed at the early
realization of fusion energy. The BA agreement came into

force in June 2007, and JAEA was designated as the agency incharge of implementing the BA activities in Japan.
The BA activities comprise three components: projects at
the International Fusion Energy Research Center (IFERC), theInternational Fusion Materials Irradiation Facility/Engineering
Validation and Engineering Design Activities (IFMIF/EVEDA) project, and the Satellite Tokamak Program (STP).
Topic 9-3 describes the results achieved in the R&D activities
toward a DEMO reactor at IFERC. Topic 9-4 describes theresults obtained by IFMIF/EVEDA. Topics 9-5 and 9-6 outline
the progress toward the construction of the Japan Torus-60
Super Advanced (JT-60SA) fusion reactor by STP. Topic 9-7
also presents results related to JT-60SA.
Fusion Plasma Research
Analysis of JT-60 experimental data was promoted, andinter-machine experiments were conducted for the purposeof achieving high economic efficiency in a fusion reactor
by means of attaining a high plasma pressure. Topic 9-8 is a
result that indicates the guidelines for overcoming instabilities
that hinder the formation of a high plasma pressure.
Fusion Frontier Research
Various R&D activities have been executed at the Rokkasho
Fusion Institute, aimed at the construction of the technologicalbasis for a DEMO reactor. Topic 9-9 describes theoreticalsimulation research that uses the high-performance computerat IFERC; the result can provide a basis for developing asecure shutdown procedure for a DEMO reactor. Topics 9-10
and 9-11 are results that are indispensable for the developmentof a breeding blanket.
JAEA R&D Review 2014109

Nuclear Fusion Research and Development

9-1 Lifetime Extension of Beam Dump by Dispersing the Deposition of Laser Energy− A High Laser Energy-Resistant Beam Dump for Thomson Scattering in ITER −(a) Conventional beam dump

Beam energy concentration(Deposition: 0.2 J/cm2)

(c)
0.016
Deposition (J/cm2)0

Laser
Return

Laser

Laser

(b) New beam dump(d)Beam energy (J)
Laser

5
1
Gradually attenuate

Beam is effectively absorbed

Laser

0.1
0

20

40

60

80

100

Depth (mm)Fig.9-2 Beam dump structures(a) The conventional beam dump has a relatively small areafor beam absorption, which induces damage; (b) the newbeam dump has a number of sheets aligned parallel to one
another which can gradually absorb the beam energy.

Fig.9-3 Beam energy deposition profile(c) Beam energy deposition profile on two of the sheets; (d) to
use the limited depth effectively, beam energy attenuation of
the latter sections is larger than that of the front sections.

When a laser beam is injected into plasma, a very small
fraction of the light is scattered by electrons in the plasma.
This phenomenon is called Thomson scattering. The electron
temperature and density of the ITER plasma are measured
from the spectrum and intensity of the scattered light,
respectively. Since the proportion of incident photons that are
scattered and detected as a signal in ITER is 10−14, it is very
important to inject intense laser beams, to collect the scattered
light effectively and to reduce the stray light of the laser beamin the Thomson scattering measurement, so that the electron
temperature and density can be accurately measured with ahigh signal to noise ratio (S/N).
Laser beam dumps are designed to terminate intense laser
beams and reduce the stray light of the laser beams. Since 109
laser pulses with a 5-J energy are to be injected throughout
the 20 year operation of ITER, the concentrated laser energy
in the valley of a conventional beam dump will cause serious
damages to the surface (Fig.9-2(a)). The accuracy of theThomson scattering measurement significantly degrades if thelaser beam dump is damaged, since the stray light from thedamaged surface of the beam dump increases. A key conceptof the new beam dump is to reduce laser energy absorption
per unit area on the surface and to gradually absorb the energyof the laser beam, so that the stray light will be reduced at thesame time.
We have applied the idea that the absorption ratio depends
on the angle of incidence and the polarization of the laser
beam to the design of a new beam dump. Fig.9-2(b) shows theschematic of the newly proposed beam dump. To graduallyabsorb the beam energy, the laser beam is injected into thebeam dump with a large angle of incidence and is polarized
parallel to the beam-exposed area (S-polarization). In general,the depth should be sufficiently large to gradually absorb the
beam energy. However, the allocated depth for the beam dump
in ITER is up to 125 mm. The new beam dump has a numberof thin molybdenum sheets with thicknesses of 0.5 mm, which
are aligned parallel to each other with a separation of 1 mm.
We optimized the angle and position of bending on the sheets
to disperse the deposition of the beam energy (Fig.9-3).
While conventional beam dump cannot withstand morethan 104 laser pulses, the new beam dump would withstandmore than 109 pulses. In addition, compatibility with ITER’s
thermal and electromagnetic loads was confirmed throughthermo-structural analysis. We have proposed the first detailed
beam dump structure in the world that complies with thesevere conditions of use, installation environment, and space
limitations of ITER.

Reference
Yatsuka, E., Hatae, T. et al., Chevron Beam Dump for ITER Edge Thomson Scattering System, Review of Scientific Instruments, vol.84, no.10, 2013,
p.103503-1-103503-6.

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Nuclear Fusion Research and Development

9-2 Investigation of Strain Using Neutron Diffraction− A Foundation for Improvement of ITER Conductor Performance −(a) Responses of the diffraction profile to strain
Stainless-steel jacket

Neutron

Sub-Cable wrapping

λ: Wave length

Bragg’s law

θ: Angle

2d sin θ=nλCentral spirald (lattice spacing)
becomes narrowerd (lattice spacing)
becomes broader

Cable wrapping
Bending

Superconducting cable
2nd (×3)

φ32.6 mm

3rd (×4)

Peak becomes lower andbroader by bending

TensileCompression

1st cabling (×3) 85 mm pitch 145 mm pitch
45 mm pitch
φ0.83 mm
φ0.83 mm

Copper strand

4th (×4)
250 mm pitch

5th (×6)
450 mm pitch

Fig.9-4 Schematics of the ITER CS conductor
The conductor comprises a superconducting cable and ajacket. Strands are cabled 5 times.
Electromagnetic loadingHigh loading side

High loading side

Low loading side
Strandbending
direction

1.2
1.0
0.8
0.6

Stress-free
Nb3Sn strand

High loadingside
Low loading
side

0.4
0.2
0.0
−0.2
−2

−1

0

1

2

Strain (%)

45 mm

Low loading side(b) Profiles measured by neutron diffraction

Intensity (a.u.)

Nb3Sn strand

Large void

Fig.9-5 Deformation of strands in the cable after testing
The cable was disassembled and visually inspected after
testing.

The central solenoid (CS) conductor for ITER, shownin Fig.9-4, is operated at 45 kA and 13 T. The conductor isrequired to withstand a 60000 cyclic electromagnetic loadingof 50 t/m. Short sample tests found that the superconducting
performance gradually degraded during these cycles.
A clarification of this degradation mechanism is needed.
The performance of the Nb3Sn strands in the CS conductor
is sensitive to strain. Thus, transverse electromagnetic loading
has been considered a major cause of the degradation of
conductors. Namely, the electromagnetic loading bends thestrand at a short pitch (around 5 mm) through contact withadjacent strands in the high loading side (HLS). However,observation of the cable showed that the deflection of strands
on the low loading side (LLS) was larger than that on the
HLS, as shown in Fig.9-5.
The mechanism behind the strain was reconsidered. Since
the fixture of the strand in the cable is relaxed by a large void,
the buckling of the strand occurs due to the axial compressive
force caused by different amounts of thermal contraction

Fig.9-6 Internal strain evaluated by neutron diffraction
Neutrons are diffracted in accordance with Bragg’s law. Theinternal strain in the conductor can be evaluated from neutron
diffraction, since the wavelength is changed by the variation of
the lattice spacing.

between the cable and the jacket.
The visual observation showed a bending strain of around
0.6% in the LLS. Since the bending strain in the HLS could
not be evaluated by visual observation, a quantitative method
was needed to evaluate it. Hence, a neutron diffraction
measurement using the engineering materials diffractometerin the Japan Proton Accelerator Research Complex (J-PARC)
was conducted to evaluate the bending strain. The averagebending strains evaluated from the variation of the profile(Fig.9-6) were 0.32% in the HLS and 0.63% in the LLS.
Consequently, bending with short pitch in the HLS was
not dominant. The bending strains in the LLS that were
evaluated by visual inspection and neutron diffraction were inagreement.
The strands’ deformation can be prevented by the
optimization of cabling, since the strand withstands bucklingby shortening the bending pitch. An optimized CS conductor
could satisfy the requirement for 60000 operation cycles.

Reference
Hemmi, T. et al., Neutron Diffraction Measurement of Internal Strain in the First Japanese ITER CS Conductor Sample, Superconductor Science andTechnology, vol.26, no.8, 2013, p.084002-1-084002-6.
JAEA R&D Review 2014111

Nuclear Fusion Research and Development

9-3 Failure Behavior of the Hot Isostatic Pressed Joint in the Fusion Blanket− The Latest Highlight of Research and Development of Reduced Activation
Ferritic/Martensitic Steel for Use in the Fusion DEMO Reactor −First wall cross section

Torsion specimen with HIP joint

Torsion test system

1700

3

11

7

6

HIP joint interface

7
164

Torsion specimen of base metal
Torsion specimen

D0=5
D1

600

1
1

0

R0.5

3

50

6

Example of fusion blanket

Fig.9-7 Schematic of miniature torsion test method (unit: mm)
The torsion test method is a brand new test technique for evaluating the HIP joint for the thin-walled fusion blanket first wall.

20
Maximum strength (τmax)
HIP-C

15

15
Base metal

HIP-B

10

Torque (Nm)Absorbed energy by torsion test (J)20

Base metal
Pilot-grade
HIP joint

10
Yield strength(τyield)

HIP-A
5

5

Base metal: τyield=425 MPa, τmax=565 MPa
HIP joint: τyield=458 MPa, τmax=564 MPa

HIP-A, B, C: HIP joint materials by various routes
Base metal: Reduced-activation ferritic steel, F82H
0

0
0

5
10
15
20
Absorbed energy by Charpy impact test (J)

0

1
2
3
4
Specific angle of twist (rad/mm)

5

Fig.9-8 Comparison between torsion and Charpy impact
tests
Similar to the conventional Charpy impact test method, thetorsion test technique enabled us to adequately evaluate thekey characteristics of HIP joints produced by various processing
routes.

Fig.9-9 Evaluation of HIP joint components as parts of thefusion blanket structure
The torsion test was the first, to our knowledge, to demonstratethat both the yield shear stress and the maximum shear
strength of the HIP joint components were equivalent to thoseof the base metal.

Reduced activation ferritic/martensitic (RAFM) steel is oneof the leading structural materials for use in the fusion DEMO
reactor, and hot isostatic pressed (HIP) RAFM joints are one ofthe primary technologies for fabricating the cooling channels
in the blanket first wall (FW). For practical application, it isvery important to identify the stability and damage toleranceof the HIP joints exposed to harsh environments, such as highenergy neutron exposure and heat from the burning plasma.
The Charpy impact test is an established method for
gauging damage tolerance with many perceived advantages.
However, it is hard to apply this technique to the thin-walled
cooling channels of the blanket FW, due to limitations on thespecimen size required in testing.
We therefore developed a brand new torsion test method
using a miniature specimen to evaluate HIP joints (Fig.9-7).
There has so far been good agreement between the fractureenergy obtained by the Charpy test and the estimated energy
obtained by the developed torsion test (Fig.9-8). We havealso introduced fracture parameters specific to torsion andidentified the effects that specimen size and geometry had on
them. Eventually, we provided a reasonable solution for the
HIP joints of the practical components.
In this study, we first applied the torsion technique to the
HIP joint components as parts of the blanket structure anddemonstrated good agreement with the properties of the basemetal (Fig.9-9).
With this achievement, we have a powerful tool that enables
us to directly evaluate the HIP joints of the fusion blanket
structure. Establishing this test technique is undoubtedly
viable in the establishment of design codes and standards forthe DEMO reactor.

Reference
Nozawa, T. et al., Determining the Shear Fracture Properties of HIP Joints of Reduced-Activation Ferritic/Martensitic Steel by a Torsion Test, Journal of
Nuclear Materials, vol.427, issues 1-3, 2012, p.282-289.

112

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Nuclear Fusion Research and Development

9-4 Toward the Construction of an Intense Neutron Source that Produces Fusion Reactor Conditions− Engineering Validation Research and Engineering Design of the Lithium Facility −Engineering Validation Research

Remote handling・Replacement of the integrated target
assembly (Japan).
・Replacement of only the backplate (the EU).

Construction, operation and tests of EVEDA
Li test loop (Installed at Oarai)(1) Functional tests of each equipment(2) Tests of high-speed fluid with a free surface undervacuum(3) Long-duration tests (the EU: Cavitation test, etc.)

Li flow

NozzleErosion / Corrosion
・The tests for 1000 to 8000 h in a small
Li loop of the EU (F82H, Eurofer 97)

Fig.9-10 Contents of engineering validation andengineering design activities of the lithium facilityof IFMIF / EVEDA
For engineering validation of IFMIF, a prototype Li test
loop was constructed, operated, and tested. Based onsome of the engineering validation tests, an intermediateengineering design report for IFMIF was created.

Diagnostics
Li Safety handling・Wave height measurement of highspeed Li flow
・Applicability evaluation of diagnostics
by a Li loop at Osaka Univ.
・Assistance analysis by a water test loop
Targetassembly

・Li handling technology, fire
extinguishing testing

Laser HeadAr GasNozzleLi purification system

・Removal of impurities in Li, impurity
monitors

EMP

N
Dump tank

H

Engineering Design

Impurity traps Cold trap Impurity monitors(N, T(D)) (O, Be, etc.) (H, N, O, C, etc.)(Unit: mm)

Li φ165

Fig.9-12 Validation tests of the welding and cutting by afiber laser
The photo shows a validation test of fiber laser welding with theco-operation of the Joining and Welding Research Institute of
Osaka University. The laser output and a welding speed weremaintained at 5 kW and 3 m / min, respectively.

Inlet Pipe

0

100

250

30

390

Double reducer
nozzle0149

FlowStraightener

SUS316 Steel(Two plates witht = 2 mm in each)

357.64200
View Port300

15 m/sOutlet
Pipe

610

Li Flow Channel
with concave552

Fig.9-11 An image of high-speed Li flow with a free surface (left) andthe flow channel (right)
Stable high-speed lithium flows are shown as the most important
validation tests conducted by the Li test loop with the world’s largest flowrate of 3000 ℓ/min (250 °C in a vacuum of about 370 Pa).

In fusion reactors, the phenomena of irradiation hardening
and irradiation embrittlement of materials occurs due to thedisplacement damage and nuclear transformation products(e.g., H and He) in materials caused by 14 MeV neutrons
produced by the fusion reaction (D–T reaction). The amount of
neutron irradiation of materials in ITER, whose construction
is currently underway, is several percentage points below thegoal for fusion DEMO reactors, and is therefore not sufficientfor the evaluation of materials developed for use in DEMO
reactors. Therefore, it is essential to evaluate the dependencyof the irradiation dose on the materials and components in
the irradiation environment for 14 MeV neutrons with a highintensity like that at IFMIF.
In IFMIF, two deuteron beams of 125 mA at 40 MeV (beamSize: 200 mm width and 50 mm height) were injected into aliquid lithium target (260 mm width, 50 mm height, 25 mm
thickness) at 250 °C, having a free surface in a vacuum(10−3 Pa), and neutrons were generated. At present, EVEDA
are being conducted at IFMIF as one of the BA activities incollaboration between Japan and the EU.
We are conducting engineering validation tests and design of
the Li facility in IFMIF, as shown in Fig.9-10. We constructed
and are operating the EVEDA Li test loop with the world’s
largest flow rate (3000 ℓ/min), which is a prototype for the
IFMIF Li facility, and we are acquiring the validation test
data required to determine the construction of IFMIF. In theevaluation of this high-speed Li flow, observation by a video
camera (Fig.9-11) and measurement by a developed noncontact
wave-height laser meter revealed that a stable flow could beformed not only at positive pressures but also at negative
pressures. As an engineering validation test of remote-control
technology, adaptability tests of cutting and welding using afiber laser were conducted to replace the equipment (called atarget assembly) in the Li facility, and some of good results
were obtained (Fig.9-12). The 6500-page intermediateengineering design report for the entire IFMIF facility wascreated in FY 2013, and was based on these engineering
validation tests.

Reference
Wakai, E. et al., Engineering Validation and Engineering Design of Lithium Target Facility in IFMIF/EVEDA Project, Fusion Science and Technology,
vol.66, no.1, 2014, p.46-56.
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9-5 Disassembly of the Japan Torus-60 Tokamak Device for the Japan Torus-60 Super Advanced Project− Disassembly of Large Radioactivated Structures −Fig.9-13 JT-60 Torus hall before disassembly
The JT-60 Torus hall was an complex narrow space. There
was the JT-60 tokamak device with auxiliary tokamak facilities,
diagnostics, and heating system.

Dry Diamond Wire Saw

pulleyDriving pulley

PFcoil support

oxygen-free copper (PFcoil)

Adjustment pulley
Driving pulley

Inconel (Vacuum Vessel)high Mn steel (PFcoil case)

Dry Diamond Wire Saw (φ10 mm)
Coil

Coil

Coil

high Mn steel

PFcoil-case

Fig.9-14 JT-60 Torus hall after disassembly
Disassembly of the JT-60 tokamak was completed
without accidents in October 2012. Assembly of JT-60SA
started in January 2013.

The disassembly of the JT-60 tokamak device (Fig.9-13)
was required for the construction of the JT-60SA tokamak
as part of the Satellite Tokamak program conducted incollaboration between Japan and the EU. The JT-60 tokamak
was a unique fusion device regulated by the Radiation Hazard
Prevention Act. This device had been radioactivated byneutrons during deuterium operations between 1991 and 2008.
This disassembly was the first such experience with a large,
radioactivated fusion device in Japan, and represented a new
approach in consideration of the application of clearance level
regulations in the future. The disassembly started in 2009
and was completed in October, 2012. The total weight of thedismantled components reached up to 5400 tons (Fig.9-14).
One of the main issues facing this disassembly was theneed to cut the several large assembled structures, such as thereinforced points between the toroidal magnetic field coils(TFCs) and the vacuum vessel (VV) with poloidal magnetic
field coils (PFCs).
Both TFCs were weld-jointed at the narrow tops andbottoms of their cases (high Mn steel) to reinforce against

Fig.9-15 Diamond wire-saw system
The diamond wire saw is an embedded diamond chip. It canefficiently cut dissimilar metals without cooling water.

electromagnetic forces. A portable milling machine wasdeveloped to be mounted at the narrow space for cutting thehigh Mn steel, which is a difficult-to-machine material. Thus,
the TFCs could be dismantled one by one.
After removing the TFCs, the main residual components
were the vacuum vessel and the PFCs. Their total weight was
about 320 tons, which was beyond the building crane capacity(250 tons). The residual component was to be separated into
two parts for lifting. The materials of the VV and the PFCs
were Inconel and oxygen-free copper supported with highMn steel, respectively. In this case, matrix structures were to
be simultaneously cut. In the mock-up test, it was found thata dry diamond wire saw could cut those materials without
cooling water. Finally, the application of the dry diamond wire
saw to the separation was successfully conducted (Fig.9-15).
The JT-60 disassembly was efficiently completed by
employing new dismantling technologies, and has provided
valuable data, which will be useful for the disassembly of
nuclear facilities as well as the other fusion devices.

Reference
Ikeda, Y., Okano, F. et al., Safe Disassembly and Storage of Radioactive Components of JT-60U Torus, Fusion Engineering and Design, vol.89, issues 9-10,
2014, p.2018-2023.

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Nuclear Fusion Research and Development

9-6 Accurate Winding of a Nb3Sn Superconducting Magnet− Development of JT-60SA Central Solenoid for Precise Plasma Control −Current Lead

CS

Dia

me

TF CoilEF Coil

ter:

2m

CS1

CS2

CS3

Fig.9-17 CS Model CoilA prototype CS model coil was manufactured with the full-sized
four-layer pancake for the verification test.

CS4

2.1

35
Current

OP1
OP2

QP
(four-layer)

OP3
QP

1.8

Voltage

25

1.5

20

1.2
Inductive
Voltage

15

0.9

10

0.6

5

0.3

Coil voltage (V)OP(eight-layer)

Operating current (kA)

30

OP4
OP5
OP6

0

0
No Voltage

−0.3

−5
0

500

1000

1500

2000

2500

Time (s)Fig.9-16 Central Solenoid for JT-60SA
The CS of JT-60SA is to comprise four electrically independent
modules (CS1-CS4). Each CS module contains 6 eight-layer
pancakes (OP1-OP6) and a four-layer pancake (QP).

Fig.9-18 Voltage Measurement of the CS Model Coil
Only inductive voltage was measured. A superconducting state
was maintained during the performance verification test.

The manufacture of superconducting magnets is in progressas one of the components of the JT-60SA project. The JT-60SA
is to comprise 18 TFCs, a CS and 6 equilibrium field (EF)coils, as shown in Fig.9-16. The TFCs are being manufacturedby the EU, and the CS and EF coils are being manufactured
by Japan.
A NbTi conductor is used for the TFCs and EF coils becauseof its low magnetic field of under 6 T. The NbTi has a highmechanical workability, but cannot be used for high magnetic
field applications. Thus, the Nb3Sn conductor was selected
for the CS since the magnetic field reaches 9 T during theoperation.
The heat-treated Nb3Sn conductor cannot be bent or wound
into a coil configuration because Nb3Sn is very brittle andloses superconductivity easily by deformation. Therefore,
the conductor is wound into the final shape before combining
Nb with Sn, and then the shaped conductor is heat-treated to
manufacture Nb3Sn. For precise plasma control, the circularityof the CS is required to be less than 4 mm per 2 m in diameter.

The technical capabilities to handle the Nb3Sn conductor andthe coil winding with a high accuracy are required for CS
manufacturing.
We developed a winding machine that controls thedeformation amount from the bending radius and the
conductor stiffness. A feedback sequence is added to thewinding process to cancel-out the variation of the bending
stiffness of the conductor. In addition, the winding accuracy
was maintained after the insulation process by appropriate
control of the insulation thickness. The winding accuracyof the preproduction was 1.6 mm, which met the minimum
requirement of 4 mm. The CS model coil shown in Fig.9-17
was manufactured using the developed winding method andwas used for the performance verification test.
The test result shows that the superconducting state was
successfully maintained up to 30 kA (Fig.9-18). This meansthat the CS can be well manufactured using the developed
winding method. The manufacturing of the real CS wascommenced as a result of this achievement.

Reference
Murakami, H. et al., Development and Test of JT-60SA Central Solenoid Model Coil, IEEE Transactions on Applied Superconductivity, vol.24, no.3, 2014,
p.4200205-1-4200205-5.
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9-7 World Record-Setting Dual-frequency Gyrotron− The Development for Electron Cyclotron Resonance Heating in JT-60SA −Electron beamdwin (2.3 mm)≒ 4 × λ/2 (110 GHz)≒ 5 × λ/2 (138 GHz)

Microwave(to be transmitted
to the target)(2) Design of cavity resonatorshape and microwave profiles
for high efficiency microwave
generation at two frequencies(3) Design of the low
loss mode convertor
for two frequencies

Triode-type electron gun
enabling optimization of
the gyration energy ofelectrons

Output energy
1 MJ

1.2

Gyrotron output power (MW)(1) Selection of two frequencies andwindow thickness (dwin) to suppress
reflection at the output window
dwin = n × λ/2
n: integer, λ: wavelength

10 MJ

Achieved in 2013
1 MW, 10 s

1.0

0.8

Target of JT-60SA
1 MW, 100 s

Short pulse
commissioningin 20120.6

Limitation on the
input energy tothe present
transmission line

0.4
110 GHz0.2

138 GHz

0
0.1

1

10

100

Pulse duration (s)Fig.9-19 Key points of the gyrotron design for obtaining
high performance at two frequencies
High efficiency and low loss were achieved by three key), and the optimum energy of electron gyration was
points (obtained with a triode electron gun.

Fig.9-20 Progress in high-power, long-pulse development
and its target value
A gyrotron output power of 1 MW for 10 s, which is a world
record for a dual-frequency gyrotron, was demonstrated.

An electron cyclotron heating (ECH) system will generateor sustain high-performance plasma in JT-60SA through theinjection of high-powered microwaves. ECH enables theplasma to be heated locally where the magnetic field strength
corresponds to the microwave frequency. However, long-pulse
operation of a gyrotron, which is a source of high-powered
microwave, was previously obtained only at a single frequency.
Consequently, it was not capable of heating the desired positionin the plasma when the magnetic field was changed. To solve
this issue, a dual-frequency gyrotron, which can independently
be operated at two frequencies, is being developed.
Three key points of the dual frequency gyrotron design(Fig.9-19) are as follows. (1) The thickness of the output
window was required to be a multiple of one half of the
wavelength for complete suppression of the reflection at
the window. (2) A desired set of operating frequencies andmicrowave profiles in a cavity resonator was required to obtainhigh conversion efficiency from the energy of the beam fromthe electron gun to the microwave energy. These conditions
have been satisfied simultaneously with a window thickness of
2.3 mm and frequencies of 110 GHz and 138 GHz. Moreover,(3) the shape of the mode convertor, which converts the
microwave profile oscillating in the cavity resonator to one

suitable for transmission, has been designed to reduce internal
loss. A gyrotron design that was sufficient for high-power,
long-pulse operations without overheating of the gyrotroninternal components was obtained.
To obtain high efficiency in practical operation, an
optimized condition of the gyration energy of the electrons in
the electron beam was required. The developed gyrotron was
equipped with a triode-type electron gun, whereby the gyrationenergy of the electrons could be controlled. This characteristic
enabled us to obtain high efficiency at both frequencies as
expected, which was not possible with a diode-type electrongun (e.g., 0.95 MW/140 GHz, 0.85 MW/105 GHz in Russia).
The developed gyrotron successfully operated at a nominal
output power of 1 MW for 10 s (Fig.9-20). This result set aworld record for dual-frequency gyrotrons.
Moreover, we confirmed that the internal loss at the mode
convertor and the temperature increases of the components
were sufficiently low, as was expected. Although the coolingcapability of the transmission line for transmitting the
microwave from the window to the target to be heated limited
the pulse length this year, it should be possible to expand thepulse length toward the target (1 MW/100 s) by improving thetransmission line for JT-60SA.

Reference
Kobayashi, T. et al., Dual Frequency ECRF System Development for JT-60SA, Transaction of Fusion Science and Technology, vol.63, 2013, p.160-163.

116

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Nuclear Fusion Research and Development

9-8 Improvement of the Stability Analysis Model for Rotating Plasmas− Introducing Rotation Effects into the Kinetic-Magnetohydrodynamic Model −

(a)

(b)
New kinetic-MHD model
Conventional kinetic-MHD model
MHD model

Central axis

Tokamakcross section

Toroidal rotation

Toroidal rotation
at cross section (km/s)

Tokamak plasma

Large rotation
shear
Small
rotation shear

25

Small shear
Large shear

20

15

10

5

Stabilize

0
0.1

0

Plasma minor radius
Plasma center
Plasma surface

Fig.9-21 Schematic of the tokamak
plasma and toroidal rotation
A tokamak plasma rotating in the toroidal(azimuthal) direction is schematicallyshown. Since the tokamak is axisymmetric,
the cross section is focused.

RWM growth rate (/m sec)

25
50

0.2

0.3

Distance between plasma surface
and wall (m)Fig.9-22 (a) Rotation profile in a tokamak cross section (b) RWM growth rate
RWM growth rates by various models. In the new kinetic-MHD model, the RWM
growth rate is reduced by the increase of rotation shear.

The JT-60SA tokamak device, which is being constructedat the Naka site of JAEA, is designed to complement ITER
research and establish a foundation for a tokamak DEMO
reactor. In particular, the JT-60SA device aims to confine ahigh-pressure plasma, because fusion output increases withthe plasma pressure. However, when the plasma pressure
increases, a macroscopic instability known as a resistive wallmode (RWM) becomes unstable, limiting the achievablepressure value. Toroidal rotation (Fig.9-21) is known tostabilize RWM. Interaction between the RWM and theconstituent particles’ motion has also been revealed to beimportant for the stabilization of RWMs.
To analyze the interaction between the RWMs and theparticles’ motion, the kinetic-magnetohydrodynamic (MHD)
model, which analyzes energy exchange between macroscopic
MHD modes (such as RWMs) and constituent particles, hasbeen formulated. In the conventional kinetic-MHD model, thebackground plasma is assumed to be static (no rotation). In
this study, we shed new light on the fact that the constituent
particles’ motion is affected by plasma rotation. For example,
particle motion is affected by the Coriolis force, which isan inertial force appearing in a rotating frame working in
the direction perpendicular to the rotational direction. In therotating frame, another inertial force, the centrifugal force,

additionally affects the particle motion. Considering theseeffects, we reformulated the kinetic-MHD model. The energyof RWMs was found to be strongly damped by resonancewith the particles’ motion affected by Coriolis and centrifugal
effects. Note that these effects were overlooked in theconventional model.
Based on the newly derived kinetic-MHD model, we haveinvestigated RWM stability. Fig.9-22(a) indicates the rotation
profile in a tokamak cross section, and shows that the rotation
shear (gradient) is increased in keeping with the rotation
amplitude at the location where the energy interaction between
RWMs and particle motion is large. Fig.9-22(b) shows
computation results of the RWM growth rate as functions of
distance between the plasma surface and the wall. Since thearrow in Fig.9-22(b) indicates the increased rotation shear,
we find that the RWM growth rates are decreased (stabilized)by the rotation shear. This result indicates that the control of
the rotation profile is beneficial for stabilizing RWMs, which
can contribute to the design of tokamaks with high-pressureplasma confinement.
The present study was partly supported by Japan Society for
the Promotion of Science (JSPS) KAKENHI Grant-in-Aid forYoung Scientists (B) (No.24760708).

Reference
Shiraishi, J. et al., On Kinetic Resistive Wall Mode Theory with Sheared Rotation, Plasma and Fusion Research, vol.9, 2014, p.3403027-1-3403027-4.

JAEA R&D Review 2014117

Nuclear Fusion Research and Development

9-9 Safe Shutdown Scenarios for Fusion Reactors− The Development of a High-Energy Runaway Electron Simulation Code −Tokamak vacuum vesselCore plasma
Vertical beam position (m)(a) 1 MeV(c) 30 MeV

0.2

0.2

0.2

0.1

0.1

0.1

0

0

0

−0.1

−0.1

−0.1

−0.2

−0.2
3.2

3.4

3.6

3.2

−0.2
3.4

3.6

3.2

3.4

3.6

Horizontal beam position (m)Runaway electron beam

Fig.9-23 Schematic of core plasmas
and the runaway electron beam
During disruptions of tokamak operation,
the voltage induced in the core plasma
accelerates electrons and yields theformation of a high-energy electron beam,
which we call “runaway electrons (REs).”(b) 10 MeV

Fig.9-24 Simulation of a high-energy electron orbit with magnetic
perturbations
Points in the figures represent the positions at which electrons pass througha tokamak minor cross section; the color of each point indicates the distance
between electrons and the beam axis. (a) A concentric circular structure indicates
well-confined electron orbits with a negligible radial excursion. (b) Magnetic
perturbations cause unstable electron orbits. (c) At high energy, the unstable orbit
becomes chaotic, exhibiting random and large radial excursion.

During tokamak operation, a quench of the plasma current,
being inherent to the confinement of core plasmas, often
occurs due to various internal and external causes, which wecall “disruptions.” In future DEMO and commercial reactors,
the operation regime must carefully be chosen to avoid suchdisruptions. On the other hand, in present machines aimed at
extending the operational regime so that high-performance
tokamaks realize fusion energy, uncontrollable disruptions
often occur in experiments.
It is important to understand the physical phenomena
involved in disruptions, which are not observed in normal
plasma operation. Particularly, the generation of a beam ofenergetic electrons (Fig.9-23), which is called as “runaway
electrons (REs),” is one of the most important topics forthe disruption study. REs often attain energies of several
tens of MeV through the conversion of magnetic to kinetic
energy, which can cause serious damages to the in-vessel
components. However, detailed mechanisms of RE generation
during tokamak disruptions have not yet been elucidated. To
improve our understanding of the interaction between REs
and magnetic perturbations, which has been conventionallyinvestigated using relatively simple models, we havedeveloped a new simulation code for analyzing high-energy
electron orbits while considering the detailed structure ofmagnetic perturbations during disruptions.
Using the developed code, we have analyzed the RE orbit
for JT-60U scale disruptions and found a novel mechanism
such that only high-energy REs above 10 MeV exhibit unstabletrajectories and significant radial transport (Fig.9-24). In
previous studies, the confinement of such high-energy REs wasconsidered to improve because high-energy REs deviate frommagnetic field lines owing to the centrifugal force, and tend
not to feel magnetic perturbations. The present calculation has
clarified the possibility that high-energy REs exhibit strongly
disturbed trajectories compared with low-energy ones. Our
simulation shows that such energy dependence is explainedby the interference between the magnetic perturbation and theperiodic motion of high-energy REs, which occurs if their own
characteristic frequencies satisfy a certain condition.
The findings obtained in this study provide a clue toward
understanding the RE generation mechanism and offers aphysical basis for the control of disruptions and REs in ITER
and DEMO reactors.

Reference
Matsuyama, A., Yagi, M. et al., Stochastic Transport of Runaway Electrons due to Low-Order Perturbations in Tokamak Disruption, JPS Conference
Proceedings, vol.1, 2014, p.015037-1-015037-4.

118

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Nuclear Fusion Research and Development

9-10 Removing Fuel from the Nuclear Fusion Blanket− Investigation of Tritium Recovery from a Future Fusion Blanket −DT neutron (JAEA-FNS）
1011 neutron / s

Lithium titanate (Li 2TiO 3)
humidity metersmolecular sieve

HTO

Container

150 ℃

Li 2TiO 3 pebble (70 g)
300〜800 ℃
CuO(250 g)
500 ℃

HT

HTO+HT

humidity meters

H2 (1%)/Helium gas

Recovered tritium (Bq/source neutron)

Beryllium

Cooling air−11(×10 )
1.4

Dry H2 / He gas
HTO
HT

1.2
1.0
0.8
0.6
0.4
0.2
0.0

300

600

800

Temperature of Li2TiO3 (℃)
Bubbler for HT

Bubbler for HTO

Fig.9-25 JAEA-FNS tritium recovery experiment apparatus
Li 2TIO 3 pebbles were installed in the blanket container in anexperimental mock-up. During DT neutron irradiation, tritium
produced in the pebbles was recovered with sweeping gas.

Tritium used as fuel in fusion reactors will be producedby nuclear reactions between lithium and neutrons generated
by DT fusion reactions, like 6Li (n, α)3T. It is important toinvestigate methods to recover the tritium produced in the“blanket,” which is a key part for the development of a fusion
reactor. To investigate the recovery of the tritium produced, we
have conducted a series of tritium recovery in situ experiments
using the DT neutron source at JAEA’s Fusion Neutronics
Source facility (FNS).
We irradiated an experimental assembly that simulated afusion blanket (Fig.9-25) and recovered the tritium producedin the Li2TiO3 pebbles that filled the assembly by sweepingwith helium gas mixed with a small amount of hydrogen(H2, 1%). During the neutron irradiation, the Li2TiO3 pebbles
were heated to temperatures ranging from 300 °C to 800 °C
by a heater set near the pebbles.
To remove water moisture from the gas emerging from thegas cylinder, a molecular sieve (MS) was inserted in the line

Fig.9-26 Result of the JAEA-FNS tritium recovery
experiment
JAEA-FNS conducted the first tritium recovery experimentwith DT neutron source. Our quantitative measurementof the recovered tritium indicated that the recovered
tritium corresponded to calculated tritium production.

to the inlet of the assembly. Among the tritium (HT and HTO)
produced in the Li2TiO3 pebbles and swept in the gas, only
the tritiated water (HTO) was collected to the first bubblers.
A copper oxide bed was set upstream of the second bubblers
to oxidize the tritium gas (HT) component to HTO, since
the second bubblers could collect only the HT component
generated in and released from Li2TiO3 pebbles. The recovered
tritium was measured with a liquid scintillation counter.
As shown in Fig.9-26, for the first time, the total amountof tritium recovered from the Li2TiO3 pebbles was found tobe almost the same regardless of the temperature and in good
agreement with the tritium production amount calculated.
This means that the tritium produced in the blanket can berecovered by sweeping with the helium-hydrogen (H2, 1%) gas
mixture. The percentage of tritium released into the sweeping
gas in the form of HT was only 40% at 300 °C, but increased
up to 95% at higher temperatures. The HT ratio at 800 °C was
almost the same as that at 600 °C.

Reference
Ochiai, K. et al., DT Neutron Irradiation Experiment for Evaluation of Tritium Recovery from WCCB Blanket, Fusion Engineering and Design, vol.89,
issues 7-8, 2014, p.1464-1468.
JAEA R&D Review 2014

119

Nuclear Fusion Research and Development

9-11 Stably Supplying Fuel Tritium to Fusion Reactors− Establishment of a New Pebble Fabrication Technology for an Advanced Tritium Breeder −Large grain

SlurryGel
SpheresContinuous
phase
(Oil)
Li2+xTiO3+y(Slurry)

Cross section

Oil

Slurry flow is cut by oil flow,
and gel spheres are obtained

Trial Fabrication
Sinteringatmosphere: N2

10 μm

T-shaped flow path

Cross section

Syringes
Solidification
Oil

Li2TiO3
Slurry withexcess Li

Small grain

Li2+xTiO3+y
Sintered pebbles of 1 mm

Gel Spheres

Improved Fabrication
Sinteringatmosphere: Vacuum

Grain size:
<5 μm

10 μm

Fig.9-27 Pebble fabrication of tritium breeders using anemulsion method
The flow of the Li2TiO3 slurry with excess Li (Li2+xTiO3+y) is cut
by an oil flow from an oil-filled syringe. The emulsion method
is suitable for mass production of Li2+xTiO3+y pebbles.

Fig.9-28 Improvement of the grain size of tritium breeder
pebbles
To prevent grain growth, Li2+xTiO3+y gel spheres were sinteredin vacuum. The average grain size of the sintered Li2+xTiO3+ypebbles was less than 5 μm.

Fusion reactors need deuterium (D) and tritium (T) as theirfuel. Since tritium does not exist in nature, it is necessary to
produce tritium in a reactor by neutron irradiation of lithium(Li). Lithium titanate (Li2TiO3) is one of the most promising
candidates among tritium breeders because of its tritium
release characteristics. However, the mass of Li in the breeders
decreases in a hydrogen atmosphere because of Li evaporationand Li burn-up.
To prevent the mass decrease of Li at high temperatures,
Li2TiO3 with excess Li (Li2+xTiO3+y) has been developed as anadvanced tritium breeder. The emulsion method as a pebble
fabrication technique for Li2+xTiO3+y was developed under the
IFERC program of BA activities. The emulsion method caneasily produce large volumes of uniform submicron particles.
For fusion reactors, tritium breeder pebbles of 1 mm are needed.
To examine whether the emulsion method is suitable for pebble
fabrication, we used the granulator shown in Fig.9-27.
This granulator comprised two syringes arranged in a
T-shaped flow path. One syringe was filled with oil and theother with a Li2+xTiO3+y slurry. The two flow lines from thesyringes were connected perpendicular to one another. This
arrangement allowed us to cut the Li2+xTiO3+y slurry flow withan oil flow from the oil-filled syringe. The size of the tritium
breeder gel spheres was controlled by the flow speeds of theoil and the slurry. The gel spheres were placed in an oil-filledcontainer.
In a series of fabrication trials, the average grain size on thesurface and cross section of the sintered Li2+xTiO3+y pebbles
was 2–10 μm. Considering the tritium release characteristics,
the optimum grain size after sintering should be less than 5 μm.
The grain growth factor was assumed to be the presence ofbinder in the gel spheres. The remaining binder reacted with
the Li in Li2+xTiO3+y and generated Li2CO3. To prevent thisreaction, the Li2+xTiO3+y pebbles were sintered in a vacuum
atmosphere (Fig.9-28). The average grain size on the surfaces
and cross sections of the sintered Li2+xTiO3+y pebbles was thus
less than 5 μm.
The results suggest that the Li2CO3 generated by theremaining binder affects grain growth and that sintering undervacuum improves the grain size of the pebbles.

Reference
Hoshino, T. et al., Development of Fabrication Technologies for Advanced Breeding Functional Materials for DEMO Reactors, Fusion Engineering and
Design, vol.87, issues 5-6, 2012, p.486-492.

120

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10 Computational Science and E-Systems Research
Advanced Computing and Simulation Technology to Support Nuclear Research and Development
Application of Computational Science and E-Systems Research
Power input

Dose rate in air(μSv/h)19.0 <
9.5 -19.0
3.8 - 9.5
1.9 - 3.8
1.0 - 1.9
0.5 - 1.0
0.2 - 0.5
0.1 - 0.2
<0.1

Heat transport

Morioka

Akita

150

km

Cs

Sendai
Yamagata

80 km

Niigata
Fukushima

20km

The TEPCO’s Fukushima
Daiichi NPS (1F)Z
NaganoUtsunomiya
Mito

X

Y

SaitamaTokyo
Chiba
Yokohama

Shizuoka

Sep. 2012

Radiation dose
distribution map

y

Si
O

Tokamak microturbulence

X

Crystal structure ofzeolite called mordenite

X

X

y

Birds-eye view of temperature profile

Fluctuating electrostatic potential (left) andThree-dimensional view of temperature profile (right)

Nuclear safety research

Nuclear fusion research

Fukushima support research

Computational science research・Prediction of nuclear material degradation
・Prediction of physical properties of nuclear fuels
・Seismic analysis

Computing technology research・ Large-scale data visualization
・ Massively parallel computing

Computational Scienceand E-Systems Research

Sustained performance(TFlops)

Maebashi

Kofu

103

102

101

K computer
BX900(JAEA)
100
102

103

104

105

Processors

Simulation for determining the effect ofhydrogen atoms on crack propagation in iron

Speed-up of plasma calculation

C
Secondarypressurizedwater cooler(SPWC)

Computer operation and management
・Supercomputer・Network system

Intermediate heatexchanger (IHX)
Primary pressurizedwater cooler (PPWC)

Analytical model for whole numericalsimulation of primary cooling system of HTTR

values of stress is large

Space axis
Time axis

The connection
Interaction model
Is setchange cycle is short

Visualization of results obtained froma vibration simulation of a nuclear facility

Fig.10-1 Advanced computing and simulation technology and its application to current nuclear research
Computational science research, computing technology research, and computer operations and management are jointly promoted,
and the results are applied to research that supports efforts to resolve the problems in Fukushima and atomic energy research.

Computational science is now widely regarded as the third
research methodology, following theoretical and experimental
research methods. Computational science is indispensable forthe elucidation and prediction of phenomena that are difficult toobserve or replicate experimentally.
The Center for Computational Science and E-Systems
promotes important research in nuclear energy fields, suchas prediction of nuclear material degradation, prediction ofphysical properties of nuclear fuels, and seismic analysis.
As a computing technology research, we promote large scale datavisualization and massively parallel computing. Additionally,
computer operation, management of supercomputers, andnetwork maintenance are promoted in conjunction with ourcore research areas. Furthermore, our research results aredeveloped further in the domains of Fukushima support
research, nuclear fusion research, nuclear safety research, andso forth, through cooperation with organizations both insideand outside JAEA (Fig.10-1).
First, we introduce the research Topics that have beendeveloped for Fukushima support. In Chapter 1 Topic 1-5,
we build a system that manages radiation dose measurement
results by environmental monitoring and make them available
to the public and administrative personnel. In Chapter 1
Topic 1-7, for more efficient decontamination, we clarify theCs adsorption-desorption mechanism on clay minerals throughsupercomputer simulations.
Second, we introduce computational science research that has
been undertaken. In Topic 10-1, the new knowledge obtainedby applying computational science to the analysis of nuclear
material degradation is shown. Topic 10-2 introduces researchand development toward a new earthquake risk evaluation
technique for a nuclear installation based on Monte Carlosimulation. Computing technology research that succeeds in thedevelopment of a high-speed computation technique is shown in
Topics 10-3 and 10-4. In Topic 10-3, a full-scale simulation of
the thermal conductivity behavior exhibited by a superconductor
is presented. This was previously impossible because it requireda large computing time in general. In Topic 10-4, we show
how a multi-scale simulation for evaluating and forecasting
plasma turbulence is made achievable by maximizing parallel
computing performance.
We work on nuclear research and development using
computer science and technology, engage in computer
technology research and development that supports nuclear
research and development, and have promoted the application ofresearch results to various fields. In the future, we will continue
to work diligently on nuclear research and development
utilizing advanced computational science.
JAEA R&D Review 2014

121

Computational Science and E-Systems Research

10-1 Helium Induces Grain Boundary Fracture of Iron
− Finding a Universal Law of Grain Boundary Strength Loss Using a Supercomputer −(b)∑17(0 -1 4)∑25(0 -1 7)∑27(1 -1 5)∑3 (1 1 1)
∑3 (1 -1 2)∑5 (0 -1 3)∑11(3 -3 2)

Normalized GB cohesive energy

(a)

1.0
0.8
0.6
0.4
0.2
0

0

1

2

3

4
2

He concentration (1/m )5× 1018

Fig.10-2 Calculation method for GB cohesive energy
GB cohesive energy is defined as the total energy increase
per unit area after separating the system at the GB, as shownin (a). The GB used for the present study is symmetrical at theboundary, as seen in (b), being called the symmetrical tilt GB.

Fig.10-3 The universal relationship between GB cohesiveenergy and He concentration
GB cohesive energy decreases as He is accumulated. The
above value is normalized to the cohesive energy without Heaccumulation.

Reduced activation ferritic steels are being considered as
structural materials of future nuclear fusion reactors. However,
it is expected that these steels will become brittle after
long-term exposure to a fusion reactorʼs environment. Thismay be caused by a decrease in the strength of the grainboundaries (GBs), which are surfaces between two adjacentgrains. We suspect that loss of the strength may be caused byhelium (He) accumulation at the GBs, which is generated by
nuclear transmutation reactions under fast neutron irradiation.
He is accumulated either in the form of bubbles or in
an atomic form without forming bubbles. The former
occurs under high temperatures and can be observed byelectron microscopes. The latter is supposed to occur at low
temperatures, but it is difficult to determine the locationsor the amount of He involved using experimental means;
therefore, the relationship between the GB cohesive energy
and the amount of accumulated He is not easily known.
For this reason, we particularly focused on the applicationof computational methodology to He embrittlement without
bubble formation under low temperatures.
In the present study, we defined GB cohesive energy as thetotal system energy increase due to ideal fracture (Fig.10-2)

at the GB divided by its area. If this total energy increases,
the cohesive energy is positive and a fracture is not expected;
otherwise, the fracture is more likely to occur. We evaluatedthis energy using an atomistic modeling method.
Note that there are plenty types of GBs. In the current study,
symmetrical tilt GBs (Fig.10-2(b)), which are suitable forour atomistic modeling, were considered. We searched for
He accumulation positions for various GBs and found thatthese varied with the type of GBs. We artificially inserted He
atoms at these positions ( in Fig.10-2(b)) and calculated
GB cohesive energy as a function of He concentration on the
GB. The results for many types of GBs are shown in Fig.10-3.
It is known that GB cohesive energy is significantly influenced
by He formation energy, which is GB-dependent; however,
we found that GB cohesive energy was not dependent on
the type of GBs, and was almost uniquely determined by
He concentration. This result is critical, because it suggeststhat investigations on a single GB are sufficient to gauge GB
embrittlement caused by He accumulation. As there exist
an infinite number of GBs, the benefit gained by this result
could significantly increase the efficiency of these types ofinvestigations for other materials.

Reference
Suzudo, T. et al., Atomistic Modeling of He Embrittlement at Grain Boundaries of α-Fe: a Common Feature over Different Grain Boundaries, Modelling and
Simulation in Materials Science and Engineering, vol.21, no.8, 2013, p.085013-1-085013-8.

122

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Computational Science and E-Systems Research

10-2 Toward Reliability Enhancement of Seismic Risk Assessment of NPP− Proposal for a Ground Motion Generation Method that Combines
a Fault Model and Frequency Prediction −Arrival time
seismic hazard curve

1E−01

target range for seismic
hazard analysis

Fault-unspecified
crustal earthquakes

1E−02

Acceleration(cm/s2)

1E+00

Mj 6.2 X = 3.2 km
0range for basic design
ground motion Ss

1E−05
0

500

1000

1100

10 s
0

1E−03
1E−04

Fault-specified
earthquakes in theSouthern Kanto

1500

acc. on the free rock surface (cm/s2)

Freq. for high-cut filter
Q-value coefficient
Asperity area ratio

fmax
CQC , CQp
CSa

Location of the asperity

aspX, aspY

Rupture starting points

startX

60

Mj 6.9 X = 43.2 km

25 s
14 s

20

40

60

Time (s)1200

Mj 7.3 X = 82.8 km
0

40 s

-1200
0

20

23 s

40

60

Time (s)Fig.10-5 Example time histories of reproduced GMs
The GMs in this figure are for the same hazard level; however, it canbe seen that GMs reproduced from different faults show different
GM properties. (Mj: Magnitude of earthquake, X: Distance fromseismic source)

1.0

log10 (Maximum acc. ratio)ατrAcceleration(cm/s2)

Fault-unspecified
Pacific plate
earthquakes

Table 10-1 SSC, considering the uncertainty
MCS was performed by setting the sizes of the uncertainties
for stress drop, high-frequency cutoff filter, among others;
these parameters were confirmed to have a large impact
on maximum acceleration by sensitivity analysis.

Rise time coefficient

40
Time (s)

0

0

∆σ

20

6s

1200

-1200

Fig.10-4 SH curve based on the attenuation relationship
The index of the GM intensity is the maximum accelerationon the free rock surface. In this study, the target of the GM
reproduction is an annual exceedance probability of 10–5,
which is the safety target for the reactor containment vessel,and corresponds to a maximum acceleration of 1100 cm/s2.

Stress drop

Duration time

1200

-1200

50 waves for each range

Acceleration(cm/s2)annual frequency

1E+01

Fig.10-6 Example of the relationshipbetween maximum acceleration ratio
and SSC (stress drop)
The horizontal axis is the deviation of SSC,
and the vertical axis is the logarithm of theratio of median acceleration to maximum
acceleration for each seismic GM.

y = 0.1365x
R2 = 0.3385
0.5

0.0

−0.5

−1.0
−3

Seismic risk assessment (SRA) of nuclear facilities iseffective for the consideration of possible ground motions(GMs) that are beyond design assumptions, allowing measures
to be reinforced to ensure better safety. Recently, we haveembarked on a trial of an enhancement to the SRA method,
utilizing a three-dimensional detailed analysis technology that
we have developed, with the objective of developing an SRA
method using the Monte Carlo simulation (MCS). Insteadof the conventional separated evaluation method, potential
damage to equipment is assessed by a response simulation
using the GM of each item. The advantages of this technique
are that it identifies the seismic source characteristics (SSC)that are likely to damage equipment of interest, and that it
directly evaluates the correlations between equipment damage.
We began by considering the generation method of the GM
as an input. The GM generation methods in conventional SRA
are divided into two types: the empirical approach based on
the seismic hazard (SH) and an attenuation relationship, andthe physical approach based on a fault model. We encountered

−2

−1

0

1

2

3

Stress drop ∆σ deviation (z)problems that defeated both methods: the SSC of the GM
could not be fully treated in the empirical approach, and thefrequency of the GM could not be considered in the physicalapproach. Therefore, to solve these problems, we devised atechnique that combines the two approaches. We proposed amethodology based on MCS to generate a set of GMs that areconsistent with the SH, and to consider the differences in thesource characteristics using a fault model. Fig.10-4 shows theSH curve at the site. Fig.10-5 shows some examples of time
histories of the reproduced GMs.
To date, a trial was conducted at the Oarai site of JAEA togenerate GMs by MCS and extract the seismic GMs according
to the frequency. Table 10-1 shows the SSC, considering theuncertainty. Consequently, it was found that the SSC (suchas high frequency cut-off filter and stress drop) had a large
impact on the maximum acceleration of the input, as shownin Fig.10-6. The generation of seismic GM with known SSC
enables the analysis of seismic GM characteristics and SSC
that affect the damage to nuclear facilities.

Reference
Nishida, A. et al., Characteristics of Simulated Ground Motions Consistent with Seismic Hazard, Transactions of 22nd International Conference on Structural
Mechanics in Reactor Technology (SMiRT-22), San Francisco, California, USA, 2013, 10p., in CD-ROM.
JAEA R&D Review 2014123

Computational Science and E-Systems Research

10-3 Simulations for Superconductors with High-Performance Eigenvalue Solver− Toward Real-Size Scale Simulations −Ideal131072 × 131072
65536 × 65536

Thermal conductivity (a.u.)

Speedup

100

10

1

4
3
2
1
0

64

128

256

512

1024

4096

Number of CPU cores

20

30

40

50

60

70

80

90

100

Inter-vortex distance (Å)

Fig.10-7 Strong scaling plot
Our method shows a perfect strong scaling even with 4096
CPU cores.

Fig.10-8 Magnetic field dependence of the thermal
conductivity in a superconductor
Thermal conductivity increases with increasing magnetic
fields.

Superconductivity is one of the most remarkable phenomena
in condensed matter systems. A typical behavior is thesudden disappearance of electrical resistance below thesuperconducting transition temperature. Presently, in the nuclear
R&D field, potential applications of the superconductors arebeing intensively studied. In fact, a superconducting coil will be
indispensable for a nuclear fusion reactor. The superconductors
are also applicable for the device in the strong radiation such asthat with the space probes.
Nano-superconductors are the candidates for replacing
the silicon-based semi-conductors. There are various kindsof applications such as the micro SQUID and single photon
detectors. Thus, the simulation technique dealing withquantum phenomena in superconductors is very important. In
the simulations for superconductors, the superconducting orderparameter has to be self-consistently determined by solving
the eigenvalue problems of the Hamiltonian matrix. However,the size of the matrix for the nano-superconductors withreal-size scale is too huge to calculate with the previous
numerical methods. Thus, we developed the simulation
technique with the high-performance eigenvalue solver
specialized in the massively-parallel supercomputers.

By using the massively-parallel Chebyshev polynomial
expansion method for the self-consistently calculations and
Sakurai-Sugiura method for calculating physical quantities, we
performed the simulations for superconductors with the largest
size in the world. Our method has the high-performance evenwith 4096 CPU cores as shown in Fig.10-7. With the use of
our method, we calculated the magnetic field dependence ofthe thermal conductivity, which is difficult to calculate in theprevious methods (Fig.10-8).
We revealed that Sakurai-Sugiura method, that the
eigenvalues in a given domain can be obtained very fast,is efficient for calculating the physical quantities in the
nano-superconductors. Our high-performance simulations
code might be efficient even in the exa-scale supercomputers.
By combining our code with supercomputers, the real-size
scale simulations for superconducting devices can be
performed and novel functional materials and devices can beconstructed in the future.
The present study was partly supported by Japan Society for
the Promotion of Science (JSPS) KAKENHI Grant-in-Aid for
Scientific Research (B) (No.24340079).

Reference
Nagai, Y. et al., Numerical Construction of a Low-Energy Effective Hamiltonian in a Self-Consistent Bogoliubov-de Gennes Approach of Superconductivity,
Journal of the Physical Society of Japan, vol.82, no.9, 2013, p.094701-1-094701-10.

124

JAEA R&D Review 2014

Computational Science and E-Systems Research

10-4 Massively Parallel Techniques for Realizing 105 Times Higher Spatio-Temporal Resolutions− The K Computer Sheds Light on Multi-Scale Turbulence in Fusion Plasma −Transpose is performed ina segment

Ion-scalefluctuations~20 mmPoint-to-Point communications
are performed between
adjacent segments

Electronscale fluc.
~0.5 mm

Reduction is performed in across section(Segment colors correspondto plasma species)

Fig.10-9 Optimized inter-node communication on 3D torus network
The 5D phase space in fusion plasma turbulence is split into small
segments, which are allocated on a 3D torus network so as to minimize
the required inter-node communications (transpose, point-to-point,
and reduction). Thanks to this segmented process mapping, thecommunication cost is significantly reduced.

The analysis of turbulent transport is one of the central
areas in magnetic fusion research, as the transport caused
by electromagnetic turbulent fluctuations has significant
impacts on the confinements of fusion fuel in the form ofplasma. Numerical simulations of plasma turbulence requiresignificant computation to solve the time evolution of theplasma distributions in a 5-dimensional (5D; three spatial andtwo velocity coordinates) phase space.
Most of the previous studies have treated the electron-scale
turbulence (wavelengths ~0.5 mm, frequencies ~400 kHz) andthe ion-scale turbulence (wavelengths ~20 mm, frequencies~10 kHz) individually by assuming scale separation. The
assumption, however, is not obvious. Hence, it is an important
issue to reveal whether there are cross-scale interactions ornot. Multi-scale simulations resolving their wavelengths andfrequencies require 105 times higher resolutions than those insingle-scale simulations. To realize their large computational
requirements, we had to wait for the appearance of petascale
supercomputers, such as the K computer (K).
K is a distributed parallel computer, which connects anumber of computation nodes with an inter-node network.

Fig.10-10 Electrostatic potential fluctuations
in multi-scale fusion plasma turbulence
We realize multi-scale turbulence simulations
that resolve both electron-scale (shown in themagnified view) and ion-scale fluctuations (shownin overall view) consistently, and enable analysisof their cross-scale interactions.

Therefore, inter-node communications often become abottleneck for performance. Investigating various types ofinter-node communications in the 5D phase space problem, we
have developed an optimized communication algorithm on K.
The key idea is the segmented mapping of the communication
processes, so that communications are performed locallyand independently on a 3D torus network (Fig.10-9). Thissignificantly reduces the communication cost by ~60%.
Thanks to the optimizations, we have achieved 0.78 PFLOPS
using ~600000 cores on K, and realized multi-scale fusion
plasma turbulence simulations (Fig.10-10). The simulation
result reveals the existence of cross-scale interactions
between the electron and ion scales and the change of the
turbulent spectrum from those in single-scale simulations.
These achievements were internationally appreciated with
the Best Poster Award in International Conference for High
Performance Computing, Networking, Storage and Analysis:
SC13 among other awards.
The present study was supported by HPCI Strategic Program
Field 4 and G8 Research Council Initiative NuFuSE.

Reference
Maeyama, S., Idomura, Y. et al., Computation-Communication Overlap Techniques for Parallel Spectral Calculations in Gyrokinetic Vlasov Simulations,Plasma and Fusion Research, vol.8, 2013, p.1403150-1-1403150-8.
JAEA R&D Review 2014125

11 Development of Science & Technology for Nuclear Nonproliferation
Development of Technology and Human Resources in the Area of Nuclear Nonproliferationand Security to Support the Peaceful Use of Nuclear Energy
The Japan Atomic Energy Agency (JAEA) has been
conducting the following technological and human resource
development activities related to nuclear nonproliferation andsecurity, in cooperation with relevant domestic and overseasorganizations (Fig.11-1).

Support for Government Policy Formulation Basedon Our Expertise
As a think tank in this area, we conduct research on measures
to ensure nuclear nonproliferation and security on the backendof the nuclear fuel cycle. In particular, for the direct disposalof spent nuclear fuel, we conducted the literature review andthe site survey for the disposal plan in Finland, and for thesafeguards and nuclear security measures of the International
Atomic Energy Agency (IAEA). Moreover, we have examined
institutional measures such as safeguards to be reviewed in the
final disposal site.

Development of Nuclear Nonproliferation Technology
for Japanese and International Applications
We have been examining the concept of nuclear proliferation
resistance and evaluation methods by participating in aninternational project on innovative reactors and the nuclear fuelcycle, as well as the Generation IV International Forum (GIF).
We have been conducting joint research toward advanced
nuclear nonproliferation and security with the U.S. Departmentof Energy (DOE). We also ratified the new project and discussed
the new research themes during the 26th permanent coordinatinggroup meeting in March 2014. We have been developing
nondestructive assay technology for material accountancy ofmolten fuels in the Tokyo Electric Power Company, Incorporated
Fukushima Daiichi Nuclear Power Station (1F). The technology
is based on measurement of passive γ-ray emission from fission
products coexisting with nuclear materials. We analyzed theproperties of γ-rays leaking from a canister storing molten fuels.
From the securing broad options perspective for final disposal,
we have initiated a program to develop the safeguard technologythat will be applied to the spent fuel direct repository facilities as
an alternative option for high-level radioactive waste disposal.

Support for JAEA’s Transportation and Duties to
Research Reactor Fuels
We offer support to the nuclear transportation conducted by
JAEA R&D Institutes/Centers, and we have coordinated the
procurement of fresh fuels and disposal measures for the spent
fuel for our research reactors. Through these activities, we
contribute to the Global Threat Reduction Initiative (GTRI),
which has strengthened global nuclear security by promoting thesystematic return of highly enriched uranium to the United States.
New Efforts on Nuclear Security
On the basis of commitments made by the Japanese
government at the Nuclear Security Summit in April 2010,
we established the Integrated Support Center for Nuclear
nonproliferation and Nuclear Security (ISCN). ISCN conducts
capacity building and infrastructure development, focusing on
the Asian region. There were approximately 500 participants(approximately 320 of whom were from Asian states) in the
training courses on nuclear security and safeguards and otheractivities in FY 2013.
Furthermore, we have been conducting the research anddevelopment of technologies for the measurement to improve
nuclear material accountancy, detection of nuclear materials andnuclear forensics in cooperation with the United States, and willcontinue to contribute to international efforts by establishinghighly precise and reliable technologies. Topic 11-1 is one of
the results of the research on nuclear security.

Contributions to the International Community Basedon Our Expertise and Experience
To establish a global verification regime for nuclear tests, wehave been provisionally operating facilities of the international
monitoring system of the Comprehensive Nuclear-TestBan Treaty (CTBT) and the National Data Center (NDC).
In April 2013, when the Takasaki radionuclide monitoring
station detected xenon isotopes that were thought to haveoriginated from a North Korean nuclear test announced twomonths earlier, we contributed to the governmentʼs evaluation
using Japan’s CTBT national operation system by conducting
analysis using the NDC’s own methods.
EURATOM

IAEA
DOE

Development of Technology for
Nuclear Non-proliferation
Proliferation resistant nuclear technology andmethodology to evaluate proliferation resistance
Advanced safeguards technologies
Measuring control technology of meltdown fuelat Fukushima Daiichi Nuclear Power Station

Policy Research
Policy Research on the current
issue on the peaceful use of
nuclear energy and non-proliferation

Transportation andResearch Reactor Fuels Support

Domestic Human Resource
Development
Cooperative relations with universities
and related organizations
Support nationwide efforts of humancultivation

Support to JAEA's transportation of
nuclear materials
Coordination of procurement anddisposal of research reactor fuels

International Contribution
Advanced development of CTBT
nuclear test verification technology andestablishment of international
monitoring system
Technical support to disposal of Pu fromdismantled Russian nuclear weapons

Cooperation

Support to Capacity Building andinfrastructure Development in
Asian States
Enhancement of capacity building throughtraining and education
Support to infrastructure development,
including the development of legal system,
regulations, and manual

Development of Technologies on
Measurement, Detection, andForensics of Nuclear Material

Demonstration of NDA system, development of
nuclear resonance fluorescence NDA Technology,
and development of He3-alternative neutron detection
Establishment and improvement of nuclear forensics
technologies, and creation of domestic database

Cooperation
WINS

126

JAEA R&D Review 2014

Domestic
organizations

Relevant
ministries

Fig.11-1 JAEA activities toward thedevelopment of science and technology
for nuclear nonproliferation
We are going to contribute even more to thereinforcement of nuclear nonproliferationand security through the developmentof technologies for safeguards, nuclear
forensics, and the detection of nuclear
materials. We are also going to engage inpolicy research and analysis, and support
the capacity building of human resources
and infrastructure development in the Asian
region, as well as provisional operation of
the CTBT international monitoring system,
among other initiatives.

ASNO
CTBT: Comprehensive Nuclear-Test-Ban Treaty
WINS: World Institute for Nuclear Security
ASNO: Australian Safeguards and Non-Proliferation Office
D O E : Department of Energy

Development of Science & Technology for Nuclear Nonproliferation

11-1 A Probabilistic Expression for Physical Protection Performance in a Nuclear Facility
Component of PI

Event tree

Probability
Detection: P(Di)Communication: P(Ci)
P(C2) Response force
P(R|A2) arrival: P(R|Ai)

1

P(D2)
1-P(D1)(1-P(D1))
× P(D2) × P(C2) × P(R|A2)
・・・(1−P(Di))2
1-P(D2)i

P(D1) × P(C1) × P(R|A1)

P(Cn)
P(R|An)P(Dn)Barrier i
Detected
Undetected

Time line

1-P(Dn)

0

Gate

④

Fig.11-3 Probability distribution functions P(D)Sensors operate in an expected fashion when a signal of strength
greater than T appears. The type II error, β, is the distributionfunction of the signal plus the noise from zero to the threshold.

Number of trials: 5000(b) We calculated probability distributions, P(Di), P(Ci), and P(R|Ai),on the basis of the scenario

Limited area

Isolation zone
(Microwave ④) Protected area

Door
(IR ⑦)
Inner
wall

0.85

i =1
i =3
i =5
i =7
i =9

Door

⑩

(IR ⑨)⑨

Target⑧i=1_IR sensors
i=2, 4, 6_Microwave sensors
i=3, 5_IR sensors
i=7, 8, 9_IR sensors

500

Controlled building
Frequency(IR ⑧)(Microwave ⑦
⑥)

1000

0.8

(IR ⑤)⑥

Accuracy
high
low

1500

1000
500

0.9
P(Di) value

0.95

Door
(IR)

0
0.85

0.2

i =2
Time margini =4 insufficientsufficient
i =6
i =8
i =10

0.4

0.6

Accuracy
high
low

0.9

0.95

1

P(Ci) value(c) We calculated the probability
distribution of PI
1000

0

i =2
i =5, 7, 9, 10

500

1

0
Outer walli =1
i =3, 4i =6, 81000

0Controlled building area

Door
⑤

10

Fence (IR ①)

Vehicle Personnel (IR ③)
gate
gate

③

R
5

Signal strength (dB)Frequency②

Probability density
function, f s+n(R)

1-β(=P(D))
T(Sum)= PI

Adversary(Microwave ②)

0.20(a) We designed an assault scenario against ahypothetical nuclear facility①–⑩PP barrier (i = 1–10)①

Probability of
malfunction,α (= P fa)

0

Fig.11-2 Schematic of the EASI method
The left side indicates a simplified event tree diagram at the i point
barriers, and the right side indicates the calculation components of
PI. PI is the total probability of these calculation components.

Adversary’s pass
(Sabotage target)

rejection

0.4

× P(Dn) × P(Cn) × P(R|An)n

Signal + Noise
Noise
Thresholdadoption

Frequency

P(D1)Probability density
function, f n (R)

0.6

0.8

P(R|Ai) value

Outer wall

Uncertainty and variability→ Probability distribution

1

Frequency

P(C1)
P(R|A1)

Probability density− A Quantification Method for Physical Protection Performance −PP performancehighlow(maximum 1)(minimum 0)

500

0
0.6

0.7

0.8
PI value

0.9

1

Fig.11-4 Example of PI assessment in an assault scenario
This figure shows (a) an assault scenario against a hypothetical facility, (b) the performance of each component of PI (P(Di), P(Ci),P(R|Ai)) at all barriers (i=1–10), and (c) the probability distribution of PI, including the data’s uncertainty and variability.

Since the September 11, 2001 attacks, it has become
increasingly necessary to consider the physical protection(PP) of nuclear materials against adversarial attacks, i.e.,sabotage against a nuclear facility. It is important to assess
PP performance when one is designing or implementing PP
measures. In general, this is qualitatively evaluated; however,
if it is quantitatively assessed, including its uncertainty andvariability, it might be possible to provide better PP measures.
In this study, we focus on establishing a quantitative
expression for the probability of an adversary being interrupted,
PI, to assess PP performance. A simple calculation describing
PI in an Estimate of Adversary Sequence Interruption (EASI)is shown in Fig.11-2. The left side of this figure indicates asimplified event tree diagram for the PI calculation at the i point
barriers, and the right side indicates the calculation components
of PI. P(D) is the probability of a detection alarm (e.g., infrared(IR) sensors) being triggered; P(C) is the probability thatthe facility guard successfully interprets the alarm condition
using the PP equipment and successfully communicates withthe response force; and P(R|A) is the conditional probability
that the response force arrives prior to the completion of theadversary’s action sequence by recognizing the alarm.

In the EASI calculation, both P(D) and P(C) values are
considered as the evaluated values without uncertainty andvariability, and the P(R|A) value is calculated using a normal
distribution. To assess the PP performance more realistically, wedevelop a new quantification method for PI that expands EASI
by expressing the influence of uncertainty and variability as aprobability distribution.
First, we express P(D), as shown in Fig.11-3, using thestatistical type II error (β). P(D) is the distribution function area
where the probability density function of the signal and noise(fs+n) exceeds a threshold value (VT). Second, we express thehuman error affecting P(C) using a log-normal distribution.
Third, we re-define P(R|A) using the Poisson distribution toshow the margin of error in the response force arrival caused byuncertainty and variability.
We design a hypothetical assault scenario and assess PI,
as shown in Fig.11-4. Furthermore, all components of PI are
calculated from data based on the scenario. Moreover, the data’s
uncertainty and variability are expressed by normal random
numbers. Consequently, we can quantitatively assess PI using
our new method.

Reference
Terao, N., Suzuki, M., A Probabilistic Extension of the EASI Model, Journal of Physical Security, vol.7, issue 2, 2014, p.12-29.

JAEA R&D Review 2014127

P romotion of Collaboration
To make the technologies, patents and other research results, and facilities and equipment of the Japan Atomic Energy Agency(JAEA) widely available to society, we are promoting activities in three areas: “academia-industry collaboration,” “utilization of
intellectual property,” and “facility usage.” We are also creating a database of patents and intellectual property information held by
the JAEA.
The following table is a list of intellectual property (patents in foreign countries) in the fiscal year 2013.

Intellectual Property Held by JAEA
Patent Information
No.

Registered Country(Representative
Applicant Country)

1

France2

Department

R&D Institutes
and Centers

PROCESS FOR PRODUCINGMETHYLIMINOBISDIALKYLACETAMIDE

Nuclear Science andEngineering Directorate

Nuclear Science ResearchInstitute1250176

Feb.8, 2013

UK

DECLADDING METHOD AND DECLADDING DEVICE FOR
SPENT NUCLEAR FUEL REPROCESSING

Nuclear Cycle
Engineering Department

Nuclear Fuel Cycle
Engineering Laboratories

1496520

Sep.11, 2013

3

FranceDECLADDING METHOD AND DECLADDING DEVICE FOR
SPENT NUCLEAR FUEL REPROCESSING

Nuclear Cycle
Engineering Department

Nuclear Fuel Cycle
Engineering Laboratories

1496520

Sep.11, 2013

4

FranceMULTIPLAYER FILM COATED DIFFRACTION GRATINGS

Quantum Beam
Science Directorate

Kansai Photon Science
Institute2009470

Jan.23, 2013

5

UK

AUSTENITIC STAINLESS STEEL EXCELLENT IN
INTERGRANULAR CORROSION RESISTANCE AND STRESSCORROSION CRACKING RESISTANCE

Nuclear Science andEngineering Directorate

Nuclear Science ResearchInstitute2143815Jan.8, 2014

6FranceAUSTENITIC STAINLESS STEEL EXCELLENT IN
INTERGRANULAR COLLOSION RESISTANCE AND STRESSCORROSION CRACKING RESISTANCE, AND METHOD FOR
PRODUCING AUSTENITIC STAINLESS STEEL

Nuclear Science andEngineering Directorate

Nuclear Science ResearchInstitute2143815Jan.8, 2014

7

RussiaHYDROGEN COMBUSTION CATALYST AND PROCESS FOR
PRODUCTION THEREOF, AND HYDROGEN COMBUSTIONMETHOD

Fusion Research andNaka Fusion Institute
Development Directorate

2494811

Oct.10, 2013

8

USA

PROCESSES FOR PRODUCING NANO-SPACE CONTROLLEDPOLYMER ION-EXCHANGE MEMBRANES

Quantum Beam
Science Directorate

Takasaki Advanced
Radiation Research Institute8232353

Jul.31, 2012

9USA

HIGHLY PROTON-CONDUCTIVE POLYMER ELECTROLYTEMEMBRANES THAT EXCEL IN MECHANICAL STRENGTH
AND A PROCESS FOR PRODUCING THE SAME

Quantum Beam
Science Directorate

Takasaki Advanced
Radiation Research Institute

8372558

Feb.12, 2013

10

USA

POWER FEEDING CONTROL APPARATUS AND CHARGINGSYSTEM

Safety Administration
Department

Head Office(Project Promotion Sector)

8421278

Apr.16, 2013

11

USA

METHOD OF SYNTHESIZING ZIRCONIUM-LOADED FIBROUS
ADSORBENT MATERIALS HAVING PHOSPHORYL GROUPSAND REMOVAL OF OBJECTIONABLE SUBSTANCES USING
THE ADSORBENTS

Quantum Beam
Science Directorate

Takasaki Advanced
Radiation Research Institute

8445404 May 21, 2013

12

USA

POLYMER ELECTROLYTE MEMBRANE COMPRISINGALKYLETHER GRAFT CHAIN AND METHOD OF PRODUCING
THE SAME

Quantum Beam
Science Directorate

Takasaki Advanced
Radiation Research Institute8470895

Jun.25, 2013

13

USA

CLOTH-LIKE RADIOACTIVE MATERIAL ADSORBENT AND
ITS MANUFACTURING METHOD

Quantum Beam
Science Directorate

Takasaki Advanced
Radiation Research Institute8476188

Jul.2, 2013

14

USA

METHOD OF MANUFACTURING GEL USING
POLYSACCHARIDES AS RAW MATERIALS

Quantum Beam
Science Directorate

Takasaki Advanced
Radiation Research Institute

8481718

Jul.9, 2013

15

USA

METHOD OF MANUFACTURING GEL USING
POLYSACCHARIDES AS RAW MATERIALS

Quantum Beam
Science Directorate

Takasaki Advanced
Radiation Research Institute

8481719

Jul.9, 2013

16

USA

APPARATUS FOR DECONTAMINATING RADIOISOTOPECONTAMINATED SURFACE VICINITY REGION BY USE OFNONTHERMAL LASER PEELING

Quantum Beam
Science Directorate

Kansai Photon Science
Institute

8518331 Aug.27, 2013

17

USA

IMAGE SENSOR, SEMICONDUCTOR DEVICE AND IMAGESENSING METHOD

Nuclear Science andEngineering Directorate

Nuclear Science ResearchInstitute

8576319

Nov.5, 2013

18

USA

GEL FOR RADIATION DOSIMETER AND RADIATIONDOSIMETER WHICH USES THE SAME

Quantum Beam
Science Directorate

Takasaki Advanced
Radiation Research Institute8586366

Nov.19, 2013

19

USA

HYDROGEN GAS DETECTING MEMBRANE

Quantum Beam
Science Directorate

Takasaki Advanced
Radiation Research Institute8591828

Nov.26, 2013

20

USA

PREPARATION OF GELS DERIVED FROM CARBOXYMETHYLCELLULOSE ALKALI METAL SALT

Quantum Beam
Science Directorate

Takasaki Advanced
Radiation Research Institute8633254

Jan.21, 2014

21

USA

METHOD FOR SEPARATING RADIOACTIVE COPPER USINGCHELATING-ION EXCHANGE RESIN

Quantum Beam
Science Directorate

Takasaki Advanced
Radiation Research Institute8647595

Feb.11, 2014

128

Title

JAEA R&D Review 2014

Publication
Date of
Number Registration

About the Design of the CoverThe cover is designed to envisage a hopeful future shining in the sky, which is a clear blue color resembling the color of the JAEA logo. This isaccompanied by white hexagons similar to the pattern on a tortoise shell, an ancient Japanese symbol of peoples’ wishes for longer lives. Coincidentally,
this shape is the same as that of the core fuel assemblies of both the prototype fast breeder reactor “MONJU” and the high temperature engineering test
reactor “HTTR.”
The images on the cover show radiation measurement from the sky using the Unmanned Airplane for Radiation Monitoring System (top left) andresults of simplified simulations on the relocation behavior of simulated molten materials (bottom right).
The top left image shows the radiation monitoring system consisting of the radiation detector that we developed and the airplane that was developed by
JAXA. The remote operational range of the system is up to approximately 100 km from the ground station, and the flight time can reach approximately
six hours. We tested the system and confirmed its capability to map the air dose rate on the ground (Topic Chapter 1 introduction, p.10).
The bottom right image shows the results of simplified simulations on the relocation and accumulation behavior of the lower plenum of simulated core
melts comprising fuels and core structural materials (Topic 4-2, p.50).

ISSN 2188-1421

JAEA R&D Review
2014

Radiation measurement from the sky using theUnmanned Airplane for Radiation Monitoring
SystemResults of simplified simulations on the relocation
behavior of simulated molten materials

JAEA R&D Review 2014

129

JAEA R&D Review 2014

Published by

Japan Atomic Energy Agency in February 2015
Editorial Board
Chief editor: Masamichi Chino
Vice editor: Hiroshi Igarashi
Editors: Kazuhiko Ogawa, Yoshio Suzuki, Hirofumi Tomikawa, Makoto Tanaka, Yoshiaki Shikaze,

Jinya Katsuyama, Toshiki Maruyama, Takamasa Mori, Ryo Yasuda, Tetsuya Kawachi,
Yasuyuki Ishii, Kazuhiro Sawa, Nariaki Sakaba, Hiroshi Takada, Nariaki Uto,
Masayuki Naganuma, Takayoshi Nakatani, Chie Oda, Hirofumi Takashita, Yuu Ishimori,
Shinichiro Mikake, Minoru Okoshi, Mitsuru Kikuchi, Kensuke KonishiThis publication is issued by Japan Atomic Energy Agency (JAEA) on a yearly basis.
Inquiries about availability and/or copyright of the contents in this publication should be addressed
to Institutional Repositor y Section, Intellectual Resources Management and R&D Collaboration
Department, Japan Atomic Energy Agency (JAEA).
Address: 2-4 Shirane, Shirakata, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195, Japan
Phone: +81-29-282-6387, Facsimile: +81-29-282-5920, e-mail: ird-seika_shi@jaea.go.jp

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