ISSN 2188-1456

JAEA R&D Review
2013

Futaba-machi

3 km from 1F

1F

Okuma-machi
Dose rate at 1 m
above the ground (μSv/h)> 30
20 - 30

Tomioka-machi

10 - 20
5.0 - 10

0

1.5

3 km

［Normalized to March 21, 2013］

1.0 - 5.0
< 1.0

Dose rate distribution map aroundthe TEPCO’s Fukushima Daiichi NPS(1F), which was traced out using
an unmanned helicopter

Cs

Z
Y

X
Atomic-scale cesium absorption on
X a type of zeolite called mordenite

SiOMessage from the President

President Shojiro MATSUURA

We sincerely thank you for your understanding and continued support of our research anddevelopment activities. Since the foundation of the Japan Atomic Energy Agency (JAEA), thispublication has been brought out every year with 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 anddevelopment, the JAEA plays a primary role in steadily promoting research and developmentin nuclear technology and, in the wake of the accident at the Fukushima Daiichi Nuclear
Power Station of Tokyo Electric Power Company, Incorporated (TEPCO), has focused oninvesting efforts toward the recovery of the damaged reactor site and the reconstruction of
the affected area. Furthermore, with serious consideration of the fact that we have not beensuccessful in the maintenance of the prototype fast breeder reactor Monju and in preventing
the radioactive material leak at the Hadron Experimental Facility of Japan Proton Accelerator
Research Complex (J-PARC), we are in the process of promoting activities for establishing asafety culture in accordance with the fundamental policies of Japanese government. Moreover,the employees are working as a unit towards a structural reform for restoring the trust ofsociety in the institute, in accordance with the plan that the JAEA prescribed.
More than two and a half years have passed since the accident at the TEPCO’s Fukushima
Daiichi Nuclear Power Station; in this time, the JAEA has been engaged in post-accident
environmental remediation and decommissioning of the damaged reactors by fully utilizing
the integrated capability of the scientific and technical experts and its research facilities. To
deal with the environmental pollution by radioactive materials, Fukushima Environmental
Safety Center, of the Headquarters of Fukushima Partnership Operations, has been set up
as the base of our operations. Here, we have performed research and development activities
such as analyses of soil and water, tests accompanying decontamination such as the controlof amount and volume reduction of the contaminated waste from the soil after the accident,
demonstration tests on a decontamination model in high-dose regions, and studies on
decontamination efficacy evaluation methods. In addition, for research and development
related to the decommissioning of units 1 to 4 of the TEPCO’s Fukushima Daiichi Nuclear
Power Station, we have promoted tests on the long-term integrity of the fuel assembly inthe spent fuel storage pool, the characterization of the fuel debris for the preparation of its
removal, and the treatment and disposal of radioactive waste caused by the accident.

For future research and development of the fast breeder reactor cycle technology, including
Monju, we will proceed in accordance with the review of the nuclear energy policy ofJapanese government. We will prioritize the safety of Monju and will also perform operations
management and work toward international standardization of the safety design requirements
of fast breeder reactors. Moreover, for the activities on the backend of nuclear energy
utilization, which are necessary irrespective of the future nuclear energy policy, we have
performed research and development on decommissioning, and the treatment and disposal ofradioactive waste.
In the research and development aimed at the practical utilization of nuclear fusion energy,
we, as a domestic agency/implementing agency of Japan under roles based on the internationalframework (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 researchand development on quantum beam science, we have been producing a variety of results inareas ranging from the fundamentals of science and technology to industrial applications,through various quantum beam facility groups owned by the JAEA. Moreover, we haverecognized that ensuring nuclear safety is an important step in the use of nuclear power, andwe have diligently conducted nuclear safety research for the improvement of safety in theutilization of nuclear energy. In addition, we have promoted a wide variety of engineeringresearch 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 planthat the JAEA has set forth. With the increased expectations toward the JAEA’s role in nuclear
technology development after the TEPCO’s Fukushima Daiichi Nuclear Power Station
accident, we are now actively involved in medium-to-long-term human resource developmentfor nuclear technology, industry–academia–government collaborations, promotion ofinternational cooperation with overseas research institutes, and the dissemination of theachievements made in research and development, without failing to prioritize operations.
We are pleased that you will gain further understanding of the accomplishments of the
JAEA through this publication, and we thank you for your continued encouragement andguidance in our research and development activities.

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

8

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

10

1. Time Series Observations in the Coastal Regions off Northern Ibaraki		─ Factors Controlling Concentrations of Radiocesium in Seabed Sediment ─
2. Creating Dose Rate Maps over Wide Areas in Collaboration with Municipalities		─ Detailed Measurements of Air Dose Rates Using Car-Borne Survey ─
3. Measurement Method without Exposure to Radiation Dose─ Radiation Measurement of Area Surrounding 1F using Unmanned Helicopter ─
4. Surface Grinding of Gravel for Waste Reduction and Reuse─ Decontamination Test of Gravel by General-Purpose Machine ─
5. Supercomputer Solves Mystery of Strong Affinity of Soil to Radioactive Cesium─ First-Principles Calculation Reveals Mechanism of Cesium Adsorption to Soil ─
6. Specific Sorption of Radiocesium by Minerals
─ Toward Remediation of Soils Contaminated by Radiocesium ─
7. Transport of Radionuclides in Depth Direction
─ Investigation on Depth Distribution of Radionuclides in the Soil of Fukushima ─
8. Retention of Mobile Radiocesium in Forest Surface Soils
─ Quantifying Microbial Uptake and Abiotic Adsorption ─
9. Water Purifier for Securing Water
─ Development of Grafted Adsorbent for Cs Removal ─
10. Assessment of Doses to Inhabitants Living in Fukushima Prefecture─ Approach to Assessing the Doses Considering Areal Differences in Contamination and Interpopulational Differences in Lifestyle Habits ─
11. For the Safe Handling of Dehydrated Sludge Contaminated by Radioactive Cesium─ Dose Estimate for Workers and Public with Incineration and Disposal of Dehydrated Sludge ─
12. Inhibiting Corrosion of Metallic Materials in Spent Fuel Pools
─ Verification of Removal Effect of Dissolved Oxygen Using Hydrazine ─
13. Characterization of Fuel Debris in Severe Accident
─ Properties of Fuel Debris for Developing Defueling Operations ─
14. What Has Formed on the Fuel Debris Surface by Sea Water Injection?
─ High-Temperature Reaction Products between Simulated Fuel Debris and Salt Deposits ─
15. Prevention of Re-Criticality in Fuel Debris in the Damaged Cores─ Criticality Characteristics of Fuel Debris Containing Concrete ─
16. Nuclear Material Accountancy for Molten Core Material byγ-ray Measurement
─ Development of Nuclear Material Measurement Technology using PassiveγSpectrometry of Molten Core Material ─
17. Calculation of Decay Heat Causing Core Meltdown─ Three-Dimensional Decay Heat Analysis for the TEPCO’s Fukushima Daiichi NPS ─
18. What’s the Point of Analyzing the Fukushima Accident?
─ Review of Five Investigation Reports on Fukushima Accident ─
19. Mechanism of Cesium Absorption on Zeolite─ First-Principles Study of Structural Effects on Absorption in Zeolites ─
20. Decontamination of Radioactive Cobalt and Manganese with Radioactive Cesium		─ Simplified Treatment Process of Radioactive Liquid Waste by Adding Hexacyanoferrate(II) Ions ─
21. Comprehension of Radioactive Contamination inside the Reactor Buildings		─ Radioanalysis of Concrete Core Boring Samples ─

JAEA R&D Review 2013

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JAEA R&D Review

2 Research and Development of Advanced Nuclear System
R&D of Fast Reactor Cycle Technology

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1. Robustness of JSFR against Earthquakes and Tsunamis─ Effectiveness of Natural Convection in a Decay Heat Removal System ─
2. Measurement of Temporospatially Fluctuating Flow─ Development of Ultrasonic Flowmeter for Fast Reactor ─
3. Aiming at Improved Visualization Technology for Reactor Structures
─ Experiment on Sodium Wetting on USV Sensor Surface ─
4. Codification of Materials for Demonstration Fast Reactor in JSME Code
─ Establishment of Standards for Material Strength and Elevated Temperature Design for 316FR and Modified 9Cr-1Mo Steels ─
5. Investigation of Sodium/Water Reaction in Collaboration with France─ Sodium/Water Reaction Test for Steam Generator Heat Transfer Tube ─
6. Controlling Oxygen Content of U–Pu Mixed Oxide Fuel
─ Oxygen Potential Measurement and Application to Fuel Technologies ─

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Research and Development on Geological Disposal of
3 High-Level
Radioactive Waste
R&D Supporting the Technology and Reliability of Geological Disposal in Japan40

1. Spatial Distribution of Fractures in Granite─ Influence of the Initial Cooling of the Granitic Pluton ─
2. Advancing In Situ Rock Stress Measurement Techniques at the Mizunami Underground Research Laboratory
─ Case Study: Successful Overcoring Tests Excluding Groundwater ─
3. Long-Term Investigation of an Excavation Damaged Zone around a Gallery
─ Seismic Tomography Survey for Investigation of an EDZ ─
4. Regional Estimate of Fracture Permeability
─ Hydrogeological Investigation of Sedimentary Rocks ─
5. Dating Fault Activity
─ K-Ar Age of Fault Clay ─
6. Estimating Long-Term Evolution of Environment and Materials around Radwaste in Deep Underground─ Development of Assessment Methodology Considering Geochemical Changes in the Near-Field ─
7. Effect of Natural Organic Matter on Migration of Radionuclides
─ Metal Ion Binding Abilities of Deep Groundwater Humic Substances ─
8. Evaluation of Bentonite Alteration in a TRU Waste Geological Repository
─ Modeling of Smectite Dissolution Rate under a Range of Chemical Conditions ─

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4		Nuclear Fusion Research and Development

4

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Toward Practical Use of Fusion Energy49

1. Multi-Parameter Measurement Using Laser Polarimetry for Fusion Reactor
─ Simultaneous Measurement of Magnetic Field, Electron Density, and Electron Temperature ─
2. Manufacturing of ITER Divertor Prototype
─ First Step toward Manufacturing of Actual ITER Divertor ─
3. Success in Beryllide Pebble Fabrication
─ Development of Granulation Technology for Neutron Multiplier Needed for Production of Fusion Reactor Fuel ─
4. Accuracy Validation of Nuclear Data for Fusion Reactor Design
─ Toward More Accurate Fusion Reactor Design ─
5. Toward Construction of Intense Neutron Source for Fusion Reactor Conditions
─ Measurement of Thickness and Stability of High-Speed Lithium Flow ─

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JAEA R&D Review 2013

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JAEA R&D Review

6. Realization of Fusion Intense Neutron Source─ Loop Antenna Development for RF Coupling in the RFQ Linac ─
7. Progress in Satellite Tokamak Program Project─ Start of Assembly of JT-60SA ─
8. Progress in Predicting the Voltage Holding Capability for Large Negative Ion Source─ Key Factor for High-Energy Accelerator toward JT-60SA and ITER ─
9. Realization of Compact Joint for Superconducting Coil
─ Development of Butt Joint Technique for Central Solenoid of Fusion Device ─
10. Clarification of Hydrogen Isotope Dependence of Plasma Heat Diffusion
─ Contribution to the Prediction of Fusion Power in ITER ─
11. Fast Ion Transport Study Using HELIOS─ Fast Ion’s Dance with Alfvén Waves in ITER Plasma ─

5		Quantum Beam Science Research

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Research & Development on Quantum Beam Technology

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1. Enhancement of Thermal Stability of Nanocomposites
─ The Role of Nano-Additives on Order-Disorder Transition of Nanocomposites ─
2. Quantitative Analysis of Conformational Changes in Proteins upon DNA Binding─ Toward an Understanding of the Molecular Mechanisms of Protein-DNA Recognition ─
3. Clear Imaging of Tumor with D-Amino Acid
─ Development of a Novel Amino Acid Tracer, D-[18F]FAMT, for PET Diagnosis of Cancer ─
4. Use of Diamonds for Single Ion Detection
─ Real-Time Detection of Ion Impact Position by Fluorescence of Nitrogen Vacancy Centers ─
5. Toward Nondestructive Assay of Fissile Materials
─ Generation of 500 keV Electron Beam from DC Photoemission Gun ─
6. Relativistic Harmonics from Tenuous Plasma─ Discovery of a New Compact Coherent X-ray Source ─
7. Heatproof Seismic Sensors Guard Aging Nuclear Power Plants
─ Application of Femtosecond Laser Processing to Maintenance Study ─
8. How Does a Cocatalyst Enhance the Activity of the Oxygen Reduction Reaction?
─ Improvement of Fuel Cell Performance ─
9. Self-Assembled Structure of Coordination Species Clarified by Quantum Beam Techniques─ Development of Reprocessing Techniques by Elucidation of Solution States ─
10. Elucidation of Electronic Excitation in Novel High-Tc Superconductor
─ Resonant Inelastic X-ray Scattering Study of Iron-Pnictide Superconductor ─

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6		Nuclear Safety Research

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Evaluation of the Safety of Various Nuclear Facilities

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1. Evaluating the Amount of Fission Gas in Fuel with Higher Accuracy
─ Development of a Rate-Law Model of Fission Gas Bubble Growth ─
2. Predicting Crack Growth in Reactor Piping under Large Seismic Loading─ Proposed Method of Evaluating Crack Growth under Random Cyclic Loading ─
3. Measuring Amount of Nuclides in Spent Nuclear Fuel
─ Development of Method of Assaying Barely Measurable Elements in Spent Nuclear Fuel ─
4. Research on Thermal Aging Embrittlement of Cast Stainless Steels
─ Study of Thermal Aging Embrittlement in “FUGEN” Nuclear Reactor ─73

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7		Advanced Science Research
For the Evolution of Nuclear Science77

1. Electricity Generation by Nanomagnet Dynamics
─ Magnetic Power Inverter Based on Spinmotive Force ─
2. Toward Experimental Determination of the First Ionization Potential of the Heaviest Actinide
─ New Ionization Potential Measurement Using Surface Ionization Process ─
3. Manipulation of Electrons by Pressure─ Coupling between Lattice and Electrons in Superconductor URu2Si2 ─
4. Novel Mechanism of Radiation Damage to DNA
─ An Advance in Elucidating the Radiation Process in DNA Damage ─
5. Evaluation of Electron Spin Using Its Antiparticle
─ Development of a Positron Beam with the World’s Highest Spin Polarization ─

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8		Nuclear Science and Engineering Research

Promoting Basic R&D on Nuclear Energy and Creation of Innovative Technology to MeetSocial Needs
1. Method of Accurately Designing New Types of Reactors
─ Advanced Method Using Experiments (Extended Cross Section Adjustment) ─
2. Accurate Prediction of Nuclear Fragmentation by Energetic Radiation
─ Fragmentation Cross Section Measurement by New Method and Theoretical Model Development ─
3. Life Prediction of Material in Nuclear Reprocessing Plants
─ Corrosion Characteristics of Stainless Steel in Boiling Nitric Acid Solution Including Np ─
4. Discovery of Thermochromic Uranium Complexes
─ Coordination Chemistry in Ionic Liquids Using Time-Resolved Laser-Induced Fluorescence Spectroscopy ─
5. Attempt to Establish a Reprocessing System That Uses Monoamides
─ Continuous Extraction of Uranium and Plutonium Using Mixer-Settler Extractors ─
6. One-Pot Biofabrication of Protein-Decorated Gold Nanoparticles
─ Facile, Rapid, and Efficient Biofabrication and Immunoassay ─
7. Analysis of Age Determination of Individual Plutonium Particles
─ Contribution of Development of Ultra-Trace Analytical Technique to IAEA Safeguard Activity ─
8. Detailed Simulation of Transfer of Radioactivity in Land Surface Ecosystems
─ Development of Nuclide Transport Model for Atmosphere-Vegetation-Soil System ─
9. Subsurface Soils Participate in Global Carbon Cycle
─ Carbon Dynamics Revealed by Tracing “Bomb” Radiocarbon over Past Half-Century ─

9		Nuclear Hydrogen and Heat Application Research

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Research and Development on Naturally Safe HTGR and Nuclear Heat Application Technologies

93

1. Small Nuclear Reactor for Multiple Heat Applications with Attractive Safety Features
─ Conceptual Design of Small High-Temperature Gas Cooled Reactor for Developing Countries ─
2. Proposal for HTGR Safety Design Criteria upon User Request
─ Allow H2 Plant Construction Coupled to HTGR under Non-Nuclear Regulations ─
3. Increased Energy Efficiency for Hydrogen Production
─ Study of Ion Permeation Mechanism in Membrane of HI Concentrator ─

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JAEA R&D Review

Development of Decommissioning and Radwaste Treatment
10		Technology
		 Executing Decommissioning of Nuclear Facilities and Treatment and Disposal of Radioactive		Waste
1. Successful Development of Nitrate Decomposition Technology
─ For Safe Disposal of Low-Level Radioactive Waste ─
2. Simplification of Analysis of Actinides in Radioactive Wastes
─ Development of an Analytical Method using Capillary Electrophoresis ─98
99

11 Computational Science and E-Systems Research		97Computational Science Activity in Nuclear Energy Research and Development

100

1. Moving Hydrogen Enhances Cracking in Iron
─ Embrittling Effect of Mobile Hydrogen Clarified by First Principles Calculations ─
2. Properties of New Materials with Topology
─ Classification of the Properties in Superconductors by Topology ─
3. Massively Parallel Fusion Plasma Simulation using 105 Processors
─ Development of Latency-Hiding Communication Technique ─
4. Technology to Speed up Fluid Structure Interaction Simulation
─ Simulation Technologies for Flow-Induced Vibration in a Nuclear Power Plant ─

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Development of Science & Technology for Nuclear
12		Nonproliferation				Technology and Human Resource Development in the Area of Nuclear Nonproliferation and
Nuclear Security to Support Peaceful Use of Nuclear Energy

Development of Experimental Techniques / Facilities at
13
		 JAEA R&D Centers
1. Challenge of Industrial Application of Repair Technologies for FBR Heat Exchanger Tubes
─ Success in High-Accuracy Wire Feeding to Laser-Irradiated Spot ─
2. Radiocesium in the Environment
─ Radiation Situation in Tokai Two Years after the 1F Accident ─
3. Neutrons in Development of High-Performance Materials
─ Role of Deformation-Induced Transformation in TRIP Steels Revealed by Neutrons at J-PARC ─
4. Development of Ultraprecise Focusing Supermirrors with Elliptical Surface
─ Focusing Gain of 52 Achieved for Advanced Instrumentation at J-PARC ─
5. Beam Acceleration by Axially Symmetric Field
─ Development of J-PARC Annular-Ring Coupled Structure ─
6. Development of Domestic 99Mo Production by (n,γ) Method
─ Fabrication Technology for High-Density MoO3 Pellet ─
7. Corrosion Behavior of Metals in Sodium─ Compatibility of Zirconium Alloy with High-Temperature Sodium ─
8. Reliable Management of Instrumentation to Ensure Reactor Safety
─ Establishment of Damage Detection Method for Neutron Detector in High-Temperature Gas-Cooled Reactor ─
9. Microfabrication of Teflon with Ion Microbeams
─ Development of New Microfabrication Techniques for Polymer Surfaces ─
10. Excavation of Galleries at -500 m Depth─ Construction of Shafts and Research Galleries for the Mizunami Underground Research Laboratory ─
11. Investigation of Properties of Radon Hot Springs
─ Radon-Induced Biological Responses and Its Biokinetics ─
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About This Publication and the Outline of Organization of JAEA
This publication introduces the latest accomplishments of research and development (R&D) in the JAEA, which are chapteredby area of study. Each chapter corresponds to the activities of one R&D Directorate. As shown in the organization chart andthe map below, the R&D Directorates perform their activities through R&D Institutes and Centers, which are located widely inJapan as the sites for R&D. Some R&D directorates are located at one site, others are at two or more. The following outlines theactivities undertaken by R&D Directorates and R&D Institutes and Centers.
1.

2.

3.

4.

The Headquarters of Fukushima Partnership Operations is engaged in R&D to recover from the accident at the Tokyo
Electric Power Company, Incorporated Fukushima Daiichi Nuclear Power Station. Partnership operations for plant restoration
have set up “Special Teams for Technologies Development” at three centers of JAEA: the Nuclear Science ResearchInstitute, the Nuclear Fuel Cycle Engineering Laboratories, and the Oarai Research and Development Center. These centersconduct R&D to restore the interior of the nuclear power plant. Fukushima Environmental Safety Center in the Fukushima
prefecture is conducting R&D concerning decontamination technologies to restore the environment, develop outreachactivities concerning radiation, and study measurement of radioactive materials in the human body by whole-body counters.
In addition, Nuclear Plant Decommissioning Safety Research Establishment has been established, and we are working on theanalysis and research of radioactive materials and development of remote controlled robot and its demonstration plant.
The Advanced Nuclear System Research and Development Directorate is carrying out R&D aimed at commercializing
the fast breeder reactor (FBR) and its nuclear fuel cycle. The R&D of plant technology using the prototype fast breeder
reactor “MONJU” is being undertaken at the Tsuruga Head Office (Fast Breeder Reactor Research and Development Center),
R&D for innovative FBR technology is being conducted at the Oarai Research and Development Center, and R&D on
manufacturing plutonium fuel and reprocessing spent FBR fuel is being conducted at the Tokai Research and Development
Center (Nuclear Fuel Cycle Engineering Laboratories).
The Geological Isolation Research and Development Directorate is carrying out multidisciplinary R&D aimed at improving
the reliability of geological isolation of high-level radioactive waste in Japan. A particular focus involves establishingtechniques for investigating the deep geological environment through research on crystalline rocks at the Tono Geoscience
Center and research on sedimentary rocks at the Horonobe Underground Research Center. At the Tokai Research andDevelopment Center, the focus is on improving technologies for designing disposal facilities and safety assessment. In addition,
work has been ongoing to develop a next generation knowledge management system based on the above R&D activities.
The Fusion Research and Development Directorate is performing fusion R&D as a domestic agency of the International
Thermonuclear Experimental Reactor (ITER) project and an implementing agency of the Broader Approach (BA) activities. The
procurement activity of the ITER project, the upgrade of JT-60 to a superconducting machine as the BA activity, fusion plasma
research, and R&D on various element technologies are carried out in the Naka Fusion Institute. Moreover, the International
Fusion Energy Research Center project and the Engineering Validation and Engineering Design Activities of the International
Fusion Material Irradiation Facility as the BA activity are performed mainly in the Aomori Research and Development Center.

Japan Atomic Energy Agency -Outline of OrganizationResearch and Development Directorates
Nuclear Safety Research Center

President
Executive Vice President
Executive Director

AuditorManagement Sector
JAEA Reorganization Headquarters
Monju Reorganization Headquarters

Policy Planning and Administration Department
General Affairs Department
Auditor Office

Advanced Science Research Center

Legal Office
Personnel Department
Financial Affairs Department

Nuclear Science and Engineering Directorate

Research and Development Institutes/Centers
Tsuruga Head OfficeContract Department

Quantum Beam Science Directorate

Fast Breeder Reactor Research and Development Center
Fugen Decommissioning Engineering Center

Safety Administration Department

Fusion Research and Development Directorate

Tokai Research and Development Center
Nuclear Science Research Institute
Nuclear Fuel Cycle Engineering Laboratories
J-PARC CenterIndustrial Collaboration Promotion Department

Oarai Research and Development Center

International Affairs Department

Advanced Nuclear System Research and Development Directorate

Naka Fusion InstituteTakasaki Advanced Radiation Research Institute
Geological lsolation Research and Development Directorate

Kansai Photon Science InstituteHoronobe Underground Research Center

Nuclear Cycle Backend Directorate

as of October, 2013

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JAEA R&D Review 2013

Project Promotion Sector

Headquarters of Fukushima Partnership Operations
Public Relations Department

Construction Department
Intellectual Resources Department

Department of Science and Technology for Nuclear Material Management
Center for Computational Science & e-Systems
Nuclear Human Resource Development Center

Tono Geoscience Center

Nuclear Emergency Assistance and Training Center

Ningyo-toge Environmental Engineering Center

Low-level Radioactive Wastes Disposal Project Center

Aomori Research and Development Center

Integrated Support Center for Nuclear Nonproliferation and Nuclear Security

5.

6.

7.

8.

9.

10.

11.

12.

13.

The Quantum Beam Science Directorate is engaged in research using neutron facilities in the Tokai Research andDevelopment Center (Nuclear Science Research Institute) and the J-PARC Center. Work using electron beams, gamma ray,and ion beam facilities is done in the Takasaki Advanced Radiation Research Institute. Research using lasers and synchrotron
radiation is performed at the Kansai Photon Science Institute.
The Nuclear Safety Research Center is in charge of safety research for supporting national nuclear safety regulatory bodies
that regulate nuclear power plants, nuclear fuel cycle facilities, and radioactive waste disposal facilities. This work is done at
the Tokai Research and Development Center (Nuclear Science Research Institute) and at the Tsuruga Head Office.
The Advanced Science Research Center conducts pioneering research in basic fields of nuclear power science, mainly
through the Tokai Research and Development Center (Nuclear Science Research Institute) and the Takasaki Advanced
Radiation Research Institute.
The Nuclear Science and Engineering Directorate is engaged in key and basic research on various elemental technologies
that support the use of nuclear power. These efforts are carried out in the Tokai Research and Development Center (Nuclear
Science Research Institute) and the Oarai Research and Development Center.
The Nuclear Hydrogen and Heat Application Research Center conducts R&D on technology for the use of hightemperature heat supplied from naturally safe, high-temperature, gas-cooled reactors and technology for using this heat toproduce hydrogen. These studies are done in the Oarai Research and Development Center.
The Nuclear Cycle Backend Directorate develops technologies for safe and rational decommissioning of nuclear power
facilities as well as measures for processing and disposing radioactive waste. This work is performed in the Tokai Researchand Development Center.
The Center for Computational Science & e-Systems develops pioneering simulation technology and basic technology incomputational science as well as operates and maintains computer equipment. This occurs mainly in the Tokai Research andDevelopment Center (Nuclear Science Research Institute).
The Department of Science and Technology for Nuclear Material Management and the Integrated Support Center for
Nuclear Nonproliferation and Nuclear Security develop technologies for nuclear nonproliferation and safeguards to ensure
the peaceful use of nuclear energy. These developments are done in the Tokai Research and Development Center (Nuclear
Science Research Institute) and the Techno Community Square Ricotti.
The R&D Institutes and Centers, located at 11 sites in Japan, manage and improve the performance of facilities andequipment to support the abovementioned R&D Directorates in safe and efficient R&D activities.

R&D Institutes /Centers of JAEA
Aomori Research andDevelopment Center

Horonobe Underground Research Center
Tsuruga Head Office

R&D of Geoscientific
Research and Geological
Disposal of HLW

Broader Approach Activities, Support
for Nuclear Fuel Cycle Business,
Decommissioning of Nuclear Ship

Prototype Fast Breeder Reactor MONJU,
Decommissioning of Advanced Thermal
Reactor FUGEN

Fast Breeder Reactor Fugen Decommissioning
R&D Center (MONJU) Engineering Center

Tono Geoscience Center
R&D of Geoscientiﬁc Research inorder to build a ﬁrm scientiﬁc andtechnological basis for Geological
Disposal of HLWTokai Research and Development Center
Nuclear Safety Research, Nuclear Science and Technology,
Neutron Science, Nuclear Fuel-Cycle Technology Development,
Radioactive Waste Treatment and Disposal Technology
Development, Human Resource Development, etc.

Nuclear Science
Research Institute

Nuclear Fuel Cycle
Engineering Laboratories

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

J-PARC Center

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

Kansai Photon Science Institute
Photon & Synchrotron
Radiation Science

Oarai Research and Development Center
Naka Fusion Institutec RIKEN/JASRI

Takasaki Advanced Radiation
Research InstituteFusion R&D, ITER Support,
Broader Approach Activities

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

R&D of Radiation
Applications

JAEA R&D Review 20139

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

Environmental Remediation and Plant Decommissioning
1.Radiation monitoring ofenvironment

2.Environmental restoration
activities

Environmental
Remediation

3.Human resource
development

4.Communications
and public relations

5.Measurement andestimation of exposed dose

Fig.1-1 JAEA’s post-accident efforts at environmental remediation for recovery of Fukushima
These activities are discussed on the web page of the Headquarters of Fukushima Partnership
Operations of the JAEA. (https://fukushima.jaea.go.jp/en/)

Environmental Remediation
As a designated public institution established under the
Basic Law on Natural Disasters, we started taking actions
immediately following the Great East Japan Earthquake on
March 11, 2011. We have undertaken diversified actions
such as dispatching technical experts. As countermeasures tothe problems that are still continuing now, we are taking thefollowing actions (Fig.1-1) for the recovery of Fukushima.
Radiation Monitoring of Environment
It is necessary to grasp the precise contamination status
and radiation dose rate of the radioactive material in order
to assess the health effects and develop a decontamination
plan. Therefore, a continuous investigation of the radioactive
cesium (Cs) density in the seabed sediment (Topic 1-1) is being
performed. In addition, under a contract with MEXT, air dose
rate maps were obtained by a car-borne survey in East Japan(Topic 1-2) and by airborne monitoring around the Tokyo
Electric Power Company, Incorporated (TEPCO) Fukushima
Daiichi Nuclear Power Station (NPS) (Topic 1-3). In addition,measurement of the soil radioactivity distribution, aerial
monitoring throughout Japan, and measurement of purified
water in the Japan Atomic Energy Agency (JAEA) branchareas are being performed.
Environmental Restoration Activities
Quick decontamination is necessary for the recovery of theenvironment. Under a trust from the Cabinet Office, guideline
formulation, decontamination catalog development, anddecontamination technology verification projects are being
performed. Waste reduction by decontamination of gravel(Topic 1-4) is being studied. By evaluating the changes in the
topographical characteristics and vegetation distribution, along-term environment change research project in Fukushima
has begun. Clarification of the soil contamination mechanism(Topics 1-5, 1-6, 1-7) and a study of the mobile Cs maintenancemechanism in forests (Topic 1-8) are being performed. In
10

JAEA R&D Review 2013

addition, the development of a drinking water purifier (Topic 1-9)
and the study of Cs behavior inside and outside of theincinerator are under way. On the basis of the above studies,
investigations toward the optimization and increased efficiency
of decontamination for environmental recovery will be
continued in the future.
Human Resources Development
Upon request from Fukushima prefecture, we responded bydispatching lecturers for a decontamination work lecture class
led by the Nuclear Human Resource Development Center.
We held the class 15 times, and 7819 persons completed
it by March, 2013. In addition, under the support systemdeveloped upon request from the Cabinet Office, Ministryof the Environment, we acted to spread knowledge about
decontamination by dispatching experts to municipalities andcooperating in decontamination plan development, performing
decontamination technology consultation and instruction, andproviding inhabitant briefing session support.
Communications and Public Relations
We dispatched experts who explained radiation data anda scientific interpretation method and conducted meetings toanswer typical questions about radiation for about 1700 groups,
for example, all the nursery schools, kindergartens, elementary
schools, and junior high schools in Fukushima prefecture. We held
these meetings in 210 places for 17286 persons by March, 2013.
Measurement and Estimation of Exposed Dose
We measured the internal radiation exposure using the wholebody counter (WBC) at the Tokai Research and Development
Center and a mobile WBC car for inhabitants upon requestfrom Fukushima prefecture. By the end of March, 2013, we
had measured 41043 people. In addition, the radiation exposure
doses for inhabitants under different conditions (Topic 1-10)
and for people who treated sludge including radioactive Cs(Topic 1-11) were evaluated.

Council for the Decommissioning of TEPCO s Fukushima Daiichi NPS
Chairman :Vice-chairman :Commissioner :Minister of Economy, Trade and Industry
Vice Minister of Economy, Trade and Industry
Senior Vice Minister of Education, Culture, Sports, Science and Technology, Japan Atomic
Energy Agency (JAEA) Chief Director, Executive President of TEPCO, President and CEOof Toshiba Corporation, Representative Executive Officer and President of Hitachi, Ltd.
Regulatory authority : Nuclear Regulation Authority, Chairman
Observer :Vice Governor of Fukushima prefecture※ The organization of the conventional “Government of Japan and TEPCO Council on
Mid-to-Long-Term Response for Decommissioning” has been
Committee on countermeasures forenhanced by the participation of concerned agencies involved in R&D.

contaminated water treatment

Notice and reporting on theconcept of facilities

Research base facility(JAEA)
Radioactive material
analysis and researchfacility
Remote-controlled
equipment and device
development facility

Notice and reporting on the
research and development plan

Organization to manage R&D
Candidate members: JAEA, National Institute of Advanced Industrial
Science and Technology, Toshiba, Hitachi GE,
Mitsubishi Heavy Industries, TEPCO, and otherpower companies (with a total of 17companies)<Advice of both domestic and overseas experts>
Collaboration−Review of the system for gathering wisdom in Japan and from abroad
andutilization−Technical advice for R&D(2 to 3 international advisors will be employed)(International Decommissioning Experts Group (with 6 experts) will be organized)

Fig.1-2 Organizational chart of research and development promotion
We are studying the establishment of research base facilities and conducting research and development based on the national
R&D plan. [Council for the Decommissioning of TEPCO’s Fukushima Daiichi Nuclear Power Station (June 27,2013)]

Plant Decommissioning
Immediately after the accident, we began providing technicaladvice to the government and TEPCO, and conducting researchfor decommissioning and radioactive waste management into
the mid- and long-term.

Diet of Japan, and the Rebuild Japan Initiative Foundation, as
well as TEPCO, to investigate the accident and published theirrespective reports. We reviewed these reports to gain insights
useful for near-term regulatory activities, including accident
investigation (Topic 1-18).

Removal of Spent Fuel and Fuel Debris
In the accident, as sea water was injected into the spent
fuel pool for emergency cooling, the possibility of corrosion
problems with various materials was considered. Therefore,hydrazine was added to salt-containing water in order to reducethe dissolved oxygen. We evaluated the effect of irradiation on
deoxidation by hydrazine (Topic 1-12).
It is important to grasp the characteristics of the molten corematerials (the fuel debris) before defueling for safe handling,storage, processing, and disposal of the fuel debris. We areconducting R&D to investigate the characteristics using
simulated debris (Topic 1-13), examine the reactions between
the sea water used for emergency cooling and the fuel debris(Topic 1-14), study criticality control in order to prevent
re-criticality even if the form of the fuel debris or the water
volume changes during defueling (Topic 1-15), and determinematerial accountancy and control measures for the fuel debris
when it is removed from the core (Topic 1-16).

Radioactive Waste Management
A massive volume of contaminated water has been generated
as a result of the tsunami and the inflow of ground water into
the reactor buildings; this contaminated water is temporarily
stored at the plant site.
Some water treatment systems are being applied to remove
radioactive nuclides, e.g., cesium and strontium, from thecontaminated water. We are addressing the long-term storage,
processing, and disposal of secondary waste produced bytreatment of the contaminated water, which presents important
challenges; making computational analyses of the cesium
absorption mechanism in zeolites (Topic 1-19); and examining
ways to simplify the water treatment process (Topic 1-20).

Estimation of Status inside Vessel and Investigationof Cause of Accident
The fuel in the reactors reached a high temperature andbecame molten when the core cooling system ceased operationin the station blackout due to the tsunami. To elucidate theprogress of the melting phenomena in the reactor core, we
evaluated the nuclide inventory at the time of the nuclear
accident (Topic 1-17).
Several committees were established by the Government, theRemote Decontamination Technology
Decontamination technology that uses remote accessdevices has to be developed to reduce the radiation dose ofworkers in the reactor buildings. It is necessary for effective
decontamination to comprehend the component nuclides, theirpenetration into the floors or walls, and the distribution of theradioactive contaminants. We conducted a radioanalysis of theconcrete core boring samples taken from the reactor buildings(Topic 1-21).
Future Plan
We will study the establishment of research base facilities
and conduct research and development based on the national
R&D plan (Fig.1-2).

JAEA R&D Review 201311

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

1-1

Time Series Observations in the Coastal Regions off Northern Ibaraki－ Factors Controlling Concentrations of Radiocesium in Seabed Sediment －39 N

Onahama38

1200

Cs concentration(Bq/kg-dry)

Miyagi

Hirakata●

The TEPCO’s
Fukushima
Daiichi NPS

●

37

Ose

●

137

Fukushima

Kujihama●

Ibaraki

36
140

141

142

143 E

Tokai

●S4 ○S5●S3 ○S6●S2 ○S7●S1 ○S8
○S9

1000
800
600
400
200
0

3 4 5 6 7 8 9 10 11 12 1 2 3 4 5 6 7 8 9 10 11 12 1 2 3

2011

2012

2013

Time (Month・Year)

Fig.1-3 Sampling area and stations
Seabed sediments were collected in the region 70–110 km
south of 1F. Bottom depths of the stations were 30–100 m. A
series of sampling campaigns was started in June 2011, and all
sampling was conducted by R/V Seikai (monitoring ship of theJapan Atomic Energy Agency).

Fig.1-4 Time series of concentration of cesium-137 (137Cs)
in upper (0~3 cm) layer of seabed sediment137
Cs concentrations in seabed sediment decreased with timeoverall, but an episodic variation was observed at several
stations. It can be considered that radiocesium-bound finesediment is transported continuously near the seabed.

For several months after the accident at the TEPCO’s
Fukushima Daiichi NPS (1F), seawater containing accidentderived radiocesium flowed into the coastal regions off
northern Ibaraki (Fig.1-3). We therefore set nine stations in
the region and have been investigating the temporal variationin the radiocesium concentration in seabed sediment andthe factors controlling the concentration by time-series
observations.
The observed concentrations of cesium-137 (137Cs) in theupper (0~3 cm) layer of seabed sediment ranged between
16 and 1020 Bq/kg (Fig.1-4). Although the concentrations
decreased with time overall, a temporarily high concentration
was observed at several stations.
In the shallow regions (less than 50 m of bottom depth),
most of the sedimentary radiocesium was adsorbed onto
coarse (>75 μm) components, but the concentrations were low.
On the other hand, a portion of the radiocesium was adsorbed
onto fine-grained sediment, which is easily remobilized bybottom currents, with higher concentrations. From thesecharacteristics, it can be considered that the episodic variationof radiocesium concentration indicates temporal accumulationof fine-grained particles near the seabed associated with thebottom currents.
Depth profiles of 137Cs in the sediment showed that, inthe shallow region, a large proportion of the 137Cs was
accumulated in the deep (>3 cm) layers of the sediment. The
higher concentrations of sedimentary radiocesium in the
deep layers are attributable to a higher contact probabilityof dissolved radiocesium with sediment, as well as efficientvertical transport of radiocesium to the deeper layer of
sediment via bioturbation.
We also found that most of the radiocesium in the coastal
sediment was incorporated into lithogenic (i.e., clay)
components, and this incorporation was almost irreversible.
These characteristics would be a primary reason for sucha slow decrease in the concentrations of sedimentary
radiocesium.
The findings of this study will be applied to a simulation
model that predicts the distribution of radiocesium on theseafloor.

Reference
Otosaka, S. et al., Sedimentation and Remobilization of Radiocesium in the Coastal Area of Ibaraki, 70 km South of the Fukushima Dai-Ichi Nuclear Power
Plant, Environmental Monitoring and Assessment, vol.185, issue 7, 2013, p.5419-5433.

12

JAEA R&D Review 2013

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

1-2

Creating Dose Rate Maps over Wide Areas in Collaboration with Municipalities－ Detailed Measurements of Air Dose Rates Using Car-Borne Survey －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

Morioka

Akita

150

km

Sendai
Yamagata

80 km

Niigata
Fukushima

20km

The TEPCO’s Fukushima
Daiichi NPS (1F)

NaganoMaebashi

Utsunomiya
Mito

SaitamaKofu

Tokyo
Yokohama

Shizuoka

Chiba

Sep. 2012

Fig.1-5 Example of large-scale car-borne survey data using KURAMA
Air dose rates measured every 3 seconds on roads are averaged in 100 m squares, and dose rate
ranges are shown by different colors.

In order to properly estimate the effects of radionuclides
released by the accident at the TEPCO’s Fukushima Daiichi
NPS (1F) on the environment and human health for a long
period and take appropriate countermeasures, it has beennecessary to obtain detailed information on the air dose rates
over wide areas. We established a system that can obtain alarge amount of air dose rate data in a short period using a carborne survey, and have periodically constructed detailed air
dose rate maps.
In the car-borne survey, the mobile survey system KURAMA-IIdeveloped at Kyoto University was employed. The system iscompact and easy to operate; moreover, it has the advantage thatthe obtained air dose rates and positional data are transferredthrough a cellular phone network, and stored on a data storage
server in real time. To ensure the quality of the measurements
made using the KURAMA-II system, we developed aspectrum-dose conversion operator (G (E) function) necessary
for accurate dose rates measurements, and have performed
calibration tests with standard radiation sources.
Measurements are intended to be performed over wide
areas where a certain amount of radiocesium could have been
deposited in the accident at 1F. Consequently, extended areas
in east Japan from Iwate Prefecture to Yamanashi Prefecture
have been covered in the car-born survey. In addition tomeasurements on main roads performed by ourselves, about
200 municipalities have been asked to perform detailed
measurements including those on small roads, resulting increation of detailed air dose rate maps over wide areas asshown in Fig.1-5.
The present study was implemented under a commission
from the Ministry of Education, Culture, Sports, Science andTechnology of Japan (MEXT).

Reference
Japan Atomic Energy Agency, Report on Establishment of Evaluation Methods for Long-Term Consequences due to Radioactive Substances Released by the
TEPCO’s Fukushima Daiichi NPT Accident. Commissioned by MEXT in FY2012, Ministry of Education, Culture, Sports, Science and Technology of Japan(MEXT), 2013, p.36-45 (in Japanese), http://fukushima.jaea.go.jp/initiatives/cat03/entry05.html
JAEA R&D Review 201313

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

1-3

Measurement Method without Exposure to Radiation Dose－ Radiation Measurement of Area Surrounding 1F using Unmanned Helicopter －CPU
Battery

Futaba-machi

Wireless
LAN

AMP, LV

3 km from 1F
PMT

Pre AMPLaBr3: Ce scintillation detector(38 mmΦ x 38 mm x 3 =0.13 L)

1F

Fig.1-6 Radiation monitoring system using an unmanned helicopter
The radiation monitoring system using an unmanned helicopter can savethe position information from DGPS and the count rates of the LaBr3: Ce
detector every 1 s. The position of the helicopter can be displayed in realtime. LAN: Local area network. PMT: Photomultiplier tube.
Okuma-machi
Dose rate at 1 m
above the ground (μSv/h)> 30
20 - 30

Tomioka-machi

10 - 20
5.0 - 10

0

1.5

3 km

1.0 - 5.0
< 1.0

Fig.1-7 Map of dose rate in area surrounding the 1F
This map was made by converting the acquired data into thedose rate at a height of 1 m above the ground. The dose rate
was normalized to that of March 21, 2013. The map includes
the dose rate of natural radiation.

After the accident at the TEPCO’s Fukushima Daiichi NPS(1F), a method by which the radiation in a large area can bepromptly measured was needed. We clarified the radiation
distribution in Japan by aerial radiation monitoring using
aircraft (a manned helicopter). However, because Japanese
regulations prohibited travel within a 3 km range of the
1F, aerial radiation monitoring using the aircraft was not
performed. Thus, we attempted to make a radiation distribution
map using a radiation detector installed in an autonomous
unmanned helicopter (AUH).
The AUH employed an R-MAX G1 made by Yamaha MotorCo., Ltd. The AUH consists of an unmanned helicopter used
for pesticide dispersion with an added capacity for autonomous
functioning (Fig.1-6). We developed a radiation monitoring
system that was installed in the AUH. We adopted the LaBr3: Ce
scintillation detector to obtain the energy spectrum of γ-rays.
This system could save not only the positional information
obtained by differential GPS (DGPS), but also the radiation
data, every 1 s. The data from the scintillator (gross countrates) were converted into the air dose rate at 1 m above theground using a pre-set parameter. A contour map based on theanalyzed dose rate at a 1 m height was made using Geographic
Information System (GIS) software. The unmanned helicopter
flew about 80 m in altitude above the ground at a speed of
8 m/s. The line spacing of the flight was 80 m for agreementwith the flight altitude. The measurement results obtained by
the AUH system exhibited a good correlation with those ofmeasurements by a survey meter on the ground.
The contour map of the dose rate around the 1F is shownin Fig.1-7. The high dose rate area extends not only in thesouthern direction but also in the western direction from the 1F.
Moreover, the high dose-rate area in the western direction was
divided into two areas. It can also be seen that the high doserate area extended in the northwestern direction. The shape ofthe distribution is attributed to the weather conditions at thetime of the accident and the topography.
This map is available to the public on the homepage of the
Nuclear Regulation Authority (http://fukushima.jaea.go.jp/
initiatives/cat03/pdf04/5-1-2.pdf).
The present study was conducted under a contract with
the Ministry of Education, Culture, Sports, Science andTechnology of Japan (MEXT) in fiscal year 2013.

Reference
Japan Atomic Energy Agency, Report of Investigation of “Study on Distribution and Migration of Radioactive Substances around Fukushima Daiichi Nuclear
Power Plant”, Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT), 2013, p.46-52 (in Japanese), http://fukushima.jaea.go.jp/
initiatives/cat03/pdf05/02-2.pdf

14

JAEA R&D Review 2013

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

1-4

Surface Grinding of Gravel for Waste Reduction and Reuse－ Decontamination Test of Gravel by General-Purpose Machine －100

Reduction rate (%)

Condition of surface asperity

80
60

Fig.1-9 Relationship between grinding
rate and reduction rate
In this test, we confirmed a reduction
rate of up to 66% relative to the surface
dose rate. We will research the correlation
bet ween the ground amount andreduction rate to determine the reduction
effect if the ground amount increases.

60 min40 min20 min40
20: Grinding time
0
0

5

10

15

20

Grinding rate (%)
Imaging graphics(Black part region indicates radioactive cesium)
BeforeDecontamination

After
Decontamination

Cross-sectional photographFig.1-8 Decontamination test by milling method
We tested gravel decontamination using a generalpurpose barrel finishing machine with 20 kg ofgravel, 20 ℓ of water, 150 rounds per min, and
20–60 min of grinding time per batch.

0

10

20

30 0

10

20

30

(unit: mm)

Fig.1-10 Distribution state of radioactive cesium before and after testing
Using an IP, we confirmed that Cs gathers mainly in the black mica parts anddecreases in the ground parts of the gravel after decontamination (grindingtime is 20 min).

In houses and public facilities, gravel (pebbles, etc.) withradioactive cesium (Cs) is laid on. To reduce the air dose rate,
the gravel is stripped and led to waste. Gravel with low-levelcontamination is reused after high-pressure water washing to
remove fine-grained soil attached to its surface. Considering
the high price of gravel compared to that of crushed stone
and the need for waste reduction, an effective method of
decontaminating gravel with high-level contamination is
expected to be selected.
Surface grinding of gravel has the potential for producing astrong decontamination effect. Therefore, we tried to find an
effective procedure for decontaminating gravel by rubbing it
against itself using a milling method in which the grinding timeand amount of water added are varied.
We observed the condition of gravel during and after thetest and confirmed that the surface of the gravel was groundproperly (Fig.1-8). However, in this test, there were no groundparts on the concave regions of the gravel because the milling
was done with only gravel and water, without grinding material.
The test results were as follows. When the grinding timewas 20 min, the mean grinding rate was about 4%, and themean reduction rate of the surface dose rate was about 37%.

Furthermore, when the grinding time was 60 min, the mean
grinding rate was about 10%, and the mean reduction rate of
the surface dose rate of the gravel was about 60% (up to 66%).
Both of these rates were low for these grinding times (Fig.1-9).
To confirm the cause of the low reduction rate, we observed the
distribution of Cs in the gravel before and after decontamination
using an imaging plate (IP). By comparing the observation ofa cross-section of gravel (granite) that was sliced in half andimaging graphics obtained with the IP (Fig.1-10), we confirmedthat the Cs was mostly attached to the black mica around thesurface of the gravel and penetrated slightly below the surface.
In this test, we used gravel that had been laid on heavily in
the garden and rooftop of Rifure-Tomioka. It took considerable
time to grind the gravel because it was slightly harder than
the gravel that was bought for a mock test. Furthermore, thereduction rate became lower than expected because the gravelcontained mica, which tended to absorb Cs. Various types ofgravel are laid on in houses. The color, shape, size, and hardnessof the gravel vary, so we must select reasonable conditions fordecontamination work. On the basis of the result of this test,
we will research decontamination methods and conditions forreusing gravel and reducing waste.

Reference
Japan Atomic Energy Agency, Decontamination Works in the Evacuation Zones concerning the Fukushima Daiichi Nuclear Power Plant Disaster【Part of
the Decontamination Pilot Projects】Report, 2012, p.107 (in Japanese).
JAEA R&D Review 201315

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

1-5

Supercomputer Solves Mystery of Strong Affinity of Soil to Radioactive Cesium－ First-Principles Calculation Reveals Mechanism of Cesium Adsorption to Soil －Weathered Clay Mineral

Weathered Part

Non−weathered Clay Minerals
Cs+

K+

Suitable
Water

Too narrow

Cs sorption is energetically unstable
Clay minerals do not adsorb Cs

Modeling of this part in supercomputer

Weathered Clay Minerals

Cs

K+

Cs+

Cs

Too wide

SuitableCs sorption is energetically stable
Weathered clay minerals
adsorb cesium

Non−weathered clay
minerals do not adsorb
cesiumClay minerals adsorb Cs

Fig.1-11 Schematic of clay mineral and model of its
weathered part
We modeled the weathered part of mica shown in the
figure and evaluated the energy difference between
the mineral before and after adsorption of cesium. As
a result, we showed that the weathered clay adsorbs
cesium, although the non-weathered clay does not.

Fig.1-12 Schematic of adsorption mechanism of cesium toweathered clay minerals
Non-weathered clay minerals have pockets whose sizes are
not suitable for cesium ions. Thus, the energy increases when
cesium ions are adsorbed by non-weathered clay minerals.
However, weathered clay minerals have pockets with sizes
suitable for cesium ions, so the energy decreases when they
adsorb cesium ions.

A large amount of radioisotopes was released into theenvironment by the accident at the TEPCO’s Fukushima
Daiichi NPS after the large earthquake and tsunami on
March 11, 2011. The government conducted large-scaledecontamination in the Fukushima area, which produceda large amount of waste. The handling of such a large
amount of waste became a new serious problem because
it is not easy to prepare storage places for it. To solve thisproblem, it is necessary to reduce the volume of the waste.
However, standard technology for such reduction has notbeen developed. One reason is the lack of knowledge about
the microscopic mechanism of adsorption. Weathered clay
minerals are known to adsorb cesium, but no one knows themechanism. To identify the mechanism, we performed firstprinciples calculations on the adsorption reaction of cesium toclay minerals.
First, we made a model of the weathered clay mineral in

the supercomputers at the Japan Atomic Energy Agency andcalculated the energy difference between the mineral before
and after adsorption (Fig.1-11). As a result, we found that
only weathered clay minerals adsorb cesium. In contrast, nonweathered ones do not exhibit adsorption. Further calculations
revealed the adsorption mechanism (Fig.1-12). Non-weathered
clay minerals have pockets with sizes suitable for thepotassium ion’s radius. On the other hand, when clay minerals
are weathered, they have wider pockets that are suitable for
cesium ions, whose radius is larger than that of potassiumions.
We revealed the adsorption mechanism of cesium to clay
minerals by first-principles calculations. We hope to develop
a simulation technique for desorption of radioactive cesium
from clay minerals on the basis of this knowledge of theadsorption mechanics and contribute to volume reduction of
the waste.

Reference
Okumura, M. et al., Mechanism of Strong Affinity of Clay Minerals to Radioactive Cesium: First-Principles Calculation Study for Adsorption of Cesium at
Frayed Edge Sites in Muscovite, Journal of the Physical Society of Japan, vol.82, no.3, 2013, p.033802-1-033802-5.

16

JAEA R&D Review 2013

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

1-6

Specific Sorption of Radiocesium by Minerals－ Toward Remediation of Soils Contaminated by Radiocesium －Elutriated< 106 µm

106 µm - 300 µm> 300 µm

Residual
Desorbed by 1 M HCl
Desorbed by 1 M KCl

1.0

Fraction of Cs desorbed

Fraction of Cs (%)

100
80
60
40
20
0

Soil A

Soil B

Soil C

Illite1.0

Kaolinite

1.0

0.8

0.8

0.8

0.6

0.6

0.6

0.4

0.4

0.4

0.2

0.2

0.2

0.0

10-9 M
10-5 M(Radiotracer)

0.0

10-9 M
10-5 M(Radiotracer)

0.0

Birnessite

10-5 M
10-9 M(Radiotracer)Fig.1-13 Size fractions of radiocesium in soils sampled in
Fukushima PrefectureRadiocesium was distributed in size fractions larger than theclay fraction, which was elutriated.

Fig.1-14 Desorbed fraction of radiocesium from minerals
at different cesium concentrations
When the Cs concentration was very low (10-9 M), more Cs
remained after treatment with desorption reagent solution.

Cesium (Cs) is known to be tightly associated with clay
minerals. In particular, radiocesium is tightly associated with
the interlayer of layered illite minerals contained in the small
particulate fraction; consequently, the associated radiocesium
cannot be desorbed by treatment with either acid or a cation
solution. This tight association resulted in immobility ofradiocesium after it was deposited to the soil in the accident at
the TEPCO’s Fukushima Daiichi NPS. For remediation of thecontaminated soil, the clay fraction was eliminated by sieving.
However, most of the radiocesium remained in some soils
after the sieving treatment (Fig.1-13).
To clarify the tight association of radiocesium with minerals
other than illite-like minerals, we conducted sorption anddesorption experiments with radiocesium on 17 minerals using
1 M KCl and 1 M HCl solutions as desorption reagent solutions.
Because the chemical concentration of radiocesium wasextremely low, two different concentrations of radiocesium

were considered: a contamination level of 8000 Bq/kg and aminor elemental level in the soil. At the contamination levelof radiocesium, 0.3 and 0.5 fractions of radiocesium remainedin kaolinite and birnessite, respectively, after the treatment(Fig.1-14).
In the natural environment, minerals are altered byweathering to generate tight association sites in the alteredminerals. Unfortunately, the number of tightly associated sites
is very limited in kaolinite and birnessite compared to thatin illite. When the concentration of radiocesium is extremely
low, most of the radiocesium is easily bound in the tightly
associated sites.
This finding of a tight association of radiocesium with nonillite minerals reveals a more effective remediation method forcontaminated soils. For instance, the soil fraction of less than
300 µm suggests a threshold for the appropriate remediation
method of the contaminated soil fraction.

Reference
Ohnuki, T. et al., Adsorption Behavior of Radioactive Cesium by Non-Mica Minerals, Journal of Nuclear Science and Technology, vol.50, no.4, 2013,
p.369-375.
JAEA R&D Review 201317

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

1-7

Transport of Radionuclides in Depth Direction－ Investigation on Depth Distribution of Radionuclides in the Soil of Fukushima －Total accumulation of 134Cs & 137Cs(Bq/m2)(converted to the values
as of April 29, 2011)

MEXT: Results of airborne monitoring by the MEXT
& U. S. Department of Energy, May 6, 2011

Table 1-1 Kd values of 137Cs and 131I onto each soil sample
Sorption experiments by a batch method were conducted for soil
samples taken at 11 locations (two different depths for each location).
Location

Depth
(cm)

pH

Kd(ml/g)Error(±)

pH

Kd(ml/g)Error(±)

Sandy

4−12

6.2

3.30E+03

2.38E+02

6.1

1.36E+00

1.34E−01

Clayish

21−38

6.5

6.10E+04

9.75E+04

6.5

9.24E−00

1.31E−01

10−20

6.5

1.30E+04

5.15E+03

7.6

1.76E+01

1.85E−02

32−43

6.2

4.24E+04

8.75E+04

6.1

2.37E+01

2.25E−01

Area without data

Teshichiro, Akougi, Namie T.
Shiobite, Akougi, Namie T.

Fukushima City

Kunugidaira, Akougi, Namie T.

Kanairo, Nihonmatsu City
Kozuka, Yamakiya,
Kawamata T.-1

Sandy

Kurabeishi, Hirusone, Namie T.

Kozuka, Yamakiya,
Kawamata T.-2

*1

5.8

1.87E+04

6.68E+03

5.8

4.44E+00

1.66E−01

35−40

*1

6.4

8.67E+03

1.63E+03

6.2

2.84E+01

2.92E−01*2

6.1

2.88E+03

1.73E+02

6.2

1.84E+01

2.08E−01*2

7.5

3.37E+03

2.40E+02

6.4

2.76E+01

2.58E−01

5−15

6.8

1.51E+04

6.52E+03

6.6

1.25E+00

8.78E−02

33−41

7.3

4.31E+04

4.14E+04

7.1

5.52E−01

1.25E−01

8−18

5.8

2.24E+03

1.19E+02

5.8

3.08E+01

5.18E−01

40−50

5.5

2.37E+03

1.26E+02

5.5

2.33E+01

4.12E−01

8−19

6.7

4.01E+03

3.56E+02

6.6

1.20E+01

1.90E−01

55−65

6.2

2.17E+04

1.60E+04

6.2

1.31E+01

1.54E−01

5−15

5.6

2.84E+03

1.86E+02

5.6

8.78E+00

1.35E−01

20−30

5.6

2.97E+03

2.03E+02

5.6

1.30E+01

1.65E−01

7−15

6.5

2.98E+03

1.92E+02

6.4

9.31E+01

6.58E−01

29−36

6.1

7.43E+03

1.25E+03

6.2

5.99E+01

5.18E−01

10−23

5.8

2.16E+03

1.10E+02

5.7

1.42E+02

1.47E+00

31−44

5.7

2.18E+03

1.07E+02

5.6

1.67E+01

3.08E−01

10−15

6.0

2.08E+03

9.85E+01

6.1

1.19E+01

1.00E−01

30−35

6.0

2.86E+03

1.85E+02

6.1

1.03E+01

1.36E−01

Sandy

10−25

6.5

2.16E+03

1.08E+02

6.6

2.08E+01

3.16E−01

Clayish

33−46

6.5

3.53E+04

4.79E+04

6.5

2.58E+00

1.47E+01

Sandy

5−15

21−30

Ohshimizu, Yamakiya,
Kawamata T.

Kozuka, Yamakiya, Kawamata T.-1

1F

Kozuka, Yamakiya, Kawamata T.-2
Matsukiyama, Shimotsushima, Namie T.-2

Mizuzakai, Tsushima, Namie T.

Matsukiyama, Shimotsushima, Namie T.-1

Fig.1-15 Locations investigated (although the distribution
map was published on May 6, 2011, the values are those
ones converted to as of April 29, 2011)
The investigation was conducted at 11 locations in
Nihonmatsu City, where 131I was detected from soil at thattime, and Kawamata and Namie Towns, where R114 runs
along the valley of Abukuma Highland. The sites are locatednorthwest of 1F in an area through which a plume of RNs is
considered to have passed.

The accident at the TEPCO’s Fukushima Daiichi NPS (1F),
which was caused by the Great East Japan Earthquake inMarch 2011, led to the release of volatile radionuclides (RNs)
that were deposited on the surrounding environment (e.g.,soils, forests, residential land) around Fukushima Prefecture.
The Ministry of Education, Culture, Sports and Technology(MEXT) initiated a project for making the distribution mapsof dose rate, etc. in cooperation with Ministry of Agriculture,
Forestry and Fisheries, JAEA, universities, etc. This work was
conducted as part of the project described above.
The Geoslicer investigation examined the depth distributionof RNs in the soil as of about three months after the 1F
accident. A total of 11 locations were selected in Nihonmatsu
City, Kawamata Town, and Namie Town (Fig.1-15). Plateshaped soil samples from a maximum depth of about 1 m(10 cm wide and 2 cm thick) were obtained, and soil samples
for RN analysis were taken after the soil was observed anddescribed. The γ decay nuclides were analyzed using a Ge
semiconductor detector.
Both cesium-134 (134Cs) and cesium-137 (137Cs) weredetected in all the locations investigated, and tellurium129m (129mTe) and silver-110m (110mAg) were detected only
in locations where the dose rates were high. For the depth
distribution of radiocesium (RC), at many of the locations
investigated, RC more than 99% of the inventory distributed
within depths of 10 cm and 14 cm in soil in the surface

Sandy

Mizuzakai, Tsushima,
Namie T.

Organic

Matsukiyama,
Shimotsushima, Namie T.-1

Sandy

Matsukiyama,
Shimotsushima, Namie T.-2

Organic

Teshichiro,
Akougi,Namie T.

Sandy

Shiobite,
Akougi, Namie T.

Sandy

Kunugidaira,Akougi, Namie T.

Organic

Kurabeishi,
Hirusone, Namie T.

137Cs(Cs+ : cation)

15−20

Kanairo, Nihonmatsu City
Ohshimizu, Yamakiya, Kawamata T.

Distribution coefficient(Kd)131I(I− : anion)

Soil

Error: counting error, ＊ 1: soil core 1, ＊ 2: soil core 2

layer (mainly sandy soil) and in soil at locations that aresupposed to have been used as farmland (mainly organicand clayish soils), respectively. The apparent diffusivities(Da) derived from penetration profiles near the surface layer
showed similar levels (10-11 m2/s) for all the RNs detected,
although the Da values tended to be higher in the soil at
locations that are supposed to have been used as farmland(Da = 0.1~1.5 × 10-10 m2/s) than in soil in the surface layer(D a = 0.65~4.4 × 10 -11 m 2/s). Because the distribution
coefficients (Kd) of cations (Cs+) and anions (I-) differedsignificantly (Table 1-1), the respective Da values were
estimated to differ significantly. However, they were at similarlevels for all the RNs; therefore, the Da values were believed
to be controlled by dispersion caused by rain water flow.
Thus, we can say that transport of RC in the soil is always
quite slow because the Kd values onto soil are quite large. The
transport of RNs in soil is strictly related to Kd, which depends
on the minerals constituting the soil, their contents, the content
of organic matter, and other factors. In particular, the clay
mineral component affects the reversibility and irreversibilityof sorption. For the long-term prediction of RC transport, it isessential to understand those details, and this remains as futurework.
The present study was sponsored by the Ministry of
Education, Culture, Sports, Science and Technology of Japan(MEXT).

Reference
Sato, H. et al., Investigation and Research on Depth Distribution in Soil of Radionuclides Released by the TEPCO Fukushima Dai-Ichi Nuclear Power Plant
Accident, MRS Online Proceedings Library, vol.1518, 2013, 6p.

18

JAEA R&D Review 2013

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

1-8

Retention of Mobile Radiocesium in Forest Surface Soils－ Quantifying Microbial Uptake and Abiotic Adsorption －Litter layer
0-3 cm3-10 cm> 10 cm

Non-fumigatedFumigated

Cs extractability (%)

15

80
60
40
20
0

10

5

137

137Cs inventory (kBq/m2)

100

0Soil 1

Soil 2

Soil 3

Soil 4

Soil 5

Soil 1

Soil 2

Soil 3

Soil 4

Soil 5

Fig.1-16 Vertical distribution of 137Cs in forest soils threemonths after the accident
At five forest sites in Fukushima prefecture, 50% –91% and6%–39% of the total 137Cs deposition were observed in theaboveground litter layer ( ) and the topmost (0–3 cm) soil layer( ), respectively.

Fig.1-17 Extractability of 137Cs in five forest surface (0–3 cm)
soils
Cesium-137 was extracted with K2SO4 from soils before (nonfumigated) and after (fumigated) chloroform fumigation. No
increase in the 137Cs extractability was observed after thedestruction of microbial biomass by chloroform fumigation.

The accident at the TEPCO’s Fukushima Daiichi NPS
released a huge amount of radiocesium into the environmentand caused serious radioactive contamination of forest
ecosystems. Our investigation, which was conducted in
Fukushima Prefecture three months after the accident,
revealed that >89% of the cesium-137 (137Cs) deposited onto
forest floors remained in the aboveground litter and topmost(0–3 cm) soil layers (Fig.1-16). In forest ecosystems affectedby the accident at the Chernobyl NPS in 1986, 137Cs hasreportedly still been observed in surface soils and has becomea major potential source for soil-to-plant transfer. Therefore,
a better understanding of the mechanisms of retention andmigration of 137Cs in forest surface soils is urgently needed.
We focused on microbial involvement in the retention
process of mobile 137Cs in forest surface soils because some
types of microorganisms in soils are believed to accumulate
137
Cs and thus influence the cycling of 137Cs in forest
ecosystems. Surface (0–3 cm) soil samples were collectedfrom five forest sites in Fukushima Prefecture one year after
the accident, and the following experiments were conducted.
First, soils were extracted with potassium sulfate (K2SO4)

solution, and the extracts were analyzed for 137Cs. The
extraction process liberated 2.1%–12.8% of the total 137Cs
as easily exchangeable ions from the soils. Two soils witha higher content of clay-sized particles and organic carbon
showed higher 137Cs extractability (Fig.1-17, Soils 4 and 5).
Next, the soils were fumigated with chloroform to destroy
microbial biomass and then extracted with K2SO4. Microbial
biomass was extracted from all of the soils; however, noincrease in the 137Cs extractability was observed (Fig.1-17).
The results indicate that the uptake of 137Cs by soil
microorganisms is less important for retention of mobile
137
Cs in the forest surface soils than ion-exchange adsorption
by abiotic components. We are now conducting a similar
experiment on litter and soil samples collected under different
meteorological conditions to fully understand the role of soil
microorganisms in controlling the mobility of 137Cs in natural
biotic systems.
The present study was sponsored in part by the Ministry of
Education, Culture, Sports, Science and Technology of Japan(MEXT).

Reference
Koarashi, J. et al., Retention of Potentially Mobile Radiocesium in Forest Surface Soils Affected by the Fukushima Nuclear Accident, Scientific Reports,
vol.2, no.1005, 2012, p.01005-1-01005-5.
JAEA R&D Review 201319

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

1-9

Water Purifier for Securing Water－ Development of Grafted Adsorbent for Cs Removal －Stability of the crosslinking
structure adsorbent (times)

16

9G10M

9G1M

(a)

4G1M
12
TAIC1M
8

4G10M

KranCsair ®
14G1M

4
Without
Crosslinking0(b)

TAIC10M

Water flow

Plastic plate
for retaining adsorbent
Activated carbon

0

0.3

0.6

0.9

1.2

Amount of Cs adsorbed (μmol/g-adsorbent)

Adsorbent

The notation shows the name and added amount ofcrosslinking reagent.
ex.)
9G 1M
Name of crosslinking
Added amountreagent(mol%/ GMA)
Fig.1-18 Efficiency of crosslinking structure for adsorbent
stability
Amount of Cs adsorbed and adsorbent stability (defined as theamount of Mo eluted from the adsorbent without crosslinking
structure divided by that with crosslinking structure). Highefficiency adsorbents are plotted in upper right corner.

Fig.1-19 (a) Water purifier for Cs removal and (b) internal
structure of cartridge
We have commercialized a cartridge-type water purifier.
This original cartridge, which incorporates the developed
adsorbent, is replaced easily.

More than two years have elapsed since the accident at theTEPCO’s Fukushima Daiichi NPS, and the radioactive cesium(Cs) concentration in environmental water has been reduced.
However, the results of a 2012 monitoring investigation of thedrinking water in Fukushima reported by the Ministry of theEnvironment showed that radioactive Cs was detected in 6 of
436 samples. Therefore, securing ‘water’ is demand.
We developed a grafted adsorbent for removing soluble
Cs immediately after this accident. The developed adsorbent
consisted of ammonium 12-molybdophosphate (AMP) as
the adsorbent functional group supported on a nonwoven
polyethylene fabric acting as the substrate by radiationinduced graft polymerization. In a previous result, thisadsorbent exhibited instability: The supported AMP released
easily from the adsorbent. In this study, the unstable adsorbent
was improved by introducing a crosslinking structure to
realize very little AMP elution. Introducing a crosslinking
structure to the grafted glycidyl methacrylate (GMA) chainof the adsorbent was attempted by adding three types ofpolyethylene glycol dimethacrylate (4G, 9G, or 14G) or

triallyl isocyanurate (TAIC) as the crosslinking reagent to a
monomer.
The efficiency of the improved adsorbent was investigatedby batch-wise adsorption tests. The adsorbent was cut into
pieces 1 cm2 (0.02 g) in size and immersed in 50 mℓ of 50 μg/ℓ
Cs solution under stirring for 24 h.
It was clarified that the amount of AMP eluted (calculated
as the eluted Mo) from the adsorbents with the introduced
crosslinking structure was less than that without the structure(Fig.1-18). Further, the amount of Cs adsorbed on mostadsorbents with the crosslinking structure was greater thanthat adsorbed without the structure.
An evaluation by an external organization verified that
this improved adsorbent satisfies the Food Sanitation Act
and Water Supply Act, and we developed a cartridge-typewater purifier called KranCsair® (Fig.1-19(a)). This original
cartridge, which incorporates the developed adsorbent,
is replaced easily (Fig.1-19(b)). A monitoring test of thedeveloped purifier was conducted in Fukushima. On the basisof the result, we will determine the final product specifications.

Reference
Mikami, T., Shibata, T. et al., Development of a Water Purifier by EB-Grafted Fiber, Setsubi to Kanri, vol.47, no.4, 2013, p.95-99 (in Japanese).

20

JAEA R&D Review 2013

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

1-10 Assessment of Doses to Inhabitants Living in Fukushima Prefecture－ Approach to Assessing the Doses Considering Areal Differences in

Contamination and Interpopulational Differences in Lifestyle Habits －

50

300

350

400

450

50

100

150

200

250

300

350

400

450

Individual doses per month (μSv)

0

Kawamata Town

250

Iitate Village(4)200

Iitate Village(3)150Minamisoma City

100

Namie Town

50

Evacuation Deliberatearea evacuation area

10

Katsurao Village

050th percentile

Tamura City

5

95th percentile

20

Hirono Town

1030

Kawauchi Village

Assessed value

Naraha Town(2)Measured value (Feb., 2012)

Assessed value

Naraha Town(1)Measured value (Feb., 2012)15

Futaba Town

20

40

Okuma Town

Population

25

Tomioka Town

Individual doses received
by outdoor workers

Individual doses received
by indoor workers

30

Estimated effective doses during one year
after the contamination occured (mSv)

35

・ The doses were calculated taking into account the contribution of cloudshine, groundshine and inhalationof radioactive nuclides. The contributions of noble gases were not considered.
・ The doses estimated for outdoor workers are shown. The doses for indoor workers were estimated to be
0.6–0.7 times those for outdoor workers.
・ The municipalities shown in this figure represent the places of residence on March 11, 2011.
・ Naraha Town (1) refers to people who evacuated from Naraha Town to Tamura City. The evacuation
facility for Naraha Town (2) is located in Aizu-Misato Town.
・ The people living in Iitate Village evacuated on May 29, 2011 (Iitate Village (3)) and June 21 2011 (Iitate
Village (4)).

Fig.1-20 Comparisons of measured and assessed valuesof individual doses
Doses to outdoor and indoor workers in Fukushima City
were assessed, taking into account areal differences incontamination and interpopulational differences in lifestyle
habits. A comparison of the measured and assessedindividual doses revealed good agreement, which indicatedthe validity of our assessment method.

Fig.1-21 Estimated effective doses during one year after
contamination occurred
Estimated doses during one year after contamination occurred
were around 1–10 mSv in many municipalities, including
the evacuated area and deliberate evacuation area. Doses
received by the residents of Namie Town and Iitate Village
were 4.9–48 mSv and 5.2–22 mSv, respectively.

The accident at the TEPCO’s Fukushima Daiichi NPS
resulted in widespread land contamination. Many residents
are exposed to radiation in their daily lives while they
continue to live in the affected areas. To manage the exposureof the population appropriately, the development of a dose
assessment method is urgently required.
The doses received by the population in daily life are
generally assessed using (1) data on the radionuclide
concentrations or exposure rates and (2) data on lifestyle
habits related to exposure pathways. If the average valuesof these data were used, it would be difficult to determine
individual doses. Thus, we developed a method of assessing
the dose distributions of a population living in the same area.
To develop the model, we measured the ambient dose
equivalent rates and prepared data on lifestyle habits, whichwere obtained from surveys on the periods of time spent insideand outside, with the cooperation of participants in various
types of occupations. After we determined the distributionsof the ambient dose equivalent rates and the lifestyle data, we
assessed the distributions of doses received by the population.

We also measured data on individual doses due to external
exposure with personal dosimeters. The validity of theassessment was confirmed by comparing the results of theassessment and the measured data (Fig.1-20).
In addition, we assessed the range of doses received by theevacuees during one year after the accident using the method
developed above. This assessment takes into account the doses
received during evacuations to multiple facilities and theevacuees’ lives after that. Fig.1-21 shows the result of the dose
assessment for outdoor workers. In most of the municipalities
included in the evacuated areas, the range of effective doses
is 1–10 mSv for outdoor workers during one year after theaccident. In Namie Town and Iitate Village, where evacuation
was delayed after contamination occurred, the effective doses
were higher than in other municipalities: 4.9–48 mSv and5.2–22 mSv, respectively. Our study provides ranges of dosesthat are more detailed than the results published by the World
Health Organization and are based on scientific grounds bytaking into account areal differences in contamination andindividual differences in lifestyle habits.

Reference
Takahara, S. et al., Assessment of Radiation Doses to the Public in Areas Contaminated by the Fukushima Daiichi Nuclear Power Station Accident,
Proceedings of International Symposium on Environmental Monitoring and Dose Estimation of Residents after Accident of TEPCO’s Fukushima Daiichi
Nuclear Power Station, Kyoto, Japan, 2012, p.212-220.
JAEA R&D Review 201321

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

1-11 For the Safe Handling of Dehydrated Sludge Contaminated by Radioactive Cesium－ Dose Estimate for Workers and Public with Incineration and Disposal of Dehydrated Sludge －Temporary Storage Scenario
Public near temporary storage
place(direct and skyshine radiation)

Unloading,transportationand landfill

Reuse of disposal site
after the closure

Disposal Scenario
Use of park
Residence

Temporary storage
and disposal

Dehydrated
sludgeDisposal ofincinerationashes

Migration ofradionuclide ingroundwater

Incineration

Ingestion of crops : adult

Park

External : children

Prohibition of reusing dehydrated sludge
and its ashes as cement material and so on

Minimum Value: 8900 Bq/kg

Ingestion of water :children

External : adult

Disposalsite use
Incineration
Scenario

Inhalation : worker
External : worker

Use of well water

Residence near
the incinerator

Direct ingestion : worker

Groundwater
migration(well)

Construction ofhouse,
agriculture

Maintenance of incinerator,transportation

Incineration of
dehydrated sludge

Landfillwork

Agriculture

Residence

Ingestion of crops : adult
External in agriculture :worker

It is necessary to forbid reusing
the disposal site for residence
and agriculture.

External : children

Necessity of restriction and management
for the disposal site reuse

External : adult
1E+1

1E+2

1E+3

1E+4

1E+5

1E+6

1E+7

1E+8

Radioactive cesium concentration in contaminatedmaterial corresponding to the effective dose criteria(Bq/kg)Fig.1-22 Scenarios for incineration and disposal of dehydrated
sludge contaminated by radioactive cesium
External and internal exposure pathways in the transportation,
incineration, and temporary storage of the sludge are specified
on the basis of information on actual work conditions andtreatment record for the sludge. Specified scenarios andpathways are guaranteed to be exhaustive through a comparison
with the pathways used in previous clearance level estimate.

Fig.1-23 Estimated results of limiting radioactive cesium
concentration to ensure safe incineration and disposal
Minimum radioactive cesium concentration in dehydrated
sludge and incineration ashes, corresponding to the effective
dose criterion (1 mSv/y for engaging in treatment of materials
contaminated by the severe accident, and 10 μSv/y for postclosure in disposal), is estimated to be 8900 Bq/kg under thecondition of limited reuse of the site.

The ground surface environment in Fukushima Prefecture
and the surrounding area has been contaminated byradionuclides released from the TEPCO’s Fukushima Daiichi
NPS. Soil particles that adsorbed radioactive cesium flow
into sewage and water resources for tap water. Dehydrated
sludge containing high concentrations of radioactive cesium
was generated by sewage and water supply processing. The
contaminated sludge should be incinerated and disposed ofwhile ensuring the radiation safety of the workers and the
public. However, there was no criterion for treatment of thesludge.
The dose to the workers and the public during incinerationand disposal was estimated to provide the technicalinformation for determining the criterion. We calculated theminimum radioactive cesium concentration from the dose
results, corresponding to the effective dose criterion indicatedby Nuclear Safety Commission of Japan (Near-term policy
to ensure the safety for treating and disposing contaminatedwaste around the site of Fukushima Dai-ichi Nuclear Power
Plants, NSC, June 3, 2011).

We focused on the similarity between the actual situationof general incineration and disposal and a series of scenarios
in previous clearance level estimation. By incorporating
additional exposure pathways peculiar to the sludge into thescenarios based on the clearance level estimation, we could
set up comprehensive and reliable scenarios and pathways
over a short period of time, as shown in Fig.1-22. According
to the information on the actual work conditions and treatment
record of the sludge, suitable parameters in each scenario wereset for a realistic estimation.
The estimated doses for the main exposure pathways areshown in Fig.1-23. Under the condition of a radioactive
cesium concentration of less than 8000 Bq/kg, generalincineration and disposal of the sludge is justifiable under the
restriction that disposal site reuse be limited to residential andagricultural use.
This result was used as technical information for
determining the criterion for sludge treatment by the Ministryof Land, Infrastructure, Transport and Tourism.

Reference
Takeda, S. et al., Dose Estimate for Incineration and Disposal of Dehydrated Sludge Contaminated by Radioactive Cesium of High Activity Concentration,
Hoken Butsuri, vol.47, no.4, 2012, p.247-259 (in Japanese).

22

JAEA R&D Review 2013

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

1-12 Inhibiting Corrosion of Metallic Materials in Spent Fuel Pools－ Verification of Removal Effect of Dissolved Oxygen Using Hydrazine －10

ASW with 32 ppm N2H4 at room temp.

5

Remarkable
DO decreaseby irradiation

5

H2O2 / ppm

DO / ppm

4
3
2
1

Low DO decrease00

Fig.1-24 Analyses of the
dissolved oxygen (DO)concentration in water
The color becomes darker
depending on the DOconcentration.

Before Test
0h

0 kGy/hAfter 1 h

7.5 kGy/hAfter 1 hFig.1-25 Changes in dissolved oxygen concentration
with γ-ray irradiation
DO in artificial seawater (ASW) containing hydrazine(N2H4) was effectively removed by γ-ray irradiation.

At the TEPCO’s Fukushima Daiichi NPS (1F), the cooling
functions of the spent fuel pools (SFPs) failed with the loss ofthe electric power supply, so sea water was injected into the
SFPs of 1F2–1F4 by fire trucks as an emergency measure foravoiding fuel exposure to the air.
Although the corrosion of metallic materials in SFPs isnormally prevented by the circulation of purified water,
accelerated corrosion is a matter of concern in the caseof water mixed with seawater. For example, the SFPs’
water might leak owing to localized corrosion (e.g., pitting
corrosion, crevice corrosion) of the stainless steel used as aliner material.
As a measure for inhibiting corrosion, hydrazine is generally
used as an inhibitor for boiler systems. Therefore, it was also
injected into the SFPs of 1F to remove dissolved oxygen(DO), which accelerates corrosion. It is well known thathydrazine can easily remove DO in hot water (over 200 ºC),
although it was unknown whether hydrazine can remove DOin water mixed with seawater at room temperature and under aradiation environment.
Therefore, we evaluated the removal of DO by adding
hydrazine to artificial seawater (ASW) using a Co-60 γ-ray

10 kGy/h at room temp.
after 1 h

ASW
0 ppm N2H4

ASW
32 ppm N2H4

Fig.1-26 Inhibition of hydrogen
peroxide (H2O2) production
The concentration of H2O2 generated
by water radiolysis was reduced byadding N2H4.

source to simulate the conditions in the SFPs.
The DO concentration in ASW containing 32 ppm of
hydrazine was measured before and after 1 h of irradiation
at a dose rate of 7.5 kGy/h. In this measurement, a coloring
reagent was used, which darkened in color with increasing DO
concentration, as shown in Fig.1-24.
As a result, as shown in Fig.1-25, for the sample that was
not irradiated, the decrease in DO concentration was minimal.
In contrast, the DO concentration in the sample exposed to
γ-ray radiation decreased significantly.
This result showed that hydrazine injected into SFP water
mixed with seawater could effectively remove DO even undera low temperature condition with γ-rays radiated from spent
fuels. Moreover, we found that the addition of hydrazine could
reduce hydrogen peroxide, which was generated by waterradiolysis and is, like DO, a corrosive species (Fig.1-26).
The above results demonstrate that DO and hydrogen
peroxide could be removed by hydrazine addition andirradiation, and it is expected that corrosion of metallic
materials in the SFPs could be inhibited.
On the basis of this research, hydrazine has been
periodically injected into the SFPs of 1F to inhibit corrosion.

Reference
Motooka, T. et.al., Effect of Gamma Ray Irradiation on Deoxygenation by Hydrazine in Artificial Seawater, Nippon Genshiryoku Gakkai Wabun Ronbunshi,
vol.11, no.4, 2012, p.249-254 (in Japanese).
JAEA R&D Review 201323

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

1-13 Characterization of Fuel Debris in Severe Accident－ Properties of Fuel Debris for Developing Defueling Operations －(a) Cutting Tools – Type A(d) Pick-and-Place Tools(b) Cutting Tools – Type B(e) Aspiration System(c) Cutting Tools – Type C( f ) Core Boringex.) Spade Bucket Tool
Working Principle: Pick and Place
Application: To collect small debrisex.) Air Chisel
Working Principle: Impact
Application: To break massive debris

Loose Debrisex.) Heavy-duty Shears
Working Principle: Shearing
Application: To cut tubes and thin plates(Particles, Gravel)ex.) Airlift system
Working Principle: Aspiration
Application: To collect powdery debrisex.) Plasma Arc Cutter
Working Principle: Fusion Cutting
Application: To break up
complex structural materials

Resolidified Debris(Massive crust)Stub-like Fuel(Partially intact fuel)ex.) Core Bore MachineWork Principle: Grinding (Press)
Application: To drill in hard crust debris

Fig.1-28 Categories of defueling tools
Defueling tools in TMI-2 categorized into six groups according
to working principle. These tools were selected according to thefeatures of the debris.

Table 1-2 Properties of fuel debris for defueling
The properties of fuel debris were selected considering theireffect on the tools’ performance.
Thermal Properties
Tools* ShapeFig.1-27 Image of fuel debris from TMI-2 accident
Fuel debris from the TMI-2 accident had various
features depending on the location.

Size

Melting
Density
Point

Mechanical Properties

ThermalHeat
Elastic Fracture
Hardness
Capacity Conductivity
Modulus Toughness

（a）
（b）
（c）
（d）
（e）
（f）
*Refer to Fig.1-28

For decommissioning of the TEPCO’s Fukushima Daiichi
NPS (1F), removal of the fuel debris, which is a mixture ofmolten fuel and structural materials, is planned to start around2020. It is necessary to clarify the characteristics of the fueldebris in order to design and develop the defueling process
and tools.
First, we surveyed the defueling process in the Three Mile
Island Unit 2 accident (hereafter, TMI-2). Several types ofdebris existed in the reactor, such as loose debris (particles,
gravel), resolidified molten debris (massive crust), and stublike fuel (partially melted fuel) (Fig.1-27). Various defuelingtools were selected according to the features of the fuel debris.
In the 1F accident, the relocation and components of thedebris were estimated to differ from those in TMI-2, e.g.,molten core and concrete interaction on the pedestal floor.
However, some types of defueling tools used in TMI-2 can beapplied to 1F defueling because the debris in 1F and TMI-2 isassumed to have similar characteristics.
The useful tools in TMI-2 defueling were categorized into

six groups according to their operating principles, such as
impact and shearing (Fig.1-28). The important propertiesof the fuel debris for defueling were selected considering
their effect on the tools’ performance (Table 1-2). Of theseproperties, the mechanical properties (hardness, elastic
modulus, fracture toughness) must be identified as preferential
items, because few data on these characteristics of fuel debris
are available in past severe accident studies.
We have to use some techniques without large samples tomeasure these mechanical properties because we can obtain
only very small actual debris samples early in the sampling
operation. The hardness is measured by the indentation
method as the Vickers hardness. The elastic modulus can bemeasured by the ultrasonic wave pulse echo method, and thefracture toughness is also measured by the indentation fracture
method.
We will suggest some types of non-radioactive surrogate
debris materials for the development of defueling tools on thebasis of these properties.

Reference
Yano, K. et al., Direction on Characterization of Fuel Debris for Defueling Process in Fukushima Daiichi Nuclear Power Station, Proceedings of International
Nuclear Fuel Cycle Conference (GLOBAL2013), Salt Lake City, Utah, USA, 2013, paper 8167, p.1554-1559., in CD-ROM.

24

JAEA R&D Review 2013

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

1-14 What Has Formed on the Fuel Debris Surface by Sea Water Injection?
－ High-Temperature Reaction Products between Simulated Fuel Debris and Salt Deposits －815 ℃-20h(U,Zr)O2 Pellet periphery

SEI
Solidified NaCl
20 μm

Crystalline MgOSalt residue(MgO+NaCl)

Salt
residue(MgO)(Ca,Na)-U-O
uranate layer

1198 ℃-12h(U,Zr)O2 pellet

50 μmPellet bare surface

Ca

10 μm

U

Zr

20 μm
Fig.1-29 SEM images of (U,Zr)O2
pellet surface heat-treated withsea salt
After heat treatment at 815 °C,
solidified NaCl covers the pellet
surface, and crystalline MgO isdeposited. After heat treatment at
1198 °C, crystalline MgO is themain residue.

Fig.1-30 SEM and X-ray images of crosssection of (U,Zr)O2 pellet heat-treated withsea salt at 1002 °C under airflow
Formation of a (Ca,Na)–U–O uranate layer
50 µm in thickness is confirmed between thepellet surface and salt residue layer.

In the accident at the TEPCO’s Fukushima Daiichi NPS,
sea water was injected into the reactor cores for emergency
cooling. Sea salt may have been deposited on the hightemperature surface of the core melt at the bottom of thereactor vessels. To investigate the possible impacts of sea
water injection on the chemical and physical state of thecorium debris, we performed experiments on the hightemperature reactions between sea salt and (U,Zr)O2 simulated
corium debris (sim-debris).
First, the decomposition behavior of sea salt at hightemperatures was observed. Fig.1-29 shows scanning electron
microscopy (SEM) images of the surface of a (U,Zr)O2 pellet
heat-treated with sea salt under an argon gas flow. The melting
point of NaCl, the main constituent of sea salt, is about
800 °C. After heat treatment at 815 °C, the pellet surface wascovered with a solidified NaCl layer, together with crystalline
MgO decomposed from magnesium salts. Because NaCl is
gradually vaporized at higher temperatures, MgO remains as
the main salt residue after heat treatment at 1198 °C.
The reaction products from sea salt and sim-debris werethen identified. Fig.1-30 shows the cross section of a (U,Zr)O2

Fig.1-31 SEM image of cross sectionof (U,Zr)O2 pellet heat-treated with sea
salt at 1002 °C under argon gas flow
Within the salt residue layer on the pellet
surface, fine pieces of (Ca,U)O2+x are
scattered (in circles). Uranate forms
temporarily, and it is reduced to thedioxide during isothermal heat treatment.

pellet heat-treated with sea salt under an airflow. We havefound that a dense layer of calcium and sodium uranate
formed on the pellet surface at 1002 °C under an airflow, witha thickness of more than 50 µm. On the other hand, under anargon gas flow, the calcium in sea salt is likely to dissolve
into the cubic sim-debris phase to form a solid solution,(Ca,U,Zr)O2+x. The diffusion depth is only 5–6 µm from thesurface after treatment at 1198 °C for 12 h. Another interestingfinding is a scattering of fine pieces of (Ca,U)O2+x in the MgO
residue layer on the pellet surface, as shown in Fig.1-31.
In the heating process, calcium uranate forms temporarily
because of an oxidizing gas (SOx) evolved from the thermal
decomposition of magnesium and calcium sulfates. Duringisothermal treatment under an argon gas flow, the uranate is
reduced to the dioxide solid solution.
From the results of these experiments, we predict that (1)MgO is the main salt residue that may be concomitant withthe corium debris to be retrieved, (2) uranate of calcium andsodium may have formed on the debris surface under highoxygen partial pressures, (3) fine pieces of (Ca,U)O2+x may be
included in the MgO layer under low oxygen partial pressures.

Reference
Takano, M. et al., High Temperature Reaction between Sea Salt Deposit and (U,Zr)O2 Simulated Corium Debris, Journal of Nuclear Materials, vol.443,
issues 1-3, 2013, p.32-39.
JAEA R&D Review 201325

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

1-15 Prevention of Re-Criticality in Fuel Debris in the Damaged Coresｒ

Fuel debris: Composite of damaged
fuel and core components

Fig.1-32 Fuel debris in damaged coreTens of thousands of kilograms of fueldebris, namely, a composite of fuel andcore components, has been generated
as a result of severe core meltdown at
the 1F.

Concretemoderator

Fuel particle

Fig.1-33 Example of calculation
model of fuel debris
This model represents an example
of supposed fuel debris consistingof small fuel particles dispersed inconcrete.

The Great East Japan Earthquake on March 11, 2011 caused
a severe accident at the TEPCO’s Fukushima Daiichi NPS (1F).
As shown in Fig.1-32, the cores of the reactors in 1F1–1F3
were seriously damaged owing to a failure of the core coolingsystem. The amount of fuel debris, which is a composite ofdamaged fuel and other core components, is estimated to betens of thousands of kilograms in each unit.
There has been no indication of re-criticality of the fueldebris since the accident. The risk of re-criticality during
removal of the fuel debris from each core should still beconsidered carefully because the minimum critical mass ofthe fresh fuel is only tens of kilograms. Although criticality
control of normal fuel handling is designed on the basis ofsome determined fuel conditions, there are many unknowns
regarding the fuel debris, such as its composition or shape.
Therefore, an exhaustive criticality database including any
conjectured conditions of the fuel debris, called a “criticality
map,” needs to be prepared to judge the criticality risk quickly
on the disclosure of its actual conditions.
We are analyzing the criticality of fuel debris considering
many parameters, such as water content or fuel burn-up, by

1.6
1.4
1.2

←

The core had melted
and dropped down

Fuel : Concrete = 1 : 7
Mirror reflection(simulating infinite
material)

Can be critical

Damaged Reactor

Infinite multiplication factor－ Criticality Characteristics of Fuel Debris Containing Concrete －1.0
0.8
235U 5wt% fresh fuelSpent fuel without FPs
Spent fuel with FPs

0.6
0.4
0.2

0.1

1

10

Heterogeneity (radius r in Fig.1-33) (cm)Fig.1-34 Infinite multiplication factor of supposed
fuel debris
It is indicated that the fuel debris modeled in Fig.1-33
can be critical under various conditions. Such fueldebris must be controlled carefully to prevent recriticality.

referring to previous studies in the literature.
An example of the calculation model is shown in Fig.1-33.
The model represents fuel dispersed in concrete, which is amajor material of the pedestal floor of the containment vessel.
The infinite multiplication factors of the fuel debris areshown in Fig.1-34. The lines in the red area indicate theconditions under which the fuel debris may be critical.
Moreover, another calculation shows that the minimum criticalmass of the fuel debris can be less than several thousand
kilograms, which is significantly smaller than the amountof fuel in each core. The fuel debris containing concrete,
therefore, must be handled cautiously to prevent re-criticality.
We continue the effort to complete the criticality map bycomputation and critical experiments considering fuel debris
under diverse conditions, such as a submerged fuel–concrete
mixture and a composite of fuel and the steel of the pressure
vessel. The results are expected to provide useful information
for judging the criticality risk for any actual fuel debris, andto contribute to the development of reliable re-criticality
prevention and detection measures.

Reference
Izawa, K. et al., Infinite Multiplication Factor of Low-Enriched UO2-Concrete System, Journal of Nuclear Science and Technology, vol.49, no.11, 2012,
p.1043-1047.

26

JAEA R&D Review 2013

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

1-16 Nuclear Material Accountancy for Molten Core Material byγ-ray Measurement－ Development of Nuclear Material Measurement Technology using Passive γ Spectrometry of Molten Core Material －Debris removal from reactor
4.0E‐03

Debris stored in canister

Segmented gamma scanning
Quantifying low-volatile FP mass154

(with correction of photon-shielding)

Eu/Pu weight ratio

Mass, average density for shielding correction

Burnup estimation

Approx. curve

3.0E‐03
2.5E‐03

Core ave. 1.93E-03

2.0E‐03
1.5E‐033 cycle1 cycle1.0E‐03

0.0E+00

cycle ave.
4.6%Eu0

１0(kgU, kgPu, kgUfis, kgPufis)

Fig.1-35 Data flow of γ spectrometryfor U, Pu quantification
U and Pu are quantified by coupling
passive γ spectrometry for radioactivity
quantification of FP with FP/U, Pu
correlation by inventory calculation, andphoton transport calculation.

Total leakage/source photon

1Completion of fuel quantification

２0

３0

４0

５0

Burnup (GWd/t)

144

Quantifying fuel by low-volatile FP/fuel ratio

4 cycle2 cycle3.0%Eu5.0E‐04

Eu/fuel, Ce/fuelfrom inventory calculations
154

Fig.1-36 Correlation between 154Eu
and Pu quantity (1F2)
154
Eu: Pu ratio increases monotonicallywith respect to burnup.

Eu/Pu(1F2)

154

3.5E‐03

¹⁰⁶Rh
（2.71

MeV）

¹⁴⁴Ce/¹⁴⁴P（
r

2.19 MeV）¹⁵⁴Eu（1.27 M
eV）
¹³⁷C（
¹⁵⁴Eu
（1.59 MeV）
s 662
keV）

0.1

¹⁵⁴Eu
（2

48 keV

0.01

0.001

）

¹⁵⁴Eu
（1

23 keV²⁴¹Am（）

59.5 ke

V）

¹⁵⁵Eu
（4

5.3 keV0

Even after the accident at the TEPCO’s Fukushima Daiichi
NPS (1F), it is required to provide assurance to the national
and international communities that uranium (U) and plutonium(Pu) in the molten fuels have not been diverted to militaryuse. We have been conducting a feasibility study of U and
Pu quantification measurement by passive γ spectrometry ofmolten core material (fuel debris) from outside of a canister
after the removal of fuel debris.
Although U and Pu themselves emit γ-rays, it is very
difficult to measure them directly because their energies aremuch lower than the energies of fission products (FPs). We
focused on the high-energy γ-emitting and low-volatile FPs
that coexist with U and Pu in the fuel melting process during
a severe accident. Cerium (Ce) and europium (Eu) are wellknown as low-volatile FPs, and 144Ce and 154Eu emit highenergy (>1 MeV) γ-rays. Experience in quantifying U and Pu
by measuring γ-rays from 144Ce in fuel debris was gained in
the TMI-2 accident.
Fig.1-35 shows the assumed data flow of passive γ
spectrometry of FPs for U and Pu quantification. Because thisis an indirect methodology for quantifying U and Pu using
FP/U and the Pu ratio by an inventory calculation at the timeof the 1F accident, the sensitivity to various parameters that
affect the quantities of FPs, U, and Pu was analyzed. Fig.1-36,

5

10

）

15

Fig.1-37 Sensitivity of leakage
photon spectrum to debris
sizeHigh-energy γ-rays leak moreeasily than low-energy γ-rays.
Self-shielding correction isessential to obtain the source
photon intensity because ofstrong attenuation in large
pieces of fuel debris.

20

Debris diameter (cm)as an example, shows a good correlation between the ratioof 154Eu to Pu and burnup, regardless of the neutron flux andwater void fraction, because the variance is offset.
In γ-ray measurement of fuel debris, it is necessary toanalyze the sensitivities to parameters such as the selfattenuation in the fuel debris and the geometry of the fueldebris and canister. A fundamental analysis was performed
using a simple spherical model with a variety of parameters,such as the debris size, composition, porosity, density,
homogeneity, or heterogeneity. It is assumed that low-energy
γ-rays (<500 keV) are sensitive to the size and composition of
the debris, and high-energy γ-rays (>1 MeV) are less sensitive.
Fig.1-37 shows the sensitivity of the energy-dependent leakage
photon ratio to the debris size. For 154Eu, strong attenuation
was observed, up to about 20% and 50% with fuel debris
2.5 cm and 10 cm in diameter, respectively. Correction of selfshielding is essential. The use of the correlation between the
difference in the photopeak flux and the energy of 154Eu has
been analyzed as an option.
We will continue the feasibility study of this methodology
by numerical analysis with the canister model and conceptual
system design, and improve the precision of nuclear material
quantification.

Reference
Sagara, H., Tomikawa, H. et al., Feasibility Study of Passive Gamma Spectrometry of Molten Core Material from Fukushima Dai-Ichi Nuclear Power
Station Unit 1, 2 and 3 Cores for Special Nuclear Material Accountancy －Low-Volatile FP & Special Nuclear Material Inventory Analysis and Fundamental
Characteristics of Gamma-ray from Fuel Debris －, Journal of Nuclear Science and Technology, vol.51, issue 1, doi: 10.1080/00223131.2014.852994.
JAEA R&D Review 201327

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

1-17 Calculation of Decay Heat Causing Core Meltdown－ Three-Dimensional Decay Heat Analysis for the TEPCO’s Fukushima Daiichi NPS －(a) Calculation method
Use of the MOSRA code systemdeveloped by JAEA and the Japaneselatest nuclear data library JENDL-4.0(b) Decay heat distribution
at the time of core shutdown（水平断面図）
1
2
3
4

13
12

9

7

7

8

6

9 10 11 12 13 14
5 4 3
2 1

6

9

8

7

7

8

6

9 10 11 12 13 14
5

4

3

2

1

0.55 0.66 0.60 0.72 0.60 0.60 0.72 0.80 0.60 0.60 0.80 0.72 0.60
0.68 0.82 0.75 0.90 0.75 0.75 0.90 1.00 0.75 0.75 1.00 0.90 0.75

220.77 0.92 0.83 1.01 0.83 0.83 1.01 1.12 0.83 0.83 1.12 1.01 0.83

1.2
1.0
0.8

0.81 0.89 1.19 0.98 1.07 1.19 0.98 1.07 1.19 0.98 1.07 0.98 1.19

A 1

5

23

0.81 0.89 1.07 1.19 0.98 1.07 1.19 0.98 1.07 1.19 0.98 1.19 1.07

2

4

11 10

24

0.81 0.98 0.98 1.07 1.19 0.89 1.07 1.19 0.89 0.89 1.19 0.98 0.89

33

0.81 0.81 0.98 0.89 0.89 0.98

0.81 0.89 0.98 1.07 1.19 0.98 1.07 0.89 0.89 1.07 1.19 0.89

42

0.36 0.43 0.39 0.47 0.39 0.39 0.47 0.52 0.39 0.39 0.52 0.47 0.39

0.81 0.81 0.81 0.89 0.89 0.98 1.07 1.19 0.89

0.81 0.89 1.19 0.98 1.07 1.19 0.89 1.07 1.19 0.98 1.07 0.98

5

1

14 13 1225

0.81 1.07 0.98 1.07 1.19 0.98 1.07 1.19 0.98 1.07

0.81 1.07 1.19 0.98 0.98 1.19 0.98 1.07 1.19 0.98 1.19

6

at the time of core shutdown
Z＼X0.81 0.98 0.89 0.89 0.98 1.07 1.19 0.89 1.07 1.19 0.89

725 axial regions

8

0.81 0.89 0.89 1.19 0.98 1.07 1.19 0.98 1.07 0.89

8

Fuel assembly

Burnup distribution

6

0.81 0.81 0.81 0.81

10Void distribution

9

14

11

The Following are calculated by region,(1) Neutron spectrum (200-group)(2) Reaction rate (e.g. number of fissions)(3) Nuclide-wise concentration (>1000 nuclides)(4) Decay heat (Sum of the nuclide-wise decay heat)5

Y＼X 14 13 12 11 100.81 0.98 0.89 1.07 0.89 0.89 1.07 1.19 0.89 0.89 1.19 1.07 0.89

after about 80 h

1
2
3
4
5
6
7
8
9 10 11 12 13 14
Y＼X 14 13 12 11 10 9 8 7 6 5 4 3 2 1

14

21

0.81 0.97 0.88 1.07 0.88 0.88 1.07 1.19 0.88 0.88 1.19 1.07 0.88

20

0.83 1.00 0.91 1.10 0.91 0.91 1.10 1.22 0.91 0.91 1.22 1.10 0.91

19

0.86 1.03 0.93 1.13 0.93 0.93 1.13 1.25 0.93 0.93 1.25 1.13 0.93

18

0.88 1.05 0.95 1.15 0.95 0.95 1.15 1.28 0.95 0.95 1.28 1.15 0.95

170.89 1.07 0.97 1.17 0.97 0.97 1.17 1.30 0.97 0.97 1.30 1.17 0.97

16

0.90 1.08 0.98 1.18 0.98 0.98 1.18 1.31 0.98 0.98 1.31 1.18 0.98

15

0.91 1.09 0.99 1.20 0.99 0.99 1.20 1.33 0.99 0.99 1.33 1.20 0.99

14

0.92 1.10 1.00 1.21 1.00 1.00 1.21 1.34 1.00 1.00 1.34 1.21 1.00

13

0.92 1.10 1.00 1.21 1.00 1.00 1.21 1.34 1.00 1.00 1.34 1.21 1.00

12

0.92 1.11 1.00 1.21 1.00 1.00 1.21 1.35 1.00 1.00 1.35 1.21 1.00

11

0.92 1.11 1.00 1.22 1.00 1.00 1.22 1.35 1.00 1.00 1.35 1.22 1.00

10

0.93 1.12 1.01 1.22 1.01 1.01 1.22 1.36 1.01 1.01 1.36 1.22 1.01

0.94 0.94 0.94 0.94

9

0.94 1.12 1.02 1.23 1.02 1.02 1.23 1.37 1.02 1.02 1.37 1.23 1.02

12

0.94 0.94 1.03 0.99 0.99 1.03

8

0.93 1.12 1.01 1.23 1.01 1.01 1.23 1.36 1.01 1.01 1.36 1.23 1.01

11

0.94 0.94 0.94 0.99 0.99 1.03 1.05 0.98 0.99

107

0.93 1.11 1.01 1.22 1.01 1.01 1.22 1.35 1.01 1.01 1.35 1.22 1.01

0.94 1.05 1.03 1.05 0.98 1.03 1.05 0.98 1.03 1.05

6

0.91 1.10 0.99 1.20 0.99 0.99 1.20 1.33 0.99 0.99 1.33 1.20 0.99

50.89 1.07 0.97 1.18 0.97 0.97 1.18 1.30 0.97 0.97 1.30 1.18 0.97

40.84 1.01 0.92 1.11 0.92 0.92 1.11 1.23 0.92 0.92 1.23 1.11 0.92

30.76 0.91 0.83 1.00 0.83 0.83 1.00 1.11 0.83 0.83 1.11 1.00 0.83

20.63 0.76 0.69 0.84 0.69 0.69 0.84 0.92 0.69 0.69 0.92 0.84 0.69

1

0.48 0.57 0.52 0.63 0.52 0.52 0.63 0.69 0.52 0.52 0.69 0.63 0.52

13

9

0.94 0.99 0.99 0.98 1.03 1.05 0.98 1.03 1.05 0.99

8

0.94 1.03 0.99 0.99 1.03 1.05 0.98 0.99 1.05 0.98 0.99

7

0.94 1.05 0.98 1.03 1.03 0.98 1.03 1.05 0.98 1.03 0.98

6

0.94 0.99 0.98 1.03 1.05 0.98 0.99 1.05 0.98 1.03 1.05 1.03

5

1.2
1.0

0.94 0.99 1.03 1.05 0.98 1.03 1.05 0.99 0.99 1.05 0.98 0.99

4

0.94 1.03 1.03 1.05 0.98 0.99 1.05 0.98 0.99 0.99 0.98 1.03 0.99

3

0.94 0.99 1.05 0.98 1.03 1.05 0.98 1.03 1.05 0.98 1.03 0.98 1.05

20.94 0.99 0.98 1.03 1.05 0.98 1.03 1.05 0.98 1.03 1.05 1.03 0.98

1

0.94 1.03 0.99 1.05 0.99 0.99 1.05 0.98 0.99 0.99 0.98 1.05 0.99

Horizontal distribution

0.8

1.37
0.86
0.36

1.00

Axial distribution(A-A’ cross-sectional surface)(normalized to core average)

Fig.1-38 Calculation method and sample result of decay heat distribution in 1F2
The 3D decay heat distribution was calculated considering the distributions of the neutron spectra, burnup, and void fraction.
The fuel loading pattern was based on that of a typical boiling water reactor.

Even after a nuclear reactor is shut down, nuclides such as
fission products (FPs) decay, and decay heat is generated. The
accident at the TEPCO’s Fukushima Daiichi NPS (1F) wascaused by a failure of the ultimate decay heat removal becauseof the loss of electric power due to the tsunami.
Understanding the core melt progression in accidents
enables the location of melted fuel to be estimated andsupports further improvements in the safety of nuclear
reactors. The detailed analysis of the core melt progression
begins with a prediction of the decay heat distribution in thecore at the time of shutdown. Although there are some simple
formulae for estimating the decay heat, they do not yield an
accurate time-dependent decay heat distribution in the core.
Thus, we newly developed data and a code to predict the
time-dependent three-dimensional decay heat distribution.
They were applied to the decay heat analysis of each core of
1F1, 1F2, and 1F3. In this method, each core is divided into
four or five horizontal regions corresponding to different
fuel assembly types or refueling batches, and each region isfurthermore divided into 25 axial regions that have different

burnup values and void fractions, as shown in Fig.1-38 (a).
For each region, the nuclide-wise concentration is calculated
for more than 1000 nuclides considering the operating
conditions of each plant until core shutdown. Then, we canobtain the nuclide-wise decay heat in each region using the
decay and heat generation constants. The region-wise decay
heat is calculated by summing the nuclide-wise decay heat.
Fig.1-38 (b) shows the obtained decay heat distribution of
1F2, the core melting time of which is estimated to be about
80 h after core shutdown. The decay heat at the time of core
shutdown is approximately proportional to the power density
under normal operating conditions because the decay heat justafter shutdown is dominated by the concentrations of shortlived FPs, which are in equilibrium according to the power
density of each region. After that, the decay heat distribution
gradually becomes flatter owing to the disappearance of theshort-lived FPs. The obtained three-dimensional decay heat
information is useful for future detailed analyses of the coremelting progression because the decay heat is an importantparameter for the movement of melted materials.

Reference
Okumura, K. et al., Nuclear Data for Severe Accident Analysis and Decommissioning of Nuclear Power Plant, Proceeding of the 2012 Symposium on
Nuclear Data, November 15-16, 2012, Research Reactor Institute, Kyoto University, Kumatori, Japan, JAEA-Conf 2013-002, 2013, p.15-20.

28

JAEA R&D Review 2013

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

1-18 What’s the Point of Analyzing the Fukushima Accident?
－ Review of Five Investigation Reports on Fukushima Accident －from Reactor Building

FO

1F3: ca. 2.0 ca. 3.5 ca. 1.3
1F4: ca. 0.1 ca. 0.5 ca. 6.7
FO

FO: fail open
FAI: fail as is

to Venting Line

to Stack(train A)
FO

FAI

FO

FO

① FAI
from Suppression Pool

Gravity Damper(NOT equipped at 1F4)SGTS Filters② to Stack

FO
FO

FAI
ca.18
(1F3)

ca. 0.8 (1F4) ca. 0.7
(1F3)
FAI
ca. 1.5

ca.1.0

1F3: ca. 1.6 ca. 3.2 ca. 3.1
1F4: ca. 0.1 ca. 0.5 ca. 5.5

FAI

FO(train B)
from Venting Line

Fig.1-39 Dosimetries at standby gas treatment system (SGTS) filters for 1F3 and 1F4 (unit: mSv/h)
Based on the dosimetries on the SGTS filters at 1F4, it is presumed that venting gases flowed from the 1F3 reactor building throughthe SGTS line(s) into the 1F4 reactor building. On the other hand, venting gas flow through the SGTS lines is assumed to be unlikely
because different trends were observed in the dosimetries on the SGTS filters at 1F3, and the dosimetry downstream of the valve
connected with the stack ( ② in the figure) was significantly low. However, the possibility of venting gas flow should be further examinedbecause the dosimetry of the filter in the center was the highest, and the containment pressure did not exceed the design pressure fora long time.

On March 11, 2011, an earthquake and subsequent tsunami
resulted in a severe accident at the TEPCO’s Fukushima
Daiichi NPS (1F). Four independent committees were
established by the Japanese government, the Diet of Japan,
the Rebuild Japan Initiative Foundation, and the TEPCO toinvestigate the accident and published their respective reports.
In addition, the Nuclear and Industrial Safety Agency analyzed
the causes of the accident to obtain the lessons learned from it.
This study reviews the five reports, clarifies the differencesin their positions, from a technological point of view, andidentifies undiscussed issues to obtain insights useful for nearterm regulatory activities, including the accident investigationby the Nuclear Regulation Authority. As a result, the different
positions identified are related to the reason of emergency
diesel generator trips, the reason of manual trip of theisolation condensers at 1F1, reactor depressurization at 1F1,
containment integrity and venting operation, and the hydrogenleak path. For example, the hydrogen explosions at 1F1 and
1F3 are assumed to be the result of hydrogen leaking directlyfrom the containments, and the explosion at 1F4 is assumed tobe caused by reverse hydrogen flow from 1F3 during venting.
This study pointed out that the contribution of venting gases
should be further examined based on the fact that containment

venting was executed shortly before the explosions at 1F1
and 1F3 and on the measured dosimetries on the standby gas
treatment system (SGTS) filters (Fig.1-39).
In addition, the following undiscussed issues are identified.
(1) Adequacy of Operating Procedures Applied at 1F1: Indepth analysis of the basic concepts for preparing theoperating procedures by comparison with those at Tsuruga
1F1 is required.
(2) Design Concepts of SGTS Valves on Loss of Drive
Forces: The design concepts of “fail open” and “fail asis” when the drive force for the valves is lost should beclarified by examining the actual designs at other plants.
(3) Design Concepts of Valves Configured in Containment
Venting Lines: The design concepts for the venting lines,
which consist of one rupture disk and two normallyclosed valves placed in series, should be studied.
(4) Operating Basis of Startup of Containment Cooling
Systems: Although suppression pool cooling was initiatedafter the earthquake at 1F1 and 1F2, it was not started
up at 1F3 considering the effects of tsunami-induced
kataseism. It is necessary to identify the causes of thisdifference.

Reference
Watanabe, N., Tamaki, H. et al., Review of Five Investigation Committees’ Reports on the Fukushima Dai-Ichi Nuclear Power Plant Severe Accident, Nippon
Genshiryoku Gakkai Wabun Ronbunshi, vol.12, no.2, 2013, p.113-127 (in Japanese).
JAEA R&D Review 201329

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

1-19 Mechanism of Cesium Absorption on Zeolite－ First-Principles Study of Structural Effects on Absorption in Zeolites －

Cs

Cs

Z

X

Si
O

Cs

Na

Al: 0

Al: 2

Al: 1

Al: 3

0 meV

X

Y

Na

100 meV

Si
Al
O

Al: 4

Fig.1-40 Crystal structure of zeolite called mordenite
Mordenite has a framework made of silicon (Si) and oxygen (O)atoms. If some of the Si atoms are replaced with aluminum (Al)atoms, mordenite can absorb cations such as cesium ions (Cs+).
The dense part of the framework represents the Cs absorption
site, which appears in Fig.1-41.

Fig.1-41 Absorption energy distribution at Cs+ absorptionsite
At the Cs absorption site described in Fig.1-40, the absorptionenergies of Cs and sodium (Na) ions are plotted for various
numbers of Al atoms (indicated after “Al:” in the figure). Dark
area corresponds to lower-energy region.

After the accident at the TEPCO’s Fukushima Daiichi
NPS, the treatment of nuclear waste water, which includes
radionuclides such as radioactive cesium, became an important
problem. Cesium is presently removed from waste water by
absorbent materials, for example, zeolite. Although more than
100 types of zeolite have been discovered so far, some ofthem absorb cesium well, and some absorb very little. If wedevelop a new zeolite that absorbs more cesium and has better
controllability of absorption and removal, treatment of waste
water will advance.
To develop desirable zeolites, we have to know how
zeolites absorb cesium. However, it is difficult to directly
observe the absorption behavior of cesium because it occurs
at the microscopic scale. In this case, numerical simulations
at the atomic scale are effective for revealing the absorption
mechanism.
The characteristics of zeolites that absorb Cs selectively are
known empirically. However, the principles of the selectivity
mechanism have been unclear. To clarify these principles, we

calculate in detail the absorption energy at the Cs absorption
site in mordenite (Fig.1-40), which is a zeolite with a highCs-selective absorbability. We adopted the first-principles
calculation method to evaluate the absorption energy because
it is the most accurate and reliable. The results are shown in
Fig.1-41. This figure shows density plots of the absorptionenergy of Cs and sodium (Na). For Na, the minimum points
are distributed, and therefore a Na ion is weakly bound at thesite. On the other hand, a Cs ion is strongly bound at only
one minimum-energy point in the center of the site. Thus, we
determined why Cs is bound more strongly than Na at thissite. By numerical simulations, we also identified various
characteristics of Cs-selective absorption, some of which werepreviously unknown.
With this deeper understanding of the absorption
mechanism, we can appropriately plan the development ofhigh-performing absorbent materials. This development
can contribute to not only treatment of waste water but also
decontamination of environmental radionuclides.

Reference
Nakamura, H. et al., First-Principles Calculation Study of Mechanism of Cation Adsorption Selectivity of Zeolites: A Guideline for Effective Removal ofRadioactive Cesium, Journal of the Physical Society of Japan, vol.82, no.2, 2013, p.023801-1-023801-4.

30

JAEA R&D Review 2013

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

1-20 Decontamination of Radioactive Cobalt and Manganese with Radioactive Cesium－ Simplified Treatment Process of Radioactive Liquid Waste by Adding Hexacyanoferrate(ⅠⅠ) Ions －Contaminated water (left) and potassium hexacyanoferrate(ⅠⅠ)
solution (right)

Cloudy solution after adding hexacyanoferrate(ⅠⅠ)
solution to the contaminated water

Solution after reaction

Fig.1-42 Simplified treatment process experiment
Hexacyanoferrate(ⅠⅠ) ions react with transition metal ions, and this reaction produces precipitation of the transition metal
hexacyanoferrate(ⅠⅠ). This precipitation product may selectively adsorb Cs. Therefore, decontamination of both radioactive transition
metals and Cs is expected to be achieved by adding potassium hexacyanoferrate(ⅠⅠ) solution to the contaminated water from 1F.
100

Cs removal rate (%)

Removal rate (%)

100

75

50

Mn
Co

25

Ni
Zｎ

80
60
40
20
0

0
0.0

0.5

1.0

1.5

[K4[Fe(CN)6]] / [transition metals] (in molality)
Fig.1-43 Effect of ratio of hexacyanoferrate/transition metals on
removal rate of transition metals
The removal rates of the transition metals exhibited some differences.
It is possible to remove every radioactive transition metal by adding
a sufficient amount of hexacyanoferrate(ⅠⅠ) ions.

At the TEPCO’s Fukushima Daiichi NPS (1F), an enormousvolume of contaminated water has been generated by reactor
cooling and groundwater inflow, and is temporarily stored
at the site. The contaminated water bears not only fission
products from the spent fuels, such as cesium (Cs), but also
activated corrosion products from the reactor materials, suchas manganese (Mn) and cobalt (Co).
Adsorption with insoluble hexacyanoferrate(ⅠⅠ) is a wellknown method of removing radioactive Cs from liquid waste. For
radioactive Mn and Co decontamination, co-precipitation withiron hydroxide is generally used. In fact, part of the contaminatedwater was actually treated by applying these methods at 1F.
It is known that the insoluble transition metal hexacyanoferrate(ⅠⅠ)
selectively adsorbs Cs among alkaline elements. Hexacyanoferrate(ⅠⅠ)
of Ni, Co, and Fe are typical compounds used to adsorb Cs, andthey are products of the reaction of hexacyanoferrate(ⅠⅠ) ions
and transition metals. In iron co-precipitation, transition metals
are co-precipitated with iron hydroxide, which is a product of
the reaction with iron ions and hydroxide ions, because of theirchemical similarity.
We developed a simplified treatment process for both
radioactive transition metals and Cs decontamination.
Fundamental experiments using both non-radioactive andradioactive solutions were conducted. As shown Fig.1-42,

0

5

10

15

20

[transition metal] / [Cs] (in molality)
Fig.1-44 Increase of Cs removal rate
The Cs removal rate increased with[transition metal] / [Cs].

the treatment process consists of two reactions. The first reaction
is in-situ generation of the transition metal hexacyanoferrate(ⅠⅠ)
by adding potassium hexacyanoferrate(ⅠⅠ) solution to thecontaminated water. The second reaction is Cs adsorption on
the first reaction product.
The properties of the in-situ generation of insoluble
hexacyanoferrate(ⅠⅠ) adsorbent were examined using nonradioactive transition metal elements. As shown in Fig.1-43,
the removal rate rose as the ratio [K4[Fe(CN)6]]/[transition
metal] increased. For [K4[Fe(CN)6]]/[transition metal] > 1, theremoval rates of transition metals were greater than 90%.
The Cs adsorption was also examined to investigate theremoval rate of Cs. As shown in Fig.1-44, the removal rate of
Cs was 95% when [transition metal]/[Cs] was 12. The removal
rate of Cs reached 99% when [transition metal]/[Cs] was 18.
These results show that this new treatment process has thepotential to remove both radioactive transition metals andCs effectively when the ratio of Cs and transition metals isappropriate for the reactions.
At 1F, contaminated water has been treated by other treatment
processes. We hope that our proposed process will be applied forthe removal of Cs and transition metals from radioactive liquid
waste in various decontamination processes in and around 1F.

Reference
Takahatake, Y., Shibata, A. et al., Decontamination of Radioactive Liquid Waste with Hexacyanoferrate(II), Procedia Chemistry, vol.7, 2012, p.610-615.
JAEA R&D Review 201331

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

1-21 Comprehension of Radioactive Contamination inside the Reactor Buildings－ Radioanalysis of Concrete Core Boring Samples －Measurement surfaces for autoradiography(grinded surfaces of the coating layer)

Coating layer(epoxy resin)

Grinding
direction

Autoradiography on the grinded surfaces of
the coating layers was carried out througheach of 0.5 mm grinding.

Measurement results of autoradiography

Floors or walls ofreactor buildings

Before grinding

0.5 mm grinded

1 mm grinded
Low

Fig.1-45 Photograph of concrete core boring sample
Radioanalysis of concrete core boring samples 10 cm indiameter was conducted. The samples were taken fromthe floors and walls of the 1F reactor buildings.

Dose rate

High

Fig.1-46  Evaluation of penetration depth of radioactive contaminant
The penetration depth of the radioactive contaminant was evaluated byautoradiograph measurements of ground surfaces of the coating layer.
Heterogeneous penetration of the radioactive contaminant to a 1 mm
depth below the surface was observed for the sample taken from 1F2.

Decontamination technology that uses remote accessdevices has to be developed to reduce the radiation dose ofworkers in the reactor buildings toward the decommissioningof the TEPCO’s Fukushima Daiichi NPS (1F). For effective
decontamination, it is necessary to comprehend the statusof radioactive contamination inside the reactor buildings,
specifically, the component nuclides, penetration into
the floors or walls, and distribution of the radioactive
contaminant. In particular, detailed radioanalysis is needed toevaluate the penetration of the radioactive contaminant into
the floors or walls. We conducted a series of radioanalysis
tests of concrete core boring samples taken from the floors
and walls of the 1F reactor buildings (Fig.1-45) at the postirradiation examination facilities in the Oarai Research andDevelopment Center. Various radioanalytical techniquessuch as α- and γ-spectrometry and autoradiography, whichhave been developed in the facilities so far, were used for theradioanalysis of core boring samples; the penetration of theradioactive contaminant was thus evaluated.
The results of the radioanalysis tests showed 134Cs and137
Cs as the dominant contaminant nuclides, with a ratioof radioactivity of 134Cs:137Cs = 2:3 in August 2012. It was
revealed that the radioactive contaminant has penetrated
heterogeneously to a 1 mm depth from the surface of thecoating layer of the floor (2.5 mm thick) at the maximum(Fig.1-46). Of note is the difference in the penetration depth
in the coating layer between 1F1/1F3 and 1F2: 0.5 and 1 mm,
respectively. Removal tests of the radioactive contaminantnear the surfaces of the coating layer were then conductedby several cleaning methods using pure water or citric acidand an ablation gel for further investigation of the differencein the penetration depth of the radioactive contaminant. The
removal tests yielded almost complete removal of the residual
radioactive contaminant that penetrated to a 1 mm depth
from the surface of the coating layer. This result indicatesthat the radioactive contaminant can be characterized notby penetration into the coating layer itself by, for example,
a chemical reaction with the coating layer material, but byadherence of the contaminant in microcracks or breaches on
the surface of the coating layer, considering the heterogeneous
penetration of the radioactive contaminant into the coating
layer. This result suggests the potential applicability of theabove-mentioned removal method without actual removal of
the coating layer.
These radioanalysis results are reflected in the remotedecontamination plan for the 1F reactor buildings in terms of
the selection of suitable removal methods for the radioactive
contaminant and optimization of the demonstration test
conditions for the decontamination apparatus.

Reference
Maeda, K., Osaka, M. et al., Results of Detailed Analyses Performed on Boring Cores Extracted from the Concrete Floors of the Fukushima Daiichi Nuclear
Power Plant Reactor Buildings, Proceedings of International Nuclear Fuel Cycle Conference (GLOBAL2013), Salt Lake City, Utah, USA, 2013, paper 7403,
p.272-277., in CD-ROM.

32

JAEA R&D Review 2013

2

Research and Development of Advanced Nuclear System

R&D of Fast Reactor Cycle Technology
R&D of FR Cycle Technology

R&D in “MONJU”◆ International Standardization of Safety
Design Requirements◆ Fundamental R&D
Experimental
Fast Reactor“JOYO”

ChemicalSodium-cooled
Fast Reactor Processing Facility
Concept

Development of SDC

Prototype Fast Breeder Reactor “MONJU”

Experience with theaccident at 1F*

International
Cooperation

Reflection of the SA measures thattake into account lessons from theaccident at 1F* to SDC

Plutonium Fuel
Production Facility◆ Demonstrating the Reliability of
an Operational Power Plant◆ Establishing Sodium HandlingTechnology

Planning of International
Joint Experiment

Enhancement of SA
measures in “MONJU”

International Workshop on
International Collaborative
Research using “MONJU”*1F: The Tokyo Electric Power Company, Incorporated Fukushima Daiichi Nuclear Power Station

Fig.2-1 Overview of research and development of fast reactor cycle technology
We are implementing activities related to international standardization of safety design requirements and conservation of the technical basis
as efforts for the near term until national policies are reconsidered. As part of these efforts, we are promoting research and development
utilizing international cooperation, such as development of the SDC and planning of an international joint experiment on safety issues.

Until FY 2011, research and development activities wereconducted toward the commercialization of fast reactor cycle
technology with the cooperation of the electric utilities; the main
target was the concept of a combination of a sodium-cooled
fast reactor, advanced aqueous reprocessing, and simplified
pelletizing fuel fabrication. Currently, national policies arebeing reconsidered because of the impact of the accident at
the Tokyo Electric Power Company, Incorporated Fukushima
Daiichi Nuclear Power Station (1F). In accordance with thesituation, we have implemented activities focused on the efforts
required for conservation of the technical basis and international
standardization of safety design requirements for the near term
utilizing international cooperation, while reflecting the experience
gained in the accident at 1F (Fig.2-1). In May 2013, the Safety
Design Criteria (SDC) that we proposed were approved at theGeneration-IV International Forum (GIF). This chapter describes
the resulting enhancement of the safety and reliability considering
the features of a sodium-cooled fast reactor and basic research on
fuel cycle technology.
The Japan Sodium-cooled Fast Reactor (JSFR) is designed
to have high resistance against earthquakes and tsunamis byadopting a seismic isolation system for the reactor building anda natural convection capability for cooling after an abnormal
shutdown (Topic 2-1).
Regarding an ultrasonic flowmeter for the cooling system, ameasure that enables precise measuring results to be obtainedeven for flows exhibiting great spatial and temporal variation
was identified. The result shows that the ultrasonic flowmeterhas the potential for adoption in the plant system (Topic 2-2).
To develop an under-sodium viewer aimed at visual andvolumetric inspection of the internal structures in a reactor vessel,
data were obtained to justify the present improvement measurewith regard to wetting of the ultrasonic sensor surface by liquidsodium, which affects the sensor performance (Topic 2-3).
The strength properties of 316FR and Modified 9Cr-1Mo
steel, which are adopted as structural materials for JSFR, were
formulated on the basis of an evaluation of the material test
data. This makes the structural design more reliable (Topic 2-4).
Regarding sodium/water reaction, which is an important
issue for ensuring the safety and protection of the steam
generator, experiments were conducted as a joint study with theCommissariat à l’Énergie Atomique et aux Énergies Alternatives(CEA), France. A reaction jet that was produced during heattransfer tube failure and the effect on the resistance characteristics
of a heat transfer tube were evaluated (Topic 2-5).
With respect to basic research on the fuel cycle, the oxygento-metal ratio, which significantly affects various physicalproperties of uranium–plutonium mixed oxide fuel, such as
the thermal conductivity, could be controlled at will. These
technologies became applicable to fuel fabrication andirradiation behavior evaluation (Topic 2-6).

JAEA R&D Review 201333

Research and Development of Advanced Nuclear System

2-1

Robustness of JSFR against Earthquakes and Tsunamis－ Effectiveness of Natural Convection in a Decay Heat Removal System －Reactor Building

Air

Heat Release

Heat Removal(Air cooler)
Rubber Bearing

Oil Damper

Fig.2-2 Advanced seismic isolation system
Thicker laminated rubber bearings and oil dampers
mitigate the seismic force to the reactor building.

Natural convection of
Na induced byvertical interval
between core and aircoolerHeat Source(Core)

Ordinary Air Path
In case of stack collapse (dashed line),decay heat removal through naturalconvection was confirmed by analysis
Outlet Air Temp.
ca.450 ℃

Air Pathin case of
Collapse

Fig.2-3 Decay heat removal with natural convection in SFR
Natural convection system achieves long-period cooling without
power supply.

lnlet Na Temp.
ca.572 ℃

Outlet Na Temp.
ca.442 ℃
lnlet Air
Temp.
ca.40 ℃

Fig.2-4 Air cooler analysis in case of stack collapse
If air path at their root is alive, air cooler can work even if all
stacks collapse.

The integrity of major components of the Japan Sodiumcooled Fast Reactor (JSFR) against the TEPCO’s Fukushima
Daiichi NPS seismic conditions is confirmed thanks to theadoption of an advanced seismic isolation system that mitigates
the horizontal seismic force to a reactor building by using
thicker laminated rubber bearings with a longer period and oil
dampers (Fig.2-2), which improved the damping performance.
Furthermore, against subsequent tsunamis, JSFR is aiming to
improve the safety characteristics by adopting a decay heatremoval system with natural convection.
Because sodium is liquid under a wide temperature range, it
exhibits remarkable heat transfer performance, which leads to alarge temperature difference between the sodium inlet and outletof the reactor. Therefore, decay heat can be removed by naturalconvection of the coolant sodium due to the density difference
that is induced by the vertical interval between a core and an
air cooler; in fact, the final heat sink is air (Fig.2-3). Thanks to
the natural convection decay heat removal system (DHRS), the
JSFR requires no active components such as pumps or blowers,
which depend heavily on a power supply. Thus, the inlet/outlet
dampers at the air cooler alone can operate the JSFR DHRS
using only an emergency DC battery in the reactor building.
Furthermore, the natural convection DHRS has an advantage
in terms of the robustness of an emergency AC power supply
against tsunamis. In the event of an emergency shutdown
in a sodium-cooled fast reactor, structural materials must be
protected against thermal shock due to a rapid temperaturetransition in the sodium. For this reason, some sodium flow
must be maintained after reactor shutdown. The conventional
forced circulation DHRS has pony motors at the main pumps.
The pony motors have to be activated within several seconds
using diesel generators (DGs) to control the sodium flow if
the off-site power supply is lost. Because the DGs require an
auxiliary system consisting of a lubricating oil system and a sea
water cooling system, the sea water system, including the water
intake components, must be protected against tsunamis. On theother hand, because the JSFR does not require this type of quick
DG initiation thanks to the natural convection DHRS, a gas
turbine generator (GT) with self-air cooling can be used as the
emergency power supply, even though it takes more than 30 s
to activate. Hence, the JSFR emergency AC power supply canwithstand even a sea water cooling system failure.
In addition, decay heat removal can be maintained even in
the event of stack collapse due to an airplane attack as long as
the vertical interval between the heat exchanger tubes of the aircooler and an air path at the bottom of the stack above the aircooler can be maintained (Fig.2-4).
In the future, the JSFR design concept will be improved tosatisfy the Generation IV SFR safety design criteria based onthese studies.
The present study was conducted as part of the “Technicaldevelopment program on a commercialized FBR plant”
commissioned by the Ministry of Economy, Trade and Industry(METI).

Reference
Hayafune, H., Kato, A. et al., Evaluation of Severe External Events on JSFR, Proceedings of International Conference on Fast Reactors and Related Fuel
Cycles; Safe Technologies and Sustainable Scenarios (FR13), Paris, France, 2013, paper INV-059, 11p., in USB Flash Drive.

34

JAEA R&D Review 2013

Research and Development of Advanced Nuclear System

2-2

Measurement of Temporospatially Fluctuating Flow
－ Development of Ultrasonic Flowmeter for Fast Reactor －Table 2-1 Comparison of main specifications of prototype
and commercial fast reactors in Japan

Piping B

In commercial reactor design, high-chrome steel with low thermal
expansion is used for piping, and the pipe drawing is shortened. Further,
the flow velocity is increased because the loop number is reduced.

Heat Exchanger

Specification
Electricity output
Piping material
Number of loops
Flow velocity

Piping A(Installation
Position)

Reference Flowmeter
0.175

Flow Velocity

Example Installed
1 Flowmeter Pair

Average flow rate (m3/s)

SensorAs one of the fundamental engineering tasks towards
commercialization of a fast reactor cycle technology, aflow rate measurement technique for the sodium coolant isdeveloped. The development requirements and investigation
results are presented below.
As shown in Table 2-1, because the piping material is anonmagnetic stainless steel, an electromagnetic flowmeter has
been used as the flowmeter in fast reactors that use a liquid
metal, such as the prototype reactor “MONJU”. However,
because the piping material in a commercial reactor is a highchrome magnetic steel, which can shorten the required piping
length, the electromagnetic flowmeter cannot be applied. In
addition, the flow velocity is increased to reduce the loop
number and simplify the heat transport systems. Because of theshort straight pipes and high velocity, the flowmeter needs tomeasure a flow that exhibits greater temporospatial fluctuation
than before. Because an examination of the flowmeter underfluctuating flow conditions and with high-chrome steel pipes
revealed that an ultrasonic flowmeter was most suitable, its
applicability was investigated experimentally.
A schematic of the test loop is shown in Fig.2-5. Because
the test loop can circulate water at a flow velocity comparable
to that in a commercial reactor, the flow condition of a

：1 pair：2 pairs

Commercial
1500 MWe
High-chrome steel
29 m/s：4 pairs：Reference

0.170
0.165
0.160
0.155
0.150
0.48

Fig.2-5 Test loop of ultrasonic flowmeter
Several pairs of ultrasonic flowmeters were set in a water
test loop to examine their applicability to a temporospatiallyfluctuating flow. The parameters are the position and number offlowmeters used to measure the instantaneous velocity profilein the fluctuating flow.

Prototype
280 MWe
SUS
3
6 m/s

0.52

0.56

0.60

0.64

Installation position (m)Fig.2-6 Measurement characteristics depending
on installation position and number of ultrasonicflowmeters
Increase in number of flowmeter pairs resulted inconstant flow rate even near the nozzle (highly
fluctuating flow).

temporospatially fluctuating flow can be simulated. Although
the flowmeters are installed in piping A and B at the outlets
of a heat exchanger, the measured flow rate should preferably
be independent of the installation position to increase thelayout design flexibility. Then, the installation position andthe number of ultrasonic flowmeters (which corresponds tothe measured instantaneous velocity profile in the pipe crosssection) were changed, and the flow rate measurement wastested.
A test result is shown in Fig.2-6. For one or two pairs offlowmeters, the measured flow rate varied with the installation
positions relative to the heat exchanger outlet. On the other
hand, for four pairs of flowmeters installed at intervals of 90 °in the circumference angle, the position dependency of themeasured flow rate was reduced, and the flow rates were thesame as that measured by the reference flowmeter. This result
indicates that the ultrasonic flowmeter can be applied to thetemporospatially fluctuating flow.
The present study was performed under a contract with
Mitsubishi Fast Breeder Reactor (MFBR) Systems, Inc., aspart of a study that was entrusted to MFBR Systems, Inc.,by the Ministry of Education, Culture, Sports, Science andTechnology of Japan (MEXT).

Reference
Hirabayashi, M. et al., Development of Ultrasonic Flowmeter for Sodium-Cooled Fast Breeder Reactor −Applicability Verification of Multi-Path UltrasonicFlowmeter−, Jikken Rikigaku, vol.12, no.4, 2012, p.391-397 (in Japanese).
JAEA R&D Review 201335

Research and Development of Advanced Nuclear System

2-3

Aiming at Improved Visualization Technology for Reactor Structures－ Experiment on Sodium Wetting on USV Sensor Surface －Liquid Sodium on Stainless SteelTransmission Sensor(9ch)

240 mmReception Sensor(256ch)

240 mm0s

30 s

Liquid Sodium on Au Plating

0s

30 s

Fig.2-7 Surface of USV sensor
The USV sensor surface was gilded to improve the wetting
between the sensor and the sodium and to heighten thesensitivity to transmission and reception of supersonic waves.

Fig.2-8 Improvement of sodium wetting by gilding
processing
The sodium wetting was improved remarkably by gilding
the sensor surface.

Liquid sodium, which is used in fast reactors, has many
advantages as a coolant; however, it is difficult to opticallyobserve the inside of the reactor vessel because of its
opaqueness. Therefore, an under-sodium viewer (USV) was
developed to inspect the structural integrity of the reactor
vessel, and the improvement of the sodium wetting on thesensor surface is studied for practical use.
The USV transmits supersonic waves into liquid sodium
and receives the reflection waves from the structures in thereactor. The structures are visualized by analyzing the period
from transmission to reception of the supersonic waves, so it
is important to improve the sensitivity to supersonic waves
for practical use of the USV. Current knowledge suggests thatthe sensitivity to supersonic waves depends on the contact
properties between sodium and the sensor surface, namely,
sodium wetting. Therefore, we tried to improve the sodium
wetting by gilding the sensor surface (Fig.2-7) and tested thesodium wetting under various plating conditions.
In the experiment, the test pieces were made under various
plating conditions, using materials such as gold (Au), which

was used in the present sensor; indium (In), which has a low
melting temperature; palladium (Pd), which is a novel metal;
and nickel (Ni), which is generally used as substrate plating.
As a result, the sodium wetting (spreading behavior)
was found to be determined by the characteristics of thesurface plating, particularly its solubility. Fig.2-8 shows thedifferences in the sodium wetting after gilding processing.
The sodium droplet spreads on the Au plating but not on thestainless steel.
The difference is attributed to a more active dissolution
reaction when the plating has a higher solubility. In thisexperiment, the sodium wetting was most substantially
improved with Au plating; the reason is that the solubility of
Au is 25mol%, which is higher than that of In, Pd, or Ni.
This study showed that gilding processing of the present
USV sensor was an appropriate method to improve thesodium wetting. We plan to advance the research on sodium
wetting, including theoretical considerations, focusing on theinterfacial chemical reaction between sodium and the sensor
surface and the effects of impurities in the sodium.

Reference

Kawaguchi, M. et al., Reactive Wetting of Metallic Plated Steels by Liquid Sodium, Journal of Nuclear Science and Technology,
vol.48, no.4, 2011, p.499-503.

36

JAEA R&D Review 2013

Research and Development of Advanced Nuclear System

2-4

Codification of Materials for Demonstration Fast Reactor in JSME Code－ Establishment of Standards for Material Strength and Elevated Temperature Design for 316FR and Modified 9Cr-1Mo Steels －1000

Mod. 9Cr-1Mo Steel

Sodium

600 ℃

CracksStressσ(MPa)250 ℃
T= 500 ℃

Specimen

T= 550 ℃

100

50

Temp
Plate Forging
(℃)

500
550
600
650

○
△
□
◇

T= 600 ℃●▲■◆

φ162.3200

45°

Liquid Penetrant Test

T= 650 ℃― : JSME Fast Reactor Code, 2012 Edition (Average)
10
10

100

1000

10000

1000001000000

Creep rupture time tR (h)Configuration (mm)The evaluatednumberof cycles forし数
the
き裂深さ
が10 mmに達した解析の繰返longest crack
to試験結果reach （約1870回）
a 10 mm depth（約1530回）
は、を(approximately
1530 cycles) wasこと
less
下回り、
解析による評価が妥当であるがthan the observed value (approximately
確認できた1870 cycles), which demonstrated theappropriateness of the evaluation.

Fig.2-9 Creep rupture time of Mod. 9Cr-1Mo steel
Creep rupture data in which the rupture time exceeded 100000 h(approximately 11 years) were obtained. The slope of a graphof rupture time and stress was found to vary below and above
half of the 0.2% yield strength at temperature. It was inferredthat the longer the rupture time, the smaller the effects of plastic
deformation.

Fig.2-10 Example of structural test of Mod. 9Cr-1Mo steel
using sodium
A thick cylindrical specimen was subjected to a sodium flow
with a temperature that varied cyclically between 600 °C and
250 °C that generated thermal loads. It was demonstrated thatthe analysis was appropriate. By performing more tests under
different conditions, the adequacy of the margins in the design
standard was confirmed.

Future fast reactors need to be more reliable, and theirdesign has to be more flexible. The development of structural
materials is one of the keys to achieving this goal, andextensive research and development have been conducted.
For the Japanese demonstration fast reactor, which is in thedesign study phase, structural materials have been selectedthat are most suitable for meeting the requirements forcomponents, such as higher strength at elevated temperatures
or lower thermal expansion rates; 316FR steel was selected
for the reactor vessels and internals, and Modified 9Cr-1Mo(Mod.9Cr-1Mo) steel was chosen for most parts of the coolant
system, including the primary piping system. The 316FR steel
was developed in Japan by the Japan Atomic Energy Agency
and partners for fast reactor use. Mod.9Cr-1Mo steel was
originally developed by the Oak Ridge National Laboratory
in the United States, and extensive R&D programs wereconducted in Japan in order to apply it to fast reactors.
The temperature during steady-state operation of thedemonstration fast reactor will be approximately 550 °C, and adesign life of 60 years is planned. To establish methodologies
to achieve these goals, material data were obtained andevaluated under various conditions such as creep, in which
tensile stress is continuously applied at elevated temperatures;
fatigue, in which thermal stresses are repeatedly imposed;and creep-fatigue, in which creep and fatigue loadings aresuperimposed. For example, long-term creep tests, in whicha heated specimen fails by tensile stress imposed by a deadweight, were conducted; the maximum rupture time so far
obtained exceeds 100000 h (11.4 years). A database has been
established, and the data were analyzed from mechanicaland metallurgical viewpoints to formulate the properties.
In the formulation, the focus for 316FR steel was on taking
advantage of its superior properties under fast reactor
conditions, and for Mod.9Cr-1Mo steel, it was on representing
the creep properties, which vary depending on the stress levels(Fig.2-9). The creep strength of 316FR steel at 300000 h
and 550 °C, evaluated using the formulated properties, was
1.7 times higher than that of the conventional material
SUS304, and the increase was 2.3 times for Mod.9Cr-1Mo
steel compared with 2 1/4Cr-1Mo steel. Moreover, structural
tests in a sodium environment that is closer to that in fast
reactor applications confirmed that the design method
developed for conventional materials is applicable to these
two materials with sufficient conservatism (Fig.2-10). The
above results led to a significant increase in the freedom of
structural design for the demonstration fast reactor.
The above achievements are codified in the Japan Societyof Mechanical Engineers Code for Nuclear Power Generation
Facilities, Rules on Design and Construction for Nuclear Power
Plants, Section II Fast Reactor Standards, 2012 (in Japanese).

Reference
Asayama, T. et al., Creep-Fatigue Evaluation Methodologies and Related Issues for Japan Sodium Cooled Fast Reactor (JSFR), Procedia Engineering,vol.55, 2013, p.309-313.
JAEA R&D Review 201337

Research and Development of Advanced Nuclear System

2-5

Investigation of Sodium/Water Reaction in Collaboration with France－ Sodium/Water Reaction Test for Steam Generator Heat Transfer Tube －Water/Steam(a) Short distance
Failed Tube(a) Short
distance

Adjacent Tube

Failed Tube

Adjacent Tube

ToroidalShape Failure
ToroidalShape Failure

Reaction Jet(b) Long distancePit Shape Failure(b) Long distance
Steam Zone Reacted Zone
Pit Shape Failure
Steam Zone Reacted ZoneNa

Na
Failed Tube

Adjacent
Tube

Na

DefectDefectReaction Products
Reaction Products

Fig.2-11 Steam generator and Na/water reaction jet
Steam generators exchange heat between sodium and water
through heat transfer tubes. If a failed tube is left in place, asodium/water reaction jet attacks adjacent tubes.

Fig.2-12 Na/water reaction jet and tube failure area
The Na/water reaction jet consists of two zones: an innerunreacted steam zone and an outer reacted zone. (a) The
annular reacted zone attacks a tube to cause a toroidal failure
at a short distance, and (b) the reacted zone attacks a tube
at a point to cause a pit-shaped failure at a long distance.

In steam generators in sodium-cooled fast reactors (SFRs),
water is turned into steam by heat from sodium (Na) to supply
steam to a turbine/generator system. Heat transfer tubes areincorporated in a steam generator; sodium flows outside
the tubes and water or steam flows inside the tubes for heatexchange (Fig.2-11).
In the event of heat transfer tube failure, water or steam leaks
into the sodium side and reacts with Na to produce a hightemperature reaction jet and an erosion or corrosion attack. If
this reaction jet continues to attack adjacent tubes, tube failureis also assumed to occur on adjacent tubes (Fig.2-11, right). To
prevent this tube failure escalation, the plant safety systemmonitors the coolant pressure changes and the concentrationof hydrogen, which is produced by Na/water reactions, in the
coolant. As soon as a Na/water reaction is detected, the safetysystem removes water or steam from the heat transfer tubes ina steam generator to terminate the reaction.
To ensure this safety function in steam generators, thereaction jet resistance of the heat transfer tubes needs to beinvestigated. Because the reaction jet resistance depends on
the water leak rate, inter-tube distance, and Na temperature,
the heat transfer tubes have to be examined under various
conditions to study the influence of these parameters. Na/water reaction tests to obtain the tube resistance data wereconducted in Japan (the Japan Atomic Energy Agency, JAEA)
and France (Commissariat à l’Énergie Atomique et aux
Énergies Alternatives, CEA), but both institutes collaborated
for efficient data acquisition.
CEA fabricated heat transfer tube test specimens for thiscollaboration, and these specimens were loaded into a sodium
vessel of JAEA’s Na/water reaction test equipment. This test
equipment provided the same environment as in real steam
generators (~500 °C, ~18 MPa), and the experiment could
be performed under a wide range of conditions appropriate
to its purpose. Water or steam was injected toward the tube
specimen in the sodium vessel, and a reaction jet was formed
here and attacked the tube specimen to penetrate the tube
wall. The tube penetration times were measured to define thereaction jet resistance characteristics. The internal structure of
the reaction jet was also revealed by the configuration on thefailed tube surface after the test and the temperature distribution
in the reaction jet. Fig.2-12 illustrates aspects of the reaction jets
and tube failure areas at different inter-tube distances.
Japan and France will continue this collaboration toaccumulate further data concerning the tube resistancecharacteristics for a few years. The collaborative study is
expected to contribute to enhancing the safety of steam
generators in both countries’ SFRs.

Reference
Beauchamp, F., Nishimura, M. et al., Cooperation on Impingement Wastage Experiment of Mod. 9Cr-1Mo Steel using SWAT-1R Sodium-Water Reaction
Test Facility, Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR13), Paris,
France, 2013, paper IAEA-CN-199-119, 10p., in USB Flash Drive.

38

JAEA R&D Review 2013

Research and Development of Advanced Nuclear System

2-6

Controlling Oxygen Content of U–Pu Mixed Oxide Fuel
－ Oxygen Potential Measurement and Application to Fuel Technologies －2.05

Temperature

at 1600 ℃
2.00

-300

-400

40％Pu

-600
1.80

333
100
22.7
1.95

PuO2

-500

1500

PH2 /PH2 O

30％Pu

1.85

20％Pu

1.90

1.95

1000

10％Pu

2.00

2.05

1.90

MOX pellet
0

2

4

6

8

Temperature (℃)

-200

2000

O/M ratio

Oxygen Potential (kJ/mol)

-100

500

Time (h)O/M ratio
Fig.2-13 Oxygen potential of MOX at 1600 °C as functions of
O/M ratio and Pu content
The O/M ratio decreases with decreasing oxygen potential
and increasing Pu content. The O/M ratio can be controlled bycontrolling the oxygen potential and temperature.

Fig.2-14 O/M change of (U0.7Pu0.3)O2-x during heat treatment
The O/M ratio of MOX changes with PH2 /PH2O ratio. The O/M
ratio reaches 1.94 at 1650 °C in atmosphere with PH2 /PH2O = 333.
Photograph shows MOX pellet set in the measurement system.

Uranium–plutonium mixed oxide (MOX) has been used as amoisture was used for oxygen potential control. In this gas

fuel for sodium-cooled fast reactors; it has a fluorite structure

system, the oxygen potential is determined by the reaction

with an oxygen-to-metal (O/M) ratio of 2.00. UO2 can take

H2O⇔H2+1/2O2 and controlled by adjusting the PH2/PH2O

extra oxygen and exist in a hyper-stoichiometric composition

ratio. Many data (more than 1000 points) were obtained andwith a range of O/M > 2.00. Pu atoms are substituted for U

evaluated, and an equation representing the oxygen potential

atoms, and a MOX solid solution is formed. On the otherwas derived as a function of the O/M ratio, temperature, andhand, MOX can exist in a hypo-stoichiometric region of

Pu content. On the basis of this equation, a technique for

O/M < 2.00 by forming oxygen vacancies. Changes in theevaluating the O/M ratio was developed.

O/M ratio significantly affect various properties of MOX,

Fig.2-13 shows the relationship between the oxygen

such as the thermal conductivity. Therefore, it is important topotential and the O/M ratio at 1600 °C. The oxygen

control the O/M ratio to develop fuel technologies.

potential increases with increasing Pu content. This

It is necessary to know the relationship among the O/M ratio,

technology for evaluating the oxygen potential has been

temperature, and oxygen potential to develop O/M control

applied to the fuel pellet fabrication process and irradiation

techniques. In this work, the oxygen potential (ΔGO2) of MOX

behavior analysis. The O/M change in (U0.7Pu0.3)O2-x during

was measured by the gas equilibrium method. The oxygen

heat treatment is plotted as a function of the PH2/PH2O ratio

potential is a quantity used in thermodynamics to describe thein Fig.2-14. The figure shows that the PH2/PH2O ratio affects

oxygen chemical stability in MOX and can be obtained using

the variation in O/M. Further, the chemical stability ofeq.(1).

fission products inside irradiated pellets can be evaluatedΔGO2 = RT lnPO2 ,⋅⋅⋅⋅⋅ (1)using this technology.

where R is the gas constant. A gas mixture of Ar/H2 to add
Reference
Kato, M., Oxygen Potentials and Defect Chemistry in Nonstoichiometric (U,Pu)O 2, Stoichiometry and Materials Science−When Numbers
Matter, 2012, p.203-218.
JAEA R&D Review 201339

3

Research and Development on Geological Disposal of High-Level Radioactive Waste

R&D Supporting the Technology and Reliability of Geological Disposal in Japan
Overpack(e.g. carbon steel)

Stable Geological
Environment(Natural Barrier)(greater than 300 m
below surface)

BufferMaterial(e.g. bentonite)

Fig.3-1 Basic concept of geological disposal of high-level
radioactive waste in Japanin earnest (Topics 3-1, 3-2, 3-3, 3-4). Underground tunnels
serve as a place for the public to experience deep geologicalenvironments first-hand and appreciate our R&D activities. In
addition, studies on tectonics, volcanic and faulting activities,
and so on have been underway for evaluating the long-term
stability of geological environments in Japan (Topic 3-5).
In parallel with this geoscientific research, we are conducting
an extensive study for performance assessment of multibarriersystems, engineered barrier systems, and long-term chemical
and migration behavior of radionuclides at Tokai to improvegeological disposal technology (Topics 3-6, 3-7, 3-8).
These studies involve the use of geological environment dataobtained at both URLs. A prototype knowledge management
system that was developed in 2009 has been improved tosystematically provide multiple lines of evidence and R&D
results. In 2012, the Information Synthesis and Interpretation
System (ISIS), which supports the implementer in planning,
performing, and evaluating the preliminary investigation,
was developed by integrating each expert system, using theexperience and know-how that experts have garnered throughR&D activities.
Horonobe UndergroundResearch Center(Horonobe, Hokkaido)

Tono Geoscience Center(Toki & Mizunami, Gifu)● Horonobe URL*(Sedimentary Rock )Cooperation and
Collaboration● Mizunami URL*(Crystalline Rock)

Other Organizations

Vitrified
Waste

JAEA Knowledge
Management System

Knowledge Base
Concept image

Tokai R&D Center(Tokai, Ibaraki)・Data・Documents
・Software・Experience and Expertise・Simulations
・Evidence・Expert Systems● Engineering Scale Test & Research Facility (ENTRY)● Quantitative Assessment Radionuclide

Concept image

Migration Experimental Facility (QUALITY)

ENTRY*URL: UndergroundResearch Laboratory

Fig.3-2 Japan Atomic Energy Agency’s R&D activities

40

JAEA R&D Review 2013

Engineered Barrier System

Geological disposal is one of the options for isolating highlevel radioactive waste (HLW) produced during nuclear power
generation from human environments over a long period oftime. This is an issue that remains critical in spite of the revisionof the national nuclear energy policy and one that must now be
dealt with by the present generation. In Japan, spent fuel frompower reactors is reprocessed to extract reusable uranium andplutonium for power generation purposes. The liquids separated
from the spent fuel during chemical reprocessing are solidified
into a stable glass form. According to the Japanese concept,
vitrified wastes are encapsulated in a thick steel overpack,which is surrounded by highly compacted bentonite and placedin a stable geological environment below 300 m (Fig.3-1).
Geological disposal of HLW will be a long-term project lasting
more than 100 years. The project begins with site selection,
continues to repository construction and operation, which
will be followed by post-closure monitoring. It is thus very
important to proceed with the project efficiently, as a national
responsibility, by continuously reinforcing the technical baseand, more importantly, building public confidence. To this end,
we have made and continue to make steady progress in researchand development (R&D) in various fields, such as geoscientificresearch, engineering technology, and safety assessment, toimprove the technology used for and reliability of geologicaldisposal in Japan. Our efforts are also focused on promoting
public understanding through the dissemination of relevant
information and on opening our R&D facilities to the public.
A particular focus of our current R&D is on projects at two
underground research laboratories (URLs)—one at Mizunami
for research on crystalline rocks and the other at Horonobe forresearch on sedimentary formations (Fig.3-2)—with the main
aim of developing a sound technical basis for the formulationand implementation of safety regulations. In March 2013,
shaft excavation reached a depth of 500 m at Mizunami and
350 m at Horonobe, and some research galleries were madeavailable for use. Multidisciplinary investigations are ongoing,as the reliability of a variety of investigative techniques shouldbe tested and confirmed before site characterization begins

QUALITY

Implementer

Regulator

Public Understanding
and ConfidenceResearch and Development on Geological Disposal of High-Level Radioactive Waste

3-1

Spatial Distribution of Fractures in Granite－ Influence of the Initial Cooling of the Granitic Pluton －
LCI

Kfs

Extent of texture
development

Myrmekite

A

SA

10 ｍ

5 μm

Distance
d₁

B

20 ｍ

10 μm

30 ｍ

C

40 μm

０（ｍ）

１．
０

-1000

０．
５

10000
（ｍ）

Region Z

in comparison

50 μm： Slow cooling

LOW
gradientC
50 (μm)Top

E-70000１．
５（μm/m）

SC-６４０００（ｍ）-６8０００

5000

W

B

large
value

Fig.3-4 Local cooling indicator
Concept of local gradient of texture development between two positions, defining the
local cooling indicator (LCI). High local gradient (small value) denotes relatively fast
local cooling.

2.0（>2.0）NLCI

4.0 μm/m^： Fast cooling

Borehole core(across: about 5-12 cm)-６6００００．
０40 μm/10 m

HIGH
gradient

small
value

0

Fig.3-3 Micrometer-scale texture(myrmekite) indicative of coolingprocess of pluton
Myrmekite occurs at plagioclase (Pl)rim in contact with K-feldspar (Kfs).

Top

−

SC-SB
d₂

Pl

0.5 μm/mSB

Depth0.25 mm

5 μm/10 m

5 μm

LCI
(Local cooling indicator)

Implying

A

−

SB-SA
Difference

Development

Local gradient of
texture development

S

Base

-６6０００E-６8０００120（>120）
N
100
80
60
40
20
0

Frequency(N/10 m)-６４０００（ｍ）-70000
０（ｍ）
-1000
10000
（ｍ）
5000

W
Region Z

S

Base

Fig.3-5 Contoured spatial distribution of LCI in the Toki
granitic body
Cool colors denote fast local cooling.

Fig.3-6 Contoured spatial distribution of fracture frequency
Warm colors indicate high fracture frequency.

Fractures in granitic rock (crystalline rock) can act as
groundwater pathways. Therefore, understanding the fracture
distribution is important for the safety assessment of thedisposal of high-level radioactive waste. The present study
proposes that the cooling process in a granitic pluton will
provide key insight for understanding the fracture distribution
on the basis of thermal stress.
Our earlier study revealed that the extent of micrometerscale textures (e.g., myrmekite; Fig.3-3) is a powerfulparameter for quantifying the cooling process in plutonic
rocks. The degree of texture development was measured at
two neighboring positions, and the difference was divided by
the distance between the two positions (Fig.3-4). The resulting
quantity is defined here as the local cooling indicator (LCI).
This is a revolutionary procedure for evaluating the thermal
stress during cooling.
The spatial distribution of the LCI was constructed on thebasis of about 670 textures from 19 boreholes in the Toki
granitic body, Central Japan (Fig.3-5). There is a relationshipbetween the distribution patterns of the LCI (Fig.3-5) and thefracture frequency (Fig.3-6); i.e., regions of fast local coolingcorrespond to regions of high fracture frequency and viceversa (e.g., the central region and region Z). This relationship
implies that thermal stress contributes to fracture genesis.
The spatial distribution of the fracture frequency in a granitic
pluton is constrained mainly by the LCI. This case study
employs a petrographical method, i.e., the LCI, and provides
a constructive study on evaluating the fracture distribution in
granite.
A further study will examine the intrusion and emplacement
process of a granitic magma. The heterogeneity of rock facies
and chemistry in the granite, which was attributed to intrusionand emplacement of the granitic magma, yields a difference inthe petrophysical properties. Fracturing is closely associatedwith this difference. In this way, we aim to elucidate thefracture formation and development processes with higher
resolution.

Reference
Yuguchi, T. et al., Three-Dimensional Fracture Distribution in Relation to Local Cooling in a Granitic Body: An Example from the Toki Granitic Pluton,
Central Japan, Engineering Geology, vols.149-150, 2012, p.35-46.
JAEA R&D Review 201341

Research and Development on Geological Disposal of High-Level Radioactive Waste

3-2

Advancing In Situ Rock Stress Measurement Techniques at the Mizunami Underground Research Laboratory－ Case Study: Successful Overcoring Tests Excluding Groundwater －

Valve

(a)

Fracture with inflow
Packer5” casing

Packer
PQ-size rod

Interception of inflow(b)

HQ-size drilling(c)

BQ-size drilling(d)Fig.3-8 Rock stress measurement
GL.-300 m gallery.

Conical-end drilling(e)Adhesion of strain cell(f)

HQ-size overcoring

Fig.3-7 Drilling tool to prevent inflow and its measurement
procedure
To protect the strain cell from groundwater inflow, an oversizedcasing was installed at the point of the borehole (HQ-size:
φ100 mm), where the strain cell is attached.

The evaluation of in situ rock stress is important in thedesign and construction of underground structures such as
underground power stations and geological disposal facilities
for high-level radioactive waste. In fact, in situ stress isrecognized as the primary factor in the development of an
excavation disturbed zone (EDZ). An EDZ is likely to be lessstable than the undisturbed rock mass and to provide higher
permeability pathways for groundwater flow near excavations.
Hydraulic fracturing tests in a borehole drilled from thesurface and overcoring tests in boreholes drilled from a gallery
were used to determine the in situ rock stresses. In 2009, theJapanese Geotechnical Society standardized the techniques
used in compact conical-ended borehole overcoring (CCBO).
However, the CCBO technique does not work if groundwater
is present because water reduces the effective adhesionof the strain cell to the rock. Furthermore, the occurrenceof disking at the conical-ended overcore does not allow

Fig.3-9 Groundwater inflow in borehole
Diameter of borehole: φ100 mm.

evaluation of the in situ stress.
As a countermeasure, we developed a tool to prevent
groundwater inflow (Fig.3-7) and adapted the tool for use
in CCBO in situ stress determination in the GL.-200 m and-300 m galleries at the Mizunami Underground ResearchLaboratory (Figs.3-8, 3-9). Numerical analysis was used tointerpret the stress concentrations around the conical-ended
core in this case study. The conclusions are as follows:(1) The in situ rock stresses were successfully determinedusing the inflow prevention tool.
(2) Numerical analysis indicates that tensile stress occurredat the ridgeline of the conical-ended overcore andmicrofractures developed at this measurement point.
(3) This work confirms the importance of determining theborehole orientation with respect to the regional stressfields to reduce the risk of core disking.

Reference
Sato, T. et al., Study of Core Disking Phenomenon on Compact Conical-Ended Borehole Overcoring Technique −A Case Study for Measurement in Boreholewith Inflow at Toki Granite−, Journal of MMIJ, vol.129, nos.2-3, 2013, p.59-64 (in Japanese).

42

JAEA R&D Review 2013

Research and Development on Geological Disposal of High-Level Radioactive Waste

3-3

Long-Term Investigation of an Excavation Damaged Zone around a Gallery－ Seismic Tomography Survey for Investigation of an EDZ －

（a）

（b）

Excavation 1（c）Excavation 2

East access shaft

Excavation 3
Excavation 4
Excavation 5
Excavation 6
Excavation 7

Bo
se rehoism leic forsource

Excavation 8

Start section
about
4.0 m

Bo
rec reho
eiv leer forSeismic source
Receiver

Seismic tomography
area

Excavation

1

Excavation

2

Excavation

3

Excavation

4

Excavation

5

Excavation

6

Excavation

7

Excavation

8

Excavation

9

0

5m

2

3m

Excavation 7
Excavation 8
Excavation 9

Excavation 9
Excavation 1001Pre-excavation

250 m W-Gallery

23m

Excavation 10

0

1

Excavation 6

（d）

（e）

2

3m

Excavation 12（ f）

12.0 mExcavation 10
Excavation 11

0

1

Driving direction

Ventilation shaft

Excavation 12

0

1

2

0

3m

150 days after excavation

1

2

3m

0

350 days after excavation

1

2

3m

550 days after excavation
Pre-excavation area

1.1

West access shaft

Fig.3-10 Layout of the seismic tomography
This figure shows a plan view of the 250 m gallery off thewest shaft (250 m W-Gallery).

In response to the excavation of an underground opening,
the host rock properties such as the rock strength andpermeability are expected to change. The resulting zone iscalled the excavation damaged zone (EDZ). In geologicaldisposal of high-level radioactive waste, an EDZ is expected
to provide critical pathways for radionuclides. Therefore, it isimportant to develop a technique for quantitative assessmentof the EDZ.
To investigate the EDZ, seismic tomography surveys havebeen widely conducted in the field. This technique investigatesthe seismic velocity in rock by measuring the travel time ofseismic waves. When a seismic wave passes through fracturesin rock, its velocity decreases. In this study, we estimate theextent of the EDZ on the basis of the decrease in seismic wave
velocity at the Horonobe Underground Research Laboratory.
Fig.3-10 shows a plan view of the layout of the seismic
tomography in a gallery at a depth of 250 m. Seismic
tomography was conducted at the 250 m W-Gallery during

1.3

1.5

1.7

1.9

Seismic velocity (km/s)

2.1

Excavated area

Fig.3-11 Results of seismic tomography survey
These figures show the velocity contour maps at each investigation
stage.

excavation (Excavations 1 to 12) and after the excavationof the gallery was complete. Fig.3-11 shows the results ofthe seismic tomography survey. In the pre-excavation andExcavation 6 stages, shown in figures (a) and (b), the seismic
velocity did not change remarkably. In Excavation 12,
shown in figure (c), a low-velocity area in which the velocity
decreased to around 1.3 km/s developed about 1.0 m into thegallery wall. After the gallery was completely excavated, thelow-velocity area was smaller at 150 days after excavation,
as shown in figure (d), but had expanded again at 350 and
550 days after excavation, as shown in figures (e) and (f),respectively. Consequently, it was found that the EDZ can be
expected to develop about 1.0 m into the gallery wall and thatthe extent of the low-velocity area changes with time.
We aim to examine the long-term change in the low-velocity
area in detail. In addition, we will conduct permeability tests andmeasure the saturation around the tomography area to investigate
the change in the hydrogeological properties in the EDZ.

Reference
Aoyagi, K. et al., Geomechanical Assessment of Excavation Damaged Zone in the Horonobe Underground Research Laboratory, Japan, Proceedings of the
13th Japan Symposium on Rock Mechanics & 6th Japan-Korea Joint Symposium on Rock Engineering, Okinawa, Japan, 2013, p.905-910, in CD-ROM.
JAEA R&D Review 201343

Research and Development on Geological Disposal of High-Level Radioactive Waste

3-4

Regional Estimate of Fracture Permeability－ Hydrogeological Investigation of Sedimentary Rocks －Permeability = Log

Permeability (Fracture+rock matrix)
Permeability (rock matrix)
100

10
8

Yuchi Fm & Koetoi Fm (upper)

Brittleness of rock

Brittleness of rock

100

10
8
Koetoi Fm (main)
With major fault
Without major fault
Wakkanai Fm
With major fault
Without major fault

Koetoi Fm (main)
Wakkanai Fm

1
-2

-1

0

1

2

3

4

5

6

1

7

-2

-1

0

Permeability of rock mass

2

1

3

4

5

6

7

Permeability of rock mass

Elevation (km)Fig.3-12 Relationship between brittleness of rock and permeability of rock mass
When the brittleness is lesser than 8, the permeability of a rock mass is low even if fractures/faults develop.

0

Zone 2
Zone 1
Zone 3
Zone 2-1

Zone 1

Zone 1

Zone 2

Zone 20

Zone 2

Yuchi Fm
Koetoi Fm
Wakkanai Fm

1 km

Brittleness Permeabilityof rock (BRI) of rock mass
Low
Zone 3
BRI≦2
Low
Zone 2 2 ＜BRI＜8High
Zone 1
8≦BRI-1

Fig.3-13 Distribution of brittleness of rock and permeability of rock mass based on geological map
Permeability of rock mass (including fractures) can be inferred from brittleness of rock from geological map.

The fracture permeability of a rock mass is an importantparameter for retardation assessment of radionuclide
migration and, consequently, for the design and constructionof an underground facility. Therefore, the regional distributionof the fracture permeability should be understood before
determining the location and layout of the facility. However,
it is impossible to measure the entire fracture permeability;
hence, some estimation of the permeability from the limited
available data is required. In sedimentary rocks, the matrix
permeability is often estimated by geophysical surveys or
stratigraphic analyses, but it is difficult to estimate the fracture
permeability.
The relationship between the strength and stress state ofrocks and the fracture permeability was studied to develop
a technique for estimating the regional fracture permeability
in sedimentary rocks. Usually in the formation of fractures,a brittle deformation tends to result in high-permeability
fractures, whereas a ductile deformation tends to result in lowpermeability fractures. Whether the deformation is brittle or
ductile depends on the strength, stress state, strain rate, andtemperature at the time of the deformation. Considering thestrength and stress state, a softer rock or a higher confined
stress relative to the pore pressure is likely related toductile deformation. For the depth (>300 m) at which highlevel radioactive wastes are disposed of, the strengths andstress states of rocks are the most sensitive and importantparameters; i.e., it is assumed that the fracture permeability isclosely associated with the strengths and stress states of rocks.
Geological, hydrological, and rock mechanical tests of
sedimentary rocks distributed in and around the Horonobe
Underground Research Laboratory (URL) revealed that, when
the brittleness of a rock is less than 8, the permeability of therock mass is low even if fractures/faults develop (Fig.3-12:
the brittleness is defined as unconfined compressive strength ÷effective vertical stress × 2). This threshold is consistent with
deformational theory assuming elastic material. The regional
brittleness can be estimated using a geological map. Fig.3-13
shows the distribution of the brittleness and permeabilityof rock masses in and around the Horonobe URL. The
relationship between the brittleness and permeability of therock masses facilitates estimation of the regional fracture
permeability in sedimentary rocks.

Reference
Ishii, E. et al., The Relationships among Brittleness, Deformation Behavior, and Transport Properties in Mudstones: An Example from the Horonobe
Underground Research Laboratory, Japan, Journal of Geophysical Research, vol.116, issue B9, 2011, p.B09206-1-B09206-15.

44

JAEA R&D Review 2013

Research and Development on Geological Disposal of High-Level Radioactive Waste

3-5

Dating Fault Activity－ K-Ar Age of Fault Clay －80
Hammer

K-Ar age of biotite

70

Fig.3-14 Outcrop of fault gauge (Sakai Toge fault,
Nagano prefecture)
The black belt indicated by the red arrow is the fault
gauge. A hammer (40 cm in length) is included for scale.

Age (million years)

Fault gauge

Fission track age of zircon

60

50

40

Fission track age of apatite

30

20

M252.9R
M403.7R＜0.1 ＜0.4

＜2

2-6

6-10

Grain size (μm)
Fig.3-16 SEM image of sample M403.7R from fault
gauge in Toki granite
Red dashed circle indicate needle illite crystals.

Fig.3-15 Age trend of sample from fault gauge in Toki
graniteA younger age is obtained for a smaller grain size sample.
The sampling depth was 252.9 m for M252.9R and 403.7 m
for M403.7R.

The activity of faults around a target locality is evaluatedto assess the safety for geological disposal and it is important
to determine the period of fault activity for specific faults.
Although the period of fault activity is generally determinedby the stratigraphy of dated formations, dating methods for
fault derived material are also developed for inapplicable
cases, e.g., when no sedimentary rocks are distributed in thelocality or when a fault is found in the tunnel.
Fault-derived material is generated by violent friction
between rocks in contact as a fault moving. A fault gauge
is a type of fault-derived material (Fig.3-14). Authigenic
clay minerals formed by alteration are contained in the fault
gauge. Methods of separating specific minerals and evaluating
the obtained age are developed to date the authigenic clay
minerals in the fault gauge, because the age of these minerals
defines the end of fault activity. Illite, a clay mineral, canbe dated by K–Ar dating, which is scaled to the decay ofpotassium-40 to argon-40.
Methods evaluating the authigenic illite age in sedimentaryrocks are developed. Finer-grained samples reportedly show
younger ages. The application of illite dating to illite in thefault gauge is under development. The low-temperature
structure illite is assumed to be authigenic in the evaluation
method of sedimentary illite age. Another evaluation method
proposes scanning electron microscopy (SEM) observation ofillite crystals in collected illite samples because illite crystals
form needles in rapid growth during brief thermal activity
after a fault moves, whereas they form sheets in slow growthduring diagenesis of sedimentary rocks.
Two samples were collected from the main pit of Mizunami
Underground Research Laboratory, which penetrates theToki granite. They showed younger ages for finer grain
fractions (Fig.3-15). SEM observation of the samples revealed
needle illite formed after the fault activity (Fig.3-16). Hightemperature structure crystals were also detected in thesamples, although the faults were in the granite host rock. This
indicates a problem with determining alteration after fault
movement; the alteration generates not only low-temperature
crystals but also high-temperature crystals. We will study theillite geochronology for presupposition of the dating method
and methods of sample preparation and observation.

Reference
Yamasaki, S., Umeda, K. et al., Constraining the Timing of Brittle Deformation and Faulting in the Toki Granite, Central Japan, Chemical Geology, vol.351,
2013, p.168-174.
JAEA R&D Review 201345

Research and Development on Geological Disposal of High-Level Radioactive Waste

3-6

Estimating Long-Term Evolution of Environment and Materials around Radwaste in Deep Underground－ Development of Assessment Methodology Considering Geochemical Changes in the Near-Field －Water transfer, change inhydraulic characteristics

HydrologicalFig.3-17 Conceptual model of coupled processes innear-field
It is believed that after an engineered barrier is set, acomplicated condition involving heat release from vitrified
waste, groundwater migration, swelling of buffer material,and geochemical reactions in pore water is induced.

Mechanical(Pore water composition,
Mineral dissolution &precipitation)(Deformation)(°C)

Temperature

120
110
100

Buffer material(Bentonite)
Overpack

90
80
70
60
50

Outputpoint
Vitrified waste

40

1m

1 yr.

10 yr.

100 yr.

Concrete support

1000 yr.
pH
13
12

1m

11
10
9

Concentration of ion-exchange site (mol/m3)

Swelling of buffer material,change in mechanicalpropertiesalms
er d ieth an ert
in n pe sio rong an al p
ha p ic
C ex anhecmChemical

C
ha
ch ng
ar e iac n hte yd
ris r
tic aus lic

C
co liqu han
nc id geen ph intra as
tio e
n

in
e l
ng a esha rm tiC the peroprChange in chemical
properties(Groundwater velocity,
saturation)
Heat transport, change inthermal properties
Change in hydrauliccharacteristics

Thermal(Temperature)

10 3

NaZ
CaZ2
10 2

10 1
0.01

0.1

1

10

100

1000

Elapsed time (Year)8

Fig.3-18 Analysis object
The red circle indicates
output point in Fig.3-20.
The layout assumes
horizontal emplacement.

pH

7
6

Fig.3-19 Analyzed example: evolution oftemperature and pH in the near-field
The disposal system is assumed to be set
in sedimentary rock 450 m in depth.

For the safety assessment of the geological disposal of highlevel radioactive waste, it is necessary to quantify the coupled
thermo-hydro-mechanical-chemical (THMC) processes in
the near-field. Because laboratory and in-situ experiments
cannot realize real phenomena, numerical models are used asa practical approach to bridging the gap between experiments
and real phenomena in long-term near-field evolution, meetingthe need for post-closure safety assessment.
This study investigated the geochemical changes arisingfrom the infiltration of groundwater into a bentonite buffer
under a thermal regime of radiogenic heating arising fromvitrified waste using a computer simulation of a developed
THMC model (Figs.3-17, 3-18).
From the analysis results, the temperature in the engineered
barrier system (EBS) reached a maximum within 20 years. The

Fig.3-20 Analyzed example: history of concentrationof ion-exchange sites in the buffer material
Z is an ion-exchange group. Na and Ca concentrations
change to half and four times their initial values,
respectively, in 10 years and are both restored to theirinitial values after 1000 years.

temperature of the buffer material on the overpack side wasapproximately 95 °C. The temperature of the EBS is foundto drop to 50 °C after 1000 years. Although the groundwater
becomes alkaline under the influence of the concrete support,
it returns to the initial pH until after 1000 years (Fig.3-19).
In the event of infiltration by saline groundwater, sulfate
precipitates as gypsum around the overpack in the buffermaterial, and the Na-type bentonite is changed to Ca-type
by exposure to Ca ions released from the concrete supports(Fig.3-20). These results were consistent with the current
scenarios assumed for geological disposal. It will be possible
to estimate the extent of the overpack corrosion in the EBS
using this developed THMC model.
The present study was sponsored by the Ministry of
Economy, Trade and Industry (METI).

Reference
Suzuki, H., Nakama, S. et al., A Long-Term THMC Assessment on the Geochemical Behavior of the Bentonite Buffer, Genshiryoku Bakkuendo Kenkyu,
vol.19, no.2, 2012, p.39-50 (in Japanese).

46

JAEA R&D Review 2013

Research and Development on Geological Disposal of High-Level Radioactive Waste

3-7

Effect of Natural Organic Matter on Migration of Radionuclides－ Metal Ion Binding Abilities of Deep Groundwater Humic Substances －

Pump

Pump
Flowmeter
10.0
μm

0.45
μm

1.0
μm

NaOH

HCl

Dando soil (Humic acid)

Cartridge filters

DAX-8
resin

Pump
Mixingtank< pH2

Groundwater sample

Dando soil (Fulvic acid)
Mixing
tank
pH6-8

Drain

Fig.3-21 Sequential extraction system used for isolation ofgroundwater humic substances (HSs)
HSs were isolated from groundwater on the basis of theirsorption affinity to hydrophobic resin (DAX-8) below pH2.

O

O

O

O

O

O

O

O

O
O

o-phthalatelogβ = 5.01

Aldrich (Humic acid)m-phthalatelogβ = 3.89

Lake Biwa (Fulvic acid)Horonobe groundwater(Humic acid)Horonobe groundwater(Fulvic acid)
10 3

10 4

10 6

10 5
K / L mol

10 7

-1

Fig.3-22 Conditional binding constants (K) of Eu 3+ togroundwater HSs and comparison with those of HSs fromsurface environments in JapanThe K values were obtained at pH5.0, 0.1 mol L-1 NaNO3, anda loading level (CEu/CL) of 0.7–76 (CEu and CL denote total
concentration of Eu3+ and carboxyl group contents of HSs,
respectively).

benzoatelogβ = 2.78

Fig.3-23 Magnitude relation between the stability constants( ) of Eu3+ to simple carboxylic acids and their structuralfeatures
The complex is more stable as the number of carboxyl groupincreases and their positions approach each other.

Humic substances (HSs) are a type of natural organic matterthat is ubiquitous in the environment. They can form soluble
and stable complexes with metal ions. Because HSs are
known to be dissolved even in deep groundwater, their effecton facilitating the migration of radionuclides in undergroundenvironments is a major concern for safety assessment ofgeological disposal systems for high-level radioactive waste.
Previous works in other countries have shown that deep
groundwater HSs have the same degree of affinity for binding
radionuclides as HSs from surface environments. However,
it is not always true that such findings can be extended to all
deep groundwater HSs because the structural feature of HSs
can depend strongly on their origin.
In this study, HSs were isolated from deep groundwater at the
Horonobe Underground Research Laboratory site (Fig.3-21),
and the conditional binding constants (K) of Eu3+ for the
groundwater HSs were then evaluated by a spectroscopic
method. It was found that the groundwater HSs had thesmallest K value among the HSs examined, which were two ordersof magnitude smaller than those of the humic acid in the Aldrichand Dando soils (Fig.3-22). On the basis of the relationshipbetween the stability constants ( ) of Eu3+ to the simple carboxylic
acids and their structural features (Fig.3-23), the low affinity for
Eu3+ might be due to the fact that the carboxylate groups in thehumic molecule are separated by a considerable distance.
These results demonstrate that some deep groundwater
HSs in Japan have binding affinities that differ from thoseof HSs obtained from surface environments. In addition, ourresults show that the existing model and database, which weredeveloped for HSs from surface environments, cannot be
useful for deep groundwater HSs. Thus, these findings suggestthat a model and database for deep groundwater HSs should
be developed for realistic modeling.
The present study was partially funded by the Ministry of
Economy, Trade and Industry (METI).

Reference
Terashima, M. et al., Europium-Binding Abilities of Dissolved Humic Substances Isolated from Deep Groundwater in Horonobe Area, Hokkaido, Japan,
Journal of Nuclear Science and Technology, vol.49, no.8, 2012, p.804-815.
JAEA R&D Review 201347

Research and Development on Geological Disposal of High-Level Radioactive Waste

3-8

Evaluation of Bentonite Alteration in a TRU Waste Geological Repository－ Modeling of Smectite Dissolution Rate under a Range of Chemical Conditions －this study (pH12, 70 ˚C)Cama et al., 2000 (pH9, 80 ˚C)Sato et al., 2004 (pH7–12, 70 ˚C)Sato et al., 2004 (pH8–13, 30 ˚C)Sato et al., 2004 (pH8–13, 50 ˚C)Marty et al., 2010 (pH9, 25 ˚C)Marty et al., 2010 (pH7, 25 ˚C)

0-1E-10-1.5E-10(x-axis) The saturation
degree becomes higher as∆Gr approaches 0-2E-10-2.5E-10-200

Log [absolute value of measured
dissolution rate (mol/m2/s)]-5E-11-9(y-axis) The dissolution
becomes slower withthe rate approaches 0

Measured dissolution rate (mol/m2/s)

pH12 (70 ˚C) (this study)
pH9 (80 ˚C) (Cama et al., 2000)
Fitting curve of the ∆Gr non-linear
function to the experimental data at pH12

-150

-100

-50

0

∆Gr (kJ/mol)

-10

-11

-12

-13

Measured vs. calculated
dissolution rates plotted on line ofy = x, showing good agreement

-14
-14

-13

-12

-11

-10

-9

Log [absolute value of calculated
dissolution rate (mol/m2/s)]

Fig.3-24 Smectite dissolution rate as a function of thedegree of saturation
Dissolution rates of smectite measured at pH12 (70 °C)
decrease (dissolution becomes slower) with an increase in
the degree of saturation (ΔGr on the x-axis is a function of
the degree of saturation and approaches 0 with an increasein the degree of saturation). Dissolution rates of Cama et al.
(2000) obtained at pH9 (80 °C) show the same tendency.

Fig.3-25 Comparison of calculated and experimentallymeasured smectite dissolution rates
Smectite dissolution rates calculated from the rate equation
developed in this study are in good agreement with thosemeasured experimentally at pH7–13 and 25–80 °C under arange of saturation conditions.

Smectite-rich bentonites have a number of attractive
physicochemical properties that make them suitable for use
as a buffer material in the geological disposal of transuranic(TRU) wastes. Cementitious materials, used for their familiar
engineering properties, can, however, generate high alkalineconditions of pH ≥ 12.5. Such a high pH can potentially cause
alteration of bentonite, including the dissolution of smectite,and consequently the loss of the attractive physicochemical
properties (e.g., the osmotic swelling capacity and consequently
the plasticity and impermeability) of the bentonite buffer.
The dissolution of smectite is, therefore, a key process in theevaluation of the long-term evolution of the bentonite buffer.
In this study, laboratory experiments were conducted toestablish the smectite dissolution behavior under high alkalineconditions, and a dissolution rate equation for smectite wasderived. A novel method using atomic force microscopy
developed by Hokkaido University and Kanazawa University
was used to observe the change in smectite particle size during

dissolution. On the basis of the observed changes in particle
size, an average dissolution rate for each experiment wasdetermined statistically. The dissolution rates of smectite at
pH12 were found to decrease with an increase in the degree ofsaturation (Fig.3-24). By combining these results with otherpublished rate data obtained at pH9, a rate equation describing
the dependence of smectite dissolution on the pH, temperature,and degree of saturation was determined. The smectite
dissolution rates predicted by the rate equation are in good
agreement with experimentally measured values at pH7–13
and 25–80 °C under a range of saturation conditions (Fig.3-25).
The rate equation can therefore be used to describe the smectite
dissolution rates in the bentonite buffer in response to changes
in the chemical conditions expected over the temporal andspatial scales of geological disposal of TRU wastes.
Part of this study was conducted under a contract with the
Hokkaido University.

Reference
Oda, C. et al., Reactive-Transport Model Analyses of Bentonite Alteration Behavior at Alkaline Condition Generated by Cement-Water Interaction in a TRU
Wastes Repository, Nendo Kagaku, vol.51, no.2, 2013, p.34-49 (in Japanese).

48

JAEA R&D Review 2013

4

Nuclear Fusion Research and Development

Toward 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
䡡㻌Exp. was completed䡡㻌Upgrade to JT-60SA started

DEMO
ReactorFusion Frontier Research

ITER Project
Domestic Agency for ITER Project
Demonstration of Fusion Energy
䡡㻌Cadarache
in France
䡡㻌Construction
Started in
Dec. 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

Broader Approach (BA) Activities
Implementing Agency for BA Activities

Satellite Tokamak

Support 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

Fig.4-1 Steps involved in the development of fusion DEMO reactor
Fusion plasma research and fusion frontier research are being pursued to develop a DEMO reactor, aiming for early realization offusion energy.

Crucial Research and Development on Fusion Plasma
Research and Fusion Frontier Research toward the practical use
of fusion energy is being pursued through intensive international
cooperation, for example, the International Thermonuclear
Experimental Reactor (ITER) project, Broader Approach (BA)
activities, and other collaborations (Fig.4-1), aiming for early
realization of a fusion DEMO reactor.
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 for the ITER projectin Japan. The JAEA proceeded with preparing the equipmentthat Japan has agreed to provide and achieved various resultsin terms of technological development (Topics 4-1, 4-2). In
particular, the JAEA began the production of a superconducting
coil for the actual machine ahead of any other country
participating in ITER.
BA Activities
The BA activities are joint projects by Japan and the EU for
conducting supporting research for ITER and research anddevelopment (R&D) for a DEMO reactor, which is the next
step of ITER, aiming for early realization of fusion energy. The
BA agreement came into force in June 2007, and the JAEA wasdesignated as the implementing agency of the BA activities inJapan.
The BA activities consist of three projects: projects in theInternational Fusion Energy Research Center (IFERC), theengineering validation and engineering design activities of theInternational Fusion Materials Irradiation Facility/Engineering
Validation and Engineering Design Activities (IFMIF/EVEDA),
and the Satellite Tokamak Program (STP). Topics 4-3 and 4-4
describe the results achieved in the R&D activities toward a
DEMO reactor in the IFERC. Topics 4-5 and 4-6 describe theresults obtained by IFMIF/EVEDA. In the STP, the constructionof JT-60SA has progressed well (Topic 4-7). Topics 4-8 and 4-9
are also results contributing to JT-60SA.
Fusion Plasma Research
The analysis of the JT-60 experimental data was promoted,
and inter-machine experiments were conducted for the purposeof achieving high economic efficiency of the fusion reactor
by attaining a high plasma pressure. Topic 4-10 is a resultthat clarifies the mechanism of the formation of high plasmapressure.
Fusion Frontier Research
Various R&D activities are executed at the BA site in Rokkasho,aiming at the construction of a technological basis for a DEMO
reactor. Topic 4-11 describes theoretical simulation research thatuses the high-performance computer at IFERC; the result can
provide a basis for controlling the plasma in a DEMO reactor.
JAEA R&D Review 201349

Nuclear Fusion Research and Development

4-1

Multi-Parameter Measurement Using Laser Polarimetry for Fusion Reactor－ Simultaneous Measurement of Magnetic Field, Electron Density, and Electron Temperature －magnetic field(T)
0.0

Propagation
direction oflaser lightElectromagnetic wave

1.6

(a)

0.5

5.0

9.5

(b)

electron temperature
(keV)
0

9

18

27

7

8

(c)

4

height (m)Ovalization owing
to Cotton-Mouton Rotation
effectowing to
Faraday
Linear polarization
effect0.8

electron density(1019m-3)

20

Polarization ellipse-2-4View from propagation direction of laser light

4

5

6

7

radius (m)

8

4

5

6

7

radius (m)

8

4

5

6

radius (m)Fig.4-2 Change in polarization state during propagation oflaser light in plasma
The ovalization of the polarization state is associated with
the Cotton–Mouton effect, and the rotation of the ellipse isassociated with the Faraday effect.

Fig.4-3 Reconstruction of laser polarimetry measurement
data(a) Magnetic field profile, (b) electron density profile, (c) electron
temperature profile. Solid lines in (b) illustrate the laser paths.

It is well known that information about the magnetic
field structure in a plasma is essential for steady-state andhigh-performance operation of tokamak devices. Laser
polarimetry is a method of measuring the magnetic fieldin the plasma. The change in the polarization state of laser
light propagated in the plasma depends on the magnetic
field and electron density. The magnetic field profile can bereconstructed from the laser polarimetry measurement data
and the information on the electron density provided by otherdiagnostics. When the electron density and temperature arehigh (e.g., in burning plasma), the change in the polarization
state becomes complicated. Therefore, it has been a concernthat laser polarimetry will not be applicable to magnetic fieldmeasurement in a future reactor.
The change in the polarization state approximates thesuperposition of the two fundamental processes, the Faraday
effect and the Cotton–Mouton effect (Fig.4-2). In plasma
with a high electron temperature, the Faraday effect isdegraded, whereas the Cotton–Mouton effect is enhanced.
Laser polarimetry is typically used when the Cotton–Mouton
effect is small, and only the Faraday effect is measured.
However, this study proposes laser polarimetry using longwavelengths to obtain a large Cotton–Mouton effect andproposes the reconstruction of the electron temperature fromthe measurement data for the Faraday and Cotton–Mouton
effects. It is reasonably expected that such reconstruction ispossible by using the different dependences of both effects on
the electron temperature.
The above expectation was tested under the conditions of
the ITER poloidal polarimeter. Using the measurement datafrom the polarimeter and information on the location of theplasma boundary, the simultaneous reconstruction of themagnetic field, electron density, and electron temperature was
successfully demonstrated. Fig.4-3 shows the reconstruction
results. These results not only show that magnetic fieldmeasurement is possible in the case of a complicated change
in the polarization state, but also pioneer a new application oflaser polarimetry to electron density measurement.
Existing fusion experimental devices use many diagnostics.
However, a future reactor needs to be operated with fewer
diagnostics. This study suggests that multi-parametermeasurement using laser polarimetry shows promise for future
reactors.

Reference
Imazawa, R. et al., Separation of Finite Electron Temperature Effect on Plasma Polarimetry, Review of Scientific Instruments, vol.83, issue 12, 2012,p.123507-1-123507-5.

50

JAEA R&D Review 2013

Nuclear Fusion Research and Development

4-2

Manufacturing of ITER Divertor Prototype－ First Step toward Manufacturing of Actual ITER Divertor －

（a）

（b）

Inner vertical target(EU)
Outer vertical target
(JA)

Dome
(RF)

〜3.5 mIn
the high heat flux test, a heat
加熱試験では上の写真の黄色枠の部
2
2 for 1000 cycles
flux
of 20
MW/m
分に20
MW/m
の熱負荷で1000回のwas
appliedto theした
tungsten armor
繰り返し加熱を実施(bounded by yellow).

Cassette body(EU)

Fig.4-4 ITER divertor (cassette structure)
The ITER divertor has a cassette structure. High heat
flux components, such as the inner vertical target, outervertical target, and dome, are installed in the cassettebody. There are 54 cassettes in the tokamak.

The Japan Atomic Energy Agency (JAEA), designatedas the Japanese Domestic Agency (JADA), is currently
manufacturing various components for ITER construction.
The ITER divertor components are being manufactured bythree parties (the European Domestic Agency (EUDA), theRussian Federation Domestic Agency (RFDA), and JADA).
The JADA is responsible for manufacturing the outer verticaltarget shown in Fig.4-4.
The divertor exhausts impurities that come from the plasma.
The divertor plate is subjected to a high heat flux from theimpact of the impurities. To withstand the high heat flux,
the surface of the divertor plate is covered with refractorymaterials. The inner target and outer target, which are subject
to severe heat flux, are covered with carbon fiber composite,
and the other components are covered with tungsten. These
armor materials are metallurgically bonded to the copper
alloy (CuCrZr) cooling tubes to achieve a high heat removal
capability. Unique technologies that have been independently
developed by the three parties are used to bond the armor
materials and the cooling tubes. The JADA currently uses
a brazing technology for the bonding. In the ITER project,
qualification of the manufacturing process for the divertor

Fig.4-5 Appearance of prototype PFU and tungsten armor(a) Test frame with four PFUs.
(b) Tungsten armor tiles after high heat flux test. Tungsten has
recrystallized. However, no degradation of the heat removal
capability was found through 1000 cycles at 20 MW/m2.

must be demonstrated by each party to validate its technicalcapability. The parties manufacture prototype componentsthat consist of the same materials that will be used for the realcomponents and perform a high heat flux test to demonstrate
the durability of the prototype components.
The JADA manufactured the first set of plasma-facing
units (PFU#1) as a first step in manufacturing the divertor, as
shown in Fig.4-5. PFU#1 has been transported to the RFDA
for a high heat flux test in the RFDA’s facility. Before the
manufacturing of PFU#1, the JADA conducted qualification
activities, such as mechanical and nondestructive testing of
the braze joints and welding joints. In addition, the brazing
process was formally witnessed by the ITER organization
during the manufacture of PFU#1. In the high heat flux testin the RFDA, PFU#1 showed excellent durability against arepetitive heat flux of 20 MW/m2 for 1000 cycles with nodegradation of the heat removal capability and no surface
crack initiation.
The second high heat flux test campaign in the RFDA isscheduled for autumn 2013. After that, the manufacturing of
the real divertor components will begin.

Reference
Suzuki, S. et al., Development of the Plasma Facing Components in Japan for ITER, Fusion Engineering and Design, vol.87, issues 5-6, 2012, p.845-852.

JAEA R&D Review 201351

Nuclear Fusion Research and Development

4-3

Success in Beryllide Pebble Fabrication－ Development of Granulation Technology for Neutron Multiplier Needed for Production of Fusion Reactor Fuel －Plasma sintering method :Rotating electrode method :This method results in powder surface activationthat enhances powder particle sinterability.

(4) Rotation of beryllide electrode(5) Discharge between beryllide & W
electrodes(6) Centrifugal ejection of molten beryllide(1) Uniaxial pressure(2) Plasma generation for
powder surface activation(3) Resistance heatingWelectrodeRaw material powder

Arc

Punch and die
unit (Graphite)

Rotation

Plasmasintereddisks(φ10x L60 mm)

Beryllide synthesis (electrode fabrication)

Insert material(raw powder)

Moltenberyllide

Beryllide electrode
Beryllide could be simultaneously
synthesized and jointed.

Be12Ti pebbles

1 mmBeryllideelectrode(Closeup)

100 µmPrototypic pebbles of beryllide 1 mm indiameter were successfully fabricated.

Fig.4-6 New beryllide synthesis by the plasma sintering
methodBeryllide could be simultaneously synthesized and jointed.
Further, a beryllide electrode was successfully fabricatedby the plasma sintering method.

Fig.4-7 Granulation by the rotating electrode method
Prototypic beryllide pebbles 1 mm in diameter weresuccessfully fabricated for the first time worldwide.

The tritium used as a fuel in fusion reactors is producedby irradiation of lithium with neutrons that are generated by
nuclear fusion reactions. A neutron multiplier is essential ineffectively generating the fuel because this material increases
the number of neutrons by reactions with the fusion neutrons.
Beryllium metal, which is a conventional material for theneutron multiplier, has a low chemical stability at hightemperatures. This has led to the development of fabrication
technology for more stable beryllium intermetallic compounds(beryllides). Beryllide can be obtained by conventional
powder metallurgy methods; however, it is very brittle.
Therefore, a plasma sintering method was applied inwhich synthesis is conducted after the surface of the raw
material powder is cleaned by electric discharge. Further,
this synthesis method was optimized in the DEMO R&D

building at the International Fusion Energy Research Centersite in Rokkasho-mura, Aomori-ken as a series of Broader
Approach activities. As a consequence, an easily processed
beryllide rod was successfully fabricated by the plasma
sintering method (Fig.4-6).
Furthermore, using the beryllide rod as an electrode in
the rotating electrode method, beryllide pebbles 1 mm indiameter were successfully fabricated for the first timeworldwide (Fig.4-7); this value of the diameter is a target
dimension for use in fusion reactors.
The present result substantiates the fuel tritium production
tests in ITER and greatly contributes to the establishment offuel production technology. Moreover, this result represents
great progress in establishing fuel production technology
toward the fusion demonstration reactor (DEMO) in the future.

Reference
Nakamichi, M. et al., Novel Granulation Process of Beryllide as Advanced Neutron Multipliers, Fusion Engineering and Design, vol.88, issues 6-8, 2013,
p.611-615.

52

JAEA R&D Review 2013

Nuclear Fusion Research and Development

4-4

Accuracy Validation of Nuclear Data for Fusion Reactor Design－ Toward More Accurate Fusion Reactor Design －(unit: cm)

Ironassembly

100φ

DT
NeutronSource20

95

Neutron flux
(n/cm2/lethargy/source neutron): Measuring pointExpt.
Calculation with JENDL-3.3
Calculation with JENDL-4.0

10

-5

10

-6

10

-7

10

-6

10

-5

10

-4

10

-3

10

-2

-1

10

10

0

10

1

Neutron energy (MeV)

Fig.4-8 Layout sketch of iron experiment (vertical crosssection)
Data for neutrons and gammas were measured inside aniron assembly 100 cm in diameter and 95 cm in thickness.

Fig.4-9 Neutron energy spectra at depth of 31 cm in ironassembly
The measured data are plotted as black circles, and thevalues calculated using the JENDL are shown by coloredlines.

Various studies toward the practical use of fusion reactors
as a future energy source have been proceeding step by step.
Fusion power plants produce 14 MeV neutrons through thereactions of deuterium and tritium in high-temperature plasma.
Nuclear reactions of these neutrons with surrounding materials
generate thermal energy (nuclear heating) and tritium (tritium
production), which is used as fuel in the plants. For safety, theneutrons have to be shielded so that they do not leak outside
the fusion power plants (shielding).
The 14 MeV neutrons initiate various nuclear reactions
with all the materials surrounding the plasma. Fusion reactor
designs estimate the nuclear heating, tritium production,
shielding, and so on by considering all the nuclear reactions
using a radiation transport code and database (nuclear datalibrary) of nuclear reaction cross sections (nuclear data)
to characterize the probability that a nuclear reaction will
occur. The typical radiation transport code is the Monte
Carlo code MCNP, which is used worldwide because of its
few calculation approximations. On the other hand, there areseveral nuclear data libraries; the Japanese Evaluated Nuclear

Data Library (JENDL) in Japan, ENDF/B in the US, and JEFF
in Europe. Nuclear data libraries are one of the most important
databases that determine the calculation accuracy of fusion
reactor design results. It is necessary to examine and improve
the accuracy of nuclear data libraries experimentally.
The Fusion Neutronics Source facility in the Japan Atomic
Energy Agency (JAEA) can produce the same 14 MeV neutrons
as those generated in fusion reactors with an accelerator. We
inject these neutrons into various assemblies, as shown in
Fig.4-8, and measure the data for neutrons and gammas inside
the assemblies. We simulate these experiments with MCNP
and the nuclear data libraries and validate the accuracy of thelibraries. Fig.4-9 shows a typical example (iron experiment).
The calculated neutron spectrum below 0.01 MeV with theprevious version, JENDL-3.3, is about 30% larger than themeasured one, but the discrepancy between the calculated andmeasured results decreases to about 15% in the calculation
result with the latest version, JENDL-4.0 (some of the crosssection data for 57Fe were dramatically modified on the basisof our study).

Reference
Konno, C. et al., Detailed Benchmark Test of JENDL-4.0 Iron Data for Fusion Applications, Fusion Engineering and Design, vol.86, issues 9-11, 2011,
p.2682-2685.
JAEA R&D Review 201353

Nuclear Fusion Research and Development

4-5

Toward Construction of Intense Neutron Source for Fusion Reactor Conditions－ Measurement of Thickness and Stability of High-Speed Lithium Flow －Confinement vessel

Fig.4-10 View of entire EVEDA Lithium Test Loop
The liquid Li target is produced in the confinement vesselat the top of the EVEDA Li Test Loop, which is the largest Li
loop in the world.

20 m

Li flow
Nozzle0
149

NozzleObservation
window
50

1650
Probe

30

30

Rod

Bellows
Movable
table

0.36

Fixed table

100

250
390

197

Direction of
D+ beamin IFMIF

The region
corresponding to
the beam footprintin IFMIF(Height: 50)

Concave flow channel(Width: 100)
552

Fig.4-11 Contact-type liquid level sensor (unit: mm)
The probe is connected to the movable table by the rod andcan be positioned with high precision by the stepping motor.

Fig.4-12 Picture of Li target flow and cross-sectional viewof Li flow path (unit: mm)
The Li flow was photographed through the observation
window at a speed of 20 m/s, which exceeds the IFMIF
condition. The red rectangle indicates the deuteron beamfootprint in the IFMIF.

To realize a fusion reactor, the development of materials
capable of withstanding fusion neutron irradiation isindispensable. The International Fusion Materials Irradiation
Facility (IFMIF) is an intense neutron irradiation facility
simulating fusion reactor conditions for the development offusion reactor materials. In the IFMIF, neutrons are generated
by the injection of deuterons into lithium (Li). The Engineering
Validation and Engineering Design Activities (EVEDA) for
the IFMIF are now being conducted as one of the Broader
Approach activities. We are performing experiments at the
EVEDA Li Test Loop (ELTL, Fig.4-10), which is a prototypeof the IFMIF Li loop, with the goal of acquiring validation data
toward the construction of the IFMIF.
The most important validation item is the stability of theliquid Li target, which flows at a speed of 15 m/s along a
concave flow channel (25 mm in thickness) at a temperature of
250 °C with a free surface (freely deformable liquid surface).
The free surface of a high-speed flow is generally unstable.
Thus, it is necessary to address the very difficult problem ofgenerating a high-speed stable flow (required flow thickness
fluctuation: less than 1 mm). Although we anticipated that thecentrifugal force that acts on a concave flow would suppress thefluctuation, we need to measure the actual fluctuation, and wehave to develop a precise measurement device.
We developed a contact-type liquid level sensor in a smaller
Li loop than the ELTL. This sensor is equipped with a sharpprobe and detects contacts between the probe and the liquid
as electrical signals. Analysis of the contact signals yields theproperties of the free-surface waves. Some issues had to beresolved before this sensor is employed as a diagnostic tool in
the ELTL. For instance, although the internal pressure of thesmaller loop is equivalent to atmospheric pressure, that of the
ELTL is a vacuum, and it is necessary to counter the differenceof 0.1 MPa between the inner and outer pressures accordingly.
Thus, we employed a strong structure and a powerful motor.
On the basis of the above design, we fabricated a sensor for
use in the ELTL (Fig.4-11). The performance test results showthat the probe can be positioned with high precision (resolution:
0.1 mm, positioning precision: 0.01 mm) as required.
Now, we are measuring the Li target thickness. The Li target
was successfully produced at a high speed of 20 m/s (Fig.4-12).
The stability of the Li target is almost within the requirement.

Probe tip

Stepping
motor

Reference
Kanemura, T., et al., Fabrication and Performance Test of Contact-Type Liquid Level Sensor for Measuring Thickness Variation of Liquid Lithium Jet in the
IFMIF/EVEDA Lithium Test Loop, Fusion Engineering and Design, vol.88, issues 9-10, 2013, p.2457-2551.

54

JAEA R&D Review 2013

Nuclear Fusion Research and Development

4-6

Realization of Fusion Intense Neutron Source－ Loop Antenna Development for RF Coupling in the RFQ Linac －Superconducting
RF LinacRadio-Frequency
Quadrupole Linac

1.00

H
140 mA 100 keV

5 MeV

9 MeV

M

Refl. Coefficient (–)

Injector

Test cell

Lithium TargetL

40 MeVRF Power system

High( > 20 dpa / y, 0.5 ℓ)Medium ( > 1 dpa / y, 6 ℓ)
Low
( < 1 dpa / y, 8 ℓ)

0.95

0.90

Operation mode

Dipole mode

0.85

0.80

170

175

180

185

190

Frequency (MHz)Fig.4-13 Configuration of IFMIF
A 40 MeV deuteron beam with at 250 mA is injected into a liquid lithium flow, anda 14 MeV neutron field similar to that of a D-T fusion reactor is produced by the
D-Li stripping reaction. Materials are irradiated with these neutrons to evaluate theirrobustness for use in fusion reactors.

Fig.4-16 Reflection coefficient from RFQ
A frequency shift of less than 0.1% for operation
mode is attained, and it is sufficient far from
dipole mode.

2.5
27 mm
40 mm
45 mm
48 mm
73 mm

Coupling factor (–)

2.0

1.5

(kA/m)
5.0

Beam center

Optimized
L=45 mm

Loop antenna

1.0

2.5

0.5
0

0

20

40

60

80

100

0.0

Loop angle (deg.)

Fig.4-14 Measured results of RF
coupling; RF coupling vs. loop
angle for the inserted depth

Fig.4-15 Magnetic field profile by 3-D
simulation code
RFQ magnetic field profile is satisfied
under the design conditions.

To realize a demonstration fusion reactor, it is indispensable
to evaluate the feasibility for use in fusion reactor materials
using 14 MeV neutron irradiation. For this purpose, theInternational Fusion Materials Irradiation Facility (IFMIF) is
being jointly planned by Japan and the European Union under
the agreement of Broader Approach Activities agreement.
In the IFMIF, a neutron field is produced by the deuteron(D)-lithium (Li) stripping reaction. The realization of a
deuteron beam that can be accelerated and injected into liquid
lithium is a key issue (Fig.4-13). In this accelerator facility,
two 125mA beam lines are used. Continuous wave (CW)
operation is also required, and this is the world’s first attempt
at developing such a high-current beam and a CW operation
linac. One beam line consists of an ion injector, a radio
frequency quadrupole (RFQ) linac and a superconducting RF(SRF) linac; their designed output energies are 0.1, 5.0 and
40 MeV, respectively. We developed a new RFQ, because
the RFQ had many technical issues to be overcome for therealization of high-current and CW operation.

Fig.4-17 Photograph of loop antenna
tip module employing an advanced
Japanese technology

In this RFQ linac, a driving RF power of 1.4 MW at 175 MHz
has to be injected to the RFQ cavity to accelerate a large
current of 125 mA in CW mode. For stable and continuous RF
power driving, it is indispensable to reduce the deformation of
the loop antenna and to avoid frequency shifts in the operation
mode. For this purpose, it is crucial to obtain an RF design byemploying a small loop antenna and a short insertion depth.
In this study, the precise RF properties (Fig.4-14) were
evaluated by measuring a real size RFQ mock-up module andanalyzing a 3-D simulation code (Figs.4-15, 4-16). The insertion
depth was successfully optimized. For the world’s first attempt,
an engineering design with cooling water channels in the loop
antenna was also developed. To make uniform cooling channels
following the loop’s curve, an advanced Japanese technology
using uniform minute grains of sand was applied. As a result, theloop antenna tip module with cooling channels was successfully
manufactured (Fig.4-17). With these results, the development of aloop antenna for CW operation is now in sight.

Reference
Maebara, S. et al., Engineering Design of the RF Input Coupler for the IFMIF Prototype RFQ Linac, Fusion Engineering and Design, vol.88, issues 9-10,
2013, p.2740-2743.
JAEA R&D Review 201355

Nuclear Fusion Research and Development

4-7

Progress in Satellite Tokamak Program Project－ Start of Assembly of JT-60SA －Rotary Crane

Cryostat Base

Vacuum Vessel
SectorAssembly Frame

Cryostat Base

Guide RailFig.4-19 Assembly frame for assembly of JT-60SA tokamak
An assembly frame (20 m in diameter) with a dedicated craneis used for effective assembly of JT-60SA tokamak.

Fig.4-18 Start of assembly of cryostat base
A ceremony marking the start of assembly of the JT-60SA
tokamak was held by participants from the Japanese andEuropean governments, local authorities, universities, institutes,
and industry in March 2013.
(a)

(b)

(c)

(d)

(e)

(f)

12 m

Assembly of cryostat baseTemporary installation oflower poloidal field coils

Assembly of vacuum
vessel up to 340 °Assembly of
toroidal field coils

Assembly ofcenter solenoid coil

Assembly of cryostat

Fig.4-20 Major assembly sequence of JT-60SA tokamak
Assembly of the JT-60SA tokamak being undertaken jointly by Japan and Europe started with the assembly of the cryostat basefabricated by Europe in January 2013. JT-60SA aims at first plasma in March 2019 after an assembly period of six years.

The design and fabrication of components for the Satellite
Tokamak Program (JT-60SA), jointly implemented by Europe
and Japan, is progressing well toward the goal of first plasma in
March 2019. Disassembly of the JT-60U tokamak at the Naka
Fusion Institute was completed in October 2012 on schedule.
The assembly of the cryostat base (12 m in diameter and 280 t
in weight), which is the first component from Europe, started
in January 2013, so the assembly of the JT-60SA tokamak
began as planned. The cryostat is a vacuum thermal insulationchamber for superconducting coils cooled at 4 K. The cryostat
base is a base structure for installation of tokamak components
such as the vacuum vessel (VV) as well as the cryostat body.
The assembly of the cryostat base was completed with a highaccuracy of ±0.5 mm in March 2013 (Fig.4-18). The VV andsuperconducting toroidal field coils (TFCs) are installed on thecryostat base one by one in the toroidal direction. For effective
Reference

movement and accurate positioning of these components, anassembly frame (20 m in diameter and 18 m in height), which
provides a dedicated rotary crane and guide rails (Fig.4-19), isused. The VV, which is composed of 10 sectors, is assembled
using the welding conditions validated by research anddevelopment up to the 340° covered by nine sectors, omitting
the last sector, which covers 20 ° (Fig.4-20(c)). Then, each
TFC is moved toroidally along the VV through the openingof 20 ° in the VV. The position of the current center of the
TFC defined during fabrication is adjusted within an error of afew millimeters using a precise metrology device with a laser(Fig.4-20(d)).
Two-thirds of the entire VV procured by Japan has alreadybeen fabricated. The fabrication of the TFC procured by
Europe started in France and Italy, so preparation for theassembly of the TFC is also progressing on schedule.

Shibanuma, K. et al., Assembly Study for JT-60SA Tokamak, Fusion Engineering and Design, vol.88, issues 6-8, 2013, p.705-710.

56

JAEA R&D Review 2013

Nuclear Fusion Research and Development

4-8

Progress in Predicting the Voltage Holding Capability for Large Negative Ion Source－ Key Factor for High-Energy Accelerator toward JT-60SA and ITER －Small electrode (16 cm diameter)● JT-60 negative ion source (1.5 m diameter)
Aperture for beam acceleration is 16 mm in diameter100

accelerator 1st stage
2nd stage
3rd stage1100 apertures

1.1 m

0.45 m1.5 m

Aperture number limit∝(aperture number)-0.15

Voltage holding capability(kV/mm0.5)

500 keV, 22 A
negative ion beam

Plasma production

Surface area limit∝(surface area)-0.13
JT-60SA (1100)
ITER (1280)
10
0

1

10

100

1000

10000

Aperture number
Fig.4-21 JT-60 negative ion source and acceleration grid
The JT-60 negative ion source is the largest negative ion source
worldwide with a height of 1.8 m, a diameter of 2 m, and aweight of 6.5 t. It accelerates 500 keV, 22 A deuterium negative
ion beams.

Fig.4-22 Relationship between aperture number andvoltage holding capability
The voltage holding capability indicates the increase rateof the sustainable voltage with increasing gap length. Red
and blue lines show the limits imposed by the surface area.
Black line shows the limit imposed by the aperture number.

In fusion devices, high-energy deuterium beams (overseveral hundreds of keV) are required to drive the plasma
current. To meet this requirement, a neutral beam injector
based on a negative ion source (Fig.4-21) was developed for
JT-60U and applied to plasma experiments for the first timeworldwide. Recently, a beam energy of 500 keV was achievedby improving the voltage holding capability of an accelerator
whose energy had been limited to below 400 keV.
The voltage holding capability of the accelerator isdetermined by the acceleration grids, which are characterized
by a large number of apertures and a large grid size. A locallystrong electric field is generated around the 1100 apertures
on a wide surface 1.5 m in diameter in the JT-60 negative ion
source, as shown in Fig.4-21. However, the voltage holding
capability was designed on the basis of results obtained witha small electrode (0.2 m in diameter) without a locally strong
electric field. The voltage holding capability with a locallystrong electric field on a large grid has not been studied yet.
First, to investigate the weak points for voltage holding, wedeveloped a visualization system for the breakdown positions.
As a result, light emission accompanied by breakdowns wasobserved around the apertures; these breakdowns occurred at
the 1100 apertures in random order. This led us to consider
that the large aperture number might restrict the voltage
holding capability. Next, the dependence of the voltage
holding capability on the aperture number was investigated,
as shown in Fig.4-22. In this figure, because the sustainable
voltage depends on the square root of the gap length, thevoltage holding capability is defined as the sustainable voltage
normalized by the square root of the gap length. We found thatthe voltage holding capability depends on not only the size but
also the number of apertures. This result also indicates that thevoltage holding capability of the JT-60 negative ion source islimited by the aperture number.
The key point of the progress is the clarification of
the breakdown positions that restrict the voltage holding
capability. This result represents progress in the predictionof the voltage holding capability of the Satellite Tokamak
Program (JT-60SA) and ITER ion sources. Moreover, it
contributes to a better understanding in terms of the vacuum
discharge between electrodes with a locally strong electric
filed.

Reference
Kojima, A. et al., Vacuum Insulation of the High Energy Negative Ion Source for Fusion Application, Review of Scientific Instruments, vol.83, issue 2, 2012,
p.02B117-1-02B117-5.
JAEA R&D Review 201357

Nuclear Fusion Research and Development

4-9

Realization of Compact Joint for Superconducting Coil－ Development of Butt Joint Technique for Central Solenoid of Fusion Device －

TF

EF3
EF2

CS1

EF1

CS2

CS3
CS4

CS

Joint for EF coils

Newly developed joint for CS

EF6
EF5
EF4
Sectional view
Joint

Superconducting
cable

Butt joint

Fig.4-24 Comparison of joints
A compact joint could be attained by making a butt joint using
diffusion bonding.

Joint
Fig.4-23 Superconducting coil system for JT-60SA andjoint part for CS
The diameter of the CS could be maximized within thelimited space by embedding the joint in the coil winding.

To obtain a high-performance plasma of long duration, asuperconducting coil system is constructed for the Satellite
Tokamak Program (JT-60SA). The system, which is one of
the largest after ITER, consists of 18 D-shaped toroidal field(TF) coils, four stacks of central solenoids (CSs), and six
plasma equilibrium field (EF) coils, as shown in Fig.4-23.
The manufacturing of these coils is shared between Japan andEurope. Japan is manufacturing the superconductors for the
CS and EF coils, and the CS and EF coils themselves. Italy
is manufacturing the superconductors for the TF coils, andFrance and Italy are manufacturing the TF coils.
Because even the smallest coil needs >3 km of superconductor,
the connection of superconductors through normal conducting
copper after coil winding is essential. If the heat caused byelectrical resistance at the joint is large, a superconducting
state around the joint cannot be sustained, so the capacityof the cryoplant has to be increased. The required electricalresistance of the joint is <5 nΩ. The conventional joint
adopted for the EF coils (Fig.4-24, left) is manufactured byoverlapping the cables and soldering. However, the diameterof the CS, which induces the plasma current, had to be
maximized as far as possible within the limited space, as shownin Fig.4-23. Therefore, the butt joint technique (Fig.4-24, right)
was developed by improving the joint technique used for the
ITER CS Model Coil. The newly developed device can connect
superconductors after coil winding and can locate joints in thecoil winding. The butt joint is obtained by diffusion bonding
with 30 MPa of bonding pressure and heat treatment at 650 °C.
A joint sample made by this technique was fabricated andtested. The joint sample reached 2 nΩ (<5 nΩ) of electricalresistance under the most severe operational condition of 2 T,
7 K, and 20 kA. Because all the manufacturing processes wereconfirmed, mass production of CSs was started.

Reference
Kizu, K. et al., Development of Central Solenoid for JT-60SA, IEEE Transactions on Applied Superconductivity, vol.23, no.3, 2013, p.4200104-1-4200104-4.

58

JAEA R&D Review 2013

Nuclear Fusion Research and Development

4-10 Clarification of Hydrogen Isotope Dependence of Plasma Heat Diffusion－ Contribution to the Prediction of Fusion Power in ITER －

(a)

(c)
Center

SurfaceHeat diffusivity (m2/s)

Temperature gradientsteepens for heavy isotopes

Hydrogen3

Deuterium

2

1

Theoreticalthreshold above
which the heatdiffusivity rises

4

Hydrogen2

Deuterium
r/a=0.60(b)

6Heat diffusivity (m2/s)

Plasma temperature (keV)

4

0

0

1

Normalized minor radius

0

1

2

3

Inverse of scale length of temperature gradient (m-1)(Indicator of the steepness of temperature gradient)

108Hydrogen

6
4

Fig. 4-25 Heat diffusivity and scale length of temperature gradientfor hydrogen and deuterium plasmas(a) Identical temperature profiles are obtained for hydrogen anddeuterium.
(b) Heat diffusivity for hydrogen is nearly twice that for deuterium.
(c) Temperature gradient becomes steeper for heavier isotopes.

Deuterium

2
0
0.8

0.2

Normalized minor radius

For a fusion reactor, the physical processes of heat andparticle diffusion have been intensively studied. Specifically,next-step devices such as ITER will operate in burning plasmausing tritium. However, current fusion research generally treatshydrogen or deuterium. Therefore, it is a critical issue to clarify
the effects of hydrogen isotopes on the process of plasma heatdiffusion.
In JT-60, the effects of the isotopes of hydrogen anddeuterium on the heat diffusion process have been examined.
As a result, identical temperature profiles were obtained forhydrogen and deuterium, as shown in Fig.4-25(a). However,
the hydrogen case requires nearly twice the heating powerof the deuterium case to sustain this temperature profile. In
other words, the heat diffusivity for hydrogen is nearly twicethat for deuterium, as shown in Fig.4-25(b). It is known that,when the plasma heating power is increased, the temperature
gradient in the plasma core becomes larger and eventually

drives an instability that rapidly enhances the heat diffusion at
its threshold. In this case, owing to its large heat diffusivity,
the temperature gradient is not changed greatly. However,
it is not known how this characteristic of the heat diffusion
depends on the hydrogen isotopes. This experiment revealed
for the first time worldwide that the indicator of the steepnessof the temperature gradient becomes larger for heavier
isotopes, whereas the heat diffusivity becomes smaller forheavier isotopes if the temperature profiles are the same, as
shown in Fig.4-25(c). This result indicates that the growth rateof the temperature-gradient-driven instability depends on thehydrogen isotopes.
This study has remarkably advanced the understanding of
the physical process of heat diffusion. The result not only
contributes to the prediction of the fusion power in ITER but
also provides information for the design of the DEMO reactor.

Reference
Urano, H. et al., Small Ion-Temperature-Gradient Scale Length and Reduced Heat Diffusivity at Large Hydrogen Isotope Mass in Conventional H-Mode
Plasmas, Physical Review Letters, vol.109, issue 12, 2012, p.125001-1-125001-5.
JAEA R&D Review 201359

Nuclear Fusion Research and Development

4-11 Fast Ion Transport Study Using HELIOS－ Fast Ion’s Dance with Alfvén Waves in ITER Plasma －

(ｂ)

(ｃ)
Neutron emission rate (1015 s -1)

Fast−ion−driven
Alfven mode

80

Vertical distance (m)Frequency ν (kHz)

1.0

60
40
20
0

Alfven spectrum

0.0

0.2

0.4

0.6

0.8

1.0

Plasma radius (× 0.95 m)

0.5

0.0

3.0

3.5

4.0

Horizontal distance (m)

(ｄ)

2.0

1.5

１．
０

Experimental
value

beforeafter
relaxation

0.5
0

1

2

3

Fast ion beta (%)

Normalized
neutron emission rate / Vol.

(a)
100

Fast ion beta = 1.3%
1.0
0.8
0.6
0.4
0.2
0.0
0.00 0.25

0.50

0.75 1.00

Plasma radius (× 0.95 m)

Fig.4-26 Properties of the fast-ion-driven Alfvén mode(a) Radial profile and frequency. The radius is normalized by
the volume-averaged plasma radius of 0.95 m. The mode
frequency of 45 kHz is consistent with experimental data, inwhich mode activity in the 30 − 70 kHz range is observed,
and lies precisely on the continuous Alfvén spectrum (whitedots).
(b) Mode structure in the poloidal cross section of the torus.
The mode has toroidal mode number 1, consistent withexperimental results.

Fig.4-27 Effect on the neutron emission rate(c) Dependence of the neutron emission rate on the fast ion
beta. The box indicates experimental values, which are
reproduced with beta values around 1% − 1.5%.
(d) Radial profile of the neutron emission rate per unit volume,
normalized by the central value of the initial profile Blue:initial, red: after relaxation. As in the experiment, the central
value decreases by 25% when the initial fast ion beta usedin the simulation is 1.3%. This is half the beta predicted byclassical calculations without Alfvén waves.

ITER is an experimental reactor intended to create theconditions required to sustain nuclear fusion reactions that canserve as a source of energy. It will confine deuterium-tritium
plasmas at 100 million degrees Celsius and produce neutrons
and helium ions via fusion reactions. The helium ions areborn with a kinetic energy 100 times larger than that of thebulk plasma and are used to heat it via collisions. However,the fast helium ions move with a rhythm that is similar to thatof plasma waves called Alfvén waves, so they dance together
and exchange energy. If the waves reach large amplitudes, thefast ions are pumped out of the confinement region. Thus, it isimportant to evaluate the fast ion transport.
Together with the National Institute for Fusion Science, we
conducted simulations of fast ion transport using HELIOS, asupercomputer operated by the International Fusion Energy
Research Center. The simulation model was validated bysimulating scenarios from the JT-60U fusion experimental
device and by comparing the results with experimental data.
To mimic the ITER conditions in JT-60U, energetic deuterium
ions were introduced by powerful beams. The fast ions wereobserved to drive intensive cyclic bursts of Alfvén wave
activity. Because the bursting phenomena depend on the formof the fast ion distribution in phase space, a novel method wasdeveloped that allows us to initialize simulations with fast ion
distributions computed by an orbit-following Monte Carlocode for realistic beam geometry and collisions.
The simulation results show that the frequency and mode
number of the instability, which triggers profile relaxationaccompanied by the burst, are consistent with experimental
observation (Fig.4-26). By varying the fast ion beta value, we
obtained a value that reproduced the fast ion redistribution
and the resulting changes in the neutron emission rate seen
in experiments (Fig.4-27). In classical calculations, the total
neutron emission rate is overestimated because Alfvén wave
activity is not taken into account. Our simulations indicate thatthe true value of the confined fast ion beta is about a factor of
two lower.
This work was partially supported by a Japan Society for the
Promotion of Science (JSPS) KAKENHI Grants-in-Aid for
Scientific Research (No.22860081, No.23360416, No.25820443).

Reference
Bierwage, A. et al., Role of Convective Amplification of n =1 Energetic Particle Modes for N-NB Ion Dynamics in JT-60U, Nuclear Fusion, vol.53, no.7,
2013, p.073007-1-073007-12.

60

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5

Quantum Beam Science Research

Research & Development on Quantum Beam Technology
interacting with the constituent atoms of a material to change
their configuration, composition, and electronic state. In
medical applications, they are used for radiotherapy, in whicha beam is focused on a cancer cell (Fig.5-1).
We own various quantum beam facilities, such as JRR-3
and J-PARC (Tokai area), TIARA, electron beam and γ-ray
irradiation facilities (Takasaki area), J-KAREN and otherlasers (Kizu area), and the SPring-8 beamlines (Harima area).
We have been conducting research and development (R&D)
on advanced beam technology and promoting a variety offundamental and applied research studies in various fields(materials science, environment and energy, medicine andbiotechnology) by using the creating and probing functionsof quantum beams. In this chapter, we introduce our recent
research topics related to advanced beam technology and theapplication of quantum beams to those fields (Fig.5-2).
In addition, we are contributing to the recovery from theaccident at the Tokyo Electric Power Company, Incorporated
Fukushima Daiichi Nuclear Power Station, for example, bydeveloping improved decontamination materials (Chapter 1,
Topic 1-9).

“Quantum beam” is a generic name for neutron beams, ion
beams, electron beams, high-intensity lasers, and synchrotron
X-rays, which are generated by accelerators, high-intensity
laser facilities, and research reactors. Recently, quantum beamtechnology has been developed considerably, with the mostadvanced manufacturing and observations being performed
using highly controlled quantum beams.
Quantum beams function as probes in observations ofatomic- or molecular-level information by control of thebeam parameters. They also allow us to process materials at
the nanometer level (i.e., the atomic or molecular level) by
Functions of Quantum Beam
Observe

CreateObservation at
atomic resolution

Atomic level
fabrication andcreation

Cure

Dr. Q

Cure andtreatment
at cell level

Fig.5-1 Characteristics of quantum beamsLife Innovation

Green Innovation
Material Science
Environment
and Energy

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

Advanced BeamTechnology
NeutronProtonSynchrotron
X-ray

JRR-3

Research Reactors

TIARA

Electron
γ-ray

Ion

PhotonJ-PARCRadioisotopes

Accelerators

SPring -8

High-Intensity Laser Facilities

LasersFig.5-2 Quantum beam facilities in Japan Atomic Energy Agency
JAEA R&D Review 201361

Quantum Beam Science Research

5-1

Enhancement of Thermal Stability of Nanocomposites－ The Role of Nano-Additives on Order-Disorder Transition of Nanocomposites －TODTIm-1 / 10-6 (nm3/e2)

(B)

TODT

Nanocomposites = impurities (a and a’) + bcps (b and b’)
(Pd)n

8K

(A)

PMMA

PMMA

(a’)

Ordered state(T＜TODT)

1.0

PS(Pd) n- (PS- -PMMA ) m

PMMA

(a)

Disordered state(T＞TODT)
0.1

2.0

(b’)
PS

PS - -PMMA

2.1

(b)

(c)

T -1 / 10-3 (K-1)
Fig.5-3 Effect of (Pd)ns on ODTIm-1 plotted as a function of T -1 for A (neat PS-b-PMMA) andB (nanocomposite). TODT is defined as the temperature at thediscontinuous change in I m-1. Black and bright patterns aretransmission electron microscopy images of the bcp template in
the disordered and ordered states, respectively.

Fig.5-4 Interpretation of how (Pd)ns increase TODT(a) Single (Pd)n tethered by m PS blocks, (Pd)n-(PS-b-PMMA)m,and (b) single PS-b-PMMA chain, and their respective coarsegrained models (a’) and (b’) for their chain conformations andspatial segment distributions in the field of the lamellar template(c). The impurities (a’) serve as anchors to thermally stabilize
the template made by the major component (b’).

Nanocomposites made of block copolymer (bcp) templates
and nano-additives have attracted much attention because
bcps are well known to organize into diverse multiphasestructures with nanoscale periodicity, and the incorporationof nano-additives is of great importance for manufacturing
optoelectronic and mechanical devices. The nano-additives
incorporated into bcps perturb the nature of the order-disordertransition (ODT) and the ODT temperature (TODT), which
influences the thermal stability of the nanocomposites. Thus,
the nanocomposite is thermally tunable, which might offer
technological opportunities in the future. In this research, we
describe how the incorporation of palladium nanoparticles,(Pd)ns, into a lamella-forming poly(styrene)-block-poly(methyl
methacrylate) (PS-b-PMMA) alters its ODT.
(Pd)ns were incorporated into a PS-b-PMMA template bythermal reduction of Pd(acac)2 at 403 K. The ODT behaviorof the neat bcp and the nanocomposites was investigated bysmall-angle X-ray scattering as a function of temperature (T).
The plot of the inverse scattering intensity maximum (Im-1)

vs. T-1 in Fig.5-3 indicates that the incorporation of ~1 wt%of (Pd)ns is sufficient to increase TODT by 8 K. This finding
should be attributed to the fact that (Pd)ns increase theeffective segregation power between the PS and PMMA blockchains; a proposed explanation for this is selectively better
incorporation of (Pd)ns into PS lamellae (~70 wt%) than into
PMMA lamellae (~30 wt%). As illustrated in Fig.5-4, theattractive interactions bind some of the PS block chains to thesurface of (Pd)ns and form impurities of (Pd)n-(PS-b-PMMA)m(a, a’) against neat bcps (b, b’) inside the lamellar template(c). The impurities serve as anchors that suppress Brownian
motion of the bcp chains in the template both parallel andnormal to the interfaces; hence, they thermally stabilize thelamellae and increase TODT.
Most importantly, the addition of (Pd) ns to the bcp
template, as a strongly correlated system normal to theinterface, is evidently able to change the ODT significantly,
which offers new opportunities for the development of thesenanocomposites.

Reference
Zhao, Y. et al., Order-Disorder Transition of Nanocomposites: Polystyrene-block -Poly(Methyl Methacrylate) with Palladium Nanoparticles,
Macromolecules, vol.46, 2013, p.957-970.

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5-2

Quantitative Analysis of Conformational Changes in Proteins upon DNA Binding－ Toward an Understanding of the Molecular Mechanisms of Protein-DNA Recognition －: DNA interface: Non-interface

DNA binding

Amino acid propensity

0.6Frequently
observed

0.4
0.2
0.0
–0.2

Infrequently
observed

–0.4

Glu

Asp

His

Arg

Lys

Cys

Gln

Asn

Thr

Ser

Tyr

Trp

Hydrophilic
Phe

Met

Ile

Leu

Ala

Pro

Val

Hydrophobic

–0.6

Gly

DNA

Amino acidFig.5-5 Conformational changes inbovine papillomavirus E2 protein
upon DNA binding
Cervical cancer is caused by infection
with papillomaviruses. This E2 protein is
directly involved in viral DNA replication.
When this protein binds to DNA, an
unstructured region (in orange) in DNAfree form adopts a stable conformation
so that it can bind to DNA.

Fig.5-6 Propensities of amino acid residues in conformationally changed
fragments
A positive propensity value indicates that an amino acid is observed more
frequently in conformationally changed fragments than in conformationallyunchanged fragments. Conformationally changed fragments in DNA interfaces andnon-interfaces have high amounts of glycine (Gly), proline (Pro), and hydrophilic
residues. Error bar indicates statistical uncertainty.

Protein synthesis is controlled by DNA-binding proteinsthat bind to specific sites in genomic DNA. Therefore, tounderstand the fundamentals of gene regulation, it is important
to elucidate the molecular mechanisms by which DNAbinding proteins recognize specific DNA targets.
Structural studies of DNA-bound as well as DNA-free forms
have been conducted to clarify the molecular mechanisms
of individual DNA-binding proteins. However, the genericmechanisms of protein-DNA recognition are not yet fully
understood.
It is often reported that the DNA interfaces of proteins
undergo conformational changes in order to recognize
specific DNA targets (e.g., Fig.5-5), but the changes are likely
to depend on individual proteins. To assess the structural
rearrangement quantitatively, we developed a new method.
In this method, protein structures are described as a onedimensional string of letters. Each letter expresses theconformation of a contiguous four-residue fragment. This
enables us to rapidly compare structures and detect theconformational changes in proteins. Using this method, weanalyzed a set of proteins whose structures were solved inboth DNA-free and DNA-bound forms.
We found that DNA interfaces have higher conformational
flexibility than non-interfaces, and that conformationallychanged fragments in DNA interfaces have a large amountof glycine, proline, and hydrophilic residues (Fig.5-6). Those
residues have previously been observed in intrinsicallydisordered regions. These characteristics of the amino acids
indicate that DNA interfaces evolve to have amino acids
that can exhibit high flexibility for accommodating DNA
conformations.
DNA recognition by proteins is one of the fundamental
processes in gene regulation. Our findings may contribute to
the development of a method that accurately predicts DNA
binding sites in proteins. Such information can be used todevelop drugs that prevent proteins from binding to theirtarget DNA.

Reference
Sunami, T. et al., Local Conformational Changes in the DNA Interfaces of Proteins, PLOS ONE, vol.8, issue 2, 2013, p.e56080-1-e56080-12.

JAEA R&D Review 201363

Quantum Beam Science Research

5-3

Clear Imaging of Tumor with D-Amino Acid
－ Development of a Novel Amino Acid Tracer, D-[18F]FAMT, for PET Diagnosis of Cancer －L[ 18F]FAMT

D[ 18F]FAMTWhole body of mouse

Whole body of mouse

MirrorKidneyKidney

Tumor

Tumor
L

18

[ F]FAMTD18[ F]FAMT

Bladder

Bladder

Low

High
Radioactivity ratio (organ or tissue/muscle)

Fig.5-7 Structure of L-[18F]FAMT and D-[18F]FAMT
L-[18F]FAMT (left) and D-[18F]FAMT (right) are enantiomers ofeach other. Their structures are mirror images of each other, as
in the relationship between the left and right hands.

Fig.5-8 Comparison of images with L-[18F]FAMT and
D-[18F]FAMTIn the image with L-[18F]FAMT (left), renal accumulation
was high and the tumor contrast was low, whereas, in theimage with D-[18F]FAMT (right), the renal accumulation
was low and the entire tumor was clearly imaged. (Thesame mouse was used for all PET imaging.)

Diagnosis with positron emission tomography (PET) isa useful tool for understanding the physiological status ofcancer, as well as for early detection. In PET diagnosis, γ-rays
emitted from the patient are quantitatively detected outside of
the body after the injection of PET tracers, and reconstructed
images are used for diagnosis. 3-[18F]fluoro-α-methyl-Ltyrosine (L-[18F]FAMT), an amino acid derivative, is clinicallyused for definitive diagnosis because it is selectively taken
up by malignant tumors. However, because L-[18F]FAMT isalso accumulated and retained in the kidney and pancreas, it
exhibits slow clearance from the blood. Therefore, the border
between a malignant tumor and benign tissue in the images is
unclear, making it hard to distinguish the region of the tumor
accurately. For accurate diagnosis, we aimed to develop a new
tracer that enables clear visualization of tumors by improving
the non-specific accumulation and retention of L-[18F]FAMT.
The new tracer should have a low affinity to both renal andpancreatic cells. The chemical structure is generally modified
to avoid uptake into the cells. However, we thought that
accumulation and retention in the kidney and pancreas could
be suppressed while retaining the advantage of L-[18F]FAMT,

not by modification of the chemical structure, but by using
the fact that D-amino acids, the enantiomers of L-amino acids,
rarely accumulate in normal tissue and are rapidly excreted
from the kidney to urine. Consequently, we developed 3-[18F]
fluoro-α-methyl-D-tyrosine (D-[18F]FAMT) (Fig.5-7).
Biodistribution studies in tumor-bearing mice showed thatD-[18F]FAMT rapidly cleared from the blood, and less of it
accumulated in the kidney and pancreas, compared to L-[18F]
FAMT. Although the amount of D-[18F]FAMT in the tumor
was reduced, the tumor-to-blood ratio of D-[18F]FAMT was
higher than that of L-[18F]FAMT. PET imaging with D-[18F]
FAMT showed high tumor-to-background contrast and low
accumulation in the kidney, indicating that D-[18F]FAMT
could exhibit lower renal accumulation and enable clear
visualization of the tumor (Fig.5-8).
If PET using D-[ 18F]FAMT is widely applied, cancer
therapy based on more accurate diagnosis would be expected.
Furthermore, D-[18F]FAMT would be an effective tracer for
renal and pancreatic cancer, which cannot be detected using
L-[18F]FAMT. In the future, we will perform a toxicologicalanalysis of D-[18F]FAMT to support practical application.

Reference
Ohshima, Y. et al., Biological Evaluation of 3-[18F]Fluoro-α-Methyl-D-Tyrosine (D-[18F]FAMT) as a Novel Amino Acid Tracer for Positron Emission
Tomography, Annals of Nuclear Medicine, vol.27, no.4, 2013, p.314-324.

64

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5-4

Use of Diamonds for Single Ion Detection－ Real-Time Detection of Ion Impact Position by Fluorescence of Nitrogen Vacancy Centers －Fig.5-9 Photograph of diamonds irradiated by ultraviolet
light
Photographs of diamonds containing dense nitrogen (left),
electron-irradiated diamonds (middle), and diamonds
containing NV centers after thermal treatment (right). The
colors are peculiar to the impurities and defects.

No irradiation

15

N 56 MeV20

Ne 75 MeV40

Ar 150 MeV

84

Kr 322 MeV

129

Signal intensity (a.u.)

6000
5000

NV 0

4000
3000

NV

Normalized counts (a.u.)

1

20 µm

‒

2000
1000
0
560

600

640

680

720

760

Wavelength (nm)

Xe 454 MeV

0

Fig.5-10 Photoluminescence spectrum of created NV
centerAfter electron irradiation and thermal treatment, thediamonds contain electrically neutral NV 0 centers and−negatively charged NV centers.

Fig.5-11 CCD images when an ion strikes a diamond
When an ion strikes a diamond containing NV centers, thevery weak luminescence light from the ion impact position is
successfully detected.

A single ion causes temporary malfunctions and permanent
failures in semiconductors, which are called single event
effects (SEEs). To use semiconductors in harsh radiation
environments such as space, it is necessary to clarify themechanism of SEEs. This requires a technique for detecting asingle ion in real time. The technique of detecting the positionof fluorescence induced by a single ion on a luminescent sheet
is currently used. Therefore, we are exploring a luminescentsheet having strong fluorescence intensity capable ofdetecting a single ion. In this study, we focus on the featuresof nitrogen-vacancy defects (NV centers), which efficiently
absorb ultraviolet light and emit high-intensity fluorescence.
First, diamonds containing dense NV centers werecreated from synthetic diamonds containing dense nitrogen
impurities. By irradiating the sample with a 2 MeV electron
beam, vacancies were introduced into the diamond by thedisplacement damage mechanism. After electron irradiation,
thermal treatment at 800 °C was performed for 2 h. By

combining nitrogen impurities with vacancies, we successfully
created a diamond containing dense NV centers (Fig.5-9).
Photoluminescence spectrum analysis clearly revealed thatelectrically neutral NV0 centers and negatively charged NV
centers were created (Fig.5-10). As shown in the figure,additional defects other than NV centers were not found. Next,
the sample was irradiated with several ions at various energies.
A measurement system consisting of a highly sensitive CCD
camera and an image intensifier was developed. Using thesystem, fluorescence was successfully detected in real time
(Fig.5-11). The spots are detected in the diamonds containing
NV centers, but not in those without NV centers. This result
is regarded as demonstrating a fundamental technology for
elucidating the mechanism of SEEs that reduce the reliabilityof semiconductors.
This work was partially supported by the Strategic
Japanese–German Joint Research Program of the Japan
Science and Technology Agency (JST).

Reference
Onoda, S. et al., Diamonds Utilized in the Development of Single Ion Detector with High Spatial Resolution, Transactions of the Materials Research Societyof Japan, vol.37, no.2, 2012, p.241-244.
JAEA R&D Review 201365

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5-5

Toward Nondestructive Assay of Fissile Materials－ Generation of 500 keV Electron Beam from DC Photoemission Gun －

600

2.0

500
400

1.5

300

Beam energy (keV)= acceleration voltage (kV)×elementary charge (e)

1.0

200

0.5
0

100
0

1

2

3

4

5

6

DC photoemission gun

Acceleration voltage (kV)Beam current (mA)

500 keV beam generation

2.5

0

ys

Energy Recovery Linac

tor

ra

ele

s

en

Int

er

mb

rc

pe

Su

g

tin

uc

d
on

c
ac

-ra
eγ

n

pto

r

se

La

g

rin

tte

a
sc

a
ch

m
Co

Time (min)Fig.5-12 Generation of a 500 keV electron beam
A DC photoemission gun generates an electron beam uponlaser irradiation. The black arrows represent the periods wherea 500 keV electron beam was generated. The blue dashedline represents acceleration voltage; and the red solid line, thebeam current.

Fig.5-13 -ray source based on energy recovery linac
An energy recovery linac is under development at KEK.
The accelerated electron beam is delivered to the Compton
scattering chamber for γ-ray generation. A proof–of-principle
experiment for the proposed nondestructive isotope identification
will be conducted next year.

The quantification of fissile materials such as uranium-235(235U), plutonium-239 (239Pu), and minor actinides in spent fuelassemblies requires isotope-specific identification. We propose
the use of nuclear resonance fluorescence (NRF) to identify
the isotopic composition of sample materials nondestructively.
A mono-energetic γ-ray beam tuned to the resonance state of
the target nuclear isotope is injected into the spent fuels, andNRF γ-rays generated spherically from the fuels are detectedwith a γ-ray detector.
The proposed nondestructive isotope identification
system requires a high-intensity mono-energetic γ-ray beam.
Although a mono-energetic γ-ray beam can be generatedusing a conventional laser Compton scattering technique, thegeneration of a high-intensity γ-ray beam requires an electron
beam of unprecedentedly high brightness, which can begenerated with an advanced accelerator system known as theenergy recovery linac (ERL). A technological challenge of the
ERL system, which we have addressed, is the development of
a high-brightness, high-current electron gun.
The generation of high-brightness beam requires a gun
exit energy of ≧ 500 keV to avoid the beam degradationthat occurs due to nonlinear space charge effect. The gun
operational voltage has, however, been limited to 350 keV

because of a discharge problem. We have successfully
solved this problem by employing a segmented insulator andoptimizing the gun configuration.
Fig.5-12 shows the result of a beam generation test. The
electron beam is generated upon laser irradiation. The beam
current is proportional to the laser power. As shown in the
figure, we successfully generated a 500 keV electron beamwith a current of up to 2 mA.
The gun was recently moved to the compact ERL (cERL)
facility at KEK, where it is connected to an adjoining
superconducting accelerator, as shown in Fig.5-13. Monoenergetic γ-ray generation using Compton scattering will becommenced next year.
This work was supported by the Ministry of Education,
Culture, Sports, Science and Technology of Japan (MEXT),
Quantum Beam Technology Program, “Development of anext-generation compact, high-brilliance light source using
superconducting accelerator technology” and partially
supported by the Japan Society for the Promotion of Science(JSPS) KAKENHI Grant-in-Aid for Scientific Research (C)No.23540353, “Study of micro-bunching instability using subpicosecond bunch measurement.”

Reference
Nishimori, N. et al., Generation of a 500-keV Electron Beam from a High Voltage Photoemission Gun, Applied Physics Letters, vol.102, issue 23, 2013,
p.234103-1-234103-4.

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5-6

Relativistic Harmonics from Tenuous Plasma－ Discovery of a New Compact Coherent X-ray Source －

(a)

(c)

Multi-TW
laser(room scale)eam

Coherent X-ray beam

10λ0 y10λ
0 x

Gas jet(b)70

0CCD Counts 250
80

λ(nm)1690 100
14

12

150
10

8

High-harmonicsourcez, ne
10λ0Laser b200 ℏω(eV)6Electron
spike

Ca

ne

Bow
wave

vity

Laser

4

Fig.5-14 New regime of high-order harmonic generation by relativistic laser focused onto gas jet target(a) Experimental setup. (b) Typical single-shot raw data. (c) 3D PIC simulations demonstrating electron density spike at the jointof the boundaries of the wake and bow waves and location of high-harmonic source.

Bright X-ray sources are necessary for fundamental research
and applications in the life sciences, material sciences,
nanotechnology, and other fields. The ultimate performancerequires temporally and spatially coherent X-ray pulsesthat are compressible to the shortest durations, focusable tothe tiniest spots, and capable of producing fine interference
and diffraction patterns. Two broad classes of bright X-ray
sources are available, those based on large-scale accelerators(synchrotrons, X-ray free-electron lasers) and those based oncompact lasers (e.g., laser plasma X-ray emission, plasmabased X-ray lasers, atomic high-order harmonics, and betatronsources). The great advantages of the laser-based sources
are their accessibility for university-scale laboratories andtheir ultrashort pulses with durations from picoseconds
down to <100 as (10-16 s). However, because of fundamental
limitations, it has been overwhelmingly difficult to achievelaser-based bright coherent X-ray sources with keV andespecially multi-keV photon energies. Our aim is to develop
a new generation of such sources to create new fields offundamental research and applications.
We discovered a new regime of high-order harmonic
generation by high-power (10–200 TW) relativistic irradiance(>1018 W/cm2) femtosecond lasers (~30–50 fs) focused on
gas jet targets, as shown in Fig.5-14(a). Comb-like spectrawith hundreds of even and odd harmonic orders (Fig.5-14(b)),

reaching a photon energy of 360 eV, including the “water
window” spectral range, were generated by either linearlyor circularly polarized pulses from the J-KAREN (KPSI,
JAEA) and Astra Gemini (CLF, RAL, UK) lasers. A 120 eV
harmonic contained up to 4 × 109 photons. Using particle-incell (PIC) simulations and mathematical catastrophe theory,
we introduced a new mechanism of harmonic generation by
sharp, structurally stable, oscillating electron spikes at thejoint of the wake wave and bow wave boundaries, as shown in
Fig.5-14(c).
Our compact bright coherent X-ray source can be built
on a university-lab-scale repetitive laser and accessible,
replenishable, and debris-free gas jet target. Importantly,
the photon energy and number of photons are scalable withthe laser power. This will affect many areas of fundamental
research and applications requiring a bright X-ray/extreme
ultraviolet source for pumping, probing, imaging, or
attosecond science.
We acknowledge the financial support from the Japan
Atomic Energy Agency (JAEA) (Research and Development
Adjustment Funding for Exploratory Research, “Relativistic
extreme ultraviolet harmonics from underdense plasma,” 2009)
and the Japan Society for the Promotion of Science (JSPS)[KAKENHI Grant-in-Aid for Young Scientists (B) No.23740413,“New high harmonic generation mechanism,” 2011].

Reference
Pirozhkov, A.S. et al., Soft-X-Ray Harmonic Comb from Relativistic Electron Spikes, Physical Review Letters, vol.108, issue 13, 2012, p.135004-1-135004-5.

JAEA R&D Review 201367

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5-7

Heatproof Seismic Sensors Guard Aging Nuclear Power Plants－ Application of Femtosecond Laser Processing to Maintenance Study －

0.02

8 μm

0.015

Wavelength shift (nm)1.6 μm

0.01
0.005
0
– 0.005– 0.01– 0.015– 0.02
– 0.5

0

0.5

1

1.5

2

2.5

3

3.5

4

Time (s)Fig.5-15 Fiber Bragg grating
Dots with 1.6 μm pitch along an optical fiber core 8 μm in
diameter processed by femtosecond laser pulses.

Fig.5-16 Vibration monitoring by FBG sensor
Damping oscillation of Cr in steel below 400 °C.

It is necessary to monitor the structural health of the coolant
piping systems in operational nuclear power plants to reduce
the risk of cracking accidents due to thermal expansion stressor intense seismic force. However, commercially available
sensors cannot be used at the operational temperature of fast
breeder reactors.
We successfully developed fiber Bragg grating (FBG)
sensors by femtosecond laser processing that can realize index
modulation along an optical fiber core. The reflectivity wasenhanced at the wavelength that corresponded to the periodic
spacing of the modulation. If the sensors are installed along
pipe lines, the wavelength shift can tell us the deformation
or oscillation caused by an intense external force. The
high sensitivity ensures structural monitoring for not only
extraordinary events but also daily operation.
When femtosecond laser pulses are focused inside silicate
glass, the locally heated zone expands far beyond the melting
point. After quenching, the central heated zone exhibits
low density, whereas the outer zone exhibits high density.
Thus, femtosecond laser processing, called the optical knife
technique, can generate index modulation in glass material
that can be maintained permanently below the melting
temperature. The wavelength of ultraviolet lasers cannot be
used because of the absorption in the clad glass around thesurface. The pulse duration of nanosecond lasers is too longto produce thermal cracks in the fiber core. Only an infrared
femtosecond laser can process the periodic dot that can act as
the grating (Fig.5-15).
In addition, the most important technology for the
installation of FBG sensors on the piping systems in nuclear
power plants is reinforcement with heatproof fabric. Here
we selected SiC fiber, which is strong enough to protect the
optical fiber from degradation caused by heat and radiation.
The combined optical fiber and SiC fabric successfully
strengthen the sensor for handling. We found the appropriate
temperature conditions for each heatproof epoxy and ceramicadhesive. The strengthened FBG sensor exhibited superior
linearity from room temperature to 600 °C. Further, it
demonstrated monitoring of the seismic vibration of a piping
system below 400 °C (Fig.5-16).
Japan has many nuclear power plants and petrochemicalcomplexes that were constructed during periods of strong
economic growth. These facilities are now facing aging
problems. We experienced a massive earthquake and hugetsunami on March 11, 2011. Now we should reconsider the
possibility that more catastrophic massive earthquakes might
hit Japan again in the near future. To maintain the piping
systems in these aging facilities by sensing technology is oneof the minimum obligations of the facilities managers.

Reference
Shimada, Y., Nishimura, A., Development of Optical Fiber Bragg Grating Sensors for Structural Health Monitoring, Journal of Laser Micro/Nanoengineering,vol.8, no.1, 2013, p.110-114.

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How Does a Cocatalyst Enhance the Activity of the Oxygen Reduction Reaction?
－ Improvement of Fuel Cell Performance －

(b)
1.2

Degree of Pt oxidation (a.u.)

Degree of Pt oxidation (a.u.)(a)Reduction of Pt oxide

0.9
0.6

Oxidation of Pt

0.3
0
-0.3

0

0.3

0.6

0.9

1.2

1.5

Electrode potential (V)

7.5
1.2

Ce3+
5

0.9
0.6

Pt
2.5

0.3
0
-0.3

Ce4+
0

0.3

0.6

0.9

1.2

0
1.5

Amount of Ce3+ and Ce4+ (a.u.)

5-8Electrode potential (V)Fig.5-17 Oxidation status of Pt and Ce
Oxidation status of (a) Pt in the conventional Pt catalyst and (b) Pt and Ce in Pt–CeOx nanocomposite catalyst. In the Pt–CeOx
nanocomposite catalyst, oxidation of the Pt surface is greatly suppressed. Blue and red data points correspond to positive- andnegative-going scans, respectively.

Fig.5-18 Schematic illustration of oxygen reduction
reaction at the surface of the Pt–CeOx nanocomposite
It was confirmed that oxidation of the Pt surface issuppressed by the CeOx/Pt interface, and the intrinsic highcatalytic activity remains enhanced.

The polymer electrolyte fuel cell (PEFC) is a clean andcompact power generation system that can perform below
100 °C, can be miniaturized easily, and emits only water as
exhaust. These features enable integration of the PEFC into
automobiles and mobile electronic devices. On the other hand,
the PEFC demands a large amount of an expensive precious
metal (platinum, Pt), which limits its wide adoption. In thelast decades, many studies for reducing the amount of Pt in
PEFCs have been carried out. For example, high-performanceelectrode materials with efficiencies close to the theoreticalvalue have been explored.
The Global Research Center for Environment andEnergy based on Nanomaterials Science (GREEN) at the
National Institute for Materials Science (NIMS) investigated
the nanocomposite catalyst Pt–CeO x , which exhibits
higher oxygen reduction reaction (ORR) activity than theconventional Pt catalyst, and established its synthetic method.
A joint research group comprising researchers from GREEN

and JAEA investigated the role of cerium oxide in the Pt–
CeOx catalyst and elucidated the mechanism by which the
ORR activity on the catalyst is enhanced.
To elucidate the valence of platinum and/or cerium atoms
in the Pt–CeOx catalyst and the conventional Pt catalyst under
ORR, in situ X-ray absorption fine structure measurements
were carried out at SPring-8. The results revealed that on theconventional Pt catalyst, a surface oxide began to form at thepotential at which ORR starts to occur. However, on the Pt–
CeOx catalyst, Pt oxide formation was suppressed, but Ce3+
was oxidized to Ce4+ at the potential at which Pt is oxidized(Fig.5-17). These results suggest that on the Pt–CeOx catalyst,
charge transfer at the Pt/Ce interface occurs, and Ce3+ in the
Ce oxide is oxidized instead of Pt (Fig.5-18). Previous studies
reported that the Pt surface oxide reduces the ORR activity.
Finally, it is concluded that the Pt–CeOx catalyst exhibits
intrinsic high ORR activity due to the suppression of Pt oxide
formation.

Reference
Masuda, T., Tamura, K. et al., Role of Cerium Oxide in the Enhancement of Activity for the Oxygen Reduction Reaction at Pt-CeOx Nanocomposite
Electrocatalyst - An In Situ Electrochemical X-ray Absorption Fine Structure Study, The Journal of Physical Chemistry C, vol.116, issue 18, 2012, p.10098-10102.
JAEA R&D Review 201369

Quantum Beam Science Research

5-9

Self-Assembled Structure of Coordination Species Clarified by Quantum Beam Techniques－ Development of Reprocessing Techniques by Elucidation of Solution States －Absolute intensity (cm-1)(a) X-ray scattering (TBP/Octane-h18)
1

Ellipsoidal TBP Assembly
PeakⅠ

Coordination spacefor heavy metal ions

φTBP =1.0

0.1(Volume fraction of TBP)

0.50

P

O
0.70

0.15

0.85

O

0.05
0

0.01

1

q(nm-1)OO
Inner core

O
P

O

O

10

1

Difference inscattering lengthdensity

0.40

ρinner core

Inner core

ρouter shell~ρoctane -h18

0.65
0.10

0.1

0.90

0.01

0.001

(b) Neutron
scattering
φTBP =1.0

Difference inscattering lengthdensity(Volume fractionof TBP)1q(nm-1)

10

TBP molecule(ρ:Scattering lengthdensity)(a) X-rayscattering(b) Neutron scattering (TBP/Octane-d18)
Absolute intensity (cm-1)

O

Outer shellρoctane -d18
Overall structure of TBP assemblyρinner core~ρouter shell
Length scale

Fig.5-19 X-ray and neutron scattering profiles obtained
for TBP/octane mixtures
Scattering maxima (peak Ⅰ) arising from the intermolecularinteractions of two TBPs were observed in (a), whereas in(b) the scattering contribution from the ellipsoidal orderedstructure was observed at lower q.

Fig.5-20 Diagram for scattering contrast and partial
structure of ellipsoidal TBP assembly in X-ray and neutron
scattering experiments
Schematic of TBP assembly, consisting of a hydrophilic innercore (light blue) and a hydrophobic outer shell (yellow), in
octane, and schematic of the spatial distribution of scattering
contrast (a) for X-rays across a TBP assembly in octane-h18
and (b) for neutrons across a TBP assembly in octane-d18.

Tri-n-butyl phosphate (TBP) is an important extractant
for separating hexavalent uranium, U(VI), and tetravalent
plutonium, Pu(IV), from used nuclear fuel by solvent
extraction. However, little was known on the structure
and characteristic features of the TBP assembly and its
coordination species with the heavy metal ions; thus, further
study is needed to identify them. We aim to address these
problems using scattering techniques; therefore, we collected
X-ray and neutron scattering data for the TBP/octane mixtures
as the first step of this study. As a result, we clarified theordered structure formed by TBPs in octane.
In Fig.5-19(a), scattering maxima (peak I) due to theintermolecular interactions of two TBPs were clearlyobserved. It is evident that TBP forms a self-assembly inoctane. In Fig.5-19(b), on the other hand, scattering from theoverall structure of the assembly was observed at q < 6.0 nm-1.
This indicates that TBPs form an ellipsoidal ordered structure,
as illustrated in Fig.5-20, where the inside core is capable ofincorporating a heavy metal ion.
The diagrams in Fig.5-20 indicate that (a) with X-rays thescattering length densities of the hydrophobic butyl groups

in TBP (ρbutyl) and octane-h18 (ρoctane-h18) are roughly equal toeach other, but that of the hydrophilic phosphate ester of TBP(ρhydrophilic) is larger than those of both hydrocarbon chains;and (b) with neutrons, the difference in the scattering lengthdensity between the hydrophobic and hydrophilic parts of
TBP (ρbutyl and ρhydrophilic) is negligible as compared with thescattering length density of octane-d18 (ρoctane-d18). Note that
X-rays and neutrons, respectively, enable the scattering to beeffectively observed from the inner core of the TBP assemblyand from its overall structure, respectively. In fact, the qdependence of the obtained scattering profiles differs greatly
between X-rays and neutrons, elucidating the structural andcharacteristic features of TBP assemblies.
The quantum beam technique is a powerful tool forinvestigating the microscopic structures of the various organic
molecules and their coordination species with heavy metal
ions. Anticipated applications include the development of not
only fuel reprocessing but also diverse research topics such as
the recycling of valuable metals from the waste products or
the polluted solutions.

Reference
Motokawa, R. et al., Microscopic Structures of Tri-n-Butyl Phosphate/n-Octane Mixtures by X-ray and Neutron Scattering in a Wide q Range, The Journalof Physical Chemistry B, vol.116, issue 4, 2012, p.1319-1327.

70

JAEA R&D Review 2013

Quantum Beam Science Research

5-10 Elucidation of Electronic Excitation in Novel High-Tc Superconductor－ Resonant Inelastic X-ray Scattering Study of Iron-Pnictide Superconductor －Synchrotron(a) Calculation
Sample0.6
750 μmIncident X-ray
Momentum : p
Energy:Eq=(0, 0)− −(pʼ, Eʼ)−−RIXS Intensity (a.u.)−(p, E)−

0.6
q=(0, 0)0.5

0.5
Emitted X-ray
Momentum : pʼEnergy: Eʼ

Electron(b) Experimentq=(π/2, 0)

0.4

0.4
q=(π, 0)0.3

0.3
q=(π/2,π/2)

0.2

0.2
q=(π,π)−q=(π, 0)q=(π,π)

0.1

0.1

−
0
Electron momentum change : pーpʼElectron energy change: EーEʼ

0

0 1 2 3 4 5 0 1 2 3 4 5Energy Loss (eV)Fig.5-21 Schematic representation of RIXS
Systematic inspection of the momentum change and energy
loss of X-rays in scattering enables us to study the momentumdependent electronic excitation spectra.

Fig.5-22 Calculated and experimental spectra
Theoretical calculation (left) was performed for non-magnetic
and magnetic ground states (thin and thick lines, respectively).
The experimental spectrum (right) at q = (π, π) (dashed line) isoverlaid with the spectra at other momenta.

Quantum beams such as neutron beams and X-rays havebeen extensively used to clarify elementary excitations invarious strongly correlated electron systems. A lot of studies
using quantum beams have been intensively carried out for the
recently discovered iron-pnictide high-Tc superconductors, toelucidate their superconducting mechanism.
Recently, resonant inelastic X-ray scattering (RIXS)
utilizing the hard X-rays at transition-metal K-edges has
become a unique technique to observe momentum-dependent
charge excitation spectra of strongly correlated electrons. In
K-edge RIXS, the incident X-ray with the 1s-4p absorptionenergy resonantly excites the 1s electron to the 4p state. In
the final state, the excited 4p electron returns to the 1s state,emitting an X-ray, where the emitted X-ray loses momentum
and energy in amounts equal to those transferred to thecorrelated electrons (Fig.5-21).
We recently succeeded in the first observation of RIXS fora typical iron-pnictide high-Tc superconductor PrFeAsO1-y.
The experimentally observed and theoretically calculated

RIXS spectra for various X-ray momentum changes q’s are
displayed in Fig.5-22. We found a characteristic momentumdependent low-energy feature around 1-2 eV, as indicated by
the ticks in the right panel of Fig.5-22. To analyze the spectral
properties, we also performed a theoretical calculation (leftpanel of Fig.5-22). From comparison between the calculated
and experimental spectra, spectral shape and momentum
dependence obtained experimentally agree well with thecalculation by assuming the magnetically ordered state rather
than the non-magnetic state. This suggests that at least shortrange local magnetic correlations exist, consistently withother experiments. By further analysis, we also succeededin verifying that the local Coulomb repulsion among the
Fe 3d electrons should be intermediately strong (2-3 eV) in
the present iron-pnictide superconductor. This evaluationof electron correlation could play a key role in discussionsof the microscopic mechanisms of the iron-pnictide high-Tc
superconductivity.

Reference
Jarrige, I., Nomura, T. et al., Resonant Inelastic X-ray Scattering Study of Charge Excitations in Superconducting and Nonsuperconducting PrFeAsO1-y,
Physical Review B, vol.86, issue 11, 2012, p.115104-1-115104-4.
JAEA R&D Review 2013

71

6

Nuclear Safety Research

Evaluation of the Safety of Various Nuclear Facilities

Fuel Manufacturing Plant

Fuel
Cycle

NSRR

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

RFEF

Severe東電 Accident
第(SA) at 1F

Reprocessing
Plant

Waste Disposal Facility
・ Safety evaluation ofradioactive waste
disposal

Decommissioning Lesson
learned(Accident survey)
Environmental
reclamation

SA Prevention
Research
Design-basis event
Prevention of core melt
due to beyond design
based accident

Prioritizing of Research

Nuclear Safety Research Center Efforts
Initiating Events

Progress of SA

Failure of components
due to outer eventsContainment vessel (CA)Pressure
Hydrogenvesselgeneration

3D
response
analyses

Thermalhydraulics

Cooling
down of CA

THYNC

LSTF

JMTR

JMTR-HL

Environmental Impact
シビアアクシデント
環境影響評価・
SA Evaluation/ Exposure Assessment
ResearchResearch
Core melt
Public exposure
Vessel failure
Nuclear disaster
Release of
prevention
radioisotopes (RIs)

NUCEF
Earthquake motion

RI emission
from molten
fuel

RI migration

Environmental Impact due to RI Emission
Atmospheric dispersion

Dry deposition

Exposure due toradioactive cloudWet deposition

Inhalation ofradionuclides
External exposure
Internal exposure

Ingestion of contaminated food, etc.

Exposure due todeposited radionuclides

Fig.6-1 Nuclear safety research areas and related JAEA
facilities
The NSRC experimentally obtains data using various Japan
Atomic Energy Agency (JAEA) facilities. These data forrisk assessment and safety evaluation of nuclear facilities
contribute internationally to the development of evaluation
methods and criteria.

Fig.6-2 Future directions of nuclear safety research
In accordance with lessons learned from the accident at 1F,
in addition to safety assessments within design basis events,
we will conduct research with a priority on severe accident
prevention efforts and nuclear emergency preparedness
activities such as environmental impact assessments.

The possibility of serious accidents in nuclear installations
and their consequences have been studied at the Nuclear
Safety Research Center (NSRC) to ensure the safe use ofcurrent nuclear facilities. A massive and severe accident
occurred at the Tokyo Electric Power Company, Incorporated
Fukushima Daiichi Nuclear Power Station (1F) as a result of
the Great East Japan Earthquake and tsunami. The NSRC has
been collaborating with the Japanese government to instituteappropriate emergency response procedures based on researchfindings and developments. Most specialists at the NSRC
regret that they were unable to prevent the accident.
Continuous efforts should be made to avoid serious damage
while we use nuclear energy. A proper understanding of
the situation and continuous improvement of developed
technologies are important for making the best and safest
use of the technologies. The International Atomic Energy
Agency, which investigated the accident at 1F, pointed out
the importance of “continuous improvement.” To contributeto “continuous improvement aimed at the highest level ofsafety,” we are committed to safety research using the various
facilities shown in Fig.6-1. Radionuclides are handled andaccident conditions are safely simulated at these special
facilities.

In this chapter, the results of recent safety research on thefollowing topics are presented: prediction of the amountof fission product (FP) gas stored in high-burnup fuel(Topic 6-1), a structural integrity evaluation method for piping(Topic 6-2), a method of evaluating isotope elements in the
spent fuel (Topic 6-3), and thermal aging embrittlement of caststainless steel (Topic 6-4).
To reduce the risk to nuclear facilities and enhance thepreparedness for emergency conditions that are postulated tolead to a severe accident, as shown in Fig.6-2, we are planningto conduct nuclear safety research with a priority on theprevention and evaluation of severe accidents as well as on
disaster prevention measures such as environmental impact
assessment.
The following topics related to the accident at 1F by the
NSRC are described in Chapter 1:(1) Assessment of Doses to the Inhabitants Living inFukushima Prefecture (Chapter 1, Topic 1-10)(2) Dose Estimate for Workers and Public with Incineration
and Disposal of Dehydrated Sludge (Chapter 1, Topic 1-11)(3) Review of Five Investigation Reports on Fukushima
Accident (Chapter 1, Topic 1-18)

72

JAEA R&D Review 2013

Nuclear Safety Research

6-1

Evaluating the Amount of Fission Gas in Fuel with Higher Accuracy

GB bubble radius (μm)

0.6
0.5
0.4

Pellet central region

0.3
0.2

Pellet peripheral region

0.1
0.0

0

10

20

30

40

50

60

Rod average burnup (GWd/tU)

Relative density of fission gas atoms－ Development of a Rate-Law Model of Fission Gas Bubble Growth －

2.0
EPMA
data

1.5

1.0

Fission gas atoms
retained in crystal
grain

Gas
releasedregion

0.5

0.0
0.0

Total amount of generated
fission gas atoms

Amount of fission gas
atoms accumulated
in gas bubbles
0.2

0.4

0.6

0.8

1.0

Relative radius of pellet

Fig.6-3 Fission gas bubble growth in central and peripheral
regions of fuel pellet
The growth of fission gas bubbles in a fuel pellet during
irradiation is shown. Growth is faster in higher-temperature
central region of pellet.

Fig.6-4 Distributions of densities of fission gas in the radial
direction of fuel pellet
The red and black curves indicate calculated values. The
hatched area enclosed by the red and black curves corresponds
to the amount of gas accumulated in gas bubbles, while somefraction is released to the outside of the pellet. The red curveand blue dots (relative values from EPMA measurement)
indicate gas atom densities retained in crystal structure of pellet.

In high burnup light water reactor fuel pellets, a large
number of fission gas atoms is retained through an extended
period of irradiation. The gas is accumulated in gas bubbles
in the fuel crystal structure and could swell the pellet, or
gas released from the bubbles could increase the internalpressure of the fuel rod and expand the cladding outward,
which has undesirable effects on the integrity of the fuel rod.
Accordingly, to accurately evaluate the growth of gas bubbles
in pellets, a new model was developed that allows better
prediction of gas bubble growth. The model was installed in
the fuel performance code FEMAXI-7.
The gas bubble grows as the excess bubble internal pressure
exceeds the external resistance from the surrounding structure.
However, because an earlier model assumes that the bubble
gas pressure is always in balance with the external resistance,
reliable evaluation has been difficult. In this study, the bubble
growth was modeled by a process in which the gas bubble
pressure is enhanced and extends the bubble outward into thesurrounding crystallite structure.
Next, a calculation was performed to predict the bubble
growth and accumulated amount of fission gas. When theradius reaches 0.5 µm, gas release occurs, and growth ceases.

Fig.6-3 shows the bubble growth (bubble radius) with burnup.
Regarding the final accumulation of gas atoms after this
irradiation process, Fig.6-4 compares the calculated andmeasured profiles of the fission gas atoms density. The latter
were obtained by an electron probe microanalyzer (EPMA).
The red curve and blue dots indicate the density of gas
atoms retained in the crystallites of the fuel pellet. The red
curve represents the distribution trend of the blue dots fairly
well. The hatched area between the red and black curves
corresponds to the accumulated amount of gas, which is
calculated to be about 18% of the total generated.
In an experiment on a simulated reactivity-initiated accident
conducted in the Nuclear Safety Research Reactor using thisfuel, the fission gas release rate was estimated to be about
22% of the total generated, which proves the satisfactory
predictability of the model. The result of another experimentindicated that the gas is released mainly from grain boundarygas bubbles. The proximity of the calculated and measured
values supports this result.
Thus, this new model has enabled us to more accurately
predict fission gas bubble growth behavior as one of thefactors related to the integrity of fuel rod.

Reference
Suzuki, M. et al., Model Development and Verifications for Fission Gas Inventory and Release from High Burnup PWR Fuel during Simulated ReactivityInitiated Accident Experiment at NSRR, Proceedings of Topical Conference on Water Reactor Fuel Performance (TopFuel 2012), Manchester, U.K., 2012, 6p.,
in USB Flash Drive.
JAEA R&D Review 201373

Nuclear Safety Research

6-2

Predicting Crack Growth in Reactor Piping under Large Seismic Loading－ Proposed Method of Evaluating Crack Growth under Random Cyclic Loading －100
Load(kN)

Excessive load
Excessive load

0

StressStressStressStress

Crack
Crack
Distance fromDistance
crack
tip from
crack tip

Increase in stressin stress
inIncrease
front of cracktip
in front of crack tip

Crack
Crack
Crack blunting
Crack blunting

Distance fromDistance
crack
tip from
crack tip

Before excessive loading
After excessive loading
Before excessive loading
After excessive loading
Crack shape and stress distribution near crack tip

Amount of crack growth (mm)-1001Measured result
Estimation result by the proposed method
Estimation result by the existing method

0.8
0.6
0.4
0.2
0

0

32

64

96
128 160 192
Number of cycles

224

256

Fig.6-5 Stress distribution shift and crack blunting at crack
tip under large seismic loading
Retardation and acceleration of the crack growth rate underlarge seismic loading are caused by crack blunting due toexcessive tensile loading and a stress distribution shift in frontof the crack tip due to excessive loading.

Fig.6-6 Amount of crack growth under simulated seismicload
The method of estimating the crack growth in reactor piping
subjected to seismic loading beyond the SSY condition was
validated by a crack growth test of pipes 114.3 mm in outer
diameter and 11.1 mm in thickness. The measured amountof crack growth is compared with that estimated by the
proposed and existing methods.

Accurately predicting crack growth is important forevaluating plant safety. The existing method of evaluatingcrack growth under constant-amplitude cyclic loading was
based on the small-scale yielding (SSY) condition. The
objective of this study is to propose a crack growth evaluation
method in pipes beyond the SSY condition under seismic
loading.
We demonstrated that the crack growth rate beyond the SSY
condition and during cyclic loading can be evaluated using
the elastic-plastic fracture mechanics parameter (J-integral).
Crack growth experiments using cyclic loads with excessive
tensile and compressive loads showed the acceleration andretardation effects of the excessive load on crack growth. This
indicates that deformation at the crack tip and the change inthe stress distribution in front of the crack tip play a key partin crack growth under large seismic loading. Finite element
method analyses revealed that the effect of acceleration andretardation on crack growth is attributed to an increase in the
stress in front of the crack tip from the excessive compressive

load and crack blunting from the excessive tensile load,
respectively, as schematically shown in Fig.6-5. A crack
growth evaluation method based on these features has been
proposed in which ΔJ is calculated in each cycle to estimate
the amount of crack growth.
The method of evaluating the crack growth in reactor piping
subjected to seismic loading beyond the SSY condition has
been validated by piping crack growth tests using a simulated
seismic load. Fig.6-6 shows the obtained amount of crack
growth ( ■) together with that evaluated by the present andexisting methods. The proposed method can accurately
estimate the amount of crack growth given by the test result
because the method considers the condition beyond SSY
and the crack growth under individual loading of irregular
amplitude, unlike the existing method. The validity of the
proposed method was confirmed for different materials andloading patterns.
The present study was sponsored by the Japan Nuclear
Energy Safety Organization (JNES).

Reference
Yamaguchi, Y. et al., The Proposal of Crack Growth Evaluation Method for Piping under Seismic Loading, Nippon Kikai Gakkai Ronbunshu, A Hen, vol.79,
no.802, 2013, p.730-734 (in Japanese).

74

JAEA R&D Review 2013

Nuclear Safety Research

6-3

Measuring Amount of Nuclides in Spent Nuclear Fuel－ Development of Method of Assaying Barely Measurable Elements in Spent Nuclear Fuel －generated per initial uranium (one-ton)

Amount of nuclides (kg),

0.6
150

Sm

0.5
0.4
0.3
0.2

present study
Measurement for BWR fuelin Foreign Country
Measurement for BWR fuelin Japan

0.1
0.0

0

20

40

60

80

Burnup (GWd/t)
Fig.6-7  Induced coupled plasma mass spectrometry(ICP-MS)
High-sensitivity, high-resolution measurement of isotopic ratio
measurement for REEs.

Fig.6-8 Comparison of measurement between past andpresent studies for 150Sm as an example
Good agreement is shown in the low-burnup region and thelinearity of the amount against burnup.

The amount of nuclides in spent nuclear fuel (SNF) is a
type of basic data used to evaluate the characteristics of SNF
and radioactive waste. To establish a method of measuring it,
we have been conducting experiments to assay the amountof fission products (FPs) in a boiling water reactor 9 × 9 fuelassembly during four years beginning in FY 2008. For handlingthe radioactive materials and because of the requirement forknowledge of the chemical separation of several elements, thisexperiment was conducted under the collaboration of several
departments in the Japan Atomic Energy Agency.
Similar experiments have usually adopted a γ-ray
measurement technique, but it is not applicable to stable
isotopes. In some cases, after the ion exchange separation ofelements, thermal ionization mass spectrometry (TIMS) wasused. It has high accuracy, but we should be careful of theradiation dose control because it requires a relatively large
quantity of the dissolution sample. Furthermore, for rare earth
elements (REEs), the chemical separation and measurement
procedure become too complicated because there are several
isotopes that have the same mass number. Thus, it has beendifficult to measure many samples or to measure repeatedly to
reduce the measurement error.
To establish a method of assaying FP elements in SNF simply
and quickly using a smaller amount of the dissolution sampleof SNF, we have been studying an efficient FP measurement
technique combining REE separation by ion chromatography
and high sensitivity inductively coupled plasma mass
spectrometry (ICP-MS) (Fig.6-7). In this study, we establisheda procedure for measuring FP nuclides by isotope dilution mass
spectrometry with ICP-MS, measuring the isotopic ratio after
chemical separation of REE.
This method enables us to separate neodymium, samarium,and gadolinium; the present results agree well with pastexperimental data obtained with TIMS (Fig.6-8). Using thismethod, we can obtain the required data by using a verysmall amount (less than 10 µg) of the resolved sample, so
the problem of radiation dose control is reduced. In addition,
it is a very efficient method that requires less than five days
from the chemical separation to the measurement. This opens
the possibility of obtaining reliable data by using repeatedmeasurements and a larger number of samples. We plan tomeasure FPs other than REEs in a future program.
This work was conducted under a research contract with
the Japan Nuclear Energy Safety Organization (JNES),“Measurement experiment of nuclide composition of fission
products in spent fuel burnt in light water reactors, F.Y. 2008 to
F.Y. 2011.”

Reference
Fukaya, H., Suyama, K. et al., Examination of Measurement Method of Isotopic Composition of Fission Products in Spent Fuel, JAEA-Research 2013-020,
2013, 81p. (in Japanese).
JAEA R&D Review 201375

Nuclear Safety Research

6-4

Research on Thermal Aging Embrittlement of Cast Stainless Steels－ Study of Thermal Aging Embrittlement in “FUGEN” Nuclear Reactor －Injection valve body

25

15
10 nm

（b）
Charpy impact energy (J/cm²)

Steel Grade:
SCS13
Operating temp:
275 ℃
Period of service: 138000 h

Injection
valve body

As collected

Recovery heated

As collected

Recirculation pump casing

35

Cr concentraion (at%)

Steel Grade:
SCS13
<40 ℃
Operating temp:
Period of service: 138000 h

Recirculation
pump casing
Recovery heated

（a）

180
160
140
120
100
80
60
40
20
0

Recirculation
pump casing

Injection
valve bodyFig.6-9 Sampling point schematics and service conditionof test specimens collected from “FUGEN” reactor
Specimens of cast stainless steels were collected fromrecirculation pump casing and injection valve body withdifferent service temperatures in “FUGEN” reactor.

Fig.6-10 Results of analyses of samples collected from
“FUGEN” reactor(a) Cr concentration distribution map obtained using
3DAP. (b) Result of Charpy impact test. Phase separationand thermal aging embrittlement have progressed inrecirculation pump casing.

Cast stainless steels are extensively used in light water
reactors to fabricate valve bodies, pump casings, and primarycoolant piping. Previous studies have shown that thermal
aging around 300 °C causes embrittlement. Predictions of thedegree of loss in toughness suffered during extended reactor
service life would be valuable for assessing the reliabilityof cast stainless steel reactor components. Therefore, somemodels for predicting changes in the mechanical properties
because of thermal aging have been developed. However, tomaintain the toughness of the components in service within areasonable time period, the aging process in the test materialmust be accelerated at high aging temperatures. Thus, it
is very important to examine the thermal embrittlement
mechanism using data obtained from materials that have aged
on site for a long period.
A program is currently underway for a decommissioning
plan that considers the progress and dismantling of the“FUGEN” reactor. The reactor contains valuable structural
materials that have been used for 25 years. In this study, thedegree of toughness and the microstructural characterizationof cast stainless steel components dismantled from the“FUGEN” reactor were investigated to determine the long-term
thermal aging effect at low temperatures (Fig.6-9). We studied
the microstructural changes due to thermal aging using thethree-dimensional atom probe (3DAP) technique, which issuitable for nanometer-scale analysis of the materials. The
analysis of the sample material, which was collected from the“FUGEN” reactor, under thermal aging at 275 °C shows thatthe initial homogeneous distribution of the Cr concentration
becomes modulated as the structure progresses into spinodal
decomposition (Fig.6-10(a)). This microstructural change isprobably related to hardening, and it causes the thermal aging
embrittlement. The results of Charpy impact tests ensure thereliability of the injection valve body. On the other hand,
the results from the recirculation pump casing show earlysigns of an ongoing thermal aging effect (Fig.6-10(b)). We
are planning to investigate the underlying mechanism of theembrittlement process to validate and improve the accuracyof the prediction model by using the valuable material dataobtained from the “FUGEN” reactor.
The present study was sponsored by the Japan Nuclear
Energy Safety Organization (JNES).

Reference
Nogiwa, K. et al., Influence of Thermal Aging on Cast Stainless Steels Used in JAEA’s Nuclear Reactor Fugen, Journal of Nuclear Science and Technology,
vol.50, no.9, 2013, p.883-890.

76

JAEA R&D Review 2013

7

Advanced Science Research

For the Evolution of Nuclear Science
Exploration of scientific

New Frontier

Advanced Science
Research

Frontier research on
heavy element systems

Advanced materials
research
Condensed matter theory,
simulation, spintronics

Nuclear Science andEngineering Directorate

Thermospintronics

Simulation

Superheavy elements and nuclei
actinides
NeutronAccelerators
Radiation fields

Radiation field effects
Hadron physics
biology and radiation
spin-polarized positron beam
Quantum Beam
Science Directorate

REIMEI program

External and foreignorganizations

J-PARC

Fig.7-1 Exploration of scientific new frontier in various collaborations

Because we are the sole institute for atomic energy researchin Japan, both basic and goal-oriented research are within thescope of our mission. Nuclear engineering is supported bymany fields of basic science and technology, and innovations
in each research field would contribute to the evolution of
nuclear engineering.
The purpose of the Advanced Science Research Center(ASRC) is to promote cutting-edge research on selected topics
in basic sciences, which would enable the exploration of new
frontiers in nuclear science.
The ASRC Visions for the midterm starting April 2010 are
to promote the most advanced basic research, to establish aninternational center of excellence, and to create new researchand technologies. The research projects being pursued canbe categorized into three fields: basic research on advanced
materials, frontier research on heavy element systems, andbasic research on radiation fields.
The research topics in the first research field include thedevelopment of molecular spin transport materials andcondensed matter theory. Frontier research on nuclear physics,
the characterization of superheavy elements, and actinide
compounds constitutes the second research field. The third
field spans from hadron physics to biology. Collaboration
among different research groups in the ASRC and with otherdirectorates is essential to explore the frontiers depicted in
Fig.7-1. The research highlights for 2012 are power generation
by applied spintronics (Topic 7-1); measurement of theionization energy of Lr, the 103rd element (Topic 7-2), andthe phase transition of URu2Si2 (Topic 7-3); a new mechanismof DNA damage (Topic 7-4); and the development of a spinpolarized positron beam (Topic 7-5). Our research activities also
include a Fukushima-related activity (Chapter 1, Topic 1-6).
ASRC members are also involved in outreach activities
regarding nuclear science. Fig.7-2 shows a 3D chart ofnuclides (the horizontal axes are the numbers of neutrons andprotons, and the heights represent the half-lives). It facilitates
understanding of the synthesis processes of elements,
including the r-process, which produces nuclear fuel elements
in supernovae.

Fig.7-2 3D chart of nuclides
JAEA R&D Review 201377

Advanced Science Research

7-1

Electricity Generation by Nanomagnet Dynamics－ Magnetic Power Inverter Based on Spinmotive Force －w1 w2

Domain wall

10.0
Nanowire
Voltmeter(b)

w1Wall energy

w2

Output voltage (μV)(a)
Magnetic field
Period d

9.0

Magnetic
field (100 mT)

8.0
7.0

Normal nanowire(DC voltage only)

6.0
5.0

Wall position

Magnetic power inverter(DC+AC voltage)

0

20

40

60

80

100

Time (ns)Fig.7-3 Schematics of magnetic power inverter(a) A magnetic field is applied to induce domain wall motion ina nanowire that is periodically patterned with wider (W1) andnarrower (W2) widths. (b) The wall energy is proportional to thewire width and becomes larger (smaller) at W1 (W2).

Fig.7-4 Output voltage of magnetic power inverter
In a normal nanowire, a DC magnetic field induces a DC output
voltage ( ― line), whereas, in a periodically patterned nanowire,
we have theoretically shown that an AC output voltage issuperposed in addition to the DC component ( ― line).

In 1831, M. Faraday, a British scientist, found that atime-varying magnetic field induces an electromotive force(EMF) in an external circuit. Faraday’s law of induction is afundamental law in electromagnetism and plays a pivotal roleas an operating principle for various electrical devices ranging
from power generators to electrical appliances.
Recent advances in nanotechnology have made it possible
to fabricate and handle magnets ultimately as tiny as one
likes. During this progress, we discovered a new mechanismthat directly converts magnetic energy to electrical energy
in a completely different manner from the conventional
inductive EMF. The mechanism is called the spinmotive
force (SMF), so called because it relies on the electron’s spin,a source of the electron’s magnetic properties. Contrary toconventional wisdom, the SMF can induce electricity from astatic (direct current, DC) magnetic field and is now regarded
as a key building block in spintronics, a research field in nextgeneration green technology.
An SMF can be induced using a magnetic domain wall
(DW), i.e., an interface between different magnetic domains,in a magnetic nanowire (a tiny magnetic strap). When an
external magnetic field is applied to the nanowire, a DW isdisplaced along the nanowire through an energy relaxation

process. The released magnetic energy is transferred to the
conduction electrons via the magnetic exchange interaction,
producing an SMF.
In this work, by further controlling the DW motion, we
propose the magnetic power inverter, a new conversion schemefrom a DC magnetic field to an alternating current (AC)voltage. Temporal oscillation of the magnetic energy is realizedby patterning a magnetic nanowire with periodically varying
widths, as shown in Fig.7-3(a). In such a patterned wire, a DW
behaves like an elastic membrane, and the DW energy changes
with the DW position, as shown in Fig.7-3(b). In this case, theoutput voltage has an AC component reflecting the alternating
DW energy in addition to a normal DC component due to the
input DC magnetic field (Fig.7-4). The characteristics of the ACcomponent, such as the amplitude (several μV) and frequency(MHz to GHz), can be tuned by the design of the wire shape,
choice of materials, and magnitude of the applied magnetic
fields.
This work introduces a novel concept that directly bridges
different types of energies with frequency modulation andopens a new route for highly efficient energy conversion
technology in nanodevices.

Reference
Ieda, J. et al., Magnetic Power Inverter: AC Voltage Generation from DC Magnetic Fields, Applied Physics Letters, vol.101, issue 25, 2012,
p.252413-1-252413-4.

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Advanced Science Research

7-2

Toward Experimental Determination of the First Ionization Potential of the Heaviest Actinide－ New Ionization Potential Measurement Using Surface Ionization Process －Filaments
Nuclear reaction
products

Counts (/10 keV)

(a)

10
8

252

Fm

256

Lr+256 No

6
4
2
0

7000

8000

9000

10000

Energy (keV)

Ion beam(b)
256

Ionizer

Lr T1/2 =27 s

α
25 2

Md

T1/2 =2.3 min

EC
25 2

α

EC
256

α

No
T1/2 =3.3 s

Fm
T 1/2 =25.4 hFig.7-5 Newly developed ion-source coupled to a He/CdI2
gas-jet transport system for JAEA-ISOL
Nuclear reaction products produced in the nuclear reaction
are quickly transported to an ionizer and ionized.

Fig.7-6 (a) Measured -particle spectrum of massseparated ions with mass number A = 256. (b)Decay scheme of 256Lr (T1/2: half-life, : -decay,
EC: electron capture)

The first ionization potential (IP) is a fundamental
quantity in physical and chemical research on an element.
Information on the IP of the heaviest elements can provide
a better understanding of relativistic effects in the electronic
configuration, which are significantly noticeable for heavy
elements. The heaviest elements, which have atomic numbers
greater than 100, can be produced at accelerators using
reactions of heavy ions with heavy target materials. Owing to
the short half-lives and low production rates of the isotopes,
they are usually available in quantities of a few atoms at atime. It has, therefore, been impossible to measure the IP
values of the heaviest elements so far.
For this purpose, we have been developing a new method tomeasure the IP value of the heaviest elements based on a surface
ionization process. In surface ionization, an atom is ionized
via an interaction with a solid surface at a high temperature.
The ionization efficiency depends on the temperature and work
function of the surface, and the IP of the atom.

To ionize the heaviest actinide element, lawrencium (Lr, Z = 103),
via surface ionization, we developed a surface-ionization typeion-source coupled to a gas-jet transport system installed in
the Isotope Separator On-Line (ISOL) system of the Japan
Atomic Energy Agency (JAEA) (Fig.7-5).
We successfully ionized and mass-separated a short-lived
Lr isotope, 256Lr, produced in the 249Cf(11B, 4n) reaction for the
first time by applying the present system. As shown in Fig.7-6,α-particles originating from 256Lr and the daughter and the
granddaughter nuclides, 256No and 252Fm, respectively, wereclearly observed. It is theoretically predicted that Lr would
have a lower IP than other actinide elements and lutetium (Lu),the lanthanide homolog of Lr. The ionization efficiency of Lr
was obviously higher than that of Lu. This result indicates that
the IP of Lr would be lower than that of Lu. We conclude thatthe surface ion-source is a promising apparatus for measuring
the IP of Lr. The IP of Lr is being determined using the present
system.

Reference
Sato, T. K. et al., First Successful Ionization of Lr (Z=103) by a Surface-Ionization Technique, Review of Scientific Instruments, vol.84, issue 2, 2013,
p.023304-1-023304-5.
JAEA R&D Review 201379

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7-3

Manipulation of Electrons by Pressure

Hidden order transition temperature (K)－ Coupling between Lattice and Electrons in Superconductor URu2Si2 －Paramagnetic state(4-fold)

20
17.5

0

Hidden order state(2-fold)5
Uniaxial pressure (GPa)Fig.7-7 Uniaxial pressure cell
Uniaxial pressure is applied vertically to the sample (back one),which breaks the four-fold symmetry.

Fig.7-8 Uniaxial pressure dependence of hidden ordertransition temperature (― line)
The transition temperature increases with increasing
uniaxial pressure.

In the superconducting state, two electrons form asuperconducting pair. An attractive interaction between theelectrons is necessary for the formation of paired electrons.
In conventional superconductors, the attractive interaction
originates in lattice vibration. In contrast, magnetic fluctuationis considered to be the origin in actinide superconducting
compounds. Because a high superconducting transition
temperature Tc can be expected for a magnetic-fluctuationinduced superconductor, it is important to clarify themechanism for the fluctuation.
In URu2Si2, the so-called hidden order state appears below
17.5 K (-256 °C). Because the superconducting state appears at
a lower temperature of 2 K (-271 °C), magnetic fluctuation in
the hidden order state is considered to induce superconductivity.
However, the hidden order was not identified for 25 years,
which is an important issue in condensed matter physics.
This compound has four-fold symmetry in the paramagnetic
state above 17.5 K. Recently, however, it was revealed thatthe electronic state exhibits symmetry breaking from the four-

fold to the two-fold state with hidden order. In this study,
the hidden order is investigated in a two-fold lattice state
artificially induced by uniaxial pressure. To apply uniaxialpressure at low temperatures, we developed a uniaxial pressure
cell (Fig.7-7). As shown in Fig.7-8, the hidden order transition
temperature increases under uniaxial pressure, indicating thatelectrons and the lattice are coupled, and we can manipulate
the magnetic fluctuation that induces superconductivity.
This technique is expected to yield an understanding ofmagnetic fluctuation and to facilitate the development of hightemperature superconducting materials.
A future new superconductor with a higher Tc is expected tobe realized in a magnetic-fluctuation-mediated superconducting
system, because the Tc value of phonon-mediated superconductors
has already reached its peak. If a new superconductor with a room
temperature Tc is found, it will be quite useful for improving theenergy efficiency.
We are going to investigate actinide compounds to clarify
the mechanism of superconductivity in detail.

Reference
Kambe, S. et al., Thermal Expansion under Uniaxial Pressure in URu2Si2, Physical Review B, vol.87, issue 11, 2013, p.115123-1-115123-6.

80

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Advanced Science Research

7-4

Novel Mechanism of Radiation Damage to DNA－ An Advance in Elucidating the Radiation Process in DNA Damage －

4
0.10
0.08

3

0.06

2

0.04
0.02

1

nitrogen K-edge

oxygen K-edge

0

0
400

450

500

550

600

650

700

Energy / eV

X-ray absorbance (arb. units)

EPR intensity (arb. units)

X-rays

5’

3’

DNA
N

N

3’

Two electrons are emitted bya normal ionization process.

5’

A photoelectron ejected from theK-shell can be captured by PCI andthen detected by EPR as an unpaired
electron species originating from a
specific excitation state.

Fig.7-9 Anomalous enhancement of EPR signal at nitrogenand oxygen K-edges
EPR signal intensity (― line, left axis) is anomalously enhanced(○) at energies slightly above the nitrogen and oxygen K-edges
when compared with the dependence of the X-ray absorption
probability of DNA (― line, right axis) on the X-ray energy(abscissa).

Fig.7-10 K-shell photoabsorption processes at specificatoms in DNA (nitrogen K-shell photoabsorption isshown as an example)
We revealed that when DNA is exposed to X-rays with anenergy slightly above the K-shell ionization threshold, anejected electron is re-captured by the DNA as an unpaired
electron.

Risk assessment for low-dose irradiation has been one of themost urgent issues. In particular, the research field of radiation
damage to DNA coding genetic information and its biologicalrepair has not yet been fully understood. In this study werevealed a novel mechanism of the radiation process in DNA
damage.
When a living cell is exposed to ionizing radiation,
particularly high-energy charged particles, energy over a wide
range is deposited from the particle to the DNA. Dependingon the deposited energy level, various DNA damage processes
might proceed simultaneously. To fully understand the damage
processes, each one should be individually analyzed. Previous
studies, however, lacked suitable techniques to realize thisgoal.
We addressed this difficulty using highly monochromatizedsoft X-rays. In an irradiated DNA molecule, an unpaired
electron remaining in the outermost shell after ionization is
one of the short-lived intermediates that cause damage. We
exposed DNA to soft X-rays obtained from a synchrotron
facility (SPring-8) while sweeping the energy. The precise
dependence of the yields of the unpaired electron species on
the soft X-ray energy was investigated, particularly around theK-shell ionization thresholds of the DNA constituent atoms

nitrogen and oxygen. Previously it was difficult to directly
observe the highly reactive and short-lived species. To realize“in situ” observation of the short-lived species, we connected
an electron paramagnetic resonance (EPR) apparatus to thesoft X-ray beamline.
The result shows that the EPR intensity (― line in Fig.7-9)
corresponding to the yield of the unpaired electron specieschanges roughly in proportion to the X-ray absorption
probability by K-shell electrons (― line in Fig.7-9). Further,
exposure to X-rays with an energy slightly above theionization thresholds causes, very interestingly, an anomalousenhancement of the EPR signal intensity (shown by circles in
Fig.7-9). A theoretical analysis revealed that the enhancement
is caused by recapturing of an ejected slow electron at theionized atom through a Post Collision Interaction (PCI)
process, which is a novel mechanism of radiation damage toDNA (Fig.7-10).
These results are expected to advance understanding of
DNA damage, particularly that originating from K-shellionization.
This work was supported by a Japan Society for the
Promotion of Science (JSPS) KAKENHI Grant-in-Aid for
Scientific Research (B) (No.21310041).

Reference
Oka, T., Yokoya, A. et al., Unpaired Electron Species in Thin Films of Calf-Thymus DNA Molecules Induced by Nitrogen and Oxygen K -Shell
Photoabsorption, Physical Review Letters, vol.109, issue 21, 2012, p.213001-1-213001-5.
JAEA R&D Review 201381

Advanced Science Research

7-5

Evaluation of Electron Spin Using Its Antiparticle－ Development of a Positron Beam with the World’s Highest Spin Polarization －Germanium-68 production

Proton beam

Positron spin:
Positrons with thesame spin direction(spin-polarized
positron beam)

Iron sample
magnetic field①Easily annihilation

GaN target
magnetic field
Electron spin:
Direction changedby the magnetic field

Cyclotron accelerator②Difficult annihilation

Difference in the intensity of䃒rays of ① and ②⇔ Electron spin changed by an external magnetic field

Fig.7-11 Evaluation of electron spins of iron sample using spin-polarized positron beam
A positron beam with a high spin polarization of 47% is irradiated on the iron sample. The spin direction of electrons in the ironsample is changed by an external magnetic field. As positrons and electrons are easily annihilate when their spin directions
are antiparallel, the state of the electron spin that relates to the magnetism can be evaluated by measuring the difference in
the intensity of γ rays in (1) and (2).

Spintronics is a new technology that uses the up and down
electron spins like the “0” and “1” values in a digital circuit.
This technology is expected to lead to the development of newdevices with features such as very low power consumption,
which cannot be achieved through conventional semiconductor
technology. In advanced research on magnetic substances for
spintronics, the evaluation of the electron spin is important.
By using positron annihilation spectroscopy (PAS), we canobtain information on the electron spin existing in the material
surface or interface more easily than by the conventional
technique.
A positron, which is the antiparticle of an electron,annihilates with the electron in the material, and γ rays are
emitted. In the conventional PAS method, only the electron
momentum is measured from the energy spectrum of theγ rays. We tried to detect the direction of the spin in addition
by developing a spin-polarized positron beam, because
positrons, like electrons, also have spin.
The positron beam can be formed by collecting thepositrons that are emitted from a radioisotope. When thespin polarization is improved by selecting the spin directionof the positrons, the beam intensity decreases. To achieveboth high spin polarization and sufficient beam intensity, weused a germanium-68 radioisotope that emits highly spinpolarized positrons. By using this radioisotope produced
by proton beam irradiation of a gallium nitride target ina cyclotron accelerator, a positron beam with a high spinpolarization of 47% was successfully produced (Fig.7-11).
This value is almost twice that of a positron beam producedby the commonly used sodium-22 radioisotope. When a pureiron sample was irradiated with this spin-polarized positron
beam, the intensity of the γ rays was increased when the spin
direction was controlled, so positrons annihilated easily withelectrons. This result shows that detailed information about
the electron spin, which causes the magnetism of magnetic
substances, can be extracted.
The PAS method using the spin-polarized positron beam is
expected to be a new evaluation technique that is necessary
for the future development of spintronics.
This work was supported by a Japan Society for the
Promotion of Science (JSPS) KAKENHI Grant-in-Aid for
Scientific Research (B) (No.24310072).

Reference
Maekawa, M. et al., Development of Spin-Polarized Slow Positron Beam using a 68Ge-68Ga Positron Source, Nuclear Instruments and Methods in Physics
Research B, vol.308, 2013, p.9-14.

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8

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 government activities
and projects both inside and outside JAEA

Providing technological platform
to industries and universities

Tackling the recovery
from Fukushima-Daiichi accident

Conservation and improvement of basis
for nuclear science and engineering

Preparing basic codes and databases

Creating innovative technology(Partitioning and transmutation, etc.)

Cultivating human resources and basic technology

Fig.8-1 Roles of nuclear science and engineering research
We play various roles in the conservation and improvement of the basis for nuclear science and engineering.

Nuclear science and engineering research aims to promotevarious activities, as shown in Fig.8-1, including recovery
from the accident at the Tokyo Electric Power Company,
Incorporated Fukushima Daiichi Nuclear Power Station (1F).
To this end, research is being conducted in the areas of nuclear
data and reactor engineering, fuels and materials engineering,
nuclear chemistry, and environment and radiation science. Joint
research with industry and academia is also being promoted via
the Nuclear Engineering Research Collaboration Center.
In the area of nuclear data and reactor engineering, theJapanese Evaluated Nuclear Data Library (JENDL) is beingedited, and nuclear data are being measured to improve it.
Calculation methods for nuclear reactor design (Topic 8-1),transmutation technologies to reduce the amount of long-lived
radioactive waste (Fig.8-2), and so on are also being studied.
Regarding the accident at 1F, the decay heat in the reactor cores
was evaluated (Chapter 1, Topic 1-17).
In the area of fuels and materials engineering, research on
the nuclear fuels and materials used in nuclear reactors andfuel cycle facilities is being promoted (Fig.8-3, Topic 8-3).
Regarding the accident at 1F, the properties of the fuel debris
and materials immersed in sea water are being investigated.
In the area of nuclear chemistry, research is being promoted
on basic data for reprocessing, separation of long-lived
nuclides from nuclear wastes, and detection of extremely small
amounts of nuclear materials (Topics 8-4, 8-5, 8-6, 8-7). These
technologies are also being applied to the recovery from theaccident at 1F.
In the area of environmental and radiation science, studiesof the behavior of radionuclides in the environment and thedevelopment of a database for radiation protection are being
promoted (Topics 8-2, 8-8, 8-9). The behaviors of 137Cs releasedby the accident at 1F in the marine sediment and forest topsoil arebeing studied in order to predict the long-term environmental
consequences (Chapter 1, Topics 1-1, 1-8).
In collaboration with industry, a production technique for
medical radio-isotopes using an accelerator-driven neutron
source has been being developed. A new corrosion-resistantmaterial is also being developed.

Proton Beam
Spallation
Neutron TargetCore(with MAs)

Fig.8-2 Accelerator-driven system (ADS) to transmute
minor actinides (MAs)
An 800 MWt ADS can be operated with a 1.5 GeV, 30 MW
proton beam at maximum power. Through operation of one
ADS, the amount of long-lived MA nuclides discharged from
10 units of a 1 GWe light water reactor can be transmuted
into short-lived or stable ones.

Manipulators

Fig.8-3 TRU High-Temperature Chemistry (TRU-HITEC)
module installed at NUCEF
Various thermochemical data are being measured for highly
radioactive trans-uranic (TRU) elements in an inert gas
atmosphere.
JAEA R&D Review 201383

Nuclear Science and Engineering Research

8-1

Method of Accurately Designing New Types of Reactors－ Advanced Method Using Experiments (Extended Cross Section Adjustment) －(a) Cross-section adjustment(conventional method)(b) Extended cross-section
adjustment

True
design value

True
design value

Designvalue

Design value calculatedby the true
Designcross-section set
value

○

Design value calculatedby the true
cross-section set
Consideration of
the target corecharacteristics○○○

Design value
without correction○○
Experimental data○
Design value
without correction

Table 8-1 Comparison of prediction accuracy
The “prediction” is the ratio of the design value to that without
correction. The “uncertainty” represents the sum of the design
value uncertainty components induced by the experiment,
analysis method, and nuclear data. The extended cross section
adjustment improves the prediction accuracy more efficiently
than the conventional method.

Design target
core
parameters(a) Cross-section
adjustment(conventional method)
Prediction○(b) Extended crosssection adjustment

Uncertainty
Uncertainty
Prediction(%)(%)

Experimental data

Fig.8-4 Features of the new adjustment method
The conventional method (a) aims to produce a design value
close to the calculation value by using the directly unknowable
true cross section set. In contrast, the present method(b) makes the design value closer to the true design value by
considering the target core characteristics.

To design a new-type of reactor, it is necessary to accurately
predict the nuclear characteristics of the reactor coreparameters, which are determined by the reactions between
neutrons and nuclei. These core parameters can be predictedby nuclear data and computer codes. The former describe thefeatures of the nuclear reactions, whereas the latter analyze thebehavior of neutrons. The accuracy of the predicted value canbe validated with a full-scale, well-simulated experiment, aso-called mock-up experiment. However, such an experiment
would be very difficult because it incurs huge costs andrequires full-scale nuclear fuels.
To solve these problems, a design method called crosssection (set) adjustment has been used in fast reactor core
design study. The cross section set consists of nuclear data
converted to be applicable to the computer codes. The crosssection adjustment method based on the Bayes theorem
enables us to improve the prediction accuracy of the coreparameters by adjusting the cross section set to existing
experimental data. In this method, the target core parameters
calculated by the adjusted cross section set are used directly as
the design values. This method, therefore, has the advantagethat all the design values are physically consistent with eachother.
On the other hand, another design method called the

Criticality

0.99770.30

0.99820.21

Powerdistribution

1.000

1.4

0.996

0.9

Control rod
worth

0.969

1.4

0.968

1.1

Improved prediction accuracy

extended bias factor method was developed in a previous
study by the Japan Atomic Energy Agency. This method has
the advantage that it can give the most accurate design values.
However, the extended bias factor method does not attempt tomaintain consistency among the design values.
In the present study, a new design method called extended
cross section adjustment is developed. The new method
theoretically integrates the two existing design methods
mentioned above. The conventional adjustment method alters
the cross section set in order to maximize its correctness by
optimizing it to the existing experimental data. In contrast,
the present method adjusts the cross section set to minimize
the uncertainty of the design value by taking into account thetarget core characteristics. Therefore, it is possible to improve
the prediction accuracy further (Fig.8-4). By applying the
new method to a typical large fast reactor design analysis, it isdemonstrated that the prediction accuracy is improved more
efficiently than in the conventional method (Table 8-1).
The new method not only clarifies the theoretical relation
between the conventional cross section adjustment method
and the extended bias factor method, but also successfully
improves the prediction accuracy. In the future, the new
method will be used for a reactor design study as a design
method that efficiently improves the prediction accuracy.

Reference
Yokoyama, K. et al., Extended Cross-Section Adjustment Method to Improve the Prediction Accuracy of Core Parameters, Journal of Nuclear Science andTechnology, vol.49, no.12, 2012, p.1165-1174.

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Accurate Prediction of Nuclear Fragmentation by Energetic Radiation

Non-equilibrium
phase

Collision

Energy: HighMulti-fragmentation

Fission andevaporation stage

Formation of
excited nucleus

Energy: Low

Formation offragments

Production cross-section(mb)－ Fragmentation Cross Section Measurement by New Method and Theoretical Model Development －Production cross-section
(mb)

8-2

10 3
10

2

10

1

10 0
10
10

Product: 24
Experiment
Simulation

Previous simulation
0
100
200
300
400
Incident carbon ion energy (MeV/nucleon)

10 2

10 1

Product: 83 Rb
Experiment

Simulation
Previous simulation
10 0
0
100
200
300
400
Incident carbon ion energy (MeV/nucleon)

Fig.8-5 Simulation flow of high-energy nuclear fragmentation
reactions
A model for simulating the multi-fragmentation process, indicatedin red, was added in this research. In the revised simulation flow,energetic nuclei form fragments through a reaction that forms
multiple fragments.

Fig.8-6 Comparison of fragment production reaction
cross sections (upper: NatPb(C,x)24Na reaction, lower:
NatPb(C,x)83Rb)
The previous simulation underestimates the cross sections
by a factor of 10 to 100 compared to the measurement data,
whereas the new calculation using PHITS as revised in thisresearch agrees with the measurement data very well.

In accelerator facilities, various radioactive fragments areproduced by spallation of nuclei by high-energy radiation.
Because radioactive fragments become radiation sources inaccelerators after the operation period, the fragments must
be considered, e.g., for safety management in maintenance.
However, the radiation transport simulation codes used forshielding calculations cannot accurately predict fragment
production from heavy elements.
To solve this problem, we measured the fragment
production cross sections of lead bombarded by carbon ionsat the National Institute of Radiological Sciences heavy ion
therapy facility, HIMAC, and revised the reaction models of
the Particle and Heavy Ion Transport code System (PHITS)
on the basis of the experimental data. In the experiment, crosssections in a wide energy range (from 50 to 400 MeV/nucleon)
were measured by a new method, in which stacked lead plates
were irradiated and the fragments in each plate were measured.
Calculation by the previous version of PHITS underestimatedthe cross sections by a factor of 100 at the maximum. We
assumed that the underestimation was attributed to the absenceof a multi-fragmentation process. In this process, nuclei aredisintegrated into multiple small fragments, whereas the old
PHITS assumes that all the excited nuclei undergo a fission/
evaporation process to form fragments. Therefore, a newmodel, the statistical multi-fragmentation model, was added tothe calculation model of PHITS; it determines the partition of
nuclei on the basis of the probabilities of various partitioning
patterns calculated using statistical mechanics (Fig.8-5). The
fragment production cross sections calculated by the revised
PHITS reproduced the measured ones accurately (Fig.8-6).
In this research, accurate prediction of the radioactive
fragment production cross sections by PHITS was realized;
this contributes remarkably to the precise evaluation ofradiation safety in high-energy accelerator facilities.
This work was partially supported by a Japan Society for
the Promotion of Science (JSPS) KAKENHI Grant-in-Aid
for Research Activity Start-up (No.24860072). This article
includes the results of the Research Project with Heavy Ions at
NIRS-HIMAC.

Reference
Ogawa, T. et al., Analysis of Multi-Fragmentation Reactions Induced by Relativistic Heavy Ions using the Statistical Multi-Fragmentation Model, Nuclear
Instruments and Methods in Physics Research A, vol.723, 2013, p.36-46.
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Nuclear Science and Engineering Research

8-3Life Prediction of Material in Nuclear Reprocessing Plants

Cooling condenser

Insulated cylindricalelectrode
Electrochemical
bridgeCorrosion potential Ecorr (V)－ Corrosion Characteristics of Stainless Steel in Boiling Nitric Acid Solution Including Np －3 mol/dm 3 HNO3
+ initial Np(VI)

SUS310Nb
Np: 5.3x10 -3 mol/dm3

1.0

3 mol/dm 3 HNO3
+ initial Np(V)0.8

3 mol/dm 3 HNO3
0.6

0.4

0.2
300

320

340

360

380

Temperature (K)Fig.8-8 Effect of temperature on corrosion potential ofstainless steel (SUS310Nb)
The increasing corrosion potential caused by a slight amountof Np accelerates corrosion; the behavior depends on thevalence of the Np and the temperature.

Optical fiber for
remote spectrum analyzer
Fig.8-7 Appearance of developed electrochemicalcell including optical cell for spectral analysis
The test cell is combined with a transparent heater,a potential probe, and an optical cell. It is operated
remotely. The amount of solution is approximately 15 cm3.

Nuclear reprocessing has been promoted for the reuseof uranium (U) and plutonium (Pu) in spent reactor fuels
worldwide. A corrosive nitric acid solution is used in nuclear
reprocessing, which involves materials containing metallic ionsthat accelerate corrosion. Therefore, accelerated corrosion ofstainless steel, one of the component materials, is an issue innuclear reprocessing plants.
It is well known that among metallic ions, Pu and neptunium(Np), which are oxidized in boiling nitric acid solution tometallic ions in higher oxidation states, accelerate corrosion
the most. Furthermore, the amount of Pu and Np that canbe handled in an experiment is restricted because they areradioactive elements.
We developed a specially designed small electrochemical test
cell integrated with an optical cell for spectroscopic analysis
in order to concomitantly evaluate the oxidation states of Np
ions and the polarization curves of stainless steel in boiling
nitric acid solution. The goal of this study is to understand thecorrosion acceleration mechanism of stainless steel.
The test cell shown in Fig.8-7 enables (1) the handling of asmall amount (ca.15 cm3) of test solution, (2) simultaneous

Np(V)
Np(VI)

100

Existence ratio (%)

Optical cell forspectral analysis
Quartz glass cell withtransparent heater

80
60
40
20
Initial Np(V): 5.3x10-3 mol/dm3

0
300

320

340

360

380

Temperature (K)Fig.8-9 Effect of temperature on valence change in Np
Pentavalent Np(V) is oxidized to hexavalent Np(VI) by
increasing the temperature of the solution. This suggeststhat the oxidized Np(VI) ions accelerate corrosion.

spectroscopic characterization and electrochemical monitoringof the solution, and (3) stable electrochemical measurements ina boiling nitric acid solution.
Fig.8-8 shows that the corrosion potential Ecorr of thestainless steel increases with increasing temperature when
the solution contains Np ions. The presence of Np ionsaccelerates the corrosion of stainless steel by increasing thecorrosion potential. Further, the corrosion potential in thesolution containing Np(VI), in a higher valence state, does notchange with increasing temperature, whereas that of a solution
containing Np(V), in a lower valence state, increases.
Fig.8-9 shows the existence ratios of Np(VI) and Np(V) in
3 mol/dm 3 HNO 3 with initial Np(V) ions at various
temperatures. The existence ratio of Np(VI) increases withincreasing temperature.
In conclusion, Np(VI) oxidized by nitric acid was found toraise the corrosion potential and accelerate the corrosion ofstainless steel.
This work was part of work commissioned by the Japan
Nuclear Energy Safety Organization (JNES).

Reference
Kato, C. et al., Electrochemical Measurements in Boiling Nitric Acid Solutions Containing Radioactive Elements by using Small Cell with a Portion forSpectral Analysis Function, Zairyo to Kankyo, vol.60, no.2, 2011, p.69-71 (in Japanese).

86

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Nuclear Science and Engineering Research

8-4Discovery of Thermochromic Uranium Complexes－ Coordination Chemistry in Ionic Liquids Using Time-Resolved Laser-Induced Fluorescence Spectroscopy －Heating
Cooling

Intensity (a.u.)

Fig.8-10 Thermochromic behavior of [C4mim]3[UO2(NCS)5]
Reversible change in color: solid (left, -196 °C) to sol (right,
90 °C). An orange uranium complex is uncommon.

450

500

550

600

650

Wavelength (nm)Fig.8-12 Time-resolved luminescence spectra of[C4mim]3[UO2(NCS)5] recorded under UV irradiation
The weak luminescence at 90 °C (thin line) becomesenhanced at -196 °C (bold line).

Inductively coupled plasma mass spectrometry or othermass spectrometric methods are used in chemical analysisof actinides in water or organic solvents instead of radiation
measurements. These techniques, however, are inapplicable
to in-situ observations of the dissolved chemical species. In
contrast, time-resolved laser-induced fluorescence spectroscopy(TRLFS) enables qualitative analysis of solutions. Investigating
the energy dissipation processes after irradiation by laser
pulses is of necessity our approach to studying the chemistry ofactinide complexes.
Observation of dissolved metal ions at high concentrations
using TRLFS has revealed the coordination structure of theuranium complex, which has remained unknown in wateror organic solvents. Ionic liquids are a series of salts that arecomposed entirely of ions and melt at room temperature. Because
they are stable at 100 °C, novel physical properties of thesesolvents at higher temperatures can be designed. We performed
a dissolution test of uranyl ions (UO22+) in 1-butyl-3-methyl
imidazolium thiocyanate ([C 4mim][NCS]; the structureis shown in Fig.8-11), which formed a liquid at very highconcentrations. Further addition of uranium affords a metal saltwith a [UO22+]:[NCS−] composition ratio of 1:5, which remains
a sol at room temperature. Surprisingly, this substance reveals

Fig.8-11 Crystal structure around [UO2(NCS)5]3- (-196 °C)−
There are five NCS ions on the equatorial plane coordinatedto a uranyl ion (UO 2 2 + ), surrounded by three 1-ethyl-3methylimidazolium cations. This salt forms a bright yellow solid
at -196 °C. The symbols are as follows: oxygen: red, nitrogen:blue, carbon: black, sulfur: yellow, and hydrogen: pink. The
chemical structure of the ionic liquid is shown above.

thermochromism, that is, temperature-dependent color changes
due to external stimuli (Fig.8-10). There have been no studies
describing this property of uranium complexes; this salt becomesyellow on solidification when cooled at the liquid nitrogentemperature, whereas it melts on heating, becoming red.
To understand this phenomenon, the thermochromic ionicliquid was crystallized at low temperature, resulting in the
crystal structure shown in Fig.8-11, where the coordination
number around uranium on the equatorial plane at low
temperature is five. In addition, the luminescence spectrum
under irradiation by ultraviolet light is shown in Fig.8-12;
it exhibits quenching of luminescence at room temperature.
The symmetrical structure around the U moiety, therefore,
affects the spectrum by reducing the coordination number onheating. The orange or red UVIO22+ compounds are uncommon,so a coordination number of four is attributed to the local
environment between the uranyl and the ligand as well as tocation–anion interactions.
This work was financially supported by a Japan Society for
the Promotion of Science (JSPS) KAKENHI Grant-in-Aid forYoung Scientists (B) (No.22760679). The Excellent Young
Researcher Overseas Visit Program (No.21-5317) of JSPS is
also acknowledged.

Reference
Aoyagi, N. et al., Thermochromic Properties of Low-Melting Ionic Uranyl Isothiocyanate Complexes, Chemical Communications, vol.47, issue 15, 2011,
p.4490-4492.
JAEA R&D Review 201387

Nuclear Science and Engineering Research

8-5

Attempt to Establish a Reprocessing System That Uses Monoamides－ Continuous Extraction of Uranium and Plutonium Using Mixer-Settler Extractors －
C2H5
C4H9

CH

CH2

C4H9

CH

CH2

N

C2H5

C

C3H7

0

【Mixer-settler extractor for U and Pu extraction】
U and Pu in a simulated dissolution solution of
spent nuclear fuel are extracted using DEHBA
by adjusting nitric acid concentration to 3 mol/ℓ.

Fig.8-13 Structure of N,N-di(2-ethylhexyl)
butanamide (DEHBA)【Mixer-settler extractor for Pu back-extraction】
Pu and some U are back-extracted by decreasing
nitric acid concentration to 0.7–1 mol/ℓ.

Distribution ratio

100

DU

10

【Mixer-settler extractor for U back-extraction】
The residual U is back-extracted by decreasing
nitric acid concentration to 0.1 mol/ℓ.

1

D Pu

0.1
0.01
0.01

0.1

1

10

Nitric acid conc. (mol/ℓ)
Fig.8-14 Effects of nitric acid concentration on the
distribution ratios of U (DU) and Pu (DPu) by DEHBA
Distribution ratios of U and Pu were obtained using
DEHBA in dodecane as the organic phase and nitricacid containing U or Pu as the aqueous phase.

Fig.8-15 Photograph of continuous extraction experiment using
mixer-settler extractors for U and Pu extraction
Extraction properties of DEHBA toward uranium and plutonium wererevealed using mixer-settler extractors in a glove box.

We have been advancing research on solvent extraction using
monoamide extractants for spent nuclear fuel reprocessing.
Because monoamides do not contain phosphorus, they can be
decomposed into gases by incineration, reducing the secondary
waste generation. In addition, monoamides can be diluted in
dodecane, which is used in the commercially operating process(PUREX process); thus, current technologies developed for
reprocessing equipment can be applied. The PUREX processuses tributyl phosphate (TBP) as an extractant, and monoamides
are as stable as TBP against radiolysis. Furthermore, thedegraded products of monoamides do not have detrimental
effects on solvent extraction.
The extraction properties of monoamides toward uranium (U)and plutonium (Pu) depend on their structures. As an example,
the effects of the nitric acid concentration on the distribution
ratios of U (DU) and Pu (DPu) by N,N-di(2-ethylhexyl)
butanamide (DEHBA, Fig.8-13) are shown in Fig.8-14. Here,the values of DU and DPu indicate the extraction power of
monoamides toward U and Pu, respectively, and the ratios of
U and/or Pu extracted by monoamides increase with increasing
distribution ratio.

As shown in Fig.8-14, DEHBA extracts U and Pu when the
nitric acid concentration is ca. 3 mol/ℓ. DEHBA still extracts
U when the nitric acid concentration decreases to 0.7–1 mol/ℓ,
whereas it extracts very little Pu. We focused on the specificfeatures of DEHBA and proposed a reprocessing system inwhich U and Pu are separated by adjusting the nitric acidconcentration. The PUREX process requires reducing agentsfor Pu. On the other hand, our system using DEHBA does not
require reducing agents, which simplifies reprocessing.
We conducted a continuous U and Pu extraction experimentusing mixer-settlers, with the goal of establishing a reprocessing
system that uses DEHBA (Fig.8-15). The results showed thatthe ratios of U and Pu extracted from a simulated dissolution
solution of spent nuclear fuel are more than 99.9%, and the ratioof Pu recovered in the U+Pu fraction is more than 97%. These
results have supported the realization of a reprocessing systemthat uses DEHBA.
Part of this study was the result of research entrusted to
the Japan Atomic Energy Agency (JAEA) by the Ministry of
Economy, Trade and Industry (METI).

Reference
Ban, Y. et al., Distribution of U(VI) and Pu(IV) by N,N-di(2-Ethylhexyl)Butanamide in Continuous Counter-Current Extraction with Mixer-Settler
Extractor, Solvent Extraction and Ion Exchange, vol.30, issue 2, 2012, p.142-155.

88

JAEA R&D Review 2013

Nuclear Science and Engineering Research

8-6

(a)

One-Pot Biofabrication of Protein-Decorated Gold Nanoparticles－ Facile, Rapid, and Efficient Biofabrication and Immunoassay －(d)Antibody-binding protein (ZZ)

Gold ions
Antibody

Thioredoxin (Trx)Gold-binding A3 peptide

One pot
Synthesis of AuNPs
Protein-decorated
Protein immobilization

ZZ-Trx-A3 fusion protein

AuNP–antibody conjugates

AuNPs

1.2

(b)

(e) 1.0 Addition of

(c)

0.8

AbsorbanceAbsorbance

1.0
0.8
0.6
0.4

Aggregation of AuNPs( f)antigen

Antigen

0.6
AuNPs
AuNPs + Antigen

0.4
0.2

0.2
0.0
400

Antigen

50 nm
500

600

700

800

Wavelength (nm)

0.0
400

50 nm
500

600

700

800

Wavelength (nm)Fig.8-16 Synthesis of AuNPs and immobilization of proteinusing ZZ-Trx-A3 fusion protein(a) The ZZ-Trx-A3 construct can simultaneously direct both thesynthesis of AuNPs and the immobilization of the protein on
the AuNPs in a single step.
(b) The synthesized AuNPs display the SPR band at 525 nm
and yield a reddish solution.
(c) TEM shows that the synthesized AuNPs exhibit sphericalmorphologies with a narrow size distribution (4.4±1.6 nm).

Fig.8-17 Immobilization of antibody on AuNPs andimmunoassay(d) AuNP–antibody conjugates are readily prepared throughspecific binding of the antibody to the ZZ domain on the
AuNPs; they can detect the target antigens based on
AuNP aggregation.
(e) Addition of target antigens to the AuNP solution yields aturbid solution and generates AuNP precipitation, which
is clearly visible to the naked eye.
( f) TEM demonstrates the formation of AuNP aggregates
through the antigen–antibody reaction.

Gold nanoparticles (AuNPs) have received increasing
attention in the field of nanotechnology because of their
attractive optical properties. AuNP–antibody conjugatesare particularly valuable in biomedical applications such as
immunoassays for the detection of target antigens and are
actually used in a pregnancy test kit. However, the typicalfabrication of AuNP–antibody conjugates is time-consuming andrequires multiple steps: AuNP synthesis followed by antibody
immobilization. In addition, the physical adsorption method,which is commonly employed to immobilize antibodies on thesurface of AuNPs, usually yields randomly oriented proteins.
To address this issue, we demonstrate a simple and new
strategy for biofabrication of AuNP–antibody conjugatesusing peptide-functionalized proteins. The strategy isbased on a one-pot direct approach that couples peptidemediated biomineralization and protein immobilization in
one simultaneous process. We prepared a tripartite fusion
protein (ZZ-Trx-A3) consisting of an antibody-binding protein(ZZ domain), a AuNP-stabilizing protein (thioredoxin) anda gold-binding A3 peptide. By mixing ZZ-Trx-A3 with gold
ions, intensely reddish solutions were obtained, and thesurface plasmon resonance (SPR) band was observed at 525

nm (Fig.8-16). This indicates the formation of AuNPs fromgold ions. Using transmission electron microscopy (TEM),
we confirmed that the AuNPs showed spherical morphologies
with a narrow size distribution (average diameter 4.4 1.6 nm).
The novel method enables facile and rapid fabrication ofprotein-decorated AuNPs within 20 min and is preferable toconventional methods requiring a few days. In addition, the
AuNPs remained suspended in the solution for at least 6 months
and in salt solutions adjusted to high concentrations, indicating
the high stability of the AuNPs in aqueous media.
As shown in Fig.8-17, protein-decorated AuNPs can bereadily derivatized with intact antibodies owing to the specificbinding to the ZZ domain on the AuNPs. The addition of target
antigens to the AuNP solution yielded a turbid solution owing to
AuNP aggregation through the antigen–antibody reaction, which
increased the background absorbance. Furthermore, AuNP
precipitation was observed after the solution stood for a few
hours, and this precipitation was clearly visible to the naked eye.
We expect that our approach can be expanded to produce avariety of biomolecule-decorated AuNPs and can be used inversatile applications in addition to biomedical applications.

Reference
Shimojo, K. et al., Facile, Rapid and Efficient Biofabrication of Gold Nanoparticles Decorated with Functional Proteins, Analyst, vol.137, issue 10, 2012,
p.2300-2303.
JAEA R&D Review 201389

Nuclear Science and Engineering Research

8-7

Analysis of Age Determination of Individual Plutonium Particles－ Contribution of Development of Ultra-Trace Analytical Technique to IAEA Safeguard Activity －Purification of Pu reference solution
10 μm

Development oftechnique for preparing Pu particles

Dissolution of single Pu particle

Addition of 243Am solution
Development of techniquefor ion-exchange separation of Pu and Am
Development of isotope ratio measurement243
Am/239Pu, 241Pu/239Pu, 241Am/243Am

Determination of Pu purification age
deduced from 241Am/241Pu ratio

1 μm

Fig.8-19 Pu particles prepared from a Pu reference solution
Micron-size age-known Pu oxide particles (right) were prepared.
A technique for isolating a single particle was developed. The
age of an isolated Pu particle (left) was determined.
Difference from the actual age (year)

Development oftechnique for isolating single Pu particles

The microneedle
attached to a SEM forthe isolation of a single
Pu particle.

0.8

Actual Pu purification age: 3.92 years

0.6
0.4
0.2

average
0.11 years(41 d)

0.0
–0.2
–0.4
1

2

3

4

5 6 7
Sample No.

8

Fig.8-18 Development of analytical procedure for Pu
purification age determination
An analytical technique for determining the purification age ofan ultrafine Pu particle was realized by combining techniques
for preparing a single Pu particle and for chemical separation.

Fig.8-20 Pu purification age of single Pu particles determinedby 243Am spike method
The average age of eight Pu particles was in good agreement withthe actual age (3.92 years) with a difference of 0.11 years (41 d).

Inspectors from the International Atomic Energy Agency(IAEA) take an environmental sample such as dust by wiping
the floors in a nuclear facility, and the amount and isotopic
composition of ultra-trace (10-15–10-12 g) uranium (U) andplutonium (Pu) are analyzed to detect undeclared nuclear
activities.
We developed a technique for isolating a single Pu particle
and an analytical technique for the precise determination of
the Pu purification age to upgrade the methods used to manage
nuclear materials in Japan (Fig.8-18). The Pu purification age
is evaluated by measuring the americium-241 (241Am)/241Pu
atomic ratio precisely. A single Pu particle of ultra-trace amount(<10-12 g) should be analyzed because Pu of different origins
may be mixed in a sample. 241Am and 241Pu must be chemicallyseparated before they are subjected to mass spectrometry
because they have the same atomic mass and cannot be
separated with a mass spectrometer. The known techniques
cannot provide the precise age of young Pu samples or ultratrace samples because of the low 241Am/241Pu atomic ratio. It isalso necessary to prepare a reference sample of young single
Pu particles for evaluating the accuracy and precision of ouranalytical results. We developed the following techniques: thepreparation of micron-diameter Pu oxide particles (Fig.8-19)
from Pu purified by ourselves; a technique for isolating a single
Pu particle by a microneedle attached to a scanning electron
microscope (SEM); and chemical separation and isotope
measurement of ultra-trace Pu and Am.
Precise measurement of the 241Am/241Pu ratio in a single Pu
particle of 10-12 g was achieved. Each Pu particle was dissolved
and spiked with pure 243Am. The 241Am/241Pu ratio was obtained
from the 243Am/239Pu ratio of a sample solution and the isotope
ratios of Pu and Am in the chemically separated fractions.
These ratios were measured with an inductively coupled plasma
mass spectrometer. The determined age was in good agreementwith the actual age (3.92 years), with a difference of 0.11 years(41 d) on average (Fig.8-20). By using this analytical technique,traces of nuclear activity can be detected. We can contributeto the IAEA safeguard activity by analyzing samples with our
excellent techniques.
The present study was sponsored by the Ministry of
Education, Culture, Sports, Science and Technology of Japan(MEXT).

Reference
Miyamoto, Y. et al., Precise Age Determination of a Single Plutonium Particle using Inductively Coupled Plasma Mass Spectrometer, Radiochimica Acta,
vol.101, issue 11, 2013, p.745-748.

90

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Detailed Simulation of Transfer of Radioactivity in Land Surface Ecosystems－ Development of Nuclide Transport Model for Atmosphere-Vegetation-Soil System －T exchanges in vegetation

Turbulent diffusion
Exchange throughTranslocation
Photosynthesis to other parts
stomata
Atmospheric
T in leaf
T in leaf
HTO
cellular waterorganic matter(a)
T concentration in leaf
cellular water (Bq m –3)

T transport in surface
atmosphere1 x 1010

T content in leaf organic
matter (MBq kg –1)

8-8

1.00

Respiration
Exchange
Condensation
Evaporation

HTO in soil air

Diffusion

Uptake by roots

HTO in soil water

T transport in soil

Diffusion Advection

Deposition of atmospheric HTO to leaf cellular water
Release of HTO fromleaf cellular water toatmosphere1 x 10 9
1 x 108

Obs:Height at 0-1.0 m
C a l: Height at 0.40 m
C a l: Height at 0.60 m
C a l: Height at 0.85 m

1 x 107

C a l: Average over the height1 x 10 6
1 x 10 5(b)Assimilation of T in leaf cellular water to organic matter
Translocation of T in leaf organic
matter to other plant parts

0.75
0.50
0.25
0.00

0

12

24

36

48

72

60

Elapsed time from 0:00 on the day of exposure (h)Fig.8-21 T transports and exchanges considered in thedeveloped model
The model simulates the transport of HTO in a multilayered
atmosphere and soil, and exchanges of T in vegetation, byconsidering the water and CO2 cycles.

Fig.8-22 Reconstruction simulation of experiment in whichgrape plants were exposed to atmospheric HTO vapor(a) The model simulates well the changes in T concentrationin leaf cellular water due to deposition of atmospheric HTO
during exposure (indicated by hatch marks) and release of
HTO from the leaf cellular water after exposure. (b) The model
also successfully predicts changes in T content of leaf organic
matter that are caused by assimilation of leaf cellular water T byphotosynthesis and translocation of T in the leaf organic matter.

Tritium ( 3H, T) and radiocarbon ( 14C) are importantradionuclides in the assessment of the radiological impact
to the public of nuclear industries because of the long halflives of T (12 years) and 14C (5730 years) and the biologicalimportance of hydrogen and carbon in all life forms. T and 14C
are discharged through the operation of nuclear power plants
and fuel reprocessing plants; in the natural environment, T
and 14C exist mainly in the form of tritiated water (HTO) and
14CO , respectively. Thus, the discharged T and 14C are soon2
incorporated into the water and C cycles in the surrounding
environment and are assimilated to vegetation as organic
matter. The assimilated T and 14C eventually deliver a dose to
the public through ingestion pathways; therefore, the transfer
of T and 14C in land surface ecosystems should be precisely
predicted to assess the dose due to nuclear facilities. However,models that dynamically consider the water and CO2 cycles
and the resultant T and 14C transfer within a land surface
ecosystem have not been developed; thus, assessments of the T
and 14C-derived dose have contained a range of uncertainties.
We developed a model that predicts the transfer of T and
14C in a land surface ecosystem in detail. The dynamics ofT (Fig.8-21) and 14C in the surface ecosystem are modeled,
and the modeled T and 14C dynamics are incorporated into aland–surface water and CO2 transport model (SOLVEG-II,developed by the Japan Atomic Energy Agency). The developedmodel successfully predicted the accumulation of T (Fig.8-22)
and 14C in leaves on plants exposed to HTO or 14CO2.
For a comprehensive understanding of the land surface
T and 14C cycle, we then performed a range of numericalexperiments using the model. The results demonstrated thatduring atmospheric T deposition to the soil, the loadingsof HTO to the leaf cellular water due to root uptake of soil
HTO significantly control the T concentration in the leaf
cellular water, and hence the T content in leaf organic matter.
Our results also quantified the increase in the 14C content ofvegetation due to below-ground 14CO2 production caused bydecomposition of 14C-containing soil organic matter originatingfrom 14C-containing plant litter. Clearly, the model can be used
to clarify the role or importance of each elemental process inthe entire land surface T and 14C cycle; this useful information
cannot be obtained from field observations or laboratory
experiments.
As a more practical application of the model, we are now
investigating the transfer of T and 14CO2 discharged at a nuclear
facility over the surrounding land surface ecosystems througha simulation coupling the developed model and an atmosphericdispersion model.

Reference
Ota, M. et al., Importance of Root HTO Uptake in Controlling Land-Surface Tritium Dynamics after An-Acute HT Deposition: A Numerical Experiment,
Journal of Environmental Radioactivity, vol.109, 2012, p.94-102.
JAEA R&D Review 201391

Nuclear Science and Engineering Research

8-9

Subsurface Soils Participate in Global Carbon Cycle

1000

(a)

Mineral-free, decadally cycling
Mineral-free, long-term resident
Mineral-associated, decadally cycling
Mineral-associated, long-term resident
Total surface soil

← Increase due to “bomb” C
14

500

100

10

Equilibrium between← production and decay of “natural” 14C

0

Carbon inventory (kgC/m2)

14
C content ofmineral-associated SOM
in subsurface soil (‰)

14
C content of
atmospheric CO2(‰)－ Carbon Dynamics Revealed by Tracing “Bomb” Radiocarbon over Past Half-Century －(b)Soil 1
Soil 2

0
–100
–200
1950

8
6
4
2
0

1960

1970

1980

1990

2000

2010

Year

Surface Subsurface
Soil 1

Surface Subsurface
Soil 2

Fig.8-23 Carbon-14 content of atmosphere and subsurface
soils
Carbon-14 content is expressed as per mil deviation of 14C/12C
ratio of sample from that of atmospheric CO 2 in 1950. (a)Carbon-14 content of atmospheric CO2 rapidly increased due
to nuclear weapons testing in the early 1960s. (b) Carbon-14
content of mineral-associated SOM in subsurface soils
remained low; however, it increased markedly from 1992 to2005 owing to incorporation of “bomb” 14C.

Fig.8-24 Stock and turnover of C in subsurface soils
Subsurface soils store as much C as surface soils (theupper 20 cm). About half of the subsurface soil C was not
associated with soil minerals, >70% of which was estimatedto be decadally cycling C. Decadally cycling C was also
found in mineral-associated SOM. Overall, the amount of
C that turns over on decadal timescales in subsurface soils
reached 3–6 kg/m2.

Soil plays an important role in the global carbon (C) cycle
through soil–atmosphere C exchange. Recent studies have shownthat soils store more C in the subsurface horizons (20–60 cm
depth) than in the surface horizons (the upper 20 cm). However,subsurface soil C has received little attention in terms of its
contribution to C exchange because it has been thought to be very
stable on the basis of its average age of centuries to millennia.
We used radiocarbon (14C) to identify the dynamic natureof C in subsurface soils. There are two main sources of
14C: “natural” 14C that is produced in the upper atmosphereby cosmic rays, and “bomb” 14C that was produced byatmospheric nuclear weapons testing during the early 1960s(Fig.8-23(a)). Similar to stable 12C, 14C in the atmosphere isfixed as organic matter by plants via photosynthesis and thenenters soils. Carbon-14 decays with a half-life of 5730 years
in soils; thus, a lower 14C content in soil organic matter (SOM)reflects a longer residence time of C in the soils. On the other
hand, the incorporation of bomb 14C into the soils increases

the 14C content in SOM. Therefore, we hypothesized that
C exchange that occurs on timescales of decades throughsubsurface soils can be identified by tracing bomb 14C
incorporation over the past half-century, if it really exists.
The 14C content was determined for subsurface soil samples
taken several times from 1958 to 2009 at two forest sites in
California. We found an increase in the 14C content in mineralassociated SOM with a time lag of >20 years after the increasein the atmosphere (Fig.8-23(b)). A model analysis showedthat ~40%–70% of the SOM turns over on decadal timescales.
We also found bomb 14C incorporation in mineral-free SOM
and CO2 released from subsurface soils collected in 2009.
The results demonstrate that subsurface soils store a large
amount of C that is exchanged with the atmosphere (Fig.8-24)
and that a lagging response of the C to climate change ispossible. The findings will improve our understanding of the
global C cycle.

Reference
Koarashi, J. et al., Dynamics of Decadally Cycling Carbon in Subsurface Soils, Journal of Geophysical Research, vol.117, issue G3, 2012, p.G03033-1-G03033-13.

92

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9

Nuclear Hydrogen and Heat Application Research

Research and Development on Naturally Safe HTGR and Nuclear Heat Application Technologies
Major features

Low-temperature
heat utilization

HTGR Commercial Plant• Hydrogen production• Steam supply• Multipurpose heat utilization• Sea water desalination• Excellent economy even for small reactor • District heating• Naturally safe without any engineered• Agriculture, fisherysafety features

High Temperature engineering
Test Reactor (HTTR): Oarai, Ibaraki
1998 Criticality
2001 Full power
2004 950 °C

Fuel particles coatedwith ceramics
UO2

Electric generation • High-efficiency
by gas turbine
power generation

Worldʼs most advanced homegrown technology• Ceramic coating (TRISO) for uranium fuelTemperature limit: 1600 °C

Diameter ~1mm
Isotropic graphite

High strength, high heat conduction,
irradiation resistance• Heat-resistant, oxidation-resistant Ni-based superalloy• Heat insulation, anti-thermal expansion, anti-vibration
structure
Service temperature: ~950 °C

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

Long term (~3 times that of LWR),
complete retention of fission products(100~1000 times that of HTGRs in USA
and Germany)• 3-dimensional isotropic, irradiation-resistant graphite
Temperature limit: 2400 °C

Heat-resistant alloy

High-temperature
heat utilization

High-temperature (950 °C) heatcan be extracted from the reactor• High-temperature, high-pressure He handling
Service temperature: ~950 °C

Chemically, nuclear, and mechanicallystable coolingFig.9-1 Outline of HTGR with features, heat utilization, major specifications, and technologies of HTTR
The HTGR is a thermal neutron reactor, helium gas cooled and graphite moderated, and can meet various heat production requirements.
In particular, the HTGR can be a naturally safe nuclear reactor, and it is strongly expected to regain the public trust in nuclear power.

We have been conducting research on the naturally safe High
Temperature Gas-cooled Reactor (HTGR) and its applications
in our contribution to building a low-carbon society.
The HTGR can supply high-temperature heat at 950 °C, wellexceeding the 300 °C of current light water reactors (LWRs),
by using inert helium gas instead of water as a coolant. The
same amount of fission energy from uranium can produce
a greater quantity of electricity and hydrogen when higher
temperature heat is obtained from the reactor.
Three other cutting-edge Japanese technologies that make it
feasible to obtain such high-temperature heat in the HTGR havebeen developed by the Japan Atomic Energy Agency. The firstis fabrication technology for ceramic-coated fuel particles about
1 mm in diameter. Ceramics remain stable even at 2500 °C andprovide superior heat-resistant coating layers to containradioactive fission products within the fuel particles. The
second is fabrication technology for isotropic and irradiationresistant graphite blocks. Graphite is generally non-isotropic
in that its material properties such as strength and thermal
conductivity are dimension-dependent, which is to be avoided
in reactor construction. The third is a manufacturing technology
for heat-resistant and oxidation-resistant superalloys withcareful composition of the alloy content. With these major
technologies and further development of expertise in hightemperature structural design and helium gas handling, we

successfully produced high-temperature heat at 950 °C from thereactor in 2004 for the first time worldwide (Fig.9-1).
The heat from the HTGR is useful not only for power
generation but also for the production of hydrogen for fuel cellcars and direct reduction iron-making, and of a steam supply
for industry. The waste heat can be used for district heating
and desalination. The thermal discharge to the environment
can be dramatically reduced to less than 30%, compared to
67% for an LWR.
It is feasible to prevent accidental overheating and oxidationof the fuel coating layers and explosive gas generation using
physical phenomena without any engineered safety features.
The self-regulating features assure no harmful release ofradionuclides to the general population and the environment inany accident. It is this unique safety advantage of the HTGR
that is expected to regain the public’s trust in nuclear power,
which is strongly needed after the accident at the Tokyo
Electric Power Company, Incorporated Fukushima Daiichi
Nuclear Power Station.
The conceptual design of a small HTGR that has excellent
safety and economic potential has been completed, and the firstdraft of the safety design philosophy for the HTGR hydrogenproduction system has been presented. Furthermore, researchon improving the efficiency of the thermochemical iodine–sulfur process is ongoing (Topics 9-1, 9-2, 9-3).
JAEA R&D Review 201393

Nuclear Hydrogen and Heat Application Research

9-1

Small Nuclear Reactor for Multiple Heat Applications with Attractive Safety Features－ Conceptual Design of Small High-Temperature Gas Cooled Reactor for Developing Countries －Design targets and results
Design items

Design target

Criteria

Reduce the number of uranium
enrichments①Core②Component

Other design results
Results

Feasibility3○< 6 (HTTR: 12)

Design items

Results

Thermal power

50 MW

Reactor inlet
temperature

325 ℃

Reactor outlet
temperature

750 ℃,
900 ℃※

Increase core power density

3.5 MW/m 3 (HTTR: 2.5 MW/m3)

3.5 MW/ m3○

Increase refueling interval

730 d (HTTR: 660 d)

730 d

○

Increase thermal duty of IHX

20 MW (HTTR: 10 MW)

20 MW

○

0 MW〜25 MW○

Demonstration test canbe performed using IHX○Primary pressure

4 MPa

Biological shielding concrete temp. at normal operation: <65 ℃

64.8 ℃○

Satisfy the acceptance criteria for accidents in safety analysis

Mn-Mo
steel

All criteria are satisfied○

Reactor pressure
vessel material③SystemSatisfy the user requirements for
multiple heat applications④Safety

Design of the engineered safety
features of the residual heat removal
system (i.e., VCS) as passive methods

Heat supply for district heating network: 0 MW〜25 MW(50% partial load operation of steam turbine)
Demonstration test of gas turbine electricity generation andhydrogen production can be performed※ demonstration of the gas turbine andhydrogen production

Example of system configuration①Core design
The core performance is enhanced by
the optimization of the fuel temperature
distribution, which is obtained by
optimizing the layout of the different
enrichment fuels and burnable poisonsin the core.
④Safety design
The engineered safety feature of theresidual heat removal system (i.e., VCS)
is designed as passive methods. The
heat removal performance is ensured bynatural circulation of the cooling waterowing to the decrease in the pressure
loss, a component layout that increases
the water head, etc.

Vessel cooling system (VCS)

ReactorIntermediate heat exchanger (IHX)

Steam turbine

Steamgenerator(SG)

Hydrogen production,
Gas turbine

Electricity
generation

Districtheating②Component design
In the IHX design, the thermal duty is
increased to twice (i.e., 20 MW) that of
HTTR, ensuring the strength of the heattransfer tube for its own weight.

③System design
The various system configurations aredesigned by applying heat utilization incascade (i.e., installation of IHX and SG inserial, steam extraction from steam turbine,etc.) to satisfy the user requirements.

Fig.9-2 Design targets and results of small HTGR system design
The conceptual design of a 50 MWt small HTGR system was completed. It can satisfy the user requirements for multiple heatapplications, and its performance is superior to that of the HTTR. The technical feasibility was confirmed by the results of the core,component, and system design, and a safety analysis in the event of representative accidents.

We tested the High-Temperature engineering Test Reactor(HTTR) to establish the high-temperature gas cooled reactor(HTGR) fundamental technologies and to obtain the datarequired for commercial HTGR design. Since 2010, we also
developed the conceptual design for a 50 MWt small HTGR
system that can satisfy the user requirements, aiming atconstruction of a demonstration plant in a developing country
in the 2020s. Its performance was improved over that of the
HTTR without significant research and development, using theknowledge obtained from HTTR design and operation (Fig.9-2).
In the core design, the number of uranium enrichments was
reduced to three (i.e., one-quarter that of the HTTR), and theaverage core power density was increased to 3.5 MW/m3 (i.e.,
1.4 times that of the HTTR), satisfying the maximum fueltemperature criterion throughout the burnup period of 730 d.
In the component design, the thermal duty of the intermediate
heat exchanger (IHX) was increased to twice (i.e., 20 MW)

that of the HTTR; the strength of the heat transfer tube for its
own weight was ensured by increasing the tube diameter todecrease the stress and by increasing the flow rate to increase
the heat transfer performance. The system configurations were
determined so as to use the nuclear heat for district heating
and process heat based on a steam turbine system, and todemonstrate a helium gas turbine and hydrogen production tosatisfy the user requirements. In the safety design, the residual
heat removal system (the vessel cooling system) was designed
as a passive system. The safety was confirmed by a safety
analysis of representative accidents.
The Republic of Kazakhstan initiated a development
program for the nuclear industry in that country in June 2011,
which includes a construction plan for the Kazakhstan hightemperature gas-cooled reactor (KHTR) based on this small
HTGR system.

Reference
Ohashi, H. et al., Conceptual Design of Small-Sized HTGR System (IV) −Plant Design and Technical Feasibility−, JAEA-Technology 2013-016, 2013, 176p.
(in Japanese).

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Nuclear Hydrogen and Heat Application Research

Proposal for HTGR Safety Design Criteria upon User Request－ Allow H2 Plant Construction Coupled to HTGR under Non-Nuclear Regulations －980

Spent fuelstorage building

Stack

Reactor building

H2 plant

Turbine building
Safe
distance

Reactor power (%)

Nuclear facility

112

Allowable variance: +7 °C

960

108
940
104

Allowable variance: +3%

-200

920

Loss of load in H2 plant

100
0

200

400

Reactor outlet
temperature (°C)

9-2

900
600

Elapsed time (s)Fig.9-3 Conceptual layout of nuclear facility and H2 plant
To ensure the integrity of safety functions in the nuclear facility
against fire and explosion and the intrusion of hazardous
chemicals into the control room owing to leakage in the H2plant, an appropriate safe distance is set between the nuclear
facility and the H2 plant.

Fig.9-4 Reactor transient response during loss of load in
H2 plant
Variations in the power and coolant temperature in thereactor can be mitigated within the value allowed in normal
operation during loss of load in the H2 plant because of acooling system installed downstream of the H2 plant.

Toward the realization of hydrogen production using heatfrom a high-temperature gas-cooled reactor (HTGR), it isnecessary to ensure the safety of the nuclear facility against
postulated events initiating in the hydrogen production plant(H2 plant) and to construct the H2 plant under non-nuclear
regulations upon user request. The present study focuses
on a H2 plant using the thermochemical water splitting
iodine–sulfur (IS) process and investigates the safety design
criteria, which consist of the requirements for constructinghydrogen production plants under conventional chemical plant
regulations as well as the requirements for collocation of a
nuclear facility and a hydrogen production plant. In addition,
design considerations to meet the criteria are suggested.
The design considerations of the nuclear facility are
identified as the assurance of a safe distance between the
nuclear facility and H2 plant, the installation of isolation
valves against leakage of combustible gas; and early detection,
the assurance of a safe distance between the nuclear facilityand H2 plant against hazardous chemical leakage (Fig.9-3).
The mitigation of fission product migration from thereactor to the H2 plant and the assurance of continuous reactor
operation against abnormal conditions in the H2 plant are
identified as the requirements for constructing the H2 plantunder conventional chemical plant regulations. In addition,

the assurance of sufficient capacity for helium purification
systems and the installation of a cooling system downstreamof the H2 plant are suggested as design considerations to meetthe requirements.
The technical feasibility of the suggested design considerations
is evaluated for the HTTR-IS system, an HTGR hydrogenproduction system coupling the HTTR and a H2 plant using
the IS process. The results showed that a safe distance of
40 m, determined by the chemical plant regulations against
fire and explosion of combustible gas, can be ensured. In
addition, the toxicity limit issued by the United States National
Institute for Occupational Safety and Health against leakage ofhazardous chemicals can be satisfied. Furthermore, the results
demonstrated the feasibility of H2 plant construction under nonnuclear regulations by showing that the tritium concentrationin the H2 plant can be maintained below the regulation limit,
and normal reactor operation can be achieved during abnormal
conditions in the H2 plant (Fig.9-4).
The proposed criteria can be used for coupling not only a H2plant but also other chemical plants such as a steam reforming
plant. The adequacy of the criteria will be assessed under an
outside committee. We also aim to standardize the criteriainternationally under the International Atomic Energy Agency.

Reference
Sato, H. et al., A Proposal for Safety Design Philosophy of HTGR for Coupling Hydrogen Production Plant, JAEA-Technology 2013-015, 2013, 68p.
(in Japanese).
JAEA R&D Review 201395

Nuclear Hydrogen and Heat Application Research

9-3

Increased Energy Efficiency for Hydrogen Production－ Study of Ion Permeation Mechanism in Membrane of HI Concentrator －Hydrogeniodidedecomposition

Heat
H2
+
I2

I2

Production of
HI and H2SO4(Bunsen reaction)

2HI

Iodinecirculation

H 2 SO4

2HI + H2SO4

0.9

1/2O2
+
SO2 +H2O

Sulfurcirculation SO

I2 + SO2 + 2H2O

1.0

O2

Heat

HTGR

+
H2O

Sulfuric aciddecomposition

2

H2O
Fig.9-5 Thermochemical water-splitting IS process
Hydrogen and oxygen are produced from water by a
combination of three chemical reactions. The process canharness the heat generated by HTGRs.

H+ selectivity (–)

H2

0.8
0.7
0.6
0.5

: Experimental value: Theoretical formula

0.4

20

40

Rich HIIon exchange membrane

t+ =
I

Anode

H+

H+

I

Selective H+ permeation

Cathode

I2

Reduction of I2

I2

60

80

100

120

Temperature (°C)

HI-I2 -H 2 O solution
Electron

We have been investigating a hydrogen production method
known as the iodine-sulfur (IS) process for heat applicationof high-temperature gas-cooled reactors (HTGRs). The IS
process is a water-splitting thermochemical cycle withoutcarbon dioxide emission that proceeds by chemical reactions ofiodine and sulfur compounds (Fig.9-5). One of the important
R&D goals for industrialization of the process is to improve thehydrogen production efficiency.
A task to improve efficiency is to concentrate the circulating
material, hydriodic acid (HI solution, an iodine compound),with low energy consumption; therefore, we have been studying
an electro-electrodialysis method. This method can enrich HI in
the HI solution with a HI concentrator employing ion-exchange
membranes that selectively permeate protons (H+) (Fig.9-6). Thusfar, we successfully prepared novel ion-exchange membranes
using a radiation-induced graft polymerization technique(radiation-grafted membranes), which are a key device for low
energy consumption in the HI concentrator.

D H + nH +

 D: diffusion coefficient of ions
D H + nH + + D I - nI -  n: ions content of membrane

Fig.9-7 Temperature dependence of H+ selectivity: comparisonof experimental results and theoretical formula
Good agreement is observed between the experimental results
and the theoretical formula based on the mathematical model ofpermeation phenomena in the radiation-grafted membrane. Thissuggests that a high-performance membrane can be prepared byadjusting the membrane properties.

Fig.9-6 Hydriodic acid concentration using HI concentrator
Enriched HI solution can be obtained by reducing iodine (I2) andpermeating protons (H+) through the ion-exchange membrane.

We are currently pursuing studies to improve the performanceof the radiation-grafted membrane. For this purpose, wehave tried to reveal the ion permeation mechanism using amathematical model to characterize the performance.
In the model, the driving force of ion movement is assumed
to be the electric potential gradient. A theoretical formula for H+
selectivity, which is an essential performance, was derived asa function of the H+ content, iodide ion content, and diffusion
coefficient. The good agreement between the experimental
results and the theoretical formula (Fig.9-7) shows that themodel we employed is appropriate.
Next, we plan to modify the model regarding the effectsof HI concentration. In addition, we will work on thedevelopment of higher-performance membranes by adjusting
the membrane properties. One way to do this is by varying themembrane’s affinity to HI using grafted polymers to introduce
new functional groups.

Reference
Tanaka, N. et al., Effect of Temperature on Electro-Electrodialysis of HI-I2-H2O Mixture using Ion Exchange Membranes, Journal of Membrane Science,
vols.411-412, 2012, p.99-108.

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JAEA R&D Review 2013

10
10 Development of Decommissioning and Radwaste Treatment Technology
Executing Decommissioning of Nuclear Facilities and Treatment
and Disposal of Radioactive Waste
Support for revising
related regulations
Planning of disposal
facilities

Support for revising
regulations related todisposal of radioactive waste

Design for shallow landdisposal facilities

Waste

Treatment andstorage of radioactive
waste

Operation and maintenance packageof existing waste treatment
facilities (Advanced volume
reduction facilities, etc.)

Waste package
inspection

Construction of waste
treatment facilities at
each site

Clearance

Developmentof relevant
technology

Waste from dismantling

Decommissioning of
nuclear facilities

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

Fig.10-1 Outline of measures for decommissioning and radioactive waste treatment/disposal
We are setting up systems for the purpose of decommissioning nuclear facilities and managing radioactive waste. In addition,
we are developing related technology (e.g., decommissioning, treatment, disposal), planning and constructing radioactive waste
treatment/disposal facilities, and providing support for the revision of related regulations.

Safe and efficient decommissioning of nuclear facilities
and radioactive waste treatment/disposal are two of ourmajor missions in advancing our research and development(R&D) activities. In these missions, we will be disposing ofradioactive waste generated not only by our research activities
but also by universities, institutes, industrial facilities, andother sites.
We are setting up systems for the decommissioning of
nuclear facilities and for managing radioactive waste, and also
developing the related technology (Fig.10-1).
Moreover, we are studying the application of our R&D
products to deal with the accident at the Tokyo Electric Power
Company, Incorporated Fukushima Daiichi Nuclear Power
Station.
R&D for Waste Treatment
In radioactive waste processing and disposal, it is importantto reduce the expense and to improve the safety of disposal.
Liquid waste, including low-level radioactive materials
resulting from reprocessing of spent nuclear fuel, is storedby the Japan Atomic Energy Agency and includes large
amounts of nitrate salts. Some liquid wastes are solidified
with cement, and the solidified cement form is disposed ofin an underground facility. Because nitrate salts are harmful

to human health, the total concentration of nitrate and nitrite
in environmental water and drinking water is required to beless than a standard value. A superior catalyst showing suchproperties as a high nitrate decomposition rate, decompositionof all nitrate ions, little by-product generation, and highdurability should be developed.
The catalyst we are developing was mixed in a sodium
nitrate solution, and the mixture was stirred and maintained
at a set temperature. We could decompose almost all thenitrate ions by slowly delivering hydradine monohydrate as a
reductant dropwise into the mixture (Topic 10-1).
R&D for Waste Disposal
The radioactivity in low-level radioactive wastes generated
during decommissioning of nuclear facilities should be
identified in order to dispose of the wastes.
The applicability of a capillary electrophoresis–laserinduced fluorescent detection method (CE-LIF) in thedevelopment of a rapid analytical method for actinides inradioactive wastes was studied. In this study, several typesof emissive probes with different chelating moieties weresynthesized and applied to CE-LIF. As a result, an acyclic
hexadentate probe achieved successful detection andseparation of americium and neptunium (Topic 10-2).
JAEA R&D Review 201397

Development of Decommissioning and Radwaste Treatment Technology

10-1 Successful Development of Nitrate Decomposition Technology

Environmental water&
Drinking water

Ingress ofground water
Nitratesalt

Nitrate
rate
salt
alt

NitratesaltRadioactive waste

Nitrate and nitrite< 10 mg/ℓ

Decomposition of nitrate ion (h-1)－ For Safe Disposal of Low-Level Radioactive Waste －

500

400

300

200

100

0
0

0.2

0.4

0.6

0.8

1.0

Ratio of Cu in catalyst

Fig.10-2 Effect of buried radioactive waste including nitrate salts
The nitrate and nitrite concentration in environmental water anddrinking water is required by law to be less than 10 mg/ℓ. Whenradioactive waste including large amounts of nitrate salts is buried,
the influence of nitrates leached by contact with groundwater is aconcern. Therefore, a technique for removing nitrates from radioactive
waste is important.

When spent nuclear fuel is reprocessed, liquid waste
including low-level radioactive materials is generated.
The liquid waste includes large amounts of nitrate salts.
Because nitrate salts are harmful to human health, the total
concentration of nitrate and nitrite in environmental waterand drinking water is required to be less than 10 mg/ℓ. Whensolidified liquid wastes including large amounts of nitrate salts
are buried underground, the nitrates in the waste are expected
to be eluted into the environment by contact with groundwater(Fig.10-2). Therefore, it is important to remove the nitratesalts before solidification. In industry in general, especiallyin the field of agriculture, several techniques have beendeveloped for removing nitrate salts at the level of several
hundreds of milligrams per liter in agricultural drainage
contaminated by fertilization. The nitrate in such waste water
is decomposed mainly by using the power of microorganisms
in nature. On the other hand, the concentration of nitrate salts
in radioactive liquid waste is high, about several hundred
thousand milligrams per liter, and a new denitration method
is required because microorganisms cannot survive in suchharsh conditions. Therefore, we developed a technique fordecomposing nitrate by a chemical reaction using a catalyst.
Nitrate ions are reduced to nitrogen using a reductant and a

Fig.10-3 Relationship between copper ratio in
catalyst and nitrate decomposition rate
The effect of the copper ratio in the catalyst on
nitrate decomposition was investigated. A catalyst
with a copper ratio of 0.4 was found to decompose
nitrate much more rapidly.

catalyst made of a palladium (Pd) and copper (Cu) alloy. Pd isa precious metal and is expensive, so the performance of thecatalyst affects the running cost in practical use. Therefore,a superior catalyst showing such properties as a high nitratedecomposition rate, decomposition of all nitrate ions, little
generation of by-products such as nitrogen monoxide andammonia, and high durability should be developed.
Here we developed a catalyst consisting of micropowdersof Pd–Cu alloy supported on the surface of carbon powderand demonstrated its performance. The catalyst was mixed
in a sodium nitrate solution including about 300000 mg/ℓof nitrate ions, and the mixture was stirred and kept at atemperature of 80 °C. We could decompose almost all thenitrate ions by slowly delivering hydradine monohydrate as a
reductant dropwise into the mixture. The relationship between
the decomposition rate and the Cu ratio in the catalyst wasinvestigated (Fig.10-3), and the maximum decomposition rate
was observed when the Cu ratio was around 0.4.
We succeeded in the development of a catalyst that was able
to decompose nitrate ions at a high concentration. We willkeep developing this technology by enhancing the catalyst’s
durability.

Reference
Kadowaki, H., Meguro, Y., Applicability of a Catalytic Reduction Method using a Palladium-Copper Catalyst and Hydrazine for the Denitration of a Highly
Concentrated Nitrate Salt Solution, Journal of Nuclear Science and Technology, vol.49, no.9, 2012, p.881-887.

98

JAEA R&D Review 2013

Development of Decommissioning and Radwaste Treatment Technology

10-2 Simplification of Analysis of Actinides in Radioactive Wastes－ Development of an Analytical Method using Capillary Electrophoresis －
H
N

O

COOH

H
N

N

S

Radiochemical analysis

Radioactivity data collection

Separation buffer
Sample solution

BeforeFused silica capillary(0.05 mm ID)

Inner wall

Emissive probe
for actinide ions
AdditionRadioactivity evaluation
for safe disposal

Emissive moiety

Injection(nL)

Radioactive
sample solution

After

Migration direction(+)

High voltage
10-30 kV(-)

Detection(Electropherogram)

Electrode
SeparationbufferCapillary electrophoresis (CE)Fig.10-4 Development of a simple and rapid analytical method using capillaryelectrophoresis
For the disposal of radioactive wastes, it is necessary to collect radioactivity datafrom radioactive waste samples. To reduce the radiation exposure of operators
during the analytical procedure, a simple and rapid analytical method for actinides
using capillary electrophoresis and new emissive probes was developed.

Chelating moiety

Fig.10-5 Emissive probe L for actinide
ions
Probe L is composed of an emissive
moiety for improved sensitivity and achelating moiety for connecting withactinide ions. Here, a probe with an acyclic
hexadentate chelating moiety is shown.

Fluorescence intensity

Radioactive waste samples

COOH

COOH

OH

Separation by difference in electrophoretic speed
due to ionic charge and size

N

COOH

O

COOH

M+

Np - L

0.035

Am - LFree L
0.025

0.015
6.0

7.5

9.0

Migration time (min)Fig.10-6 Typical electropherogram ofactinide complexes of Am and Np
Am and Np were successfully detected
and separated using probe L. “Am-L”
represents an Am complex of probe L.

For the disposal of radioactive wastes, it is necessary tocollect radioactivity data from radioactive waste samples.
To reduce the radiation exposure of operators during theanalytical procedure, a simple and rapid analytical method of
detecting actinides using capillary electrophoresis (CE) has
been developed (Fig.10-4). Although CE has been employed
for the analysis of ionic species, including actinide ions,and has the advantages of high separation efficiency andsimplicity, there is no report of a high-sensitivity CE method
amenable to practical use for radioactive waste samples, to ourknowledge.
This study focuses on the capillary electrophoresis–
laser-induced fluorescent detection method (CE-LIF) forthe purpose of dramatically improving the sensitivity. New
emissive probes for the actinide ions, americium (Am) and
neptunium (Np), suitable for CE-LIF were developed forthe first time. The emissive probes designed for detecting
actinides by CE-LIF are composed of an emissive moiety for
improved sensitivity, a chelating moiety for connecting withactinide ions, and a spacer between them. A chelating moiety,the more important moiety, is required for stability andprevents dissociation during migration. In this study, seventypes of newly synthesized emissive probes with different
chelating moieties were applied to CE-LIF. As a result, an
acyclic hexadentate probe (Fig.10-5) successfully detected
and separated Am and Np (Fig.10-6). Detection limits of
11 ppt and 4.7 ppt were obtained for Am and Np, respectively.
The sensitivity of this fluorescent detection method was about
90000–200000 times greater than that of a conventional UV
detection method. This method, with its short analytical time(within 10 min), has great potential for application to analysisof radioactive waste samples for the reduction of the radiation
exposure of operators.
Our research was performed as a part of a collaborative
study with Saitama University, “Separation and Detection
System for Actinide Ions by Capillary Electrophoresis Using
Novel Fluorescent Probes.”

Reference
Haraga, T. et al., Capillary Electrophoresis with Laser-Induced Fluorescent Detection Method using Highly Emissive Probes for Analysis of Actinides
in Radioactive Wastes, Proceedings of the ASME 14th International Conference on Environmental Remediation and Radioactive Waste Management(ICEM2011), Reims, France, 2011, ICEM2011-59092, 5p., in CD-ROM.
JAEA R&D Review 201399

11
11 Computational Science and E-Systems Research
Computational Science Activity in Nuclear Energy Research and Development

Sustained performance(TFlops)

Seismic analysis

Nuclear Fusion

Nuclear Safety

Fukushima Reconstruction

Flow around oscillating pipe

R&D on Computational Science

Material degradation･Predicting degradation of
nuclear materials･Predicting physical propertiesof nuclear fuels･Seismic analysis

10

3

102

Physical propertiesof nuclear fuels

101

100
102

K computer
BX900(JAEA)
103

104

105

Processors

Adsorption of radioactive
materials

Speed-up of plasma calculation

R&D on Computing Technology

Operation & Management･Visualization techniques･Parallel computation techniques･Supercomputer･Network system

Fig.11-1 Computational science activity in nuclear energy R&D
The CCSE unites computational science research, computing technology research, and computer operation and management,
and is creating cutting-edge knowledge and data in order to contribute to nuclear energy R&D and Fukushima reconstruction.

Computational science, which is known as the third type of
research methodology after the theoretical and experimental
approaches, is indispensable for understanding and predicting
phenomena that cannot be resolved theoretically because of theircomplexity and/or that are difficult to observe experimentally
because of cost and/or safety. Recently, the Ministry of
Education, Culture, Sports, Science and Technology, which
recognized the importance of computational science, officiallyannounced a plan for developing a new supercomputer with acomputing speed 100 times faster than that of the K computer.
In addition, the news that China had successfully developed thefastest supercomputer in the world astonished the world. Suchunremitting progress in computers is based on a background ofincreasing importance for computational science.
The Center for Computational Science and E-Systems (CCSE)
manages and operates rapidly progressing supercomputersand promotes advanced research and development (R&D) on
computational science using them. It grapples with important
problems in nuclear energy R&D that require complicatedlarge-scale computation, such as material degradation by aging,
the high-temperature behavior of nuclear fuel, and evaluation of
the earthquake resistance of structures. Furthermore, to create
new knowledge about the mechanism of cesium soil pollution,
100

JAEA R&D Review 2013

the CCSE recently began a remarkable acceleration of advanced
R&D by computer simulation (Fig.11-1).
In Topic 11-1, which aims at understanding the degradation
mechanism of nuclear structural materials, the new conceptthat mobile hydrogen promotes crack growth is reported as afinding from a first principles calculation and thermodynamic
analysis. In Topic 11-2, the strange thermal conductivity
behavior exhibited by a new functional material, the topologicalsuperconductor, is introduced as a result obtained by a stateof-the-art computational science method. Topic 11-3 describes
R&D for the international thermal nuclear fusion reactor inwhich the computation speed of the code for predicting plasma
turbulence was materially improved using an innovative
calculation method. In Topic 11-4, the effect of pipe oscillationon coolant flow is estimated as an example of seismic analysis.
In Topics 1-5 and 1-19, the first principles calculation results
for understanding the cesium adsorption mechanism of clay
minerals and zeolites are reported as R&D for Fukushimareconstruction.
To contribute to Fukushima reconstruction and nuclear
energy R&D, the CCSE will develop a state-of-the-art computer
simulation technology as a key to resolving problems and strive
continuously to be the leading force in computational science.

Computational Science and E-Systems Research

11-1 Moving Hydrogen Enhances Cracking in Iron

－ Embrittling Effect of Mobile Hydrogen Clarified by First Principles Calculations －Mobile hydrogen
during fractureH

Grainboundary
(GB)

H

H

H

H

H

Newly generated
trapping site formobile hydrogen

Fracturesurface(FS)

Fracturesurface
(FS)

H

H

H

H

Mobile hydrogen
during fracture

Energy difference
between GB and FSs (J/m2)

5
Immobile hydrogen
during fracture

Immobile hydrogeneffect4
3

Temperature: 300 K
2
1

Limit of realistichydrogen content

010-18 10-16 10-14 10-12 10-10 10-8

Mobile hydrogeneffect

10-6

10-4

10-2

100

Hydrogen content in iron bulk lattice (atomic fraction)

Fig.11-2 Concept of mobile hydrogen
Immobile hydrogen is segregated along grain boundary before
fracture. In contrast, mobile hydrogen is segregated at newly
generated segregation sites on fracture surfaces as they form
during fracturing.

Fig.11-3 Energy difference before and after fracture
The energy difference is calculated between the grain boundary
and its fracture surfaces affected by mobile and immobilehydrogen. In a realistic range of hydrogen content in the ironbulk lattice (<10-6 atomic fraction), mobile hydrogen plays adominant role in embrittlement.

Hydrogen-induced embrittlement of steel has been studied
for many years, but the mechanism is still not well understood.
The wide variety of hydrogen-induced embrittlement implies
that there is more than one mechanism. This study focuseson the hydrogen-induced grain boundary embrittlement ofhigh-strength steels. Metallic materials such as steel generally
consist of many crystal grains, which are roughly several tensof micrometers in size. The boundary between two grains iscalled a grain boundary and has disordered atomic structure.
It is well known that diffusing hydrogen atoms in the ironlattice can be trapped along the grain boundaries, and thenthe aggregation (segregation) of hydrogen atoms causes
cracking. However, the microscopic (atomistic) mechanismof hydrogen-induced grain boundary embrittlement is still
unclear.
We investigated the mechanism by using first principles
electronic structure calculations based on quantum mechanics.
The results showed that hydrogen atoms can be trapped
along grain boundary and fracture surfaces, and then promotecracking along the grain boundaries. Furthermore, wesuggested that moving hydrogen in the iron lattice during

fracturing assisted the formation of fracture surfaces by being
trapped at newly generated atomic sites. This is called themobile hydrogen effect.
A schematic illustration of the embrittling effect of mobilehydrogen is shown in Fig.11-2. Grain boundary embrittlement
is believed to be caused by weakening of atomic bonding
across the grain boundary plane. The strength of the bonding
is characterized well by the energy difference between thegrain boundary and its two fracture surfaces before and after
fracturing. This energy difference can be calculated from firstprinciples. We showed that hydrogen segregation along thegrain boundary decreases its energy difference, indicating thathydrogen causes embrittlement. Furthermore, hydrogen iniron can diffuse very rapidly and thus can be trapped at newly
generated atomic sites on freshly forming fracture surfaces.
This can assist the breaking of atomic bonds at the iron grainboundary. Our first principles calculations with a statisticalmechanics analysis estimated the embrittling effect of mobilehydrogen for the first time and showed that mobile hydrogengreatly enhances the embrittling effect along grain boundaries
in a realistic range of hydrogen content, as shown in Fig.11-3.

Reference
Yamaguchi, M. et al., Mobile Effect of Hydrogen on Intergranular Decohesion of Iron: First-Principles Calculations, Philosophical Magazine, vol.92, no.11,
2012, p.1349-1368. (with Corrigendum, vol.92, no.24, 2012, p.3121-3124).
JAEA R&D Review 2013101

Computational Science and E-Systems Research

11-2 Properties of New Materials with Topology

China

Hong Kong

departing China
departing Hong Kong

Thermal conductivity(a.u.)－ Classification of the Properties in Superconductors by Topology －x-direction
y-direction

600
500
400
300
200
100
0

0

0.1

0.2

0.3

0.4

0.5

Temperature (T/∆0)

Fig.11-4 Non-trivial topological bridge
Mainland China and Hong Kong drive on opposite sides of
the road. The above bridge can naturally connect the two. Thisbridge has a non-trivial topology.

Fig.11-5 Angular dependence of the thermal conductivity
A certain type of topological superconductor has strongly
anisotropic thermal conductivity due to the non-trivial topology.

Solids are characterized by the Metal, Semi-metal, Insulator
in the conductive-order. Recently, the 4th solids where thebulk is insulating but the surface is metal called Topologicalinsulator have been found. These novel materials can beclassified by the mathematics called the Topology so thatmany scientists are attracted. In addition, this group of thematerials has the high thermoelectric power.
Topology is the mathematical study of properties that arepreserved under continuous deformations. For example, thebride shown in Fig.11-4 has the non-trivial topology. The
topology distinguishes the topological insulators from theconventional solids.
Recently, the researchers have been found the fact thatthe topological insulator Bi2Se3 becomes the superconductor
whose transition temperature is 3 K. This kind of thesuperconductors might be novel functional materials, since
superconductivity can transfer the electricity without theenergy lost so that it is promising material as the standalone device workable in the extreme environment wherea human can not enter. However, the physical properties ofthis superconductor called the topological superconductor
have been unknown. Therefore, we investigate the difference
between Bi2Se3 and the conventional superconductors. Then,by focusing on the fact that the electrons are running ultra
fact in Bi2Se3, in terms of the theory of the relativity, we cansuccessfully classify the superconductors by the topology. The
theory of the relativity hunted the non-trivial topology in the
topological superconductors.
In terms of the description about ultra-fast particles in thetheory of the relativity, we construct the theory of the thermal
conductivity in this material. Then, we find that a certain typeof the topological superconductor has the strong anisotropyof the thermal conductivity due to the non-trivial topology
as shown in Fig.11-5. This phenomenon is originated fromthe fact that the twist in the space-time due to the non-trivial
topology makes a certain crystal axis special.
We show that the mathematical study of the topology isimportant even in the superconductors. Novel functional
materials and devices can be constructed in terms of the new
topological materials in the future.

Reference
Nagai, Y. et al., Rotational Isotropy Breaking as Proof for Spin-Polarized Cooper Pairs in the Topological Superconductor CuxBi2Se3, Physical Review B,
vol.86, issue 9, 2012, p.094507-1-094507-5.

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Computational Science and E-Systems Research

11-3 Massively Parallel Fusion Plasma Simulation using 105 Processors

Tcalc +Tcomm /n

comm.
latency

New method

Tcalc

Tcomm

Conventional method

1234 … ntime 1 2 3 4 … nprocessing corescomputation

processing timeis reduced

communication

Sustained performance(TFlops)－ Development of Latency-Hiding Communication Technique －

103

102

101

K computer
100
102

BX900(JAEA)
103

104

105

Processors

Fig.11-6 Scheduling computation and communication
tasks on a processor with n processing cores
In the conventional method, computations are processed
after communications, and most of the processing cores areidle during the communication latency. In the new method,these tasks are simultaneously processed on different
processing cores, and the computing performance isimproved by hiding the communication latency.

Fig.11-7 Computational performance (TFlops = 1012
operations per second) of fusion plasma simulations
on the BX900 (JAEA) and the K computer
The maximum sustained performance of the BX900 was~19 TFlops using 4098 processors with four processing
cores. When the new latency-hiding communication
technique was applied, the K computer achieved ~682
TFlops using 73728 processors with eight processing
cores.

Predicting plasma turbulence is a key issue in estimating theperformance of fusion core plasmas in ITER, which is determinedby turbulent transport phenomena. A calculation model of fusion
plasmas is described in five-dimensional (5D) phase space (threedimensional position × two-dimensional velocity), which isestimated as requiring 3003 × 128 × 32 ~ 1011 grids for existing
devices such as the JT-60U. The development of modern
parallel computers enabled large-scale turbulence simulations of
the 5D problem, and computational plasma turbulence researchhas been significantly advanced. However, ITER is a few
times larger than existing devices, and simulating its turbulent
transport requires an order of magnitude larger computational
resources. Although a next-generation massively parallel
computer such as the K computer is a promising solution, thelatency due to data communications among ~105 processors has
been a bottleneck in improving the computational performanceof massively parallel simulations.
To resolve this issue, a novel communication technique that
dramatically reduces the communication latency is developed.

Turbulence simulations based on grids process calculation
models in parallel by assigning decomposed small domains
to each processor. This computational process, which requires
boundary data in the neighboring domain, is normally performed
after the boundary data is communicated. In the new technique,the computational performance is dramatically improved byassigning different tasks, computation and communication, tomultiple processing cores on a single processor and processing
them simultaneously (Fig.11-6). By applying this technique to73728 processors on the K computer, massively parallel fusion
plasma turbulence simulations were accelerated by ~35 times(Fig.11-7). This speedup enables ITER-size simulations, whichwere estimated to take more than a year on previous machines,
to be completed in a week.
This work was supported by a Ministry of Education, Culture,
Sports, Science and Technology of Japan (MEXT) grant for
the HPCI Strategic Program Field No.4, “Next-Generation
Industrial Innovations.”

Reference
Idomura, Y. et al., Communication-Overlap Techniques for Improved Strong Scaling of Gyrokinetic Eulerian Code beyond 100 k Cores on the K-Computer,
International Journal of High Performance Computing Applications, 2013, doi: 10.1177/1094342013490973.
JAEA R&D Review 2013103

Computational Science and E-Systems Research

11-4 Technology to Speed up Fluid Structure Interaction Simulation－ Simulation Technologies for Flow-Induced Vibration in a Nuclear Power Plant －Vertex Judgment by MC (Front view)Set boundary

Flow

Fluid mesh

Flow

Flow

T=12.44 s

T=12.47 s

Vertices outside ofa structure

Boundary
Vertices ina structure

Structure meshFig.11-8 Setting of boundaries using MC
Using the results of a judgment regarding whether vertices areincluded in the structure mesh, the boundaries in a calculation
domain are set. The processing cost is significantly reduced
because the shape is calculated approximately.

T=12.39 s

Fig.11-10 Time variations in cylinder deformation
Red shows high-stress region; blue shows low-stress region.
The figure indicates that the cylinder deforms because of
the pressure difference around it. The deformation amount isenlarged 100 times.

（a）（b）
Drag coefficient (Cd)

Cylinder

Flow

1.4
1.2
1.0

Experiment Cd = 1.0

0.8
0.6
5

6

7

8

9

10

Time (s)Flow-induced vibration is a very important phenomenon in a
nuclear power plant because it causes breaks in the instrumentsor pipes. To analyze this phenomenon, fluid structureinteraction simulation technologies have received considerable
attention in recent years. In a fluid structure interaction
simulation, a mesh consisting of small polyhedra is generally
used to describe the fluid and structure regions. To date, amethod that describes accurately the shape of the boundary
between the fluid mesh and the structure mesh has been used.
In this method, when the structure shape is highly deformed,
the mesh shape also becomes awkward. Consequently, thesimulation accuracy and speed are reduced. One of the methodsof overcoming this problem is the immersed boundary method.
However, setting the boundary conditions in the calculation
domain requires considerable time.
In this study, a new algorithm that uses Marching Cube (MC)to set the boundaries approximately was invented (Fig.11-8).
The processing speed is more than 20 times faster than that of

Fig.11-9 Pressure and time variations in drag coefficient(a) Pressure distributions. High pressure (red) occurs at the
upstream side, and low pressure (blue) occurs at the downstreamside. (b) Drag coefficient versus time. The simulation result
exhibits good agreement with the experimental data.

traditional methods. This makes it possible to finish a simulationthat requires half a year by traditional methods within 10 days.
Consequently, the method enables us to perform more practicalsimulations.
For verification, the method was applied to a simulation of theflow-induced vibration around a circular cylinder. Figs.11-9(a)and 11-10 show that the simulation can compute the deformationof a cylinder due to a pressure difference around it. The dragcoefficient (Fig.11-9(b)) and amount of deformation of thecylinder are in good agreement with the experimental data. This
study confirmed that the proposed method maintains adequate
accuracy for application to practical problems.
This method enables us to perform large-scale complex
simulations of flow-induced vibration with adequate accuracy
within a practical timeframe. It is hoped that the method willcontribute to the development of technologies to increase thesafety of nuclear power plants.

Reference
Kino, C. et al., Flow Analysis around an In-Line Forced Oscillating Circular Cylinder using IB-Method, Nippon Kikai Gakkai Ronbunshu, B Hen, vol.78,
no.796, 2012, p.2113-2126 (in Japanese).

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12
12 Development of Science & Technology for Nuclear Nonproliferation
Technology and Human Resource Development in the Area of Nuclear Nonproliferation
and Nuclear Security to Support Peaceful Use of Nuclear Energy
The Japan Atomic Energy Agency (JAEA) is conducting
the following technology and human resource development
activities related to nuclear nonproliferation and nuclear security,
in cooperation with relevant domestic and overseas organizations(Fig.12-1).

Support for Government Policy Formulation Based on
Our Expertise
As a think tank in this area, we conduct research on measures
to ensure nuclear nonproliferation and nuclear security on thebackend of the nuclear fuel cycle.

Nuclear Nonproliferation Technology Development forJapanese and International Applications
We have been developing proliferation resistant nuclear
technologies, methodologies for evaluating proliferation resistance,
and advanced safeguard technologies through cooperation withinternational partners such as the U.S. Department of Energy(DOE). In February 2013, we commemorated the twenty-five year
anniversary of cooperation with the DOE in the field of nuclear
nonproliferation. On the occasion of an annual bilateral meeting,
the JAEA was presented with a medallion with a message from
Daniel B. Poneman, Deputy Secretary of Energy, which expresses
his appreciation and gratitude for the JAEA’s cooperation in thisfield (Fig.12-2).

Strict Management of Nuclear Material at Our Own
Facilities and Utilization of the Experience Gained fromManagement of Nuclear Material
We strictly manage the nuclear material that the JAEA
possesses. Moreover, we assist in increasing the efficiency
of inspections by providing technical support to the Japanese
government and the International Atomic Energy Agency (IAEA).
We also provide support to the IAEA in the field of physicalprotection and respond appropriately in the event of revisions to
the domestic legislation in this area.

Development of Measures to Account for and Control
Nuclear Material in Response to the Accident at 1F
We are examining the technologies applicable to thenondestructive measurement of nuclear material that remains
within the molten core of the reactors at Tokyo ElectricPower Company, Incorporated Fukushima Daiichi Nuclear
Power Station (1F). As one such candidate technology, we aredeveloping a technology for measuring nuclear material using
γ-rays emitted by fission products accompanying nuclear material(Chapter 1, Topic 1-16).
Contributions to the International Community Based on
Our Expertise and Experience
With respect to activities relating to the Comprehensive
Nuclear-Test-Ban Treaty (CTBT), we operate radionuclide
monitoring stations and other related facilities, and thus contributeto the establishment of an international monitoring system forthe detection of nuclear tests. The JAEA’s activities in this area
attracted international attention when radionuclides that aresuspected to originate from the nuclear test announced by North
Korea in February 2013 were detected in April at Takasaki
Radionuclide Monitoring Station.

EURATOM

Proliferation-resistant nuclear
technology and methodology toevaluate proliferation resistance
Advanced safeguards technologies

Domestic Human Resource
Development

Management/Protectionof Nuclear MaterialInternational Contribution
Advanced development of CTBT
nuclear test verification technology andestablishment of international
monitoring system
Technical support for disposal of Pu fromdismantled Russian nuclear weapons

Domestic
organizations

Enhancement of capacity building throughtraining and education
Support for infrastructure development,
including the development of a legal
system, regulations and manuals

Cooperative relations with universities
and related organizations
Support for nationwide efforts toward
human resource development

Effective and efficient management
Physical protection of nuclear materials
Cooperation with/support of IAEA, etc.

DOE

Support of Capacity Building andinfrastructure Development in Asian States

Policy Research
Policy research on the current
issue in the peaceful use of
nuclear energy and nuclear
nonproliferation

IAEA

Cooperation

Development of Technology for
Nuclear Nonproliferation

New Efforts toward Nuclear Security
On the basis of commitments made by the Japanese government
at the Nuclear Security Summit in April 2010, we establishedthe Integrated Support Center for Nuclear Nonproliferation and
Nuclear Security (ISCN) in December 2010 and began conducting
its mission of capacity building and infrastructure development,
focusing on the Asian region.
Approximately 630 participants (approximately 430 from Asianstates) participated in the training courses on nuclear security andsafeguards and other activities organized by the ISCN in FY 2012.
We are proud of the contributions that we are making in the areaof capacity building in primarily the Asian region.
Furthermore, we are conducting research on and developmentof the technology on the measurement of nuclear material, which
will contribute to the advancement of accounting for and control
of nuclear material, as well as the technology on nuclear detection
and forensics technology in cooperation with the United States.
We will continue to provide support to the Japanese government
in the area of international contributions by establishing accurate
technologies and sharing these technologies with the international
community.

Development of Technologies
for Measurement, Detection, andForensics of Nuclear Material

Demonstration of NDA system, development of nuclear
resonance fluorescence NDA technology,
development of He3-alternative neutron detection
Establishment and improvement of nuclear forensics
technologies and creation of domestic database

Cooperation
WINS

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

Relevant
ministries

ASNO

Fig.12-2 Medallion to commemorate
twenty-five year anniversary ofcooperation between JAEA-DOE in
the area of nuclear nonproliferation
presented by DOE
Fig.12-1 JAEA activities in development of science andtechnology for nuclear nonproliferation
We have two primary missions: to support the government indeveloping nuclear nonproliferation policies through researchand study, and to support government and internationalorganizations by developing nuclear nonproliferation technology.
In addition, we appropriately implement the management of
nuclear material and the development of related technology
and conduct capacity-building activities.
JAEA R&D Review 2013

105

13
13 Development of Experimental Techniques/Facilities at JAEA R&D Centers
Tsuruga Head Office
We are engaged in checking the unchecked equipment
and improving the safety culture and so on to remedy the
overdue status of the components inspection schedule for“MONJU” and to return to a normal state as soon as possible
from the present status of deviation from the Nuclear Reactor
Regulation Law.
We are also working steadily toward the decommissioningof “FUGEN” and related technical development as well as
selection tests to determine suitable cutting methods for thedismantling required for the decommissioning of the Tokyo
Electric Power Company, Incorporated Fukushima Daiichi
Nuclear Power Station (1F) as a result of the accident.
In addition, technical cooperation, such as promotion ofresearch and development (R&D), including that on laser
technology, by industry–government–academia collaboration
is being advanced substantially. In particular, the industrial
application of lasers to fast reactor heat exchanger tube repair
technology is in progress (Topic 13-1).
The FBR Safety Technology Center was established in April
2013. The center conducts R&D toward commercial deploymentof the fast breeder reactor, mainly in the area of safety studies,such as severe accident analysis and evaluation for “MONJU”.
Tokai Research and Development Center, Nuclear
Science Research Institute (NSRI)
The NSRI facilities, such as research reactors (JRR-3, JRR-4,
and NSRR), accelerators (e.g., Tandem), critical assemblies(e.g., STACY, FCA), and hot laboratories (e.g., WASTEF,
BECKY) were also damaged during the Great East Japan
Earthquake. Because the damaged lifelines were recovered
quickly, hot laboratories were able to begin supporting thenuclear fuel R&D activities, and accelerators started operation
for common use by researchers from JAEA, universities, andother organizations. In addition, the Fukushima Project Team,
which was launched on October 1, performed R&D activity
using these facilities (Chapter 1, Topics 1-12, 1-14, 1-15).
In the recovery of NSRI security, a new building with anearthquake-resistant structure was completed and establishedas the urgent activity base in the event of atomic accidents.
Regarding the use of NSRI technology, NSRI cooperated
with companies in the development of their technologies. For
example, a dosimeter system that can simultaneously measurevarious weather conditions was produced by a manufacturer
using radiation detection and electronics technologies
developed at NSRI.
Tokai Research and Development Center, Nuclear
Fuel Cycle Engineering Laboratories
At the Plutonium Fuel Development Center, basic data on
mixed oxide (MOX) fuels was acquired, and treatment testsof MOX fuel powder were conducted as part of a technicalcollaboration with Japan Nuclear Fuel Limited.
At the Tokai Reprocessing Technology Development
Center, cold tests on the vitrification of high-level radioactive
liquid waste and cementation of low-level radioactive liquid
waste were conducted.
The Waste Management Department developed decommissioning
technologies for old nuclear facilities, planned and built the Tokai
Waste Treatment Facility for treating uranium and trans-uranic
wastes, conducted fundamental and applied studies to increase
the reliability of waste disposal technologies, and developed
safety assessment methods.
At the Radiation Protection Department, the concentrationsof radioactive materials and the ambient dose rate were
measured on and off site, and analytical methods forenvironmental samples were developed (Topic 13-2).
The Department of Fukushima Technology Development
has supported the activity of the Fukushima Project Team andhas conducted R&D for decommissioning of the damaged
Fukushima reactors (Chapter 1, Topics 1-13, 1-20).
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JAEA R&D Review 2013

Ceremony launching the FBR Safety Technology Center(April 2, 2013)

Japan Research Reactor-3 (JRR-3)

Nuclear Fuel Cycle Safety Engineering
Research Facility (NUCEF)

Nuclear Safety Research Reactor (NSRR)

Reactor Fuel Examination Facility

Main research facilities at Nuclear Science Research InstituteSampling of seabed soil with a monitoring boat

Development of Experimental Techniques/Facilities at JAEA R&D Centers

Oarai Research and Development Center
Restoration work was performed and safety measures weretaken for the facilities and equipment damaged by the Great
East Japan Earthquake.
In the Japan Materials Testing Reactor (JMTR), theconfirmation of the integrity of the facilities after the Great East
Japan Earthquake was completed, and the results were reported
to the regulatory agency with the goal of re-operation. A lightwater reactor water environment demonstration test facility was
fabricated and installed with funding from the Leading-edge
Research Promotion Fund, which was provided by the Ministryof Education, Culture, Sports, Science and Technology (MEXT)
in Japan. A real-time system for monitoring radiation workers’
locations, health conditions, and exposed doses using portableradio devices was developed and put into operation. To enable
domestic 99Mo production, a technique for fabricating a highdensity pellet was developed (Topic 13-6).
In the Experimental Fast Reactor “JOYO,” equipment for
use in repair work was designed and fabricated. A mockup test using those pieces of equipment was started. The
compatibility of a zirconium alloy with high-temperature
sodium was examined (Topic 13-7).
In the High Temperature engineering Test Reactor (HTTR),
an inspection of the facilities and a comprehensive evaluationof the integrity of the equipment were conducted with the goalof resuming operation. A damage detection method for theneutron detector was established (Topic 13-8).
Naka Fusion Institute
The Naka Fusion Institute conducts R&D toward thepractical use of fusion energy.
Currently, it is involved mainly in the procurement anddevelopment of the JA-shared apparatuses as a domestic
agency of the ITER project and in the upgrade of the large
tokamak device JT-60 to the satellite tokamak device JT-60SA,
which will be used for research supporting and complementing
ITER as Broader Approach activities in cooperation with theEuropean Union.
In 2012, the production of the superconducting conductor for
ITER was advanced, and the production of the superconducting
coil was begun. Moreover, the disassembly of the main bodyof JT-60 was completed; the cryostat base, which was procured
from Europe, was accepted, and the assembly of JT-60SA began.
The production of JT-60SA components was proceeded, andthe smallest superconducting poloidal magnetic field coil (EF4)
was completed. In addition, three 40° sectors of the vacuum
vessel for the actual machine were welded, and a total of six
were completed.

6.5×1013 n/pulse
300

Beam power (kW)J-PARC CenterIn fiscal year 2012 (FY2012), J-PARC provided neutron
beams to users for 8 run cycles (176 d). We successfully
achieved steady and reliable operation at the target beam
power of 200 kW, which was one of our goals for FY2012, bySeptember, and also increased the power to 300 kW in Octoberat the Materials and Life Science Facility (MLF). As a result of
this power increase at the MLF the world’s strongest neutron
intensity of 6.5×1013 neutrons per pulse, was generated. Thisintense neutrons enabled many experimental outcomes, suchas the development of high-performance structural materials(Topic 13-3) and the development of high-performance
elliptical super-mirror (Topic 13-4). In addition, the MLF muonsource produced 250×106 muons per pulse, which exceeds
3×104 muons at a similar facility in the UK and became world’s
record. A total of 526 proposals were submitted to the MLF
user programs which was much more than in any year before
the Great East Japan Earthquake. Other major progresses on
the dominantly JAEA projects at J-PARC is as follows: (1) the
new neutron instrument, the Energy Resolved Neutron
Imaging System, which aims to measure elements, magnetic
fields and crystalline structure in matter, is under construction
at the MLF; and (2) Acceleration cavities with an annular-ring
coupled structure were fabricated (Topic 13-5).

Target power for 2012
200

100

0

2009

2010

2011

2012

2013

Time (Year)
Operational history of J-PARC

Radio portable devices

Worker’ s position

Location detector tag

Wireless dosimeter

Worker’ s health condition and exposed dose

Development of real-time personal exposure and healthcondition monitoring system in the JMTR
The system can monitor workers’ locations, health conditions,and exposed radiation doses in real time using wirelessportable devices.

Celebration of the delivery of the first component from the EU
and the beginning JT-60SA tokamak assembly (March 25, 2013)

JAEA R&D Review 2013107

Development of Experimental Techniques/Facilities at JAEA R&D Centers

Takasaki Advanced Radiation Research InstituteTakasaki Ion Accelerators for Advanced Radiation
Application (TIARA) consisting of four ion accelerators,
an electron accelerator, and gamma irradiation facilities are
made available to researchers in JAEA and other organizations
for R&D activities on new functional and environmentallyfriendly materials, biotechnology, the radiation effects ofmaterials, and quantum beam analysis. Practical technology
development activities that are currently in progress involve
microbeams, single ion hits, techniques for uniform wide-area
irradiation at the cyclotron, three-dimensional in-air PIXE
analysis, and a three-dimensional microfabrication technique(Topic 13-9) at the electrostatic accelerators. In 2012, a beam
shape control technique was developed that transforms acomplicated transverse intensity distribution into a Gaussianlike one through multiple scattering using a thin metallic foil.
A 70 mm × 40 mm beam with ±10% uniformity was obtainedby focusing the transformed beam with multipole magnets.

Kansai Photon Science InstituteIn the Kizu District, we are in the process of improving
the laser facilities, e.g., improving the quality of the highintensity lasers. Regarding high-intensity short-pulse lasers,
we worked on improving the position stability of the beamand, by exchanging the mirror mount and reducing the
wobbling in the position, improved the reproducibility ofexperimental results. In addition, in the Consortium for Photon
Science and Technology, we are in the process of developing
ultrawideband light sources ranging from terahertz to X-rays
to quantum beams.
In the Harima District, we have been developing andimproving a state-of-the-art analysis technique for thefunctional expression mechanism and reaction mechanismof materials using the JAEA synchrotron radiation beamlines
at SPring-8. They are also applied to studies related tonanotechnology, energy, and the environment, for instance, adecontamination technique for revitalization in Fukushima.
Further, under the Nanotechnology Platform Project entrusted
to us by MEXT, external researchers are being supported.

Horonobe Underground Research CenterWe are currently constructing the Horonobe UndergroundResearch Laboratory for conducting geoscientific researchand R&D on geological disposal technologies targeted at
sedimentary rocks.
In FY2012, the west shaft was excavated to depths ranging
from 50 m to 350 m (the east and ventilation shafts alreadyreached 350 m in depth in the last fiscal year). In addition, the350 m research gallery passed through the ventilation shaft
to join the east shaft. About 400 m of the 700-m-long gallery
have been completed.
In geoscientific research, we have continuously developed
techniques for investigating the geological environment andfor deep underground engineering, and studied the long-term
stability of the geological environment. On the other hand, in
our R&D on geological disposal, we investigated the impactof low-alkaline cement materials, including high fly-ash andsilica fume content cement, to the surrounding groundwaterand rock mass. In addition, the use of materials in spraying
techniques was tested in situ during the construction of the
350 m research gallery.
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JAEA R&D Review 2013

A container of 22 radiochromic dosimetry films for measuring
the size and uniformity of a beam, installed in the chamber
for formation and irradiation of a large-area uniform beam

JAEA Kansai Advanced Relativistic ENgineering (J-KAREN)
laser system

Penetration of 350 m research gallery between the ventilationand east shaftsDevelopment of Experimental Techniques/Facilities at JAEA R&D Centers

Tono Geoscience Center (TGC)
The TGC’s task is to provide the scientific and technicalfoundations for developing a safe technique for the geologicaldisposal of high-level radioactive waste. This involves research
into the long-term stability of the geological environment,
the development and improvement of techniques for thecharacterization of the deep geological environment, and awide range of engineering techniques for deep undergroundapplications at an underground research laboratory in crystalline
rock, referred to as the Mizunami Underground ResearchLaboratory. Approximately 150 m of horizontal excavations(Sub Stage, Access/Research Galleries-North and South) werecompleted for the GL −500 m stage (meters below groundlevel). While planning the construction, it was necessary
to obtain reliable preliminary information on the bedrock
conditions in terms of the rock mass stability and hydrogeology.
Therefore, site characterization borehole investigations wereconducted before excavation began (Topic 13-10). Researchand development activities, such as geological mapping during
excavation and borehole investigations, were also conducted.
Ningyo-toge Environmental Engineering Center
The Ningyo-toge Environmental Engineering Center has
advanced the environmental remediation of the Ningyotoge and Togo Uranium Mines after decades of mine-related
activities, including uranium exploration and mining, were
terminated. The main purposes of the remediation are to takemeasures to ensure the safety and protection from radiation
via exposure pathways to humans in the future, and to preventthe occurrence of mining pollution.
As part of the remediation, the upstream part of the Yotsugi
Mill Tailings Pond, the highestprioritized facility among allof the mine-related facilities, was remediated to FY2012. In
the remediation, a multi-layered capping was constructedon the ground surface after the specifications and the entire
remediation procedure were determined. Monitoring data
will be accumulated to confirm the effectiveness of theremediation, and the results will be used for remediation of
the downstream part of the Pond.
The Center also performed the required maintenance ofrelated facilities and promoted technology development.
In addition, a sense of security among local residents andactive use of local features (Topic 13-11, “Investigation of
Properties of Radon Hot Springs”) have been sought as anotherindispensable aspect of the Center’s activities.
Aomori Research and Development CenterIn the Rokkasho District, R&D toward developing thefusion DEMO reactor is in progress, for example, a fusion
simulation using a supercomputer that has the second-best
performance in Japan, and the development of the fusion
reactor structural material using various material analysis
and testing devices. In addition, the injector for the prototype
accelerator of the International Fusion Materials Irradiation
Facility (IFMIF) was delivered, and commissioning is planned
to begin this autumn.
In the Mutsu District, a research survey of reasonable andeconomical large assembly dismantling methodologies for theoperation of a waste disposal plant for research facilities, anddecommissioning of the nuclear-powered ship Mutsu (e.g.,surveying the contaminant content and developing technology
to analyze ultra-trace elements using accelerator mass
spectrometry) are conducted continuously.

Breaking through at part of the GL −500 m Sub Stage(July 30, 2012)

Before remediation

After remediation

Upstream part of the Yotsugi Mill Tailings Pond

Delivery of the injector for the prototype accelerator to the
IFMIF/EVEDA (Engineering Validation and Engineering Design
Activities) accelerator building
JAEA R&D Review 2013109

Development of Experimental Techniques/Facilities at JAEA R&D Centers

Tsuruga Head Office13-1 Challenge of Industrial Application of Repair Technologies for FBR Heat Exchanger Tubes－ Success in High-Accuracy Wire Feeding to Laser-Irradiated Spot －(b)Heat exchanger tube

Laser torch

Laser-irradiated spot

23 mm

Vis-NIR mirror

(a)

Wire

Focusing lens
Ar gas1 mmWire feeder

Wire

Rotation stage

Coupling
deviceHeat source(QCW laser)

Laser
cladding

Laser-irradiated spot

Composite-type optical fiber

Vertical linear stage

23 mm
Before laser cladding

Artificial
thinning
After laser cladding

Fig.13-1 Schematic view of laser cladding system in heatexchanger tube
This system was designed to work around the edge of a
23 mm tube. A laser beam was focused on the tube wall by
optical lenses and a vis-near-infrared (NIR) heatproof mirror.
A wire and Ar gas were supplied to the laser spot. QCW:
quasi continuous wave.

Fig.13-2 Laser cladding on wall of 1 in. tube(a) Wire moves toward the center of the laser spot. (b) Line
cladding deposited on wall was about 1 mm in width and 0.2 mm
in height. It is possible to suppress wall thinning by using erosion–
corrosion-resistant wire.

We developed an inspection and repair system for use ina heat exchanger tube in a fast breeder reactor (FBR). The
key technology in this system was a composite-type opticalfiber consisting of a central fiber to deliver the laser beam andsurrounding fibers to transmit the visible image. We could
observe a crack in the tube wall in the visible image and repair
it by laser welding.
Inspection and repair technology for use in a limited tubularspace is expected to have wide industrial applications outsideof the field of nuclear energy. We are promoting the technicaldevelopment of a method of repairing wall thinning in heatexchanger tubes in a petrochemical plant. Wall thinningappears as a dimple on the tube wall and is repaired bymelting and bonding wires to the base metal. Arc welding
is typically used for repairing metal parts. However, in thelimited tubular space, the excess heat of the arc often breaks
the welding devices and causes inferior welding. A laser beam
can be delivered by an optical fiber and concentrate the energy
in a local spot, so it is suitable for use in a narrow space.
Precise wire feeding to a small laser-irradiated spot is very

important for successful laser cladding. A composite-type
optical fiberscope is useful for this purpose. It is necessary
to use a welding wire thinner than the typical one because anallowable bending radius is about 10 mm in a 23 mm tube We
developed a new wire feeder in cooperation with an excellent
machining company in Fukui (Fig.13-1). This device cancontinuously feed 0.4 mm wire by adjusting the roller gripping
force and varying the feeding speed. Its size is about 10 cm3.
The developed wire feeder, laser torch, and composite-type
optical fiber were assembled (Fig.13-1), and the laser cladding
in a 23 mm carbon steel tube was examined using this system.
As a result, we succeeded in continuously feeding the wire to
the laser-irradiated spot and producing a line cladding layer that
was about 1 mm in width and 0.2 mm in height (Fig.13-2).
In the future, we will control this laser cladding system
automatically so that it will fit inside a tube for practical use
in petrochemical plants. In this experiment, the wire wascomparable in hardness to the base metal. We will try to use
a wire that has high erosion–corrosion resistance to suppress
wall thinning.

Reference
Nishimura, A., Terada, T. et al., Instrumentation Device and Surface Control Technology for Coolant Piping System of Nuclear Power Plants, Proceedingsof the 2012 20th International Conference on Nuclear Engineering Collocated with the ASME 2012 Power Conference (ICONE20), California, USA, 2012,ICONE20-POWER2012-54406, 5p., in DVD-ROM.

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Nuclear Fuel Cycle Engineering Laboratories

13-2 Radiocesium in the Environment－ Radiation Situation in Tokai Two Years after the 1F Accident －

(a)

(c)
Cs concentration (one-month average)
in airborne dust (Bq/m3)

30

10000
3100 nGy/h

25

1000

20
15

(32〜45) nGy/h

10
13

14

15 16

17 18

ST1
ST3
ST5
P2
P4
P6
P8

Rainfall
ST2
ST4
P1
P3
P5
P7

100

19 20

21

22

23 24

25 26

10
5

137

2100 nGy/h

Rainfall (mm)Dose rate (one-hour average)(nGy/h)4800 nGy/h010
On site (MP7)
Off site (ST2;Tokai-mura)

1

0.1

0.01

0.001

0.0001

Detection limit
0.0001 Bq/m3

Jan-2011

Jul-2011

Jan-2012

Jul-2012

Jan-2013

Date in March, 2011

(d)

(b)

1000
540 nGy/hOn site (MP7)
Off site (ST2;Tokai-mura)

340 nGy/h(May)

One-month average in March, 2013

190 nGy/h100

Cs concentration in foodstuff(Bq/kg fresh)

One-month average in March, 2011

160 nGy/h

99 nGy/h(May)59 nGy/hDose rate before Fukushima accident 30-40 nGy/h

Fish
100

Shellfish

limit in foods

10

1

0.1
Detection limit : 0.08 Bq/kg (vegetable)
0.04 Bq/kg (seafood)
0.01

10
Jan-2011

SeaweedLeafy vegetable

137

Dose rate (one-month average)(nGy/h)1000

Jul-2011

Jan-2012

Jul-2012

Apr-2010

Jan-2013

Apr-2011

Apr-2012

Apr-2013

Fig.13-3 Chronological change in dose rate and 137Cs concentration observed in Tokai for two years around 1F accident

Because of the release of radioactive materials during theaccident at the TEPCO’s Fukushima Daiichi NPS (1F), an
increase in the dose rate and radiocesium concentration wereobserved in the Nuclear Fuel Cycle Engineering Laboratories(Tokai, Ibaraki Prefecture), which is located about 115 km
from 1F, after March 15, 2011. This was reported by the Japan
Atomic Energy Agency (JAEA-Review 2011-035). The dose
rates (1-h averages) just after the 1F accident are presented
in Fig.13-3(a). We report the chronological change in theradiation in Tokai for two years after the 1F accident.
The dose rates (one-month averages) are presented
in Fig.13-3(b) for monitoring post-7 (MP7; on site) andmonitoring station-2 (ST2; off site) as representative locations.
Because of the differences in the environments (e.g., thepresence or absence of trees), the observed dose rates
varied by a factor of two or three. Because the half-lives ofcesium-134 (134Cs) and cesium-137 (137Cs) are about 2 and30 years, respectively, they contributed most of the dose
rate. From May 2011, when the contribution from shortlife nuclides almost disappeared, to March 2013, the dose
rate decreased by around 60%. Thus, 25% of the 134Cs and137

Cs were transported elsewhere or shielded by immersion,in addition to the decrease by radiological decay (at a rate of
35% in 22 months).
The 137Cs concentrations in airborne dust (one-month averages)
also tended to decline (Fig.13-3(c)). The concentration off site
is likely to be slightly higher than that on site. The origin ofrecent 137Cs in airborne dust is believed to be not 1F itself but
soil on which 137Cs was deposited two years ago. Outside of
the Laboratories, strong winds could resuspend and transportsoil particles from cultivated fields.
The 137Cs concentrations in foodstuffs are presented in
Fig.13-3(d). Taking into account 134Cs, the concentrationof 134Cs plus 137Cs was less than half of the limit in foods(100 Bq/kg fresh for 134Cs plus 137Cs, 500 Bq/kg fresh before
April 2012). The 137Cs concentration generally decreased,
but the concentration in demersal fish, which could be
influenced by 137Cs in seabed sediment, did not decrease as
much. Because the transfer of Cs from the soil to vegetables
is smaller than that from sea water to seafood, the 137Csconcentration in vegetables was less than 1 Bq/kg fresh.

Reference
Fujita, H. et al., Special Environmental Monitoring around Tokai-Mura after the Accident of the Fukushima Dai-Ichi Nuclear Power Station, Proceedings ofthe 13th International Congress of the International Radiation Protection Association (IRPA-13), Glasgow, Scotland, U.K., 2012, P12.31, 7p.
JAEA R&D Review 2013111

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J-PARC Center13-3 Neutrons in Development of High-Performance Materials－ Role of Deformation-Induced Transformation in TRIP Steels Revealed by Neutrons at J-PARC －high carbon concentrated
retained austenite (A)ferrite (F)(matrix)

plastic deformation

0.2%C-TRIP steel
0.4%C-TRIP steel

3000

applied stressapplied stress

2500

1.2
0.25
1.0
0.20

Rel. vol. frac. of austenite
0.2%C-TRIP steel
0.4%C-TRIP steel

0.8

0.15

0.6

0.10

Applied S-S curve
0.2%C-TRIP steel
0.4%C-TRIP steel

0.4
0.2

0.05

0.0

Macroscopic true strain

Relative volume fraction of retained austenite

Fig.13-4 Illustration of TRIP effect
The M phase is formed during plastic deformation, and thematerial’s strength and elongation are increased.

0.00
0

200

400

600

800

1000

1200

Applied true stress (MPa)Fig.13-5 Changes in relative volume fraction of retained
austenite and stress–strain curve of TRIP steels
Phase ratio does not change during deformation in common
steels, whereas it varies in TRIP steels.

Transformation-induced plasticity (TRIP) is an important
effect in steel that improves the strength and ductility andrealizes excellent high-speed deformation behavior. Steels with
0.2%C-TRIP and 0.4%C-TRIP are known to have excellent
shock absorption in collisions and are expected to be usedin car bodies. The TRIP effect, as shown in Fig.13-4, occurs
in steel having a metastable phase (the A phase) that cantransform into a stronger phase (the M phase) during plastic
deformation. However, there is almost no quantitative researchon the behavior causing the TRIP effect that directly shows
the contribution of the M phase to the strength. We attempt tounderstand this behavior using neutron diffraction.
We performed in-situ neutron diffraction experiments during
tensile loading at room temperature using the Engineering
Materials Diffractometer TAKUMI at the Materials and Life
Science Experimental Facility of J-PARC. TAKUMI can

Phase stress (MPa)martensite (M) formed by transformation

martensite (M)

2000

1500

austenite (A)

1000

500

ferrite (F)

0
0

200

400

600

800

1000

Fig.13-6 Stress partitioning among constituent phases of
TRIP steel during deformation
The stresses were calculated from the phase lattice strainsobtained from an in-situ neutron diffraction study during
tensile deformation. We successfully separated the diffraction
peaks to elucidate the stress partitioning between the M andF phases, which was not possible using the conventional
instruments of ISIS and JRR-3 with low d-spacing resolution.

perform an in-situ neutron diffraction experiment during
deformation without interruption of the load or displacement
because of its high intensity and instantaneous data collection
system. The relative volume fractions of the A phase obtained
from neutron diffraction, as shown in Fig.13-5, are gradually
decreased with increasing applied stress after the onset ofplastic deformation, showing that a deformation-induced
martensitic transformation occurs. Fig.13-6 shows the phasestresses that are shared by the constituent phases in the TRIP
steels during tensile deformation. The M phase stresses arefound to be the highest and the ferrite phase stresses are thelowest among the constituent phases. Moreover, the bulk
stresses that are estimated using these phase stresses andthe volume fractions show good agreement with the appliedstresses.

Reference
Harjo, S. et al., TRIP Steel Deformation Behavior by Neutron Diffraction, Materials Research Society Symposium Proceedings, vol.1528, 2013, 7p.

112

JAEA R&D Review 2013

1200

Applied true stress (MPa)Development of Experimental Techniques/Facilities at JAEA R&D Centers

J-PARC Center

13-4 Development of Ultraprecise Focusing Supermirrors with Elliptical Surface－ Focusing Gain of 52 Achieved for Advanced Instrumentation at J-PARC －(a) unfocused direct beam(b) focused beam

A’

5 mm

0

5

100

400 mm
B

50

B’

lnset

5 mm

0

0

A

100

50

FWHM = 0.128 mm
Gain = 520BA’

2012/2/9

Neutron intensity (a.u.)A

Neutron intensity (a.u.)

100

B’

2012/2/9

1

Fig.13-7 Neutron focusing images obtained by large-scale elliptical supermirror ((a): unfocused direct beam; (b): focused beam)
The FWHM of the focused beam was 0.128 mm, and a focusing gain of 52 was observed using an imaging plate. Inset photographshows elliptical NiC/Ti supermirror with m = 4 coated on 400-mm-long substrate.

Slow neutrons, which exhibit quantum mechanical wave
properties, are used to investigate crystal and molecular
structures and dynamics in solid-state physics, polymer andbiological sciences, and so on.
Neutron focusing devices play the important roles offocusing neutrons on samples and increasing the available
neutron intensities or improving the experimental resolutions
in various experiments such as small-sample diffraction andsmall-angle scattering. Therefore, high-performance focusing
mirrors have been aggressively developed to accelerate
the exploration of new scientific phenomena at neutron
experimental facilities worldwide, such as J-PARC.
We successfully developed focusing supermirrors with alarge critical angle, high reflectivity, and low diffuse scatteringby using ion beam sputtering (IBS). A precise surface figure
was obtained on the focusing mirror by using the numericallycontrolled local wet etching (NC-LWE) on the mirror
substrate. The NC-LWE technique enables an asphericalshape to be figured deterministically with sub-micrometer
form accuracy because it uses a noncontact chemical removal
process that is insensitive to external disturbances such as
vibration and/or thermal deformation.

We designed and fabricated one-dimensional neutron
focusing with a large beam divergence by a 400-mm-longelliptical supermirror. A NiC/Ti supermirror with 1200 layers
was deposited on the substrate using the IBS instrument.
We successfully obtained a figure error of 0.43 μm peak-tovalley with a surface roughness of less than 0.2 nm root meansquare. The neutron reflectivity was measured using the BL17
SHARAKU instrument at J-PARC/MLF. The reflectivity ofthe m = 4 supermirror was ≃0.6.
The focusing performance of the fabricated ellipticalsupermirror was evaluated at BL10 NOBORU. The wavelengthof the neutron beam was varied from 3.5 to 10 Å. The upstream
width of the slit in this experiment was 0.10 mm.
Fig.13-7 shows neutron profiles of the unfocused direct
beam and the focused beam. The full width at half-maximum(FWHM) of the focused beam was 0.128 mm, and a focusing
gain of 52 in terms of the peak intensity was achieved
compared with the unfocused direct beam. Note that the tail ofthe focused beam is no larger than that of the unfocused beamprofile, which suggests that the focusing mirror produces little
diffuse scattering.

Reference
Nagano, M., Yamazaki, D. et al., One-Dimensional Neutron Focusing with Large Beam Divergence by 400mm-Long Elliptical Supermirror, Journal of
Physics: Conference Series, vol.340, 2012, p.012034-1-012034-6.
JAEA R&D Review 2013113

Development of Experimental Techniques/Facilities at JAEA R&D Centers

J-PARC Center

13-5 Beam Acceleration by Axially Symmetric Field－ Development of J-PARC Annular-Ring Coupled Structure －Coupling cell

Vacuum port

Beam axis

Beam axis
Electric field

Fig.13-8 Side-coupled structure (SCS, left) vs. annular-ring
coupled structure (ACS, right)
The coupling cell of the ACS surrounds the accelerating cell
circularly. Compared to the SCS, the ACS has the advantage ofaxial symmetry around the beam axis.

(a)

700
Design pulse width 600 µs

6600

Designaccelerating field
4.1 MV/m

5

500

4

400

3

300

2

200

1

100

0

0

5

10

15

20

25

30

35

40

0

Conditioning time (h)Fig.13-10 Power test result of ACS
The present ACS module was successfully conditioned up to an
accelerating field of 5.3 MV/m (30% higher than the designed
field of 4.1 MV/m), a pulse width of 600 μs, and a repetition rateof 50 Hz. This corresponds to an input power of 600 kW.

(b)

HAccelerating field (MV/m)

7

Accelerating cell

Pulse width (µs)

Coupling slot

H
Beam
E

Beam
E

E: Electric field
H: Magnetic field

The J-PARC linac is a 181 MeV linear accelerator that
injects protons generated by an ion source to the Rapid Cycling
Synchrotron (RCS) with a repetition rate of 25 Hz. The goalof J-PARC is to achieve a 1 MW output beam power, which
would be the largest worldwide, for the Materials and Life
Science Experimental Facility.
Minimizing the beam loss to maintain machine activation
within the permissible level is one of the most important issues
for high-intensity proton linacs, including J-PARC. The spacecharge effect arising from Coulomb repulsive forces among
beam particles is one of the causes of beam loss in the RCS.
This effect can be reduced by increasing the injection energyof the proton beam. Thus, an injection energy upgrade of the
J-PARC linac from 181 to 400 MeV is planned. An annularring coupled structure (ACS) has been developed for thisenergy upgrade (Fig.13-8).
Because of the axially asymmetric structure of the ACS, it

Fig.13-9 Schematic comparison of accelerating field of
SCS (left) and ACS (right)
The accelerating field of the SCS contains a transverse fieldcomponent of 1% due to the configuration of the couplingcell. In contrast, this component of the ACS is negligibly
small because the coupling cell is axially symmetric.

has a negligibly small transverse accelerating field component,
which kicks the proton beam perpendicularly to the beamaxis and is smaller in the ACS than in the side-coupledstructure (SCS) (Fig.13-9). Proton linacs use several types ofaccelerating structure depending on the beam energy. The ACS
realizes energy efficiency and stability comparable to those of
the SCS, which has been used for the same energy region as
the ACS. Consequently, the injection energy can be increased
while reducing the beam loss of the linac.
A prototype module for the ACS for the J-PARC linac wasdesigned and fabricated, and it was successfully conditioned up
to the designed accelerating field (Fig.13-10). This is considered
to be a big step toward the energy upgrade of the linac andrealization of 1 MW beam operation. A total of 25 ACS
modules have been completed to date and are being prepared
for installation. In FY2013, we will finish the installation andstart the world’s first beam acceleration by an ACS.

Reference
Ao, H. et al., First High-Power Model of the Annular-Ring Coupled Structure for Use in the Japan Proton Accelerator Research Complex Linac, PhysicalReview Special Topics - Accelerators and Beams, vol.15, issue 1, 2012, p.011001-1-011001-13.

114

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Oarai Research and Development Center

13-6 Development of Domestic 99Mo Production by (n, ) Method－ Fabrication Technology for High-Density MoO3 Pellet －Pressure㺃Pulse current

Graphite punch

Graphite mold
Thermocouple

Molybdenum trioxide
powderGraphitespacerElectrode (–)

Sintering density (%T.D.)

Electrode (+)
Graphitespacer100
95

Atmosphere: Air
Pellet size:φ20 × 10 mm

90

1 μm

85
1 μm

80
75
70

1 μm

400

450

500

550

600

Sintering temperature (℃)

Pressure
Fig.13-11 Schematic diagram of spark plasma sintering
methodStarting powder is installed in graphite mold. Sintering andmolding are simultaneously performed by pulse electricity
under pressure. During this operation, powder surfaces
are cleaned and activated, and local temperature increase
produces high-density pellet under lower temperature.

Fig.13-12 Relationship between sintering temperature andsintering density of MoO3 pellet
The sintering density increases with increasing sintering
temperature. The target density (90%T.D., T.D.: theoreticaldensity) was attained by sintering at about 500 °C. On theother hand, changes in grain growth were not observed below
a sintering temperature of 550 °C.

In Japan, about 1400000 cases are diagnosed using nuclear
medicine every year. About 900000 of these nuclear medicine
diagnoses use technetium-99m (99mTc), whose half-life is 6 h.
99mTc is produced as a daughter nuclide from molybdenum-99( 99Mo), whose half-life is 66 h. The quantity of 99mTc
demanded in Japan is ranked as second in the world, after
only the U.S.A., but all of the 99mTc is imported from abroad.
A stable supply of 99Mo in Japan might be difficult to obtainbecause of troubles with transportation and other problems.
Therefore, the establishment of domestic production of 99Mo
is an important issue for Japan. Two methods are used to
produce 99Mo by nuclear reactors, the nuclear fission method
and the neutron activation method [(n, γ) method]. The Japan
Materials Testing Reactor (JMTR) is developing the (n, γ)
method for 99Mo production from the viewpoints of physicalprotection of nuclear materials and the reduction of highly
radioactive waste.
99 Mo is produced in molybdenum trioxide (MoO ,3
sublimation temperature: 795 °C) by the (n, γ) method in
nuclear reactors. Next, irradiated MoO 3 is resolved into
6M sodium hydroxide (NaOH), and then 99mTc is extracted
from the solution. The technical issues in this method are
the establishment of a high production rate of 99Mo andthe extraction of 99mTc with a high purity and radioactivity
concentration. For that purpose, a fabrication method forhigh-density MoO3 pellets was also developed in this study toincrease the amount of 98Mo per unit volume and to optimize

the characteristics such as solubility.
As the first step, fabrication tests of high-density MoO3
pellets were conducted using the spark plasma sintering
method, as shown in Fig.13-11.
This method results in a high-purity pellet because no binder
is added, and the sintering temperature is low. The pellet
properties were measured for various sintering temperaturesand sintering densities. In this result, the target density wasattained by sintering at a temperature above 500 °C, as shownin Fig.13-12. Scanning electron microscopy observation of
the MoO3 pellet revealed that grain growth did not occur
below a sintering temperature of 550 °C, and the grain sizeof the MoO3 pellet was almost the same as that of the starting
powder. Next, MoO3 pellets were oxidized in atmosphere asan oxidization treatment, and the solubility into 6M NaOH
was measured. The result confirmed that the MoO3 pellet
dissolved within the target time, and the obtained solution was
clear and exhibited high purity.
The above results demonstrate that a technique forfabricating high-density MoO3 pellets from an irradiated target
was established, and the prospects are bright for domestic
99Mo production by the (n,γ) method.
The present study was performed as part of the program“Technology development of domestic 99Mo/99mTc productionby (n, γ) method in JMTR” commissioned by the Ministry of
Education, Culture, Sports, Science and Technology of Japan(MEXT).

Reference
Nishikata, K. et al., Fabrication and Characterization of High-Density MoO3 Pellets, Proceedings of the 2012 Powder Metallurgy World Congress &Exhibition (PM 2012), Yokohama, Japan, 2012, 8p., in CD-ROM.
JAEA R&D Review 2013115

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Oarai Research and Development Center

13-7 Corrosion Behavior of Metals in Sodium－ Compatibility of Zirconium Alloy with High-Temperature Sodium －
(a)

(b)

as-received

Fig.13-13 Zr alloys exposed to sodium at 500 °C for 1000 h
The surface of the specimen exposed to static sodium (b) was
the same as that of the as-received specimen (a). On the other
hand, the specimen immersed in flowing sodium (c) became
black because dissolved oxygen in sodium penetrated.

(c)in static Na

in flowing Na

200

10 ｰ⁴

Tensile strength (MPa)650 ℃, in flowing Na
650 ℃, in static Na
500 ℃, in static Na

10 ｰ⁵

10 ｰ⁶
10 ｰ⁷

650 ℃, in static Na
650 ℃, in flowing Na

150

100

100

50

50
0

1000

2000

3000

Time (h)

150

0

1000

2000

Total elongation (%)

10 ｰ³(g/mm²)

Weight gain per unit surface

10 mm

0
3000

Time (h)Fig.13-14 Weight change of Zr alloys in high-temperature
sodiumThe weight of the specimens increased in proportion to the
1/2 power of the elapsed time by oxygen diffusion into themetals. The amount of weight gain increased with increasing
temperature.

Fig.13-15 Tensile properties of Zr alloys after sodium
exposure
The data for “time = 0” show the tensile properties of the asreceived material. It was understood that the tensile strength
increased and the total elongation decreased as the elapsed
time increased.

The materials used in sodium-cooled fast breeder reactors
undergo long-term exposure to high-temperature sodium.
Therefore, it is necessary to estimate the compatibility of thesematerials with sodium, that is, the corrosion behavior andits effect on the mechanical strength, as well as the inherent
mechanical properties of the materials.
For the experimental fast reactor “JOYO”, the application of
zirconium (Zr) alloy as a neutron reflector is being examined.
The alloy does not contact sodium in the normal operating
condition because it is inserted into a protective tube madefrom stainless steel. However, for reactor safety, it is necessary
to ensure that a significant effect does not occur even if theprotective tube is ruptured.
In this study, the compatibility of the Zr alloy with hightemperature sodium was examined. It was found that thesurface became blackened only after exposure to flowing
sodium, as shown in Fig.13-13. Moreover, a weight gain wasobserved with increasing temperature and elapsed time, as
shown in Fig.13-14. It was due to oxygen diffusion into theZr alloy from dissolved oxygen in the sodium. In the static

sodium condition where the fill ratio of sodium was limitedby the container size, the progress of oxygen diffusion was
suppressed as the elapsed time increased because the amountof dissolved oxygen was also limited. On the other hand,
oxygen diffusion into the Zr alloy continued in the flowing
condition because dissolved oxygen was continually suppliedby the flowing sodium. The behavior occurred because theaffinity of Zr with oxygen was much higher than that of othersteels such as stainless steel, so Zr tends to form a stable oxide
thermochemically.
The mechanical properties were also changed by theenvironmental effect. Fig.13-15 shows the relationshipbetween the tensile properties and the elapsed time. In flowing
sodium, it was observed that the tensile strength increased andthe total elongation decreased as the elapsed time increased.
These were the typical embrittlement behaviors; however, theproperties changed sluggishly.
We are advancing the investigation towards the applicationof the Zr alloy as the neutron reflector in “JOYO”.

Reference
Furukawa, T. et al., Compatibility of Zirconium Alloys in High-Temperature Sodium, JAEA-Research 2011-039, 2012, 20p. (in Japanese).

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Development of Experimental Techniques/Facilities at JAEA R&D Centers

Oarai Research and Development Center

13-8 Reliable Management of Instrumentation to Ensure Reactor Safety－ Establishment of Damage Detection Method for Neutron Detector in High-Temperature Gas-Cooled Reactor －TDR cable tester

Lead wire

MI cable

Detector element

Electrodebbaa

WRM outline drawing●: Connection

Impedance (–)

Abnormal WRM
Normal WRM
Occurrenceof damageConnection“a~b”(25 mm/div)

WRM

HTTR bird’s eye view

Position of internal WRM (–)

Fig.13-16 Wide-Range Monitor (WRM)
The WRM is a reactor safety instrument that is installedin the core to measure the neutron flux.

Normalized CT value

1.2

Wave patterns of characteristic impedance observed by TDR

Fig.13-17 Application of electrical inspection method to WRM
Wave patterns of characteristic impedance observed by TDR in normal
and abnormal WRMs are compared. For the abnormal WRM, damage
was identified by recognizing a significant change in the wave patternof the characteristic impedance around connection “a” in the figure.

Unbroken lead wire

1.0
0.8
0.6

Lead wire in abnormal WRM

0.4
0.2

Lowered CT value

0.0
−1.5

0.0

1.5

Connection
“a”
3.0

4.5
[mm]

6.0

Distance from connection “a” (mm)Breaking
position oflead wireinabnormal
WRM

7.5

Breaking position of lead wirein abnormal WRM

Fig.13-18 Non-destructive and destructive
inspections of abnormal WRM
Non-destructive inspection showed a local
decrease in the CT value at the damaged part(a broken lead wire around connection “a” asshown in outline of WRM, Fig.13-17), which
was identified by the newly proposed electricalinspection method. On the other hand, thebreaking of a lead wire was directly observed in
the specified part by destructive inspection.

Confirmatory result of position of damage

Wide-Range Monitors (WRMs) are installed inside thereactor pressure vessel of the High-Temperature engineering
Test Reactor (HTTR). The WRMs are very important safety
instruments that measure the neutron flux to transmit thereactor scram signal and prevent such abnormal power
excursion events in the HTTR (Fig.13-16). Because the WRM
is activated by neutrons, it is difficult to examine its intactness
from outside of the core. For this reason, it has been expectedthat a reliable inspection method would be developed toguarantee its intactness in operation by detecting signs ofdamage to the WRM. Therefore, a new damage detection
technique was examined using an electrical inspection method.
An abnormal WRM, which became malfunctioning in the
HTTR, was inspected in detail to verify the proposed method
and also to investigate the cause of the malfunction.
Here, the selected electrical inspection method was TimeDomain Reflectometry (TDR), which is generally used todetect the breaking of cable wires. First, the observed wave
patterns of the characteristic impedance of normal andabnormal WRMs were compared. As a result, the existenceof damage in the lead wire of the WRM was successfully
identified by recognizing a significant change in the wave
pattern of the characteristic impedance (Fig.13-17). Second,non-destructive and destructive inspections were applied to theabnormal WRM to verify the result of the electrical inspection
method. A high-energy X-ray Computed Tomography (CT)inspection confirmed that the CT value was lower at the same
part where the damage was detected by the proposed electricalinspection method. The destructive inspection directly verified
that the lead wire was broken at the specified damaged part(Fig.13-18).
Thus, a new damage detection method was established thatuses electrical inspection while keeping the WRM installedin the core. The results contribute to upgrading of the HighTemperature Gas-cooled Reactor (HTGR) technology andcould be a promising technology for future commercial
HTGRs.

Reference
Shinohara, M. et al., Investigation on Cause of Malfunction of Wide Range Monitor (WRM) in High Temperature Engineering Test Reactor (HTTR)−Sample Tests and Destructive Tests−, JAEA-Technology 2012-032, 2012, 29p. (in Japanese).
JAEA R&D Review 2013117

Development of Experimental Techniques/Facilities at JAEA R&D Centers

Takasaki Advanced Radiation Research Institute13-9 Microfabrication of Teflon with Ion Microbeams－ Development of New Microfabrication Techniques for Polymer Surfaces －3 MeV
H microbeam+

Diameter of the beam
1 µm

3 MeV
H+ microbeam

250 keV
N+2 beam

Pre-irradiation

Fabrication of
micro-protrusions

50 µm

Area not irradiatedwith H+ microbeam

130 µm130 µmArea pre-irradiated withscanned H+ microbeam

50 µm

Protrusions

50 µm

Area pre-irradiated withscanned H+ microbeam

Smooth10 µm

Fig.13-19 Microstructures on Teflon surfaces fabricatedwith scanned MeV H+ microbeams
When a 3 MeV H+ microbeam is scanned spirally from thecenter of a square or a circle, a pyramid or a cone is formed,
respectively.

Fig.13-20 Checkerboard pattern consisting of smooth andmicro-protrusion areas fabricated with a scanned H +microbeam before uniform N2+ beam irradiation
A 3 MeV H + microbeam is scanned horizontally to make acheckerboard pattern; subsequently, the entire surface isirradiated with a 250 keV N2+ beam. Only the H+-irradiated areas
become smooth.

At the TIARA ion accelerator facility at the Japan Atomic
Energy Agency/Takasaki, a microfabrication technique called
proton (H+) beam writing (PBW) has been established on thebasis of the techniques of focusing down an ion beam to a
1 µm spot size and scanning it arbitrarily over a target surface.
Concerning the resist materials, we have demonstrated PBW’s
specific ability to fabricate high-aspect-ratio microstructureswith chemical etching after irradiation. We have attempted toapply the technique to Teflon® (fluoropolymer), on which isdifficult to fabricate microstructures because of its chemical
inertness, and started research on microfabrication to berealized only with ion microbeams.
First, a revolutionary technique using a MeV H+ microbeam
was developed: fabrication of microstructures on a Teflon
surface through a process of local rising (Fig.13-19). A 3 MeV
H+ beam travels in a straight line to a depth of approximately
100 µm in Teflon. Along the trajectories, main-chain scission
is induced, and consequently degradation gases are formed.
The amount of the gases depends on the irradiation conditions

such as the beam current, irradiation time, and scanning
pattern. When the gases leave the inside of a Teflon target
as bubbles, they raise the irradiated area; as a result, themorphology is changed, as shown in Fig.13-19.
When a Teflon surface is irradiated with keV nitrogenmolecular ion (N+2 ) beams, an initial flat surface changes intoa lawn-like one covered with dense micro-protrusions. In
addition, we found the conditions under which H+ microbeam
irradiation does not induce any surface morphological change
due to the degradation gases but instead reduces the molecularweight. Subsequent N+2 beam irradiation makes the H+irradiated surface evaporate uniformly and very smoothly.
This morphological change was also applied to make a smoothsurface with micro-protrusion areas at the same ground level(Fig.13-20).
It is shown that only ion beam irradiation can enable themicrofabrication of chemically inert materials without any
chemical processes. We intend to expand this technique tovarious materials other than Teflon.

Reference
Kitamura (Ogawa), A. et al., Microfabrication on Teflon Surface by MeV-Proton-Microbeam and keV-Nitrogen-Ion-Beam Irradiation, Nuclear Instruments
and Methods in Physics Research B, vol.307, 2013, p.610-613.

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Development of Experimental Techniques/Facilities at JAEA R&D Centers

Tono Geoscience Center

13-10 Excavation of Galleries at -500 m Depth－ Construction of Shafts and Research Galleries for the Mizunami Underground Research Laboratory －Existing as of March 31, 2013
Planned future development-200 m Measurement Niche
off Main Shaft
Sub Stage(Every 100 m deep)-200 m Measurement
Niche off Ventilation
Shaft-300 m Stage-300 m Measurement
Niche off Ventilation
Shaft-500 m Access/Research
Gallery-South
Borehole-300 m Access/Research Gallery

Fig.13-22 -500 m Access/Research Gallery-North
View is from -500 m A/R Gallery-North (5 m wide, 4.5 mhigh) toward Main Shaft.

-500 m Access/Research
Gallery-North

Borehole-500 m Stage
Ventilation Shaft :-500.2 m (φ4.5 m)

Main Shaft :-500.4 m (φ6.5 m)

Fig.13-21 Layout of the Mizunami Underground ResearchLaboratory
Layout shows existing Shafts and Research Galleries on
March 31, 2013 and planned future development.

Fig.13-23 Borehole investigation site
View shows borehole investigation in Measurement Niche
in -500 m A/R Gallery-North.

The Mizunami Underground Research Laboratory iscurrently under construction in Mizunami City. As of March
2013, approximately 150 m of horizontal excavations werecompleted for the GL -500 m stage (meters below groundlevel, Figs.13-21, 13-22).
To mitigate the potential excess groundwater inflow, preexcavation grouting was conducted before excavating theshafts and galleries. Grouting is the injection of material suchas cement into open fractures in a rock mass to stabilize andseal the rock. While planning the construction of the URL,
it was necessary to obtain reliable preliminary informationon the bedrock conditions in terms of the rock mass stability
and hydrogeology. Therefore, site characterization borehole
investigations were conducted prior to any excavations. These
investigations indicated that potentially large groundwater
inflows could occur during excavation of the Ventilation Shaft
at around GL -200 m, from GL -400 m to -460 m and near the-300 m Access/Research Gallery.
Borehole investigations were also conducted before
excavation of the Research Galleries at GL -500 m to obtainreliable information on the bedrock conditions (Fig.13-23).
These investigations indicated that high conductivity of therock mass, in the range of 10-7 -10-5 m/sec, could occur during
excavation of the -500 m A/R Gallery-South. Therefore preexcavation grouting was conducted before the excavation of
the shafts and galleries at this depth.
In planning the construction, the target for reduction
in water inflow was established by theoretical analysis of
the groundwater flow in terms of the bedrock conditions.
Observations during the excavations indicate that the preexcavation grouting was successful and that the targeted
reduction in inflow was achieved. Thus, the results indicatethat this methodology is effective in reducing water inflow.

Reference
Ishii, Y., Ikeda, K. et al., Results and Considerations on the Pre-Excavation Grouting below Four Hundreds Meter Depth of the Ventilation Shaft, JAEATechnology 2010-044, 2011, 92p. (in Japanese).
JAEA R&D Review 2013119Development of Experimental Techniques/Facilities at JAEA R&D Centers

Ningyo-toge Environmental Engineering Center

13-11 Investigation of Properties of Radon Hot Springs－ Radon-Induced Biological Responses and Its Biokinetics －(a) Saline+Sham(b) Alcohol+Radon

10 μm(c) Alcohol+Sham

10 μm

10 μm

Fig.13-25 Radon inhalation inhibits alcohol-induced oxidative injury inmouse liver(e) STZ+Radon(d) Saline+Sham

50 μm

Fig.13-24 Animal test chamber for radon
inhalation( f ) STZ+Sham

50 μm

50 μm

Fig.13-26 Radon inhalation inhibits Streptozocin (STZ)-induced diabetes
in mouse pancreas

According to legend, one of the hot springs in Misasa,
Tottori was first found in 1164, about 850 years ago. Misasa
means “third morning” in Japanese and was named foranother legend that people who went there for a hot spring
cure were healed completely by the third morning. The hotsprings in Misasa are famous worldwide due to the highradon concentration of the water. The Misasa hot springs
are believed to stimulate the body’s healing mechanism andenhance immunity, especially against reactive-oxygen- or
free-radical-related diseases. There is some possibility thatradon, a radioactive noble gas, plays an important role in theseproperties.
Okayama University and JAEA Ningyo-toge have conducteda collaborative study of the physiological effects of inhaledradon in the low-dose range. The main assignments were as
follows. On the basis of clinical knowledge, Misasa Medical
Center (Okayama University Hospital) clarified the issues thatshould be addressed. The Graduate School of Health Sciences(Okayama University) supervised the research and studied thebiological responses. The JAEA Ningyo-toge developed andcontrolled a facility for radon inhalation experiments and theinvestigation of the biokinetics and exposure doses of radon.
From 2007 to 2011, the following results were obtained.
●

The first Japanese large-scale facility was developed forradon inhalation experiments with small animals (Fig.13-24).
● The relationships between the radon concentration andinhalation time, which are the most basic parameters, were
widely examined to understand the change in antioxidative
functions due to radon.
● The inhibitory effects of radon on oxidative damage wereobserved using model mice with reactive-oxygen- or freeradical-related diseases such as alcohol-induced oxidative
damage and type I diabetes (Figs.13-25, 13-26).
● To discuss the biological responses to radon inhalation
quantitatively, the biokinetics of inhaled radon was
examined, and a model for the calculation of absorbed doses
for organs and tissues was obtained.
A comprehensive discussion, including considerations of
the radiation risk and safety, is required.

Reference
Ishimori, Y. et al., Animal Study on Biological Responses by Radon Inhalation Making Use of Waste Rock Which Contains Feeble Activity of Uranium (Joint
Research), JAEA-Research 2013-005, 2013, 60p. (in Japanese).

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About the Design of the CoverThe cover is designed to envisage a hopeful future shining in the sky, which is a clear blue resembling the color of the JAEA logo. This is accompanied
by white hexagons similar to the pattern in a tortoise shell, an ancient Japanese symbol of people’s wish for longer lives. Coincidentally, this shape is thesame as that of core fuel assemblies for both the prototype fast breeder reactor “MONJU” and the high-temperature engineering test reactor “HTTR”.
The images on the cover show the dose rate distribution map around the Tokyo Electric Power Company, Incorporated (TEPCO) Fukushima Daiichi
Nuclear Power Station (NPS), which was traced out using an unmanned helicopter (top left), and atomic-scale cesium absorption on a type of zeolite
called mordenite (bottom right).
The former shows the results of the dose rate at 1 m above the ground, which were converted from monitoring data obtained by a flight at 80 m above theground around the TEPCO’s Fukushima Daiichi NPS (from 3 km to 5 km from 1F) by a radiation monitoring system using an unmanned helicopter.
The dose rate was normalized to that on March 21, 2013, which was the last day of measurement (Topic 1-3, p.14).
The latter shows that mordenite, a type of zeolite, is effective for cesium decontamination. We numerically simulated cesium absorption on mordenite at
the atomic scale (Topic 1-19, p.30).

ISSN 2188-1456

JAEA R&D Review

2013

Futaba-machi

3 km from 1F

1F

Okuma-machi
Dose rate at 1 m
above the ground (μSv/h)> 30
20 - 30

Tomioka-machi

10 - 20
5.0 - 10

0

1.5

3 km

［Normalized to March 21, 2013］

1.0 - 5.0
< 1.0

Dose rate distribution map aroundthe TEPCO’s Fukushima Daiichi NPS(1F), which was traced out using
an unmanned helicopter

Cs

Z
Y

X
Atomic-scale cesium absorption on
X a type of zeolite called mordenite

Si
O

JAEA R&D Review 2013

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JAEA R&D Review 2013

Published by

Japan Atomic Energy Agency in February 2014
Editorial Board
Chief editor: Kazuo Minato
Vice editor: Hiroshi Igarashi
Editors: Tsutomu Kawamoto, Takumi Nemoto, Makoto Tanaka, Yoshiaki Shikaze, Masaru Watahiki,

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Takayuki Ozawa, Yasuo Sasaki, Mitsuo Tachibana, Tetsuro Ishii, Akihiro Onose,
Yoshihisa Murokawa, Hiroshi Takada, Masanori Kaminaga, Junsaku Nakajima,
Fumiaki Yamada, Kazumasa Narumi, Shinichiro Mikake, Yasufumi Uechi
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