2012

Decontamination on pavement using shot blasting

Decontamination of soil by using poly-ion complex inOguni, Ryozen, Date-city, Fukushima

Message from the President

President Atsuyuki Suzuki

It has now been over a year since the nuclear accident at the Tokyo Electric Power
Company, Incorporated (TEPCO) Fukushima Daiichi Nuclear Power Station (NPS). Of note,
only be started a lengthy process of the environmental decontamination and remediation
work, which has involved the efforts of the entire country. Despite our efforts, however, wemust be prepared for a long way to go. In addition, while ensuring the long-term steadiness ofsafety condition of the damaged facilities on site, and within the power plant, the tenacious
work pursuing the completion of the stepwise tasks necessary for decommissioning continues
on a daily basis.
I believe that, as an organization that is the only comprehensive R&D institute involved innuclear power in the country, the Japan Atomic Energy Agency (JAEA) must understandably
direct all of its technological expertise toward resolving the Fukushima matter with theutmost priority.
Tasks such as the decontamination of areas outside of the power plant and environmental
monitoring have been possible, and efforts to clean up the environment continue to be madethroughout the country. In addition, in the future, based on the policies put forward by thegovernment-TEPCO s Fukushima Daiichi NPS med-long-term response council R&D
headquarters, we would like to invest significant resources in providing technical support forthe decommissioning activities within the power plant.
For example, with respect to environmental remediation, following an integrated
investigative analysis focused on the long-term environmental fate of Fukushima prefecture
and the surrounding area, we want to help support the comprehensive integration such
scientifically reliable data from around the country. Regarding the decommissioning of thepower plant, we have initiated basic and fundamental R&D programs concerning the handlingand management of the complex and diverse forms of molten fuel collectively referred to asfuel debris. We plan to further develop these activities in the future.
Furthermore, because the decommissioning of the TEPCO s Fukushima Daiichi NPS is achallenge concerning nuclear power, which is of the utmost importance for the country, I feelthat we need to accelerate the current decommissioning plans set by the JAEA as soon as
possible and consider ways to effectively capitalize on the experience and knowledge gained
from these to be used for Fukushima.
The seriousness of the accident at the TEPCO s Fukushima Daiichi NPS serves to newlyraise the importance of ever more severe measures for the safety of nuclear power. In order tomake this tragedy a lesson that should never be forgotten, we are increasingly pushing ahead
with R&D that is able, for example, to explain the phenomena associated with severe
accidents, providing measures to consider the event of such accidents, and substantiating
safety through material and reactor irradiation tests. We are also planning to study ways tomake innovative safety improvements in future reactors, such as the sodium-cooled reactor
and the high-temperature gas reactor. Safety research remains one of the main, primary
themes that JAEA must address, and we are in the process of strengthening the system towardthat end.
The Framework for National Nuclear Energy Policy regarding R&D for the fuel cycle,
which target the prototype fast-breeder reactor MONJU , reprocessing, and geologicaldisposal, are presently under review, and any specific future planning must await theconclusions of that review. However, I do not expect the importance of the R&D associatedwith spent fuel that actually exists today to be lessened. The management of spent fuel is a
global problem, and therefore, JAEA can play an important role internationally.
We are also earnestly engaged in R&D activities related to fusion energy technologies andquantum beam applied technologies, and moreover, in a broader sense, cutting-edge basic andfundamental R&D. In addition, we are actively pursuing innovations that can serve as
catalysts for future advances and the creation of more sophisticated basic and fundamental
technologies.
With regard to fusion energy technologies R&D, the JAEA, having been appointed as thesole national institute in our country for the planning of the International Thermonuclear
Experimental Reactor (ITER), has played a central role in implementing that plan.
Our R&D activities in quantum beam applied technologies have proceeded in cooperation
with other facilities, such as conventional nuclear reactors and accelerators, and researchusing Japan Proton Accelerator Research Complex (J-PARC) has been developed on aninternational level, with many notable accomplishments already achieved.
Furthermore, with the principal objective of cultivating promising young researchers andthrough continued external support, we have encouraged groundbreaking research in cuttingedge fields. These activities have been highly praised both at home and abroad, and I lookforward to their continued and evolving implementation.
We believe it is important to proactively share and utilize the achievements of thisorganization s R&D.
This publication constitutes a review of our achievements in the fiscal year 2011. It
provides you with a look at some of the work that has been carried out, and also invites you tocheck the references listed and contact the researchers if there are any topics that you wish tolearn more about. I would be most gratified if I could hear from you with any comments on
this publication.
I hope that you enjoy this publication. Thank you for your interest.

J 
8

About This Publication and the Outline of Organization of JAEA

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

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1. Radionuclides Discharged into the Atmosphere−Source Term Estimation and Atmospheric Dispersion Simulation Using WSPEEDI−
2. Radionuclides Released into the Ocean−Diffusion Simulations for Regions Adjacent to Japan−
3. Collection and Provision of Environmental Radioactivity Data
−Publication of Data for Environmental Reconstruction−
4. How Far Did Radionuclides Diffuse from the Accident?
−Measurement of Atmospheric Anthropogenic Radionuclides Using the CTBT IMS Network−
5. Detailed Investigation of Radionuclide Distributions in the Environment−Construction of Soil Deposition Map and Dose Rate Map−
6. Radiation Dose Map of Japan Using an Aircraft−Visually Grasping the Influence of the Accident−
7. Effects of Forest Decontamination on the Reduction in External Exposure Doses−Field Demonstration of Forest Decontamination−
8. Creating Decontamination Guidelines Based on Practical Actions−Decontamination Pilot Project−
9. Cleanup of Cesium-Contaminated Soil
−Investigation of Removal of Radiocesium from Soil by In Situ Ground Heating−
10. Decontamination of Soil Using a Poly-Ion Complex and Clay
−Decontamination of Soils Polluted by Radioactive Cesium and Prevention of Recontamination−
11. Decontamination of Playground Lot−Trial Decontamination and Its Effects on Air Radiation Dose Rate−
12. Tool for Planning Rational and Effective Decontamination−Development of Computer Software for Efficiently Predicting Dose Rates after Decontamination−
13. A Simple Method for Selective Measurement of Cesium 134 and 137−Utilization of Common NaI(Tl) Scintillation Spectrometers−
14. Fibrous Grafted Adsorbents Help in Cleanup of Contaminated Water−Development of Fibrous Adsorbent for Removal of Radioactive Cesium−
15. Development of a Highly Cesium-Selective Adsorbent for Volume Reduction−Molecular Design of Crown Ether Using New Chemical Bond Property−
16. Verification of the Exposure Reduction Effect of Tap Water Restrictions−Evaluation of Averted Doses to Infants in the Implementation after the Accidents at the TEPCO s Fukushima Daiichi NPS−
17. Visualization of Distribution of Fallout Radioactive Cs in Plants−Toward Unveiling the Present Status of Forest Contamination−
18. Data Acquisition for Analysis of the Melt Progress in the Reactor Core
−Thermal Properties of Molten Fuel from Three Mile Island Unit 2−
19. Prediction of the Core Meltdown of the TEPCO s Fukushima Daiichi NPS Unit 1 Reactor−Evaluation of the Influence of an Isolation Condenser on Core Cooling Using the TRAC Code−
20. How to Avoid Core Damage in Unit 2 of the TEPCO s Fukushima Daiichi NPS−Evaluation of the Efficiency of Alternative Water Injection Timing through Accident Analysis−
21. Toward Safe Management of Accident Waste−Challenges to Long-Term Storage, Processing, and Disposal of Waste−
22. Decontamination of Radioactive Strontium to Make Fresh Water−Application of a Strontium Adsorbent Prepared for Use with Radioactive Waste Solutions from Nuclear Fuel Reprocessing−
23. Evaluation of Hydrogen Production from Waste Zeolite Adsorbents−γ-Radiolysis of Mixtures of Zeolites and Seawater−
24. Development of Special Measuring Equipment Using JMTR Irradiation Techniques−A New Water-Level Indicator Applicable under Severe Conditions−

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

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1. Japanese Fast Reactor Technologies Draw the Attention of Electricit de France−Safety Improvement and Mass-Volume Reduction of the JSFR−
2. Containment of Core Disruptive Accidents−Fuel Discharge Behavior in a Fuel Assembly with Inner Duct Structure−
3. Prediction of the Flow Structure in Large Diameter Elbows−Evaluation of Flow-Induced Vibration in Large Diameter Pipes in the JSFR−
4. For Enhancement of the Reliability of SFR Pipe Designs−Creep Strength Assessment of Mod.9Cr-1Mo Steel Welds Taking Type-IV Damage into Account−
5. The MONJU Core Can Be Cooled during a Station Black-Out−Core Cooling Capability during an Earthquake and Subsequent Tsunami−
6. Toward Highly Efficient Production of MOX Fuel
−Confirmed Feasibility of Production from Agitated Granular Powder−

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and Development on Geological Disposal of
3 ResearchHigh-Level Radioactive Waste
R&D Supporting the Technology and Reliability of Geological Disposal in Japan

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1. Study of the Groundwater Flow in Coastal Areas−Optical Measurement of the Salinity Distribution on the Saltwater Wedge−
2. Understanding the Impact of Nitrates on Geological Disposal Systems−Modeling of the Chemical Transition of Nitrates Accompanied by Corrosion of Carbon Steel−
3. Challenges of Seepage Control at a Repository for Radioactive Waste−Applicability of a Grout Penetration Model−
4. Looking at Huge Earthquakes through a 0.1 mm Grain
−Thermochronology with a Single Grain Using the (U-Th)/He Method−
5. Fault Zone Determination through Multivariate Analysis−Case Study Using a Dataset from a Deep Borehole−
6. Technology for Estimating Hydrogeological Characteristics Deep Underground−Evaluation of Hydrogeological Characteristics Using Water Pressure Data−
7. Development of a Methodology for Reconstructing Very Long-Term Geosphere Evolution−A Very Long-Term Groundwater Flow Analysis Using Numerical Simulation−
8. Examination of the EDZ Developed around a Tunnel−Estimation of the EDZ by Seismic Tomography Investigation−
9. Evaluation of the Initial Stress Distribution Deep Underground−Three-Dimensional Initial Stress Measurement by Hydraulic Fracturing−

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4 Nuclear Fusion Research and Development
Toward Practical Use of Fusion Energy

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1. Manufacture of Superconductors for ITER Coils−Progress in Procurement of Superconductors for TF Coils, with Japan Leading in the World−
2. A Laser System for Plasma Diagnostics Has Achieved the World s Best Performance−Development of YAG Laser for the Edge Thomson Scattering System in ITER−
3. Attainment of Large-Scale 3D Multibeam Analysis−Compensation for Beam Deflections in the Accelerator of ITER NBI−
4. Overview of IFERC Project−Activities of DEMO Design and R&D Coordination Centre, and Computational Simulation Centre−

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4 JAEA R&D Review 2012

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5. Achieving Safety in the Deuteron Accelerator Facility−Development of a Safety Protection System for the IFMIF/EVEDA Accelerator−
6. Progress in Satellite Tokamak Programme Project−Construction Activities of JT-60SA Tokamak Progressing toward Assembly−
7. Accurate Fabrication of Large Superconducting Coil−Fabricating the Superconducting Coil for JT-60SA−
8. Manufacturing the Doughnut-Type Vessel for JT-60SA−High-Precision Structure of Double-Wall Vacuum Vessel−
9. Progress in Stability Theory for Rotating Plasmas−New Matching Theory for Resistive Wall Modes−
10. Lithium Isotope Separation for Fusion Reactor Fuel
−Development of Innovative Technology for Lithium-6 Enrichment Using an Ionic Liquid−
11. Toward Clarifying Complex Plasma Behavior−Development of an Integrated Simulation Code for Entire Tokamak Plasma−

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5 Quantum Beam Science Research
Development of Quantum Beam Technology

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1. Large Increase in the Maximum Energy of a Laser-Driven Proton Beam
−40 MeV Proton Generation Using a Ti:sapphire/OPCPA Hybrid Laser System−
2. High-Power γ-ray Generation via the Radiation Reaction Effect Using an Ultra-Intense Laser Pulse−Proposal for a New Laser-Driven γ-ray Source−
3. High-Precision Measurement of Nuclear Materials−Proposal for the Advanced HKED System−
4. Neutron Observation of Deformation Behavior of Materials at Cryogenic Temperature−Neutron Strain Measurement of a Superconducting Wire under Cryogenic Tensile Loading−
5. Discovery of Novel Rare-Earth Metal Hydrides under High Pressure−Formation of NaCl-Type Monohydrides by Pressure-Induced Phase Separation−
6. Microscopic Magnetization Process in a Tb43Co57 Amorphous Film Using Magnetic Compton Scattering−New Measurement Method for Spintronics Materials−
7. Development of a Microfabrication Technique for Utilization of Environmentally Friendly Plastics−Micro-/Nano-Fabrication of Crosslinked Poly(L-lactic acid) Using EB Nanoimprint Lithography−
8. Contribution to the Elucidation of the Mechanism of O2 Generation from Water Using Ce(IV)−Discovery of Dinuclear Complexes of Tetravalent Cerium in an Aqueous Solution−
9. Developing a Technique for Operating a Protein Assembly−Identification of the Key Residue Controlling Protein Assembly via X-ray Crystallography−
10. Phase Retrieval from Sparse Diffraction Data
−Coherent X-ray Diffraction Imaging of Single Particles−
11. Drastically Altering the DNA of Plants−Ion Beams near the Range End Frequently Induce Large Deletions−

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6 Nuclear Safety Research
Evaluation of the Safety of Various Nuclear Facilities

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1. Accurate Evaluation of the Fuel Failure Limit−Improved Application of NSRR Test Data to Power Reactor Conditions−
2. Assessing the Failure Resistance of Reactor Pressure Vessels−Structural Integrity of Reactor Pressure Vessels Based on Probabilistic Fracture Mechanics
Considering the Effect of Residual Stress−
3. Evaluation of the Confinement Safety of Radioactive Materials during a Fire−Acquisition of Combustion Data and Evaluation of the Clogging of the HEPA Filter−
4. Reactions for Determining Chemical Environments in Radioactive Waste Repositories−Corrosion Reaction of the Steel Overpack for High-Level Radioactive Waste−

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7 Advanced Science Research
For the Evolution of Nuclear Science

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1. First Observation of 239Pu Nuclear Magnetic Resonance−Elucidation of the Electronic State in Pu Compounds−
2. Generating Magnetic Flows with Sound−Discovery of a Link between Sound and Modern Spintronics−
3. Development of a New Method for Preparing Atomically and Electronically Uniform Graphene−Significant Progress for Spintronic and Nanoelectronic Applications−

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8 Nuclear Science and Engineering Research
Formation of Basis for Nuclear Energy R&D and Creation of Innovative Technology

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1. Challenge of Obtaining Accurate Nuclear Data for Radioisotopes−Measurement of Neutron Capture Cross Sections of 244Cm with ANNRI−
2. Research on Boiling Heat Transfer Mechanism−Development of Simultaneous Measurement Technology for Surface Heat Flux and Surface Temperature−
3. More Precise Estimation of Reactor Decay Heat−JENDL/FPD-2011 and JENDL/FPY-2011−
4. Development of Database for a Practical Fuel Design−Formulation of Thermal Conductivity of Zr-Based TRU Nitride Fuels−
5. Analysis for Safety Assessment of Radwaste Disposal−Validation of Calculated Values of Content of Difficult-to-Measure Nuclides 79Se and 135Cs−
6. Removal Mechanism of Radiation-Induced Nucleotide Damage−Simulation Study of the Mechanism of Molecular Recognition by the Enzyme MutT−
7. Prediction of Radiation Damage in Accelerator Facilities and Nuclear Power Plants−Development of Radiation Damage Model for Various Particles−

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9 Nuclear Hydrogen and Heat Application Research
Research on a Naturally Safe HTGR for a Low-Carbon Society94

1. Explore the Ultimate Nuclear Reactor That Is Safe Even If Accidents Occur−Conceptual Design Study of a Naturally Safe HTGR−
2. Spontaneous Stabilization of HTGRs without Reactor Scram and Core Cooling−Safety Demonstration Tests Using the HTTR: Loss of Reactivity Control and Core Cooling−
3. Design Criteria of HTGRs for Withstanding Extreme Events−Investigation of Technical Feasibility for Naturally Safe HTGRs−

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of Decommissioning and Radwaste Treatment
10 Development
Technology
Execution of the Decommissioning of Nuclear Facilities and the Treatment and Disposal ofRadioactive Waste
1. Analysis of Transuranium Nuclides in Radioactive Waste−Analysis of Am and Pu in Contaminated Metal Surface−
2. Prevention of Heavy Metal Elution from Solidified Forms of Incinerated Ash and Cement−Confinement of Heavy Metals Using Insolubilizers−

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11 Computational Science and E-Systems Research
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Role and Potential of Nuclear Computational Science
1. How Does Impurity Diffuse in Irradiated Materials?
−Evaluation of Impurity Segregation by Diffusion Rate Equations Based on the Atomic Migration
Mechanism−
2. J-PARC and Supercomputer, a Collaborative Challenge to Solve the Mechanism of Superconductivity:−A Quest for Electron Pairing through Computer Simulation of Neutron Scattering in Iron-Based
Superconductors−
3. How Do Earthquake Waves Propagate into a Nuclear Facility?
−Realization of Precise Simulation of Building-Equipment Interaction during Earthquakes−

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

of Experimental Techniques / Facilities at
13 Development
JAEA R&D Centers
1. Reuse and/or Recycle of Dismantled Objects−Wet Blast Decontamination Technology−
2. Improved Reliability of Hand Dosimetry−Development of a Ring-Type Dosemeter Using Optically Stimulated Luminescence Elements (OSLRD)−
3. Effective Utilization of Concrete Debris Generated from Nuclear Reactors−Clearance of Concrete Debris Generated from Modification Work on JRR-3−
4. Separation and Recovery of Long-Lived Nuclides for Reduction of Radioactive Waste−Behavior of Oxidation and Extraction of Neptunium in Simplified Solvent Extraction Process−5. Scintillation Neutron Detectors with a High Spatial Resolution−Position-Sensitive Detectors for Single-Crystal Diffractometer in J-PARC−
6. Visualizing Magnetic Fields Using Neutrons−Magnetic Field Imaging Technique Using Polarized Pulsed Neutrons−
7. Realizing a 1 MW High-Power Proton Beam
−Beam Loss Reduction by Injection Painting in J-PARC RCS−
8. First HTGR Demonstration−First Achievement of a Continuous Supply of High-Temperature Nuclear Heat−
9. Development of a Dissimilar Metal Joint of Zircaloy-2 and 316L Stainless Steel−Application to Power Ramping Test of LWR Fuel in JMTR−
10. Corrosion Issues in Sodium Environment−Effects of Sodium on Structural Materials for Fast Breeder Reactors−
11. Chemical-State Observation of Microparticles−Microscopic Imaging of Luminescence Induced by Scanning Ion Microbeam−
12. Shafts Excavated to -500 m Depth−Construction of Shafts and Research Tunnels and Confirmation of Reduced Groundwater Inflow
at the Mizunami Underground Research Laboratory−
13. Progress in Methodology Considering γ-ray Shielding Effects−Development of Simplified Calculation Method for Heterogeneously Distributed Radioactive Waste Drums−

JAEA R&D Review 2012

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About This Publication and the Outline Organization of JAEA
This publication introduces our latest research and development (R&D) results in each field. Each chapter presents theactivities of one R&D Directorate. As shown in the organization chart, the various R&D Directorates perform their activities
through R&D centers or institutes. Some of these consist of only one site, while others are located at two or more sites,
depending on their R&D activities. The R&D centers and institutes are located throughout Japan, as shown in the map below.
The following brief introduction outlines the research undertaken by each R&D Directorate in various R&D centers andinstitutes.
1.

2.

3.

4.

The Headquarters of Fukushima Partnership Operations is engaged in R&D to recover from the accident at theTokyo 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
Research Institute, the Nuclear Fuel Cycle Engineering Laboratories, and the Oarai Research and Development Center.
These centers conduct 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 outreach activities concerning radiation, and study measurement of radioactive materials in the human body bywhole-body counters.
The Advanced Nuclear System Research and Development Directorate is carrying out R&D aimed atcommercializing the fast breeder reactor (FBR) and its nuclear fuel cycle. The R&D of plant technology using theprototype fast breeder reactor MONJU is being undertaken at the Tsuruga Head Office (Fast Breeder Reactor Researchand Development Center), R&D for innovative FBR technology is being conducted at the Oarai Research andDevelopment 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
establishing techniques 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 and Development Center, the focus is on improving technologies for designing disposal facilities and safetyassessment. In addition, work has been ongoing to develop a next generation knowledge management system based on theabove R&D activities.
The Fusion Research and Development Directorate is performing fusion R&D as a domestic agency of the 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

Japan Atomic Energy Agency -Outline of OrganizationResearch and Development Directorates

Nuclear Safety Research Center

President
Executive Vice President

AuditorExecutive Director

Management Sector
Policy Planning and Administration Department
General Affairs Department
Auditor Office‫ޓ‬
Legal Office
Personnel Department‫ޓޓ‬

Advanced Science Research Center

Financial Affairs Department‫ޓ‬Contract Department‫ޓޓ‬
Nuclear Science and Engineering Directorate

Quantum Beam Science Directorate

Fusion Research and Development Directorate

Advanced Nuclear System Research and Development Directorate
‫ޓޓޓ‬
‫ޓ‬

Geological lsolation Research and Development Directorate
‫ޓ‬
‫ޓ‬
‫ޓ‬
‫ޓ‬
‫ޓ‬

Nuclear Cycle Backend Directorate

as of December, 2012

8 JAEA R&D Review 2012

Research and Development Institutes/Centers
Tsuruga Head Office
Fast Breeder Rector Research and Development Center
Fugen Decommissioning Engineering Center

Project Promotion Sector
Headquarters of Fukushima Partnership Operations
Safety Administration Department

Tokai Research and Development Center
Nuclear Science Research Institute
Nuclear Fuel Cycle Engineering Laboratories

Public Relations Department

J-PARC Center‫ޓ‬International Affairs Department

Oarai Research and Development Center

Construction Department

Naka Fusion Institute‫ޓޓ‬Intellectual Resources Department

Takasaki Advanced Radiation Research InstituteCenter for Computational Science & e-Systems

Kansai Photon Science Institute‫ޓ‬

Department of Science and Technology for Nuclear Material Management

Horonobe Underground Research Center

Nuclear Human ResourceDevelopment Center

Tono Geoscience Center‫ޓ‬

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

Industrial Collaboration Promotion Department

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 Fusionmaterial Irradiation Facility as the BA activity are performed mainly in the Aomori Research and Development Center.
5. 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 andsynchrotron radiation facilities is performed at the Kansai Photon Science Institute.
6. 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 doneat the Tokai Research and Development Center (Nuclear Science Research Institute) and at the Tsuruga Head Office.
7. 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.
8. 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.
9. 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.
10. 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
Research and Development Center.
11. The Center for Computational Science & e-Systems develops pioneering simulation technology and basic technology
in computational science as well as operates and maintains computer equipment. This occurs mainly in the Tokai
Research and Development Center (Nuclear Science Research Institute).
12. 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.
13. The R&D Centers and Institutes, located at 11 sites in Japan, manage and improve the performance of facilities andequipments to support the abovementioned R&D Directorates in safe and efficient R&D activities.

R&D Centers of JAEA
Horonobe Underground Research Center
Tsuruga Head Office
Prototype Fast Breeder Rector MONJU,
Decommissioning of Advanced Thermal
Reactor FUGEN

Aomori Research andDevelopment Center

R&D of Geoscientific
Research and to
Geological Disposal
for HLW

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

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

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

Ningyo-toge Environmental
Engineering Center

Nuclear Fuel Cycle
Engineering Laboratories

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

J-PARC Center

Kansai Photon Science InstituteMaterials and Life Science using
Neutrons and Muons, Particle
and Nuclear Physics using Kaonsand Neutrinos at Japan Proton
Accelerator Research Complex.

Photon & Synchrotron
Radiation Science

Naka Fusion Institutec RIKEN/JASRI

Takasaki Advanced Radiation
Research InstituteR&D of Radiation
Applications

Fusion R&D, ITER Support

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

JAEA R&D Review 20129

1  '  
Environmental Remediation and Plant Decommissioning
Radiation monitoring

Environmental
remediation

Emergency responses
On-site Accident
Mitigation

Communications
and public relations

Human resource
development
Fig.1-1 JAEA s post-accident efforts
This is the homepage of the Headquarters of Fukushima Partnership
Operations of the JAEA. (https://fukushima.jaea.go.jp/en/)

JAEA has undertaken diverse action for post-accident
Fukushima (Fig.1-1, Fig.1-2).
Emergency Response to the Great East Japan
Earthquake
JAEA, as a designated public institution according to the
Basic Law on Natural Disasters, took action in response to
the Great East Japan Earthquake that occurred on March 11,
2011. In response to a request by the government, JAEA
dispatched seven technical experts equipped with monitoring
devices on March 12; the technical experts were scrambled
by the Japan Self-Defense Force helicopter from the Hyakuri
air base in Ibaraki Prefecture. JAEA sent a total of 45318
staff members to Fukushima between this first action andMarch 31, 2012.
Radiation Monitoring
Reliable and accurate data relating to environmental
radioactivity are required for assessment of health effects anddecontamination plans. Our monitoring- and mapping-related
activities include air radiation dose rate mapping (Topic 1-1),
simulation of radionuclide transport in seawater (Topic 1-2),
collection and publication of environmental radioactivity
monitoring data (Topic 1-3), and measurement of atmosphericradioactive materials by IMS (Topic 1-4). Additionally, under
the contract with MEXT, JAEA has also conductedmeasurement and mapping of radionuclides deposited on the
ground surface in Fukushima (Topic 1-5) and aerial radiationdose monitoring throughout Japan (Topic 1-6); the resulting
maps are available on the MEXT homepage.
Environmental Remediation
Some of the major contributions of JAEA include
decontamination - or cleanup - related activities for theenvironmental remediation of Fukushima. JAEA released
decontamination guidelines and catalogs of decontamination
techniques to assist in the cleanup of environment
10 JAEA R&D Review 2012

Fig.1-2 Logo created by Fukushima Prefecture,symbolizing rebuilding of Fukushima with thecooperation of all Fukushimans

contaminated with radioactive materials (Topic 1-7). The
decontamination pilot project (Topic 1-8) involved acomprehensive demonstration, providing practical information
on how to clean up radioactive contamination that had spreadthroughout residential areas. Based on the experience andtechnical expertise obtained as a result of these activities,
which were conducted under contract with the government,
JAEA has been able to offer various forms of technicalassistance to local authorities in Fukushima and adjacentprefectures, and has supported environmental remediation
planning in these prefectures.
JAEA, as a R&D institute, has been engaged in varied
R&D: soil washing for farmland reparation (Topic 1-9,
Topic 1-10), trial decontamination of playground lots inresidential areas (Topic 1-11), computer simulation to predictand estimate the effects of areal decontamination (Topic 1-12),
simple and convenient methods for measurement ofradioactivity (Topic 1-13), cesium adsorbents (Topic 1-14,
Topic 1-15), verification of the effectiveness of post-accident
preventive measures (Topic 1-16), and investigation of forest
contamination (Topic 1-17).
Communications and Public Relations
JAEA has held question and answer sessions discussing
radiation and health: JAEA technical members have been
dispatched to schools in Fukushima Prefecture (fromkindergartens to junior high schools) upon request. After
briefing on radiation, JAEA members talked face to face withparents and teachers, answering their questions about
radiation and its health effects. Since July 2011, 178 sessions
have been held and a total of around 16746 people attended.
Human Resources Development
Upon request from the government (the Agency for
Natural Resources and Energy) and Fukushima Prefecture,
JAEA initiated training sessions to foster development ofradiation management staff and decontamination workers.

Secretariat

Working team for preparation of fuel debris removal

SWT for corestatus assessment
and analysis

Sub-working team (SWT) for
equipment/device development, etc.

SWT for fueldebris property
assessment
and treatment
preparation

Workingteam forradioactive Joint task force
waste
for remotetreatment
technologies
anddisposal

FY2011- FY2013-

FY2011- FY2011- FY2011- FY2011- FY2011- FY2013- FY2015- FY2013- FY2012-

FY2011-FY2011- FY2011- FY2011- FY2016-

Examination of waste treatment
and disposal

Stability of contaminated water
processing

Analysis of fuel debris properties

Fuel debris measuring andmanagement measures

Fuel debris treatment technologies

Assessment of simulated fuel debris
characteristics

Accident development analysis

Fuel debris criticality control

Fuel debris storage technologies

Removal of fuel debris and structuresin the reactor

Investigation of the RPV interior

Investigation of the PCV interior

Sealing and repair in the reactor
building/PCV

Identification of leak areas in thereactor building/PCV

PCV/RPV soundness evaluation

Remote decontamination of thereactor building interior

Damaged fuel processing
technologies

ReportExamination

Working teamfor spent fuel poolcounter-measures

Long-term stability of fuel assemblies‫ޣ‬Specific R&D projects‫ޤ‬‫ޣ‬Overall management‫ޤ‬R&D Headquarters

FY2011- FY2011-

Fig.1-3 Government and TEPCO mid-to-long-term R&D framework
JAEA will dominate R&D relating to long-term fuel assembly management and remotely controlled decontamination technologies forthe power plants. R&D road map for decommissioning of Units 1F1~1F4 at the TEPCO s Fukushima Daiichi NPS, Nuclear Emergency
Response Headquarters Government-TEPCO Med- and-long-Term Response Council (December 21, 2011).

Onsite Accident Mitigation
Immediately after the accident, JAEA has been providing
technical advice to the government and the Tokyo ElectricPower Company, Incorporated (TEPCO), while simultaneously
initiating research into mid- and long-term post-accident
mitigation. Data acquisition and analysis are underway forthe progression of core meltdown (Topic 1-18), and analysis isbeing conducted on reactor core behavior in Unit 1 (Topic 1-19)
and Unit 2 (Topic 1-20) of the TEPCO s Fukushima Daiichi
Nuclear Power Station (NPS). Additionally, generation ofradioactive waste by treatment of contaminated water used forcooling pools (Topic 1-21), strontium absorption (Topic 1-22),and hydrogen gas generation from spent zeolite (Topic 1-23)
are also being studied. Water gauges that can work under
extremely high radiation fields are required, and theirdevelopment is also one of JAEA s R&D topics (Topic 1-24).
JAEA s future R&D efforts will follow the government s and
TEPCO s mid-to-long-term R&D roadmap (Fig.1-3). JAEA
will be extensively involved in fuel assembly management
and remotely controlled decontamination technologies for the
TEPCO s Fukushima Daiichi NPS.
Response from and Involvement of JAEA Branches
Technical support, including staff attachment, has involved
the cooperation of all JAEA branches. JAEA colleagues
played various roles, including those in the Emergency
Operations Center (EOC) of MEXT: environmental radiation
monitoring, decontamination pilot projects, providing local
authorities with technical assistance for decontamination,communication and public relations, support for temporary
return of evacuees, and so on.
Tsuruga Head Office
The head office sent staff to Fukushima Medical
University to join medical teams prepared for highexposure to radiation.

Tokai Research and Development Center, Nuclear
Science Research Institute
The institute conducted analyses of environmental
samples such as soil, contaminated water generated byreactor core cooling, and radionuclide-adsorbed zeolite.
Tokai Research and Development Center, Nuclear
Fuel Cycle Engineering Laboratories
The laboratories conducted analyses of treatment ofaccumulated water, molten fuel, and cesium behavior
during incineration of cesium-contained wastes.
Oarai Research and Development Center
The center conducted salt immersion tests utilizing storage
capsules for spent fuel at their Fugen Decommissioning
Engineering Center, in order to evaluate long-term safetyof fuel assemblies taken from the spent fuel pools of the
TEPCO s Fukushima Daiichi NPS. Remotely controlleddecontamination was attempted at the Fugen plant.
Naka Fusion Institute
The institute conducted measurements of the radioactivityof water in the region of Fukushima.
Takasaki Advanced Radiation Research Institute
Using their γ-ray irradiation facility, the instituteconducted an irradiation test for Quince, a disaster relief
robot introduced to survey the inside of power plantbuildings. They also performed simulation tests of hydrogengas generation by radiolysis of seawater.
Horonobe Underground Research Center
The center dispatched staff members to the Headquartersof Fukushima Partnership Operations and EOC of MEXT.
Tono Geoscience Center
The center conducted analyses of groundwater flow
around the Fukushima Daiichi site.
Ningyo-toge Environmental Engineering Center
Computer simulation was made to analyze cesium
behavior during incineration of contaminated debris. They
also studied on the effects of ground cover plants for land
restoration.
JAEA R&D Review 201211

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

1-1

Radionuclides Discharged into the Atmosphere=Source Term Estimation and Atmospheric Dispersion Simulation Using WSPEEDI=

42°

0.097

40°

Release rate (Bq h-1)

1.E+16

34

1100

I-131

1.E+15

Cs-137

1.E+14

10000

38°

1.2

1.E+133340000

1.E+120.21

1.E+113

4

12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 1 2 3 4 5

36°

4700
8100

170
0
0

Date in 2011
00

Fig.1-4 Estimation results of I-131 and Cs-137 release rates
It can be seen that the release rates drastically increased on March 12, 14,
and 15.

1200

0.095

5700

17000

0

0

3400

5300
4900

540

34°
138°

136°
100

1000

10000

140°
100000142°
1000000

Bq m-2

10000000

Fig.1-5 Surface deposition of Cs-137 over landin eastern Japan calculated using WSPEEDIThe calculations agreed well with the observations
at each prefectural and city government (numbersin the figure).

A significant amount of radionuclides were discharged into
the atmosphere owing to the accident at the TEPCO s
Fukushima Daiichi NPS (1F) on March 2011. Evaluation of
the effects of the accident on the environment and assessmentof the radiological dose to the public is the urgent issue thatneeds to be addressed.
Since the accident at the 1F, we have continued to estimate
the release amount to the atmosphere and analyze theatmospheric dispersion of radionuclides using the Worldwide
version of System for Prediction of Environmental Emergency
Dose Information, WSPEEDI. WSPEEDI provides accurate
predictions of the atmospheric movements, dispersion, anddeposition of radionuclides and the external exposed dose ona regional (approximately several 10 km) to a hemispherescale.
By comparing the results of WSPEEDI simulations withenvironmental monitoring data, the rate of radionuclide
release to the atmosphere was estimated. Fig.1-4 shows the
temporal changes in the I-131 and Cs-137 release rates fromMarch 12 to April 5, 2011. The increases in the release rates
on March 12, 14, and 15 correspond to the plant events at the
1F, such as the hydrogen explosion of Unit 1 and Unit 3 anddecreases in the reactor pressure in Unit 2. The release rate
gradually decreased after March 23, implying that thedamaged reactors became stable and the atmospheric releaseof radioactive materials was reduced.
Fig.1-5 illustrates the distribution of the surface depositionof Cs-137 accumulated from March 12 to April 1 calculatedusing WSPEEDI with use of the estimated release rate andmeteorological data. A detailed analysis of the simulation
results revealed that current areas of high surface depositionand dose rate were formed mainly as a result of dry
deposition from March 15 to 16 and wet deposition fromMarch 20 to 21.
The results of the present study were provided to the
Nuclear Safety Commission of Japan and were used by thegovernment in the Report of the Japanese Government to the
IAEA Ministerial Conference on Nuclear Safety.

Reference
Katata, G. et al., Atmospheric Discharge and Dispersion of Radionuclides during the Fukushima Dai-Ichi Nuclear Power Plant Accident. Part I: Source
Term Estimation and Local-Scale Atmospheric Dispersion in Early Phase of the Accident, Journal of Environmental Radioactivity, vol.109, 2012, p.103-113.

12 JAEA R&D Review 2012

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

1-2

130°

132°

Radionuclides Released into the Ocean=Diffusion Simulations for Regions Adjacent to Japan=

134°

136°

138°

140°

142°

144°

146°

148°

150°

152°

154°

130°

132°

134°

136°

138°

140°

142°

144°

146°

148°

150°

152°

154°

130°

132°

134°

136°

138°

140°

142°

144°

146°

148°

150°

152°

154°

46°

46° 46°

46° 46°46°

44°

44° 44°44° 44°44°42°

42° 42°42° 42°42°40°

40° 40°40° 40°40°

38°

38° 38°38° 38°38°36°

36° 36°36° 36°36°34°

34° 34°34° 34°34°

32° 32°

32°
130°

132°

134°

136°

138°

140°

142°

144°

146°

148°

150°

152°

154°

32° 32°
130°

132°

March 16, 2011

134°

136°

138°

140°

142°

144°

146°

148°

150°

152°

154°

32°
130°

132°

134°

136°

March 18, 2011
0.001

0.010

0.100

138°

140°

142°

144°

146°

148°

150°

152°

154°

April 28, 2011
1.000

10.000Concentration (Bq/Ͱ)

Fig.1-6 Concentration of Cs-137 at the sea surface predicted using a simulation assuming direct ocean release andatmospheric deposition
The simulation results illustrate that Cs-137 was deposited on the sea surface along the Fukushima and Miyagi prefectures and theoffshore area northeast of Miyagi prefecture via atmospheric release immediately after the accident. Additional Cs-137 was also
directly released into the ocean from 1F. Cs-137 in the ocean was transported by ocean currents, and the mixing process of the ocean
currents diluted its concentration.

A large amount of radionuclides was discharged into theatmosphere by the TEPCO s Fukushima Daiichi NPS (1F)
accident. Many of them were transferred toward the east anddeposited into the ocean. Moreover, some water used forcooling of the damaged nuclear reactors leaked from theplant and flowed into the ocean. It is important to understandthe actual conditions of the release of radionuclides into theocean in order to assess the influence of the accident on themarine environment. Therefore, an ocean diffusion
experiment based on a numerical simulation was performed
using a general ocean circulation model developed at Kyoto
University and a diffusion model for radionuclides in theocean (SEA-GEARN) developed by JAEA.
To perform a radionuclide migration simulation in theocean, it is necessary to predict the ocean currents. In thisstudy, the accuracy of ocean current data was improved byapplying the four-dimensional variation method, which is ahighly efficient data assimilation technique that assimilates
observational data to a numerical model.
Determining the amount of released radionuclides
presented the most difficulty, because information about thesource term, e.g., the released radionuclides, amounts, andduration, which is essential for the numerical experiments,
was not available. Thus, the source term for the radionuclides
released directly into the ocean from 1F was estimated using
the concentrations of radioactive materials in the ocean
reported by TEPCO. Regarding the atmospheric deposition ofradionuclides, the results calculated with WSPEEDI wereused.
Fig.1-6 shows the concentration of Cs-137 at the sea
surface, as determined using the simulation. A peakatmospheric release occurred on March 15. Cs-137 was
deposited on the sea surface along the Fukushima and Miyagi
prefectures and offshore of the area northeast of Miyagi
prefecture. The concentration of the radionuclides depositedon the sea was diluted by the Oyasio water and then they
moved southward. The radionuclides that arrived in the
Kuroshio region were quickly carried eastward by the
Kuroshio extension, whereas the radionuclides separated
from the Kuroshio extension showed a tendency to stagnatein that region for a while.
A numerical simulation of the long-term dispersal ofradionuclides in the Pacific Ocean will be performed in order
to assess their impact on the marine environment.

Reference
Kawamura, H., Kobayashi, T. et al., Preliminary Numerical Experiments on Oceanic Dispersion of 131I and 137Cs Discharged into the Ocean because of the
Fukushima Daiichi Nuclear Power Plant Disaster, Journal of Nuclear Science and Technology, vol.48, no.11, 2011, p.1349-1356.
JAEA R&D Review 201213

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

1-3

Collection and Provision of Environmental Radioactivity Data
=Publication of Data for Environmental Reconstruction=

Fig.1-7 Radioactivity distribution map system
The system provides intuitive images of radionuclide
concentration and air dose rate for qualitative estimation.
(http://ramap.jaea.go.jp)

Fig.1-8 Radioactivity distribution database system
The system is designed to provide radionuclide
concentration and air dose rate data for detailed analysis.
(http://radb.jaea.go.jp)

Table 1-1 Examples of registered data
This table shows the registered environmental radioactivity data. The registration of other data is also scheduled.
Registered data

Map

Database

Concentration of Radionuclide in Soil Samples and Air Dose Rate at around 2200 Locations within
Approximately 100 km from the TEPCO’s Fukushima Daiichi NPS

registeredregistered

Vehicle-borne Monitoring Survey within Approximately 100 km from the TEPCO’s Fukushima Daiichi NPS

registerednil

Air-borne Monitoring Survey by MEXT and DOE

registerednil

Survey on the Movement of Radioactive Substances in River Water and Well Waternilregistered

Present Depth Analysis of Cs Radioactivity at 77 Locations

nilregistered

JAEA has conducted extensive monitoring work to takeappropriate measures against the environmental consequencesof radionuclides released by the accident at the TEPCO s
Fukushima Daiichi Nuclear Power Station (NPS).
Radioactivity distribution maps and database systems havebeen developed based on the results of this monitoring work.
The radioactivity distribution map system (Fig.1-7)
provides visualization of the distribution of the dose rate in
air and of radionuclide concentration in soil. A zooming-up
function helps to check the distribution status ofenvironmental radionuclides intuitively. The radioactivity
database system (Fig.1-8) provides not only quantitative air
dose rate and radionuclide concentration data but alsodetailed information relating to the measurement methods,
analysis methods, precision of the measured data, and so on.
Examples of registered data are shown in Table 1-1.
Before providing data to the public, access loads of both
systems were evaluated through stress tests. It was found thatthe map system suffered from connection failure owing toheavy loads, because this system was designed to create a
distribution map on demand according to user requests. To
solve this problem, we redesigned the system to provide
maps prepared in advance. Subsequently, the number of
failures was reduced from 20% to 0.000001%. The map
system was accessed by more than 300000 users over the first
10 days with no failures.
The most recent monitoring project has been in operation
since June 2012. Environmental radioactivity data havecontinuously been collected and we intend to make these data
available through maps and database systems.
This research project has been conducted as The 1st and2nd Study on Distribution of Radioactive Substances aroundthe TEPCO s Fukushima Daiichi Nuclear Power Station
funded by the Ministry of Education, Culture, Sports, Scienceand Technology (MEXT).

Reference
The Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT) et al., The Study on Distribution of Radioactive Substances
around Fukushima Daiichi Nuclear Power Plant, The report of the FY2011 Strategic Funds for the Promotion of S&T, 2012, p.1 - 97 - 1 - 104 (in Japanese),
http://radioactivity.mext.go.jp/ja/contents/6000/5235/26/5600_201203131000_report1-2.pdf

14 JAEA R&D Review 2012

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

1-4

How Far Did Radionuclides Diffuse from the Accident?
=Measurement of Atmospheric Anthropogenic Radionuclides Using the CTBT IMS Network=

109

Activity Concentration (ǴBq/m3)

108

All

Cs-137

I-131

Te-132

107
106
105
104
103
102
101

Fiji (Nadi)

Malaysia (Kuala Lumpur)

Papua New Guinea (New Ireland)

Cameroon (Douala)

Mexico (Guerrero Negro)

Panama (Panama City)

Kiribati (Kiritimati)

France (Pointe-a-Pitre, Guadeloupe)

Philippines (Tanay)

France (Kourou, French Guiana)

Kuwait (Kuwait City)

Germany (Schauinsland, Freiburg)

Russia (Zalesovo)

Mauritania (Nouakchott)Sweden (Stockholm)

Portugal (Ponta Delgada, Azores)

U.S.A. (Ashland, KS)

Mongolia (Ulaanbaatar)

Iceland (Reykjavik)

Norway (Spitsbergen)

Japan (Okinawa)

Canada (St. John's, N.L.)

U.S.A. (Upi, Guam)

Russia (Kirov)

U.S.A. (Wake Island)

Russia (Dubna)

Russia (Ussuriyk)

Canada (Vancouver, B.C.)

U.S.A. (Charlottesville, VA)

Russia (Petropavlovsk-Kamchatskkiy)

Canada (Yellowknife, N.W.T.)

U.S.A. (Salchaket, AK)

U.S.A. (Midway Islands)

Canada (Resolute, NU)

U.S.A. (Sacramento, CA)

U.S.A. (Oahu, HI)

U.S.A. (Melbourne, FL)

Japan (Takasaki)

U.S.A. (Sand Point, AK)100

Fig.1-9 Comparison of the integral activity concentration of particulate radionuclides detected at each IMS station
Data for each station is presented in descending order of activity concentration for all radionuclides. It can be seen that the activity
concentration has a tendency to be high in North America, Russia, and Japan and low in equatorial regions and the Southern
Hemisphere.

The Comprehensive Nuclear-Test-Ban Treaty (CTBT) bans
all nuclear explosions. Although it has not been in effect yet,the International Monitoring System (IMS) is underconstruction around the world and approximately 80% of thesystem is provisionally operated in order to monitor nuclear
tests. The IMS network consists of 337 monitoring stations,
including 80 radionuclide (RN) stations. In Japan, we havealready built IMS RN stations for the continuous monitoringin Takasaki, Gunma and Onnason, Okinawa.
The accident at the TEPCO s Fukushima Daiichi NPS (1F)in March 2011 emitted copious amounts of anthropogenic
RNs into the environment. The RNs emitted into the ambient
air were detected by the IMS RN stations. Part of theradioactive plume containing particulate RNs derived fromthe accident diffused eastward owing to westerly winds andbegan to be detected at the IMS stations one after another on
the North American and Eurasian Continents. Eventually,
the particulate RNs were detected at 39 out of 63 operating

stations: all stations in the Northern Hemisphere and two
stations in the Southern Hemisphere. Fig.1-9 shows acomparison of the integral activity concentration of threemajor RNs and all of the anthropogenic RNs found at eachstation from the detection start date to September 30, 2011,
when the detection subsided. Although the integral activity
concentration itself means nothing, it is used as a guide to
determine any tendencies in the spread of RNs among the
IMS stations.
As mentioned above, the network of IMS RN stations
illustrates the global extent of the RN contamination in not
only the Northern Hemisphere but also a part of the Southern
Hemisphere. Therefore, in addition to its primary role, the
IMS network contributed greatly to the measurement of theanthropogenic RNs released from the accident at 1F.
The present study was sponsored by the Japan Institute ofInternational Affairs (JIIA).

Reference
Yonezawa, C., Yamamoto, Y., Measurement of Anthropogenic Radionuclides in the Atmosphere with a Radionuclide Monitoring Network for Nuclear
Tests, Bunseki, vol.2011, no.8, 2011, p.451-458 (in Japanese).
JAEA R&D Review 201215

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

1-5

Detailed Investigation of Radionuclide Distributions in the Environment=Construction of Soil Deposition Map and Dose Rate Map=

10

0k

m

80

km

3

Minamisoma-city 0km
20km

The TEPCỎ
s
Fukushima
Daiichi NPS

Activity per area
Cs-137㧔Bq/㨙2㧕
3000 k㧨1000 k - 3000 k
600 k - 1000 k
300 k - 600 k
100 k - 300 k
60 k - 100 k
60 k
30 k 30 k
10 k ҇10 k

Iwaki-city

Kitaibaraki-city

0

20

40 km

Fig.1-10 Cs-137 deposition map
Ranges of activity per area are shown by colors according to the results averaged over five samples collected at one
location and normalized to that on June 14, 2011.

In the accident at the TEPCO s Fukushima Daiichi NPS,
large amounts of radionuclides were released into theatmosphere and deposited on ground over wide areas. To
properly estimate the effects of released radionuclides on theenvironment and on human health over long time periods andimplement appropriate countermeasures, it is necessary toobtain detailed information about regional distribution ofdeposited radionuclides and dose rates in air. Since theaccident occurred, a large number of environmental
monitoring data have been accumulated by many organizations,
including the Ministry of Education, Sports, Science andTechnology (MEXT). However, variations in method,
accuracy, location, and dates exist between different sets ofmonitoring data, making it difficult to integrate the data into
maps.
Considering these factors, both MEXT and researchers
suggested that radionuclide deposition maps should beconstructed as soon as possible based on reliable and largescale monitoring activities. Thus, JAEA conducted a mapping
project in collaboration with many universities and research
institutes. More than 400 people took part in soil sample
collection and more than 300 people participated in analysisof the collected samples.
Radionuclide deposition maps were constructed for Cs-137,

Cs-134, I-131, Te-129 m, Ag-110 m, Pu-238, Pu-239+240,
Sr-89, and Sr-90 according to analyses of soil samples
collected at about 2200 locations. Furthermore, a dose rate
distribution map at 1 m height (based on measurements at
soil sampling locations) and dose rate distribution maps forroads based on carborne surveys conducted over 17000 km
were also prepared.
Fig.1-10 presents the completed Cs-137 deposition map,
indicating Cs-137 activity per ground area (Bq/m2). It is clearthat the region to the northwest of the Fukushima site ishighly contaminated and that the middle part of Fukushima
Prefecture (referred to as Nakadori) is highly contaminated.
Furthermore, we were able to construct a novel I-131
deposition map, even though statistically significant data
were obtained for I-131 at fewer locations than for cesium.
These maps have been utilized as basic data to evaluate theconsequences of the accident and related countermeasures.
JAEA will continue to manage a database, which will keep
the obtained data secure; JAEA will also provide the data indesired formats and contribute to the investigation ofcountermeasures.
The work was implemented under commission of the
Ministry of Education, Culture, Sports, Science andTechnology of Japan (MEXT).

Reference
The Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT) et al., The Study on Distribution of Radioactive Substances around
Fukushima Daiichi Nuclear Power Plant, The report of the FY2011 Strategic Funds for the Promotion of S&T, −Report on Construction of Maps Indicating
Air Dose Rate Distribution etc.−, 2012, p.1-1-1-64 (in Japanese), http://radioactivity.mext.go.jp/ja/contents/6000/5235/25/5600_201203131000_report1-1.pdf

16 JAEA R&D Review 2012

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

1-6

Radiation Dose Map of Japan Using an Aircraft=Visually Grasping the Influence of the Accident=

Data logger

Air dose rate at 1 m above
ground level (ǴSv/h)
Converyed info the value
as of May 31.2012
㧨19.0 ‫ޓ‬
9.5 㧙 19.0
3.8 㧙 9.5
1.9 㧙 3.8
1.0 㧙 1.9
0.5 㧙 1.0
0.2 㧙 0.5
0.1 㧙 0.2
҇ 0.1 ‫ޓ‬

Detectors

Fig.1-11 Airborne radiation monitoring system
The airborne radiation monitoring system operates to save thedata that synchronizes the calculation rate of the large NaI
detector and positional information with the GPS every second.

The TEPCỎ
s
Fukushima
Daiichi NPS
A boundaryline of 80 km

Fig.1-12 Radiation dose map for the whole of Japan
We installed a large NaI detector on a plane, measured NaI
concentration at intervals of 3~5 km (or 1.8 km in the vicinityof the nuclear power plant), and converted the data collected
into dose rates at a height of 1 m above the ground. The
influence of the natural radiation is included in this map.
0 50 100

200

300

400

500

km

Since the accident at the TEPCO s Fukushima Daiichi
NPS (1F), a method of environmental radiation monitoringthat can quickly measure large areas has been required. It isalso desirable for the results of this monitoring to bevisualized. We conducted monitoring of the radiation in
conjunction with JAEA, using a plane (helicopter) to grasp
the full extent of the effects of the accident.
For this airborne radiation monitoring, we installed large
detectors for γ-rays (NaI detectors) on a helicopter andmeasured the count rate of γ-rays, which we synchronizedwith the location and altitude information collected by GPS(Fig.1-11). The survey was conducted at 3 km intervals ineastern Japan (1.8 km intervals in the vicinity of 1F and at
5 km intervals in western Japan and Hokkaido. The data wegathered in this manner were compared with the datameasured on the ground; then, we corrected for the influenceof factors such as altitude and converted the data into dose
rates at 1 m above ground level. We deducted backgroundnatural radionuclides based on an energy spectrum of the
gamma beam that we measured at the same time; we werealso able to convert the dose rate into rates of radiocesium
deposition (Bq/m2).
The dose rate map is shown in Fig.1-12. Because
measurement days vary according to location, an equivalencemap (dose rate and radiocesium) was created as a result ofapplying decay correction to data from May 31, 2012. A highdose area was found to extend from 1F in the northwest towardTochigi and Gunma. In addition, a relatively high-dose area wasfound in southern Ibaraki and northern Miyagi. It is thought thatradioactive Cs released by 1F was deposited in these areas.
Conversely, relatively high doses were found in the central partof Chugoku District and Kyushu. However, these areas wereshown to have been influenced by the background radiation.
This map is available to the public on the homepage of the
Ministry of Education, Culture, Sports, Science and Technology(http://radioactivity.mext.go.jp/ja/contents/6000/5847/view.html).
This work was carried out under a contract with the
Ministry of Education, Culture, Sports, Science andTechnology of Japan (MEXT) in fiscal years 2011 and 2012.

Reference
Torii, T., Sanada, Y. et al., Report of Investigation of Radionuclide Distribution using Aircraft for Surrounding Environmental Survey from Fukushima
Daiichi Nuclear Power Plant, JAEA-Technology 2012-036, 2012, 182p. (in Japanese).
JAEA R&D Review 201217

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

1-7

Effects of Forest Decontamination on the Reduction in External Exposure Doses=Field Demonstration of Forest Decontamination=
Before decontamination
Coniferous tree area

Measurement points

10

10

3
one

Z
Broad-leaved tree area

20 m

e2
Zon 1
e
Zon

ԛ

m

Ԛ

After decontamination

m
0m

1

ԙ

Ԙ

© AAS & NNK
Fig. 1-13 Aerial view of test location
Decontamination tests were conducted in the following order:
Zone 1, Zone 2, Zone 3. Each zone is arranged parallel to theforest boundary and has a width of 10 m. Air dose rates at theforest boundary were measured after completing eachdecontamination work.

Fig. 1-14 Before and after photos of the test location
In the broadleaved tree area, grass mowing, raking of fallen
leaves, and removal of the litter layer were applied as
decontamination methods.

Table 1-2 Forest decontamination depths measured from the forest edge and the resultant reductions in air dose rates (1 m)
at the forest edge
A 50% reduction in the air dose rate at the forest boundary was obtained by decontamination in Zone 1, except at measurement point(4), but no significant effects were obtained by further decontamination in Zones 2 and 3.
(ǴSv/h)Area

Measurement
point
numberAfter decontamination of Zone 1

After decontamination of Zones 1 and 2

After decontamination of Zones 1, 2, and 3

Before decontamination(Air dose rates
Mowing and fallen leaves removal Litter layer removal Branch trimming at forest edge Mowing and fallen leaves removal Litter layer removal Mowing and fallen leaves removal Litter layerremoval
at a height of 1 m)
Surface dose rates㧖1
Air dose rates (at a height of 1 m)

Forest edge
(coniferous
tree area)

Ԙ

2.6

2.2

1.4

1.3

1.2

1.3

1.3

1.2

ԙ

2.5

2.3

1.6

1.4

1.5

1.4

1.2

1.3

Forest edge
(broad-leaved
tree area)

Ԛ

2.4

1.7

1.4

1.5

1.4

1.4

1.6

2.2

2.2

1.5

1.9

ԛ

2.7

2.3

2.0

-㧖2

＊

1 The dose rate values for Zone 1 after mowing and collection of fallen leaves reflect measurements taken at a height of 1 cm above the ground surface. The values at a height of 1 m would
be approximately 80% of those at a height of 1 cm resulted from the average ratio of the values at a height of 1 m to those at a height of 1 cm at other measurement points.
＊
2 For broadleaved trees, no branch trimming was implemented because all the leaves had fallen.

One of the most urgent issues following the accident at theTEPCO s Fukushima Daiichi NPS was the remediation of theextensive areas contaminated by the radioactive materials
discharged. In response to this situation, JAEA performed
decontamination experiments at two test sites located in the
Date and Minamisoma municipalities, combining a range ofbuildings and different types of land use, in order toaccumulate knowledge and data for full-scale decontamination
activities to be performed by local governments. We
established a plan involving practical decontamination
methods that can be easily implemented at each siteaccording to the decontamination targets (e.g., forests,
agricultural land, residential buildings, and roads). Here, we
introduce the results of this field demonstration of forest
decontamination.
It is considered important to reduce the exposure doses ofpeople living in houses adjacent to contaminated forest, sincemost of Fukushima is covered with forest. In addition, it isnecessary to obtain data related to the effectiveness of
decontamination methods and appropriate decontamination
conditions, e.g., the determination of the distance from theforest boundary that is sufficient for the development ofeffective and efficient decontamination.
To this end, mowing of grass, removal of fallen leaves andthe litter layer, and trimming of branches were conductedsequentially in Zone 1 (up to 10 m from the forest boundary)of two areas: one covered with coniferous trees and anothercovered with broadleaved trees. Air dose rates at the forest
boundary were measured after completion of eachdecontamination activity. This process was extended to a
distance of 30 m from the forest boundary in increments of
10 m (i.e., Zone 2 and then Zone 3), but no branch trimming
was implemented for Zones 2 and 3 because trimming of all
the trees in these zones had been already done before theexperiment (Fig.1-13, Fig.1-14).
As shown in Table 1-2, litter layer removal was mosteffective in reducing the air dose rate at the forest boundary.
Overall, up to 50% reduction in the air dose rate at the forest
boundary was obtained by decontamination in Zone 1,
although no significant effects were obtained by further
decontamination in Zones 2 and 3.
These results indicate that removal of litter layer within adistance of 10 m from the forest boundary can be mosteffective in reducing the exposure doses of people living inhouses adjacent to a contaminated forest.
This work was commissioned by the Cabinet Office,
Government of Japan.

Reference
Japan Atomic Energy Agency, Decontamination Model Projects to Establish Guidelines of Municipal Remediation Work for Residential Areas
Contaminated with Radioactive Materials Discharged from the Fukushima Daiichi Nuclear Plant, 2012, p.166-168 (in Japanese), Available at JAEA Library,
http://www.jaea.go.jp/fukushima/kankyoanzen/guideline_report.html

18 JAEA R&D Review 2012

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

1-8

Creating Decontamination Guidelines Based on Practical Actions=Decontamination Pilot Project=

Fig.1-15 Shot blasting(the Okuma town government office area)
Shot blasting, as the term indicates, involves blasting by small
iron balls (typically around 1 mm in diameter) to remove a thinlayer of the flat surface. We applied this technique, which has
been used in nuclear facilities, to pavement, and found that it
reduced the surface contaminant density at Okuma town by
more than 90%.

Fig.1-16 Guide to decontamination work
(extract from the report)
We evaluated each decontamination method from the viewpointof cleanup efficiency, waste treatment, and effectiveness.

The central government and local authorities of Japan havebeen planning and undertaking decontamination to ensure
remediation of environments contaminated with radioactivity
after the accident at the TEPCO s Fukushima Daiichi NPS.
For this unprecedented environmental pollution in Japan,
guidelines for swift and effective decontamination are
required to provide technical information that describes how
to perform a series of decontamination processes including
dose rate measurement, removal of contaminated soil,
cleanup of roads and buildings, safe storage of wastes
generated, and protection of workers from radiation. We,
under contract with the government, were in charge of thedecontamination pilot project for this purpose in 11 cities,
towns, and villages within the restricted area (i.e., areas withannual dose rate >20 mSv) and planned evacuation area. The
area of the project sites totaled to more than 200 ha. The
obtained data were analyzed and interpreted, and the related
documents were filed at the end of June, 2012.
A variety of existing techniques were applied in the pilot
project depending on the materials to be decontaminated(buildings, farmland, road, forest), the topography, and thestructure. The applicability of the techniques employed wasevaluated in terms of cleanup efficiency (speed, timerequired), waste treatment (volume of waste generated bycleanup, recovery of used water), and effectiveness (dose rate
reduction, decontamination factors) (Fig.1-15, Fig.1-16).
We introduced a computer simulation system known as the
Calculation System for Decontamination Effect (CDE),

which calculates dose rates of the area of interest on the basisof the surface contamination density. Therefore, CDE can beused to predict dose rates after decontamination for given
values of decontamination factor; in other words, CDE
provides information for decontamination planning, such as
indicating the amounts of decontamination factor required forcertain areas. The applicability of the system wasdemonstrated in our pilot project, indicating that CDE could
be a potent tool for decontamination planning. CDE wasdeveloped by JAEA and can be downloaded from the JAEA
website (Topic 1-12).
As part of the decontamination pilot project, we designed
and constructed temporary storage structures for waste
generated by decontamination. The topography of the site,
isolation with impervious sheets, and radiation shielding withsandbags were taken into account in the design, in addition to
the configuration or positioning of combustible/noncombustible
wastes. We conducted periodic inspection to confirm normal
function; this involved measuring radiocesium concentrationsin leachate tanks, surrounding dose rates, and temperature.
The results of our pilot project were reported at the
progress debrief session in Fukushima City on March 26,
2012 (Reference). The pilot project was finished at the end ofJune. Slides from the presentation at the March 26 session as
well as the final documents are available from JAEA s
homepage.
This work was commissioned by the Cabinet Office,
Government of Japan.

Reference
Japan Atomic Energy Agency, The Decontamination Demonstration Project for the Evacuation Zones Affected by the Accident at the Tokyo Electric Power
Company Fukushima Daiichi Nuclear Power Station【Decontamination Pilot Project】Report, 2012, 667p.(in Japanese), Available at JAEA Library,
http://www.jaea.go.jp/fukushima/kankyoanzen/d-model_report.html
JAEA R&D Review 201219

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

Cleanup of Cesium-Contaminated Soil
=Investigation of Removal of Radiocesium from Soil by In Situ Ground Heating=

20

600 ͠
800 ͠

‫ޓ‬㧦SiO2

1300 ͠

‫ޓ‬㧦NaAlSi3O81000 ͠
1300 ͠

0‫ޓ‬㧦CsAlSiO4

Intensity (a.u.)

Rate of change in the amount of radiation (%)

1-9‫ޓ‬㧦CsAlSi2O6
or CsAlSiO41000 ͠

‫ޓ‬㧦Cs2CO3㨯2H2O

800 ͠
Before heating

㧙20

5

10

30

10

60

Heating time (min)

20

30

40

50

2ǰ (deg)

Fig.1-17 Rate of change in the amount of radiation
Rate of change in the amount of radiation was calculated fromthe measurement of radiocesium activity in the soil before andafter heating.

Fig.1-18 X-ray diffraction patterns of the soil at eachheating temperature
Stable compounds, e.g., CsAlSiO4, were formed by thereaction of NaAlSi3O8, SiO2, and so on, during heating of
Cs-contaminated soil.

1200

Temperature (͠)

soil1
1000

soil2

800

soil3
soil4

600

soil5
400

soil6

Fig.1-19 Changes of temperature at soil surface
More than 15 min of heating was necessary to reach
relatively high soil temperatures using a portableburner. Maximum temperature was about 700 ℃.

200
0
0

5

10

15

Time (min)After the accident at the TEPCO s Fukushima Daiichi NPS(1F), an extensive area was contaminated with radionuclides
from 1F; accordingly, various methods of remedying suchcontaminated soil have been investigated. Because theprimary material contaminating soil is radiocesium, it has
been postulated that such contamination could be removed byadopting a heating method that focuses on the high volatilityof cesium (Cs) compounds at relatively low temperatures. To
investigate the possibility of removal by in situ groundheating, the behavior of radiocesium in the soil environment
was evaluated using an electric furnace. A burner was used
for in situ ground heating because it avoided the need to
introduce large machinery.
Fig.1-17 illustrates the rate of change in the amount of Cs-137
in response to heating using the electric furnace. The rate ofchange of the amount of radiation varied from -12.0 to+13.5% in response to heating to temperatures of 600~1300 ℃,
with no noticeable volatility of radiocesium.
To confirm the reactivity of the soil and cesium due toheating, the crystal structure of a mixture of soil samples andcesium carbonate reagent (weight ratio 2:1) was analyzed
after heating. The results indicate that stable compounds suchas cesium aluminum silicate (CsAlSiO4) were formed by thereaction of sodium aluminum silicate (NaAlSi3O8), silicon
dioxide (SiO2), and others during heating of Cs-contaminatedsoil (Fig.1-18).
Fig.1-19 illustrates changes in the temperature at the soil
surface as a result of heating with a burner. More than 15 minof heating was necessary to achieve a relatively high soil
temperature with the portable burner, and the maximum
temperature of the soil was about 700 ℃. Moreover, it was
found that this technique resulted in spot heating; aconsiderable amount of time was required to heat a wide area.
Although it is possible to heat the soil to relatively hightemperatures, radiocesium would then form stable compoundswith materials in the soil, causing the radiocesium to remainin the soil.

Reference
Japan Atomic Energy Agency, Investigation of Radiocesium Volatilization from Soil by In-Situ Ground Heating, JAEA-Research 2011-026, 2011, 74p.
(in Japanese).

20 JAEA R&D Review 2012

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

1-10 Decontamination of Soil Using a Poly-Ion Complex and Clay
=Decontamination of Soils Polluted by Radioactive Cesium and Prevention of Recontamination=

Fig.1-20 Solidification of soil flakes
After spraying a poly-ion solution on field soil
followed by drying, the soil was solidified andcould be removed as flakes.
Fig.1-22 Decontamination test in Date-City, Fukushima
Test of the decontamination technique using the poly-ion complex was
performed in Oguni, Ryozen, Date-city, Fukushima in July 2011.

Fig.1-21 Swelling of the poly-ion complex
Dried poly-ion complex gel is swollen owing to water absorption.

The poly-ion complex and clay method (PCCM) is asurface soil treatment technique for the decontamination ofsoil (removal of surface soil) and the prevention ofrecontamination of the cleaned soil (containment ofradioactive cesium in heavily polluted soil). This method iseffective for long-term decontamination of broad areas suchas fields and village forests, where decontamination withtime and care is needed.
In the PCCM, the poly-ion complex is a gel material
formed by mixing cationic and anionic polymers (called
poly-ions ) that can solidify and stick to surface soils andinhibit soil dust discharge and mud water runoff, while theclay strongly adsorbs and firmly fixates cesium ions and canprevent radioactive cesium from migrating and diffusing into
deeper soil layers. Both the poly-ions and clay are non-toxic
and safe substances. In addition, large quantities can beprocured, and their use was demonstrated following theChernobyl nuclear power plant accident. Notably, the polyions cannot firmly fixate cesium, while the clay is easily
dried and broken and thus can fly apart as dust. Therefore,
the PCCM combines both poly-ion complexes and clay inorder to overcome the limitations of each.
The most important feature of the PCCM is its long-term
durability in the environment. The poly-ion complex is solidwhen dried. Although it becomes soft and swollen when
exposed to water, it does not flow out from the soil owing torainfall because it is very sticky and has the properties of thegel. In addition, the poly-ion complex can effectively inhibit
soil dust discharge in both dry and wet areas. With thesecharacteristics, radioactive cesium in heavily polluted soil canbe contained for a long time, and thus, recontamination withsoil dust and mud water, even in heavily polluted areas, canbe avoided. Fig.1-20 presents a photograph of soil flakes
solidified with the poly-ion complex. Fig.1-21 shows thepoly-ion complex dried (lower right) and swollen after
absorption of water (upper left). Meanwhile, the clay canfixate radioactive cesium for a long time. In addition, a clay
suspension can draw up cesium from deeper soil layers to thesurface during the course of drying as a result of thephenomenon known as suction. Suction is the effect of
capillary action as the suspended solution percolates into thesoil voids. When the fine clay particles in the surface layer
are exposed to sunlight and wind, they dry and shrink, andthe wet and swollen particles in the suspension within the
deeper soil layers can be siphoned off to the surface. The
swollen fine particles that percolate into the deeper soil layers
can trap radioactive cesium that is weakly bound by coarse
non-clay soil particles and bring radioactive cesium to thesurface via suction. The soil decontamination technique using
the poly-ion complex was tested in Oguni, Ryozen, Date-city,
Fukushima in July 2011. The grounds of two community
centers, 1500 m2 in total area, were decontaminated byremoving a 2 cm soil surface layer, and 85~90% of theradioactive cesium was cleared off (Fig.1-22).

Reference
Naganawa, H., Removal of Radioactive Cesium from Contaminated Soil by using Poly-Ion Complex with Clay, Nendokagaku, vol.50, no.2, 2011, p.52-57(in Japanese).
JAEA R&D Review 201221

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

1-11 Decontamination of Playground Lot=Trial Decontamination and Its Effects on Air Radiation Dose Rate=i h g f e d c b a
1
2
3
4
5
6
7
8
9
10
11
12
13

Measurement Point
Decontamination
Not Decontamination

Fig.1-23 Decontaminated playground lot
and air radiation dose rate monitoring
posts
The playground equipment was wiped off
using paper towels. Surface soil was removed
and the area was backfilled with clean soil.
Side gutters were washed with a scrubbing
brush and high-pressure water jet. The white
2 × 2 m lattice in the photograph and circles
in the attached drawing (red in the area of
decontamination, blue outside) indicate themonitoring points.

Table 1-3 Air radiation dose rate before and after decontamination
The air radiation dose rate was measured at heights of 1 m, 50 cm, and 1 cm above the ground surface. Values were high close to theground surface (1 cm) and lower at higher positions before decontamination as a result of soil contamination. Decontamination
reversed this trend, such that the higher positions exhibited higher dose rates. Although surface soil contamination was removed, thecontribution of radiation from the surroundings dominates the remaining dose rate and accounts for the reversal.

Measurement Height

Before (ǴS v/h)After (ǴS v/h)

Ave.̪

Max.

Min.

Ave.̪

Max.

Min.

1m

1.5

1.9

1.0

0.6

1.1

0.4

50 cm

1.8

2.4

1.2

0.6

1.0

0.3

1 cm (without shielding)

2.4

5.0

1.6

0.4

1.1

0.3

1 cm (with shielding)

-

-

-

0.2

0.4

0.1

̪Average value of 2 m×2 m measurements mesh in park

After the nuclear accident at the TEPCO s Fukushima
Daiichi NPS, living spaces were contaminated withradionuclides and many people were forced to evacuate andlive in anxiety. Children, who are sensitive to radiation,
were able to play outdoors only for limited periods.
Decontamination of schoolyards has been prioritized and has
progressed rapidly on the basis of investigation by JAEA in
May 2011. However, playground lots in residential areas
were not necessarily prioritized in the same way. To open andexpand outdoor spaces for children, we conducted a trial
decontamination of a playground lot in a residential area inthe city of Fukushima. Data acquisition was conducted to
study the available techniques and their effectiveness incleanup efforts and to investigate the effectiveness of areal
decontamination in reducing air radiation dose rate. The trial
also served another purpose: it allowed nonprofessionals andthe general public to learn how to decontaminate theirresidential areas using commercially available tools andcommodities.
A relatively small (150 m2) playground lot was chosen for
the trial decontamination; playground equipment, groundsurfaces, and side gutters were of interest in this trial. The
selected playground lot had no enclosing fences and did notoffer any protection from radiation coming from thesurroundings (Fig.1-23).
Decontamination work proceeded from higher to lower
locations, i.e., from playground equipment such as slides,
swings, and horizontal iron bars to soil and side gutters.
Three centimeters of surface soil were removed and the areas

were backfilled with clean soil. Side gutters were washedwith a high-pressure water jet (7 MPa). Generated wastes
such as removed soil, contaminated tools, and commodities
used for decontamination were buried in an underground pit
and covered with 30 cm of clean soil.
Decontamination produced several notable effects. The airradiation dose rate at 1 cm from the ground surface was reduced
from an average of 2.4 µ Sv/h to 0.4 µ Sv/h, as shown in
Table 1-3. When the instruments were equipped with leadshielding, the radiation dose rate was further reduced to
0.2 µ Sv/h; this rate represents radiation from the groundsurface only, and the difference (i.e., between 0.4 µ Sv/h and0.2 µµSv/h) represents the contribution from the surroundings.
This observation indicates that decontamination of a wider
area (i.e., areal decontamination) is required to significantly
reduce the air radiation dose rate.
The soil cover of 30 cm shielded the radiation emitted byburied waste such that the buried waste did not increase the
air radiation dose rate in its vicinity.
The techniques described above were incorporated into the
Ministry of the Environment s decontamination guidelines
and the Cabinet Office s decontamination catalog, to whichpeople who are involved in cleanup work can refer.
Video clips of this trial decontamination work can be foundon JAEA s homepage (h t t p : / / f u k u s h i m a .jaea.go.jp/en/ ).
These clips were presented at Fukushima Prefecture
decontamination seminars and are available at Josen Plaza of
the Ministry of the Environment in the city of Fukushima(http://josen-plaza.env.go.jp/).

Reference
Tagawa, A., Effect of Ground Surface Decontamination on the Air Radiation Dose Rate −Results of a Decontamination Trial at a Playground Lot in a
Fukushima Residential Area−, Nippon Genshiryoku Gakkai Wabun Ronbunshi, vol.11, no.2, 2012, p.111-117 (in Japanese).

22 JAEA R&D Review 2012

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

1-12 Tool for Planning Rational and Effective Decontamination=Development of Computer Software for Efficiently Predicting Dose Rates after Decontamination=
Set input data(radioactivity, decontamination
method)(1) Database on
dose contribution
per radioactivity(built-in)

1

2
Set geometry data

3
Repeat execution
after changing
decontamination method
Dose rate(2) Radioactivity
distribution(input)Executecalculation

Dose ratewith the updated method“Before”(3) Dose rate
distribution(output)

For the reduction of dose rates in residential areas, it isnecessary to remove the radioactive cesium deposited on the
ground. Thus, a computer software program named
Calculation System for Decontamination Effect (CDE) has
been developed for planning the rational and effective
decontamination of areas through the quick and accurate
evaluation of dose rate reduction.
To achieve such a quick and accurate evaluation, thefollowing calculation method was developed. First, the dose
contributions per unit radioactivity to the surrounding region
were evaluated using the three-dimensional particle transport
code PHITS, which accurately treats the absorption andscattering reactions in air and on the ground, and the results
were compiled in a response matrix constructed as a built-in
database (Fig.1-24 (1)). Next, by multiplying the radioactivity
distribution (Fig.1-24 (2)) by the response matrix, the dose
rate distribution is calculated (Fig.1-24 (3)). While significant
computation time is required to obtain accurate calculation
results with PHITS, the CDE completes the calculation in aHighχ
Dose rate
ω
Low“After”“After”
Fig.1-24 Concept of the dose calculation method in the
CDE
Multiplying a user-input radioactivity distribution by the dose
contributions per unit radioactivity to the surrounding region,
which have been compiled in a built-in database, the CDE
quickly and accurately calculates the dose rate distribution.
By reconstructing the radioactivity distribution after
decontamination according to the decontamination factors
input by the user, the CDE similarly predicts the dose rate
distribution after decontamination.

4

The lower-right figure indicates that thechanging decontamination methods
extends the area of lower dose rate.

Fig.1-25 CDE for establishing a decontamination plan
The CDE runs on Microsoft Excel. Once the radioactivity
distribution and decontamination methods are set through its
graphical user interface, the CDE visualizes the dose rate
distributions before and after decontamination. It is applicableto the planning of decontamination through the use of efficient
case studies involving changing decontamination methods.

few seconds with comparable accuracy to PHITS. Therefore,
the CDE successfully achieves a drastic reduction in thecomputation time.
The CDE runs on Microsoft Excel with a graphical userinterface for easily proceeding through the evaluation steps,
such as setting the input data, executing the calculation, andvisualizing the evaluation results on a personal computer, as
shown in Fig.1-25. Basic operations, such as copying digital
data and saving evaluation results, are performed with thefunctions in Microsoft Excel.
The CDE has been distributed free as an open source
software program on the Japan Atomic Energy Agency website
beginning on November 2, 2011 (http://nsed.jaea.go.jp/josen/
en/index.html). The number of users was 452 at the end of
March, 2012. The CDE was also utilized to predict the dose
reductions used to assess the possible decontamination methods
that could be used in Cabinet Office decontamination model
projects for rural residential areas and evacuated zones.

Reference
Satoh, D., Kugo, T. et al., Development of Calculation System for Decontamination Effect, CDE, JAEA-Research 2012-020, 2012, 97p. (in Japanese).
JAEA R&D Review 201223

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

1-13 A Simple Method for Selective Measurement of Cesium 134 and 137=Utilization of Common NaI(Tl) Scintillation Spectrometers=
9600
720 keV
662 keV 730 keV 850 keV

300

100
0
500

Cs

134

7

400

200

137

8

Count rate (cps)Counts (count)

500

137

134

Cs

base line

Cs

5
4
3
2
1

base line

0

600

700

800

900

Cs

6

1000

0

200

400

600

800

1000

Values obtained with Ge detector (Bq)

Energy (keV)
Fig.1-26 Scheme for selective quantification of 134Cs and 137Cs
The colored parts were specified for use in the analysis in thismethod. The regions under the baselines were considered as parts
that represent γ-rays from natural radiation sources and wereomitted.

Fig.1-27 Correlation between nuclide concentrations instandard 100 mL-container and areas of specified parts
It was confirmed that the areas of the specified parts areproportional to concentrations of 134Cs and 137Cs in dilution
series.

Table 1-4 Results of environmental samples
Measurement was conducted in January-February 2012. It should be noted that the abundances and their ratios change because
both nuclides decay year by year according to their own half-lives.
137Cs

134Cs

NaI (TI) spectrometer(Bq kg-1)

Ratio to resultswith Ge detector

NaI (TI) spectrometer(Bq kg-1)

Ratio to resultswith Ge detector

Leaf mold

6.0×103 ±60

0.98

4.4×103 ±38

0.95

Grassland soil

1.7×104 ±72

Cow manure

5.9×102 ±11

0.97

4

1.3×10 ±47

0.98

0.87

4.8×102 ±7.4

1.06

There is an increasing demand for analysis of the huge
numbers of foods, soils, fertilizers, and so on contaminatedby radiocesium (134Cs, 137Cs). The NaI(Tl) scintillation
spectrometer (NaI(Tl) spectrometer) andgermanium
semiconductor detector (Ge detector) are typical devices used
for this purpose. The NaI(Tl) spectrometer has many merits,
such as its lower price, smaller size, and easier handlingcompared to the Ge detector; as a result, its use is becoming
common. However, it has a critical disadvantage in that it
cannot distinguish completely between the γ-rays emitted
by 134Cs and 137Cs. In this study, we aimed to establish a
simple and open method by which producers and consumers
can measure amounts of 134Cs and 137Cs separately using a
NaI(Tl) spectrometer. We consider it is important to provide acommon basis for more reliable analysis for use by many
nonprofessional people.
Common NaI(Tl) spectrometers exhibit energy spectrawith overlapping gamma ray peaks for 134Cs and 137Cs when acontaminated sample is analyzed (Fig.1-26). We determined
two regions within the peaks, each specific to either 134Cs or

137

Cs (Fig.1-26, colored parts). It was theoretically estimatedthat, approximately, the area of each region is affected byneither the other nuclide nor 40K, a natural radionuclide. The
output of the spectral data file is available as a spreadsheet inmany commercially produced NaI(Tl) spectrometers. We
summed the total counts in the specified regions andcalculated the count rate (cps) by dividing the total counts bymeasurement time (seconds), and found that the results
accurately reflected the amounts of 134Cs and 137Cs in thesample (Fig.1-27). Several environmental samples were
subjected to this method and the results exhibited highconsistency with values obtained using a Ge detector for therespective nuclides; this demonstrates the applicability of our
method (Table 1-4).
The linearity of the dilution series (Fig.1-27) and theabundance ratio between 134Cs and 137Cs (Table 1-4) may
reflect the accuracy of the measurements; thus, users cancheck their preparation procedures. We expect the method
described here to contribute to reliable analysis conducted bya wide variety of people.

Reference
Yin, Y.-G. et al., A Simple Method for Selective Measurement of Cesium 134 and Cesium 137 by NaI(Tl) Scintillation Spectrometer, Nippon Dojo
HiryoGaku Zasshi, vol.83, no.3, 2012, p.296-300 (in Japanese).

24 JAEA R&D Review 2012

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

1-14 Fibrous Grafted Adsorbents Help in Cleanup of Contaminated Water=Development of Fibrous Adsorbent for Removal of Radioactive Cesium=

134

SampleEffluent conc. of Cs (C)Feed conc. of Cs (Co)

1.0
0.8

Cs

137

Cs

Total

No.1

14

20

34

No.2

19

24

43

No.3

17

22

39
(Bq/Ͱ)

0.6

Grafted adsorbent for Cs removal

filtered with membrane(pore size 0.45 µm)

0.4
134

Breakthrough capacity
54 g-Cs/kg-adsorbent

0.2
0
0

2000

4000

6000

Effluent vol. (mͰ) / Adsorbent vol. (mͰ)

Cs

137

Total

Cs

Total

2.3

2.7

5

4.5

6

10.5

ND

4.8

6.3

11.1

ND

filtered withgrafted adsorbent

ND(Bq/Ͱ)

Fig.1-28 Breakthrough curve of grafted adsorbent for Cs
adsorption
Fibrous Cs adsorbent could collect 99~100% of Cs from 3000
times its packed volume of 1 ppm Cs solution.

Fig.1-29 Field tests of grafted adsorbent forradioactive removal in the Fukushima area
Fibrous Cs adsorbent could remove radioactive Cs
completely from pond water.

The large tsunami caused by the Great East Japan
Earthquake on March 11, 2011 damaged the electricity
supply and halted the circulation of the cooling water of thenuclear reactor in the TEPCO s Fukushima Daiichi NPS. As
a result of the meltdown of the nuclear reactor, radioactive
materials were dispersed into the surroundings, including
numerous water bodies.
To selectively remove radioactive cesium (Cs) from suchcontaminated water, we have developed a fibrous absorbent
with ammonium 12-molybdophosphate (AMP) by radiation
grafting of suitable monomer onto polyethylene nonwoven
fabric. An evaluation of the effectiveness of this adsorbent in
Cs removal under continuous flow was conducted bypumping 1 ppm Cs solution into a column (7 mm innerdiameter) in which 47 mg of AMP-type adsorbent was packed.
The flow rate was expressed by space velocity (SV [h-1]),
which was adjusted at 300 h-1 and calculated by dividing theflow rate of the solution by the volume of adsorbent packed
in the column, referred to as the bed volume (BV). The
outflow solution was fractionally collected each minute using

a fraction collector. The breakthrough point was defined as
the point at which the concentration at the outlet reached 1%(C/C0 = 0.01) of that at the inlet. Fig.1-28 illustrates thebreakthrough curve of Cs for AMP-type adsorbent. The
breakthrough capacity of the adsorbent at C/C0 was
54 g-Cs/kg-adsorbent (4 mol/kg), indicating that a column
packed with AMP-type adsorbent can process about 3000
times its packed volume. The grafted absorbents, folded in a
cartridge, were tested for a pond in Iitate-mura in Fukushima
Prefecture (Fig.1-29). The results indicated that the radioactive
Cs was adsorbed successfully, and removed to Not Detected(ND) level. An additional advantage of this technique is thatno sludge was produced as a result of such direct collectionof the toxic component, which would otherwise require
additional handling and purification.
The present study was sponsored by the Ministry of
Education, Culture, Sports, Science and Technology of Japan(MEXT). We thank the Iitate-mura municipal office forassistance with field experiments.

Reference
Iwanade, A., Seko, N. et al., Hybrid Grafted Ion Exchanger for Decontamination of Radioactive Cesium in Fukushima Prefecture and Other Contaminated
Areas, Journal of Radioanalytical and Nuclear Chemistry, vol.293, no.2, 2012, p.703-709.
JAEA R&D Review 201225

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

1-15 Development of a Highly Cesium-Selective Adsorbent for Volume Reduction=Molecular Design of Crown Ether Using New Chemical Bond Property=
Cs

4.0 Å

expectation 4.8 ÅComplex structure

Cs

6.0 Å

extending thecrown size ofcrown ether

DB18C65.0 Å

expectation

Complex structure

DB20C6

Fig.1-30 Molecular design concept of crown ether
Design of new crown ether was conducted with the aim of extending crown size in order to unite crown size with
the ion radius of cesium.
0.005

B3LYP of DET th eory
Lan L2DZ for Cs
6-311G(d) for H. C. O

0

0

-5

-5

-10

-10

-15

-15

-20

-20

-25

-25

-30

-30

0.004
0.003
0.002

FT

K

Cs

0.001
0
-0.001

Binding energy Eo:‫ ޓ‬E iot (Cs+-ether) = -244.036568 Ha‫ ޓޓ‬E iot (ether) = -223.993667 Ha‫ ޓޓޓ‬E iot (Cs+)= -19.985162 Ha‫ ޓ‬E o = -1.57 eV = -36.23 kcal/mol‫ޓޓޓ‬c.f. E o (Li+-H2O) = -33.6 kcal/mol‫ޓޓޓޓޓ‬Eo (Cs+-H2O) = -13.2 kcal/mol-0.003

0

1

2

3

4

5

6

7

8

R/A

Fig.1-31 The Cs(I)-DB20C6 complex is formed by electron donationfrom benzene π to Cs+ 5d(and 4f) orbital
According to the results of MO calculation, dibenzo-20-crown-6-ether(DB20C6) offers the capability to identify the ionic size of Cs. Furthermore, thecalculated binding energy of Cs+-ether is smaller than that of Li+-H2O, whichindicates that the Cs-DB20C6 complex is more stable than the Li hydrated
complex.

Radioactive cesium (Cs) contamination occurred in theenvironment owing to the accident at the TEPCO s
Fukushima Daiichi NPS. Clay minerals adsorbed radioactive
Cs; in order to recover this radioactive Cs, continued
development of new materials and new methods will be
required. Previously, it was established that crown ether is
one of the most popular absorbents for alkaline metals suchas Cs.
In this study, we conducted the following: (1) developmentof new crown ether, (2) evaluation of absorption properties of
new crown ether for alkaline metals, and (3) field investigationof the new crown ether in Fukushima. The development of a
new Cs-selective adsorbent used potassium (K)-selective 18crown-6-ether as a starting compound. The design of the newcrown ether was conducted with the aim of extending thecrown size of the crown ether in order to unite it with the ion
radius of Cs and promoting the complexation properties of
the crown ether over those of Cs (Fig.1-30). Moreover,chemical bonding properties between Cs and six oxygen
atoms of the crown ether were evaluated by MO calculation.
The results of the MO calculations indicate that dibenzo-20crown-6-ether (DB20C6) is capable of identifying the ionicsize of Cs. Furthermore, it was also established that the pi
electron of the benzene ring of DB20C6 and the d-f hybrid
orbital electron of Cs were combined (Fig.1-31). Although

RCs-H2O: 2.92Å
DWF: 0.001Å2-0.002

Fig.1-32 Radial structural function of
Cs-DB20C6 complex and Cs-crown ether
complex
From the results of EXAFS measurements,
the bond distance of Cs-O(OH2) in the
DB20C6-Cs complex was determined.

Cs made an orbit of d-f hybrid orbital, it did not exist in
sodium (Na) or K, which are some of the most common
alkaline metals in the environment. For the interaction
between the d-f hybrid orbital electron of Cs and the pi
electron of the benzene ring of DB20C6, DB20C6 isolated
Cs when high concentrations of Na and K existed in the same
system. According to the results obtained by EXAFS
measurements in SPring-8, the bond distance of Cs-O(OH2)in the DB20C6-Cs complex is 0.05~0.1 shorter than that ofCs-O(OH2) in a CsCl hydrated complex (Fig.1-32). This
result indicates that the DB20C6-Cs complex exhibits astrong capability to draw water molecules compared to Cs.
As part of the field examination in Fukushima, we also
developed a tornado adsorber tower and conducted samplingin Iitate village. The combined use of DB20C6 and theadsorber tower allowed removal of the water-soluble
radioactive Cs from agricultural water with about 100%efficiency. The adsorbent consisting of DB20C6 can beincinerated completely, in contrast to inorganic materials
such as zeolite. Therefore, the volume of radioactive waste(such as contaminated soil) could be drastically reduced by
the use of an organic adsorbent based on DB20C6.
The present study was sponsored by the Ministry of
Education, Culture, Sports, Science and Technology of Japan(MEXT).

Reference
Seko, N., Suzuki, S. et al., Countermeasures Against Radioactive Materials Pollution after the Great East Japan Earthquake, NTS Inc, 2012, p.204-210 (inJapanese).

26 JAEA R&D Review 2012

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

1-16 Verification of the Exposure Reduction Effect of Tap Water Restrictions=Evaluation of Averted Doses to Infants in the Implementation
after the Accidents at the TEPCO s Fukushima Daiichi NPS=

IitateKoriyama
Kawamata
Minamisoma

Iwaki
Tokai
Tokyo

IitateMinamisoma

10-4

Iwaki

101

Tokai
Kawamata

10-5

100
Tokyo

10-5

10-6

Effective dose (Sv)

102

10-4

Thyroid equivalent dose (Sv)

10-3

Koriyama131

I concentration in tap water (Bq/kg)

10-2

103

Fig.1-33 Concentration of 131I in tap water andprojected doses
The concentration of 131I in tap water at selectedareas in Fukushima, Tokai and Tokyo depended on
the sampling points and decreased with anapparent half-life of 131I of 2.8±1.2 days. Equivalentdoses to the thyroids and projected effective doses
for 1-year-old children also decreased.

10-7

10-6

10-1
3/17

3/21

3/25

3/29

4/2

4/6

4/10

4/14

Sampling date

Fig.1-34 Averted doses by implementing tap water
restrictions
In an emergency situation with the potential for internal
exposure due to intake of radionuclides, the averted equivalentdoses to the thyroids and the effective doses were a maximum
value of 8.3 mSv and 4.1 × 10-1 mSv, respectively, in Iitate. Tap
water restrictions were an effective protective action for
reducing the internal doses due to ingestion of 131I.

Averted doses (Sv)10-1

1-year-old children

Thyroid equivalent dose
Effective dose

10-2
10-3
10-4
10-5
10-6
Iitate

The accidents at the TEPCO s Fukushima Daiichi NPS
resulted in considerable land contamination, and thus tap
water restrictions were implemented to reduce the internal
doses due to ingestion of radionuclides such as iodine 131(131I). After the concentration of 131I in tap water decreased,
local governments were required to make a decision
concerning the termination of tap water restrictions.
However, because there was no clear provision for thetermination of tap water restrictions, social confusion was
induced. When considering appropriate protective action foravoiding unnecessary exposure, it is important to review theeffectiveness of the tap water restrictions by examining the
temporal change in the concentration of 131I in tap water andevaluating the dose prevented by the implementation of the
tap water restrictions -, i.e., the averted dose. To contribute to
the development of criteria for the termination of tap water
restrictions, the concentration of 131I in tap water was
examined using monitoring data published by several
authorities in Fukushima, Ibaraki and Tokyo, and this data
was applied to the subsequent evaluation of the averted doses
for members of the public -1-year-old children- due to theintake of 131I through tap water restrictions.
Fig.1-33 shows the concentration of 131I in tap water in Iitate,
Koriyama, Kawamata, Minamisoma, Iwaki (Fukushima),
Tokai (Ibaraki), and Tokyo using data from March 16 to April

Koriyama Kawamata Minamisoma Iwaki

Tokai

Tokyo

16, 2011. The concentration of 131I varied with the sampling
point, but showed an overall tendency to decrease with time.
It was found that the apparent half-life of 131I in tap water was
2.8±1.2 days, which was shorter than the physical half-lifeof 131I (8.0 days).
Fig.1-34 shows the averted doses for 1-year-old children
by implementing tap water restrictions in Fukushima, Ibaraki
and Tokyo. The doses were evaluated by multiplying thedaily 131I intake amount from tap water for 1-year-old children
with the dose coefficient of the International Commission ofRadiation Protection (ICRP). In Iitate, the maximum averted
equivalent dose to the thyroids and the effective dose were
found to be 8.3 mSv and 4.1 × 10-1 mSv, respectively. This
result indicated that the tap water restrictions were aneffective protective action for reducing the internal doses due
to ingestion of 131I. In Tokai, the dose due to ingestion of 131I
was averted fortuitously because it took substantial time torestore the water supply owing to repeated earthquakes andcontinuous blackouts. In Tokyo, the concentration of 131I in
tap water was relatively low and tap water restrictions were
implemented only for one day. Nevertheless, the averted
equivalent dose to thyroids was found to be 1.3 × 10-1 mSv,
confirming that the tap water restrictions were implemented
effectively.

Reference
Kinase, S., Kimura, M. et al., Evaluation of Averted Doses to Infants by Tap Water Restrictions after the Fukushima Daiichi Nuclear Power Plant Accident,
Nippon Genshiryoku Gakkai Wabun Ronbunshi, vol.10, no.3, 2011, p.149-151 (in Japanese).
JAEA R&D Review 201227

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

1-17 Visualization of Distribution of Fallout Radioactive Cs in Plants=Toward Unveiling the Present Status of Forest Contamination=

(a)

(b)

(c)

(d)

Ground levelFig.1-35 Autoradiograph image of Torreya
nucifera(a) Photograoph of Torreya nucifera, (b)autoradiograph image. Black spots in (b) showed
the radiation source in branches and leaves,indicating the presence of radioactive Cs. Red
circles indicate the leaves (light green colored
leaves in (a)) that grew after the accident. Blackspots are practically zero within the red circles,
indicating that the old leaves were rarelytransported to the young leaves. See details in thetext.

Roots and soil

Fig.1-36 Autoradiograph image of meadow grass
Yellow and red lines indicate ground level and roots and soil, respectively, in
optical photograph (c). Black spots are practically zero within the red circles
in the autoradiogarph image illustrating radionuclide distribution (d).

How were radionuclides attached to and transported inforest plants and soil after fallout of radionuclides as a resultof the accident at the TEPCO s Fukushima Daiichi NPS. This
question must be resolved to quantify present and future
forest contamination. Distributions of radionuclides in plants
and soil samples are usually measured by cutting small
fractions from the samples. Unfortunately, this techniquerequires considerable time and effort. Thus, we have applied
an autoradiography technique to analyze the distribution ofradionuclides within the plants and soil.
We collected Torreya nucifera, Cryptomeria japonica, andsoil in Iitate-mura, Soma-gun, Fukushima in May, 2011. We
placed the samples on an imaging plate (IP) coated withfluorescent paint that reacts to radiation and photographed
the spatial distribution of radionuclides.
The distribution of radionuclides in the branches andleaves of Torreya nucifera (Fig.1-35(a)) was analyzed by the
autoradiograph image where many black spots were present(Fig.1-35(b)). These black spots appeared on the IP in
response to radiation originating from radioactive Cs. Notethat 131I was decayed out in May. Fig.1-35 (a) illustrates dark
and light green leaves that grew before (old) and after(young) the accident, respectively. Black spots were present
on the dark green leaves but not on the light green leaves,indicating that radionuclides attached to the old leaves wererarely transported to the young leaves. The same result was
obtained from analysis of Cryptomeria japonica.
The autoradiograph image (Fig.1-36(d)) in roots and soil
collected from the grass meadow (Fig.1-36(c)) shows many
black spots on the leaves, but not in the roots and soil. This
indicates that radioactive Cs is not transported from theleaves to the roots and that rain did not, in general, dissolve
the radioactive Cs attached to the leaves to penetrate to thesoil.
The samples used in the present study were collected twomonths after the accident at Fukushima. To study long-termchange in forest contamination, we intend to collect samples
continuously and analyze the long-term behavior ofradioactive Cs attached to these plants.

Reference
Sakamoto, F., Ohnuki, T. et al., Local Area Distribution of Fallout Radionuclides from Fukushima Daiichi Nuclear Power Plant Determined byAutoradiography Analysis, Nippon Genshiryoku Gakkai Wabun Ronbunshi, vol.11, no.1, 2012, p.1-7 (in Japanese).

28 JAEA R&D Review 2012

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

1-18 Data Acquisition for Analysis of the Melt Progress in the Reactor Core
=Thermal Properties of Molten Fuel from Three Mile Island Unit 2=

5

UO2

Fe, Cr, Ni,
Zr, SiAg, In, Cd

25 µm

Metallic phase(b) VIP-12A(Zr, U, O)

10 mm

Thermal diffusivity (×10-6 m2/s)

Zr, Ni, Fe, Sn, Si(a) VIP-11A

VIP-11A (Metallic)

4

3
UO2 (TD95㧑)2

1

VIP-9D
VIP-12B
VIP-12A

SIMDEBRIS

VIP-9E

10 µm(U, Zr, O)

Fig.1-37 Molten fuels from Three Mile Island Unit 2
Appearance and microstructure of two typical types of molten fuels(debris). VIP-11A contains relatively large amounts of metallic
components originating from stainless steel, Inconel and Zircaloy,
whereas VIP-12A is ceramic debris that mainly consists of oxides ofuranium and zirconium.

During the accident at the TEPCO s Fukushima Daiichi
NPS (1F), the reactor cores reached very high temperatures
and the fuels were molten, resulting in a severe accident.
Decommissioning of the power stations, including removal ofthe molten fuels (debris) is underway. Information on theproperties of the debris is important for analyzing the
progress of the accident, estimating the status inside thedamaged reactors, and developing a plan for debris removal.
The most extensive examinations and analyses wereconducted on debris sampled from the reactor core of Three
Mile Island Unit 2 (TMI-2) after the accident in 1979. We
obtained approximately 60 pieces of TMI-2 debris andconducted appearance observation, density measurement,
microstructure observation and chemical analysis. In
addition, thermal properties, including the thermal expansion,
specific heat capacity, thermal diffusivity, and melting
temperature, were measured. Simulated debris that had asimilar chemical composition to that of the TMI-2 debris was
also used in the measurements.
Fig.1-37 shows the appearance and microstructure of two

VIP-11B

05001000 1500 2000 2500
Temperature (K)Fig.1-38 Temperature dependence of the thermal
diffusivity of debris
Thermal diffusivities of the TMI-2 debris and thesimulated debris (SIMDEBRIS) are lower than that of
UO2 below 1500 K, indicating lower heat conductivity inthe debris.

typical types of debris. The VIP-11A has relatively larger
amounts of metallic components, whereas the VIP-12A
debris is ceramic debris that mainly consists of uranium
and zirconium oxides. Fig.1-38 shows the temperaturedependence of the thermal diffusivity of the TMI-2 debris
and simulated debris. The thermal diffusivities of the ceramicdebris and the simulated debris (SIMDEBRIS) were lower
than that of UO2 below 1500 K; however, that of the metalrich VIP-11A was not. The presence of zirconium oxides
with lower diffusivities and pores that tend to decrease thediffusivity is probably the main cause of the obvious
difference between the TMI-2 debris and UO2.
It was also shown that the melting temperature of thesimulated debris was about 2840 K and that other corematerials such as iron, chromium, nickel, and silver have very
little influence on the melting temperature.
We are conducting various experiments to study the behaviorof the fuels under accident conditions and analyses to simulate
the accidents in order to support safe and efficient progress inthe decommissioning of 1F.

Reference
Nagase, F. et al., Thermal Properties of Three Mile Island Unit 2 Core Debris and Simulated Debris, Journal of Nuclear Science and Technology, vol.49,
no.1, 2012, p.96-102.
JAEA R&D Review 201229

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

1-19 Prediction of the Core Meltdown of the TEPCO s Fukushima Daiichi NPS Unit 1 Reactor
Operation of IC

10

Main steam
isolation valve

Isolationcondenser

Ǿ4.8 mPressure vessel

Ǿ4.4 mWater-vapor
separator

Main steam

Water

Pressure
Pressure (MPa)
Water level (m)

Steam

5

Water level0

Upper limit of heating section offuel rods

-10

Core

0

3000

2000

Top of core-5

Feed water

Safety relief valve

Stop of IC
4000

5000

10000

1000

Fuel rod surface temperature (͠)=Evaluation of the Influence of an Isolation Condenser on Core Cooling Using the TRAC Code=

0
15000

Time after earthquake (s)Ǿ3.9 mControl rods(a) Results of Case 1

Recirculation

Operation of IC

10

4000

Jet pump
Pressure (MPa)
Water level (m)Fig.1-39 Outline of the analytical components
To perform reliable accident analyses, details of the reactor
components, such as the core, pressure vessel, containment
vessel, isolation condenser, jet pump, valves, were simulated.

Pressure
5

Water level

0

Table 1-5 Analytical conditions
To validate the reactor core cooling using an IC, two analyticalconditions were provided: Case 1 simulates the actual condition,
and Case 2 simulates the assumed condition.

Case 1

Actualcondition

IC operates for 420 s after 12660 s
progress from an earthquake

Case 2

Assumedcondition

IC operates continuously after 8400 s
progress from an earthquake

In the TEPCO s Fukushima Daiichi NPS Unit 1 reactor(1F1), after the station blackout due to the tsunami following
the Great East Japan Earthquake, the isolation condenser (IC)was operated. However, because the cooling water could not
be supplied to the reactor core, a core meltdown occurred. To
validate reactor core cooling using an IC, accident analyses
for 1F1 were performed using TRAC-BF1, which is a safety
evaluation analysis code for light water reactors.
Fig.1-39 shows the outline of the analytical components.
The analytical conditions are shown in Table 1-5. Case 1
describes the conditions that simulate the actual event. The
IC operates for 420 s after 12660 s progress from theearthquake occurrence. On the other hand, in Case 2, the IC
is assumed to operate continuously after 8400 s progressfrom the earthquake occurrence.
Analytical results for the pressure, water level, and fuel rod
surface temperature are shown in Fig.1-40. Here, the waterlevel is the value presumed on the basis of the amount ofwater from the bottom end of the pressure vessel. The fuel2000

Top of core
Upper limit of heating section offuel rods

-5

-10
0

3000

5000

10000

1000

Fuel rod surface temperature (͠)

Containment
vessel0
15000

Time after earthquake (s)(b) Results of Case 2
Fig.1-40 Predicted results using the TRAC-BF1 code
In Case 1, because the starting time of the IC is late,
avoidance of a core meltdown is difficult. On the other
hand, in Case 2, because the IC is operated at an earlystage after the earthquake, a core meltdown can be
avoided.

rod surface temperature indicates the maximum outer surface
temperature of the fuel rod cladding.
In Case 1, exposure of the fuel rods occurs owing todecrease in the water level in the core at 11000 s after theearthquake. Because the fuel rod temperature was alreadyabove 1000 ℃ when the IC was started up after 12660 s
progress from the earthquake, it was found that avoidance ofa core meltdown was difficult, even if the IC continued tooperate.
In Case 2, because the upper limit of the heating section of
the fuel rods was not exposed, it was confirmed that the fuelrods could be cooled. Therefore, it was judged that a core
meltdown could be avoided if the IC had been operated at theearly stages after the earthquake.
The present results were reported to the accident cause
research team of the committee on accidental investigation of
the Japanese government, and were used for creation of theinterim report published on December 26, 2011.

Reference
Tamai, H., Takase, K. et al., Study on Influence of Isolation Condenser in Fukushima Daiichi Nuclear Plant Unit 1 Accident with TRAC-BF1 Code,
Nippon Genshiryoku Gakkai Wabun Ronbunshi, vol.11, no.1, 2012, p.8-12 (in Japanese).

30 JAEA R&D Review 2012

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

1-20 How to Avoid Core Damage in Unit 2 of the TEPCO s Fukushima Daiichi NPS=Evaluation of the Efficiency of Alternative Water Injection Timing through Accident Analysis=

Tmax : Maximum temperature of cladding
8time length from RCIC termination
to AWI initiation (h)9measuredpressure (MPa)

7
6
5
4

calculated
terminationof RCIC

3
2
1
0
3/11
12:00

3/12
00:00

3/12
12:00

3/13
00:00

3/13
12:00

3/14
00:00

3/14
12:00

3/15
00:00

date and hour

65
4

AWI start
relief valve open
Tmax above1500 K:
Core damaged

3
2
1

Tmax below 1500 K:
Core not damaged

0
3/11 3/12 3/12 3/13 3/13 3/14 3/14 3/15 3/15
12:00 00:00 12:00 00:00 12:00 00:00 12:00 00:00 12:00

date and hour

Fig.1-41 History of the primary pressure after theearthquake
The primary pressure was rather stable until approximately
12:00 on March 14 but began to increase drastically thereafter
because RCIC water injection stopped. The TRAC-BF1 code
effectively simulated such a pressure response, particularly
after the termination of water injection.

Fig.1-42 Time margin for beginning alternative water
injection to avoid core damage
The horizontal axis represents the time when RCIC water
injection is terminated. The vertical axis shows the estimatedtime length until the beginning of alternative water injection(AWI) using a fire engine after RCIC water injection
termination. The blue region indicates the estimated range ofthe time margin for avoiding core damage.

A long-term station blackout (SBO) accident occurred at
the TEPCO s Fukushima Daiichi NPS (1F) following theGreat East Japan Earthquake and resulted in severe coredamage in Units 1~3 owing to loss of adequate core cooling.
We simulated the accident in Unit 2 (1F2), in which coredamage occurred last among the three units, using thethermal-hydraulic best estimate code TRAC-BF1, whichanalyzes coolant behavior and core cooling. An input modelthat had been used for the analysis of a BWR with an electric
power of 1100 MW was adapted for the analysis of 1F2 with
an electric power of 780 MW in order to rapidly initiate theanalysis. The reactor core of 1F2 was long cooled using
continuous water injection from the Reactor Core Isolation
Cooling (RCIC) system while the SBO condition persisted.
Therefore, we studied the reactor response to the RCIC water
injection, its termination, and alternative water injection(AWI) using a fire engine in the accident analysis.
Fig.1-41 compares the measured and calculated primarypressure in 1F2 from the reactor scram to the start of AWI.
After the reactor scram, the pressure decreased significantly
once and a rather mild response followed. The pressure thenincreased drastically after termination of RCIC on March 14
and stayed high until the system depressurized through use ofa relief valve. The TRAC code underestimated the initialpressure response but provided a good simulation of thefollowing transient, particularly of the pressure aftertermination of RCIC water injection. This result suggests thatthe code can properly simulate the 1F2 accident.
The code analysis results obtained by changing the timingof RCIC termination and AWI initiation are summarized in
Fig.1-42 for discussion of the efficiency of AWI. Core
damage occurs when the cladding temperature of the fuelrods exceeds 1500 K. Fig.1-42 indicates that core damage
can be avoided within the blue region, where the maximum
cladding temperature is lower than 1500 K. Core damage
may occur in the region above the red line. While theinitiation of AWI using a fire engine on March 14 was toolate for 1F2, this diagram suggests that core damage could
have been avoided if the AWI was begun approximately 4 hearlier. These results should contribute to the improved
evaluation of accident management measures.

Reference
Watanabe, T., Ishigaki, M. et al., Analysis of BWR Station Blackout Accident −Thermal-Hydraulic Behavior up to Severe Core Damage in Fukushima
Daiichi Power Plant Unit 2−, Nippon Genshiryoku Gakkai Wabun Ronbunshi, vol.10, no.4, 2011, p.240-244 (in Japanese).
JAEA R&D Review 201231

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

1-21 Toward Safe Management of Accident Waste=Challenges to Long-Term Storage, Processing, and Disposal of Waste=

Hydrogen molar
fraction
1.5e-3 (mol/mol)(Hydrogen is discharged
by diffusion effect)(mol/mol)
1.22e-02
1.16e-02
1.10e-02
1.04e-02
9.77e-03
9.16e-03
8.55e-03
7.94e-03
7.33e-03
6.72e-03
6.11e-03
5.50e-03
4.89e-03
4.28e-03
3.67e-03
3.05e-03
2.44e-03
1.83e-03
1.22e-03
6.11e-04
0.00e+00(Hydrogenmixture gas
outflow)

Cross section of
the vessel withvent tube

0.0 (mol/mol)(air inflow)

Cross section of
the vessel withline of water
inlet

Cross section of
the vessel withline of wateroutletCs-137 Activity in Waste (Bq/t)

1.00E+16
0.012(mol/mol)

Half life of Cs-137: 30.17
Secondary cesium
years
adsorption apparatus
spent zeolite
Cs-137 limit for near-surface
disposal (concrete pit): 1.00E+14 Bq/t

1.00E+15
1.00E+14
1.00E+13

Cesium adsorption
apparatus
spent zeolite

1.00E+12
1.00E+11
About 90 About 140
years
years

1.00E+10

Decontamination
apparatus
sludge

1.00E+09
0

50

100

150

200

250

300

Storage Time (y)Fig.1-43 Analytical results for hydrogen diffusion in aspent zeolite vessel
Analytical results obtained for a hydrogen production rate of
18.3 /day and decay heat of 237 W. Estimated hydrogenconcentration is well below 4% volume (the lower explosive
limit).

Fig.1-44 Relationship between activity of secondary
wastes and storage time
Spent zeolites are categorized as sub-surface disposal waste.
Radioactivity of spent zeolites will decrease with storage timeand spent zeolites may be categorized as near-surface
disposal waste.

We are currently addressing the challenges presented by
the decommissioning of the TEPCO s Fukushima Daiichi
NPS. In particular, R&D regarding long-term storage,
processing, and disposal of secondary waste produced bytreatment of contaminated water present important challenges.
Zeolite adsorption and coagulation-sedimentation methods
are applied to remove radioactive nuclides, mainly
radioactive cesium, from contaminated water. Currently, weare studying the characteristics of hydrogen generation invessels, corrosion of the vessels, and related subjects toensure appropriate long-term storage of spent zeolite andsediment. We are also investigating waste categorization,
waste conditioning technologies, and so on to clarify future
prospects for waste processing and disposal.
A spent zeolite vessel adsorbing cesium is detached fromthe radioactive water decontamination system. Althoughwater is drained from the vessel before storage, combustiblehydrogen could be produced via radiolysis of water absorbed
in the zeolite and the small amount of water remaining at thebottom of the vessel.
Analytical simulation of hydrogen diffusion was conducted

to estimate hydrogen concentrations under uniform rates ofhydrogen production and heat generation without steam
generation. The analytical results presented in Fig.1-43
illustrate that the air is inducted slowly into the vessel bottom
through the outlet water pipe as a result of buoyancy effects,
which in turn results from heat generation and density
differences due to hydrogen generation in the vessel. The
mixed gas (air and hydrogen) is released to the outside
through the inlet water pipe and the vent tube. This thermalhydraulic behavior prevents the hydrogen concentration fromexceeding 4%, i.e., the lower hydrogen explosive limit.
Fig.1-44 illustrates the decrease of activity with storage
time for waste generated by contaminated water treatment.
Two types of spent zeolites are categorized as subsurface
disposal waste based on their Cs-137 activities. Activities ofsuch wastes will decrease with storage time and the zeolite
wastes may be categorized as near-surface disposal waste in
future. It is important to evaluate storage time, radioactivity,and chemical composition of waste form comprehensively toensure safe and rational management of waste produced by
the accident at Fukushima.

Reference
Nakamura, H., Research for Treatment and Disposal of Secondary Waste Produced by the Processing of Contaminated Water, Kankyo Gijutsu, vol.41, no.6,
2012, p.365-370 (in Japanese).

32 JAEA R&D Review 2012

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

1-22 Decontamination of Radioactive Strontium to Make Fresh Water

Contaminated waste water
Treated water(Freshwater)

Radioactive material treatment unit
(cesium adsorption,
decontamination)

Desalination unit
(Reverse osmosis,
Concentration)
Concentrated waste water
Vessels
Fig.1-45 Treatment of contaminated water at the 1F
The fresh water obtained from the contaminated water isreused as a coolant. The concentrated waste water resulting
from the treatment contains radioactive nuclides such as
strontium.

At the TEPCO s Fukushima Daiichi NPS (1F), an
enormous volume of contaminated water has been generated
and temporarily stored at the plant site. As shown in Fig.1-45,
contaminated water bearing radioactive nuclides such as 137Cs
is treated by adsorbing Cs into a zeolite using Cs adsorption
units, followed by purification using a reverse osmosis
membrane. The water is then used for cooling the fuel inside
the reactors. The resulting waste water is stored in newly
fabricated vessels. The amount of waste water will increase
with consecutive treatments; therefore, treated fresh water
will be released to the environment after confirmation that theconcentrations of radioactive substances are sufficiently low.
Among the radionuclides in the waste water, 90Sr is importantbecause it has a low concentration limit and requires aneffective decontamination technology.
We developed a treatment process for the low-level liquid
waste from the Tokai Reprocessing Plant that utilizes atitanium oxide adsorbent (READ-Sr) for 90Sr decontamination.
The contaminated water at the Fukushima site contains
seawater, and thus, Sr must be adsorbed in the presence ofvarious metallic cations. Because the READ-Sr adsorbent

7
6
5
4
3
2
1
0
0

20

40

60

80

100

120

Time for contact (min)Fig.1-46 Increase in the Sr concentration in the adsorbent
Sr was adsorbed from the water and its concentration
increased with time, reaching a constant value after 30 min.

1000

Kd
Distribution coefficient ‫ޓ‬

ReactorSr concentration inadsorbent (g/L)=Application of a Strontium Adsorbent Prepared for Use with
Radioactive Waste Solutions from Nuclear Fuel Reprocessing=

Ca
100

Sr

10

Mg

1

K

0.1
0

25

50

75

100

Seawater fraction (㧑)Fig.1-47 Distribution coefficients (Kd) for some elements
The variable Kd is the ratio of the concentration in the solid andliquid phases for an objective component. Sr is selectively
adsorbed with Ca over Mg and K.

was developed for a sodium nitrate solution, its application to
the waste water with seawater was investigated.
The properties of READ-Sr for adsorption were examinedusing simulated solutions. As shown in Fig.1-46, for Sradsorption from water containing 50% seawater, the Srconcentration in the adsorbent reached equilibrium within
30 min. Thus, it was verified that a practical adsorption rate
was obtained for Sr adsorption. Sr is an alkaline earth
element, and seawater contains chemicals resembling Ca andMg; therefore, to efficiently remove Sr, the adsorbent must be
selective for the element. As shown in Fig.1-47, thedistribution coefficient (Kd) is affected by the fraction ofseawater, although Sr adsorption and selectivity for Sr and Caover Mg and K were observed.
The adsorbent was also examined for batch-wise
adsorption combined with the Cs adsorbent and in a column
test. The Sr decontamination expected based on the
distribution coefficients was observed in both tests.
We will continue this study with the aim of applying thistechnology at the 1F.

Reference
Takahatake, Y., Koma, Y. et al., Strontium Decontamination from the Contaminated Water by Titanium Oxide Adsorption, Proceedings of International
Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011), Makuhari, Japan, 2011, paper no.462855, 5p., in CD-ROM.
JAEA R&D Review 201233

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

1-23 Evaluation of Hydrogen Production from Waste Zeolite Adsorbents=γ-Radiolysis of Mixtures of Zeolites and Seawater=

Seawater
Zeolite H, 53wt㧑
Zeolite EH, 49wt㧑
Natural mordenite, 58wt㧑

Fig.1-48 Photograph of the zeolite adsorbents used inthis study
Three types of zeolite adsorbents (from left zeolite H, zeolite
EH, and natural mordenite) were used in this study.
Table 1-6 Evaluated hydrogen production and assumed
condition of the waste zeolites
Hydrogen production was evaluated assuming that theradiation from the adsorbed Cs was fully absorbed.

Zeolite adsorbents

1t

Seawater

1t

Adsorbed Cs / zeolite adsorbents

0.1wt㧑

Radiation-chemical yield of hydrogen

3.5×10-8 mol/J

Absotbed fraction of the radiations

100㧑

Hydrogen production rate

1.5 Ͱ/hSafe storage of waste zeolites is important for stable water
treatment operations at the TEPCO s Fukushima Daiichi
NPS (1F). Because the treatment uses zeolite adsorbents forremoval of radioactive materials from contaminated water,
highly radioactive waste zeolites are generated. A similarwater treatment method was applied after the accident atThree Mile Island (TMI), and the obtained experience serves
as a useful reference. However, the water in 1F contains salts
from seawater, which makes the situation different from thatof TMI.
Hydrogen (H2) control is necessary for the safe storage of
the wastes, because the radiolysis of residual water in thewastes produces H2. However, the prediction of H2 production
is difficult, because porous ceramics such as zeolites are
considered to affect water radiolysis. The reaction scheme forH2 production in the presence of ceramics has not been
established. Hence, further studies are required for evaluating
the effect of zeolites on the radiolysis of water containing
salts.
Mixtures of zeolites and seawater were irradiated with
γ-rays, and the quantity of generated H2 was measured.

Hydrogen production (vmol/g)

0.4

0.3

0.2

0.1

0.0
0.0

2.0

4.0

6.0

8.0

Absorbed Dose (kGy)
Fig.1-49 Hydrogen from mixtures of zeolites and seawater
The mixtures of zeolites and seawater were irradiated with
γ-rays from 60Co, and the produced quantities of hydrogenwere measured.

Three types of zeolites (Fig.1-48) were used. Zeolites H andEH were supplied by KURION. Zeolite H has actually beenused at 1F. Natural mordenite was supplied by SHINTOHOKU Chemical Industry.
Fig.1-49 presents the results. The produced quantities of H2
were comparable for the three zeolites. Based on the results,
the H2 production from the waste zeolites was estimatedassuming the conditions shown in Table 1-6. The radiationchemical yield of H2 indicates the H2 production per unit
energy deposition by the γ-rays. The highest value of theyield in the experiment, 3.5 × 10-8 mol/J, was used, and theH2 production was determined to be 1.5 /h at standard
temperature and pressure.
A more conservative evaluation can be obtained using theH2 yield of the radiolysis of seawater. As shown in Fig.1-49,
seawater in the absence of the zeolites produces more H2 thanthe mixtures with zeolites.
In future studies, we are planning to investigate H2
production under various conditions in order to provide more
precise evaluation of the H2 production from the waste
zeolites.

Reference
Kumagai, Y. et al., Measurement and Evaluation of Hydrogen Production from Mixtures of Seawater and Zeolite in Decontamination of Radioactive Water,
Nippon Genshiryoku Gakkai Wabun Ronbunshi, vol.10, no.4, 2011, p.235-239 (in Japanese).

34 JAEA R&D Review 2012

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

1-24 Development of Special Measuring Equipment Using JMTR Irradiation Techniques=A New Water-Level Indicator Applicable under Severe Conditions=

MIcableA schematic of thesensor unit
Ǿ4.8 mm

Movement of positionof the water level indicator

Ǿ1 mm

100 m

Ǿ4 mm

350

Water temperatureǾ4.8 mm

+

㧙

Water

Temperature (͠)

HeaterAir

AlumelSheath (SUS316)

Measurement Chromel
point

200 mmMeasurementpoint

K-type Thermocouple

Sensor unit

300

Distancefromwater
surface‫ع‬㧦Measurement point

Fig.1-50 Schematic diagram of the water levelindicator
A heater and a thermocouple are installed in thesensor unit, which is connected to a radiationresistant MI cable. The sensor unit is heated with an
installed heater, and the water level is detected on
the basis of the temperature difference in air or inwater.

The water level in a light water reactor is monitored using
water-level indicators, such as float-electromagnetic-inductionand electrode-type indicators. During the accident at theTEPCO s Fukushima Daiichi NPS, measurement of the waterlevel in the spent fuel pools and in the pressure vessels was
not available owing to the station blackout, which caused adelay in information gathering. Thus, it is necessary todevelop a new, reliable water-level indicator that is available
under station blackout conditions.
The target indicator is small, portable, has high radiation
resistance, and operates on a battery-level power supply. To
achieve downsizing and high radiation resistance, a method
for measuring the water level using an installed heater andthermocouple was adopted.
A heater wire and a K-type thermocouple are installed in astainless sheath. This structure enables achieving a small
diameter. An MI cable is used as the lead wire and can beused in a high-radiation environment.
The water level is measured using the indicator, whichmeasures the temperature difference in the water or in the aircaused by the different heat transfer coefficients of the waterand air. To measure the water level with high accuracy, thetemperature difference is increased by the heater in theindicator using a suitable impressed voltage (Fig.1-50).
A characterization test was performed. The impressed

27 ͠
90 ͠

250
200
150
100
50
0
-150

-100

-50

0

50

100

150

Distance from water surface (mm)Fig.1-51 Results of water-level measurements using thedeveloped water-level indicator
The developed water-level indicator shows a large temperature
difference in water and air, regardless of the water temperature. The
measurement results for the change in the distance between thetemperature measurement point and the water surface indicate thehigh accuracy of the water-level indicator.

voltage was changed from 1 to 5 V, and the temperaturechange was measured in the water and in the air by moving
the detector in the vertical direction. From the test, it was
confirmed that the water level can be detected using thesesmall voltages. Fig.1-51 shows the results of water-levelmeasurements performed with water at 27 ℃ and 90 ℃. A
voltage value of 5 V was impressed to the heater. The
developed water-level indicator showed a large temperature
difference in water and air, and the accuracy of themeasurement of the water level was approximately 20 mm. If
the water-level indicator is used in an actual reactor facility,
the measuring person and the apparatus should be kept awayfrom the sensor unit in order to avoid severe conditions.
Therefore, a water-level indicator with a 100 m MI cable wasalso manufactured, and a demonstration test was performed.
The results of this test indicated that the developed waterlevel sensor is sufficiently enough to be portable and that it ispossible to measure the water level using a small battery
power supply.
Importantly, the developed sensor unit functions in hightemperature and high-radiation severe environments, where
no one can access the measurement apparatus, because thesheath-type thermocouple and heater have high radiation
resistance.

Reference
Miura, K., Shibata, A. et al., Development of Reactor Water Level Sensor for Extreme Conditions, Proceedings of the 4th International Symposium on
Material Testing Reactors, Oarai, Japan, JAEA-Conf 2011-003, 2011, p.193-195.
JAEA R&D Review 201235

2    
R&D for Commercialization of Fast Reactor Cycle Technology
FY2010Confirmation of Technical Issues
for Innovative Technology

Determination of Adoptionof Innovative Technology

Fast Reactor Cycle Technology Development(FaCT Project)‫ ع‬Design Study‫ ع‬Experimental Study‫ ع‬International Standardization of‫ ޓ‬Safety Design Requirements‫ ع‬Fundamental R&D

R&D in MONJU‫ ع‬Demonstrating the Reliability of‫ ޓ‬an Operational Power Plant‫ ع‬Establishing Sodium Handling‫ ޓ‬Technology

Domestic Cooperation
International Development of the 4th Generation Reactor

Fast Reactor Cycle Systemthat Achieves Performance Targets of
Safety, Economy, etc., as a Future Energy System

FY2006

Fig.2-1 Overview of research and development efforts targeting commercialization of a fast reactor cycle
R&D activities have been implemented toward the commercialization of fast reactor cycle technology, which has excellent performancein terms of safety, economy, environmental conservation, efficient utilization of resources, and nuclear proliferation resistance. In
FY2011, investigations aiming at international standardization of safety design requirements were in progress to realize top-tier safety
goals for 4th-generation reactor systems.

Various activities, such as the FaCT project and R&D in
the MONJU reactor are being implemented to commercialize
fast reactor cycle technology, as shown in Fig.2-1. Regarding
the FaCT project, innovative technologies to be adopted in
the fast reactor cycle have been determined, and the firstphase of the project has been completed. With respect to the
R&D in the MONJU reactor, the system start-up test was
resumed and the core confirmation test was completed. Given
that a review of the nuclear energy policy is underway by thegovernment following the accident at the Tokyo ElectricPower Company, Incorporated Fukushima Daiichi Nuclear
Power Station, the second phase of the FaCT project, which
was scheduled to start in FY2011, was postponed. Currently,international standardization of safety design requirements isunderway in order to realize top-tier safety goals for 4thgeneration reactor systems. The topics introduced in thischapter include subjects regarding further improvement ofsafety. The summary of the topics is as follows.
Electricit de France (EDF) reviewed innovativetechnologies designed to improve safety and reduce the useof materials for the Japan Sodium-cooled Fast Reactor(JSFR) from the perspective of an investor-operator of future
French SFRs, and evaluated that these technologies are also
attractive for future French commercial SFRs (Topic 2-1).
In the JSFR, the goal is to contain the consequences of acore disruptive accident inside the reactor vessel. It was
confirmed that the introduction of a fuel assembly with an
36 JAEA R&D Review 2012

inner duct structure was an effective measure for achieving invessel retention based on fundamental experiments (Topic 2-2).
In addition, bent pipes with a large diameter, which are
adopted for a compact component layout, have coolant
flowing through them at high speeds. The flow in the pipe elbow
was evaluated using flow experiments in order to prevent
pipe failure caused by flow induced vibration, and the resultsof this study are reflected in the piping design (Topic 2-3).
Regarding the reduction in the creep strength of modified
9Cr-1Mo steel, which is utilized as the piping material, atelevated temperatures for long periods, a creep lifetime
prediction expression was prepared through analysis of creep
test data corresponding to the failure mechanism, and it was
confirmed that the piping design of JSFR is assured(Topic 2-4). Even in the case of a station blackout caused bya tsunami, after shutting down the MONJU reactor, it was
successfully shown that the natural circulation of the coolant,
and thus the reactor core cooling, is possible as long as theflow path of the coolant is ensured (Topic 2-5).
With respect to the R&D of fuel fabrication techniques, thefeasibility of a fuel fabrication process based on a simplified
pelletizing method, which enables simplification of the fuelfabrication process, was experimentally confirmed after
testing the tumbling granulation and die lubricating-typepelletizing methods as well as the ability to adjust the oxygen
to metal ratio for sintering (Topic 2-6).

Research and Development of Advanced Nuclear System

2-1

Japanese Fast Reactor Technologies Draw the Attention of Electricitde France=Safety Improvement and Mass-Volume Reduction of the JSFR=

Highly effective
technologies for safety
improvement

Promising technologies for mass
and volume reduction,
while they have further
development subjects

Secondary pump

Steam
Generator

Self-actuated shutdown system(SASS)

Compacting reactor vessel

When coolant temperature rises in caseof anticipated transient without scram(ATWS), bearing capacity of temperature-sensitive electromagnet holding thecontrol rod will decrease and the control
rod falls without manual operation, andthe reactor can be shutdown passably.

Space required for refueling is reduced byadopting upper inner structure with verticalslit and pantograph-type fuel handlingmachine (FHM). Reactor vessel iscompacted by adopting a hot-vessel
concept without overflow system.

Fuel assembly with inner duct
structure (FAIDUS)
Concept in which a steel duct is installed ineach fuel assembly in order to discharge
melting fuel at an early stage and avoid
recriticality, when core disruptive accident(CDA) occurred.

2 loop cooling system
Reactor
VesselIntegrated
pump-IHX

Number of primary loops is reduced fromconventional 3 to 2 in JSFR, without
degradation of safety level. Two-loop
concept allows simplification of the system
and reduction of mass and volume.

Fig.2-2 JSFR innovative technologies draw the attention of Electricitde FranceElectricitde France has paid attention to the safety improvements and mass-volume reduction technologies included in the innovativetechnologies adopted for the JSFR, and evaluated them to be highly effective technologies for safety improvement and promisingtechnologies for mass and volume reduction.

We are pursuing the international standardization of theJapan sodium cooled fast reactor (JSFR) concept that has ahigh safety level. For this purpose, we have concluded anarrangement for technical cooperation on sodium cooled fast
reactors (SFR) with Electricitde France (EDF) and carriedout exchange of information. Under the framework of thiscooperation agreement, EDF reviewed the innovativetechnologies adopted for the JSFR design from theperspectives of an investor and operator of future Frenchcommercial SFRs.
EDF remarked that the technologies related to safety
shown in Fig.2-2, including the self-actuated shutdown
system (SASS) and the fuel assembly with inner duct
structure (FAIDUS), are potential technologies for safetyenhancement based on reliable experimental results, andcould be effective for commercial SFR safety enhancement.
In the past, a passive safety system (the auxiliary shutdown
system) with a function similar to that of the SASS, wasdeveloped in France and installed in the French SFRs Ph
nix
and Superph
nix, but a reduction in the electromagnetic
bearing capacity was observed over time, and the possibilityof an unexpected fall of the control rod was feared.
Development of the SASS adapted to the JSFR was done withextensive experiments, and the appropriate electromagnetic
material was selected on the basis of past experiences. The
performance of the SASS structural material under elevatedtemperature, the stability of the bearing capacity over an
extended period, and the response characteristics wereconfirmed. An irradiation test of the SASS was conducted in
the Japanese experimental SFR Joyo under actual reactoroperation condition and neutron fluence corresponding to 60
years of use in a large-scale SFR, and the stability andreliability of the SASS were demonstrated. Given thisbackground of recent developments and experimental resultsin Japan, JAEA and EDF will continue the exchange oftechnical information regarding the application of the SASS
in future SFRs.
Concerning FAIDUS, EDF remarked that the FAIDUS
concept is a potential system for avoiding re-criticality in thecase of a core disruptive accident. Its effectiveness was
confirmed via an in-core test conducted by JAEA in theimpulse graphite reactor (IGR) in Kazakhstan.
In addition, EDF remarked that the technologies forcompaction of the reactor vessel and for system
simplification in the JSFR design could be effective for mass
and volume reduction, and that further development of thesetechnologies is worth pursuing.
We will continue our discussions with EDF, including
comparison of technologies and perspectives on safetybaselines in order to gain a mutual understanding in
anticipation of achieving safety improvements for future
SFRs.

Reference
Uematsu, M. M., Pr
le, G. et al., Comparison of JSFR Design with EDF Requirements for Future SFR, Proceedings of 2012 International Congress on
Advances in Nuclear Power Plants (ICAPP 12), Chicago, USA, 2012, paper 12354, 13p., in CD-ROM.
JAEA R&D Review 201237

Research and Development of Advanced Nuclear System

2-2Containment of Core Disruptive Accidents=Fuel Discharge Behavior in a Fuel Assembly with Inner Duct Structure=

Molten fuel simulant discharged into theupper plenum through the inner duct

Molten fuelFissiongas
AA’

Coolant
vapor
Coolant(Sodium)

Reactor vessel

Inner duct

(Top end is open,
whereas the bottom
is closed.)

A-A’ crosssectional view

Fuel assembly with
inner duct structure

Molten fuel simulant
moving upward in
the inner duct
Coolant channel
simulating the innerduct structure(The equivalent diameter
was approximately 30 mm.)

Injection nozzle(The molten-fuel simulant
was injected into the innerduct filled with water fromthe lateral side.)0s(before melt injection)

0.47 s0.63 s

1.0 s

2.3 s

Discharge behavior of molten-fuel simulant observed in the experimentFig.2-3 Fundamental experiment on upward fuel discharge in a fuel assembly with inner duct structure
In an experiment using a high-density melt and water as simulants for molten fuel and coolant respectively, it was found that pressurebuildup due to coolant vapor behaved as a driving force for upward fuel discharge in the inner duct structure.

In order to ensure the safety of FBRs, it is important toprevent the occurrence of core disruptive accidents (CDAs)
and, in case one does occur, to mitigate its consequences andthus contain the impact within the reactor vessel.
In CDAs, there is the possibility that the compacting
movement of a large-scale molten-fuel pool within the core
may induce severe power excursions and degrade theintegrity of the reactor vessel. The reactor vessel of existing
FBRs has been designed to withstand such power excursions.
However, for next-generation FBRs, in order to ensure thecontainment of CDAs in the reactor vessel, we are aiming toeliminate the potential for such power excursions by
preventing the formation of a large-scale molten-fuel pool. A
design measure for achieving this target is the introduction ofa fuel assembly with inner duct structure (FAIDUS) thatfacilitates molten-fuel discharge from the core region.
In the present study, in order to clarify the fuel discharge
behavior in the FAIDUS, we conducted a fundamental
experiment with a high-density melt and water as simulants
for the molten fuel and coolant, respectively. In the current
reference design for the FAIDUS, the top end of the innerduct is open, while the bottom end is closed. Therefore it isexpected that the molten fuel will be driven by the pressure in
the reactor core region, which increases with fission gas
release upon fuel melting, and thus discharged toward theupper plenum. In the experiment, we employed an apparatus
whose configuration and size was similar to that of the innerduct structure. Fig.2-3 shows that the molten-fuel simulant
was driven upward along the duct and discharged into theupper plenum. The experimental data suggest the following: (1)the liquid coolant, which may have provided resistanceagainst the upward discharge, was excluded from the duct
due to the coolant vaporization during the early phase of thedischarge, and (2) the coolant vapor flow drove the upward
discharge of the molten-fuel simulant against gravity.
Based on the results of this fundamental experiment, it was
found that both the coolant vapor and fission gas can behave
as driving forces for fuel discharge from the core region in
the inner duct structure. Thus, the experimental result
confirmed that the FAIDUS is an effective measure for
ensuring the containment of CDAs in the reactor vessel.

Reference
Matsuba, K. et al., Experimental Study on Upward Fuel Discharge during Core Disruptive Accident in JSFR: Results of an Out-of-Pile Experiment with
Visual Observation, Proceedings of 19th International Conference on Nuclear Engineering (ICONE 19), Chiba, Japan, 2011, ICONE19 - 43993, 6p.,
in CD-ROM.

38 JAEA R&D Review 2012

Research and Development of Advanced Nuclear System

2-3

Prediction of the Flow Structure in Large Diameter Elbows=Evaluation of Flow-Induced Vibration in Large Diameter Pipes in the JSFR=
LASER
t=0.0(ms)t=0.6(ms)

Velocity calculatedby the two pictures

RD

short-elbow
Radius of curvature (R) equals to diameter of piping (D)Fig.2-4 Schematic diagram of the PIV (Particle Image Velocimetry) method
Fine particles in a transparent elbow made of urethane were irradiated using a laser, and images were captured using a high-speed
camera. The velocity was calculated based on the movement of the particles, and the time difference was obtained by comparing two
pictures.

OutsideOutsideFlow direction

Velocity 0

4.5(m/s)

InsideInsideVelocity 02(m/s)

Fig.2-5 Distribution of the average velocity
Low velocity (blue) and high velocity (red or yellow) regions
were observed in the elbow. The velocity of the boundary
between them fluctuated periodically. A reverse flow, or flow
separation, was observed in the low velocity region, which is
known as the separation region.

Fig.2-6 Instantaneous velocity distributions in the crosssection of the pipe at the elbow outlet
Circumferential secondary flows, alternating from both lateralsides, moved to the inside of the elbow. This flow interactedwith the flow separation depicted in Fig.2-5, and their mixing
caused velocity fluctuations at the inside wall of the elbow.

In the design of the Japan Sodium-cooled Fast Reactor(JSFR), a large diameter short elbow (D1.2 m) was adopted
as a part of the primary hot leg pipe. The short elbow causes
high turbulent flow due to changes in the flow direction overa short distance. When the frequency of the flow fluctuation
is close to the characteristic vibration frequency of the piping,
the piping may vibrate destructively. This phenomenon iscalled Flow-Induced Vibration (FIV). For the design of the
JSFR, it was therefore necessary to understand the flow
structure and flow fluctuation in the elbow in order toevaluate the structural integrity of the hot-leg piping withrespect to FIV. Thus, water experiments were conductedusing a 1/8-scale elbow in order to investigate the flow
structure and hydrodynamic phenomena in the elbow forclarification of the FIV mechanism and to evaluate theintegrity of the design.
In these experiments, the flow structure was investigatedusing a PIV (Particle Image Velocimetry) method. Two
successive images of a fine particle tracer in the fluid were

captured using a high-speed camera, and the velocity fields
were measured on the basis of the movement of the tracer inthe fluid calculated using the two images (Fig.2-4). The
velocity field in the elbow using PIV can be seen in Fig.2-5.
We succeeded in measuring the flow structure inside an
elbow, including the flow separation, and quantitatively
determined the frequency of the flow fluctuation. In addition,
we succeeded in measuring the secondary flow in the crosssection of the pipe, which has not been measured in previous
studies (Fig.2-6). At the elbow outlet, it was found that thecircumferential secondary flow moved to the inside of theelbow, alternating from both lateral sides, and the velocity
fluctuated intensely near the inside wall. Using theseexperimental results, it was possible to explain themechanism of the flow structure and develop not only the
proposed guidelines for the FIV evaluation of the hot-leg inthe primary cooling system of sodium-cooled fast reactors
but also the design of piping in the JSFR.

Reference
Ono, A. et al., Influence of Elbow Curvature on Flow Structure at Elbow Outlet under High Reynolds Number Condition, Nuclear Engineering and Design,
vol.241, issue 11, 2011, p.4409-4419.
JAEA R&D Review 201239

Research and Development of Advanced Nuclear System

2-4

For Enhancement of the Reliability of SFR Pipe Designs=Creep Strength Assessment of Mod.9Cr-1Mo Steel Welds Taking
Type-IV Damage into Account=

Mod.9Cr-1Mo steel(welded joints)

Integrated pump-IHX component

equivalent to 550 㷄×60 years

Weld (a)Primary hotleg pipe(in red)Weld (b)Reactor vessel

Primary cold-leg pipe (in yellow)

ElongationǬ(%)

30

Sold symbols meandata obtained understress smaller thanhalf of 0.2% proof
stress of base metal

20

10

ٌ
٤

‫غ‬
٠

0
25000

500 ͠
550 ͠
600 ͠
650 ͠

30000

35000

Time-Temperature parameterFig.2-7 Schematic view of the JSFR RV and primarysystems
Employing Mod.9Cr-1Mo steel, a simple piping layout will beachieved. On the basis of results of creep damage evaluations
for welds (a) and (b), it was confirmed that the stress at thewelds was less than the allowable stress, and that the SFR
pipe was designed with enough reliability.

Fig.2-8 Observed creep rupture elongation for Mod.9Cr-1Mo
steel welds
Significantly small ductility was observed in long-term creep
tests, which were performed with a stress less than half of the
0.2% proof stress of the base metal, and the failure
mechanism shifted from ductile creep to Type-IV failure.

For Sodium cooled Fast Reactors (SFR), a compact plantdesign will be achieved by shortening the pipe length andemploying Mod.9Cr-1Mo steel (Fig.2-7), which has a low
thermal expansion rate and excellent creep strength.
However, it is believed that the creep strength of welds ofhigh-chromium ferritic steels, such as Mod.9Cr-1Mo steel,degrade compared to those of the base metals, particularly
after long-term use and in high-temperature regions. In theJapanese technical standard for thermal power plantcomponent design the allowable stress for these steel welds
has been reduced because of Type-IV damage. In this study,
available creep test results for welds were analyzed
considering the possible failure mechanisms, and the validityof the SFR pipe designs with acceptable margins was verified.
As shown in Fig.2-8, significantly small ductility wasobserved in the long-term creep tests, which were performed
with a stress less than half of the 0.2% proof stress of thebase metal. In addition, the failure position observed in thecreep tests shifted from the base metal to the heat-affectedzone with an increase in the creep rupture time. As a result,in Mod.9Cr-1Mo steel welds, it was clarified that the failure
mechanism shifts from ductile creep to Type-IV failure withan increase in the creep rupture time. Therefore, creep
characteristic equations for Mod.9Cr-1Mo steel welds were
proposed on the basis of the individual polynomial
regressions for the long-term and short-term regions.
In addition, the distribution of the ratio of observed creep
lives to calculated creep lives in the long-term region was
remarkably smaller than that in the short-term region. Thus,
individual safety factors were proposed, and 60 year creep
damage was evaluated for two welds in the hot-leg pipe of an
SFR, as shown in Fig.2-7. As a result, it was confirmed thatthe stress at the welds was smaller than the allowable stress,
and that the SFR pipe was designed with enough reliability.

Reference
Wakai, T. et al., Creep Strength Evaluation of Welded Joint Made of Modified 9Cr-1Mo Steel for Japan Sodium Cooled Fast Reactor (JSFR), Proceedingsof the ASME 2010 Pressure Vessels & Piping Division/K-PVP Conference, Washington, DC, USA, PVP2010-26014, 9p., in CD-ROM.

40 JAEA R&D Review 2012

Research and Development of Advanced Nuclear System

The(MONJU)Core Can Be Cooled during a Station Black-Out

2-5=Core Cooling Capability during an Earthquake and Subsequent Tsunami=

Natural circulationin 3 loops

AC
Heat sink(Air cooler)

IHX

17 m

(Approximately)

Atmosphere

RV

7m

Heat source

Start ofpump operation

RV inlet temp.

Circulation ofcold sodiumPrimary sodium flow rate

0 s. Earthquake(Loss of off-site power)

AC
Reactor

Emergency DG

Above sea level

RV outlet temp.

(Approximately)(Reactor core)(from Tokyo Peil)+0 m

Flow rate(%)

Heat release to
the atmosphere

Temperature (͠)

Circulation of
hot sodium

Forced circulationin 1 loop+21 m+6.4 mBreakwater Curtain wall

SHTS

PHTS

SG
Floodwall

Sea-water pumps

13 m. Tsunami(Station black out)

Time (d)3 d. Recovery of asingle diesel generator

Fig.2-10 RV inlet/outlet temperatures and primarysodium flow rate (analytical results)
The plant will be cooled down to a cold shut-down state(coolant temperature below 250 ℃) after approximately
three days via natural circulation of the three heattransport systems.

Fig.2-9 Schematic view of the height difference and natural circulation mechanism at the MONJU plant
The reactor equipment is located at 21 m or more above sea level. The difference in height between the core andthe air coolers is approximately 24 m, which enables the coolant in the heat transport system to circulate naturally
on the basis of the temperature (density) difference, even after a circulation pump shut-down.

The decay heat of the core in the MONJU reactor should
be safely removed, even if the reactor is shut-down due toinsertion of the control rods during an earthquake.
Therefore, the coolant in the heat transport system (HTS)
is forced to be circulated by rotating the circulation pumps,
for which the power will be supplied by alternating-current(AC) fed by the emergency diesel generator, even if the offsite power supply is lost. The decay heat of the core can thenbe safely released to the atmosphere by switching the heatsink from the steam generators to the air coolers. However,the loss of all AC power (SBO: Station Black-Out) can be
imagined in the case of a beyond-the-design-basis hugetsunami that would destroy the sea-water pumps of thecomponent cooling water system and result in a diesel
generator shut-down. Therefore, the MONJU plant isdesigned to be able to cool down the core via natural
circulation induced by the temperature difference of thecoolant without the forced convection caused by thecirculation pumps. A height difference of approximately 24 m
between the core and the air cooler makes this coolingpossible. In addition, the major equipment is located 21 m or
more above sea level (Fig.2-9).
A detailed investigation of the capability to cool the coreusing natural circulation during an SBO induced by anearthquake and a subsequent tsunami was conducted on thebasis of these design considerations and considering theaccident at the TEPCO s Fukushima Daiichi NPS (1F). The
plant dynamics computer code Super-COPD was employed,
which was validated via an analysis of preliminary test results
for the natural circulation at MONJU .
The results showed that, during an SBO event in which atsunami attack is assumed 13 minutes after an earthquake
under a rated power initial condition (Fig.2-10), the plant will
be cooled down to a cold shut-down state after approximately
three days via natural circulation. In addition, it was clarified
that the core can be successfully cooled down even undervaried conditions, such as different tsunami delay times,
initial reactor powers, core-decay heats, and atmospherictemperatures. It was also confirmed that even if one HTS
fails, the two remaining HTSs have sufficient core coolingcapabilities.
Therefore, the reactor core of the MONJU reactor will
be safely cooled down even during an SBO event similar to
the accident at 1F. Natural circulation of the sodium coolant
will enable decay heat removal as long as the sodium-coolant
flow circuits are intact and secured.

ReferenceFBR Plant Engineering Center et al., External Evaluation on the Logical Adequacy of the Safety Measures and Coolability of the Reactor Core and othersin Monju Considering Earthquake and Tsunami (the Technical Committee on Severe Accident Countermeasures and Others Considering Tohoku District off the Pacific Ocean Earthquake), JAEA-Evaluation 2011-004, 2012, p.55-76 (in Japanese).
JAEA R&D Review 201241

Research and Development of Advanced Nuclear System

2-6

Toward Highly Efficient Production of MOX Fuel
=Confirmed Feasibility of Production from Agitated Granular Powder=

13wt%H2O

raw powder

14wt%H2O

15wt%H2O

Fig.2-11 SEM photomicrograph of MOX powder obtained using agitating granulation
Granulation was initiated when the added water fraction exceeded 10wt%. Addition of 13〜17wt% water resulted in goodflowability.
1.950 1.958 1.966 1.974 1.982 1.990

pelletejectionO/M ratio

2.00

core rodgreen pellet obtainedby die-wall lubrication
method

1.99
O/M ratio

54

average for sintering tray 1 㨪 8
calculation (1973 K)
calculation (2023 K)

die

50

1.98

Sintering
tray 1

4h

8h

12 h

18 h

㨪

Density (㧑T.D.)

58

1.97
side vent

46
0

200

400

600

Pelletizing pressure (MPa)

1.96lubricantlower punch

Sintering
tray 80

10
15
5
holding time(h)20

Fig.2-12 Relationship between pelletizing pressure anddensity
47〜55% of theoretical density (T.D.) was obtained by 220〜
550 MPa of pelletizing pressure. 93〜96%T.D. was obtained
after sintering.

Fig.2-13 O/M ratio control of MOX pellet in a small-scale
test
An O/M ratio of approximately 1.97 was obtained after heat
treatment at 2023 K under a control of PH2/PH2O ratio.

Low decontamination fuel has been developed for FBRs. A
highly efficient and simplified fuel production process has
been developed that requires one-third as many steps as theconventional process. The core techniques include agitating
granulation, die-wall lubrication, and control of the oxygen tometal (O/M) ratio.
Agitating granulation leads to the formation of a goodflowable powder that can be smoothly poured into a metallic
die. Water is added to a uranium-plutonium mixed oxide(MOX) powder obtained using the microwave de-nitration
method. With agitating granulation, the conventional
granulation process can be omitted. Investigation of therelationship between the granulation conditions and thepowder characteristics revealed that a large diameter
granulated powder with good flowability can be obtained byadding 13〜17wt% water (Fig.2-11).
The die-wall lubrication method involves the application ofa lubricant on the die-wall in order to decrease the friction
between the MOX powder and die wall. With this technique,conventional processes, such as the addition and mixing of
lubricant and presintering, can be omitted. As shown in
Fig.2-12, the density can be controlled without a nonuniform

compressive load, by applying the lubricant in a spiralmanner from the bottom up. The amount of lubricant can be
decreased from 0.2wt% to 0.017wt% using this method.
The O/M ratio can be adjusted to 1.97 during sintering inorder to prevent chemical reactions between the fuel andcladding. The O/M ratio depends on the oxygen potential,
which is affected by the PH2/PH2O in the H2/Ar gas used forsintering. With a conventional sintering furnace, a lack of
uniformity of the O/M ratio among different pellets is arecognized problem. It has been suggested that the H2 gas
remaining between the sintering plates cannot react
uniformly with the pellets. Therefore, a mesh plate wasdeveloped. In a small-scale (a few kilogram scale) test usingmesh plates, it was confirmed that the gas was equally
supplied to all of the pellets. The temperature andatmospheric conditions for control of the O/M ratio were thendetermined on the basis of the laboratory-scale test results.
As can be seen in Fig.2-13, the O/M ratio can be accurately
controlled.
Based on the results of this investigation, an efficient
production process was confirmed that involves agitating
granulation, die-wall lubrication, and heat treatment.

Reference
Hirooka, S. et al., Oxide Fuel Fabrication Technology Development of the FaCT Project(3) - Analysis of Sintering Behavior for MOX Pellet Production,
Proceedings of GLOBAL 2011, Makuhari, Japan, 2011, paper no.445139, 6p., in CD-ROM.

42 JAEA R&D Review 2012

3    
R&D Supporting the Technology and Reliability of Geological Disposal in Japan

Tono Geoscience Center(Toki & Mizunami, Gifu)

Cooperation and
Collaboration

Other Organizations

٨Mizunami URL㧖(Crystalline Rock)Overpack(e.g. carbon steel)

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

Vitrified
Waste

BufferMaterial

Engineered Barrier System

Geological disposal is aimed at isolating high-level
radioactive waste (HLW), which is produced during nuclear
power generation, from human environments over long
periods of time. This is a critical issue that the present
generation must take responsibility for dealing with and thatremains crucial irrespective of the revision of national nuclear
energy policy. In Japan, spent fuel from power reactors is
reprocessed to extract reusable uranium and plutonium for
power generation purposes. The liquids separated from thespent fuel during chemical reprocessing are solidified into astable glass form. In the Japanese concept, vitrified wastes
are encapsulated in a thick steel overpack, which issurrounded by highly compacted bentonite and then placed ina stable geological environment below 300 m (Fig.3-1).
Similarly, part of the low-level radioactive waste (LLW) will
also be disposed of in the geological environment.
Geological disposal of HLW tends to be a long-term
project, lasting over 100 years. The project begins with site
selection and continues to repository construction andoperation, which will be followed in turn by post-closure
monitoring. Thus, it is essential to proceed with the project
efficiently, treating it as a national responsibility, bycontinuously reinforcing the technical base and, more
importantly, enhancing public confidence. To this end, we
have made, and will continue to make, steady progress inresearch and development (R&D) in various fields, such asgeoscientific research, engineering technology, and safetyassessment, to improve the technology used for and thereliability of geological disposal in Japan. Our efforts are also
focused on promoting public understanding through thedissemination of relevant information and on opening our
R&D facilities to the public.
A particular focus of our R&D at present is on projects at
two underground research laboratories (URLs)−one atMizunami for research on crystalline rocks and the other at
Horonobe for research on sedimentary formations (Fig.3-2) −with the primary aim of developing a sound technical basis
for the implementation and the safety regulations. In March
2012, shaft excavation reached a depth of 500 m at Mizunami
and 350 m at Horonobe, and some research galleries weremade available for use. Multidisciplinary investigations are(e.g. bentonite)

Fig.3-1 Basic concept of geological disposal of highlevel radioactive waste (HLW) in Japan

ongoing as the reliability of various investigation techniquesmust be tested and confirmed before site characterization
begins in earnest. Underground tunnels serve as a place forthe public to experience deep geological environments
firsthand and appreciate our R&D activities. In addition,
studies of, for example, tectonics and volcanic and faulting
activity have been in progress in order to evaluate the longterm stability of geological environments in Japan.
In parallel with such geoscientific research, we areconducting an extensive study to assess the performance ofmultibarrier systems, engineered barrier systems, and longterm chemical and migration behavior of radionuclides at
Tokai in order to improve geological disposal technology.
These studies involve the use of geological environment dataobtained at both URLs. In 2011, particular focus was placedon the developing of sorption and diffusion databases for
cement and soil to contribute to the analysis of radionuclide
migration at the surface in relation to the accident at theTokyo Electric Power Company, Incorporated Fukushima
Daiichi Nuclear Power Station. The prototype knowledge
management system that was developed in 2009 has been
under improvement in order to systematically provide
multiple lines of evidence and R&D results.
Horonobe UndergroundResearch Center(Horonobe, Hokkaido)

JAEA Knowledge
Management System

٨Horonobe URL㧖(Sedimentary Rock)Knowledge Base

Tokai R&D Center(Tokai, Ibaraki)
Concept ImageENTRY

㧖URL: Underground‫ޓޓ‬Research Laboratory

Concept Image٨ Engineering-scale Test & Research Facility (ENTRY)٨ Quantitative Assessment Radionuclide‫ ޓ‬Migration Experiment Facility (QUALITY)QUALITY࡮Data
࡮Documents
࡮Software
࡮Experience and Expertise࡮Simulations
࡮Evidence࡮Expert Systems

Implementer

Regulator

Public Understanding
and Confidence

Fig.3-2 JAEA s R&D activities
JAEA R&D Review 201243

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

3-1

Study of the Groundwater Flow in Coastal Areas=Optical Measurement of the Salinity Distribution on the Saltwater Wedge=

Fig.3-3 Optical measurement of salinity distribution(e.g., saltwater intrusion)
The chamber is 250 mm high, 500 mm wide, and 100 mm deep. Salt
water dyed red is supplied from the left side and nondyed fresh
water is supplied from the right side. A saltwater intrusion
phenomenon appears in the medium.

Fig.3-4 Experimental results for saltwater intrusion(salinity distribution)
The white part at the lower left is the saltwater region and theblack part at the upper right is the freshwater region. The
contour lines are isosalinity lines. Salinity was converted fromreflected light intensity.

Fig.3-5 Simulation results for saltwater intrusion(salinity distribution)
This figure illustrates a result of the numerical simulation for thesame boundary conditions as in the laboratory experiment.
The white mixing zone appears between the saltwater region(red) and the freshwater region (blue).

A safety assessment for the geological disposal ofradioactive waste is based on the groundwater scenario ,which assumes transportation of radioactive nuclides from a
repository to the human living environment by groundwater.
For the simulation of radionuclide migration, an appropriate
model is selected on the basis of the characteristics of atarget. For targets in coastal areas characterized by thecoexistence of salt and fresh water, the parameters modeled
include saltwater intrusion caused by the density difference,
upconing of salt water with tunnel excavation, and change inthe position of the salt-fresh water mixing zone as a result ofsea level change. Although efforts have been made tounderstand these phenomena in situ, this understanding islimited owing to extremely complex groundwater flow andsalinity distribution caused by heterogeneity in rock facies
and fracture distribution. Consequently, numerical simulation
coupling multiple phenomena consisting of advection,diffusion, and density flow is essential to complement in situ
assessment.
To verify such groundwater simulation codes including
Dtransu , we constructed experimental equipment

incorporating a chamber filled with glass beads, which
simulates a homogeneous media for saltwater intrusion, andan optical measurement device with a white LED light sourceand high-resolution CCD camera to observe salinity
distribution (Fig.3-3). In the experiments, we measured theintensity of light reflected from the side surface of the salt
water, which is colored with dye of sufficiently low
concentration that acts as a tracer, using the optical device.
We quantified the salinity distribution by applying the
Kubelka-Munk formula to convert reflected light intensities
into tracer concentrations (Fig.3-4). The optical measurement
device can simultaneously quantify the salinity distribution
and temporal variation for the entire section without
disturbing the flow in the chamber. In addition, the spatial
resolution of the measurement has been greatly improved.
At present, we are comparing the results of ourexperiments with those of numerical simulations (Fig.3-5)
and verifying and upgrading the simulation codes. After
improving the experimental techniques, we intend to study
the phenomena of sea level change and upconing.

Reference
Oda, Y. et al., Optical Measurement of the Salinity Distribution by Saltwater Intrusion Experiment, Doboku Gakkai Ronbunshu Division C: Geotechnics,
vol.67, no.2, 2011, p.186-197 (in Japanese).

44 JAEA R&D Review 2012

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

3-2

Understanding the Impact of Nitrates on Geological Disposal Systems=Modeling of the Chemical Transition of Nitrates Accompanied by Corrosion of Carbon Steel=Fig.3-6 Reaction container (glass ampul)
and apparatus for analysis of gas phase(ampul opening member)
Experimental apparatus for measuring gaseous
reaction products during immersion of steel
specimens in a test solution.
Nitrite (Calc.)
Ammonia (Calc.)
Hydrogen (Calc.)

Nitrite (Measured)
Ammonia (Measured)
Hydrogen (Measured)

Metal Fe

Surface Film3n࡮Fen࡮Fe3O4

Aqueous Solution of NaNO3m࡮NO3-

+

m࡮H2O

2m࡮e-

m࡮NO2-+

2m࡮OH-

6p࡮e-p࡮NO2-+

5p࡮H2Op࡮NH3+

7p࡮OH-

2.67x3n࡮e-2q࡮e

Amount of species (mol/ampul)

1.0×10-4

1.0×10-5

1.0×10-6

1.0×10-7

1.0×10-8

1.0×10-9-02q࡮H2O
q࡮H2

100

200

300

400

Time (d)+2q࡮OH-

Fig.3-7 Conceptual model of nitrate evolution accompaniedby corrosion of carbon steel
Chemical reaction scheme among nitrate ions, carbon steel (Fe),and water, including electron transfer.

A number of low-level radioactive wastes contain large
amounts of nitrate salts. Nitrate ions are oxidizing species
that can be reduced to nitrite and ammonia by the corrosionof metals inside a repository for radioactive wastes. It is
possible that nitrate ions and their reduction products affectthe mobilities of radionuclides owing to changes in theirsolubilities and sorption distribution coefficients. Therefore,
estimation of the chemical evolution of nitrate ionsaccompanied by metal corrosion is necessary to assess thesafety of repositories for the geological disposal ofradioactive wastes.
Carbon steel, which is likely to be used extensively anddisposed of in repositories, was immersed in a closed
aqueous solution of NaNO3 (Fig.3-6) to observe the chemicalinteraction between metal and nitrate ions. The experimental
pH range of the NaNO3 solution was 12.5~13.5, which was
designed to simulate conditions imposed by the pore fluid of
cementitious material. Despite the accelerated cathodic
reaction due to the existence of nitrate and nitrite ions, theresults of the immersion experiments showed no accelerationof the corrosion rate of carbon steel in this pH range. Thisimplies that the system is controlled by an anodic reaction.
The nitrate reduction accompanied by the corrosion of carbon

Fig.3-8 Variation of reaction products (NO2−, NH3, and H2)
with time in an ampul
Measured and calculated variation of reaction products at
323 K, pH 12.5, and 5.0 mol/dm3 of NaNO3.

steel is considered to be a reaction series of NO3−→NO2−→NH3.
The nitrate reduction reaction competes with the water
reduction (hydrogen evolution) reaction under the anodic
controlled conditions. Therefore, the nitrate ions strongly
reduce the hydrogen generation rate in a 1.0 mol dm-3 NaNO3
solution.
Accordingly, the sum of the current of the reaction seriesNO3− →NO2− →NH3 and that of the water reduction reaction
was assumed to be balanced with the anodic current ofcarbon steel (conceptualized in Fig.3-7). Examples of thesimulation results are shown in Fig.3-8. To develop asystem for estimating nitrate impacts on the chemical
conditions inside and outside of repositories for radioactive
waste, this model will be combined with a model of nitratechemical evolution caused by microbial activity and mineralreactions outside a repository.
The present study was sponsored by the Ministry of
Economy, Trade and Industry (METI) through the program
Combined development of nitrate salt removal technology
and an assessment system for the impact of nitrate on the colocational disposal of TRU waste and HLW (2007, 2008,2009, 2010).

Reference
Honda, A. et al., Understanding and Modeling of Chemical Transition of Nitrate Accompanied with Corrosion of Carbon Steel under Hyper-alkaline andHigh Nitrate Concentration Conditions, Zairyo to Kankyo, vol.60, no.12, 2011, p.541-552 (in Japanese).
JAEA R&D Review 201245

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

3-3

Challenges of Seepage Control at a Repository for Radioactive Waste=Applicability of a Grout Penetration Model=
One of 4 pressure gauges

Fig.3-9 Flat parallel plate apparatus (top view)Two flat parallel plates form a narrow aperture. Infiltration
distances of the injected grout from the inlet were observed
through acrylic windows and infiltration pressures weremonitored by pressure gauges.

Permeation distance (mm)One of 8 observation windows

1000
800
600
400
Visual measurement

200

Pressure measurement point
Theoretical solutions

0

0

5

10

15

20

25

30

Aperture

Pg
Injectionpressure
produced by
the grout
pump

Grout

Pw
Groundwater pressure

L
Actual penetration depth

Rock

Permeation distance (mm)Time (s)
1000
800
600
400
Visual measurement

200

Pressure measurement point
Theoretical solutions

0

0

10

20

30

40

50

60

Time (s)Fig. 3-10 Grout flow in an aperture
Grout is injected into apertures in the rock with injection
pressures higher than groundwater pressure. Penetration
distance of the injected grout is referred to as L .

Fig.3-11 Results of modeling and laboratory tests ofpenetration
The modeling is in agreement with the results of laboratory
tests: the penetration fronts were solved theoreticallyaccording to the model plot close to the averages of thoseacquired by observation and from the pressure gauges in thelaboratory tests.

Grouting open apertures in a rock mass improves thestrength of the rock mass and limits the ingress ofgroundwater during the construction and operational phasesof facilities for the geological disposal of radioactive waste.
Ordinary Portland cement is conventionally used for
grouting, but which can generate a high pH plume that has
generally uncertain effects on the performance of otherfacility components and of the rock mass. A low pH cement
grout material and appropriate grouting techniques are
therefore being developed to mitigate these effects.
Assessments of penetration depth are particularly importantbecause they can be used to provide information for longterm safety assessment.
In the current study, the grout penetration depth of a low
pH cementitious material was tested in the laboratory using 1Dflat parallel plate apparatus and compared with predictions
using the Gustafson & Stille model (Fig.3-9). Fig.3-11 showsthat the modeling predictions are in agreement with theresults of the laboratory tests: the evolution of the penetration
distance solved theoretically according to the model wasapproximately consistent with that of the average distancemeasured visually and by pressure gauges at the observation
points.
The model provides the average distance of injected groutin the aperture (Fig.3-10). Grout material is simulated as aBingham fluid, which exhibits a nonzero (i.e. positive) yieldstrength on the stress vs. rate of strain relationship. Bingham
fluids exhibit heterogeneous flow velocity distributions in the
cross section of the aperture; in particular, the flow velocity isslower closer to the wall of an aperture.
The current study was conducted in FY2008 as a part of
the Project for Grouting Technology for the Geological
Disposal of High Level Radioactive Waste funded by the
Ministry of Economy, Trade and Industry (METI).

Reference
Fujita, T. et al., Fundamental Study on a Grout Penetration Model for a HLW Repository, Journal of Energy and Power Engineering, vol.6, no.8, 2012,p.1191-1203.

46 JAEA R&D Review 2012

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

3-4

Looking at Huge Earthquakes through a 0.1 mm Grain
=Thermochronology with a Single Grain Using the (U-Th)/He Method=
Fig.3-12 Mass spectrometer for helium dating
Helium extracted from the zircon heated by the laser beam in thesample chamber is quantified using the mass spectrometer.

Laser projection lensSample chamberStandard gas
Heating laser

Vacuum pump
& pressure meter

Mass spectrometer
Vacuum pump
& pressure meter

0.1 mm1 mm

Fig.3-13 Sample grain for dating(Right) A grain of zircon handpicked under a microscope after
separation by heavy liquid.
(Left) A handling packet containing a grain of zircon forlossless analysis.

Fig.3-14 Fault rock in the Median Tectonic Line
Extraordinarily thick pseudotachylyte, which usually exhibits
thicknesses of around 1 mm in active Japanese faults, isidentified.

Radiometric dating is used to determine when geologicalevents occurred. Some dating methods can determine thelength of time that has passed since a rock cooled to aparticular temperature, defined as the closure temperature.
Such methods are referred to as thermochronologicalmethods. The combination of methods with different closure
temperatures can reveal the cooling history of a rock;additionally, denudation history can be determined becausetemperature within the Earth tends to increase with depth.
We are studying a single-grain (U-Th)/He dating method.
This method is characterized by its low closure temperature,
which enables the comprehension of shallow geomorphology.
This method has only recently been introduced and our mass
spectrometer (Fig.3-12) is the only instrument in Japan.
Dating targets, mainly zircon grains, have extremely small
dimensions, usually around 0.1 mm (Fig.3-13).
Fig.3-14 illustrates a fault rock collected from the Median
Tectonic Line, the longest active fault in Japan, in Mie

Prefecture. Pseudotachylyte was formed by frictional meltingof wall rocks during rapid fault movement associated with aseismic event and was followed by rapid cooling. Mylonite
was formed by fault movement without frictional melting. A
combination of several radiometric ages, including the(U-Th)/He age, indicates that this rock intruded at 90 Ma andcooled relatively quickly to 200 ℃ by 60 Ma; then thepseudotachylyte was formed at depths of around 5 to 10 km
by huge earthquakes and cooled slowly to the present
temperature (Yamada et al., 2012). Such a cooling history is abasic piece of information that can help in the interpretationof the present topography of the Japan arc and the conditions
that lead to earthquakes along faults.
We will use this method for smaller samples or otherminerals and we intended to develop dating techniques tohelp improve the understanding of geological events, whichis essential for the prediction of future changes in thegeological environment.

Reference
Yamada, K. et al., The First (U-Th)/He Thermochronology of Pseudotachylyte from the Median Tectonic Line, Southwest Japan, Journal of Asian Earth
Sciences, vol.45, 2012, p.17-23.
JAEA R&D Review 201247

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

3-5

Fault Zone Determination through Multivariate Analysis=Case Study Using a Dataset from a Deep Borehole=

0

ԛ
Ԛ

ԙ
-5

5

0

-5

-10

-15

-20

First component scores
Ԙ

ԙ

Ԛ

ԛ

Ԝ

ԝ

70.2
100

100

200

200

300

300 311.0

400

400

500

500

600

600

109.1

195
220

700

800

800

691
720
721

825
873

900

1000

1100

900

1000

0 5 -5 01015

Fault Zone Section

Results ofCluster Analysis
1

2

3

4

5

919
983

6

Moving Average(20 m intervals)

Section 1

589
593
654

700

Second Component
Score

First Component
Score

Fault Core

Strata Name

-20 -10

Mizunami
Upper highlygroup
fractured domain

Cluster1
Cluster2
Cluster3
Cluster4
Cluster5
Cluster6

Results of Multivariate Analysis

Section 2
Lower sparsely fractured domain

Ԙ5Toki Granite

ԜDriling Depth (m)10

Fig.3-15 Scatter plot of the firstand second component scoresand clustering results
The first component score, inwhich resistivity, density, andelastic velocity are dominant, isclosely related to the rigidity of rock.
The second component score, inwhich self-potential, resistivity, andborehole diameter are dominant,is closely related to water content.
Numbers shown with circles
correspond to numbers in Fig.3-16.

Lithostratigraphic
Columns

Core Description

ԝ

Vertical Depth (m)Second component scores

15

Section 3
Section 4

Section 5

Section 6

1100 1095
1117
1154
1168
1200

Section 7
Section 8

1200

Cluster1499㨪501 m

Cluster2
190㨪192 m

Cluster3
667㨪669 m

Cluster4
930㨪932 m

Cluster5970㨪972 m

Cluster6967㨪969 m

Non-altered㨪
weak-altered rockWeak
altered rockModerate
altered rock

Strongaltered rock

Strongaltered rock

Core loss(Unrecovered core)Fig.3-16 Borehole cores corresponding to the clusters
Rock condition becomes worse toward the right. Clusters
3~6 correspond to fault zones. Drilling depths are shown at
the top of each core image.

A fault with a crush zone can strongly affect themechanical, geochemical, and hydrological properties of arock mass. Consequently, fault zone determination is treatedas an essential element in evaluating the undergroundgeological environment and the engineering performance ofrocks. Nonetheless, the results may differ depending on thedetermination accuracy, because of the limited quantity andquality of geological data obtained, and/or depending on thegeological knowledge and skill of the person in charge. As acase study for reducing such uncertainty, we tested a fault
zone determination method (incorporating multivariate
analysis composed of principal component analysis andclustering) on data from a deep borehole investigation in the
Mizunami Underground Research Laboratory.
A 1300-m-long borehole, which encountered several fault
zones, was selected for analysis. Quantitative data was
acquired along the borehole by means of core description,
geophysical logging, and a borehole TV. A principle

1300.00

Extracted fault sections

Fig.3-17 Fault sections in the borehole
Fault sections 1~8 show fault zones extracted bymultivariate analysis.

component analysis, which enables the evaluation ofcorrelations between data from different sources, wasconducted; then, a scatter plot was produced with the first andsecond component scores clustered according to the k-means
method (Fig.3-15). Fig.3-16 illustrates the rock conditions
corresponding to the clusters in Fig.3-15. Because clusters
3~6 have been identified as fault zones from thecorresponding core descriptions, sections of these clusters
were extracted as fault sections along the entire borehole(Fig.3-17). According to examination of the geologicalfeatures of the extracted sections, the sections were
accurately identified as fault zones (Fig.3-17). Thus, thismethod is assumed to be effective for fault zone
determination and for providing useful geological andengineering information for 3-D geological modeling andgroundwater flow analysis.
Our research was selected as an excellent research paper in
Geoinformatics in 2011.

Reference
Abumi, K., Tsuruta, T. et al., Fault Zone Determination and Bedrock Classification through Multivariate Analysis −Case Study using a Dataset from a
Deep Borehole in the Mizunami Underground Research Laboratory−, Joho Chishitsu, vol.22, no.4, 2011, p.171-188 (in Japanese).

48 JAEA R&D Review 2012

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

3-6

Technology for Estimating Hydrogeological Characteristics Deep Underground=Evaluation of Hydrogeological Characteristics Using Water Pressure Data=
MSB - 1(#3-#5)DH - 15-10.0

MSB - 3 (#3)
NNW fault

MSB - 3 (#4)

MSB - 1 (#1-#2)
MSB - 2 (#1-#3)
MSB - 3 (#1-#2)
MSB - 4 (#1-#4)

Group 1-B

10.0Group 1-A

MSB - 2 (#4)

30.0

MSB - 2 (#5)

MIZ - 1

Group 2-B

50.0

Log (t/r 2)
70.0

Group 1-B
Group 1-A

90.0

MSB-2 (#6-#10)
DH - 2

Group 2-B
s

130.0

150.0
1.00E+01

05ME06

Group 2-A-

1.00E+02

1.00E+03
2

2

t/r (sec/m )

MSB-2

MSB-3

05ME06

DH-2

MSB-1

1.00E+04

1.00E+05

̪Borehole name (#Observation section number)

DH-15

MSB-4

MIZ-1
Akeyo/Hongo Formation
Toki Lignite-bearing Formation

Mudstone
layer

MSB-4
MSB-1

MIZ-1

LAFZ
UHFD

Ventilation
Shaft
Main Shaft

MSB-3
05ME06
Ventilation
Shaft

LSFD

Fault A

Sectional view

MSB-2

DH-2
0m

100 m

200 m

Plan top viewA

Researchlaboratory lot

Group 1-A in s-Log (t/r2) plot
Group 1-B in s-Log (t/r2) plot
Group 2-A in s-Log (t/r2) plot
Group 2-B in s-Log (t/r2) plot
Categorized not clearlyCenter of a measurement section
Confirmed fault

Researchlaboratory lot
Main Shaft

ult
Fa

1328 m (MIZ-1 Perpendicular depth)

MSB - 3 (#5-#7)

Mizunami Group110.0s㧦drawdown
Group 2-At 㧦time (since2005/1/1)
r 㧦distance of a vertical shaft toan observation boreholeGrouping by s-log (t/r 2)

Toki Granite

drawdown (m)MSB - 4 (#5-#7)

Fig.3-18 s(drawdown)-log(t/r2)
plot
The x-axis is the logarithm of t/r2
and the y-axis shows thedrawdown in each observation
section.
Group 1-A exhibits a small
amount of drawdown. Group 1-B
does
notexhibitobviousdrawdown. Group 2-A exhibits alarge amount of continuing
drawdown. Group 2-B exhibits alarge amount of drawdown,
which has already been balanced.

DH-15

Fig.3-19 Borehole grouping based on s-log(t/r2) plots (section view, plan top view)

Estimation of groundwater flow and hydrogeologicalcharacteristics deep underground is important for assessmentof the safety of the geological disposal of high-level
radioactive wastes (HLW).
For the Mizunami Underground Research Laboratory(MIU) project, we are constructing a scientific andtechnological basis for the geological disposal of HLW.
MIU is now under construction in Mizunami City, Gifu
Prefecture. Long-term groundwater pressure monitoring isbeing conducted in and around the MIU construction site inorder to observe the groundwater pressure changes due toshaft excavation.
Such changes in groundwater pressure are assumed to be
an analogous to those caused by large-scale pumping tests. In
this study, hydrogeological characteristics were estimatedusing monitoring data and s-log(t/r2) plots based on the
Cooper-Jacob straight line method.
The groupings of the monitoring section, based on trends
in groundwater pressure changes, are shown in Fig.3-18.
Fig.3-19 illustrates the locations of the monitoring sections
and the result of groupings in and around the MIU
construction site. The results indicate that the groundwaterflow field is separated into areas, which can be further
subdivided into individual groups, bounded by low permeable
structures such as Fault A and the mudstone layer in the
Mizunami Group.
The study suggests that an evaluation method usings-log(t/r2) plots is an effective method for hydrogeologicalcharacterization of heterogeneous fractured rock.

Reference
Daimaru, S., Takeuchi, R. et al., Hydrogeological Characterization Based on Long Term Groundwater Pressure Monitoring, Proceedings of the 13th
International Conference on Environmental Remediation and Radioactive Waste Management (ICEM2010), Tsukuba, Japan, 2010, ICEM2010-40064, 10p.,
in CD-ROM.
JAEA R&D Review 201249

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

Development of a Methodology for Reconstructing Very Long-Term Geosphere Evolution=A Very Long-Term Groundwater Flow Analysis Using Numerical Simulation=

East

West
(a) Coastal area(b) Hill area

Present

present
seawater=1.0

0.8
0.6
0.4

Salinity0

Present
seawater =1.0

400 kilo years ago(b) Hill area(a) Coastal area

1.0

Salinity

3-7

20

40

60

80

100

120

Time (x10 kilo years ago)

Present

140
150 x 10kilo years ago

1.0

Ground surface at (a) coastal area

Ground surface at (b) Hill area

Sea level

800 kilo years ago

1500 kilo years ago

5 km

0.5

Elevation (m)

100
50
0

-50
-100
-150

10 km
0.0

Location of the shoreline̪Black lines show geological boundaries and faults

Fig.3-20 Evolution of groundwater salinity
The combined effects of the sea-level change and crustal
movement play an important role in the evolution ofgroundwater salinity in the Horonobe coastal area.

It is crucial to demonstrate how the geosphere evolves overtens of thousands of years to ensure acceptance of the very
long-term safety of geological disposal by society at large.
One of the main methods for describing such future evolution
is the extrapolation of past evolution based on phenomenologicalunderstanding; another method is numerical simulation byconsidering the mechanism of geosphere evolution. In thisstudy, we reconstructed the evolution of groundwaterflow/salinity over the past 1500 ka in Horonobe, northern
Hokkaido, using numerical simulation developed at JAEA
toward the development of a methodology for describing
geosphere evolution over very long time periods in coastaland sedimentary environments.
We developed a methodology involving numericalsimulation of the evolution of groundwater flow/salinity byconsidering geometric changes of topography/geologicalstructures caused by uplift/subsidence as well as changes ingroundwater recharge and sea level as a result of climate

0

20

40

60

80

100

Time (x10 kilo years ago)

120

140
150 x 10kilo years ago

Fig.3-21 Evolution of groundwater salinity at 300 m depth
Groundwater salinity has been decreasing with uplift from 800 ka(upper figure) but differences in the evolution of altitude cause
differences in long-term trends in salinity for (a) coastal and (b) hill
areas (lower figure).

change. Application of the methodology to the Horonobe
area suggests groundwater salinity decreases gradually in
uplifted land areas, which were originally on the sea floor at
1500 ka (the uplifted area at the eastern end of Fig.3-20). The
results also indicate the wide distribution of groundwater ofvery low salinity, i.e., fresh water, under the sea floor atpresent (the coastal area in uppermost Fig.3-20). This fresh
water distribution was formed by the infiltration of rainwater
when the present sea floor was land. The groundwatersalinity at 300 m depth decreases with uplift in both coastal
and upland areas, exhibiting different trends depending on theelevation change in each area (Fig.3-21).
The simulation results are confirmed by geophysical andborehole investigations. Furthermore, this study contributes
to improving the technical basis for describing evolution of
the geosphere over very long time scales in coastal andsedimentary environments.

Reference
Niizato, T. et al., Development of a Methodology for the Characterization of Long-Term Geosphere Evolution (1) Impact of Natural Events and Processeson the Geosphere Evolution of Coastal Setting, in the Case of Horonobe Area, Proceedings of 19th International Conference on Nuclear Engineering(ICONE19), Chiba, Japan, 2011, ICONE19-43163, 11p., in CD-ROM.

50 JAEA R&D Review 2012

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

3-8

Examination of the EDZ Developed around a Tunnel=Estimation of the EDZ by Seismic Tomography Investigation=

Excavation

Ventilation shaft side

6.5

6.5

6.5

6

6

6

6

5.5

5.5

5.5

5.5

5

5

5

4.5

4.5
0.5

Excavation 20
Excavation 45

Excavation 41

1

1.5

2

2.5

Excavation 1

Excavation line
Steel support
Shotcrete

4.0 m
0

1.5

2

2.5

0.5

3

When a tunnel in an HLW repository is excavated, an
excavation damaged zone (EDZ) is formed around the tunnel.
Inside this zone, changes in the characteristics of a rock mass
occur; for example, the degree of damage depends on therelationship between the strength of the rock and the effective
in situ stresses. Such changes will affect the migration ofradionuclides; therefore, estimation of the various
characteristics of EDZs is essential for the geologicaldisposal of HLW. The Horonobe Underground ResearchLaboratory (URL) Project, involving research anddevelopment for the geological disposal of HLW in
diatomaceous mudstone, concerns modeling EDZ behavior
based on the understanding of the mechanisms ofdevelopment and evolution of EDZs.
Seismic tomography is a geophysical survey method that
can image conditions inside a rock mass, similar to medical
X-ray CT scans. Thus, it is an effective technique for2

2.5

Excavation 1

2

2.5

3

0.5

6

6

5.5

5.5

5

2

2.5

1

1.5

2

2.5

3

0.5

68 days later
6.5

1

1.5

2

2.5

0.5

3

85 days later
6.5

6

6

6

5.5

5.5

5.5

5

5

5

4.5

1.5

2

2.5

1

1.5

2

2.5

3

1

1.5

2

2.5

3

0.5

6.5

6.5

6

6

6

5.5

5.5

5.5

5.5

5

5

5

2.5

3

0.5

1

1.5

2

2.5

2.5

3

1

1.5

2

2.5

3

6.5

5

4.5

4.5
2

2

301 days later 350 days later

6

1.5

1.5

4.5
0.5

6.5

1

1

5

4.5
0.5

3

3

6.5

6

1

2.5

149 days later

5.5

4.5

2

4.5

4.5
0.5

3

1.5

5

5

4.5

1.5

1

6.5

6

1

3

Excavation 9

5.5

0.5

4.5
0.5

3

1

1.5

2

2.5

3

0.5

1

1.5

2

2.5

3

505 days later 553 days later

407 days later 469 days later

Velocity variation
drawing area
Hammer location
Receiver
Seismic survey line

1.5

6.5

6.5

4.5

1m

1

Excavation 7

Excavation 6

2.5

4.5
0.5

3

6

6.5

6.5

Fig.3-22 Layout of the EDZ experiment
Receivers are installed in boreholes to investigate theinfluence of excavation and a hammer source is installed
for each recording operation.

1.5

211 days later 254 days later0

140 m Gallery (Plan view)

1

2

5

4.5

4.5
1

1.5

6.5

5.5

3m

East Access shaft side

6.5

1

Excavation 4

5

Excavation 4

Ǿ86 mm0.5

Excavation 3
6

0.5

Start section

3

5.5

5

Excavation 2

2.5

5

4.5

Excavation 3

2

6

6.5

Seismic tomography‫ޓ‬Hammer location‫ޓ‬Receiver‫ޓ‬Borehole

1.5

5.5

Excavation 11

Excavation 1

1

5.5

0.5

Excavation 5

4.5
0.5

5.5

6.5

Seismic
tomography
Excavation 5 investigation area

3

Excavation 2

Excavation 5

Excavation 6

2.5

6

0.5

(Bird’s-eye view)

2

6.5

5

Excavation 10

1.5

6

Excavation 35

Seismic
tomography
investigation area

1

6.5

4.5

Excavation 15 Excavation 31

5

4.5
0.5

3

6.5

6

6

5.5

5.5

5

5

0

1m

4.5

4.5
0.5

1

1.5

2

2.5

3

0.5

1

1.5

2

2.5

3

Velocity variation (㧑)602 days later 650 days later0-5-10 -15 -20 -25 -30

Fig.3-23 Variation of seismic velocity
Each diagram illustrates ratios of velocity variation to initial velocity
corresponding to the progress of excavation. Decreases in seismic
velocity and their distribution are observed.

investigation of EDZs. It is known that seismic velocity isslower in rocks with cracks than in those without cracks.
Seismic tomography was applied to examine the influenceof excavation at the 140 m Gallery of the Horonobe URL.
Fig.3-22 illustrates a plan view of the layout of an EDZ
experiment. Fig.3-23 is an example of the results of theseismic tomography surveys. It is apparent that regions withvelocity decrease appeared in the area surrounding the tunnel
as excavation progressed. It is also noted that the seismic
velocity decreases by 10% further from the tunnel but
25~30% closer to the tunnel, and regions exhibiting velocity
decreases extended to depths 0.5~1.0 m from the tunnel wall;such regions expanded even after excavation ceased.
We intend to examine the influence of EDZs at different
depths in order to understand the mechanisms underlying thedevelopment and evolution of EDZs and to model EDZ
behavior.

Reference
Sugita, Y. et al., Development of a Simplified Seismic Tomography Survey System, JAEA-Research 2011-043, 2012, 25p. (in Japanese).
JAEA R&D Review 201251

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

3-9

Evaluation of the Initial Stress Distribution Deep Underground=Three-Dimensional Initial Stress Measurement by Hydraulic Fracturing=(a) 140 m gallery

ǻ1
ǻ2
ǻ3
ǻ1
ǻ2
ǻ3

N

2.1

2.3

W
2.4

E
1.1

2.2

HDB-1
HDB-3
HDB-6
HDB-11

140 m niche off
Ventilation Shaft No.1
140 m niche off
East Shaft No.1

140 m niche off Ventilation Shaft No.1
140 m niche off East Shaft No.1
250 m niche off Ventilation Shaft No.1
250 m niche off East Shaft No.1
250 m niche off West Shaft No.1

0

The normal vector of the geologicalboundary plane

100

The value expresses themagnitude of stress
Unit : MPa

200S(b) 250 m gallery

ǻ1
ǻ2
ǻ3
ǻ1
ǻ2
ǻ3
ǻ1
ǻ2
ǻ3

N
1.6

2.6
2.2

1.8

W

2.2

E
2.8

1.4
4.0

250 m gallery off theVentilation Shaft
250 m niche off
East Shaft No.1

Depth (m)

1.0

300
400

Geological boundary(HDB-6)500

250 m niche off
West Shaft No.1

The nomal vector of the geologicalboundary plane

The value expresses themagnitude of stress
Unit : MPa

2.1

S

600
700
-90

W

-60

-30

0

30

60

N

90

E

Direction of the maximum
horizontal principal stress (q)Fig.3-24 Lower hemisphere projection (Wulff net) of eachprincipal stress in the 140 m and 250 m galleries
σ1, σ2, and σ3 are the maximum, intermediate, and minimum
principal stresses, respectively. The Wulff net is an equal-angle
projection in which the plotted direction of each principal stresson the lower half of a sphere is projected onto the equatorial
plane toward the zenith of the sphere.

Fig.3-25 Depth distribution of the direction of the
maximum horizontal principal stress
The data of HDB-1, 3, 6, and 11 represent the direction of the
maximum horizontal principal stress obtained in the surfacebased investigation. The data from the galleries represent thedirection of the maximum horizontal principal stress calculated
from the three-dimensional principal stresses.

Information about initial stress distribution is importantwhen designing the specifications of support members andthe direction of drifts in an underground facility. In the
Horonobe Underground Research Laboratory Project, theinitial stress was measured by hydraulic fracturing using deep
boreholes (520~1020 m) in the surface-based investigationreferred to as Phase 1. Each principal stress was obtained
under the assumption that one of the principal stresses isalong the borehole direction, i.e., the vertical direction.
However, this assumption remains to be investigated, because
the site is located in the vicinity of an anticline axis affectedby folding deformation. To evaluate the initial stress obtainedin Phase 1, spatial variation in initial stress was investigatedby hydraulic fracturing at several depths in the undergroundlaboratory. Measurements were conducted using threeboreholes with different orientations at each measurement
point, such that the characteristics, magnitudes, anddirections of the three-dimensional principal stresses could beobtained without any assumptions regarding the directions of

the principal stresses. Fig.3-24 illustrates the Wulff net
projecting each principal stress in the 140 m and 250 m
galleries. In the 140 m gallery, one of the principal stresses isclose to the vertical direction and this was valid according to
the assumptions of Phase 1. Conversely, even the principal
stress closest to the vertical direction dipped at 35~39°fromthe vertical direction in the 250 m gallery. Because the 250 m
gallery is developed near a geological boundary, thedistribution of the principal stress is probably affected by
local factors. Fig.3-25 illustrates the direction of the
maximum horizontal principal stress. The results of Phase 1
agreed with the results for the 140 m and 250 m galleries.
Therefore, initial stress measurements from the surface of theground may be used to evaluate the direction of the
maximum horizontal principal stress correctly.
We intend to conduct initial stress measurements at
different depths in order to elucidate the influence ofgeological structure and depth on the initial stress distribution.

Reference
Nakamura, T., Sanada, H. et al., In Situ Stress Measurements at 250 m Gallery off the Ventilation Shaft, JAEA-Research 2011-012, 2012, 217p. (in Japanese).

52 JAEA R&D Review 2012

4  
Toward Practical Use of Fusion Energy
Fusion Plasma Research

DEMO
ReactorFusion Frontier Research

Toward ITER and DEMO٤Develop Plasma‫ޓ‬Performance for DEMO
٤Support ITER
٤Cultivate Human‫ޓ‬Resources
Large Tokamak Device JT-60‫࡮ޓ‬Exp. was accomplished‫࡮ޓ‬Upgrade to JT-60SA started

ITER Project
Domestic Agency for ITER Project
Demonstration of Fusion Energy
࡮Cadarache‫ޓ‬in France࡮Construction‫ޓ‬Started in‫ޓ‬Dec. 2007

JT-60

Develop Technological
Basis for DEMO
٤R&D of Breeding Blanket
٤DEMO Design, R&D
٤Simulation of Burning‫ޓ‬Plasma, Theoretical‫ޓ‬Model٤Remote Experiment٤Engineering Validation‫ޓ‬and Engineering Design‫ޓ‬Activities 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 Development steps toward the Fusion DEMO Reactor
Fusion plasma research and fusion frontier research are being pursued to develop a DEMO reactor, aiming for the early realization offusion energy.

Crucial research and development on fusion plasma
research and fusion frontier research toward the practical use
of fusion energy are being pursued through intensive
international cooperation such as the International
Thermonuclear Experimental Reactor (ITER) project,
Broader Approach (BA) activities, and other collaborations(Fig.4-1), aiming for the early realization of a fusion DEMO
reactor.
ITER project
The ITER project is an international cooperative project todemonstrate 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 JAEA was designated as the domestic
agency of the ITER project in Japan. JAEA has proceededwith preparing the equipment that Japan was committed toprovide. In particular, JAEA has continued mass productionof a superconducting coil conductor (Topic 4-1) ahead of any
other country participating in ITER, and has made various
technological developments (Topics 4-2, 4-3).
BA activities
The BA activities are joint projects by Japan and EU for
executing support research for ITER and the research anddevelopment for a DEMO reactor, which is the next step of
ITER, aiming for the early realization of fusion energy. The
BA agreement came into force in June 2007, and JAEA wasdesignated as the implementing agency of BA activities inJapan.
BA activities consist of three projects: (a) projects on theInternational Fusion Energy Research Center (IFERC), (b)Engineering Validation and Engineering Design Activities of
the International Fusion Material Irradiation Facility(IFMIF/EVEDA), and (c) the Satellite Tokamak Program(STP). Topic 4-4 reports the progress made on the IFERC
project. Topic 4-5 describes the results achieved along withthe researches related to IFMIF/EVEDA. In STP, theconstruction activities of JT-60SA have progressed well,including the design and manufacture of components(Topics 4-6, 4-7, 4-8).
Fusion plasma research
Analysis of JT-60 experimental data has been performed
and inter-machine experiments have been conducted for the
purpose of achieving high economical efficiency of fusion
reactor by attaining high plasma pressure. Topic 4-9 is theresult to contribute to overcoming the instability that is thedominant factor to hinder the achievement of high plasmapressure.
Fusion frontier research
In the fusion reactor, tritium is produced (breeding) fromlithium by using the neutron produced by the fusion reaction.
Topic 4-10 is intended to efficiently acquire the lithium
necessary for fuel breeding. Topic 4-11 is intended to obtain
results that will form the basis for controlling the plasma in a
DEMO reactor.
JAEA R&D Review 201253

Nuclear Fusion Research and Development

4-1

Manufacture of Superconductors for ITER Coils=Progress in Procurement of Superconductors for TF Coils, with Japan Leading in the World=

Central Solenoid (CS) coil
(Nb3Sn, 6 modules,
40 kA,13 T)

Toroidal Field (TF) coil
(Nb3Sn, 18 coils,
68 kA,11.8 T)

Radial Plate

14 m

9.0 m

Conductor
Winding Pack and Coil Case
Superconducting
Cable
Insulation
JacketFinal Cable Wrapping
Central Cooling
ChannelTF coil

Poloidal Field (PF) coil (NbTi, 6 coils, 45 kA, 4 㨪 6 T)

Fig.4-2 ITER superconducting coil system
The system consists of 18 Toroidal Field (TF) coils, six
Central Solenoid coils and 6 Poloidal Field coils.

Relative Frequency (㧑)30

Sub Cable Wrapping

Superconducting Conductor for TF coilFig.4-3 Structure of TF coils and their superconductor
A circular, multistage superconducting cable, consisting ofaround 1000 superconducting strands, is inserted into a
circular stainless steel jacket. The superconductor issubsequently inserted into the groove of a radial plate.

25
20
15
10
5
0
200

210

220

230

240

250

260

270

Critical Current (lc) at 4.2 K, 12 T (A)Fig.4-4 Histogram of critical current (Ic) of mass-produced
strands
Critical current (Ic) is the maximum current at which strands
can be maintained in a superconducting state at 4.2 K and amagnetic field of 12 T.

Fig.4-5 Completed superconductor with a length of 760 m
The superconductor is wound at a diameter of 4 m.

To confine and control a very high-temperature plasma, the
ITER device being constructed by Japan, China, the EU,
India, Korea, Russia and the US, will have the maximum
magnetic field of 10 T. Huge superconducting coils arenecessary to achieve such a high magnetic field (Fig.4-2).
The height, width and weight of a Toroidal Field (TF) coil,which generates the magnetic field, are 14 m, 9 m and about
300 tons, respectively (Fig.4-3). Japan is producing nine TF
coils and 25% of all TF superconductors, and is the first of all
ITER countries to begin manufacturing the TF conductors.
A circular superconducting cable, consisting of around
1,000 superconducting Nb3Sn strands, is inserted into a
circular stainless steel jacket. The jacket is composed oftubes, each 13 m in length, that are butt-welded with anautomatic welding machine to create a continuous 780 mlong jacket. Following cable insertion, the jacket iscompacted in a single step to the specified diameter (43.7 mm).
The conductor is manufactured under a high-level quality
assurance plan according to the ISO standard. Fromexperience, it is known that if the fabrication processes are
not tightly monitored and controlled, the performance ofsuperconducting strands can widely vary. Therefore,
Statistical Process Control (SPC) is applied during strand
fabrication to monitor all key parameters including Ic,
hysteresis loss, residual resistance ratio and outer diameter.
This monitoring allows us to detect, at an early stage, any
problems in the manufacturing process and any deviations ofparameters from design specifications. The Ic values ofmanufactured strands show a Gaussian distribution, as shownin Fig.4-4. The standard deviation of the distribution is 6.7 A,
which is around 3% of the average value (233 A). This result
indicates that the quality assurance plan using SPC is very
useful for the manufacture. The completed TF conductor isshown in Fig.4-5. We are manufacturing conductors at therate of approximately one conductor per month. We havecompleted 26 conductors, which corresponds to about 80% of
the Japanese share. In other ITER countries, they have juststarted conductor fabrication, although some amount of
Nb3Sn strands has been manufactured. Japan is leading theway as the first ITER country to provide TF conductors
according to the highest ISO standards.

Reference
Takahashi, Y. et al., Technology Development and Mass Production of Nb3Sn Conductors for ITER Toroidal Field Coils in Japan, Nuclear Fusion, vol.51,
no.11, 2011, p.113015-1-113015-11.

54 JAEA R&D Review 2012

Nuclear Fusion Research and Development

4-2

A Laser System for Plasma Diagnostics Has Achieved the World s Best Performance=Development of YAG Laser for the Edge Thomson Scattering System in ITER=

Newly developped laser
7.66 J, 100 Hz, 766 W

W

W

Japan(JT-60)
7.46J, 50Hz, 373W

ag
er
Av

2006
2003

e
r
we
po
=

1

10

m

2001 Russia
Japan(LHD) 2000
Korea(KSTAR)
UK(MAST),USA(NSTX)
China(HT-7)
Czech(COMPASS)
USA(Alcator C-MOD)
Germany(ASDEX-U) USA(DIII-D)

2

W

Performance ofcommercial laser

2002

4.

0

EU(JET)

10

Output energy (J)

60

0

0

5

W
0
80 W
0

40

20

10

Target for ITER
5 J, 100 Hz, 500 W

1.5 m

100
Repetition rate (Hz)Fig.4-6 Comparison of plasma diagnostic lasers in theworld
Letters in parenthesis indicate the name of fusion experimental
device.

Fig.4-7 Laser system in operation
Bright light of pumping flash lamps in laser amplifiers is visible.
Because this YAG laser emits an infrared ray, the laser beam isinvisible.

We are responsible for developing the edge Thomson
scattering system that will measure electron temperatureand electron density profiles in an edge region of ITER
plasmas. The Thomson scattering is a plasma diagnostic
technique that evaluates electron temperature and density byanalyzing laser light scattered by electrons when a highpowered pulsed laser light is injected into the plasma.
Photons scattered from electrons are extremely weak, and foreach hundred billion photons injected into the plasma, only
about one is scattered. To attain measurements with sufficientaccuracy, the pulsed laser light must be fired with highenergy. Moreover, to determine the plasma state, which varies
from moment to moment, the laser pulses must be fired at ahigh repetition rate. In this diagnostic system, the Nd: YAG
laser needs to provide 500 W of average output power; thatis, 5 J/pulse of energy needs to be fired at 100 Hz. Meetingthese design requirements for both high output energy andhigh repetition rate had been the biggest issues.
We have developed a laser system that provides 373 W(7.46 J, 50 Hz) of average output power; so far, this system
has been shown the best performance than any other systemin the world (Fig.4-6). However, to develop the laser system

required by ITER, the repetition rate must be doubled.
Moreover, the heat created in the laser amplifier must be
controlled, typically by controlling the input energy to thelaser amplifier. However, decreasing the input power also
decreases the gain in amplification; this has posed a serious
problem in that energy in an amplifier cannot be efficiently
extracted as laser light. However, by using samarium-doped
special glass tubes in the laser amplifier, the noise of thelight, which was interfering with the improvement in
amplification gain, is selectively absorbed. Therefore, even at
half-input energy, the amplifier gain has been approximately
doubled, compared to conventional performance. As a result,it becomes possible to efficiently extract energy from theamplifier, leading to an average output power of 766 W (7.66 J,
100 Hz). This corresponds to twice the conventional average
output power (Fig.4-6, Fig.4-7). Thus, we have succeeded indeveloping a laser system that exceeds the target performancein ITER, and electron temperatures and densities can bemeasured with high accuracy.
A part of our research was accomplished by the support of
the Institute of Laser Engineering, Osaka University.

Reference
Hatae, T. et al., Development of a YAG Laser System for the Edge Thomson Scattering System in ITER, Review of Scientific Instruments, vol.83, issue 10,2012, p.10E344-1-10E344-3.
JAEA R&D Review 201255

Nuclear Fusion Research and Development

4-3

Attainment of Large-Scale 3D Multibeam Analysis=Compensation for Beam Deflections in the Accelerator of ITER NBI=

X (m)Grid (Yellow)

Aperture to extract beam

Metal ridge attachedat the back side ofthe extractor

One aperture group0.16 my(c)

200 kVZ

Y

5×16
apertures(Ǿ14 mm)

Extractor
Grid support
structure(Blue)
Region for
3D beamanalysis0

Grid support

1.6 m
Direction ofbeam acceleration

1 MV

0.

(d)

Grid support

Negative ion source
X
m
64

x

Negative ion source

0.4 m

0.8

Direc
beam tion ofacce
leratio
n

Grid
0.5

0.32 mZ (m)(a)

Beam trajectories(80 beams  4 aperture area
㧩 320 beams)(b)

Fig.4-8 A cross section of the 1 MeV accelerator for the
ITER NBI
One grid is composed of 4 × 4 aperture groups having 5 × 16
apertures each. Beams are extracted from 1280 apertures intotal and accelerated up to 1 MeV by the stack of the fiveacceleration grids.

Fig.4-9 Deflection of 320 beams in 1/4 accelerator model
View on right-hand side shows superposition of apertures at
the accelerator exit and intensity distribution of beams at 2.5 m
downstream from the accelerator exit. Beams in insets (a) and(b) are deflected in the directions of an arrow.

In the accelerator of the ITER neutral beam injector (NBI)
shown in Fig.4 - 8, a deuterium negative ion (D − ) beam of
1 MeV, 40 A is produced from a large extraction area of 0.64
× 1.6 m2 with 1280 apertures. However, the beam produced
from each aperture is deflected owing to repulsion between
beams having the same electric charge and because of electricfield distortion formed by a grid support structure. The
deflected beam directly hits grids and beamline components.
Then excessive heat loads limit the pulse length. To
compensate for these beam deflections, we have carried out athree-dimensional (3D) beam analysis.
The beam trajectory is much influenced by electric fielddistortions at low energy. Thus, a fine mesh is necessary to
precisely calculate the electric field in the extractor. Because
this requires a huge number of meshes and a large amount ofcomputer memory, the number of beams in the calculation
has been limited to five. To analyze many beams
simultaneously, we developed the following technique: a twodimensional beam analysis was used to obtain ion positions
and energies at the exit of the first grid. These positions andenergies were then used as the set of input data for the 3D
beam analysis. This technique allows us to use larger meshes,
thereby reducing the number of meshes required for

calculations. As a result, the 3D beam analysis applied to the
MeV accelerator in JAEA with 15 beams showed good
agreement with experimental measurements.
The analysis was applied to a 1/4-accelerator model having
320 beams for ITER, as shown in Fig.4-9. This is the firstattempt in the world to perform such a large-scale 3D beam
analysis. The right-hand side of Fig.4-9 shows the intensity
distribution of the beam produced from the aperture groupnear the grid support structure. The beams from theperipheral aperture lines (insets (a) and (b)) were deflected
outward owing to repulsion between the beams. To
compensate for these deflections, a metal ridge was attachedaround the apertures at the backside of the extractor. This
metal ridge distorted the electric field, steering the deflectedbeams inward. For the beam near the grid support structure(insets (c) and (d)), the grid support structure, instead of themetal ridge, was used to compensate for beam deflections.
Thus, this analysis allowed us to identify and compensate
for beam deflections in advance of experiments. The
compensation method has been tested in the MeV accelerator
and applied to the ITER design.
The present study was sponsored by the ITER organization
as an ITER R&D task, No.C53TD48FJ.

Reference
Kashiwagi, M. et al., Beam Optics in a MeV-Class Multi-Aperture Multi-Grid Accelerator for the ITER Neutral Beam Injector, Review of Scientific
Instruments, vol.83, issue 2, 2012, p.02B119-1-02B119-3.

56 JAEA R&D Review 2012

Nuclear Fusion Research and Development

4-4

Overview of IFERC Project=Activities of DEMO Design and R&D Coordination Centre, and Computational Simulation Centre=
Sintering temperature㧦1273 K

2.5
TPC

2

PROCESSIntensity (cps)

3000

Be
Ti
2000

Be12Ti
Be17Ti2
10 mm

Be2Ti

1000

1.5

0
10

20

40

60

80

100

2ǰ (deg)1

Fig.4-11 Appearance and result of X-ray diffraction
analysis of a plasma-sintered specimen
A small amount of unreacted beryllium remains.
Radiation loss rate

Inductive operation
period (hours)

Self-driven
current ratio

Normalized pressure

Confinement factor

Density limit ratio

Plasma density
20-3(10 m )

Plasma current(10 MA)

0

Fusion power
(GW)

0.5

Fig.4-10 Comparison of tests by JA and EU
system codes (JA: TPC, EU: PROCESS)
Good agreement obtained except for the
inductive operation period. (Radiation loss rateis radiative loss power normalized by 100 MW.)The International Fusion Energy Research Centre(IFERC), one of the Broader Approach activities, supports
three subprojects: (1) the DEMO Design and R&D
Coordination Centre, (2) the Computational Simulation
Centre (CSC), and (3) the ITER Remote Experimentation
Centre. These activities are a joint collaboration between
Japan and the EU extended over the 10 years starting from2007. The IFERC is located in Rokkasho, Aomori prefectureand has the mission of contributing to ITER and an early
realization of fusion energy. With the recent operations in the
DEMO R&D Building and the supercomputer, the IFERC
project shifted in 2012 from a preparatory phase to an actual
research phase.
An integrated project team, consisting of the EU, JA hometeams, and the project team, has been established in the
DEMO Design activity in IFERC. Joint work has been
performed on divertor, in-vessel components, operation
scenarios, and system codes. Benchmarks of system codes
show good agreement, except for the inductive operation
period, as shown in Fig.4-10. The physics model underlying
this deviation is under investigation.
As for DEMO R&D activities, R&D for the DEMO
blanket is progressing steadily in five tasks: SiCf/SiC
composites (T1), tritium technology (T2), materials
engineering (reduce-activation ferritic/martensitic steel) (T3),advanced neutron multiplier material (T4), and advanced

Time 0.51 ms
Fig.4-12 TAE mode driven
by alpha particles in ITER
steady state scenario
Magnetic fluctuations in the
TAE mode driven by alphaparticles do not seriously
degrade the confinement of
the alpha particles.

tritium breeder materials (T5). These tasks are based on thecommon interests of the EU and Japan in DEMO blanket
development. One of the most important recent results is thefeasibility of using a newly developed plasma sintering
method to synthesize a beryllium intermetallic compound(Be12Ti) for use as an advanced neutron multiplier material, as
shown in Fig.4-11. Investigations to determine the optimum
synthesis conditions continue, including efforts to reduce theamounts of unreacted beryllium and other undesirable
compounds. Furthermore, the DEMO R&D Building has
been completed recently as a radioisotope (RI) handlingfacility. There, researchers from various fusion-blanket fields
can participate in collaborative research. This facility will
stimulate significant synergetic interactions among researchers
engaged in R&D activities for the DEMO blanket.
In the CSC, the supercomputer with 1.52 PFlops (1.237
PFlops in the Linpack test) has been in operation since last
January. The Light House Project was performed through the
middle of March to explore the frontier of magnetic fusion
confinement. An example of the results is shown in Fig.4-12,
which has been obtained by a simulation of alpha particle
transport in the ITER under a steady state scenario. It isfound out that instabilities driven by alpha particles do notseriously degrade alpha particle confinement.
With these new facilities in operation, new research results
are expected to increase.

Reference
Nishitani, T. et al., Recent Progress in Blanket Materials Development in the Broader Approach Activities, Journal of Nuclear Materials, vol.417, issues
1-3, 2011, p.1331-1335.
JAEA R&D Review 201257

Nuclear Fusion Research and Development

4-5

Achieving Safety in the Deuteron Accelerator Facility=Development of a Safety Protection System for the IFMIF/EVEDA Accelerator=

High-speed transmission
for interlock signal

IFMIF/EVEDA Accelerator Building
Beam-Inhibit Module

PPS Operation Console

MPS Equipment
Beamshut off

PPS PLC（a）MPS Equipment
Controller

PPS PLC（b）

Accelerator
Vault

IFMIF/EVEDA Accelerator

MPS Unit

Radiation Monitor

MPS Equipment

MPS Equipment

Emergency Stop Button/ Monitorring Camera

Hige-speed transmission
for interlock signal

Equipments managed by PPS
Door monitored & controlled by PPS
Door monitored by PPS

Emergency stop button

Monitoring camera (Warning light㧦radiation-hardened)

Radiation monitor

Fig.4-13 Configuration of the safety protection system for the IFMIF/EVEDA accelerator(1) Personnel protection system (PPS) for safety of workers: Highly reliable protection can be achieved by component redundancy(a and b) and hard-wired connections with monitored devices. The PPS is linked with the radiation monitoring system to prevent
exposure owing to radiation leakage.
(2) Machine protection system (MPS) to secure equipment: Beam-induced thermal damage can be avoided by transmitting interlock
signals from equipment to beam inhibit-modules at a high speed.

In the Engineering Validation and Engineering Design
Activities (EVEDA) for the International Fusion Material
Irradiation Facility (IFMIF), a prototype accelerator(IFMIF/EVEDA Accelerator) will continuously accelerate
deuteron ions to 9 MeV/125 mA. Thus, radiation shielding
and reducing activation of equipment are crucial issues forsafe operation. To achieve a safety in the IFMIF/EVEDA
Accelerator facility, we are developing the two independent
systems shown in Fig.4-13: (1) a Personnel Protection
System (PPS) and (2) a Machine Protection System (MPS).
The system concept and design of PPS are consideredprimarily to realize using commercially available equipment
to protect workers. Nevertheless, an integrated system has
been established that employs more conservative, mechanicalequipment with high reliability to impose safety in important
areas; examples include personal keys to control entering andleaving, door interlocks, duplicated signal paths, and use of a
Programmable Logic Controller (PLC) as a signal processor.
Moreover, we developed a system for responding flexibly andquickly to the various operational demands for thisaccelerator. The approach, based on our experience, uses an
in-house design and fabrication of the PLC s logic sequence.
In addition, we established a radiation safety management
method through a central control system by connecting the
PPS to a radiation monitoring system whose performance isevaluated by a shielding analysis using experimental datafrom 9 MeV deuterons.
MPS is required to exclude malfunctions due to signal
noise and to inhibit a beam by transmitting only those signalsthat show abnormal states of accelerator equipment. Signals
are sent at high speeds to beam-inhibit modules. For thispurpose, we configured the MPS to use a high-speed signal
processor with a FPGA (Field-Programmable Gate Array)and MOS-type IC. This MPS satisfactorily tolerates signal
noise and achieves a high-speed signal transfer of 5 µ s, whichis half the required performance. Therefore, we haveconfidence that the MPS can protect equipment from beam
damage.

Reference
Takahashi, H. et al., Evaluation of Gamma-ray and Neutron Energy for Area Monitoring System in the IFMIF/EVEDA Accelerator Building, Fusion
Engineering and Design, vol.87, issues 7-8, 2012, p.1235-1238.

58 JAEA R&D Review 2012

Nuclear Fusion Research and Development

4-6

Progress in Satellite Tokamak Programme Project=Construction Activities of JT-60SA Tokamak Progressing toward Assembly=

2.84 m

11.95 mFig.4-14 Bird's eye view of cryostat and cryostat base
The left image shows the cryostat vessel body and the cryostat base. The right image shows the dimensional confirmation test for the
120°sector of the lower structure of the cryostat base.
TF CoilIn-cryostat Feeder

Spool
CTB

CTB

CS

BenderSand Bluster

EF4

Conductor
HTS Current Lead

CTB

Turn Table

Fig.4-15 Bird s eye view of feeder system design
The structure of the feeder system connecting the coil terminalbox (CTB), the high temperature superconducting current lead
(HTS-CL), and the superconducting coils (TF coil, CS, and EF4).

Fig.4-16 Central solenoid manufacturingWinding test for the central solenoid using a copper dummy
conductor.

For the Satellite Tokamak Programme, jointly implemented
by Europe and Japan, construction of the JT-60SA
progressed well; this includes designing and manufacturingof components. Newly concluded were the Procurement
Arrangements (PAs) for the cryostat base assembly and theremote handling equipment for welding and cutting, which
are to be contributed by Japan. PAs were also concluded forthe superconducting toroidal field (TF) coil test, which is tobe contributed by Europe. A total of 18 PAs were concluded
by February 2012 between the two Implementing Agencies(IAs), the Japan Atomic Energy Agency (JAEA) for Japanand Fusion for Energy for Europe.
For the cryostat base to be contributed by Europe, weldingof the lower structure sectors (three 120°sectors) wascompleted. After passing dimensional confirmation tests(Fig.4-14), highly precise machining, which finishes surfaces
within ±0.5 mm, was started. As for the TF coil, the contract
for coil manufacturing was awarded in 2011 following thecontract of the NbTi strand in 2010. Furthermore, the contractof the quench protection circuits for the superconducting
coils was also awarded and its detailed design was completed.
For the vacuum vessel (VV) to be contributed by Japan,
the third 40°sector was completed at the JAEA Naka site.
Regarding the poloidal field coils, the equilibrium field coilNo.4 (EF4), which is to be installed in the lowest part of thetokamak, was delivered to the Naka site. A feeder system wasdesigned to connect the coil terminal (CTB), the hightemperature superconducting current lead (HTS-CL), and thesuperconducting coils (TF coil, CS, and EF4) (Fig.4-15). As
for the CS with the Nb3Sn conductors, its winding test wasstarted using copper dummy conductors (Fig.4-16).
In addition, the disassembly of JT-60 at the Naka site isprogressing well and is on schedule. All these activities aretaking us toward the assembly of JT-60SA, which will bestarted in January 2013.

Reference
Yoshida, K. et al., The Manufacturing of the Superconducting Magnet System for the JT-60SA, IEEE Transactions on Applied Superconductivity, vol.22,
issue 3, 2012, p.4200304-1-4200304-4.
JAEA R&D Review 201259

Nuclear Fusion Research and Development

4-7

Accurate Fabrication of Large Superconducting Coil=Fabricating the Superconducting Coil for JT-60SA=
12 m

Toroidal field (TF) coil

Central solenoid (CS)EF3

Joint partEF2
Fig.4-18 Double pancake coil for EF4

EF1 Two-layer coil consists of a conductorthat is 434 m in length.

EF6
EF4

EF5

Equilibrium field (EF) coilFig.4-17 Superconducting coil system in JT-60SA
PF coils (CS and EF coils), which are developed by
JAEA, are set on the TF coils.

Fig.4-19 Winding pack for EF4 coil
Winding pack was fabricated by stacking 10 DP coils,
then wrapping it with ground insulation.

Confinement of high-density, high-temperature plasmas forlong durations is an important issue that must be resolved
before the Tokamak fusion reactor is realized. For the longtimes needed for plasma control, the magnet system in JT-60SA
is to be made of superconducting materials. A poloidal field(PF) coil system, which consists of a central solenoid (CS)and plasma equilibrium field (EF) coils, has been developed
by JAEA (Fig.4-17). These coils are fabricated by stacking
several pancake coils (Fig.4-18), and then connecting the
conductor joints of the stacked pancake coils. The EF4 coil,the first superconducting coil in JT-60SA, consists of 10
double pancake (DP) coils.
To sustain high-performance plasmas, the configurationand position of the plasmas must be effectively controlled;
this requires that the PF coil system must be manufactured
with high precision. In particular, the EF4 coil forms thedivertor configuration that needs to be sensitively controlled,
and so errors in manufacturing the EF4 must be small comparedto those for other coils. Simulation of plasma control showsthat the required manufacturing error of circularity (in-plane
ellipticity) in EF4 should be less than 6 mm. Measured errorsof circularity for 10 DP coils, which makeup the EF4 coil,were no more than 3.8 mm. Circularity errors tend to be large
around joint parts. Thus, joint parts were not gathered in a
toroidal section; instead, they were toroidally distributed byeach DP coil. By adopting this structure, circularity errors forthe winding pack were averaged after stacking the DP coils.
Therefore, for the EF4 winding pack, the circularity error was
reduced to 0.6 mm (Fig.4-19). This is an order of magnitude
smaller than the design requirements; this should reduce thestrength of the error field, which limits precise control of theplasma.

Reference
Tsuchiya, K. et al., Manufacture of the Winding Pack and Development of Key Parts for the JT-60SA Poloidal Field Coils, IEEE Transactions on Applied
Superconductivity, vol.22, issue 3, 2012, p.4202304-1-420234-4.

60 JAEA R&D Review 2012

Nuclear Fusion Research and Development

4-8

Manufacturing the Doughnut-Type Vessel for JT-60SA

6.6 m=High-Precision Structure of Double-Wall Vacuum Vessel=

Actual Size
(Dimension +/- tolerance)

Rib
Reinforce
10 m
Inner Wall

One-turn
ResistanceOuter Wall
One-turn
Resistance

Closure Weld
Inner Wall
(18 mm)Inside double wall

Outer Wall
(18 mm)Rib Reinforce(22 mm)Inner Wall Height6244 mm(6245.0 +6/-4)
Inner Wall Width
3124 mm(3119.9 +5/-5)

Fig.4-20 Vacuum vessel and double wall structure
Manufacturing the doughnut-shaped vacuum vessel from adouble-wall welded structure. The vessel has a 10-m diameter
divided in 40°units.

Fig.4-21 Product vacuum vessel
The vacuum vessel was manufactured so that its major
dimensions were within the few millimeters of tolerance
required by the design.

Upgrading the JT-60 to the JT-60SA tokamak machinewith fully superconducting coils is the Japan-EU satellite
tokamak program; this project is under both the Broader
Approach program and Japanese domestic program. Japan is
to manufacture a doughnut-shaped vacuum vessel (VV) with aheight of 6.6 m and a diameter of 10 m, as shown in Fig.4-20.
The VV is a double-wall welded structure with 18-mm-thickplates. Openings for plasma heating devices and diagnostic
devices complicate the VV structure. Segments of the vessel
are manufactured in a factory in 40°units; these units arethen assembled on-site into the complete 360°structure. Thiswelded vessel must endure enormous electromagnetic forces;
however, the high precision required by the design means thatthe structure must undergo minimal distortion during welding.
The JT-60SA VV must have the following features: (a) ahigh vacuum for the core plasma, (b) a high one-turnresistance (16 µ Ω) for plasma breakdown, (c) must be able
to endure electromagnetic forces of up to 7.5 MN verticallyand 2.5 MN horizontally, and (d) must be shielded against
neutron penetration. Austenitic stainless steel 316L wasselected as the structural material; each of the inner and outerwalls have a thickness of 18 mm and the two walls are
separated by 22 mm of rib reinforcement. Shielding water
circulates between the two walls. The total weight of thevessel, including the water, is 400 tons. The vessel issupported by nine legs. The operational temperature is 323 K;
however, a baking operation at 473 K is used to improve thequality of the vacuum. The baking is done by circulating high-temperature nitrogen gas.
In manufacturing, the double wall must be precisely weldjointed into the 10-m diameter doughnut-shaped VV. To meet
design tolerances, distortion must be minimized during
welding, which in turn requires that heat input must be
minimized. Welding inspections are hampered because it is
impossible to watch, from inside the double wall, welding
penetration at closures. Some structural members were
designed with a partial penetration weld-joint, which is
focused on the rib reinforcements. Partial penetration
welding, which requires less heat input and therefore causes
less distortion, was adopted for rib welds at the inner andouter walls; full penetration welds were not needed for the
structure to endure large electromagnetic forces. A limit of heat
input was determined before manufacturing and a 20°upperhalf mock-up was fabricated to test the manufacturing
procedure. Feasibility of manufacturing and integrity of thestructure were validated. Major dimensions on the productwere within the few millimeters of allowed tolerances; thus,
high precision was achieved in manufacturing the productvessel (Fig.4-21).
Manufacturing of the VV started in late 2009. The first
40°unit (two sectors) was completed in May 2011, as shownin Fig.4-21, and three 40°units were completed by the end of
March 2012. Another three sectors will be completed in fiscal
year 2012. Manufacturing of all sector units will be finishedin fiscal year 2013.

Reference
Masaki, K., Shibama, Y. K. et al., Design and Manufacturing of JT-60SA Vacuum Vessel, Fusion Engineering and Design, vol.86, issues 9-11, 2011, p.18721876.
JAEA R&D Review 201261

Nuclear Fusion Research and Development

4-9

Progress in Stability Theory for Rotating Plasmas=New Matching Theory for Resistive Wall Modes=

8

Numerical solution
Generalized
matching theory

0.8

0.6

Matching region

0.2

0rot. shear < mag. shear

4

unstable
2

0.5

1.0

Normalized plasma minor radius

destabilize
stabilize

0

-2

0

ǩ＝0
ǩ>0

6

resonance0.4

ǩ<0rot. shear > mag. shear

Normalized RWM growth rate

Normalized plasma displacement

1.0

stable

-4
1.03

1.04

1.05

Wall location(wall radius/plasma radius)Fig.4-22 Plasma displacement by generalized matchingtheoryUsing a matching region with finite width, the resonant
structure of RWM is captured. In matching theory, physicalquantities controlling RWM stability can be identified bychanging the physical model in the matching region.

Fig.4-23 RWM stability analysis by analytic dispersion
relation
Dependence of RWM growth rates on the wall location isshown by the analytic dispersion relation via generalized
matching theory. Here, α denotes a physical parameter that
depends on profiles of rotation and magnetic field.

The JT-60SA device that is being constructed in the Naka
Fusion Institute aims to realize advanced (high-beta, steady)plasmas to contribute to ITER and DEMO research. In
advanced plasmas, one of the most important issues is thatthe achievable plasma beta value is limited by the occurrenceof instability called resistive wall modes (RWMs). It is
known that the use of plasma rotation is an effective way tostabilize the RWMs. However, the stabilizing mechanism andthe most essential physical quantities (magnetic field,
temperature, density, etc.) have not yet been clarified.
In stability analysis, there is a method called matchingtheory that can extract the most essential physical quantity
controlling stability. In matching theory, the asymptotic
matching method has been established. In this method, theregion governed by the most essential physics (physically it
appears as a resonance and the region is called the resonantsurface) is asymptotically connected to the region that isapart from the resonant surface. Without rotation, the locationof resonance is determined by background magnetic field,
thus asymptotic matching theory is valid because we know
the behavior of the solution around the resonant surface.
However, with rotation, the resonant surface splits due toDoppler shift. Furthermore, because the RWM rotates withfinite frequency and resonates, the location of resonanceshifts away furthermore. The locations of these resonantsurfaces are not known a priori, hence, the asymptotic
matching theory fails to capture the resonance. To overcome
this difficulty, the matching theory is generalized. The
generalized matching theory invokes a matching region withfinite width to include the resonant surface, because thelocation of resonance is limited to a finite range. With the aidof the matching region with finite width, we can solve thematching problem for rotating plasmas (Fig.4-22).
Generalization of matching theory enables us to derive ananalytic dispersion relation that describes the RWM stability(a relation among the RWM wave number, growth rate, andfrequency). By analyzing the dispersion relation, the mostinfluential physical quantity for RWM stability has beenclarified. This physical quantity is determined by profiles ofplasma rotation and magnetic field. By studying how thisphysical quantity affects the RWM stability, it is found thatwhen the rotational shear is smaller (larger) than themagnetic one, the rotation stabilizes (destabilizes) the RWM(See Fig.4-23, where the physical quantity is denoted by α).
This result is applicable for designing the profiles of rotationand magnetic field that are suitable for stabilizing the RWMs.

Reference
Shiraishi, J. et al., Analytic Dispersion Relation for Resistive Wall Modes in Rotating Plasmas by Generalized Matching Theory, Nuclear Fusion, vol.51,
no.5, 2011, p.053006-1-053006-9.

62 JAEA R&D Review 2012

Nuclear Fusion Research and Development

4-10 Lithium Isotope Separation for Fusion Reactor Fuel

Ionic liquid
PP13-TFSI

1.5

PP13-TFSI
1.4

N+

1.3

6
Li isotope separation[(CF3SO2) 2N]- coefficient㧩

Final 6Li/7Li ratio
Initial 6Li/7Li ratio

1.2
1.1
1.0

6

Ionic-liquid impregnatedorganic membrane Depleted 6Li
6Enriched Li

Li isotope separation coefficient=Development of Innovative Technology for Lithium-6 Enrichment Using an Ionic Liquid=

10

100

1000

10000

100000Li yield (mg/m2/h)

6

Li

+

6

Li

+
Electrodialysis

Li+

7

Anode

6

CathodeLi isotope separation efficiency

6

Solution without Li

Li solution ofnatural isotope ratio

100

PP13-TFSI80

N60

Calculation of
optimum
6
Li isotope separation
efficiency[(CF3SO2) 2N]-

1.05㨪1.15
6

Li isotope separation
efficiency㧩
20
6
Li yield×
6
0(Li
isotopeseparation
0.9
1.0
1.1
1.2
1.3
1.4
1.5
coefficient-1)
6Li isotope separation coefficient40

Fig.4-24 New lithium isotope separation technique using
an ionic liquid
An ionic liquid is impregnated into an organic membrane that
repels water and is used in sportswear. 6Li has a higher mobility
than 7Li and will be enriched on the cathode side.

Fig.4-25 6Li isotope separation coefficient and efficiency
The 6Li isotope separation coefficient of this method (1.05〜
1.15) is the same as or better than that of the amalgamation
process using mercury (1.06).

Tritium needed as a fuel for fusion reactors is produced via
neutron capture by lithium-6 (6Li). However, natural Li
contains only about 7.6% 6Li, and enrichment of 6Li up to
40〜90% is required for adequate tritium breeding in fusion
reactors.
The amalgamation process using mercury is the only 6Lienrichment technology in practical use overseas; however,
because mercury is toxic, this method cannot be industrializedin Japan. Other methods have very low separation efficiencies
and are unfit for mass production. Because it is difficult toimport 6Li from overseas, the establishment of a 6Li
enrichment technology that is unique to Japan is an issue of
top priority for the realization of fusion reactors.
Therefore, we have proposed a new and original processthat uses an ionic liquid with electrodialysis, thereby
establishing an innovative Li isotope separation technology.
The new approach demonstrates excellent environmental
resistance, is amenable to mass production, and is highly
energy efficient (Fig.4-24, Fig.4-25). Lithium ions move byelectrodialysis between the cathode and the anode in lithium
solutions. Because the ionic mobility of 6Li ions is higher

than that of 7Li ions, 6Li can be enriched on the cathode sideof a cell.
Using PP13-TFSI (C11H20OF6N2O4S2) as the ionic liquid,
we obtained a maximum of 1.4 for the 6Li isotope separation
coefficient (Fig.4-25, top). Furthermore, we found that it ispossible to obtain the separation efficiency for the 6Li isotopethat ranges from 1.05 to 1.15 (Fig.4-25, bottom). These
results show that the 6Li isotope separation coefficient of thismethod is the same as or better than that of the amalgamation
process using mercury (1.06).
Currently, Li recovery technology from seawater and usedbatteries is developing this technology, which is sponsored by
the Funding Program for Next Generation World-Leading
Researchers (NEXT Program) in the Cabinet Office,
Government of Japan. In addition, this technology has
become a focus of great attention by the electric vehicleindustry, which depends on Li ion batteries.
The present study was sponsored by the Ministry of
Education, Culture, Sports, Science and Technology of Japan(MEXT).

Reference
Hoshino, T. et al., Basic Technology for 6Li Enrichment using an Ionic-Liquid Impregnated Organic Membrane, Journal of Nuclear Materials, vol.417,
issues 1-3, 2011, p.696-699.
JAEA R&D Review 201263

Nuclear Fusion Research and Development

4-11 Toward Clarifying Complex Plasma Behavior=Development of an Integrated Simulation Code for Entire Tokamak Plasma=

18

without pellet
Perturbation
by pellet
Collapse by
instability

Pedestalregion00
2.50Normalized plasma radius 1

40

0
100

0
0.88

1 0.88

Normalized plasma radius

1

Normalized plasma radiusElectron temperature(eV)

Diffusivity Normalized(m2/s)
gradient

Plasma pressure(kPa)

Pedestal region

Ion temperature
(keV)

10

In front ofdivertor plate

0
-5

0

15

Distance along divertor plate (cm)Fig.4-26 Profile evolution when a pellet is injected during an
increase in pressure gradient in pedestal region (left) and when nopellet is injected (right)
An ablated pellet absorbs plasma energy, makes perturbation in→) and localized pressure gradientpressure profile (top,(middle), and triggers an instability that enhances diffusivity (bottom) andcollapses the profile (). Without a pellet, the pressure gradientincreases in the entire pedestal region and triggers an instability. Pelletinduced diffusivity is so localized that energy loss is reduced.

In fusion plasmas, many phenomena interact, resulting incomplex behavior that must be understood and controlled. To
do so, we divide the tokamak plasma into three regions: thecore, the pedestal and its peripheral divertor regions. Then,for each region, we developed physical models for plasmaheat/particle transport, instabilities, and so on. We then
combined the models to obtain an integrated model thatcovers the entire region from the core to the divertor; such acomplete model is required for self-consistent evaluation of
the entire plasma. By using this model, we clarified thevarious behaviors discussed below, thereby contributing to
the design of ITER, JT-60SA, and DEMO.
(1) Core model development and integration. The
radial transport of alpha particles, enhanced by instabilities
induced by alpha particles themselves, was modeled andincluded in a core integrated code. We evaluated the
reduction of fusion performance in ITER. We developed aplasma rotation model and clarified the mechanism ofrotation driven by alpha particles. Then, we evaluated theplasma rotation in DEMO. We developed a pinch model ofhigh-Z impurity and found large inward pinch when theplasma rotates in the direction opposite to that of the plasma

Fig.4-27 Temperature profile evolution incore/pedestal regions (top) and in front ofdivertor plate (bottom)
While transport is reduced in the pedestal region
and core temperature increases (), temperature
decreases once () and then increases () in
front of the divertor plate. Integrated code
covering entire plasma can clarify this type ofdynamics over the entire plasma.

current. Using this model, we succeeded in explaining
Tungsten accumulation in JT-60U experiments, which had
been an unresolved issue.
(2) Pedestal model integration. We developed apedestal integrated code by coupling the core integrated codewith an instability code, a model of solid pellet injection, anda simple divertor model. This model clarified the onset
mechanism of the instability induced by the pellet, and it
showed reduced energy loss by the pellet-induced instability(Fig.4-26). These results can serve as a guide to reduce theheat load on the divertor plate in ITER.
(3) Integration of entire plasma region. We coupled
the core integrated code with a divertor integrated code
composed of plasma fluid code and neutral/impurity Monte
Carlo code, instead of the above simple divertor model, anddeveloped an integrated code that can treat the entire plasma
region self-consistently. The code enabled us to clarify the
dynamics of the entire plasma (Fig.4-27) and to study
detailed operational scenarios that would yield highperformance in the core plasma compatible with low heatload on the divertor plate.

Reference
Hayashi, N. et al., Integrated Modeling of Whole Tokamak Plasma, Plasma and Fusion Research, vol.6, 2011, p.2403065-1-2403065-8.

64 JAEA R&D Review 2012

5    
Development of Quantum Beam Technology

Green Innovation

Life Innovation

Materials Science
Environment
and Energy

Advanced R&D using Quantum Beams

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

Advanced BeamTechnology
NeutronSynchrotron
X-ray

JRR-3

Research Reactors

ElectronProton

TIARA

γ-ray

Ion

PhotonJ-PARCRadioisotopes

SPring-8

High-Intensity Laser Facilities

Accelerators

LasersQuantum Beam Platform
Fig.5-1 JAEA quantum beam facilities

Quantum beam is a generic name for neutron beams, ion
beams, electron beams, high-intensity lasers, and synchrotron
X-rays, which are generated from accelerators, high-intensity
laser facilities, and research reactors. Recently, quantum
beam technology has been greatly developed, with the mostadvanced manufacturing and observations being made using
highly controlled quantum beams.
We own various quantum beam facilities (Quantum Beam
Platforms), such as JRR-3 and J-PARC (Tokai area), TIARA,
and electron beam and gamma-ray irradiation facilities(Takasaki area), J-KAREN and other lasers (Kizu area), andthe SPring-8 beamlines (Harima area). We are carrying out
R&D on advanced beam technology and promoting many
fundamental and applied research studies in various fields(materials science, environment and energy, medicine andbiotechnology) by utilizing the create and probe functions ofquantum beams. In addition, we are making efforts tocontribute to the recovery from the accident at the Tokyo
Electric Power Company, Incorporated Fukushima Daiichi
Nuclear Power Station (1F) by developing improved
decontamination materials, etc. (Topics 1-13, 1-14, 1-15 in
Chapter 1) (Fig.5-1).
Quantum beams have a probe function, by which we canobtain atomic- or molecular-level information throughobservation of the alterations in beam parameters. They also
allow us to process materials on a nanometer level (atomic ormolecular level) by interacting with the constituent atoms ofa material to change their configuration, composition, andelectronic state. In medical applications, they are used forradiotherapy in which a beam is focused on a cancer cell(Fig.5-2).

Observe

CreateObservation at
atomic resolution

Atomic level
fabrication andcreation

Cure

Dr. Q

Cure andtreatment
at cell level

Fig.5-2 Characteristics of quantum beams

In this chapter, we introduce our newest topics related toadvanced beam technology and its application in the fields ofmaterials science, environment and energy, and biotechnology.
JAEA R&D Review 201265

Quantum Beam Science Research

Large Increase in the Maximum Energy of a Laser-Driven Proton Beam

100
10 µm

10-2

Intensity (arb.)

(a)

10-4
10-6
10-8

1.0

Main
Pulse

36 fs

0.0
-200

0

(b)

200

Time (fs)
Prepulse Level(Noise Level)

10-10
10-12

(c)

Maximum Proton Energy (MeV)=40 MeV Proton Generation Using a Ti:sapphire/OPCPA Hybrid Laser System=

Intensity (arb.)

5-1

Prepulse

-400

-200

0

200

Time (ps)Fig.5-3 Focal spot and temporal profile(a) Typical spatial intensity distribution at the focal spot, (b) measured
third-order cross correlation in the 50 TW operation mode (the crosscorrelator temporal resolution of 120 fs causes up to a 3-foldunderestimation of the contrast on the ps and ns time scales), and (c)Transient-grating frequency-resolved optical gating (TG-FROG)
measurement of the main pulse temporal profile.

It is well known that when a laser beam is focused on ametal foil or a plastic foil target, hot electrons are produced
and ejected from the back of the foil target. When these hot
electrons leave the thin foil target, a strong electric field isgenerated owing to charge separation. Protons fromcontaminants, such as water or oil on the back side of the foil
target, are accelerated by this electric field. To apply thislaser-driven proton source to ion cancer therapy, etc., it isnecessary to produce high-energy protons by tightly focusing
the laser beam. A 70 MeV proton beam was reportedly
produced by a single-shot base large glass laser system. In
addition, a 25 MeV proton beam was produced using arepetitive Ti:sapphire laser system.
In this study, we used the J-KAREN laser system (JAEA
Kansai Advanced Relativistic Engineering Laser). To
efficiently focus the laser beam on a target, deformation of
the wavefront of the laser beam transmitted from the laser
system to the target was suppressed as much as possible. We
focused a laser beam with an energy of 7.5 J onto a target in
a 2 µm spot (Fig.5-3(a)). The laser intensity on the target
reached 1021 W/cm2. As far as we know, there have been noreports of the observation of any physical phenomena

200
100

CR-39
NuclearTrack Detector
Protons
10 mm〜40 MeV (d)10Lasers that are notsuitable for applications
Lasers that are suitable
for applications

1
0.1

1

10
100
Laser Energy (J)

(e)
1000

Fig.5-4 Results for this experiment and other studies(d) Example of tracks formed by protons. (e) Maximum
proton energy vs. the on-target laser pulse energy indifferent experiments. The star denotes the result for thepresent study.

corresponding to a 1021 W/cm2 interaction, although thegeneration of over 1022 W/cm2 has been achieved. Using an
optical phase control technology for the laser system and apulse-shape measurement technique, we produced a low
prepulse level of 10-10 (Fig.5-3(b)) and a pulse width of 36 fs(Fig.5-3(c)). If the prepulse level is very high, it affects thefoil target before the main pulse arrives. In this experiment,
because the contrast was approximately 10-10, the foil target
was not damaged before the main pulse arrived. As a result,
we could irradiate the foil in spite of high intensity. Thus, we
generated a 40 MeV laser-driven proton beam (Fig.5-4(d)(e)).
The energy of the generated proton beam was estimated fromthe range of the protons using a CR39 nuclear track detector.
Because of the deep stopping range for small animals, thisnew proton source can now be used in preclinicalexperimental studies of in vivo proton therapy. We believethat this laser-driven proton generation technique reduces thesize of the particle accelerator system required for cancer
therapy. It is also expected that this laser-driven proton source
can be applied not only to other medical applications but also
to other proton acceleration applications.

Reference
Ogura, K. et al., Proton Acceleration to 40 MeV using a High Intensity, High Contrast Optical Parametric Chirped-Pulse Amplification / Ti:Sapphire Hybrid
Laser System, Optics Letters, vol.37, no.14, 2012, p.2868-2870.

66 JAEA R&D Review 2012

Quantum Beam Science ResearchHigh-Power -ray Generation via the Radiation Reaction
Effect Using an Ultra-Intense Laser Pulse

5-2=Proposal for a New Laser-Driven -ray Source=
expanding plasma
foil

120

high-energy
electrons

90

60
30

150

laser pulse

180high-energy
electrons

10 30
10 25

210

γ-raysRadiation power (PW)

330
240

(/m3)

270

300

Fig.5-6 Angular distribution of the emitted γ-rays
The γ-rays are collimated toward the laser propagation
direction (The laser propagation direction corresponds to 0°).

Fig.5-5 Simulation results showing the electron density
distribution of a solid target irradiated by an intense laser
pulse in the unit of m-3
Expanding plasma, which is generated by a prepulse, exists infront of the foil. High-energy electrons escape from the target
front and back side, and γ-rays are emitted from the rear side.

20 PW, 15 fs
10 PW, 30 fs
5 PW, 60 fs

6

Fig.5-7 Dependence of the γ-ray power on the plasma scale
lengthThe three lines correspond to different laser power with a fixedlaser energy. There exists an optimum scale length for effective
γ-ray generation.

4
2
0

0

0

5

10

15

20

Expanding plasma scale length (µm)

The radiation reaction (RR) effect, which is a back reaction
acting on a charged particle that emits radiation due to its
acceleration, is important such as for electrons in synchrotron
accelerators or high energy charged particles in the universe.
For electrons irradiated by an intense laser pulse, the RR
effect has been neglected to date, but as laser intensities
increase beyond approximately 1023 W/cm2, it is expected toplay an important role. The power working on electrons by
the RR effect can be easily estimated for a single electron
performing a simple trajectory, but for plasmas, in which acollective behavior is important, it is not fully understood.
In this study, we used a newly developed particle
simulation code incorporating the RR effect, which enables
us to evaluate the amount of energy transferred from the laser
pulse to γ-rays through the motion of electrons via the RR
effect.
Fig.5-5 shows the electron density distribution of a thinlayer carbon target irradiated by a 10 PW, 30 fs laser pulse
from the left hand side. The maximum energy of theaccelerated electrons reaches 400 MeV, and some portion ofthem interact with the laser pulse and emit γ-rays. The pulse
duration of the γ-ray is as short as the laser pulse, and thetotal energy is 96 J which is 32% of the incident laser pulse,
resulting in aγ-ray power of 2.8 PW which is much higher thanthose generated by any other devices. The angular distributionof the emitted γ-rays is shown in Fig.5-6. The γ-rays arecollimated toward the laser propagation direction into two
lobes. The average γ-ray photon energy is 1.1 MeV, and thetotal photon number is estimated to be 5.5×1014.
The γ-ray characteristics depend on the laser and plasma
parameters, such as the laser intensity, duration, and scale
length of the expanding plasma. The γ-ray power dependence
on these parameters is summarized in Fig.5-7, where thethree lines correspond to different laser powers with a fixedlaser energy of 300 J. The γ-ray power increases with theincident laser power and has an optimum value of the plasma
scale length which is roughly the same as the laser pulse
length. We have also revealed the dependences of the total
photon number and the average photon energy on theseparameters, which could lead to the possibility of controllingthe γ-ray characteristics.
In conclusion, we have shown that γ-rays are generated byan intense laser pulse and solid target via the RR effect. This
γ-ray source has high power, which is several orders ofmagnitude higher than those generated by other devices. The
unique laser-driven γ-ray source could open up various new
applications, such as laboratory astrophysics, photonuclear
reactions, copious generation of electron-positron pairs.

Reference
Nakamura, T. et al., High Power γ-ray Flash Generation in Ultraintense Laser-Plasma Interactions, Physical Review Letters, vol.108, issue 19, 2012,
p.195001-1-195001-5.
JAEA R&D Review 201267

Quantum Beam Science Research

5-3

High-Precision Measurement of Nuclear Materials=Proposal for the Advanced HKED System=

10

5

10

4

10

3

10

2

10

1

10

0

Counts (/keV)

UKα2

5m

UKα1
UKβ1
UKβ3
UKβ2
PuKα1

90

UKβ4
PuKβ1

100
110
Energy (keV)

120

Fig.5-8 Schematic view of the advanced HKED system
The advanced HKED system is composed of an energy
recovery linac (ERL), a laser supercavity, and KED/XRF
detectors. Mono-energetic X-rays generated by collision ofhigh-energy electrons and laser quanta are used for irradiationof sample solutions. Transmitted and scattered X-rays aredetected by the KED and XRF detectors, respectively. The sizeof the ERL accelerator can be reduced using a 6-loop
configuration.

Fig.5-9 X-ray response spectrum
A Monte Carlo simulation code was used to investigate the X-ray
responses from a sample including U (200 g/) and Pu (2 g/).
The results for mono-energetic and bremsstrahlung incident
X-rays are shown with red and blue lines, respectively. A highpeak intensity of the Pu characteristic X-rays (PuKα1) can beobtained through efficient measurement of the PuKα1 line using
mono-energetic X-rays.

Nuclear material accountancy is an important issue fornuclear safeguards. Generally, destructive analysis with mass
spectrometry is used for density determination of nuclear
materials. However, non-destructive analysis is desired
because of simplicity and rapidity of the analysis, as well as
for the reduction of discharged nuclear wastes. Hybrid
K-edge densitometry (HKED) is a non-destructive analyticalmethod of nuclear materials dissolved in solution. This
technique can be applied for quantification of uranium (U)and plutonium (Pu) concentrations.
In the existing HKED system, bremsstrahlung X- raysfrom an X-ray tube are used as the incident photon source.
The broad energy distribution of the bremsstrahlung X-rays
produces background counts due to inelastic scattering that
affect the counting precision. However, if mono- energetic
X-rays are used, efficient KED and X-ray fluorescence(XRF) measurements are possible through reduction of the
background counts. Thus, we have proposed an advanced
HKED system with mono-energetic X-rays. Fig.5-8 shows aschematic view of the advanced HKED system.
To evaluate the performance of the advanced HKED

system, X-ray responses were investigated using a Monte
Carlo simulation code. The result is shown in Fig.5-9. The
peak intensity of the Pu characteristic X-rays (PuKα1) can beenhanced using mono-energetic X-rays. Consequently, both
the ratio of the peak to the background count and ratio of thepeak to the total count are improved in comparison with thoseobtained when using bremsstrahlung X-rays. The counting
statistics in the region of the U K-absorption edge in the KED
spectra are also increased using mono-energetic X-rays.
Analysis of low concentrations of neptunium (Np) isdifficult with the existing HKED system because of the highbackground counts from the Pu characteristic X-rays.
However, with the advanced HKED system, by adjusting theenergy of the incident X-rays between the K-absorption
edges of Np and Pu, it is possible to measure the Np
characteristic X-rays without any influence from the Pu
characteristic X-rays.
The present technique is expected to facilitate the nondestructive rapid, and precise density determination of
nuclear materials in solutions at nuclear power plants andnuclear reprocessing facilities.

Reference
Shizuma, T. et al., Proposal for an Advanced Hybrid K-edge/XRF Densitometry (HKED) using a Monochromatic Photon Beam from Laser Compton
Scattering, Nuclear Instruments and Methods in Physics Research A, vol.654, issue 1, 2011, p.597-603.

68 JAEA R&D Review 2012

Quantum Beam Science Research

5-4

Neutron Observation of Deformation Behavior of Materials at Cryogenic Temperature=Neutron Strain Measurement of a Superconducting Wire under Cryogenic Tensile Loading=

GM refrigerator

ctor
Dete

0.4

Detector
Vacuumchamber

Diffraction

Loading
driverSlit
Incident neutron

Specimen

θ0

ctio

n

Principle of neutron strain measurement
Tension

fra

on
utr
Ne

Te
ns
io
n

Vacuum chamber
Loading driver

Neutron window

Dif

GM refrigerator
Conductor cabled0

θ1(strain-free)

0.1

(200)
(020)

Tension0Compression

- 0.1
- 0.2

0

Ǎd

Fig.5-10 Cryogenic load frame for neutron diffraction
The cryogenic load frame consists of a Gifford-MacMahon (GM) refrigerator,
a vacuum chamber, and a loading driver. The change in the lattice strain of a
specimen placed inside the vacuum chamber can be measured by neutron
diffraction. The strain is determined by measuring the change in thediffraction angle, 2θ , associated with the change in the lattice spacing, Δd .

The neutron diffraction method can be commonly used todetermine the internal strain in materials at centimeter-order
depth. Furthermore, it can be used to investigate themicroscopic deformation behavior of materials under various
environmental conditions, such as different loads andtemperatures, which is difficult to accomplish when using
mechanical measurement methods with strain gauges.
Fig.5-10 shows the cryogenic load frame for a neutron
diffraction system developed in the neutron engineering
diffractometer RESA-1, which is located at the JRR-3
research reactor. The deformation behavior of materials at
cryogenic temperature can be measured via neutron diffraction
using the cryogenic load frame under tensile loading.
Fig.5-11 shows the results of the strain measurement of asurround Cu stabilized YBCO (YBa2Cu3O6+δ)-coated conductor
using the cryogenic load frame. The mechanical and criticalcurrent properties in a superconducting composite wire are
known to be affected by the local strains generated in the
superconducting phase. Therefore, it is important to know the
local strains at cryogenic temperatures corresponding to its
operating temperature. As shown in Fig.5-11, it was clarified
that the lattice strains of the 200 and 020 reflections in the
YBCO phase at 77 K were approximately zero at an applied

0.2

A ff

d0

Crystal structure

77 K

Stabilized layer
Protective layer
Superconducting layer‫(ޓޓޓޓޓ‬YBCO)Intermediate layer
Metal substrate

0.3
Lattice strain (%)

Diffraction

0.1 0.2 0.3 0.4 0.5
Applied strain (%)

0.6

Fig.5-11 Change in the lattice strains of asurround Cu stabilized YBCO-coated conductor
A surround Cu stabilized YBCO (YBa2Cu3O6+δ)-coatedconductor consists of multiple layers, including a
superconducting layer. This result shows the changes
in the lattice strains for the 200 and 020 reflections of
the YBCO phase measured at 77 K under tensileloading. The variable A ff in the figure is the force-freestrain.

strain from 0.19 to 0.21%, which is the force-free strain (Aff).
Compressive strains in the YBCO phase induced during the
manufacture of a YBCO-coated conductor can increase thematerial strength up to the corresponding force-free strain.
Furthermore, it was clarified that the conventional idea thatthe maximum critical current agrees with the Aff cannot beapplied to a YBCO-coated conductor, because its maximum
critical current appeared at 0.035% of the applied strain.
As described above, strain measurement using neutron
diffraction with a cryogenic load frame is a very useful toolfor developing practical superconducting composite wires,and it is also expected to be widely used in cryogenic
materials engineering research.
This research was partially supported by a Grant-in-Aid for
Scientific Research (B),Basic study of mechanoelectromagnetic properties and development of highperformance HTc superconducting composites and Strain
effect on critical current under high magnetic fields forpractical superconducting wires −a new development instress/strain problems, from the Ministry of Education,
Culture, Sports, Science and Technology of Japan (MEXT),
Grant Nos.19360289 and 21360312, respectively.

Reference
Osamura, K., Suzuki, H. et al., Force Free Strain Exerted on a YBCO Layer at 77 K in Surround Cu Stabilized YBCO Coated Conductors, Superconductor
Science and Technology, vol.23, no.4, 2010, p.045020-1-045020-7.
JAEA R&D Review 201269

Quantum Beam Science Research

5-5Discovery of Novel Rare-Earth Metal Hydrides under High Pressure=Formation of NaCl-Type Monohydrides by Pressure-Induced Phase Separation=(1) monohydride(2) dihydride(3) trihydride

Fig.5-12 Structures of three hydrides with different H concentrations and the fcc metal lattice
Yellow, light-blue, and dark-blue spheres indicate metal atoms, H atoms at O sites, and H atoms at T sites, respectively. (1) Monohydridewith only the O sites occupied; (2) dihydride with only the T sites occupied; and (3) trihydride with both O and T sites occupied.

Intensity (a.u.)

LaD2
XRD (SPring-8)
13.1 GPa

111 fct

200 fct
222 fct

202 fct
311 fct

400 fct
004 fct

220 fct

002 fct

113 fct
311 LaD

111 LaD

200 LaD

220 LaD

222 LaD

NPD (J-PARC)
13 GPa1.5

diamond(materials for anvils)

2.0

2.5

3.0

d spacing (Å)
Fig.5-13 SR XRD (upper) and NPD (lower) profiles after phase separation of LaD2 at 13 GPa
Intensities are on a logarithmic scale in arbitrary units. The reflection peaks of LaD are shown in red.

Rare-earth metals have a very high affinity for hydrogen(H) and form various H-concentration states, starting from asolid solution and ending with the trihydride RH3, through thedihydride RH2. Ideally, RH2 has a face-centered-cubic (fcc) metallattice in which the tetrahedral (T) sites are filled with H atoms
and the octahedral (O) sites remain empty (Fig.5-12(2)).
Further absorbed H atoms occupy the O sites, and RH3 iseventually formed (Fig.5-12(3)). However, rare-earth metalmonohydrides RH (Fig.5-12(1)), in which only the O sites
are occupied by H atoms, has not yet been reported.
Previously, through synchrotron radiation (SR) X-ray
diffraction (XRD) measurements at BL22XU, SPring-8, wefound that LaH2 decomposed into two different fcc phases atpressures above 11 GPa. The observed decomposition
suggests that H-poor and H-rich phases are spontaneously
formed by pressurization. To understand this phase separation,
information on the interstitial H atoms is necessary.
In this study, we attempted to determine the concentration
and atomic positions of deuterium (D) after phase separationof LaD2 by neutron powder diffraction (NPD) at highpressure. First, we developed a high-pressure setup with a
Paris-Edinburgh press and then installed it in the high-intensity total diffractometer NOVA at BL21 in the MLF,
J-PARC. With this setup, we succeeded in measuring NPD
patterns up to 17 GPa, the highest pressure for domestic NPD
experiments, at ambient temperature.
Fig.5-13 shows the XRD and NPD profiles obtained at~13 GPa. The Bragg peaks of the D-poor phase with an fcc
metal lattice are shown in red. One may notice that several
Bragg peaks indexed with odd numbers are missing in the
NPD pattern for the D-poor phase, while these peaks arepresent in the XRD pattern. The lack of odd-number-indexed
Bragg peaks suggests the formation of an NaCl-type structure
due to the comparable neutron scattering length of the La andD atoms. The NaCl-type hydride can be regarded as an fcc
monohydride, in which only the O sites are occupied by H(D)atoms. The present result indicates that only rare-earth metals
can form a series of stoichiometric hydrides, such as RH,
RH2, and RH3, with the fcc metal lattice.
This work has been partially supported by the New Energy
and Industrial Technology Development Organization (NEDO)
under Advanced Fundamental Research on Hydrogen Storage
Materials.

Reference
Machida, A. et al., Formation of NaCl-Type Monodeuteride LaD by the Disproportionation Reaction of LaD2, Physical Review Letters, vol.108, issue 20,
2012, p.205501-1-205501-5.

70 JAEA R&D Review 2012

Quantum Beam Science Research

5-6

Microscopic Magnetization Process in a Tb43Co57 Amorphous
Film Using Magnetic Compton Scattering=New Measurement Method for Spintronics Materials=
Magnetic recording material(a) Macroscopic magnetism

Tb-Co amorphous alloy

Magnetic moment (emu/cm3)

200

(c)
100

(b) Microscopic magnetism

Tb43Co57

Orbital

0

Spin
Total
Spin
Orbital

-100

-200

Tb
-2

-1
0
1
Magnetic field (T)Co

2

Fig.5-14 Magnetic hysteresis loops measurement(a) Illustration of the macroscopic magnetism of an amorphous Tb43Co57 film. (b) Illustration of the microscopic magnetism of Tb43Co57.
Magnetic moments of Tb and Co are oriented in anti-parallel to each other. Each magnetic moment consists of spin and orbital
magnetic moments. (c) Solid curve: Macroscopic magnetic hysteresis loop measured using SQUID. Dark green circles: spin-specificmagnetic hysteresis (SSMH) loop. Light green circles: orbital-specific magnetic hysteresis (OSMH) loop. The shape of the SSMH isdifferent from that of the OSMH.

Amorphous rare earth-transition metal alloys are candidate
materials for high-density recording media and spinelectronic devices. Traditionally, the materials for magnetic
devices are developed on the basis macroscopic magnetic
property measurements, e.g., using superconducting quantum
interference devices (SQUIDs) (Fig.5-14(a)). Because
macroscopic properties originate in microscopic properties,
such as the magnetic spin and orbital components (Fig.5-14(b)),
it is also valuable to investigate the microscopic magnetic
properties to design a high-performance magnetic recording
material.
While the Compton profile directly corresponds to the
double integral of the electron moment in materials, themagnetic Compton profile (MCP) reflects only the net spin
magnetic moment of a magnetically active electron. Recently,
we developed a method for measuring the spin-specificmagnetic hysteresis (SSMH) loop using the magnetic fielddependence of the MCP intensity.

The dark green circles in Fig.5-14 (c) show the SSMH
loop of a Tb43Co57 amorphous film measured at thesynchrotron radiation facility SPring-8. In addition, we
obtained the orbital selective magnetic hysteresis (OSMH)
loop for the first time, which is shown as light green circles in
Fig.5-14(c), by combining the SSMH with the results for themacroscopic magnetic hysteresis loop. The results show that
the SSMH, OSMH, and macroscopic hysteresis loops havedifferent line shapes. The spin component is softer than theorbital component under a magnetic field, indicating that the
speed of magnetic switching is different between the spin andorbital components. In addition, it was found that thecontribution to the SSMH from Co was different from thatfrom Tb.
This study demonstrates that it might be important to use
separately microscopic information, i.e. the magnetic
hysteresis of the spin and orbital components, to develop
magnetic devices such as spin electronic devices.

Reference
Agui, A. et al., Microscopic Magnetization Process in Tb 43 Co 57 Film by Magnetic Compton Scattering, Applied Physics Express 4, 2011,
p.083002-1-083002-3.
JAEA R&D Review 201271

Quantum Beam Science Research

5-7

Development of a Microfabrication Technique for
Utilization of Environmentally Friendly Plastics=Micro-/Nano-Fabrication of Crosslinked Poly(L-lactic acid) Using EB Nanoimprint Lithography=(a) Coating(b) Forming(e) PLLA imprinted by thermal-NIL& Imprinting
PLLA+TAIC

70 ℃, 2 MPa

(nm)
(nm)

(nm)

(nm)

(µm)
(µm)

(µm)

(µm)

PET film

Si mold(c) Crosslinking

Si mold

Deformed
Heating at 70 ℃
BeforeAfter
10 minAfter
1 hourAfter
24 hour(d) Releasing( f ) PLLA imprinted by EB-NIL (100 kGy)

EB

0

[nm]
(nm)

Si mold

Si mold

Fig.5-15 Steps of the electron beam (EB)nanoimprint lithography process
EB nanoimprint lithography was proposed for
fabricating micro-/nano-structures of crosslinkedpoly(L-lactic acid) (PLLA). PLLA and triallyl
isocyanurate solutions were dropped into asilicon mold, shaped by pressing, fabricated bycrosslinking using EB irradiation, and thenremoved from the silicon mold.

500
500
0

10

20

30

40

50

0

[nm]
(nm)

500
10 5000[µm]10(µm)
60 [µm]

20

30

40

50

0

[nm]
(nm)

500
10 5000[µm]1060 [µm]
(µm)

20

30

40

50

0

[nm]
(nm)

10

500
500
0

60 [µm]
(µm)

10

20

30

40

50

10
60 [µm](µm)

Retaining finestructure

Fig.5-16 Thermal durability of PLLA imprinted pattern using conventional
and EB nanoimprint lithography methods.
Imprinted micro-/nano-sized green polymer patterns on PLLAs wereobserved before and after heat treatment using atomic force microscopy. The
designed structure on PLLA using conventional thermal lithography (e) wascompletely deformed after 24 h at 70 ℃. The imprinted pattern of PLLA using
EB crosslinking (f) was retained owing to the high resistance to thermal
deformation.

Research and development on the preservation of the
global environment is progressing in various fields. It is
desired to replace petroleum-based plastics with plant-basedalternatives with the goal of establishing a recycling-oriented
society. Poly(L-lactic acid) (PLLA), prepared from lactide
produced in the fermentation of sugar or corn starch, is themost promising candidate for replacement of petrochemicalplastics because of its high strength, transparency, andmelting point. However, its low thermal deformation
temperature is a critical issue for industrial applications.
Therefore, we improved the heat resistance of PLLA using aradiation crosslinking technique with triallyl isocyanurate(TAIC) as the crosslinking agent and applied the crosslinking
technology to fabricate fine structures on PLLA.
PLLA fine structures are expected to be used for fine
optical components and lab chip materials, etc. Thermal
nanoimprint lithography (T-NIL) is the conventional and costefficient method for the mass production of nanostructures inpolymers. However, the obtained PLLA fine structures were
deformed by heating at 70 ℃, which is the softening
temperature of PLLA. Electron beam nanoimprint lithography(EB-NIL) was proposed as a novel method for fabricating
micro-/nano-structures and improving the heat resistance of
PLLA via crosslinking. As shown in Fig.5-15, the EB-NIL
method consists of four steps: (a) coating of a 3%

PLLA/TAIC (95/5) solution in chloroform by dropping into asilicon (Si) mold, (b) forming and imprinting the sample via
hot-pressing at 2 MPa and 70 ℃ for 20 min, (c) crosslinking
using EB irradiation with doses from 10 kGy to 500 kGy atroom temperature under vacuum, and (d) releasing the PLLA
fine structure from the Si mold.
Fine structures were imprinted on PLLA using T-NIL and
EB-NIL and observed via atomic force microscopy, as shownin Fig.5-16. The micro-/nano-sized fine structures of both
imprinted patterns were maintained at room temperature.
However, the fine structure produced on the PLLA using
T-NIL was completely deformed after heating at 70 ℃. On theother hand, the patterns imprinted on PLLA using EB-NIL(100 kGy) were retained, even after heat treatment at 70 ℃
for 24 h, due to its high resistance to thermal deformation.
The nanoimprinted, environmentally friendly plastics are
expected to be used for fine optical components and lab-on-achip materials, etc.
Our research was accomplished by the joint research of
Waseda University and Osaka University. A part of thisproject was supported by the Nanotechnology Network Project(Handai Multi-functional Nanofoundry, Osaka University) of
the Ministry of Education, Culture, Sports, Science andTechnology of Japan (MEXT).

Reference
Okubo, S., Nagasawa, N. et al., Micro-/Nanofabrication of Cross-Linked Poly(L-Lactic Acid) using Electron Beam Nanoimprint Lithography, Applied
Physics Express, vol.5, issue 2, 2012, p.027303-1-027303-3.

72 JAEA R&D Review 2012

Quantum Beam Science Research

5-8

Contribution to the Elucidation of the Mechanism of
O2 Generation from Water Using Ce(IV)=Discovery of Dinuclear Complexes of Tetravalent Cerium in an Aqueous Solution=
(B)

(C)

Peak Intensity (a.u.)(A)

Ce (IV) aq.
soln.

Dimer1

Dimer1
Bridged structure by
OH group

Dimer2

Dimer2
Bridged structure by
two OH groups

Dimer3

Trimer0

1

2
3
4
R +ΔR /Å

5

6

Fig.5-17 Radial structural functions of synchrotron
based Ce-K XAFS for aqueous Ce(IV) solution
Radial structural function (RSF) of the Ce-K EXAFS for+
Ce4 in an aqueous solution (black line) and simulated
RSFs for di- and tri-nuclear complexes optimized using
the DFT theory (colored line, each color corresponds to abond distance in each column).

Dimer3
Bridged structure byoxoacid

Fig.5-18 Chemical structure of dinuclear Ce(IV) complexes
determined by XAFS and molecular orbital calculation
Di- and tri-nuclear complex structures optimized using molecularorbital calculations based on the DFT theory (yellow: Ce(IV), blue:
oxygen, white: hydrogen). Dimer 3 is most probable.

Tetravalent cerium (Ce(IV)) is one of the widely known
inorganic oxidizing agents, particularly in the field of organic
synthesis. More recently, Ce(IV) has attracted considerable
attention as an electron sacrificial oxidant in studies of thephotocatalytic oxidation of water to produce H2 and O2 forenergy sources; thus, a continuous demand for Ce(IV) is still
anticipated in various basic and applied research fields. As apowerful chemical reagent, Ce(IV) is often employed inaqueous solutions, and thus, Ce(IV) interacts with the target
substances to be oxidized or activated as aquo species.
However, the nature of the Ce(IV) aquo species is far lesswell understood. Therefore, the detailed characterization of
the Ce(IV) aquo species is indispensable for elucidating theappropriate reaction schemes in which Ce(IV) species play animportant role and for discussing the origin of thefunctionality of Ce(IV) in applied systems. Thus, we carriedout the present EXAFS and DFT calculation study to identify
the speciation and complex structure of the Ce(IV) aquo
species in an aqueous perchloric acid (HClO4) solution.
The EXAFS Fourier transform (FT) of the Ce(IV) aquo
species (black data in Fig.5-17) exhibits two well-definedpeaks at approximatley R +Δ= 1.8 and 3.7. The first peak at
R +Δ= 1.8  corresponds to the coordination of watermolecules and possibly hydroxide ions. Meanwhile, thesecond peak at R +Δ= 3.7  is interpreted as scattering fromthe Ce(IV) atoms.
To probe the probable structural arrangement of polymericCe(IV) species, geometry optimization based on the DFT
theory was carried out for di- and trinuclear Ce(IV)complexes in the aqueous phase, as shown in Fig.5-18. The
geometry optimization was converged for three dinuclear and
one trinuclear complexes, as shown in the inset illustrationsin Fig.5-17. Based on a detailed data analysis of the EXAFS
results and comparison of the data with the DFT results, it
was concluded that the single oxo-bridging dinuclear
complex (Dimer3 in Fig.5-18) was the most dominant formof Ce(IV) species in the HClO4 solution. The dinuclear
complexes shown in Fig.5-17 are energetically close to one
another, implying that all of these species are potentiallyinvolved in the dimerization process of Ce(IV), although thefinal species is likely to settle into the single oxo-bridging
structure. The bridging-oxo groups of the dinuclear complex
are chemically active. Therefore, this species would also be
effective as an oxygen source, in addition to its significant
role as an oxidant.

Reference
Ikeda-Ohno, A., Yaita, T. et al., Dinuclear Complexes of Tetravalent Cerium in an Aqueous Perchloric Acid Solution, Dalton Transactions, vol.41, issue 24,
2012, p.7190-7192.
JAEA R&D Review 201273

Quantum Beam Science Research

5-9Developing a Technique for Operating a Protein Assembly=Identification of the Key Residue Controlling Protein Assembly via X-ray Crystallography=(a)

The E134A mutant HaNDK tetramer
The wild-type HaNDK dimers superposed on
the E134A HaNDK tetramer
Electrostatic repulsion between oxygen atoms

Oxygen atom
E137

(b)
E134

Dimer / dimer interface

E134

E137

Oxygen atom

Fig.5-19 Three - dimensional structures HaNDKs(a) wild-type HaNDK dimer, and (b) E134A mutant
HaNDK tetramer. Orange circles indicate thelocation of the 134th residue responsible fortetrameric association.

In some proteins, the formation of a higher-orderedassembly is essential for enabling their biological function.
Such an assembly formation is known to be effective for
improving the stability and function of proteins. To develop atechnique for operating protein assembly, we attempted tochange the assembly state of nucleoside diphosphate kinase(NDK).
NDK exists in the cells of various organisms, and while
NDKs from most organisms assemble into a tetramer or ahexamer, NDK from the halophilic bacteria Halomonas
sp.593 (HaNDK) exceptionally forms a dimer. To identify thereason for the lack of dimer-dimer association of wild-type
HaNDK, its three-dimensional structure was determined via
X-ray crystallography (Fig.5-19(a)). It was found that, when
the structure of the wild-type HaNDK was superposed on the
structure of a known tetrameric NDK, the 134th and 137th
glutamic acids (E134, E137) directly contacted the E134 andE137 positions in the neighboring HaNDK molecule (Fig.5-20).
The fact that the overlay provides at least four negative
charges at the interface of the two dimer units suggests thatthe electrostatic repulsion prevents the dimer-dimer

Fig.5-20 The tetramer is the interface newly created by two
dimer units of the E134A mutation
The wild-type HaNDK molecules (pink and blue sticks) aresuperposed on the E134A mutant HaNDK tetramer (black stick).
Bold sticks show the E134 and E137 glutamic acids of the wild-type
HaNDK.

association.
Given these results, we expect that the HaNDK dimer will
be associated into a higher-order assembly form if the E134
glutamic acid is replaced with alanine (E134A), which has an
uncharged short side chain. In fact, the mutation of E134 toalanine (E134A mutant HaNDK) resulted in conversion of
the assembly state of HaNDK from a dimer to a tetramer, as
shown in (Fig.5-19(b)). Moreover, the specific activity of the
HaNDK tetramer was 19% higher than that of the HaNDK
dimer.
This study demonstrated that the assembly state of aprotein can be controlled with a minimum mutation if
detailed structural information is available. The investigationof the structural differences in the dimeric and tetramericassemblies will provide useful information for creating
higher-ordered assemblies, which will contribute to thecreation of proteins with increased enzymatic activity andstability.
Our research was accomplished with the support ofKagoshima University.

Reference
Arai, S. et al., A Structural Mechanism for Dimeric to Tetrameric Oligomer Conversion in Halomonas sp. Nucleoside Diphosphate Kinase, Protein Science,
vol.21, issue 4, 2012, p.498-510.

74 JAEA R&D Review 2012

Quantum Beam Science Research

5-10 Phase Retrieval from Sparse Diffraction Data

=Coherent X-ray Diffraction Imaging of Single Particles=
(a)

(b)

(c)

(d)

X-ray

Single Particle(molecule)

Diffraction PatternFig.5-21 Schematic picture of single
particle diffraction imaging
A single particle is irradiated with an intense,
coherent x-ray and the diffraction pattern isrecorded by a detector.

Fig.5-22 Electron density of a lysozyme protein and thediffraction image(a) Electron density. (b) Diffraction image. (c) Reconstructed image
using the conventional method. (d) Reconstructed image using the
new method.

Conventionally, x-ray crystallography has been the principal
tool for determining the high-resolution 3D structures ofproteins, nucleic acids, and their complexes. In this method,
crystallization is a critical process by which a sufficiently
large crystal must be prepared. However, it is known thatapproximately half of proteins encoded in the humangenome, particularly membrane proteins, do not crystallize.
X-ray free-electron lasers (XFELs), a fourth-generation lightsource, are extremely powerful new x-ray sources and canpotentially provide a novel means for determining the threedimensional (3D) structure of biomolecules from thediffraction images of single molecules (Fig.5-21), because
coherent x-ray diffraction imaging (CXDI) using XFELs
does not require a crystal. However, to realize single
molecule imaging, some difficulties have to be resolved, suchas the phase problem: the phase information is lost in theexperimentally obtained diffraction data (Fig.5-22(b)). The
phase information is needed to reconstruct the image in real
space. Under certain conditions, this phase problem has been
computationally solved. However, strong diffraction and priorknowledge of the size and shape of the molecule are required,
but are usually unknown. This condition would be difficult tomeet for single molecule CXDI, because the target of the x -rays
is a single molecule, and the number of photons received byeach pixel on the detector is very limited (Fig.5-22(b)), evenwith XFELs. Therefore, the reconstructed image with theconventional method significantly depends on the initialcondition of the calculation.
Thus, we developed a new method for retrieving the phasesof sparse diffraction images. The method is based on the
maximum a posteriori (MAP) estimation of the Bayesianstatistics and is very robust against quantum noise. Comparedto the widely used conventional method, our method
succeeded in reconstructing the images from diffraction datathat were 10~20 times more sparse (Fig.5-22(c)(d)).
This method is a step closer to the realization of CXDI
from single particles. The realization of CXDI will strongly
contribute to the elucidation of the molecular mechanism ofproteins and to the development of new drugs.
The present study was partly sponsored by the x-ray FreeElectron Laser Utilization Research Project of the Ministryof Education, Culture, Sports, Science and Technology ofJapan (MEXT).

Reference
Ikeda, S., Kono, H., Phase Retrieval from Single Biomolecule Diffraction Pattern, Optics Express, vol.20, issue 4, 2012, p.3375-3387.
JAEA R&D Review 201275

Quantum Beam Science Research

5-11 Drastically Altering the DNA of Plants=Ion Beams near the Range End Frequently Induce Large Deletions=

Carbon ions thatpenetrate seeds

Carbon ions near
the range end

Arabidopsis seeds

Frequency of plants with mutations

2.7％

Equivalent level

㧩

2.7％

Frequency of large deletions

0.11㧑 Six times higher 0.66㧑
(4% of total)㧨 (24% of total)
Fig.5-23 Detection of mutations by utilizing trichomes
Arabidopsis has many trichomes (hairs) on the leaf surface(arrows). Mutation of the GL1 gene, which is indispensable fortrichome development, can be visually detected, because leaf
tissue with a mutated GL1 gene is glabrous (area encircled
with the dotted line).

Accelerated ion beams have been used as useful mutagens
for plant breeding, because they induce mutations different
from those induced by γ-rays or chemical mutagens. For
plant breeding, ion beams that penetrate seeds or cultured
tissue are typically used. While the ion beams deposit most of
their energy close to the end of their range, the types ofmutations that are induced by ion beams near their range end
have not been studied. Thus, we established an experimental
method that enables efficient detection of mutations in plants
and examined the characteristics of the mutations induced byion beams near the range end, with an emphasis on large
deletions.
Many trichomes (hairs) can be seen on Arabidopsis
leaves. The GLABRA1 (GL1) gene is indispensable fortrichome development. Arabidopsis plants that have amutation of the GL1 gene in either one of a pair ofchromosomes generate glabrous leaf tissue when the otherFig.5-24 Carbon ions near the range end frequently
induce large deletions
Two types of carbon ion beams with different energies
were used to irradiate Arabidopsis seeds to compare
their mutagenic effects. While the overall frequencies ofthe mutations were equivalent, the carbon ions near therange end induced large deletions at a frequency six
times higher than that of the second type.

GL1 gene is mutated (Fig.5-23). This change enables visual
detection of the induced mutations and analysis of thecharacteristics of the mutations induced in the GL1 geneusing DNA isolated from the glabrous leaf tissue.
Arabidopsis seeds were irradiated with two types ofcarbon ion beams: carbon ions that penetrate seeds, andcarbon ions near the range end. The frequency of plants with
mutated leaf tissue was equivalent for both types of carbon
ions at doses that had the same effect on survival reduction.
However, the ratio of large deletions among all of themutations was six times greater for the carbon ions near therange end (Fig.5-24). This result suggests for the first timethat ion beams near the range end frequently induce large
deletions. Thus, ion beams near the range end are expected tobe effective for drastic alteration of DNA to create useful
mutations for plant breeding.

Reference
Hase, Y. et al., Mutagenic Effects of Carbon Ions near the Range End in Plants, Mutation Research/Fundamental and Molecular Mechanisms of
Mutagenesis, vol.731, issues 1-2, 2012, p.41-47.

76 JAEA R&D Review 2012

6  
Evaluation of the Safety of Various Nuclear Facilities

Fuel Manufacturing Plant

Fuel
Cycle

NSRR
Nuclear Power Plant

RFEF

Reprocessing
Plant

Fuel Cycle Facility

࡮Fuel safety
࡮Safety evaluation of࡮Thermohydraulics safety ‫ޓ‬fuel cycle facilities
࡮Evaluation of material aging
࡮Degradation mechanism of‫ޓ‬fuel and material under‫ޓ‬neutron irradiation(4) Assessment ofenvironmental impact
and radiation exposure(3) SA assessment

࡮Latest knowledge

࡮Reactor system safety assessment

Radiation exposure
to public
Waste Disposal Facility
࡮Safety evaluation of‫ޓ‬radioactive waste‫ޓ‬disposal

Prevention of
nuclear disasters

Accident termination

Radioactive material release
Phase II AM(SA mitigation)

LSTF

JMTR

JMTR-HL

Core melt,
Pressureboundary failure࡮Analysis method࡮RPV, PCV integrity
࡮Damaged fuel behavior
࡮Radioactive material release

Phase I AM
(SA prevention)(2)SA prevention
࡮Thermohydraulic safety‫(ޓ‬Experiments by‫ޓޓ‬LSTF+New facility)
࡮Fuel safety

THYNC

SA: Severe Accident
AM: Accident Management

The accident
at 1F

Progress
to SA(1)
Investiga
tion ofdamagedreactorBeyond Design Base
Events(Multi-, External)
Initiating↑
Design Base Events Events

NUCEF

Fig.6-1 Nuclear safety research areas and related JAEA
facilities
The Nuclear Safety Research Center experimentally obtains
data using various JAEA facilities. These data for risk
assessment and safety evaluation of nuclear facilities
contribute internationally to the development of evaluation
methods and criteria.

Fig.6-2 Future direction of nuclear safety research
In addition to safety assessments of various nuclear facilities(mainly within design base events), we will conduct researchwith a priority on severe accident prevention efforts andnuclear emergency preparedness, 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) in order to ensure the safeuse of current 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 institute appropriate emergency response
procedures based on research findings and developments.
Most specialists at the NSRC regret that they were unable toprevent the accident.
Continuous efforts should be made to avoid serious
damage while we utilize nuclear energy. Because humantechnologies are not perfect, a proper understanding of thesituation and continuous improvement of developed
technologies are important for making the best and safest useof the technologies. The International Atomic Energy
Agency, which investigated the accident at the 1F, pointed out
the importance of continuous improvement. The different
phenomena that could threaten nuclear safety have beenstudied at the NSRC using the various facilities shown in
Fig.6-1. Radionuclides are handled and accident conditions
are safely simulated at these special facilities.
In this chapter, the results of recent safety research on thefollowing topics are presented: the fracture limit of fuelcladding and the integrity of reactor pressure vessels (Topics
6-1, 6-2) for reactor safety; confinement in the case of fire at
reprocessing facilities (Topic 6-3) for nuclear fuel cycle
facility safety; and corrosion of high-level waste storage
tanks (Topic 6-4) for radioactive waste disposal safety.
We believe that continuous research on these technicalsubjects is indispensable for promoting the continuous
improvement of nuclear technologies, including those forsafety regulation and electric-power generation. As shown in
Fig.6-2, to reduce the risk of nuclear facilities and enhance
the preparedness for emergency conditions that are postulated
to lead to a severe accident, we are planning to conduct
nuclear safety research with a priority on advanced evaluation
methods for accidents and external events so that severe
accidents with significant impact occur even less frequently.
The following topics related to items (2)〜(4) indicated in
Fig.6-2 have been described in Chapter 1:(2) prediction of reactor core damage avoidance in the 1F2
Unit (Topic 1-20)(3) data acquisition for the analysis of core melt progression(Topic 1-18)(4) effect of protective measures taken after the accident(Topic 1-16)

JAEA R&D Review 201277

Nuclear Safety Research

6-1

Accurate Evaluation of the Fuel Failure Limit

J-integral㧖 at cladding crack (kN/m)=Improved Application of NSRR Test Data to Power Reactor Conditions=
Crack generation at the cladding
surface in the early RIA stage

Cladding temperature (℃)

Displacement
and heat
Pellet thermal
expansion

Hydride rim Oxide layer

500
400

Wide power pulse

300
200

Crack tip

100

20

Calculations for wide power pulse cases.
'X' denotes the anticipated failure limit.

Failure10
Failure

0

NSRR
test

Room temperature

0

0 NSRR power pulse
0
0.5
1.0
Relative radial position in cladding (-)

㧖

NSRR
test

indicating a level‫ޓ‬of mechanical load

200

erion
e critfailur ated fromestim ata
test dHigh temperature

400

600

Temperature at the radial positionof incipient crack tip, T tip (℃)

Fig.6-3 Radial cladding temperature profiles
Radial cladding temperature profiles at the same displacement
are shown for two different power pulse widths. With a wide
pulse, the crack tip temperature and consequently the local
ductility increase. Hence, further crack growth is suppressed.

Fig.6-4 Estimation of the fuel failure limit considering thepower pulse width and temperature conditions
Evolution of the mechanical state is shown as a function of thecrack tip temperature. The anticipated failure limit for a wide
pulse is obtained as the intersection of the evolution curve withthe linearly estimated failure criterion based on the NSRR test
data.

Nuclear reactors must be safe, even if an accident occurs.
One of the accidents assumed for safety design confirmationis the reactivity initiated accident (RIA), which is a power
excursion typically caused by control rod ejection. To
determine whether fuel rods fail in RIAs, fuel failure criteria
based on experimental data from the Nuclear Safety Research
Reactor (NSRR) are used in Japan. Pulse power operation,
which produces high power for a short time, enables RIA
simulation tests, but the pulse power has a shorter time widthand a higher power peak than those anticipated in powerreactors. In addition, RIAs in power reactors are assumed tooccur at a high temperature (ca . 280 ℃), but most of the
NSRR test data were obtained at room temperature (20 ℃).
Despite these differences in the test conditions, the NSRR
data have been directly applied to the safety criteria, because
the differences lead to conservative results. However,quantification of the conservativeness and failure limit inpower reactors is necessary for evaluating the accurate safetymargin of the present safety criteria.
Therefore, we developed a procedure for converting thefuel failure limits obtained from the NSRR tests to thoseunder power reactor conditions by utilizing the accumulated
knowledge on fuel failure and recently obtained data at 280 ℃(Topic 6-1, JAEA R&D Review 2010).

On the cladding tube of long-term used fuel rods, the outersurface is oxidized and a layer with dense hydride
precipitation, so-called hydride rim, is formed. Our previous
studies clarified that cladding cracks are generated from thesurface into the hydride rim owing to pellet thermal
expansion in an RIA, and the crack begins growing further
when the mechanical state at the crack tip exceeds a certain
limit and finally penetrates.
In this study, we calculated the cladding deformation andtemperature during an RIA (Fig.6-3) and determined theevolution of the mechanical state at the crack tip. In addition,
we estimated the failure criterion from the NSRR test data forroom and high temperatures. Finally, we evaluated the fuelfailure limit for the wide power pulse (Fig.6-4). This
procedure reflects the mechanism in which the pulse width
affects the crack tip temperature, local ductility, crack growth
conditions, and finally the fuel failure limit and enables
conversion of the NSRR test data for power reactors andimprovement of the safety criteria with an accurately
evaluated safety margin.
This study was performed using data from experiments
sponsored by the Secretariat of the Nuclear Regulation
Authority (former the Nuclear and Industrial Safety Agency,
the Ministry of Economy, Trade and Industry of Japan).

Reference
Sugiyama, T. et al., Influence of Coolant Temperature and Power Pulse Width on Fuel Failure Limit under Reactivity-Initiated Accident Conditions,
Proceedings of 2011 Water Reactor Fuel Performance Meeting, Chengdu, China, 2011, T3-022, 6p., in CD-ROM.

78 JAEA R&D Review 2012

Nuclear Safety Research

Assessing the Failure Resistance of Reactor Pressure Vessels=Structural Integrity of Reactor Pressure Vessels Based on Probabilistic
Fracture Mechanics Considering the Effect of Residual Stress=

Emergency core cooling water

Emergency core cooling water

Reactor Pressure Vessel

with considering the residual stress

Temperature in RPVStress in RPV

Weld-overlay
cladding

Base
metal

Conditional failure probability

6-2

without considering the residual stress
-3

1×10

1×10-4
1×10-5
1×10-6
1×10-7
1×10-8
1×10-90

5

10Fluence (1019 n/cm2, E>1 MeV)

Fig.6-5 Pressurized thermal shock event in an RPVHigh tensile stress in the inside wall of an RPV is generated
owing to rapid cooling caused by the injection of emergency
core cooling water in response to a loss of coolant accident.

Fig.6-6 Comparison of conditional failure probabilities
with and without weld residual stressConditional failure probabilities as a function of the fast neutron
fluence with and without residual stress. The conditional failure
probabilities with residual stress are higher than those without
residual stress.

For the safe operation of nuclear power plants, thestructural integrity of reactor pressure vessels (RPVs),
regardless of the situation, has to be maintained. RPVs
composed of low alloy steel with a thickness ofapproximately 200 mm are exposed to neutron irradiationfrom the reactor core. Therefore, it is necessary to ensure thestructural integrity of RPVs while considering neutron
irradiation embrittlement associated with decrease in thefracture toughness and the occurrence of reactor transientsthat subject the vessel to severe thermal shock as well as thenormal pressure loading conditions. Pressurized thermal
shock (PTS) is one of the most severe events that increases
the possibility of failure of the RPV in a pressurized water
reactor (Fig.6-5). PTS, which is characterized by rapid
cooling of the inner surface of the RPV due to injection of
emergency core cooling water that causes high tensile stressin the inner surface, has the most significant impact on the
integrity of the RPV. PTS analysis is performed as a series ofsequential steps with a postulated crack at the inner surface inorder to identify the margins against failure.
There are two analytical approaches for PTS events:
deterministic and probabilistic fracture mechanics (PFM)

methods. Safety factors and/or margins are used for thedeterministic approach in order to conservatively assess the
structural integrity of an RPV. For the PFM approach, theuncertainties in the material properties of the structuralcomponents, the distribution of the flaw sizes, among otherfactors are incorporated into the analysis. The PFM analysis
code PASCAL developed by the JAEA can evaluate theconditional probabilities of crack initiation and failure in an
RPV under PTS events. We have updated PASCAL to version
3 (PASCAL3) in order to reflect the latest findings, such as
the residual stress effect due to welding. Fig.6-6 shows theconditional failure probabilities of an RPV as a function of
the fast neutron fluence with and without the residual stresscaused by the weld-overlay cladding. The results show thatconsideration of the weld residual stress increases theconditional failure probability. We continue to work to
improve the accuracy of structural integrity assessments byapplying advanced analytical methods and the PFM approach.
This study was sponsored by The Secretariat of the
Nuclear Regulation Authority (former Nuclear and Industrial
Safety Agency, the Ministry of Economy, Trade and Industryof Japan).

Reference
Masaki, K. et al., User s Manual and Analysis Methodology of Probabilistic Fracture Mechanics Analysis Code PASCAL3 for Reactor Pressure Vessel(Contract Research), JAEA-Data/Code 2010-033, 2011, 350p. (in Japanese).
JAEA R&D Review 201279

Nuclear Safety Research

6-3

Evaluation of the Confinement Safety of Radioactive
Materials during a Fire=Acquisition of Combustion Data and Evaluation of the Clogging of the HEPA Filter=

Combustion Gas Analyzer
Release behavior of soot and radioactive
materials, Particle size of sootHalogen radiation heater(Simulating external radiation heat)
HEPA filter

ΔP of HEPA filter (Pa)

10000

8000
PC

6000

eco-material
cable sheath
4000
PMMA
2000

0
Variation of ΔP of HEPA filter

Weight Measuring Device
Evaluation of combustion rate

Fig.6-7 Outline of the Fire Source -term Evaluation Experimental
Apparatus
Experimental data, such as the variation in the mass of the
combustion materials, the released mass of the simulated
radioactive materials and soot, and the rise in ⊿P of the HEPA filter
due to clogging by soot, can be measured simultaneously using thisapparatus.

During a fire in a fuel cycle facility, soot and radioactive
materials, which are produced from combustible materials,
will be released to the outside of the facility after filtration
using a high-efficiency particulate air (HEPA) filter.
Therefore, increase in the differential pressure (⊿P) andbreakage of the filter, which would result in the deteriorationof the confinement capability, may be caused by clogging of
the filter with soot. To quantitatively estimate the time until
breakage for the HEPA filter, the Fire Source-term Evaluation
Experimental Apparatus (FSEEA) (Fig.6-7) was designed andfabricated. Combustion tests using glove-box panel materials(polymethylmethacrylate (PMMA) and polycarbonate) andcable sheath materials (eco-materials and non-halogen flame
retardant cables) were performed using the FSEEA. In thetest, combustion property data, such as the combustion rateand soot generation ratio (ratio of released soot mass to
combustion mass of the material), and clogging property data
for the HEPA filter were measured under various experimental
conditions, including variation of the flow rate and externalradiation heat supplied to the materials.
From the test results, it was found that the properties wereconsiderably different depending on the type of combustible
materials. For example, PMMA had the largest combustion
rate but the smallest soot generation ratio and caused thenon-halogen flame
retardant cable sheath
0

1

2

3

4

5

6

Time (h)Fig.6-8 Estimation of the progress of ⊿P increase using
the test data
The change in ⊿P was considerably different depending on
the type of combustible materials. This estimation cancontribute to the quantitative evaluation of confinement safety
during a fire.

fastest rise in ⊿P per clogging mass of soot, because theparticle size of the soot was the smallest. Moreover, thecombustion of the non-halogen flame retardant cable sheathcaused a gradual rise in ⊿P with a relatively small clogging
mass of soot, but a rapid rise with increasing clogging mass
when there was a relatively large mass of clogging soot.
The change in ⊿P was estimated by applying the test data,
and the results are shown in Fig.6-8. The estimation wascompleted assuming that the combustion area for eachmaterial was a circle with a constant diameter of 50 cm andusing test data that was subjected to an external radiant heatof 40 kW/m2, except for PMMA, for which non-radiant heatcondition data was used. Although PMMA has the largest
combustion rate, ⊿P rose most gradually after 1.5 h. On theother hand, ⊿P for combustion of the non-halogen flame
retardant cable sheath rose rapidly with increasing mass ofthe clogging soot on the HEPA filter.
By applying the experimental data, we were then able toquantitatively estimate the time until breakage of the HEPA
filter.
The reference report includes the experimental results
which were carried out under a contract with Japan Nuclear
Energy Safety Organization, in fiscal year 2009.

Reference
Abe, H. et al., Experiment on Evaluation of Confinement Capability of Fuel Cycle Facility under Combustion of Glove-box and Cable Sheath Materials(Contract Research), JAEA-Reserch 2011-015, 2011, 27p. (in Japanese).

80 JAEA R&D Review 2012

Nuclear Safety Research

6-4

Reactions for Determining Chemical Environments
in Radioactive Waste Repositories=Corrosion Reaction of the Steel Overpack for High-Level Radioactive Waste=(a)1.35 m(b)

300 m or deeper

2.2 m

(c)
Vitrified waste
Steel overpack
Bentonite buffer

Host rock
Natural barrier

Engineered barrierFig.6-9 Steel overpack for high-level radioactive waste ina geological disposal system
Vitrified high-level radioactive waste will be encapsulated in asteel overpack surrounded by an engineered bentonite buffermaterial and placed in a deep geological repository. The steel
overpack is expected to retard the contact of the vitrified wastewith groundwater.

Fig.6-10 Iron specimens sealed in glass ampules andtheir surfaces before and after corrosion(a) Iron specimens were sealed and corroded in glass ampules
with aqueous solutions. Gaseous corrosion products, such ashydrogen sulfide, methane, and hydrogen, were analyzed. (b)Glossy surfaces were (c）covered with rust, such as magnetite,after corrosion.

In a deep geological repository for high-level radioactive
waste, as shown in Fig.6-9, the steel overpack will retard thecontact of the vitrified waste with groundwater for a certain
period but will be continuously corroded and thus will affectthe chemical environment in the bentonite buffer.
Corrosion reactions of iron under anoxic deep undergroundconditions are written as follows:(Eq.1)
Fe + 2H2O → Fe(OH)2 + 2H+ + 2eand(Eq.2)
3Fe + 4H2O → Fe3O4 + 8H+ + 8eThese reactions are oxidization reactions for iron, whichshould be accompanied by some reducing reactions. Sulfate
and hydrogen carbonate anions are naturally occurring
candidates as reductants. The reduction of hydrogencarbonate anions yields methane, as shown in Eq. 3, andkeeps the groundwater redox potential near -350 mV versus anormal hydrogen electrode.
(Eq.3)
HCO3- + 9H+ + 8e- → CH4 + 3H2O
The reduction of sulfate anions keeps the redox potential
near -300 mV with a similar reaction.
Understanding the extent of the redox potential of thegroundwater is essential, because high-level radioactive waste
contains elements such as Np and Se, whose mobility deep
underground is sensitive to the redox potential. Knowing

whether the hydrogen carbonate anions are reduced tomethane is also important, because elements such as U, Np,
Pu, and Am become mobile in the presence of hydrogencarbonate anions.
Therefore, we performed corrosion experiments with iron
specimens over 160 days using water containing sulfate andhydrogen carbonate anions at 60 ℃ in glass ampules without
contamination of atmospheric oxygen (Fig.6-10) andanalyzed the corrosion products. Sulfate and hydrogencarbonate anions were not reduced; however, the reduction ofwater, as shown in Eq. 4, was observed.
(Eq.4)
2H2O + 2e- → H2 + 2OHThis reaction occurs at extremely low redox potentialsof - 500 mV.
The corrosion of iron thus generated a strong reducing
environment and kept the hydrogen carbonate anions
unchanged in the bentonite buffer. These results revealed an
important process that can affect the safety of radioactive
waste disposal.
This study was sponsored by The Secretariat of the
Nuclear Regulation Authority (former Nuclear and Industrial
Safety Agency, the Ministry of Economy, Trade and Industryof Japan).

Reference
Otsuka, I., Yamaguchi, T. et al., Identification of the Cathode Reaction Accompanied with Overpack Corrosion, Scientific Basis for Nuclear Waste
Management XXXV (Materials Research Society Symposium Proceedings Volume 1475), 2012, p.507-512.
JAEA R&D Review 201281

7   
For the Evolution of Nuclear Science
Exploration of scientific

New Frontier
Nuclear Science andEngineering Directorate

Advanced Science
Research
Advanced materials
research
Condensed matter theory,
simulation, spintronics

Frontier research on
heavy element systems
Thermospintronics

simulation

Superheavy elements and nuclei
actinides
Neutron, accelerators
radiation fields

Radiation field effects
Hadron physics,
biology and radiation,
spin-polarized positron beam

REIMEI program

Quantum Beam
Science Directorate

External and foreignorganizations

J-PARC

Fig.7-1 Exploration of new scientific frontiers through various collaborations

As the sole institute for atomic energy research in Japan,
both basic and goal-oriented studies are within the scope ofour mission. Nuclear engineering is supported by many fields
of basic science and technology, and innovation in eachresearch field can contribute to the evolution of nuclear
engineering.
The purpose of the Advanced Science Research Center(ASRC) is to promote cutting-edge research on selected basic
science topics, which will enable the exploration of new
frontiers in nuclear science.
For the midterm starting at April 2010, ASRC Visions
are to promote the most advanced basic research in order toestablish an international center of excellence (COE) andcreate new research and technologies. The research projects
being pursued can be categorized into three fields: basic
research on advanced materials, frontier research on heavyelement systems, and basic research on radiation fields.
Specific topics in the first research field include thedevelopment of molecular spin-transport materials andcondensed matter theory. Frontier research in 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 for exploring the frontiers depicted in
Fig.7-1. The research highlights for 2011 are the detection of
NRM signals from 239Pu (topic 7-1), the discovery of an
innovative mechanism for spin-flow creation (topic 7-2), and
82 JAEA R&D Review 2012

the fabrication of a high-quality graphene film (topic 7-3).
These highlights are described in detail in the following
pages. Some ASRC researchers are also engaged in studies
related to the accident at the Tokyo Electric Power Company,
Incorporated Fukushima Daiichi Nuclear Power Station.
Results on the visualization of the fallout radioactive Cs inplants are presented in topic 1-17.
Collaborations with other institutes and universities are
essential for ASRC activities, particularly those through the
REIMEI research program for promotion of international
collaborations. A new ferromagnetic semiconductor Li(Zn, Mn)As(Fig.7-2) has been developed through collaborative research
efforts with the Chinese Academy of Sciences, Columbia
University, and the University of Tokyo.

LiAs

Zn (Mn)Fig.7-2 Crystal structure of the newly developed
semiconductor Li(Zn, Mn)AsAdvanced Science ResearchFirst Observation of 239Pu Nuclear Magnetic Resonance

7-1=Elucidation of the Electronic State in Pu Compounds=(a) PuO2

Pu

Spin-Echo Intensity (a.u.)O(b) PuO 2-x
Large ShiftPu
O

Oxygen Defect

5.2

5.4

5.6

5.8

6.0

6.2

6.4

6.6

6.8

7.0

Magnetic Field (T)Fig.7-3 239Pu NMR signals observed for the first time in the world
239
Pu NMR spectra in (a) pure PuO2 and (b) PuO2-x, which has oxygen defects. Insets of (a) and (b) show the crystal structuresof PuO2. The solid and dotted blue curves represent the deconvoluted Gaussian curve fit to the data and each component,
respectively.

Nuclear magnetic resonance (NMR) can be used to explore
electronic states via the interaction between nuclear andelectron spins. NMR has been applied to a wide range offields, such as molecular structure analysis in chemistry andbiology, for materials science, and for magnetic resonanceimaging (MRI) in medicine. Since an NMR signal was firstobserved in 1946, it has become possible to detect more than
90 types of nuclear isotopes. Nevertheless, NMR analysis of
the actinide nuclei has not been possible, except for 235U.
Among the actinide elements, Pu has attracted much attention
owing to its noble and mysterious character. Moreover, Pu is
important as an atomic-energy fuel, but the electronic state ofPu ions has not been well understood. For such reasons,scientists around the world have been searching for the Pu
NMR signal over the past 50 years or more, but until recently,
it remained elusive.
We therefore initiated a 239Pu NMR signal search throughan international collaboration with the U.S. Los Alamos
National Laboratory (LANL). After establishing a detailed
experimental design, we succeeded in observing the 239Pu
NMR signal for the first time. This success stems from theprecise control of many experimental parameters and the useof a very pure sample synthesized by chemists at LANL.
Fig.7-3(a) shows the 239Pu NMR spectrum measured forhigh purity PuO2. From the field dependence of the central
NMR frequency, the nuclear magnetic moment, which is theimportant physical constant, has been implied to be 0.15 µ N(µ N is the nuclear magneton). Fig.7-3(b) shows the 239Pu
NMR spectrum measured in PuO2-x, which has oxygen
defects. When the two spectra are compared, it can be seenthat the resonance frequencies and the structures are different.
This result implies that the 239Pu NMR analysis has highresolution and is very sensitive to the differences in thecoordination state of oxygen.
With this success, it is expected that the research regarding
the electronic state of various Pu compounds using 239Pu
NMR analysis will now progress. In particular, because thestability of the Pu ionic state in oxides can be judged byelucidating the electronic state of Pu, it is expected that 239Pu
NMR will be useful for constructing a safer, prolonged
storage method for used atomic-energy fuel containing Pu,
which is a globally important issue today.

Reference
Yasuoka, H., Chudo, H. et al., Observation of 239Pu Nuclear Magnetic Resonance, Science, vol.336, no.6083, 2012, p.901-904.
JAEA R&D Review 201283

Advanced Science Research

Generating Magnetic Flows with Sound=Discovery of a Link between Sound and Modern Spintronics=

(d)

(f)
V

Resonance frequency = 7.2 MHz(rotation of electron)

ChargeMagnet(b) Charge current
Up-spinelectronPiezoelectric
actuator(e)(c) Spin current

V
Voltage

Up-spinelectron

Platinum
MagnetSpincurrent

Sound
wave

5.4 MHz
3.6 MHz

7.2 MHz
5.4 MHz
3.6 MHz0

Down-spinelectronDown-spinelectronFig.7-4 Electrons, charge currents, and spin currents(a) Charge and spin, the two properties of electrons. (b)Up-spin and down-spin electrons move in the same
direction for charge current generation. (c) Up-spin anddown-spin electrons move in the opposite direction forspin current generation.

Electronics form the basis of the modern informationsociety. Spintronics, a new electronic technology, has
drawn global attention as a new way to achieve further
reduction in both size and power consumption of electronic
devices. We have found a new phenomenon linking suchmodern spintronics and ubiquitous sounds.
Electrons in a material possess two aspects: charge as asource of electricity and spin as a source of magnetism(Fig.7-4). While conventional electronics have been developedby making use of the charge currents in devices, spintronics
aim to develop an innovative functionality of devices bymaking use of the spin currents , or the flow of spins. The
spin current is indispensable for spintronics, but it is difficult
to drive the spin current in a device. Therefore, only a limited
number of ways for generating spin currents have been
identified to date.
In this context, we have discovered a new method forgenerating spin currents by merely injecting sound waves
into a material. Our method uses a spin current device,
which is composed of a magnet with a platinum terminal on

Piezoelectric
actuator

0

2

4

2
4
6
8
Frequency (MHz)

6

8

10

Frequency of sound waves (MHz)Fig.7-5 Spin current generation using sound waves(a) Schematic view of the spin current device used in thepresent study. (b) Schematic illustration of the spin current
generation process. (c) Dependence of the spin current signal on
sound wave frequencies. Inset: Piezoelectric actuator oscillates
strongly at a particular frequency.

the top, attached to a piezoelectric actuator that acts as asound wave generator (Fig.7-5). After injecting sound waves
into the spin current device , we measured the spin current
signal induced in the platinum terminal and found that whena sound wave with a certain frequency is injected, a spincurrent is generated in the spin current device .
The following phenomenon is known as the spin pumping
effect : When the spin current device is excited by radio
frequency (rf) waves, a spin current is generated in thedevice. In the present study, we have clarified boththeoretically and experimentally that a similar spin pumping
effect can be observed even if a sound wave is used to excite
the spin current device instead of rf waves.
Sound waves can propagate through materials that canneither conduct electricity nor possess magnetism, and thus
are used as device substrates. Therefore, the method
developed in the present study enables the extraction ofelectromagnetic energy from a substrate material and createsa new route for the development of innovative spintronic
devices.

Reference
Uchida, K., Adachi, H. et al., Long-Range Spin Seebeck Effect and Acoustic Spin Pumping, Nature Materials, vol.10, issue 10, 2011, p.737-741.

84 JAEA R&D Review 201210

Spin current signal due to sound waves

Platinum

Spin

Oscillation amplitudeof the actuator(a) Electron

Electric voltage induced in Pt

7-2

Advanced Science Research

7-3

Development of a New Method for Preparing
Atomically and Electronically Uniform Graphene=Significant Progress for Spintronic and Nanoelectronic Applications=Single layer
graphene
Nucleation

15

0.4 layer

0

Opening of the second
layer growth0
1020
Benzene exposure time (min)Peak position (cm-1)

Electron beam intensity of
specular reflection (a.u.)

30

2710

Coverage (layer number)
0.4 (en rote to graphene formation)
1.0 (single layer graphene)
Single layer graphene by exfoliation
Uniform

2700

2690

Large distribution

1584

1590

1596

1602

-1

Peak position (cm )Fig.7-6 Evolution of the electron beam intensity during
the growth of graphene
Change in the microstructure of the nickel surface during
benzene exposure was observed via in situ analysis of theelectron beam intensity of the specular reflection. It wasdemonstrated that graphene grows in a layer-by-layer fashionand that a graphene sheet with a uniform number of carbon
layers can be obtained through precise control of the benzene
exposure time.

Fig.7-7 Uniformity of the atomic and electronic structureof graphene
Micro-Raman analysis was performed on randomly selectedareas of graphene sheets (size 5 × 5 mm2). The large distributionof the peak positions on the vertical and horizontal axes reflect
the non-uniformity of the electronic state and the atomic
structure in graphene, respectively. It was found that singlelayer graphene (coverage: 1.0) has an extremely small
distribution of the peak positions. This result indicates that thegraphene film has a highly uniform electronic state and atomic
structure.

Since the discovery of a convenient method for the
fabrication of graphene via micromechanical exfoliation frombulk graphite (2010 Nobel Prize in physics), graphene has
attracted world-wide attention as an innovative nanoelectronicmaterial. In the field of spintronics, graphene is expected to
be an ideal spin-transport material owing to its extremely
long spin diffusion length and high carrier mobility. However,the non-uniformity in both the number of carbon layers andthe electronic states of exfoliated graphene is making it
difficult to control the spin transport properties. Therefore, a
new fabrication method that enables the tailoring of graphene
is indispensable for the development of spintronic applications.
In this study, we focused on a chemical growth method forgraphene. A graphene film was synthesized by exposing a
metal (nickel) to precursor (benzene) vapors. The growth
process for graphene was monitored via in situ analysis of theelectron beam intensity of the specular reflection (Fig.7-6),
and it was successfully demonstrated that the nucleation andstep flow of the graphene growth process can be
simultaneously monitored by the change in the specular beam
intensity. This approach makes it possible to obtain graphenewith a uniform layer number through precise control of thegrowth conditions. In addition, it was found that a remarkably
high uniformity of the electronic states of our graphenecompared to that of the exfoliated graphene can be achievedby adjusting the exposure conditions used to complete thegrowth of a single carbon layer, as shown in Fig.7-7.
Use of this tailored synthetic method should enableefficient control of the spin transport properties of graphene
and lead to the development of graphene-based spintronic
and nanoelectronic devices.

Reference
Entani, S. et al., Precise Control of Single- and Bi-Layer Graphene Growths on Epitaxial Ni(111) Thin Film, Journal of Applied Physics, vol.111, issue 6,2012, p.064324-1-064324-6.
JAEA R&D Review 201285

8   
Formation of Basis for Nuclear Energy R&D and Creation of Innovative Technology
Nuclear Science and Engineering Research
Supporting various activities of the government
and projects both inside and outside JAEA

Providing a technological platform
to industries and universities

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

Conservation and improvement of basis
for nuclear science and engineering

Preparing basic codes 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 at thepromotion of various activities, as shown in Fig.8-1,
including recovery from the accident at the Tokyo ElectricPower Company, Incorporated Fukushima Daiichi Nuclear
Power Station (1F). To this end, research is being conductedin the areas of nuclear data and reactor engineering, fuels andmaterials engineering, nuclear chemistry, and environmentand radiation science. Joint research with industry andacademia is also being promoted via the Nuclear Engineering
Research Collaboration Center (NERCC).
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(Fig.8-2). Calculation methods for nuclear reactor design are
also being studied (Topics 8-1, 8-2, 8-3). Regarding theaccident at 1F, a simple calculation tool to estimate thedecontamination effect on land was developed and released tothe public. Analysis of the accident is also under way (Topics
1-12, 1-19).
In the area of fuels and materials engineering, research on
nuclear fuels and materials used in nuclear reactors and fuelcycle facilities is being promoted (Fig.8-3) (Topic 8-4).
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 longlived nuclides from nuclear wastes, and detection of
extremely small amounts of nuclear materials (Topic 8-5).
These technologies are also being applied to the recovery
from the accident at 1F (Topics 1-10, 1-23).
In the area of environment and radiation science, studies ofthe behavior of radionuclides in the environment and thedevelopment of a database for radiation protection are being
promoted (Topics 8-6, 8-7). The behavior of radionuclides
released by the accident at 1F is being studied in order to predictthe long-term environmental consequences (Topics 1-1, 1-2).
In collaboration with industry, a production technique for
86 JAEA R&D Review 2012

medical radio-isotopes using an accelerator-driven neutron
source has been developed. A new corrosion-resistantmaterial is also being developed.

Fig.8-2 Beam line developed for nuclear data measurement(ANNRI) at the Materials and Life Science Experimental
Facility of J-PARC
To measure nuclear data for various nuclides such as minor
actinides, the Accurate Neutron-Nucleus Reaction measurement
Instrument (ANNRI) is under operation.

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

Nuclear Science and Engineering Research

8-1

Challenge of Obtaining Accurate Nuclear Data for Radioisotopes=Measurement of Neutron Capture Cross Sections of 244Cm with ANNRI=

100000This WorkJENDL-4.0 (Broadened)
M. S. Moore, et. al. (1971)

Cross Section (b)

10000
1000
100
10
1
New Data

0.1

2

20

200

Neutron Energy (eV)Fig. 8-4 Photo of an array of Ge detectors
in ANNRI
Using the detectors, energies of prompt γ-rays
are recorded together with a flight time of anincident neutron. An energy of the incident
neutron is calculated from the flight time.

Fig. 8-5 Measured neutron capture cross sections of 244Cm (●)
The results are compared with the data obtained by Moore (▲) andevaluated values in JENDL-4.0 ( ). The resonances below 20 eV
were observed in the capture reactions for the first time.

Accurate neutron capture cross section data for minor
actinides (MAs) and long-lived fission products (LLFPs) are
required in order to estimate the production and transmutation
rates for developing innovative nuclear systems. To satisfy
this requirement, the Accurate Neutron-Nucleus Reaction
measurement Instrument (ANNRI, Fig.8-4) has beendeveloped in the Material and Life Science Experimental
Facility of the Japan Proton Accelerator Research Complex.
A series of neutron capture cross-section measurements havebeen begun with ANNRI.
244
Cm is one of the most important MAs. However, there is
only one reported experimental data set, which was obtainedin 1969 using a nuclear explosion. The difficulties associatedwith experiments are as follows: Because strong decay γ-rays from 244Cm produce a severe
dead time, deduced cross sections have a large error due tothe dead time correction. (The radioactivity of our sample
was 1.8 GBq.)
 Because it is difficult to obtain and handle 244Cm samples,
uncertainties in the sample amounts become quite large.
(The uncertainty of our sample was about 16%.)
 Because the detectors are sensitive to fission events, a ratioof the sensitivity to fission events to the sensitivity tocapture events must be evaluated.
To overcome these difficulties, we have developed thefollowing new techniques: An accurate dead time correction method using randomtiming pulses
 A normalization technique at the first resonance of 240Pu,

which is the daughter nuclide of 244Cm, to reduce theuncertainty in the sample amount A method of evaluating the sensitivity ratio at the firstresonance of 245Cm, which has a large fission cross section
Using these techniques, we obtained the neutron capture
cross sections of 244Cm, as shown in Fig.8-5. The resonancesat around 7.7 and 16.8 eV were observed in the capture
reactions for the first time. An uncertainty of 5.8% wasachieved at the top of the first resonance of 244Cm.
Currently, in addition to 244Cm, analyses of 246Cm and 237Np
have been completed, and analyses of 241Am, 129I, 107Pd, 99Tc,
and 93Zr are in progress. These results will make significant
contributions in the development of innovative nuclear
systems.
The present study includes the results of Study on nuclear
data by using a high-intensity-pulsed neutron source for anadvanced nuclear system entrusted to Hokkaido University
by the Ministry of Education, Culture, Sports, Science andTechnology of Japan (MEXT). This project is partly
supported by a JSPS Grant-in-Aid for Scientific Research(S), Grant No.22226016, and by a Grant-in-Aid Young
Scientific (B), Grant No.22760675.

Reference
Kimura, A. et al., Neutron-Capture Cross-Sections of 244Cm and 246Cm Measured with an Array of Large Germanium Detectors in the ANNRI at J-PARC/MLF,
Journal of Nuclear Science and Technology, vol.49, no.7, 2012, p.708-724.
JAEA R&D Review 201287

Nuclear Science and Engineering Research

8-2

Research on Boiling Heat Transfer Mechanism=Development of Simultaneous Measurement Technology for Surface Heat Flux and Surface Temperature=
several µmjunctions (measuring points)
common metal film (copper)

1

2

heated surface
heated block

(1)

(3)

(4)

(1) bubble
generation(3)(4) bubble
detachment

3
heatedsurfaceheated(2)
1 mm

bubbleblock(copper)heated block

Fig.8-6 Principle of new temperature sensors
New temperature sensors with multiple individual metal
wires (constantan) and a common metal film (copper)
were developed.
(a) outer appearance(b) Microscope photograph of the surface center(before the common copper film was formed.)

nucleustemperature
sensors

sensor diameter
50 µm

500 µm

Fig.8-7 Boiling test piece with new temperature
sensors installed
Manufacturing technology was developed for the newsensors. A technology for placing the sensors in a highdensity of up to six sensors per 1 mm at a depth of lessthan 2 μm from the surface has been established.

Boiling has an extremely high heat transfer coefficient andis widely used in equipments such as heat exchangers,
boilers, and boiling water reactors. It has been studied
extensively, but its mechanism still remains unknown. This
research developed a measuring technology for the surface
heat flux and surface temperature. Pool boiling experiments
were performed to elucidate the heat transfer mechanism.
The following technologies are required for the measurementof the surface heat flux and temperature below a boiling bubble.
(1) A non-contact measurement technique without the
installation of sensors on the surface(2) A high spatial resolution technique to make measurements
at least to three points per 1 mm, considering that the contact
area of a boiling bubble with its heated surface has a radius ofonly about 1 mm(3) A rapid response technique to trace boiling, which is afast phenomenon with the highest frequency component at
more than 500 Hz
To achieve (1), we introduced an analytical approach toheat conduction in which the surface temperature and surface
heat flux are calculated simultaneously using the measured
body temperature. To achieve (2), we need to installtemperature sensors inside the body at a high density of threeor more points per 1 mm because the spatial and time
resolutions of the calculated values depend on those of thedetachingprocess

growing
process108

4
bubble
(air)

2

104
1

3

three phase line
temp. sensors

1

100

(2)

0

5

water
(liquid)2 3 4heated surface (solid)

ｰ V1 +

1.6 µmindividual
metal wire(constantan)

Measured temperature (℃)

112
ｰ V3 +ｰ V2 +

1 r(mm)0nucleus Boiling test piece

10

15

20

25

time (ms)Fig.8-8 Experimental results of temperature distribution
just below a boiling bubble
We succeeded in deriving the temperature distribution justbelow a boiling bubble from generation to detachment.

measured temperatures. Moreover, the sensors must be
located in close proximity to the heated surface to achieve (3).
To meet these requirements, we developed a newtemperature sensor. The concept is shown in Fig.8-6.
Multiple temperature junctions are formed in a copper block.
Each junction has an insulated constantan wire as its negativepole, but shares a common positive pole made of copper. The
common pole can be formed by sputtering technology, so thejunctions can be located in close proximity beneath thesurface. Fig.8-7 shows a boiling test piece with the newtemperature sensors installed. We established the manufacturing
technology for placing the new sensors at a high density of
up to six points per 1 mm and in close proximity to thesurface (less than 2 µm beneath the surface).
We performed pool boiling experiments. Here, a test piecewith four temperature sensors installed within 1 mm at a
depth of 1.6 µm beneath the surface was used. As shown in
Fig.8-8, we succeeded in deriving the temperature distribution
just below a bubble. The sensors show their lowest
temperatures when the solid-gas-liquid three-phase line
reaches their positions, and this confirms the reasonablenessof the hypothesis that heat transfer near the three-phase line
dominates the boiling heat transfer.
The present study was entrusted by the Ministry of Education,
Culture, Sports, Science and Technology of Japan (MEXT).

Reference
Liu, W. et al., Development of Measurement Technology for Surface Heat Fluxes and Temperatures, Nuclear Engineering and Design, vol.249, 2012, p.166-171.

88 JAEA R&D Review 2012

Nuclear Science and Engineering Research

8-3

More Precise Estimation of Reactor Decay Heat=JENDL/FPD-2011 and JENDL/FPY-2011=

235 U Thermal Neutron Fission

235 U Thermal Neutron Fission (Total)

25.0

1.60
1.40

Energy (%)
Half-life (%)
Yield (%)
Total (%)

Uncertainty (%)

f (t) × t (MeV/fission)

20.0
1.20
1.00
0.800
0.600

JENDL_2011
JENDL_2011 (+)
JENDL_2011 (−)
ORNL

0.400

15.0

10.0

5.00
0.200
0.00 -1
10

0

10

1

10

10

2

10

3

4

10

5

10

Time after fission burst (s)0.00 -3
10

-2

10

-1

10

10

0

1

10

2

10

3

10

10

4

5

10

Time after fission burst (s)Fig.8-9 Decay heat from 235U thermal neutron fission
Although the decay heat power decreases exponentially withtime, the curves in the figure are multiplied by the time on thevertical axis for easy recognition. Therefore, the decay heat
has no peak structure, as shown above. Please do not
misunderstand. Symbols with error bars are the experimental
data, and solid line is the calculated result with the error region(broken lines).

Fig.8-10 Origins of error for decay heat calculation
The error arises from the decay radiation energies, half-lives,and FPYs. The error due to the energies is dominant after 0.1 s.

Many radio-isotopes (RIs) exist in nuclear reactors, evenafter they are shut down. These RIs emit radiation, whichchanges into decay heat. Most of these RIs are fission products(FPs), which are produced from fission reactions of uranium(U) and plutonium (Pu). There are more than 1000 FP nuclides,
including isotopes with very short lives. To estimate FP
production rates and decay heat, nuclear data such as the halflives, emission energies of radiation, FP yields (FPYs, theratio of FP produced per fission reaction) are necessary. The
databases that can be used for decay heat calculations ofspent nuclear fuels are JENDL (Japanese Evaluated Nuclear
Data Library) FP Decay Data File 2011 (JENDL/FPD-2011)
and JENDL Fission Yields Data File 2011 (JENDL/FPY2011). JENDL/FPD-2011 includes the FP half-lives, emission
energies of radiation, and intensities. The FPYs for manyfission reaction systems are contained in JENDL/FPY-2011.
These databases were produced with the latest knowledge aswell as error data, which are the first in the world.
The decay heat of a 235U fission burst with thermal neutronvs. the decay time is shown in Fig.8-9. The calculated result
reproduces the experimental data well at shorter periods, andthe error region covers almost all of them. Fig.8-10 indicatesthe origin of the error. Overall, the error arising from theemission energies of radiation from decaying FPs is dominant.
The reactor decay heat has been investigated since the very
early stage of peaceful nuclear power use. Although anempirical equation based on experimental data was used atthat time, the summation method, which accumulates theindependent contribution of each FP nuclide, has beendeveloped. Some decay data have been unclear, sinceconsidering FPs should consist of more than 1000 nuclides.
With the development of the database and summation
method, the accuracy and reliability of decay heatcalculations are expected to be greatly improved.

Reference
Katakura, J., JENDL FP Decay Data File 2011 and Fission Yields Data File 2011, JAEA-Data/Code 2011-025, 2012, 73p.
JAEA R&D Review 201289

Nuclear Science and Engineering Research

8-4

Development of Database for a Practical Fuel Design=Formulation of Thermal Conductivity of Zr-Based TRU Nitride Fuels=

Thermal conductivity (W/mK)

Fig.8-11 (Zr0.58Pu0.21Am0.21)N sintered compact
Appearance of high-density sintered compact with a diameterof about 3 mm. Photograph was taken in a glove box.

80
ZrN

60
40

(Zr0.80Pu0.10Am0.10)N(Zr0.58Pu0.21Am0.21)N20(Pu0.50Am0.50)N
UO2

0

500

1000
Temperature (K)1500

Fig.8-12 Thermal conductivity of Zr-based TRU nitrides
Thermal conductivity was corrected for porosity to theoreticaldensity for comparison with that of other materials.

Spent fuels from nuclear power plants contain minor
actinides (MAs: Np, Am, Cm) which are long-lived radioactive
nuclides; thus, the management of MAs will be a key issue
for sustainable nuclear energy use. To solve this problem,
nuclear transmutation technology using an accelerator-driven
system (ADS) is an option in future nuclear fuel cycles. A
nitride solid solution containing MA and Pu (transuranium[TRU] nitride) diluted by ZrN is proposed as a candidate
material for the ADS fuel. Therefore, it is necessary todevelop a thermal conductivity database for a practical fueldesign. To this end, we prepared Zr-based TRU nitrides andmeasured their thermal diffusivity and heat capacity todetermine their thermal conductivity. Furthermore, thethermal conductivity was fitted to equations as functions of
the temperature and ZrN concentration.
A high-density sintered compact of Zr-based TRU nitride
was prepared as follows (Fig.8-11). (1) The TRU nitride wasprepared via carbothermic reduction of dioxide. (2) ZrN
powder was synthesized from Zr metal through its hydride.
(3) The nitride mixture was pressed into a disk and sinteredin a stream of N2+4%H2 gas. Using the sintered compact, thethermal diffusivity and heat capacity were measured by a

Thermal conductivity (W/mK)80

MA nitride fuel region for ADS
Curve by equation

ZrN

60
40

(Zr0.80Pu0.10Am0.10)N(Zr0.58Pu0.21Am0.21)N

20 (Pu0.50Am0.50)N
0
0.0

0.2
0.6
0.8
Molar fraction of ZrN

1.0

Fig.8-13 ZrN concentration dependence of thermal
conductivity at 873 K
Thermal conductivity was fitted as a function of ZrN
concentration.

λ＝A+Bexp (x/C)
λ: Thermal conductivity, x: ZrN concentration, T: Temperature
A＝4.7624+1.3937x10-2T-1.4543x10-5T2+5.6365x10-9T3
B＝-0.15962-4.2325x10-4T+9.1965x10-7T2
C＝-9.8103x10-3+2.3662x10-4T-3.2471x10-8T2

Fig.8-14 Fitting equation of thermal conductivity for(ZrxPu(1-x)/2Am(1-x)/2)NCoefficients in equations A, B, and C were determined as afunction of temperature.

laser flash method and drop calorimetry, respectively.
The thermal conductivity of Zr-based TRU nitride increased
with temperature, unlike that of oxide fuels (Fig.8-12). The
increase in thermal conductivity with temperature is believed
to be derived from the increase in the electronic contribution.
The results prove that a Zr-based TRU nitride is a goodcandidate fuel with superior thermophysical properties because
ZrN, used as a diluted material, effectively raises the thermal
conductivity. A fitting equation for the thermal conductivityof (ZrxPu(1-x)/2Am(1-x)/2)N was obtained by evaluating the ZrN
concentration dependence at each temperature (Fig.8-13,
Fig.8-14). The predicted values agreed well with theexperimental ones, indicating that the thermal conductivity ofa nitride fuel for an ADS can be predicted in a ZrN
concentration range from 60 to 80mol%.
In this study, we obtained the thermal conductivity andheat capacity of Zr-based TRU nitrides for the first time andprepared the fitting equation for the thermal conductivity.
These results, which can be included in a database forpractical design study, will contribute to ADS fueldevelopment.

Reference
Nishi, T. et al., Thermal Conductivities of Zr-Based Transuranium Nitride Solid Solutions, Journal of Nuclear Science and Technology, vol.48, no.3, 2011,
p.359-365.

90 JAEA R&D Review 2012

Nuclear Science and Engineering Research

8-5

Analysis for Safety Assessment of Radwaste Disposal=Validation of Calculated Values of Content of Difficult-to-Measure Nuclides 79Se and 135Cs=

From
hot cellto hood

Dissolution of spent
nuclear fuel pellet in
hot cell facility

Cationexchange
resin

Preparation of diluted
sample solution(in laboratory hood)

Simplified chemical
separation(approximately 30 min)Measurement by inductively
coupled plasma-mass
spectrometry (ICP-MS)(1 M HNO3)

0.25 mℓ1 M HNO3
2.75 mℓ

Non-separation
measurement

1 M HNO3
10 mℓ

Cation exchange resin column (1 mℓ)

Se

Se

Br removal(If necessary)Se

Cs

ICP-MS

82Se/79Se, 133Cs, 135Cs ( 137Cs)

ǫ spectrometry 137Cs

10( calculation of Cs
recovery percentage)

Fig.8-16 Analytical procedure for 79Se and 135Cs
This procedure enables mutual separation of Se/Cs
along with removal of highly radioactive coexisting
components and sources of interference in ICP-MS
measurement.

8

Experimental
value
ORIGEN2(JENDL-4.0)

79

Se

6
4
2
0
0

10 20 30 40 50
Fuel burnup (GWd/t)

Content in initial U (g/MTU)

Sample solution

Content in initial U (g/MTU)

Fig.8-15 Procedure for determination of 79Se and 135Cs in spent nuclear fuel sample
Sample solution was prepared by dissolving a spent nuclear fuel pellet in a hot cell, which is a shielded facility forworking with highly radioactive substances by remote operation. The preparation steps were significantly simplified toreduce the radiation exposure time.

500
Experimental
value
ORIGEN2(JENDL-4.0)

400

135

300

Cs

200
100
0

0

10 20 30 40 50
Fuel burnup (GWd/t)

Fig.8-17 Comparison of measured values and calculated values
obtained by ORIGEN2
79
Se and 135Cs contents in spent nuclear fuel were calculated using
ORIGEN2. The measured values proved that ORIGEN2 provided reliablecalculation.

High-level radioactive waste (HLW) produced in a nuclear
fuel reprocessing plant contains long-lived radionuclides withhalf-lives of more than 104 years. Among such radionuclides, 79Se
and 135Cs will govern the public dose originating from the
HLW repository 104 to 106 years after disposal is implemented.
Therefore, inventory estimation of 79Se and 135Cs in HLW isessential for realizing safe HLW disposal. However, thereliability of the nuclear data used for dose evaluation has notbeen sufficiently verified by experimental values because ofdifficulties in their measurement. In 79Se analysis, a sample
containing a large amount of radioactive substances isnecessary for accurate and precise determination because
only a small amount of 79Se is present in HLW.
In this study, we developed a simple and reliable
separation procedure to reduce the radiation exposure time
(Fig.8-15). Mutual separation of Se/Cs along with removal ofhighly radioactive coexisting components and sources ofinterference in inductively coupled plasma mass spectrometry(ICP-MS) measurement was achieved by a single cationexchange step (Fig.8-16). The analytical method presented inthis study provided the first experimental values for the 79Se
and 135Cs contents of spent nuclear fuel in Japan. In addition,
the measured values showed good agreement with the values
calculated using ORIGEN2 (an isotope generation anddepletion code) based on the latest evaluated nuclear datalibrary, JENDL-4.0 (Fig.8-17). This confirms that ORIGEN2
is applicable to the estimation of the 79Se and 135Cs contents ofspent nuclear fuel.
The present study is the results of Experimental study on
the inventory estimation of long-lived and scarcely analyzed
radionuclides in high-level radioactive waste (Stage I) fundedby Japanese 11 electric power companies.

Reference
Asai, S. et al., Determination of 79Se and 135Cs in Spent Nuclear Fuel for Inventory Estimation of High-Level Radioactive Wastes, Journal of Nuclear Scienceand Technology, vol.48, no.5, 2011, p.851-854.
JAEA R&D Review 201291

Nuclear Science and Engineering Research

8-6

Removal Mechanism of Radiation-Induced Nucleotide Damage=Simulation Study of the Mechanism of Molecular Recognition by the Enzyme MutT=

(a)

dGMP

8-oxo-dGMP
different
structure

B’

(b)

A’

B

When MutT binds
8-oxo-dGMP,
loop A and loop B
move like A’→A
and B’→BAloop A andloop B are
flexible8-oxo-dGMP

Distance (1×10-10m)

20

10

0
0

MutT

5

10

15

20

Time (ns)Fig.8-18 Structure of dGMP, 8-oxo-dGMP, and MutT(a) Structure of dGMP and 8-oxo-dGMP.
(b) Conformational difference between substrate-free
MutT (red) and 8-oxo-dGMP-MutT complex (blue).

Fig.8-19 Time dependence of the distance between thetips of loop A and loop B
Red line represents 8-oxo-dGMP-MutT, and green line
represents dGMP-MutT.

Radiation can induce oxidization damage in nucleotides.
Oxidized nucleotides cause strong mutagenesis when they are
erroneously incorporated into nucleic acid. MutT carries out
its enzymatic function of removing oxidized nucleotides fromthe nucleotide pool before incorporation.
An example of an oxidized nucleotide is 8-oxo-dGMP. Its
chemical structure does not differ significantly from that of
the non-damaged nucleotide, dGMP (Fig.8-18(a)). An
enzyme generally binds strongly to a substrate when both
structures fit each other. It is difficult for an enzyme todistinguish between the similar structures of the substrates;
hence, MutT binds both 8-oxo-dGMP and dGMP because of
their similar structures. However, 8-oxo-dGMP is bound far
more strongly than dGMP. How can the large difference inthe binding constants arise from the small difference in thechemical structures?
A comparison of the substrate-free and 8-oxo-dGMPbound X-ray crystal structures reveals a remarkabledifference. Two loops surrounding the substrate binding site

are closed in the complex with 8-oxo-dGMP, whereas they
are open in the substrate-free structure (Fig.8-18(b)). On thebasis of this, we conducted a molecular dynamics simulationof the 8-oxo-dGMP-MutT and dGMP-MutT complexes.
The results for the 8-oxo-dGMP-MutT complex showedthat the closed structure is more stable, and the structuralfluctuations are small. On the other hand, two loops are open
in the dGMP-MutT complex, and the closed structure isunstable (Fig.8-19). More specifically, when a hydrogenbond, which plays a crucial role in maintaining the closed
substrate in the 8-oxo-dGMP-MutT complex, is absent in thedGMP-MutT complex, the closed structure is no longer
maintained. This study revealed the molecular mechanism
distinguishing the oxidized nucleotide from the normal
nucleotide. MutT removes the oxidized nucleotide effectively
using this mechanism. In this way, an understanding of themolecular mechanisms of DNA repair plays an important rolein the investigation of biological responses to radiation.

Reference
Higuchi, M. et al., Enhanced Resolution of Molecular Recognition to Distinguish Structurally Similar Molecules by Different Conformational Responses ofa Protein upon Ligand Binding, Journal of Structural Biology, vol.173, issue 1, 2011, p.20-28.

92 JAEA R&D Review 2012

Nuclear Science and Engineering Research

8-7

Prediction of Radiation Damage in Accelerator Facilities and Nuclear Power Plants=Development of Radiation Damage Model for Various Particles=

Coulomb scattering Cascade damage
approximation
Incident particle

Old radiation
damage model

Displacement
cross section

Incident particles

New PHITS
Coulomb scattering
Secondary particle
Secondary particle
production

Old radiationtransport code

Cascade damage
approximation

Displacement
cross section
Displaced atom

5

Proton beam, Cu target
10

4

10

3

10

2

10

0

10

1

10

2

10

3

Incident proton energy (MeV)

Fig.8-20 Overview of radiation damage model in PHITS
Radiation damage results from displaced atoms produced by Coulomb
scattering between particles and a target atom.

Radiation damage of materials is usually evaluated as afunction of the average number of displaced atoms for all
atoms (DPA) in a material. DPA is defined as the integral ofthe displacement cross section and the irradiation fluence.
The accuracy of the evaluation of DPA values affects themaintenance periods of accelerator facilities and nuclear
power plants. Monte Carlo radiation transport codes havebeen used to study radiation damage of materials. However,these codes did not calculate the displacement cross section
because the radiation damage model was not implemented in
the codes.
To evaluate the radiation damage of materials by various
particles in a wide energy range, we developed a calculation
model to estimate the displacement cross section, including theeffects of nuclear reactions and Coulomb scattering. Fig.8-20
shows an overview of the radiation damage model in theParticle and Heavy Ion Transport code System (PHITS). The
conventional radiation damage model treated only the
Coulomb scattering between the incident particle and the10

Displacement cross section (b)Displaced atom
Particle transport

Nuclear reaction

This work (Incident+Secondary)This work (Incident)
Expt., Greene
Conventional model, Jung

Fig.8-21 Displacement cross section as afunction of incident proton energy for the
proton irradiation of copper
Improved PHITS can be applied to evaluate thedisplacement cross sections for various particles
over a wide energy range.

target atom. In this work, we integrated the nuclear reaction
model, Coulomb scattering model, and cascade damage
approximation to calculate the contribution of various
secondary particles such as protons, heavy ions, and neutrons
to the DPA values. As shown in Fig.8-21, the improved
PHITS could reproduce the experimental data in the highenergy region when the displacement cross section was
calculated for irradiation of protons over a wide energy
range.
Thus, the improved PHITS can be applied to evaluate thedisplacement cross sections precisely for various particles
and to calculate DPA values over a wide energy range fromeV to GeV. PHITS version 2.30, which includes thedeveloped calculation model, has been used to design the
Facility of Rare Isotope Beam (FRIB) and Transmutation
Experimental Facility (TEF) in J-PARC. The code will be
used to develop a database of displacement cross sections fordesign of nuclear power plant and nuclear fusion researchfacility.

Reference
Iwamoto, Y. et al., Improvement of Radiation Damage Calculation in PHITS and Tests for Copper and Tungsten Irradiated with Protons and Heavy-Ionsover a Wide Energy Range, Nuclear Instruments and Methods in Physics Research B, vol.274, 2012, p.57-64.
JAEA R&D Review 201293

9   
Research on a Naturally Safe HTGR for a Low-Carbon Society

Major features

Low temperature
heat utilization

࡮Seawater
࡮Multi-purpose heat utilization
࡮Excellent economy even small sized reactor‫ޓ‬desalination
࡮District heating
࡮Naturally safe without any engineered
࡮Agriculture, fishery‫ޓ‬safety features

HTGR Commercial Plant

High
temperature
heat utilization

࡮Hydrogen‫ޓ‬production
࡮Steam supply
Electric generation ࡮High efficiency
by Gas turbine‫ޓ‬power generation

High Temperature engineering Particle fuel coated
World’s most advanced homegrown technology
with ceramics
Test Reactor (HTTR):
Oarai, Ibaraki
࡮Ceramics coating (TRISO) for Uranium fuelUO2
Temperature limit: 1600 ℃ Long term (~3 times of LWR),

1998 Criticality
2001 Full power

Diameter ~㧝mm

2004 950 ℃

Isotropic graphite

complete retainment of fission
products (100~1000 times of HTGRs
in USA and Germany)࡮3-dimensional isotropic, irradiation resistant graphite
High strength, high heat conduction,

Temperature limit: 2400 ℃ irradiation resistance࡮Heat resistant, oxidation resistant Ni-base superalloy
࡮High-temperature structure design on thermal insulation,
Heat resistant alloy‫ޓ‬thermal expansion, anti-vibration
High temperature (950 ℃) heat

Temperature limit: ~950 ℃ can be extracted from the reactor
Thermal power‫ޓ‬:
Outlet temperature :Coolant pressure‫ޓ‬:

30 MW
950 ℃
4 MPa

࡮High temperature, high pressure He gas handlingTemperature limit: ~950 ℃ Chemically, nuclear, andmechanically stable coolingFig.9-1 Outline of HTGR: features, heat utilization, major specifications, and technologies of HTTR
The HTGR is thermal-neutron reactor that is cooled by helium gas and is moderated with graphite. It can meet various heat production
requirements; in particular, the HTGR can be a naturally safe nuclear reactor that is strongly expected to regain public trust for
generating nuclear power.

We have been conducting research on a naturally safe hightemperature gas-cooled reactor (HTGR) and its applications
toward building a low-carbon society.
The HTGR can supply heat at temperatures of 950 ℃; thisexceeds the 300 ℃ heat currently provided by light water
reactors (LWRs). This is accomplished using inert helium
gas, instead of water, as the coolant. When high-temperature
heat is obtained from a nuclear reactor, the same amount offission energy from uranium can produce greater quantities ofelectricity and hydrogen.
Three other cutting-edge Japanese technologies, developed
at the JAEA, make it feasible to obtain such hightemperatures from an HTGR. The first is a fabrication
technology for producing ceramic-coated fuel particles ofapproximately 1 mm diameter. Ceramics remain stable evenat 2500 ℃ and provide superior heat-resistant coatings thatcan contain radioactive fission products within fuel particles.
The second is a fabrication technology for isotropic andirradiation-resistant graphite blocks. Generally, graphite is
non-isotropic; that is, material properties of graphite, such as
strength and thermal conductivity, are dimension-dependent,
and materials having dimensionally dependent properties are
typically avoided for reactor construction. The third is amanufacturing technology for a heat-resistant and oxidationresistant superalloy whose contents have been carefully
controlled. With these major technologies, together with
94 JAEA R&D Review 2012

further developments of know-how for high-temperature
structural design and helium-gas handling, we used the HTGR
to successfully produce high-temperature heat at 950 ℃. Thiswas done in 2004, for the first time in the world. (Fig.9-1).
The heat from the HTGR is useful for power generation,
for hydrogen production in car fuel cells, for direct-reduction
ironmaking, and for steam supply to various industries. Waste
heat can be used for district heating and desalination. The
thermal discharge to the environment can be drastically reduced
from 67% using an LWR to less than 30% using an HTGR.
HTGRs have inherent self-regulating features that, in any
accident, prevent a harmful release of radionuclides to theenvironment and surrounding populations. Specifically, noengineered safety features are required to prevent accidental
overheating, oxidation of fuel-coating layers, and generationof explosive gases. This unique safety advantage of HTGRs
is expected to regain the public trust in nuclear power
following the accident at the Tokyo Electric Power Company,
Incorporated Fukushima Daiichi Nuclear Power Station(Topics 9-1, 9-2, 9-3).
Elsewhere in the world, national projects are being carriedout for constructing commercial HTGRs. China has started
construction of the first unit. The US has completed theconceptual design of a Next Generation Nuclear Plant andhas proceeded to the next design stage. Kazakhstan isplanning a feasibility study on HTGR construction.

Nuclear Hydrogen and Heat Application Research

9-1

Explore the Ultimate Nuclear Reactor That Is Safe Even If Accidents Occur=Conceptual Design Study of a Naturally Safe HTGR=

Confinement function

ReactorPhenomena that degrade theconfinement function

Causes ofeventsInherent safety
feature㧖1

Diffusive release

Doppler effectCore heat up

Sublimation

Fuel coating

㧖1㧘
㧖2

Thermal conduction,
radiation,
natural convection

Fissionproduct
Uranium

Corrosion

Oxidation

Oxide layer
formation

Flammable
gas explosion

Flammable gas
oxidation

Attain astable state‫ޓ‬Retainradionuclides
withinconfinement

㧖2

RPV

FailureCV

Fuel-coating layers

Fig.9-2 Concept of the naturally safe HTGR
The goal is to protect people and the environment from the harmful effects of radionuclides, even if an accident leads to the loss ofreactor safety systems. This goal can be achieved by utilizing physical phenomena to suppress the progression of any loss ordegradation of the confinement function of fuel-coating layers, thereby retaining radionuclides within coated fuel particles. (＊1 topic 9-2,
＊2
topic 9-3)

It is important to regain the public trust in nuclear reactor
safety following the accident at the TEPCO s Fukushima
Daiichi NPS. Questions and comments from the public, suchas how to deal with unanticipated events and even if theprobability of an accident is small, use of nuclear reactors is
unacceptable because the consequences of an accident are
extremely grave , should be addressed. We proposed a newconcept for an HTGR, referred to as a naturally safe HTGR,
to ensure no harmful release of radionuclides occurs under
any accident conditions. This reactor has self-regulating
characteristics that suppress the progression of accidents via
physical phenomena even in the absence of engineered safety
features.
A coated particle fuel is used in the HTGR; the integrity ofthis coating can be adequately maintained up to 1600 ℃,
thereby confining radionuclides to the fuel particles, evenunder extreme conditions. Possible physical events that leadto the loss or degradation of the confinement function of thefuel-coating layers include (a) core heat up caused by anincrease in heat generated or a decrease in the amount ofcooling, (b) oxidation caused by air admitted during an airingress accident, and (c) flammable gas explosion ofcarbon monoxide, which is produced by the oxidation ofgraphite with air (Fig.9-2). The core heat up can besuppressed by the Doppler effect and by the removal of heat
via thermal conduction, radiation, and natural convection, as
has been demonstrated using the HTTR. Oxidation can besuppressed by the formation of a stable oxide layer (i.e., SiO2)
on the surface of the fuel-coating layer. Flammable gas
explosion can be avoided by maintaining the flammable gas
concentration below the explosion limit, taking advantage ofthe mild oxidation of the flammable gas by air. Thoughengineered safety features, such as the decay-heat-removal
system and the reactor containment vessel, are installed in theformer HTGR, such as the HTTR, the naturally safe HTGR is
intended to ensure safety only utilizing physical phenomena,
without any engineered safety features. This goal will betested in demonstrations using the HTTR after further R&D
studies.
We aim to assist in regaining the public trust in nuclear
reactor systems by ensuring that, under any accident
conditions involving the naturally safe HTGR, the people andthe environment will remain safe.

Reference
Ohashi, H. et al., Concept of an Inherently-Safe High Temperature Gas-Cooled Reactor, The 3rd International Conference on Advances in Nuclear Scienceand Engineering 2011 (ICANSE 2011), American Institute of Physics Conference Proceedings, vol.1448, 2012, p.50-58.
JAEA R&D Review 201295

Nuclear Hydrogen and Heat Application Research

9-2

Spontaneous Stabilization of HTGRs without Reactor Scram and Core Cooling=Safety Demonstration Tests Using the HTTR: Loss of Reactivity Control and Core Cooling=

Table 9-1 Conditions and results of safety demonstration tests using the HTTR
In the safety demonstration tests using the HTTR, HTGRs were demonstrated to spontaneously stabilize despite a simultaneous lossof reactivity control and core cooling.
The Accident at the TEPCO’s
Fukushima Daiichi NPSHTTR Safety Demonstration Tests
Loss of reactivity control by disabling movement of all control rods ⇒Reactor power could decrease effectively to zero spontaneously

Active ⇒ Reactor power could decrease to almost zeroautomaticallyCore cooling

Loss of core cooling by tripping all gas circulators ⇒ Core could be cooled
spontaneously (Decay heat could be removed successfully)

Loss of core cooling ⇒ Decay heat could not be removed

Final state

HTTR could stabilize spontaneouslyCore meltdown, hydrogen generation, reactor building
explosion, release radioactive materials

Reactivity control

Trip all gas circulators

100
Flow rate (%)

50

Disable insertion of control rods
Control rods
Water pump

RPV

VesselcoolingsystemFailure to insert any control rod into the core

1.0
Percentage ofinitial valueof reactor power (-) 0.5

Reactor power could decrease effectively to zero
spontaneously

Experiment result

0.0
1600

Water-coolingCore

Experiment result

0

Fuel temperature could stabilize spontaneously
without a significant increase

1200

Trip
all gas circulators

Maximum Fuel
Temperature (℃)

800

Analysis result

4000Decay heat removal by thermalradiation and air natural convection

Helium gas circulators

Containment vessel
Fig.9-3 Outline of safety demonstration tests using the
HTTR
At an initial power of 30%, the test was performed by tripping
all gas circulators and disabling the insertion of all control rods
into the core. As a result, the coolant flow rate immediately
dropped to zero, causing the reactor transient to start.

In a safety demonstration test involving a loss of bothreactor reactivity control and core cooling, HTGRs such as
the HTTR, which is the only HTGR in Japan, demonstrate thatthe reactor power would stabilize spontaneously (Table 9-1).
In the test at an initial power of 30%, when the insertion of
all control rods was disabled and all gas circulators were
tripped to reduce the coolant flow rate to zero, a reactor
transient was initiated and examined. The results confirmedthat the reactor power would decrease immediately andbecome effectively zero (Fig.9-3, Fig.9-4). This reactor
behavior occurred because the temperature of U-238 in thefuel increased, which increased the absorption rate ofneutrons caused by the Doppler effect, immediately stopping
the fission chain reaction.
Although the reactor coolant flow rate dropped to zero andthe decay heat was being generated, the fuel temperature did
not show a large increase because the large heat capacity of
the graphite core absorbed heat from the fuel in a shortperiod. Over a longer period, some decay heat was conducted
through the reactor pressure vessel wall and was emitted bythermal radiation from the outer surface of the reactor

0

1

2

3

4

5

6

7

Elapsed time (h)Fig.9-4 Transients of coolant flow rate, reactor power,
and maximum fuel temperature
At the HTTR, even though no control rods were inserted into
the core, the reactor power effectively decreased to zero
because of the reactor’s negative temperature coefficient ofreactivity. Moreover, despite tripping all gas circulators andlosing core cooling, we demonstrated, by using the real HTGR,
that the decay heat could be removed without a significant
increase in the fuel temperature and that the reactor power
would stabilize spontaneously.

pressure vessel.
In the accident at the TEPCO s Fukushima Daiichi NPS
Unit 1 caused by the Great East Japan Earthquake andTsunami, the control rods were successfully inserted into thecore, and the reactor was shut down; however, the decay heat
could not be removed because core cooling was lost as soon
as the station blacked out. This resulted in the core meltdown,hydrogen generation, and explosion of hydrogen in thereactor buildings, causing a release of a large amount ofradioactive material into the atmosphere.
In contrast, even though no control rods were inserted into
the core, the reactor power of the HTTR can be effectively
reduced to zero. Moreover, despite tripping all gas circulatorsand losing core cooling, real HTGRs such as the HTTR,
demonstrate that the decay heat can be removed; the core wasnever exposed to the danger of a meltdown, and the reactor
power stabilized spontaneously (Fig.9-4).
We plan to perform more safety tests, including tripping
the vessel cooling system, to test the removal of decay heatfrom the core to the ground as the final heat sink.

Reference
Takamatsu, K. et al., Reactor Kinetics in a Loss-of-Forced-Cooling (LOFC) Test of HTGRs, Proceedings of the 20th International Conference on Nuclear
Engineering (ICONE 20), California, USA, 2012, ICONE20-54641, 10p., in CD-ROM.

96 JAEA R&D Review 2012

Nuclear Hydrogen and Heat Application Research

9-3

Design Criteria of HTGRs for Withstanding Extreme Events=Investigation of Technical Feasibility for Naturally Safe HTGRs=

Decay heat

Conduction
Conduction

Oxidation
heatCore power density(MW/m3)16

Naturalconvection

Decay heat
12
8
4

CO
productionCOoxidation
RPV

Air ingress

Surrounding
structures

Coolant channel diameter(mm)

0

Radiation

Decay heat + oxidation heat

100
80
60

Design envelope
2.0Core radius (m)1.0

Design envelope forremoving
oxidation heat

Designenvelope

40
20
0
0.01

Design envelope
for avoiding
CO combustion

0.1
1
Coolant channel flow area (m2)10

Fig.9-5 Physical phenomena in HTGR during extreme
eventsDuring extreme events that cause loss of coolant and massive
air ingress in HTGRs, the decay and residual heat in the core
are removed by physical phenomena such as thermal
conduction, radiation, and natural convection because of the
intrinsic temperature differences. Carbon monoxide, generated
by the oxidation of graphite in air, is converted into carbon
dioxide by reacting with oxygen in the air.

Fig.9-6 Design envelope for self-stabilization during
extreme events
Design envelope which can remove core heat and suppress
carbon monoxide concentrations below the lower flammability
limit for extreme events that cause loss of coolant and massive
air ingress in HTGRs is clarified.

The Naturally Safe HTGR concept aims to intrinsicallymaintain safety; that is, under accident conditions, physicalphenomena ensure that radionuclides are retained withincoated fuel particles (Fig.9-5). We examined the conditions
that can suppress physical phenomena that would otherwise
degrade the integrity of fuel-coated layers. In particular, we
focused on core heat up due to radioactive decay, theoxidation of graphite, and the production of carbon monoxideby the oxidation of graphite with ambient air.
The temperature and carbon monoxide concentration in thereactor core were investigated as factors that can influencecore design parameters, such as core diameter, size of coolant
channels. The goal is to set HTGR core design parameters sothat, during a loss of coolant and massive air ingress, thetemperature and carbon monoxide concentration in thereactor core remain within safe limits.
A set of simulations clarify the range of average power

densities and diameters of reactor cores that can suppress
core peak temperatures within the allowable limit of 1600 ℃.
Such control of core temperatures is to be purely performed
by inherent means, such as thermal conduction, radiation fromthe reactor pressure vessel (RPV) to surrounding structures,and natural convection in the reactor cavity (Fig.9-6). In
addition, the results show that indeed there are ranges ofdesign parameters that support the removal of oxidation heatand suppress the concentration of carbon monoxide in thecore below the lower flammability limit of 12.5% (Fig.9-6).
The simulation results indicate that, by fully utilizing
inherently safe characteristics, HTGRs have the potential toeliminate massive, uncontrolled radioactive releases from
nuclear plants to the environment.
We will explore the technical feasibility of Naturally Safe
HTGRs through research and development with the goal of
fostering public understanding.

Reference
Sato, H. et al., Analysis of Core Heat Removal Capability under DLOFC Accidents for HTGRs, Proceedings of 6th International Topical Meeting on High
Temperature Reactor Technology (HTR2012), Tokyo, Japan, 2012, paper HTR2012-8-003, 9p., in CD-ROM.
JAEA R&D Review 201297

10
10  
Execution of the Decommissioning of Nuclear Facilities
and the 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
package

Treatment andstorage of radioactive
waste

Operation and maintenance ofexisting waste treatment facilities

Waste package
inspection(Advanced volume reduction facilities, etc.)
Construction of waste
treatment facilities at
each site

Developmentof relevant
technology

Clearance

Waste from dismantling

Decommissioning of
nuclear facilities

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

Fig.10-1 Outline of measures for decommissioning and radioactive waste treatment/disposal
We are establishing systems for decommissioning nuclear facilities and for radioactive waste management. In addition, we aredeveloping related technology (decommissioning, treatment, disposal, etc.), planning and constructing radioactive waste
treatment/disposal facilities, and providing support for revising related regulations.

Safe and efficient decommissioning of nuclear facilities
and radioactive waste treatment/disposal are two of our major
missions for advancing our R&D activities. We will bedisposing of radioactive waste generated not only from ourresearch activities but also from universities, institutes,
industrial facilities, etc.
We are establishing systems for decommissioning nuclear
facilities and for managing radioactive waste, and are also
developing the related technology (Fig.10-1).
In addition, we are studying the applicability of our R&D
products for dealing with the accident at the Tokyo ElectricPower Company, Incorporated Fukushima Daiichi Nuclear
Power Station.
R&D for Waste Disposal
Radioactivity in low-level radioactive wastes, such as thosegenerated during decommissioning of nuclear facilities,
should be identified in order to be able to properly dispose ofthem.
Methods for the dissolution of radioactive contaminants
attached to the surface of metal wastes and the chemical
98 JAEA R&D Review 2012

separation of an infinitesimal quantity of Pu and Am in a
dissolved solution containing a large amount of Fe(III) weredeveloped for the quantification of Pu and Am. Approximately
100% of Am was recovered from a solution containing 1 g ofFe(III) using a transuranium (TRU) resin with the addition of
1.3 equivalents of ascorbic acid. The developed method wasapplied to the analysis of Pu and Am in the pipe waste
sampled from the advanced thermal reactor Fugen (Topic 10-1).
R&D for Waste Treatment
For radioactive waste processing and disposal, it isimportant to reduce the expense and to improve the safety ofdisposal. Incinerated ash generated from burnable wastes isstored by the JAEA and contains a small amount of heavy
metals such as Pb and Zn. The incinerated ash is solidified
with cement, and then the solidified cement form is disposedof in an underground facility. However, the heavy metals are
eluted from the solidified form when it comes in contact withgroundwater. Therefore, we have developed a technique forpreventing heavy-metal elution using additives (Topic 10-2).

Development of Decommissioning and Radwaste Treatment Technology

10-1 Analysis of Transuranium Nuclides in Radioactive Waste=Analysis of Am and Pu in Contaminated Metal Surface=

Pipe waste sampled from Fugen(before dissolution)

Dissolved solution of thecontaminated surface and thedecontaminated pipe

Dissolution of thecontaminated surface

Fig. 10-2 Dissolution of the contaminated surface of pipe waste sampled from Fugen
Contaminants on the surface were dissolved by soaking in an HNO3/HCl solution for 5 h.

120

Conditioning
20 mℓ 2 M HNO3
-0.3 M ascorbic acid
Waste
Waste

Waste

TRU
resin
5 mℓ

Rinse 20 mℓ 2 M HNO3
- 0.3 M ascorbic acid
Pu(III)→Pu(IV)5 mℓ 0.1 M NaNO2-1 M HNO3

80
60
40

1 g Fe(III)20

Rinse 20 mℓ 2 M HNO3Am

30 mℓ 4 M HCl

Pu

30 mℓ
0.02 M TiCl3-4 M HCl

α-spectrometry

0.5 g Fe(III)100
Am recovery (%)

Dissolved surface contamination
20 mℓ 2 M HNO3
+1 g ascorbic acidFig.10-3 Separation of Pu and Am using atransuranium resin

The radioactivity of transuranium (TRU) nuclides such as
neptunium-237; plutonium-238, 239, 240, 242 (238,239,240,242Pu);
americium-241,242m,243 (241,242m,243Am); and curium-244 shouldbe evaluated before the disposal of various types of low-level
radioactive wastes generated during the decommissioning of
nuclear facilities.
In this study, an analytical method was developed forevaluating the amount of Pu and Am attached to the surfaceof the pipe sampled from the prototype advanced thermal
reactor Fugen. The pipe surface was in contact with coolant
water and thus contaminated by radionuclides. Pu and Am must
be separated from the main components of the pipe and preparedin the form of a thin film in order to measure the α-rays fromthe infinitesimal quantity of Pu and Am in the contaminatedspecimen.
First, the contaminated pipe surface was dissolved by soaking
it in an acid solution comprised of 12 M hydrochloric acid(HCl) and 13 M nitric acid (HNO3) (HCl/ HNO3/ H2O = 1:1:4)
for 5 h (Fig.10-2). Next, Pu and Am were rapidly separated
from the main components of the pipe via extraction
chromatographic separation using a TRU resin. The TRU
resin selectively extracts tri-, tetra-, and hexavalent actinides(An) from HNO3 solutions, but elutes An(III) from HCl
solutions. Therefore, the dissolved solution was evaporated todryness, and the residue was dissolved in 2 M HNO3 in order

0

0

0.5

1

1.5

2

2.5

Added amount of ascorbic acid for reaction equivalentFig.10-4 Effect of the addition of ascorbic acid on Am
recovery from a solution containing Fe(III)Trivalent iron was reduced to Fe(II) by using ascorbic acid, andapproximately 100% of Am was recovered by the addition of
1.3 equivalents of ascorbic acid.

to load the TRU resin. The TRU resin extracted Pu and Am
and eluted the main components of the pipe. Pu was thenoxidized to Pu(IV), which remained on the resin, and Am
was recovered in an HCl solution. Next, Pu(IV) was reduced
to Pu(III) using TiCl3 and recovered in an HCl solution(Fig.10-3).
In this scheme, the trivalent iron (Fe(III)) in the sample
solution must be reduced to Fe(II) because a large amount of
Fe(III) will interfere with the extraction of Am(III). Thus, westudied the effect of the addition of ascorbic acid on Am
recovery from 20 mof a 2 M HNO3 solution containing 0.5
or 1 g of Fe(III) (prepared using Fe(NO3)36H2O). Approximately
100% Am recovery was achieved with the addition of 1.3
equivalents of ascorbic acid (Fig.10-4).
Finally, the Pu and Am attached to the pipe (sixspecimens) sampled from Fugen were analyzed using theabove-described method. In one of the specimens, 1.6 × 10-5
Bq/g of 239+240Pu was detected, but the other isotopes were
present in quantities lower than the detection limit (< 2.8 ×
10-6 Bq/g of 238Pu, < 3.0 × 10-6 Bq/g of 239+240Pu, and < 2.9 ×
10-6 Bq/g of 241Am). In addition, a high recovery for Pu andAm (94%〜101% and 85%〜93%, respectively) was
achieved in the experiments using samples containing known
amounts of radioisotopes.

Reference
Shimada, A. et al., Analysis of Th, U, Pu, and Am in Radioactive Metal Waste using Extraction Chromatography, Journal of Radioanalytical Nuclear
Chemistry, vol.286, no.3, 2010, p.765-770.
JAEA R&D Review 201299

Development of Decommissioning and Radwaste Treatment Technology

10-2 Prevention of Heavy Metal Elution from Solidified
Forms of Incinerated Ash and Cement

Penetration ofgroundwater

Chromium

Leaching concentration of Cr(VI) (mg/ℓ)=Confinement of Heavy Metals Using Insolubilizers=

Soil andwater source

Lead
Cadmium
Solidified
cement form

Fig.10-5 Confinement of heavy metals in the solidified form of
cement
The hazardous heavy metals present in minor amounts in incinerated
ash must be confined into a solidified form of ash with cement, because
these metals may be eluted from the solidified form when it comes incontact with groundwater.

We incinerate combustible radioactive wastes, and plan tosolidify the obtained incinerated ash with cement and bury itin the ground. Cement solidification has several advantages
such as the low price of cement and resistance of cement todeterioration caused by radiation. However, when thesolidified form comes in contact with water, a small amountof the form may be dissolved in it.
Incinerated ash contains a minor quantity of hazardous
heavy metals such as Pb, Cd, and Cr(VI), which are includedin plastics and paints (which are combustible wastes). These
heavy metals must be confined in the solidified cement formof incinerated ash, because they may be eluted when thesolidified form comes in contact with groundwater (Fig.10-5).
We investigated the elution behavior of these metals byusing a solidified cement form. A mock form was preparedby mixing 1wt% each of PbCl2, CdCl2, and Na2CrO4 withincinerated ash, cement, and water. The mock form wassoaked in a large amount of water, and the metal
concentrations in the water were determined. Target values
for the leaching concentrations of the heavy metals were2
1.6

Target value

1.5

1

0.82
0.56

0.5

0
No additive

Iron sulfate

Sodium sulfide

Insolubilizer

Fig.10-6 Leaching concentration of Cr(VI)Iron sulfate and sodium sulfide insolubilizers show thepotential to maintain the leaching concentration of
Cr(VI) below the target value.

determined on the basis of the Waste Management and Public
Cleaning Law. The values for both Pb and Cd were found tobe 0.3 mg/and that for Cr(VI) was 1.5 mg/.
The leaching concentration of Pb was less than 0.1 mg/and that of Cd was less than the detection limit (0.01 mg/),
both of which were less than their target values. It was thus
found that Pb and Cd can be confined in the solidified cement
form. Because solidified cement forms are alkaline, the Pb
and Cd ions form hydroxide complexes, which are hard todissolve in water.
However, the leaching concentration of Cr(VI) was greaterthan its target value, and Cr(VI) was difficult to confine in thesolidified cement form. It is known that Cr(VI) is a strongoxidizer and can be easily reduced to Cr(III) by using
reductants. Therefore, we prepared solidified cement forms
containing iron sulfate or sodium sulfide insolubilizers,
which are reductants, and conducted leaching tests using
these additives. As a result, the leaching concentration of
Cr(VI) was less than its target value (Fig.10-6).

Reference
Meguro, Y., Nakayama, T. et al., Elution Behavior of Heavy Metals from Cement Solidified Products of Incinerated Ash Waste, Proceedings of the 14th
International Conference on Environmental Remediation and Radioactive Waste Management (ICEM2011), Reims, France, 2011, ICEM2011-59102, 5p.,
in CD-ROM.

100 JAEA R&D Review 2012

1111  
Role and Potential of Nuclear Computational Science

Nuclear Fusion

Nuclear Reactor

High-intensity Proton Accelerator
dislocation

Extension to Project Research
Nuclear Fusion
Nuclear Reactor
High-intensity Proton Accelerator

Seismic Analysis

S

P

Aged Material Deterioration

R&D for Computational Science
Seismic Analysis, Nuclear Material
Property Analysis, Prediction, etc.

R&D for Computer Science
Large-scale Parallel Computing, Visualization,
Large data processing, etc.

Radioactive Substance
Absorption

Operation & MaintenanceSupercomputer, Network

Visualization
Radiation Dose Database࡮
Distribution Map
Enhancement and Reinforcement of University-Industry Research Collaboration
Promotion of the Combination of Facilities and Equipment

Fig.11-1 Development of nuclear computational science research in JAEA
Nuclear computational science research is utilized not only for research projects in JAEA but also for the enhancement andreinforcement of university-industry research collaboration. In addition, we are promoting its application to research and development
for Fukushima support.

Computational science is the third research methodology,
following theory and experiment, and has maderemarkable progress by becoming a field that is indispensablefor the elucidation and prediction of phenomena difficult toobserve or replicate in experiments. The next-generation
supercomputer K was developed as a national project andhas begun to be used. A change in the environment
surrounding computational science has allowed its growth toaccelerate.
We study important issues in the nuclear field, including
those related to earthquake-resistance strength and deteriorationof aged material, which cannot be sufficiently investigatedwithout making full use of advanced computational
technology, as shown in Fig.11-1. In addition, we are
involved in various research and development activities andwe develop and maintain the super-large-scale parallel
computational technology required as a generic technology
for such research. We ensure that the results of this researchsupport broad research themes, such as fast reactor or nuclear
fusion research, and allow for further development in this field.
The elucidation of the mechanism and the prediction of thedeterioration of nuclear reactor structure materials areimportant problems faced in the research on aged material
deterioration. Therefore, it is necessary to predict the
segregation of atomic impurities that lower the material
strength. In Topic 11-1, we introduce an example study that
evaluates the segregation quantity at the grain boundary ofphosphorus in irradiated materials by solving the diffusion
rate equation using the coefficient obtained by the firstprinciples calculation.

In the research cooperation with the Quantum Beam
Science Directorate, it is an important problem to analyze thedata for the neutron dispersion strength of an iron-based hightemperature superconductor observed in J-PARC and to
elucidate the superconducting structure. We introduce a
research example of a virtual experiment on a scale that only
a supercomputer can handle using the modeling by the firstprinciples calculation in Topic 11-2.
In the earthquake-resistance research for nuclear facilities,
improving the precision to simulate the dynamic behavior ofan extra-large earthquake is an important problem. First, it isimportant to model the elastic-plastic behavior of aconnection and the neighborhood structure expected toachieve the plastic state. In Topic 11-3, we introduce one of
the research examples of the elastic-plastic modeling of theconnections between a nuclear building and the equipment.
In the Fukushima support, we build a database that canmanage the radiation dose measurement results and a
distribution map system that can confirm them on a map
based on the accident at the Tokyo Electric Power Company,
Incorporated Fukushima Daiichi Nuclear Power Station, in
cooperation with the Fukushima Environmental Safety
Center, and make this available to the public. In Topic 1-3,
we summarize these systems.
In future, we will continue to work hard on nuclear
research and development utilizing advanced computational
science and generic computer technology. We will also
deepen our cooperation with organizations inside and outside
JAEA as we progress in the field of nuclear computational
science research.
JAEA R&D Review 2012101

Computational Science and E-Systems Research

11-1 How Does Impurity Diffuse in Irradiated Materials?

=Evaluation of Impurity Segregation by Diffusion Rate Equations
Based on the Atomic Migration Mechanism=
2

16

2

Case excluding C：k ＝5.8×10 /m2
162
Case including C ：k ＝2.0×10 /m
experimental data ٤ ٨  ٌ ً ‫ ڎ ع‬ٟ

C affects both vacancy and SID.
C affects only vacancy.
C affects only SID.
experimental data ٤ ٨  ٌ ً ‫ ڎ ع‬ٟDose (dpa)
0.05

0.10

0.15

0.20

0.25

15

10

ٌ
ً
‫ڎ‬

5

٨

٨

٤
ٟ‫ع‬

0
0.0

2.8e+23 5.7e+23 8.5e+23 1.1e+24
Neutron fluence (n/m2)

0.00
30
GB P Segregation (at%)

GB P Segregation (at%)

0.00
20

Dose (dpa)
0.10
0.15

0.05

0.20

0.25

25
20
15
10
ً
‫ڎ‬

5 ٟ‫ع‬
00.0

٤

2.8e+23ٌ٨

5.7e+23٨

8.5e+23

1.1e+24
2

Neutron fluence (n/m )Fig.11-2 GB P segregation vs. dose
The simulation conditions are as follows. P content: 0.013at%,
temperature: 290 ℃, and dose rate: 1.8 × 10-10 dpa/s. The
curve of the case including C reproduces the experiments witha sink strength (k2) lower than that of the case excluding C.

Fig.11-3 Influence of C on GB P segregation
The results for the case where C affects only V are almost thesame as those for the case in which C affects both V and SID.
Thus, C affects GB P segregation primarily through its effects
on V.

Several impurity elements at grain boundaries (GBs) insteel cause degradation of the toughness of steel by
weakening the cohesive strength of the GBs. Thisphenomenon is known as GB embrittlement. Because
irradiation accelerates GB impurity segregation, the evaluationof this segregation in the structural materials of nuclear
reactors is an important avenue of research.
Most impurity elements are stable in steel because they arelocated on crystal lattice points instead of iron atoms.
However, when a vacancy (V, a lattice point without any
atoms) or self-interstitial dumbbell (SID, formed by two ironatoms that are not on a lattice point) comes to the next lattice
point of an impurity atom, the impurity atom can migrate to
GBs owing to its interaction with V or SID. As a result, GB
impurity segregation is accelerated in irradiated steel, where
many Vs and SIDs are generated by the displacement of ironatoms.
We estimated GB segregation of phosphorus (P), whichcauses GB embrittlement of nuclear reactor pressure vessels,by simultaneously solving diffusion rate equations for V,
SID, and P. V and SID decrease when the sink strength in theequations of V and SID is large, such that GB P segregation

is suppressed. While empirical values have been used for the
diffusion speeds of V, SID, and P in previous studies, we
adopted values that were theoretically derived from several
energies and evaluated by atomistic calculation; the energies
investigated included the migration energies of V and SID,
and the association and dissociation energies between P and
V and between P and SID. In addition, because carbon atoms
are inevitably included in steel, we estimated the interactions
between a V and a C atom and between a SID and a C atom
by atomistic calculation and incorporated the effect of C on
suppression of diffusion of V and SID into the equations. In
the case including C in Fig.11-2, the experimental results
were reproduced with a sink strength lower than that in thecase excluding C. Thus, we can conclude that C suppresses
GB P segregation. We can also conclude that C suppresses GB
P segregation primarily through its effects on V (Fig.11-3).
Our method can be applied to GB P segregation of various
irradiation conditions and to GB segregation of other diverse
impure elements.
This study was sponsored by the Japan Nuclear Energy
Safety Organization (JNES).

Reference
Ebihara, K. et al., Effect of Carbon on Irradiation-Induced Grain-Boundary Phosphorus Segregation in Reactor Pressure Vessel Steels using First-PrinciplesBased Rate Theory Model, Journal of Nuclear Materials, vol.414, issue 2, 2011, p.328-335.

102 JAEA R&D Review 2012

Computational Science and E-Systems Research

11-2 J-PARC and Supercomputer, a Collaborative Challenge to
Solve the Mechanism of Superconductivity:=A Quest for Electron Pairing through Computer Simulation of Neutron Scattering in Iron-Based Superconductors=

N

N
NeutronS

S

Table 11 - 1 Comparison between the intensities ofexperiments and theoretical values for BaFe2(As, P)2
Two electrons form a pair in superconductors because of
attractive forces; different attractive forces result in different typesof electron pair. We have shown that the pair in case III isconsistent with experimental results.

Property

Intensity

Experiment at
J-PARC

Relatively lower intensity thanthat in other iron-based
superconductors.

㨪1.7

Case I (S±-wave)

Pairs originated from magnetic
forces㨪2.1

Case II (dxy-wave)

Magnetic pairs in non-iron-based
superconductors<1

Case III (S±-wave
with horizontal nodes)

Pairs originated from relatively
weaker magnetic forces

㨪1.6

N
Iron-based
S Superconductor
Fig.11-4 Schematic figure of a neutron scatteringexperiment
Neutrons can detect magnetic fields originating fromelectrons.

Superconductivity is a phenomenon that typically involves
sudden disappearance of electric resistance below thesuperconducting transition temperature. In the nuclear R&D
field, a superconducting coil is indispensable for nuclear
fusion reactors and accelerators.
A new superconducting material, an iron - based
superconductor, was discovered in 2008. This type ofmaterial exhibits a very high transition temperature and arichly varied chemical composition. Neutron scatteringexperiments for the iron-based superconductor BaFe2(As, P)2
have been conducted at the Japan Proton Accelerator
Research Complex (J-PARC) in JAEA (Fig.11-4). The resultsof these experiments indicate that the magnetism of ironatoms is important in the mechanism of superconductivity.
Further evidence is required to clarify this mechanism
because it has proven difficult to ascertain from previous
experiments how the pairing of electrons results insuperconductivity.
To address this, we attempted to identify the relevant
electron pairs in a superconductor in order to simulate
neutron scattering using a supercomputer. Previous studies
have been unable to offer a reasonable explanation owing to

insufficient accuracy; to overcome this problem, wedeveloped a method on the basis of the random phaseapproximation and constructed a model using first-principle
calculations in order to consider the electron correlations. To
reduce the computational cost, we changed the algorithm anddeveloped the hybrid parallel codes with OpenMP+MPI.
Some candidates exist for possible electron pairs, withdifferent magnetic origin, in iron-based superconductors. We
simulated neutron scattering using what we considered to begood candidates for neutron scattering experiments andcompared our results with those of previous experiments(Table 11-1). We have shown that the pairing in case III canreproduce the experimental results and that the magnetism of
the iron atom is important for the mechanism of
superconductivity. Our results can help clarify the mechanism
involved because this magnetism is also important in otheriron-based superconductor compounds.
This work is supported by the CREST Framework
Development for Multiscale and Multiphysics Simulations
toward Novel Application of Superconductivity from theJapan Science and Technology Agency (JST).

Reference
Nagai, Y. et al., Theoretical Analysis for Inelastic Neutron Scattering on BaFe2(As1-x ,Px)2 using Realistic Three-Dimensional 10-Orbital Tight-Binding
Model, Journal of the Physical Society of Japan, vol.80, Supplement B, 2011, p.SB021-1-SB021-4.
JAEA R&D Review 2012103

Computational Science and E-Systems Research

11-3 How Do Earthquake Waves Propagate into a Nuclear Facility?

=Realization of Precise Simulation of Building-Equipment Interaction during Earthquakes=
Cnnection

Equipmentside

The connection
interaction model
is setEquipment

Building
Non-linear spring arranged
distributedly
Present
The connection interaction
model is set at theconnection surface

Building
ConventionalBuilding-equipment
connection are
assumed as fixed

0
-50

B

-100
-10

-5

0

Analyticalresult510

変位(mm)
Displacement(mm)

Intermediate heatexchanger (IHX)

50

Primary pressurizedwater cooler (PPWC)

0

-50

-100
-10

-5

Experimental
results05
10

Analytical model for whole numerical simulation ofprimary cooling system of HTTR

載荷点変位(mm)Displacement (mm)Reponces are reduced

6

Acceleration (cm/s2)×104

Applicability of the model was confirmed bycomparing analytical results with experimental results
Resonance frequency was changed
during vibration

12

A
――
モデル無
モデル有
―― B

8
4

A: elastic
B: elastic-plastic

10

20

30

40

50

Frequency (Hz)

Y
Z

5
4
3
2
1
0
-1
-2
-3
-4
-5

X
Y
Z

Time (s)With the connection model

The time response of acceleration of non-linear
springs of the connection interaction model at point C

Applicability of the model was confirmed by checking theresonance frequencies

Fig.11-5 Connection interaction model between
buildings and equipment
We compared the analysis results from the suggested
connection interaction model with experimental results
and confirmed the applicability of the suggested modelto the real connection structure.

X

Time (s)Without the connection model

0
0

5
4
3
2
1
0
-1
-2
-3
-4
-5

Acceleration (cm/s2)×104

Force (kN)A

荷重（ ｋN

50

Fourier Amplitude ×10

Secondarypressurizedwater cooler(SPWC)

100

100

Force (kN)C

Fig.11- 6 Real entire -plant analysis with connection interaction model
We incorporated the connection interaction model into the real plantmodel and conducted earthquake response analysis. The results
indicate that the equipment response was reduced by absorption ofenergy as a result of the nonlinearization imparted by connection; thus,
we were able to confirm the effects of adopting the connection
interaction model.

Nuclear facilities are composed of a large number of
equipment, and it is necessary to evaluate realistic responsesof each of these equipment to earthquakes in order toascertain earthquake resistance safety. In particular, it isimportant to know which types of wave propagate into thebuilding through the connection; in other words, it isimportant to know the detailed wave propagation propertiesof the connection (wave transmission and reflection), because
equipment response is evaluated using building response.
Because the connection between buildings and equipment is aconcrete structure, it is usually assumed to be a fixedboundary to simplify the elucidation of its wave propagation
properties. However, recent experimental results havesuggested that a slight displacement or gap can occur at theconnection for large inputs, and it has become clear that theassumption of a fixed boundary is not appropriate in such
instances.
Therefore, we have attempted to express wave propagation
properties at the connection to model force-displacement
relationships based on experimental data. We devised forcedisplacement relationships for each micropart directly byarranging nonlinear springs dispersively on the boundary
surface in order to consider historical properties such as thepartial gap at the connection. We confirmed the applicabilityof the connection interaction model by comparing ouranalytical results with experimental results and confirmed theresonance frequencies before and after the nonlinearization of

the connection, as shown in Fig.11-5.
Furthermore, we conducted earthquake response analysisof the entire cooling structure of a nuclear facility
incorporating the connection interaction model using a large
earthquake input, for which part of the building-equipment
connection becomes nonlinearized, to confirm the applicabilityof the connection interaction model to the real plant. Fig.11-6
shows a model of the entire cooling structure of the hightemperature engineering test reactor (HTTR) and our analysis
results. We were able to confirm that the equipment response
was reduced because of historical energy absorption by
nonlinearization of the connection in the case with theconnection interaction model. On the basis of these results,
we were able to assess the potential for use of the connection
interaction model in earthquake resistance evaluation.
It is expected that more detailed analysis will be necessaryfor evaluation of the earthquake resistance of nuclear plants.
Furthermore, we expect that as the representation of the
behavior of nuclear facilities during earthquakes improves,
we will be able to share more detailed information with thegeneral public.
The present study is the result of the project entitled
Development of physical model describing the dynamic
interaction characteristics of component connections for the
analysis of an entire nuclear facility entrusted to the Japan
Atomic Energy Agency by the Ministry of Education,
Culture, Sports, Science and Technology of Japan (MEXT).

Reference
Nishida, A. et al., Elastic-Plastic Connection Model Describing Dynamic Interactions of Component Connections, Progress in Nuclear Science andTechnology, vol.2, 2011, p.576-581.

104 JAEA R&D Review 2012

1212    
Technology and Human Resource Development in the Area of Nuclear Nonproliferation
and Nuclear Security to Support Peaceful Use of Nuclear Energy
Nuclear Nonproliferation Technology Development
for Japanese and International Applications
We have been developing proliferation-resistant nuclear
technologies, methodologies for evaluating proliferation
resistance, and advanced safeguard technologies throughcooperation with international partners such as the U.S.
Department of Energy (DOE). In the field of environmental
sample analysis for safeguards, we have been providing
technical support to the International Atomic Energy Agency(IAEA) as a part of the IAEA Network Labs.

Support for Government Policy Formulation Basedon Our Expertise
As a think tank in this area, we conduct policy research,
including research on the issues to be addressed if the JapanU.S. Nuclear Cooperation Agreement is amended.
Strict Management of Nuclear Material at Our Own
Facilities and Utilization of the Experience Gained
from Management of Nuclear Material
We have had strict management of our nuclear material.
Moreover, we assist in the streamlining of inspections andproviding technical support to the Japanese Government andthe IAEA. We also provide support to the IAEA in the fieldof physical protection, and respond appropriately in case ofrevisions to Japanese legislation in this area.

Development of Measures to Accounting for andControl of Nuclear Material in Response to theaccident at the Tokyo Electric Power Company,
Incorporated Fukushima Daiichi Nuclear Power
Station (1F)
We have initiated an investigation of the measures required
to account for and control nuclear material that remains
within the core of the reactors, because conventional
measures of accounting and control cannot be applied to such
nuclear material.

New Efforts on Nuclear Security
On the basis of commitments of 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
carrying out its mission of capacity building and infrastructuredevelopment, focusing on the Asian region.
Approximately 600 participants (approximately 280
participants mainly from Asian states) participated in the
training courses and other activities organized by the ISCN in
FY 2011. We are proud of the contributions that we aremaking in the area of capacity building
Furthermore, in cooperation with the United States, wehave started developing technology for the measurement of
nuclear material, which will contribute to the advancement ofaccounting for and control of nuclear material, as well as fornuclear detection and forensics technology utilizing cooperation
with the United States. We will continue to provide support to
the Japanese government in the area of international
contributions by establishing an accurate technology by theend of 2013 and sharing such technology with theinternational community.

Contributions to the International Community Basedon Our Expertise and Experience
With respect to activities relating to the Comprehensive
Nuclear-Test-Ban Treaty (CTBT), we have operated radionuclide
monitoring stations and established a national data center forthe analysis and evaluation of data collected at worldwide
radionuclide monitoring stations, and thus have contributed
to the establishment of an international monitoring system forthe detection of nuclear tests.
We have also played a significant role in increasing theunderstanding of the atmospheric diffusion of radioactivity
caused by the accident at 1F by using the network of CTBT
radionuclide monitoring stations.

EURATOM

IAEA

Cooperation
Development of Technology for
Nuclear Nonproliferation

DOE
Domestic
organizations

Proliferation-resistant nuclear technology
and methodology for evaluating proliferation
resistance
Advanced safeguard technologies

Support for Capacity Building andInfrastructure Development in Asian States

Policy Research
Policy research on the current issues
regarding peaceful use of nuclear energy
and nuclear nonproliferation

Enhancement of capacity building throughtraining and education
Support for infrastructure development,
including the development of legal systems,
regulations, and manuals
Domestic Human Resource

Development
Management/Protectionof Nuclear Material

Cooperative relationships with
universities and related organizations
Support nationwide efforts for humancultivation throughout Japan

Effective and efficient management
Physical protection of nuclear materials
and nuclear facilities
Cooperation/support to IAEA, etc.

Development of technologies
for the measurement, detection, andforensics of nuclear material

Demonstration of an NDA system, developmentof nuclear resonance fluorescence NDA technology,
development of He3-alternative neutron detection
Establishment and improvement of nuclear
International Contribution
forensics technologies and creation of a
Advanced development of CTBT nuclear
domestic database
test verification technology and establishmentof an international monitoring system
Technical support for the disposition of Pu
from dismantled Russian nuclear weapons

Relevant
ministries

Cooperation
CTBT : Comprehensive Nuclear-Test-Ban Treaty
WINS : World Institute for Nuclear Security
ASNO : Australian Safeguards and Non-proliferation
DOE : Department of Energy
EURATOM: European Atomic Energy Community

ASNO

WINS

Fig.12-1 JAEA activities for the development of science and technology for nuclear nonproliferation
We have two primary missions: to support the government in developing nuclear nonproliferation policies through research, and tosupport the government and international organizations by developing nuclear nonproliferation technology. In addition, we
appropriately implement the management of nuclear material and the development of related technology, and conduct capacitybuilding activities.
JAEA R&D Review 2012

105

13
13     
Tsuruga Head Office
Following the accident at the Tokyo Electric Power
Company, Incorporated Fukushima Daiichi Nuclear Power
Station (1F) on March 11, 2011, emergency safety measures
have been implemented at both the MONJU and FUGEN
Power Stations.
At MONJU , there have been no confirmed effects on thereactor vessel or the core internals after the replacement of
the in-vessel transfer machine (IVTM) and its function checkfor the recovery work of the IVTM falling accident that
occurred in August 2010. The suspended 40% power
confirmation test, which is the second stage of the three-stage
system start-up tests, awaits the government policy decision
giving the green light to restart, while taking utmost care to
maintain the quality level of the facilities and equipment.
At FUGEN , the decommissioning and its related
technical development have been steadily conducted. The
implementation of decontamination facilities has beenlaunched on a trial basis, in particular, to treat low-level
radioactive waste as general waste (Topic 13-1).
Also, the technical collaboration among industry, government,
and academia, including research and development projects
in the field of laser development and laser applications, hasbeen actively promoted.
Tokai Research and Development Center, Nuclear
Science Research Institute (NSRI)
The NSRI facilities, such as research reactors (JRR-3, JRR- 4,
and NSRR), accelerators (Tandem etc.), critical assemblies(STACY, FCA, etc.), and hot laboratories (WASTEF,
BECKY, etc.) also suffered damage during the Great East
Japan Earthquake. Because of the quick recovery of thedamaged lifelines, research activities at NSRI were able tobegin supporting the restoration from the accident at 1F froman early stage.
To continue the development of technology, the ring-type
dosimeter, which has advantages of long-term stability andrepeatability in the readout process (Topic 13-2), wasdeveloped. In addition, a handy-type dosimeter was producedby manufacturers, employing electronics technologies thatwere developed at NSRI. Investigations of the practical(rational) and efficient disposal of radioactive waste havebeen carried out, and the results were applied to the system toclassify the very-low-dose concrete generated from JRR-3
modification construction work as industrial waste (Topic 13-3).
Tokai Research and Development Center, Nuclear
Fuel Cycle Engineering Laboratories
At the Plutonium Fuel Development Center, physicalproperties of MOX fuels were studied, and the test for the
preparation of standard plutonium materials was executed asa part of technological cooperation with the Japan Nuclear
Fuel Limited (JNFL).
At the Reprocessing Technology Development Center, we
carried out seismic reinforcement of the Tokai Reprocessing
Plant and technological cooperation with the JNFL for theoperation of the Rokkasho Reprocessing Plant. In addition,
safety measures were executed to prevent against the loss ofthe cooling functions.
We also performed various other R&D activities, such as
basic studies to maintain FBR cycle technologies (Topic 13-4)
and tests to enhance the reliability of geological disposal
technologies. In addition, we developed a basic design for theconstruction of an incineration facility for radioactive waste
treatment.
Furthermore, as a part of the activities for thedecommissioning of the damaged Fukushima reactors, themimetic molten debris were fabricated by mixing zircaloy
and uranium oxide, and these were then investigated.
106 JAEA R&D Review 2012

Securing cooling function

Measures for tsunami

࡮Auxiliary cooling seawater‫ޓ‬pump backup motor‫(ޓ‬By the end of March 2013)

࡮Reinforcement of floodwall‫ޓ‬to protect pump
࡮Additional sealing to prevent‫ޓ‬seawater intrusion into the‫ޓ‬building࡮Auxiliary cooling seawater‫ޓ‬pump Alternative pump and‫ޓ‬motor

Securing power source࡮Power-supply car
࡮Power connecting boards
࡮Emergency diesel generator‫(ޓ‬auxiliary air-cooled power‫ޓ‬supply equipment)‫(ޓ‬By the end of March 2013): Underworking

Reactorbuilding
Dieselbuilding

T.P.
+42.8 mT.P.+21 m
T.P.+5.2 m 6.4 m▽

T.P.+0 m ▽IntakeBreakwater

T.P.-6.5 m

Curtain wall

Air cooler
outlet valve

Reactorauxiliarybuilding

Screen pump room

Securing cooling function
࡮Modularized thermal insulator for four valves‫ޓ‬to make detachment easier at SBO

MONJU Safety Measures

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 Institute (NSRI)

Plutonium Fuel Development Facility

The mimetic molten debris (right hand) fabricated by theapparatus for measuring melting points (left hand)

Mimetic molten debris fabricated at the Plutonium Fuel Development Facility

Development of Experimental Techniques / Facilities at JAEA R&D Centers

J-PARC Center
The extensive damage to facilities and infrastructurecaused by the Great East Japan Earthquake on the 11th ofMarch 2011 led to the shutting down of the J-PARC.
However, because of the tremendous effort of the competent
J-PARC staff, the Linac accelerator was restarted on the 9thof December, 2011, and the neutron beam production wasverified at the Materials and Life Science Experimental
Facility (MLF) on the 22nd of December. After performing
various commissioning operations, user programs in both the
MLF and neutrino experimental facilities resumed on the
24th of January, 2012. The supplies of proton beam
intensities to those facilities were recovered to their preearthquake values of 210 kW and 180 kW, respectively. Since
January 2012, the number of J-PARC users also quickly
recovered to more than 2500 persons per month, which was
the same as the number before the earthquake. Many exciting
results were reported from all experimental facilities,
including the neutron structure analysis of a single protein
crystal and a technical development for visualizing themagnetic field in matter taking place at the MLF.

Oarai Research and Development Center
Restoration work and safety measures were performed to
the facilities and equipment damaged by the Great East Japan
Earthquake.
Immersion tests of materials in sodium chloride solution
and cesium decontamination trial tests, etc. were carried out
as part of the R&D into the decommissioning of 1F.
In the Japan Materials Testing Reactor (JMTR), facilities
were inspected aiming at re-operation, and various irradiation
facilities and equipment and a simulator for the irradiation
test reactors were fabricated by the Leading-edge Research
Promotion Fund, which was provided by the Ministry of
Education, Culture, Sports, Science and Technology (MEXT)
in Japan.
In the Experimental Fast Reactor JOYO, there wasprogress in the design of pieces of equipment used in the
repair work. The dummy for a mock-up test using theseequipment was also fabricated.
In the High Temperature Engineering Test Reactor(HTTR), facilities were inspected with the aim of resuming
operation. The verification runs were conducted in the coldstate, and a comprehensive evaluation of the integrity of theequipment was also carried out.
The construction of the Advanced Technology Experiment
Sodium (Na) Facility (AtheNa) to carry out sodium tests, etc.,
in the FaCT project was also completed.
Naka Fusion Institute
The Naka Fusion Institute conducts research anddevelopment toward the practical use of fusion energy.
Currently, it is mainly involved 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 the research supporting andcomplementing ITER as the Broader Approach activities incooperation with the European Union.
In 2011, the Japan Atomic Energy Agency (JAEA) brought
forward the planned production of the superconducting
conductor of the ITER. In the disassembling operation of
JT-60, the disassembly and removal of TF coils as the major
components of JT-60 was continued, and the removal of all
18 coils was completed by the end of the fiscal year.
The component production of JT-60SA was carried out.
Three 40-degree sectors of the vacuum vessel for the actualmachine were delivered and completed with the continuationof the production of the superconducting conductor for themagnetic field coil of the actual machine.

Staff members celebrating the resumption of proton beam
delivery to the Materials and Life Science Experimental Facility(December 22, 2011)

Training using a simulator for irradiation test reactors

Disassembly operation of TF coils in JT-60 (December 2011)
JAEA R&D Review 2012107

Development of Experimental Techniques / Facilities at JAEA R&D Centers
Takasaki Advanced Radiation Research InstituteTakasaki Ion Accelerators for Advanced Radiation
Application (TIARA), which consists of four ion accelerators,
the electron accelerator, and the gamma irradiation facilities,
are made available to researchers in the JAEA and otherorganizations for R&D activities regarding new functionaland environmentally friendly materials, biotechnology, theradiation effects of materials, 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, and threedimensional in-air PIXE analysis, and three-dimensional
microbeam writing technology with proton beams of different
energies at the electrostatic accelerators. In 2011, a techniqueused to measure the transverse intensity distribution of a
uniform wide-area beam by a radiochromic dosimetry film
was developed for 520 MeV Ar and 490 MeV Xe.

Two-dimensional Gaussian-like intensity distribution of a 520 MeV
Ar beam measured with a radiochromic film

Kansai Photon Science InstituteIn the Kizu District, we are in the process of improving
laser facilities, e.g., the improvement in the quality of highintensity lasers. With regards to high-intensity short-pulse
lasers, we successfully reduced the pulse width and improved
the laser wavefront. We achieved an intensity two times thatobtained previously (1021 W/cm2). Also, in the Consortiumfor Photon Science and Technology, we are in the process ofdeveloping ultra wideband light sources ranging fromterahertz-x-rays to quantum beams, as a management agency.
In the Harima District, we have been developing andimproving a state-of-the-art analysis technique for thefunctional expression mechanism and the reaction mechanismof materials using JAEA synchrotron radiation beamlines atSPring-8. They are also applied to studies related tonanotechnology, energy, and the environment and are used
for external applications. In the JAEA soft X-ray beamline,
we successfully developed an ultrahigh-precision, elementselective measurement technique for the magnetization ofnanoscale thin films.

JAEA Kansai Advanced Relativistic ENgineering (J-KAREN)
laser system

Horonobe Underground Research CenterWe are currently constructing an underground researchlaboratory for carrying out investigations involving Geoscientificresearch and Research and development of geologicaldisposal technologies, focusing on sedimentary rocks.
In the fiscal year 2011, both the east and ventilation shafts
were excavated from depths ranging from 250 m to 350 m indepth, and the west shaft was excavated from the groundlevel (GL) down to a depth of 50 m. The construction of the
250 m drift has been completed, and the construction of the
350 m drift has begun.
In geoscientific research, we have continuously carried out
the development of techniques for the investigation of thegeological environment, for the long-term monitoring of thegeological environment, and for the development ofengineering techniques for use in the deep undergroundenvironment and studies regarding the long-term stability of
the geological environment.
In our R&D into geological disposal technologies, weinvestigated the degree of influence of low-alkaline cement
materials, including HFSC (Highly Fly-ash containing Silicafume Cement) in the surrounding groundwater and rockmass. Also, the use of the materials in spraying techniques
was tested in situ during the construction of the 250 m
Gallery.
108 JAEA R&D Review 2012

Relative vertical position (mm)Fluence low

high

150

100

50

0
0

50

100

150

200

Relative horizontal position (mm)Spraying test of low-alkaline cement in the 250 m Gallery off
the west shaft

Development of Experimental Techniques / Facilities at JAEA R&D Centers
Tono Geoscience Center (TGC)
The TGC s task is to provide scientific and technicalfoundations for the development of a safe technique for thegeological disposal of high-level radioactive waste. This
involves research into the long-term stability of the geologicalenvironment and the development and improvement oftechniques for the characterization of the deep geologicalenvironment, and a wide range of engineering techniques fordeep underground applications at an underground researchlaboratory in crystalline rock, referred to as the Mizunami
Underground Research Laboratory (MIU). The Main Shaft
was excavated from 481.3 m to 500.4 m below GL, and theVentilation Shaft was excavated from 497.7 m to 500.2 m
below GL (meters below ground level). Then, approximately
5-m-long horizontal tunnels were excavated at the GL -500 m
stage (Topic 13-12). Research and development activities,
such as geological mapping during excavation and borehole
investigations, were also carried out.
Ningyo-toge Environmental Engineering CenterWe are currently developing an engineering approach forthe decommissioning of a uranium refinement and conversion
plant (URCP) facility and a uranium enrichment plant facility
in the Ningyo-toge Environmental Engineering Center. The
URCP facility decommissioning project is the first trial for
any commercial-scale nuclear fuel facility in Japan. We
began the decommissioning project of the URCP facility in
FY 2008 and have since completed the dismantling of themain (large size) equipment. In the process, we carried out
the application tests by using the commercially available
decontamination reagent for economic purposes andconvenience. Also, we obtained successful results for metal
surface decontamination, and it has satisfied the standard value＊.
From now on, the application to the decontamination oflarge-size equipment will be considered, including 1F.
＊the standard value which carried out from a radiation control area: α-ray 0.4
Bq/cm2, βγ-ray 4.0 Bq/ cm2

Aomori Research and Development CenterIn the Rokkasho district, the use of radioisotopes in the
DEMO R&D building was permitted by the relevant
authorities as one of three research facilities of the Broader
Approach Activities in the Field of Fusion Energy Research.
The supercomputer (made by Bull, France) was installed in
the Fusion Computational Simulation Centre (CSC), and thesupercomputer had been in service to the users from April 2012.
In the Mutsu district, we continuously carried out aresearch survey for a reasonable/economical large-assemblydismantling methodology for the operation of a waste
disposal plant for research facilities, the decommissioning of
the nuclear-powered ship Mutsu (such as a survey of thecontaminant contained), and the development of a technology
for the analysis of ultra-trace elements using accelerator mass
spectrometry.

View of the intersection of the Main Shaft and the horizontal
tunnel at the GL -500 m stage

This photograph shows the state of decontamination of a tankobtained using a commercially available decontamination
reagent. The commercially available decontamination reagent,
which became a film, is peeled off in the figure.

Exterior of the supercomputer at the CSC
JAEA R&D Review 2012109

Tsuruga Head Office ■Development of Experimental Techniques / Facilities at JAEA R&D Centers13-1 Reuse and/or Recycle of Dismantled Objects=Wet Blast Decontamination Technology=

Radioactive density Co-60 (Bq/g)

Main steam pipe
Drain pipe of moisture separator
Inner pipe of condenser
100
10
1

Clearance level

0.1
0.01
0.001
0.0001
0

5

10

15

20

25

30

Decontamination duration (s)Fig.13-1 Wet blast decontamination device (manual type)The system is composed of a decontamination booth, a blast
material sucker, a wastewater treatment system, an exhauster,and an air compressor. Decontamination is to be done by ablast gun in the booth.

Blasting grit‫ޓޓޓޓޓ‬: High-hardness stainless-steel grid
Diameter of grid‫ޓޓޓ‬: 50 㨪 200 µm
Density of grid‫ޓޓޓޓ‬: 12 vol%Blasting nozzle‫ޓޓޓޓ‬: Round type (Ǟ11 mm)
Blasting Pressure‫ޓޓޓ‬: 0.4 MPa
Blasting standoff/angle: 100 mm/90°
Blasting duration‫ޓޓޓ‬: 3 sec for one treatment‫ޓޓޓޓޓޓޓޓޓޓޓ‬on surface area (20 cm2) of specimen

Fig.13-2 An example of decontamination performance(carbon steel used at FUGEN )For contaminated pipes used in turbine facilities at FUGEN ,the system can reduce contamination to less than theclearance level using one treatment of 3 s processing time.

The Fugen Decommissioning Engineering Center ( FUGEN )is now decommissioning the advanced thermal reactor
facilities. To do so, radioactive waste that has extremely low
levels of radioactivity generated during decommissioning
must be treated so that the waste can be either reused and/or
recycled as clearance resources or disposed of as industrial
waste.
Turbine facilities are now being dismantled and removed at
FUGEN , and we are preparing to apply a decontamination
process to those metals that are dismantled. To enter into
clearance, technology is required to remove radioactive
substances from metal surfaces, thereby reducing theconcentration of residual radioactive substances to acceptablelevels.
We developed a decontamination technology that provides
the necessary containment of dust created by decontamination.
The technology is a manually operated wet blast method(Fig.13-1) that removes radioactive substances adhering tometal surfaces. The blast material is water, pressured to a few
atmospheres at droplet sizes of some tens to hundreds ofmicrons.
Before beginning the decontamination work, tests wereconducted to confirm that the required level of decontamination
could be achieved. Using the specimens collected in thedismantled materials at FUGEN , we were able to establish
values for process parameters that enable the technology tosuccessfully reduce radioactivity to the required levels.
By removing rust and paint, residual radioactivity on
carbon and stainless steel can be reduced to levels below theclearance level in a short period (Fig.13-2), and decontamination
factors (the ratio of residual radioactivity before and after
decontamination) of not less than 100 can be obtained for anyof the materials. Liquid waste generated by decontaminationis treated in an existing liquid waste treatment facility. Thisfacility filters blast material from the wastewater so that no
dedicated disposal device is needed.
This technology is now being used to decontaminate metalparts from the turbine facility. This work will enable us toaccumulate knowledge that will help us refine the process,including scale-up and automation.

Reference
3 Decontamination Examination Result of Hand-Operated Decontaminant−,
Hamada, N., Preparatory Situation for Operation of Clearance System −○Document Collection of the 23rd Technical Special Committee on Fugen Decommissioning, JAEA-Review 2011-026, 2011, p.25-31(in Japanese).

110 JAEA R&D Review 2012

■ Nuclear Science Research Institute

Development of Experimental Techniques / Facilities at JAEA R&D Centers13-2 Improved Reliability of Hand Dosimetry=Development of a Ring-Type Dosemeter Using Optically Stimulated
Luminescence Elements (OSLRD)=

Entrance window

Endcap
Filter (aluminium)
Filter (copper)
Fig.13-3 Composition of the OSLRD
To measure dose equivalents for β- and γ(X)-rays separately, theOSLRD employs metal filters placed between two OSL elements toprevent incident β-rays from passing through the second OSL
element. This enables the second OSL element to measure onlyγ(X)-rays.

1st OSL element
2nd OSL element

2.0

2.0

1.5

1.5
Response

Response

Holder

1.0
0.5
0.0
0.5

1.0
0.5
0.0

1

1.5

2

2.5

Energy of β-rays (MeV)

10

100
1000
Energy of γ(X)-rays (keV)

10000

Fig.13-4 Energy dependence for β-rays
Energy dependence was within ±15% for energies between
0.88 and 2.06 MeV.

Fig.13-5 Energy dependence for γ(X)-rays
Energy dependence was within ±15% for energies between
48 and 1250 keV.

Measurements of hand exposure doses are required forradiological decontamination work, because exposure doses
to hands are often higher than those to human bodies. Hand
exposure doses had been measured with ring - typethermoluminescence dosemeters (TLDs) in the Nuclear
Science Research Institute. However, TLDs have somedisadvantages, such as short-term stability (only one month)and a destructive readout process in which all dose
information is lost in a single readout.
In recent years, an optically stimulated luminescence(OSL) technique has been successfully applied to many
dosimetry problems. OSL dosemeters offer the advantage oflong-term stability and allow repeated readouts. To overcome
the difficulties associated with hand exposure measurements
using TLDs, we developed a new ring-type OSL dosemeter(OSLRD) using compact water-resistant OSL elements
produced by LANDAUER INC. (Fig.13-3). The shape of thering was designed to maintain workability and ensure

performance of the OSLRD.
Irradiation tests using β- and γ(X)-rays in standard radiation
fields and Monte Carlo simulations were conducted to evaluate
the performance of the OSLRD. The energy dependence of
the OSLRD was within ±15% for both β- and γ(X)-rays(Fig.13-4, Fig.13-5). These values are lower than the values(±30%) required by the Japanese Industrial Standards for theenergy dependence of OSL elements. We also confirmed that
the OSLRD discriminates between β- and γ(X)-rays forvarious types of radionuclides and that the OSLRD operationis stable for three months.
The reliability of hand dosimetry was improved using the
OSLRD developed in this work. The OSLRD has sufficientperformance for a dosemeter and allows repeated readouts ofdose information. This OSLRD has already been placed inpractical use in JAEA, and it is expected to be widely used inmany different types of radiation work.

Reference
Miyauchi, H. et al., Development of Ring Type Dosemeter Using OSL Elements, JAEA-Technology 2010-050, 2011, 17p. (in Japanese).
JAEA R&D Review 2012111

Development of Experimental Techniques / Facilities at JAEA R&D Centers

Nuclear Science Research Institute ■13-3 Effective Utilization of Concrete Debris Generated from Nuclear Reactors=Clearance of Concrete Debris Generated from Modification Work on JRR-3=

Packing of less than 1 ton
concrete debris into a
storage container(Storage management)
Storage ofconcrete debris

Removal of
concrete debris

Segregation offoreign matterfrom concrete debris

Confirmation of nonlocalization of radioactivity
distribution of 60Co in each 100
kg concrete debris

Confirmation of
the correctness ofmeasurement andevaluation resultsby the government

Process for recycling

Radioactivity measurementof 3H, 60Co, 137Cs, and 152Eu

RecyclingFig.13-6 Flow of procedures to measure and evaluate radioactivity concentration
We established procedures to measure and evaluate radioactivity concentration in concrete debris generated frommodification work of JRR-3.
Before process for recycling

After process for recycling

Fig.13-7 Processed concrete debris for recycling
We crushed concrete debris to satisfy the quality requirements of
RC40 after the regulatory authorization of clearance.

We support effective utilization of materials through theclearance system. The purpose of the clearance system is toallow recycling of valuable materials generated from nuclear
facilities so as to contribute to the establishment of arecycling-oriented society. The clearance system sets upperbounds on radioactivity concentration of materials, so that
below those bounds the materials pose no radiation hazard tohuman health.
We have been storing about 4000 tons of very low-level
radioactive concrete debris as radioactive wastes generated
from the modification of JRR-3 between 1985 and 1990. At
that time, the clearance system was not a part of ourregulatory law. To recycle the debris, we conducted apreliminary survey of contamination levels, and thenestablished procedures for measuring and evaluating theradioactivity concentration. These procedures include theselection of radionulides to be measured, choice ofradioactivity measurement methods depending on characteristicsof the debris, methods for evaluating the measured results,
and management of the debris after measurement. In 2008,
the government authorized our procedures for clearance.
Since FY 2009, we have measured and evaluated theFig.13-8 Recycling of the concrete debris
We recycled the processed concrete debris as
backfilling material for depressed areas aroundbuildings, which were due to the earthquake on
March 11, 2011.

radioactivity concentration of concrete debris, using thegovernment-authorized procedures. By the end of FY 2011,
the government had confirmed the correctness of ourmeasurements and evaluation results on about 1900 tons ofconcrete debris (Fig.13-6).
We processed the concrete debris that was approved by thegovernment for recycling, and then the Ibaraki Construction
Technology Center verified that the debris met the quality
requirements in RC40; this required the debris to pass sieve
tests, abrasion tests, and foreign matter contamination tests(Fig.13-7).
By the end of FY 2011, we had reused about 600 tons ofconcrete debris as backfilling material in depressed areas
around buildings in the Nuclear Research Institute. Those
depressed areas were caused by the Great East Japan
Earthquake on March 11, 2011 (Fig.13-8).
This is the first Japanese case of recycling cleared concrete
debris. It is expected that our procedures will contribute tofuture decommissioning of nuclear facilities in a reasonable
way and will promote formation of a recycling-oriented
society.

Reference
Satoyama, T. et al., Clearance of Concrete Generated from Modification Activities of JRR-3 −Method for Measuring and Evaluating of Radioactivity
Concentration−, JAEA-Technology 2011-003, 2011, 53p. (in Japanese).

112 JAEA R&D Review 2012

■ Nuclear Fuel Cycle Engineering Laboratories

Development of Experimental Techniques / Facilities at JAEA R&D Centers13-4 Separation and Recovery of Long-Lived Nuclides for Reduction of Radioactive Waste=Behavior of Oxidation and Extraction of Neptunium in Simplified Solvent Extraction Process=Np (IV)Np (V)Np (VI)

10 2

100

Np (V) ratio (%)

Distribution coefficient (㧙)

80
10 1

10 0

60

40

10-1
20

10-2

0
0

2

4

6

8

10
3

HNO3 concentration (mol/dm )

3

3.5

4

4.5

5

5.5

6

3

HNO3 concentration (mol/dm )Fig.13-9 Distribution coefficient of Np in HNO3-TBP system
In an HNO3 medium, Np normally has three valences: Np(VI),
Np(V), and Np(VI). The distribution coefficients of Np(IV) andNp(VI) are high, and they are extracted into TBP. However, thedistribution coefficient of Np(V) is low and it is not extractable.

Fig.13-10 Relationship between Np(V) ratio and HNO3
concentration
Effect of HNO3 concentration on Np(V) ratio in a dissolver
solution of nuclear fuel was evaluated. The Np(V) ratio
decreased with increasing HNO3 concentration in a dissolver
solution of nuclear fuel.

Neptunium (Np) has a long half-life isotope (237Np:
2.14×106 y), and in conventional nuclear fuel reprocessing,it is disposed as radioactive waste with fission products in astable geological environment. In fast reactor fuel reprocessing,uranium (U), plutonium (Pu), and Np have to be recovered at
the same time from a solution, which is produced bydissolving irradiated nuclear fuel in nitric acid (HNO3) using
tri-n-butyl phosphate (TBP) as an extractant. Because theseelements are loaded again as nuclear fuels to fast reactors, theamount of radioactive waste will decrease.
The extractability of Np into TBP varies with its valence inthe HNO3 medium; Np(IV) and Np(VI) are extracted into
TBP, while Np(V) is not (Fig.13-9). Therefore, the valence of
Np needs to be changed to Np(IV) or Np(VI) in order toextract Np into TBP. It is known that Np(V) is oxidized toNp(VI) by HNO3, so the variation of Np(V) ratio with HNO3
concentration was evaluated (Fig.13-10). The results showthat the Np(V) ratio decreased with increasing HNO3
concentration in a solution, and high HNO3 concentrations

have beneficial effects on Np oxidation. In the simplified
solvent extraction process where Pu partitioning andpurification sections are deleted, Np(V) has to be oxidized toNp(VI) in the feed solution as much as possible.
To confirm the Np(V) oxidation effect, countercurrent
experiments were carried out using centrifugal contactors at
the Chemical Processing Facility (CPF). A dissolver solutionof irradiated fast reactor fuel was used as the feed solution.
Using 3 mol/dm3 of HNO3 in the feed solution, 9.9% of Np
leaks to the raffinate, while the leakage was less than 1.3%
using 5 mol/dm3 of HNO3 in the feed solution. The
experiments showed that Np(V) was easily oxidized toNp(VI) in a feed solution containing 5 mol/dm3 HNO3; this inturn promoted the extraction of Np into TBP in thecentrifugal contactors.
In this study, U, Pu, and Np co-recovery was demonstrated
by using a high HNO3 concentration (5 mol/dm3) of feedsolution.

Reference
Nakahara, M. et al., Influence of Nitric Acid and Nitrous Acid on Oxidation and Extraction of Neptunium with Double Scrub Flow Sheet in Simplified
Solvent Extraction Process, Journal of Chemical Engineering of Japan, vol.44, no.5, 2011, p.313-320.
JAEA R&D Review 2012113

J-PARC Center ■Development of Experimental Techniques / Facilities at JAEA R&D Centers13-5 Scintillation Neutron Detectors with a High Spatial Resolution=Position-Sensitive Detectors for Single-Crystal Diffractometer in J-PARC=(b) Schematic view
To photomultipliers(c) Measured beam profile

Wavelengthshifting fibers

Clear fibers

9913.0Counts(a) picture

WavelengthShifting fiber(WLSF)

0.9 mm(in full widthat half maximum)

0.0

y

.0
els
pix

Scintillator (top)
Scintillator (back)

B4C Block27

neutron

Scintillator(133×133 mm)

34.0

pixelsx

Fig.13-11 A high spatial resolution neutron image detector for iBIX
Wavelength shifting fibers that are sandwiched between scintillator screens collect scintillation light efficiently. Bending
the fibers backward with a radius of 3 mm helps decrease the neutron-insensitive area. The detector measured acollimated beam with a width of 0.3 mm as 0.9 mm in FWHM, demonstrating a high spatial resolution.

(d) Detector module(e) Detector efficiency
Detector efficiency (%)

300 m

m

Detector head

m
300 m

(f) SENJU

80

Signal processing & Dataaquisition system

Experiment
Theory60

40

20Detectors
0

Detector inside(WLSFs)Thermal neutrons(1.8Å)

0

1

2

3

4

5

6

Neutron wavelength (Å)

Fig.13-12 A large area neutron image detector for SENJU
One detector module has a neutron-sensitive area of 256 × 256 mm2. The SENJU accommodates 31 detectors,
covering a total scattering angle of 30%. The state-of-the-art detectors make the SENJU a unique and highly efficient
instrument.

Neutron diffraction on a single crystal of protein is aunique technique for identifying the hydrated position of thesample. However, acquiring high-quality data from the protein
has inherent difficulties because the diffraction intensity is
weak and scattered Bragg peaks are very close. A twodimensional neutron detector that has a spatial resolution lessthan 1 mm and a detector efficiency higher than 50% with aminimum dead area is required. To fulfill such specifications,
we have developed a detector that consists of scintillator andwavelength shifting fibers (WLSF) (Fig.13-11). With thisdetector, a neutron is detected through scintillation light. A
spatial resolution of 0.9 mm is achieved by confining thescintillation light to a 0.4-mm thick scintillator and by reading
the light using a 0.5-mm diameter fiber.
We also developed a ZnS/10B2O3 ceramic scintillator thathas a four-fold larger neutron absorption cross section than aconventional ZnS/6LiF scintillator. The detector, which
employs two scintillator screens and the dedicated photon
counting electronics, exhibited a detector efficiency of 50%
for thermal neutrons. This is higher than the conventional
detector efficiency of 35%. The IBARAKI Biological Crystal
Diffractometer (iBIX), populated with 30 such detectors, hada significantly reduced measurement time: only three days fora 1-mm3 protein sample. In contrast, more than a month was

required with a conventional instrument built in the Japan
Research Reactor-3 (JRR-3).
We have also developed a new image detector for theSingle-crystal Neutron Diffractometer under Extreme Condition(SENJU) project. The aim of SENJU is to study crystal
structures of materials under extreme environmental conditions,such as low temperatures and high magnetic fields. The
SENJU project requires unique and unprecedented detectorsthat have a spatial resolution of 4 mm, high detector
efficiency, and a large detector to scan a momentum space as
efficiently as possible. We designed a dedicated detector headthat incorporates a 1-mm diameter WLSF placed with aregular pitch of 4 mm (Fig.13-12). With this head, thedetector exhibited a detector efficiency of 40% for thermal
neutrons and the neutron-sensitive area was four times that of
the iBIX detector. The SENJU shortened the measurement
time to only a day for a small sample (1 mm3); this compares
to a few days required for the measurement with a similar
instrument in the JRR-3.
These new scintillation detectors significantly add to thehigh-performance neutron scattering instruments in the
J-PARC. These new types of detectors would also be very
useful in any application that requires measurement of theincident positions of neutrons.

Reference
Nakamura, T. et al., A Large-Area Two-Dimensional Scintillator Detector with a Wavelength-Shifting Fibre Readout for a Time-of-Flight Single-Crystal
Neutron Diffractometer, Nuclear Instruments and Methods in Physics Research A, vol.686, issue 1, 2012, p.64-70.

114 JAEA R&D Review 2012

■ J-PARC Center

Development of Experimental Techniques / Facilities at JAEA R&D Centers13-6 Visualizing Magnetic Fields Using Neutrons=Magnetic Field Imaging Technique Using Polarized Pulsed Neutron=

Polarization Analysis Setup
PulsedNeutrons

Image Detector

Spin Rotator

Spin Rotator
ObjectPolarizer

Analyzer

Spin Flipper(b) Y component of magnetic field vector

14

14

12

12

1.0

10

10

0.5

8
6

4

2

2

-0.5
-1.0

0
0

X

0.0

6

4

0

Y

8

Field strength (T)(a) X component of magnetic field vector

Y (mm)Y (mm)

Fig.13-13 Schematic of magnetic field imaging system using polarized neutrons
The magnetic field imaging apparatus is composed of a polarization analysis device, which consists of a polarizer, an analyzer, a spin
flipper, spin rotators, and an image detector. The maximum polarization of the incident neutron beam is 98% in this system.

48
X (mm)

120

4

8
X (mm)

12

Fig.13-14 Observation of magnetic field distribution inside a soft magnetic foil
Magnetic field distribution of (a) X-component and (b) Y-component of magnetic field vectors were visualized for a soft magnetic foil
(left picture) using the magnetic field imaging system. The arrows in (b) show the direction of the magnetic field inside the stripes.

Because a neutron has a magnetic moment, it interacts
with a magnetic field directly and precesses around themagnetic field. The resulting rotational angle of the neutron
spin reflects the magnetic field strength and the direction inwhich the neutron moves. Therefore, by analyzing theneutron spin rotation at each position, we can obtain imagesof spatial distribution of magnetic fields. Because interactions
between neutrons and magnetic fields depend on theirvelocities, measurements of the spin rotation as a function ofvelocity enable us to quantify the magnetic field. Thus, using
pulsed neutrons is very suitable because they make available
a wide range of neutron velocities; precision measurements
can then be obtained by applying the time-of-flight method.
In this work, we developed a new technique for visualizing
magnetic field distribution by combining the neutron
polarization analysis with the pulsed neutron imaging method(Fig.13-13). Neutron spin rotation is recorded as the change
in polarization at each position over a wide range of neutron

velocities. In addition, by using spin rotators to select bothincident and transmitted neutron spin directions, vector
components of the magnetic field can be obtained; then the
direction of the field can be determined. The images in
Fig.13-14 show the magnetic field distribution of a soft
magnetic foil obtained at the instrument NeutrOn Beamline
for Observation & Research Use (NOBORU) of J-PARC
MLF. These images contain stripe-shaped magnetic domains
with magnetic field directions in neighboring domains being
opposite one another (see arrows in Fig.13-14(b)). The
magnetic field strength was evaluated to be 1.2 T. A high-flux
neutron beam is required for analyzing the neutron
polarization with fine spatial resolution. J-PARC s highintensity pulsed neutrons enabled us to not only visualize but
also quantify the magnetic field with a spatial resolution ofless than 1 mm.
The present study was partially supported by a JSPS Grantin-Aid for Scientific Research (C), Grant No.22604009.

Reference
Shinohara, T. et al., Quantitative Magnetic Field Imaging by Polarized Pulsed Neutrons at J-PARC, Nuclear Instruments and Methods in Physics Research
Section A, vol.651, issue 1, 2011, p.121-125.
JAEA R&D Review 2012115

J-PARC Center ■Development of Experimental Techniques / Facilities at JAEA R&D Centers13-7 Realizing a 1 MW High-Power Proton Beam
1
0.9
0.8
0.7
0.6
0.5
0.4
0.3
0.2
0.1
0

Without transverse
injection painting

With transverse
injection painting

End ofinjection
0

50

0 ms

Transverse position (mm)Charge density (a.u.)

Extraction(3 GeV)(181 MeV)-50

Fig.13-15 Transverse beam profiles obtained with andwithout transverse painting
The charge density peak in the transverse direction decreases
by transverse injection painting.
1
0.9
0.8
0.7
0.6
0.5
0.4
0.3
0.2
0.1
0
-100

Beginning of
injection

Without longitudinal
injection painting
With longitudinal
injection painting

Circulating beam intensity(×1013)Charge density (a.u.)=Beam Loss Reduction by Injection Painting in J-PARC RCS=

0.5 ms

20 ms

4
3.5
3
2.5
2

Without
injection
painting

1.5
1
0.5
0
10

-1

With
injection
painting

1

10Time (ms)

-50

0

50

100

Longitudinal position (m)Fig.13-17 Beam loss observed for the 420 kW intensity
beam with and without injection painting
The beam loss observed just after injection decreases byinjection painting.

Fig.13-16 Longitudinal beam profiles obtained with andwithout longitudinal painting
The charge density peak in the longitudinal direction
decreases by longitudinal injection painting.

The J-PARC Rapid Cycling Synchrotron (RCS) has twofunctions as a proton driver to produce pulsed muons andneutrons at the Materials and Life Science Experimental
Facility and as an injector to the following Main Ring
Synchrotron. RCS accelerates protons injected from the linac
up to 3 GeV with a repetition rate of 25 Hz. Our final goal isto achieve 1 MW output beam power, which would be thehighest level in the world.
The most important issues in increasing output beam
power are the control and minimization of beam loss tomaintain machine activation within the permissible level.
There are many sources of beam loss, but the most criticalone is the space-charge effect arising from Coulomb
repulsive forces among beam particles. This effect is mostserious at low energies. Therefore it causes beam loss justafter injection. To alleviate this effect, RCS adopts aninjection painting scheme, that reduces the charge density
peak by distributing injected beam particles as uniformly as
possible in both the transverse (direction at right angle to thebeam orbit) and longitudinal (direction along the beam orbit)
phase spaces. In transverse phase space, the injection beam isuniformly distributed by controlling the amplitude of betatronoscillation, namely by varying the phase-space offset between
the centroid of the injection beam and the ring closed orbit
during injection (0.5 ms), as shown in Fig.13-15. On theother hand, in longitudinal phase space, a uniform particle
distribution is formed through emittance dilution by large
synchrotron oscillation excited by a momentum offset during
injection. By combining a dynamical control of the synchrotron
potential well with the momentum offset injection scheme,
the charge density peak in the longitudinal direction was
significantly reduced, as shown in Fig.13-16. Using
transverse and longitudinal injection painting, the large beamloss of 15%, observed for the 420 kW intensity beam, wassuccessfully reduced to less than 1% , as shown in Fig.13-17.
At present, the injection energy is 181 MeV, but it will be
upgraded to 400 MeV in 2013. After that, we will aim at our
final goal of 1 MW output beam power. The above result
achieved for the 420 kW intensity beam with an injectionenergy of 181 MeV is considered to be a big step towardrealizing 1 MW design beam operation with the higher
injection energy of 400 MeV, because the above two beam
operations give an equivalent space-charge effect at each
injection energy.

Reference
Hotchi, H. et al., Beam Loss Reduction by Injection Painting in the 3-GeV Rapid Cycling Synchrotron of the Japan Proton Accelerator Research Complex,
Physical Review Special Topics − Accelerators and Beams, vol.15, issue 4, 2012, p.040402-1-040402-15.

116 JAEA R&D Review 2012

■ Oarai Research and Development Center

Development of Experimental Techniques / Facilities at JAEA R&D Centers13-8 First HTGR Demonstration=First Achievement of a Continuous Supply of High-Temperature Nuclear Heat=
Reactor
30 MW

UO2
kernelHeatfrom reactor

CFP

Fuel block

-4

10

-5

10

Water
cooler20 MW

in FSV (USA)-610

in AVR (Germany)
-7

10

1000

Primary

750

Secondary

500
250

Reactor power

0

100
75

HTTR

Supplying heat to thesecondary cooling system

-8

10

50
25

High-temp. cont. operation

High temp. cont. operation

-9

10

1/5

1/15

1/25

2/4

2/14

2/24

3/6

3/16

Date (2010)

0
1/5

1/25

2/14

Reactor power (%)

Fuel rod

IHX helium temp. (℃)

Fuel compact

Watercooler

IHX
10 MW

Heat
exchange
at IHX

Ceramic
coating layer

R/B of 88Kr (-)

Heat utilization system(in planning)

Supply tosecondary
the coolingsystem

3/6

Date (2010)

Fig.13-18 Confinement performance of HTTR fuels
R/B of 88Kr demonstrates the FP confinement performance of
CFPs used as fuel; a lower R/B indicates a higher confinement
performance. R/B was almost constant during operation andwas much lower than that in a previous HTGR in anothercountry.

Fig.13-19 Heat exchange performance of the IHXHeat generated in the reactor is transferred to the secondary
cooling system through the IHX. The supplied heat andprimary/secondary helium temperature at the IHX were almost
constant during operation. It was demonstrated that the IHX
could supply the generated heat to the heat utilization system
for a long period.

The high-temperature gas-cooled reactor (HTGR) is
attractive worldwide because of its inherent safety features,
multipurpose heat utilization, and high-efficiency power
generation. The high-temperature engineering test reactor(HTTR) was constructed as the first HTGR in Japan andachieved a reactor outlet coolant temperature of 950 ℃,
which was the highest temperature reported anywhere in
April 2004.
To use an HTGR as a commercial reactor for power
generation or hydrogen production, it was necessary todemonstrate that an HTGR could supply high-temperature
heat to a heat utilization system stably for a long period. In
the HTTR, 50 days of high-temperature operation at a
maximum reactor outlet coolant temperature of 950 ℃ was
achieved in 2010 for the first time anywhere. The fuelperformance and heat-supplying characteristics of the hightemperature equipment for long-term continuous hightemperature operation were evaluated during operation.
Coated fuel particles (CFPs) are used as HTTR fuel. It wasnecessary to confirm the integrity of the CFPs [their fission
product (FP) confinement performance] for long-term

continuous high-temperature operation. The release/birth rate(R/B) of 88Kr, which is a typical FP, during operation was
estimated. The results confirmed that the R/B during
operation was almost constant and was the same as that inprevious operation. Moreover, the integrity of the CFPs is
ensured and the CFPs exhibited high FP confinement
performance for a long duration (Fig.13-18).
The intermediate heat exchanger (IHX) is part of the hightemperature equipment in the HTTR. It is very important for
stably supplying heat generated in the reactor to the heatutilization system. The primary/secondary helium temperatureand heat supplied to a secondary cooling system during thepresent operation were evaluated. The results confirmed thatthey were in a stable state during operation, and the IHX
could stably supply high-temperature heat to the secondary
cooling system for a long period (Fig.13-19).
These results demonstrated that an HTGR could stably
supply heat for a long period.
We were awarded the Award for Distinguished Technology
Development 2012 of the Atomic Energy Society of Japanfor these results.

ReferenceTochio, D. et al., Result of Long-Term Operation of HTTR −High-Temperature/Parallel-Loaded 50-Days Operation−, JAEA-Technology 2010-038, 2010,
57p. (in Japanese).
JAEA R&D Review 2012117

Oarai Research and Development Center ■Development of Experimental Techniques / Facilities at JAEA R&D Centers13-9 Development of a Dissimilar Metal Joint of Zircaloy-2 and 316L Stainless Steel=Application to Power Ramping Test of LWR Fuel in JMTR=

Joint technique of fuel sample-Development of dissimilar metal joint-

Pressure
vesselIn-pile tube㩷

Fuel sample

Instrumentation

BOCA

Tensile strength (MPa)(a) Relationship between friction time and tensile strength
600
Rotation speed N:1500 rpm
Friction Pressure P1:147 MPa
Upset puressue P2:245 MPa
Upset time t 2:6.0 s

500

Test temp:294 K

400
Zircaloy2

316 L SS
110°

Test temp:573 K

300
Pressure direction

200

0

1

2

3

4

Friction time (s)(b) Photographs of dissimilar metal joint
Zircaloy2

316L SS

Zircaloy2

316L SS

Fuel sample
Core

HAZ

Inner structure of BOCA

Welding part
(Friction time: 3 s)

Cross section of welding partFig.13-20 Power ramping test of LWR fuel in JMTR
The fuel center temperature and pressure in the fuel sample
are measured by the instrumentation attached to the fuelsample. The sample and instrumentation cannot be welded
directly because they are made of dissimilar materials.
Therefore, a dissimilar joint is first fabricated, and then, thejoint is attached to the tip of the instrumentation. In this way, it
is possible to form a weld in the same metal in the hot cell.

Fig.13-21 Joint test of the dissimilar metal joint(a) Relationship between friction time and tensile strength was
revealed by tensile tests using a dissimilar metal joint of
Zircaloy-2 and 316L SS.
(b) The heat-affected zone and white discoloration due toetching were observed near the weld interface, and the weld
flash was also observed. Moreover, no cracks or holes wereobserved in the weld interface.

A power ramping test facility using a boiling water capsule(BOCA) in JMTR (Fig.13-20) has been prepared for integrity
tests of high burn-up light water reactor (LWR) fuels under
the power transient condition. In the BOCA, different
material bonding has been applied to the fuel sample andinstrumentation components, which are made of Zircaloy-4
and 304 SS, respectively. However, each metal has
disadvantages: For Zircaloy-4, it is difficult to obtain thematerial for manufacturing a tube having the same diameteras the fuel sample. Moreover, 304 SS is prone to stresscorrosion cracking under neutron irradiation, and its use
increases the amount of radioactive waste owing to the highdose after irradiation.
To overcome these disadvantages, two materials, Zircaloy-2
and 316L SS, were selected, and a new dissimilar metal joint
was developed. In the development process, the condition of
Zircaloy-2 and 316L SS joints obtained by friction welding
was investigated. Joint tests were conducted to examine
parameters such as the rotation speed, friction pressure, andfriction time at room temperature. A suitable rotation speed
and friction pressure were first determined, and then, thefriction time was determined.
Tensile tests at room temperature showed that the tensile
strength increased with increasing fraction time and ranged
from 529 to 536 MPa at friction times greater than 3 s.
Therefore, it is possible to obtain a minimum tensile strengthof more than 480 MPa as regulated for 316L SS by thestandard of the Japan Society of Mechanical Engineers(Fig.13-21(a)); note that 316L SS has a smaller strength thanZircaloy-2 at room temperature. Furthermore, a tensile test at
573 K (the maximum temperature during the power ramping
test in JMTR) showed a tensile strength ranging from 277 to
281 MPa, which is almost the same as that of Zircaloy-2, 286 MPa
at 573 K. Note that Zircaloy-2 has a smaller strength than
316L SS at 573 K.
Fig.13-21(b) shows an optical observation of the weldinterface of the dissimilar metal joint. The heat-affected zone
(HAZ) near the weld interface was observed in the crosssection of the joint. However, the outer part of the HAZ was
smaller than the center part, and the HAZ was forced out withthe weld flash by friction welding. In addition, no cracks orholes were observed in the weld surface, thus confirming the
integrity of the welding joint.
The above results clarify the relationship between thefriction time and tensile strength for a dissimilar metal jointof Zircaloy-2 and 316L SS, and the joint condition wasdetermined. In future, the technology developed for obtaining
the dissimilar metal joint by friction welding will be applied
to the fabrication of the BOCA.
This study was sponsored by The Secretariat of the Nuclear
Regulation Authority (former Nuclear and Industrial Safety
Agency, the Ministry of Economy, Trade and Industry of Japan).

Reference
Kitagishi, S. et al., Performance Evaluation of Friction Welded Joint of Zircaloy-2 to Type 316L Stainless Steel, Nippon Kikai Gakkai Rombunshu, A Hen,vol.78, no.788, 2012, p.564-570 (in Japanese).

118 JAEA R&D Review 2012

■ Oarai Research and Development Center

Development of Experimental Techniques / Facilities at JAEA R&D Centers13-10 Corrosion Issues in Sodium Environment=Effects of Sodium on Structural Materials for Fast Breeder Reactors=(a) 525 ℃(b) 575 ℃(c) 625 ℃(d) 420 ℃
Corrosion
products

10 µm

Sodium flow

Fig.13-22 Corrosion of 304 SS exposed in flowing sodium for 82000 h
Corrosion due to dissolution of alloy elements in structural materials is observed in higher-temperature regions, anddissolved products are deposited in lower-temperature regions.

Start-up

Steady-state corrosion

Fig.13-23 Schematic diagram of sodium
corrosion characteristics
Metal loss in sodium occurs in two stages: startup corrosion caused by selective dissolution ofalloy elements and steady-state corrosion,
which is the corrosion reaction of iron. Start-up
corrosion continues for up to a few thousands ofhours; subsequently, steady-state corrosion
proceeds at a constant rate.

Metal loss

Total loss (A) + (B)For sodium-cooled fast breeder reactors (FBRs), the effects
of a sodium environment on the corrosion and mechanicalstrength properties of structural materials have to beevaluated in order to maintain the material integrity
throughout the design lifetime.
Because the oxides that form on the materials are easily
reduced by sodium, a liquid-phase metal (sodium) comes into
direct contact with solid-phase metals (structural materials).
When this happens, because of the difference in the chemical
potential depending on the temperature and time, either thealloy elements dissolve from the metal surface into thesodium or the inverse process (i.e., deposition) occurs. Thesephenomena would be essentially controlled by the saturation
solubility of the alloy elements and the oxygen concentration
in sodium.
In a non-isothermal sodium circulation loop such as that in
FBRs, the alloy elements dissolve in the higher-temperature
region, and the dissolved elements are deposited in the lowertemperature region because of supersaturation (Fig.13-22).
Start-up corrosion is caused by the selective dissolution ofalloy elements such as nickel and chromium into sodium. In
contrast, steady-state corrosion is dominated by the corrosion
reaction of iron (Fig.13-23). Furthermore, because the carbon

Bulk loss (A) ∝ t
Selective loss (B) ∝ t 1/2

Sodium immersed time, tcontained in steels plays an important role in maintaining thematerial s strength properties, considerable emphasis shouldbe placed on the decarburization and carburization phenomena
in sodium. In a bimetallic sodium circulation loop consistingof stainless and ferritic steels, because of the difference incarbon activity between the two materials, decarburization
occurs in ferritic steel, which has higher carbon activity,
whereas carburization occurs in stainless steel.
In addition, because FBRs are used at higher temperatures
than light water reactors, it is necessary to understand thecreep properties in addition to the tensile and fatigue
properties. Furthermore, the damage caused by thesuperposition of creep and fatigue becomes dominant inresponse to repetitive loading due to thermal stress.
Therefore, it is also important to understand the creep-fatigue
strength properties. On the other hand, it is understood thatthe effects of decarburization/carburization and sodium
corrosion of materials are not a dominant factor in impuritycontrolled sodium.
This knowledge has been summarized to describe theenvironmental effects of high-temperature sodium and isreflected in the material strength standard for the design of
sodium-cooled FBRs.

Reference
Yoshida, E. et al., Corrosion Issues in Sodium-Cooled Fast Reactor (SFR) Systems, Nuclear Corrosion Science and Engineering, no.22, 2012, p.773-806.
JAEA R&D Review 2012119Development of Experimental Techniques / Facilities at JAEA R&D Centers

Takasaki Advanced Radiation Research Institute ■13-11 Chemical-State Observation of Microparticles=Microscopic Imaging of Luminescence Induced by Scanning Ion Microbeam=

Characteristic X-rays:
Elemental Compositions

X-ray Intensity (a.u.)104+

3 MeV H
microbeam

103

PIXE (Si Kα)

Ca

Low

Ca

10

20 µm
0 1 2 3 4 5 6 7 8 9 10Energy of X-ray (keV)
PIXE Spectrum

Analysis Targets
aerosols, asbestos,or othermicroscopic targets

20 µm

Elemental Distribution by micro-PIXE
Wavelenghcorresponding to
CaPO4

Wavelenghcorresponding to
CaSiO3
IL Intensity

IL: Chemical Compositions

Photon Intensity (a.u.)

200

Collimator Optical Lens for ILUMISHigh

PIXE (P Kα)

102

1

X-ray Semiconductor Detector
for Micro-PIXE

X-ray Intensity

Si P

150
100
50
0

High

CaSiO3 :λ=556 nm

I L (ǳ= 426 nm)

Low

CaPO4 :
λ=426 nm

250

450

20 µm
650

850

Wavelength ǳ(nm)ILUMIS Spectrum

Fig.13-24 Schematic illustration of ILUMIS system combined withthe micro-PIXE system (left) and spectra of PIXE (right upper) and of
IL (right lower)
Compositions of elements, such as silicon, phosphorus, and calcium,
were analyzed by PIXE. Those of chemical compositions (i.e., CaSiO3
and CaPO4) can now be analyzed by ILUMIS.

Air-borne particles (aerosols) are abundantly floating in
the atmosphere, providing surfaces that can transport harmful
chemical substances and microbes. Large-scale transmission
and expansion of chemical and biological toxins is a global
environmental issue. To study the adhesive nature ofparticulate surfaces for such toxic materials, we must be able
to visualize distributions of not only elements but also theirchemical states in micrometer-sized structures.
At the ion accelerator facility of TIARA in JAEA/Takasaki,
Particle-Induced X-ray Emission analysis using an ion
microbeam (micro-PIXE) has been developed and utilized toanalyze elemental distributions in individual microscopic
samples. Micro-PIXE analyses detect characteristic X-rays
emitted from elements heavier than Mg during the scan of an ion
microbeam on a sample with typical spatial resolution of 1 µ m.
During PIXE attributed to the ionization of inner-shellelectrons, low-energy photons are also emitted due to theexcitation of outer-shell electrons in the target. Such photon
emission, which is called ion luminescence (IL), is strongly
related to the chemical state of the elements or molecules;
thus, measurements of IL enable us to determine chemicalstate on the same sample area in which elemental
compositions are analyzed by micro-PIXE analysis at thesame time.
As shown in Fig.13-24, we have newly developed a

I L (ǳ= 556 nm)

20 µm

Compound Distribution by ILUMIS
Fig.13-25 Examples of microscopic imaging of
the elemental distribution obtained by PIXE
(upper) and the chemical-state distribution by IL(lower) from aerosol samples with the scanningion microbeam
While the main constituent elements of aerosols,
such as silicon and phosphorus show similar
distributions in micro-PIXE imaging, IL imaging
shows that only CaSiO3 is present.

wavelength-dispersive IL analysis system named Ion
Luminescence Microscopic Imaging and Spectroscopy(ILUMIS) to obtain the chemical state distribution frommicroscopic target. The system consists of a large optical lensfor high detection efficiency and an electrically cooled
photon sensor for high signal-to-noise ratios with a grating
effective between 200 and 850 nm.
Results obtained using the combined system of ILUMIS
and micro-PIXE is shown in Fig.13-25 for individual
aerosols. The particulate samples were scanned over with aproton microbeam in the atmosphere. The elemental
compositions of the particles were obtained for silicon andphosphorus by the micro-PIXE, as shown in the upper panelof the figure. On the other hand, chemical state of the target
was clearly obtained by ILUMIS. The result shows that theparticulate samples consist not of calcium phosphate (CaPO4)
but of calcium silicate (CaSiO3), as shown in the lower panelof the figure. These results demonstrate that ILUMIS
combined with micro-PIXE is a compact and powerful tool
for analyzing and imaging chemical compositions of
microscopic samples.
This research was partially supported by a MEXT/JSPS
Grant-in-Aid for Young Scientists (B) No.24710097 and by a
JAEA Grant-in-Aid for Exploratory Research.

Reference
Kada, W. et al., Development of Analysis System of Micro-IBIL Combined with Micro-PIXE, International Journal of PIXE, vol.21, nos.1-2, 2011, p.1-11.

120 JAEA R&D Review 2012

■ Tono Geoscience Center

Development of Experimental Techniques / Facilities at JAEA R&D Centers13-12 Shafts Excavated to -500 m Depth=Construction of Shafts and Research Tunnels and Confirmation of Reduced
Groundwater Inflow at the Mizunami Underground Research Laboratory=

Borehole
Large volume ofgroundwater inflow
50 m

2500

-200 m Measurement
Niche off Ventilation
Shaft

Prediction: A/R Gallery (Without grouting)-200 m Measurement Niche
off main Shaft

Prediction: A/R Gallery (Grouting)Groundwater inflow (ℓ/min)

Sub Stage(Every 100 m deep)-300 m Access/Research
Gallery-300 m Stage-300 m Measurement
Niche off Ventilation
Shaft

Measured: A/R Gallery (Grouting)

2000

Measured: Borehole

1690 ℓ/min
2.4 MPa

1500

1000

400 ℓ/min
500-500 m Stage

100 ℓ/min
0
0

10

20

30

40

50

60

70

80

90

100

Distance from Main Shaft (m)Ventilation Shaft:-500.2 m (Ǿ4.5 m)

Main Shaft:-500.4 m (Ǿ6.5 m)

Fig.13-26 Layout of the Mizunami
Underground Research Laboratory
Depth of the shafts as of March 31, 2012.

Fig.13-27 Predicted and measured groundwater inflow during
excavation of A/R Gallery and drilling borehole
Upper panel is the -300 m A/R Gallery and borehole, and location of
the large volume of groundwater inflow. Lower panel shows predicted
and measured groundwater inflow during excavation of -300 m A/R
Gallery and drilling borehole.

The Mizunami Underground Research Laboratory iscurrently under construction in Mizunami City. As of March
2012, the Main and Ventilation Shafts have been excavated to
GL -500 m (meters below ground level, Fig.13-26).
To mitigate potential excess groundwater inflow, preexcavation grouting was conducted before the excavation of
the shafts and tunnels. Grouting refers to the injection ofmaterial such as cement into open fractures in a rock mass tostabilize and seal the rock. In planning the construction, it
was necessary to obtain reliable preliminary information on
the bedrock conditions in terms of the rock mass stability andhydrogeology. Therefore, site characterization borehole
investigations were conducted before excavation began.
These investigations indicated that large groundwater inflow
could be expected during excavation.
Fig.13-27 shows the results of grouting around the -300 m
Access/Research (A/R) Gallery. In this area, maximum
groundwater inflows in excess of 1000 /min were expected
at several locations during borehole drilling. This area also
has relatively high groundwater pressure (2.4 MPa). Grouting
using ordinary Portland cement was performed in thisgeological environment, and the measured total groundwater
inflow after excavation of the -300 m A/R Gallery was
actually about 100 /min. Consequently, we conclude thatexisting grouting technology is applicable in deep undergroundenvironments and under high groundwater pressure conditions.
The grouting technology was improved and optimized
through several grouting trials during excavation, and themethod can be applied to rock masses with a wide
conductivity range under relatively high groundwaterpressure conditions. Existing grouting technology using
ordinary Portland cement and super-fine cement wasdemonstrated to reduce the conductivity of a rock mass in therange of 10-7 m/s. Furthermore, grouting methodology using
liquid-type grout material was also tested in the -300 m
Measurement Niche. The methodology will be applied for
grouting rock masses under 10-7 m/s. The excavations haveshown that the pre-excavation grouting was successful andthe targeted reduction in inflow was achieved. Thus, theresults indicate that this methodology is effective in reducing
water inflow.

Reference
Mikake, S., Ikeda, K. et al., Studies on Planning and Conducting for Reducing Water Inflow due to Underground Construction in Crystalline Rock, JAEATechnology 2010-026, 2010, 146p. (in Japanese).
JAEA R&D Review 2012121

Development of Experimental Techniques / Facilities at JAEA R&D Centers

Ningyo-Toge Environmental Engineering Center ■13-13 Progress in Methodology Considering -ray Shielding Effects=Development of Simplified Calculation Method for Heterogeneously
Distributed Radioactive Waste Drums=
1.0

1.0

0.9

0.9

γ-ray count rate (s-1)

0.8

0.8

0.7

0.7

0.6

New evaluation model

0.6

0.5

Shielding effect will bequantifiable by calculating the
Xgeometry. The variation ofcounting rate will correspondto the shielding effect.

0.5

0.4

0.4

0.3

0.3
0

0.2

1 2 3 4frequency

5

Conventional method
Variations in the counting rate
will be reflected in the amountof uranium.

0.1
0.0
1

10

100

1000

XgeometryFig.13-28 γ-ray measurement test using simulated waste
We produced a simulated waste and used it to measure the γ-ray
counting rates at 1001 keV and 766 keV.

Fig.13-29 Comparison of conventional methods and the
new model
We calculated the Xgeometry with a count rate for simulated waste
and plotted the results. We compared this result with theconventional method. Some of the data were obtained during
research funded by RANDEC.

It will be necessary to confirm the radioactive concentrationof uranium-bearing wastes drums generated at nuclear fuels
facilities to provide disposal validation. Although gamma- (γ-)
ray measurement is expected to be used as the measurement
technique, a correction for the effects of the heterogeneous
waste distribution was needed.
During measurement of the uranium waste, the γ-ray
counting rate will change significantly depending on thewaste placement and distribution of uranium. For the sameradioactivity, the γ-ray counting rate will vary greatly
depending on conditions between the detector and theshielding from the radiation source. This is why the error islarge.
We proposed a new theory on passive gamma assay for U-238
determination of uranium-contaminated waste drums. We
also proposed a new evaluation function that yields a uniquerelationship between the source positions and γ-ray countrate. In equation (1), we define the shielding effect, where kis the emission ratio between two peaks, and R is the ratio of
the two γ-ray counting rates.

According to this theory, regardless of the inhomogeneityof the waste matrix density or uranium source distribution,
we can accurately determine the amount of U-238 containedin drums non-destructively using count rates of γ-rays of
two energies (1001 keV and 766 keV) emitted from the
U-238 progeny nuclide Pa-234m. The two γ-ray peaks(766 keV/1001 keV) are used to analyze the difference in theattenuation rate through matrices in this methodology.
We verified the theory by tests under various wasteconditions using simulated waste drums (Fig.13-28). Tests
were performed against various patterns of source distributions
and matrices. The results verified the suitability of thismethodology using plotting coordinates based on equation(1) (Fig.13-29). We estimated the relative error to be less than
20% and confirmed that this new assay system is efficient for
the rational classification of uranium wastes requiring
disposal. We plan to introduce this method to an existingmeasuring device. From now on, this methodology will beapplied to the uranium determination measurements in orderto improve their accuracy. Furthermore, this methodology has
the potential for application to other measurements tools
using nuclides with multiple γ-ray emission wavelengths.

X geometry =1⎛ ⎛ k ⎞⎞⎜ln ⎜⎜ ⎟⎟ ⎟⎜⎟⎝ ⎝ R ⎠⎠

(1)
2

Reference
Yokoyama, K. et al., Passive γ ray Measurements for a Simple Determination of Radioactivity of Uranium in Waste Drums, Radioisotopes, vol.59, no.12,
2010, p.707-719 (in Japanese).

122 JAEA R&D Review 2012

About the Design of the CoverThe cover is intended to symbolize a hopeful future shining in the sky. The sky is clear blue, similar to the color of the JAEA logo. The
white hexagons are arranged in a pattern similar to that in a tortoise shell, which symbolizes the wish of the people for long lives sinceancient times in Japan. Coincidentally, this shape is the same as that of core fuel assemblies in both the prototype fast breeder reactor
MONJU and the high temperature engineering test reactor HTTR .
The images on the cover show decontamination of pavement using shot blasting (top left) and decontamination of soil using a poly-ion
complex (bottom right).
Shot blasting uses small iron balls (typically, around 1 mm in diameter) to remove a thin layer from a flat surface. We applied thisprocess, which has been used in nuclear facilities, to pavement at Okuma town, and found that it could reduce the surface contaminantdensity by more than 90% (Topic 1-8, p.19).
Soil decontamination using a poly-ion complex was tested on a 1500 m2 recreation ground at a community center in Oguni, Ryozen, Date-city,
Fukushima. The test removed 85~90% of radioactive cesium by stripping 2 cm of surface soil from the ground (July, 2011) (Topic 1-10, p.21).

JAEA R&D Review 2012

123

JAEA R&D Review 2012

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Chief editor: Mikazu YuiVice editor: Keizo Itabashi
Editors: Takumi Nemoto, Akihiro Tagawa, Yoshio Suzuki, Masaru Watahiki, Kunio Onizawa,

Naofumi Kozai, Takamasa Mori, Taiju Shibata, Hiroyuki Yamamoto,Shimpei Matsuhashi, Mitsuru Kikuchi, Kazunori Hirao, Yoshikazu Koma,
Fumiaki Yamada, Naotaka Shigeta, Hideyuki Funabashi, Tetsuro Ishii,
Yoshihisa Murokawa, Hiroshi Takada, Michihiro Naka, Masahiro Ishihara,
Junsaku Nakajima, Kazumasa Narumi, Shinichiro Mikake, Hideaki Yoshida
Published by Japan Atomic Energy Agency (JAEA)
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