Message from the President

President Toshio Okazaki

Welcome to the third issue of the “JAEA R&D Review”, which will inform you of thecurrent R&D activities of the Japan Atomic Energy Agency (JAEA).
Three years have passed since the establishment of JAEA. We are now reaching a very
important stage for the achievement of our goals for our R&D activities in the first mid-term
plan. In addition, the importance of nuclear energy was recognized at the G8 Summit held in
Tohyako, Hokkaido, Japan in July, 2008, because of its stable supply of energy and potential tosolve global environmental problems. JAEA is the sole comprehensive R&D institution aiming
at facilitating both basic research and practical application with regard to nuclear energy inJapan. In other words, our research areas cover a wide range of sciences and technologies
needed for peaceful and safe uses of nuclear energy. Through our activities, JAEA has been
contributing significantly to creating and improving various innovative technologies related tonuclear energy.
In the field of Fast Breeder Reactor (FBR) cycle technology designated as a “Key Technology
of National Importance” by the government of Japan, JAEA has been making great efforts torestart “MONJU” (a prototype FBR) and also has been promoting R&D of the FBR cycle
for commercialization. In the field of R&D for utilization of applied quantum beams, JAEA
continues to put great effort into the operation of “J-PARC” (high-intensity proton accelerator
project) which utilizes the most advanced quantum beam technologies. In the field of fusionenergy, JAEA is strongly pursuing the ITER Project and Broader Approach Project which arerecognized as a science and technology of strategic priority. Furthermore, JAEA is constructing
two underground research laboratories where R&D projects for disposal of high-level
radioactive waste are being undertaken. In addition, JAEA is steadily executing a new mission,for disposal of radioactive wastes which are generated by research facilities.
The mission of JAEA is to improve the quality of life for all people by innovativeadvancements in nuclear energy. I strongly believe that JAEA will achieve this goal and become
one of the world’s most authoritative and comprehensive institutes, what we call a Center of
Excellence (COE), for creative and reliable nuclear energy R&D.
This publication constitutes a review of our achievements in the fiscal year 2007. It provides
you with a flavor of some of the work that has been carried out, and also invites you to checkthe references listed and contact the researchers if there are any topics that you wish to learn
more about. I hope that you will take a few minutes to peruse some of the following pages. 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 very much for your interest.

J AEA R&D Review
1 Research and Development of Advanced Nuclear System
Steady Implementation of FaCT Project－Assessing Innovative FBR Cycle Technology－

9

1. Coolant Pipes Made Economically Competitive by Shortening
10－Development of High Chromium Steel for FBR Components Which Has Excellent High		
Temperature Strength, Toughness and Microstructural Stability－
2. Evaluation of the Structural Integrity of Reactor Vessels
11－Measurement of the Thermal Deformation near the Fluid Surface Using the Laser Speckle Method－
3. Determining Transfer / Damping of Temperature Fluctuation of Fluid to Structure
12－Evaluation of High Cycle Thermal Fatigue at Core Outlet in FBR－
4. Development of a Fuel Handling Method for a Compact Reactor Vessel
13－Fuel Handling Machine Bearing Endurance Test in Sodium Vapor－
5. Improved Analysis of Hypothetical Core Disruption Behavior in FBRs
14
－3-D Simulation of Complex Thermal-Hydraulic Phenomena under Very High Temperature－
6. Experimental Study on Molten Core Material Behavior
15－The EAGLE Project as a Collaborative Study with the Republic of Kazakhstan－
7. Analysis of Complex Phenomena upon Tube Failure in SG
16
		 －Development of a Numerical Simulation Program Evaluating Sodium-Water Reaction Phenomena
		for Safety Assessment－
17
8. Development of Uranium Recovery System by Crystallization Technologies－R&D Status on Uranium Crystallization System for Future FBR Recycle System －9. Toward Engineering Scale Minor Actinides Recovery
18
		 －Development of Extraction Chromatography Techniques－
10. Toward Innovative MOX Fuel Production by Simplified Process
19－Developing a Unified Process Combining De-Nitration by Microwave Heating with Granulation－
11. Heat Removal during TRU Fuel Fabrication
20－Design Study of Heat Removal for TRU Fuel Fabrication System－

Research and Development on Geological Disposal
2 of High-Level Radioactive Waste
R&D Supporting the Technology and Reliability of Geological Disposal in Japan21

1. For the Utilization of Knowledge on Geological Disposal Technology－Detailed Design of Knowledge Management System－
2. Development of Buffer Material Database－Compilation of Characteristic Data of Buffer Material－
3. Estimation of Long-Term Metallic Corrosion Using Archaeological Sample－Application of Analogize Phenomena in Nature on Geological Isolation Research－
4. Realistic Application of Safety Assessment Methodology to Disposal Environments－Arrangement and Development of Basic Information Using FepMatrix Tools－
5. Development of Knowledge Base for HLW Disposal Technology－Technical Know-How and Analysis of Decision Criteria for Characterization of Geological		Environment－
6. Development of Quality Assessment Method for Hydrochemical Data
－Proposal of Methodology Based on Evidence Support Logic－
7. Study of Features of Active Faults Based on Measurement of Hydrogen Gas Emission from		 the Underground－Development and Application of Portable Hydrogen Gas Equipment－
8. Strategy for Long-Term Predictions for Geological Disposal R&D－Long Term Prediction of Groundwater Flow Changes Based on the Geological History－

22
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25
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J AEA R&D Review
9. Tracing Water Movement from Surface to Underground－Surface Hydrological Investigation at Horonobe Area, Hokkaido, Japan－
10. Tunneling Underground to Investigate Initial Stress Conditions－3-D Initial Underground Stress Measurement by Hydraulic Fracturing－30
31

3 Nuclear Fusion Research and Development
For Practical Use of Fusion Energy
1. Establishment of Long-Pulse Operation for “ITER” Burning Plasmas
		 －Steady State Sustainment of High Confinement and High β Plasmas－
2. Intrinsic Rotation Due to Plasma Pressure Discovered
		 －Toward the Understanding of High Pressure / Highly Self-Regulating Plasma－
3. Development of the Plasma Movie Database System for JT-60
		 －Movie Data Handling Technology for Remote Experiment in the ITER-BA Project－
4. Achieving High Power in the JT-60U RF System
		 －World Record for Gyrotron Oscillation (1.5MW,1s)－
5. World Record Achievement in Plasma Heating Device
		 －Demonstrated High-Power and Steady-State Operation in Gyrotron toward ITER Construction－
6. Development of a Large Negative Ion Source
		 －R&D for Maintenance Free Ion Source for Uniform Negative Ion Beam－
7. Demonstration of Final Installation to within 0.5mm of a 4-ton Blanket by Remote Handling		 －Improvement in a Robotic Positioning Control Scheme for ITER－
8. Materials Necessary for Stably Supplying Fuel Tritium to Fusion Reactors
		 －Development of Tritium Breeding Materials Resistant to Reduction in Hydrogen－
9. Tritium Extraction Using a Fuel Cell
		 －Tritium Recovery from Blanket Sweep Gas－
10. Torus Configuration and Maintenance Concept of Compact Fusion DEMO Reactor
		 －Torus Configuration for High Operation Rate－

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35
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38
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40
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4 Quantum Beam Science Research
The Future Explored by Quantum Beam Technology

43

1. Making a Mirror Moving Nearly at the Speed of Light by Lasers
44
		 －Demonstration of Light Reflection and Frequency-Upshifting by Electron Density		Modulation in Plasmas－
2. Observation of a Key Hydrogen Atom in an Enzymatic Reaction
45
		 －Neutron Structure Analysis of a Compound Modeling the Reaction Center of [NiFe]Hydrogenase－
3. Why Do Heavy Ions Have Great Effect on a Cell? Chemistry Elucidates the Reason46
		 －Quantitative Analysis of Reactive Species Which Induce DNA Damage－
4. Biodegradable Elastic Gel from Cellulose Derivatives
47
		 －Combination of Radiation-Induced Crosslinking and Aggregation－
5. Medical Applications for the Laser-Driven Proton Beam48
		 －High Quality Beam Is Produced with High Conversion Efficiency－
6. Local Structure in a Material Shrinking with Heat49
		 －Broadening of Magneto-Volume Effect (Invar Effect) Caused by Local Lattice Distortion－
7. Observation of Electron Dynamics Utilizing Synchrotron Radiation X-ray
50
		 －Charge Excitations in Ladder Cuprates by Resonant Inelastic X-ray Scattering－
8. Observation of Curious Long-Period Structures in Metal Hydride51
		 －A Clue to Understand the Hydrogen-Metal Interactions and Their Role－
9. A Breakthrough in Development of High-Temperature Polymer Electrolyte Membrane Fuel Cells
52
		 －Successfully Making a Highly Heat-Resistant Proton-Conductive Aromatic Polymer Electrolyte		Membrane－

JAEA R&D Review 2008

10. Strain Measurement on Rebars in Reinforced Concrete
		 －Application of Neutron Diffraction to Nondestructive Strain Measurement－
11. Compact and Low Noise Ultra-High Intensity Laser
		 －Good Prospects for a Laser Driven Proton Accelerator for Cancer Therapy－
12. Mechanism for Biological Sensing of Ionizing Radiation
		 －The Role of γ Subunit of G Protein in Caenorhabditis elegans－

53
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55

5 Nuclear Safety Research
To Support Safety Regulations and to Ensure Nuclear Safety and Confidence56

1. To Learn from Operating Experience
		 －Analysis of Primary Water Stress Corrosion Cracking Events at PWRs－
2. Simulation of MOX Fuel Irradiation Behavior for LWRs
		 －Analysis of MOX Fuel Behavior in Halden Reactor by FEMAXI-6 Code－
3. Does Pellet Microstructure Change Affect Fuel Behavior during a RIA?
		 －Fission Gas Release in High Burnup BWR Fuel under RIA Conditions－
4. Reactor Instruments to Prevent PWR Severe Accident
		 －LSTF Test to Estimate Usefulness of Core Exit Thermocouple (CET)－
5. Measurement of Rapid Boiling and Vapor Void Behavior in Coolant among Fuel Rods
		 －Toward Improved Prediction of Power Increase during RIAs－
6. Estimating Fracture Resistance of Reactor Pressure Vessel Steels
		 －Estimating Toughness against Combination of Cleavage and Intergranular Fracture－
7. Taking Advantage of Spent Fuel Being Harder to Make Critical than Fresh Fuel
		 －Acquisition of Verification Data for Burn Up Credit Analysis Methods－
8. Evaluating the Radioactivity Concentration of Radioactive Wastes below Which They Can Be Disposed
		 －Dose Calculation to Derive the Upper Bound of Radioactivity Concentration in Disposal of		Transuranium Wastes－
9. How Long Does a Repository Confine Radionuclides?
		 －Modeling of Long-Term Alteration of Bentonite Buffer Material－

57
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60
61
62
63
64

65

6 Advanced Science Research
Advanced Basic Research to Create the Future66

1. Chemical Characterization of Superheavy Elements
		 －Fluoride Complexation of Element 104, Rutherfordium－2. Heavy Fermion Superconductivity in Neptunium Compound		 －A New Material Advances Actinide Science－
3. Microscopic Investigations of Uranium and Trans-Uranium Oxides
		 －Nuclear Magnetic Resonance Studies of the Electronic States－
4. Atomic Scale Observation of Metal Degradation
		 －Development of a Positron Scanning Microscope－
5. Biological Effects of Clustered DNA Damage
		 －Mutation Induction by DNA Lesions in Close Proximity－

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69
70
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7 Nuclear Science and Engineering Research
Formation of Basis for Nuclear Energy R&D, and Creation of Innovative Nuclear Energy
Utilization Technology
1. Compilation of Reliable Nuclear Reaction Data
		 －Development of Comprehensive Code for Nuclear Data Evaluation, CCONE－

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J AEA R&D Review
2. Detailed Measurement of Boiling Flow
		 －Monitoring 3D Change of Vapor / Water Distribution by Neutron Beam－
3. Demonstration of Excellent Thermal Properties of Minor Actinide Nitride Fuel		 －High-Quality Thermal Conductivity Data of MA Nitride Obtained from Small Samples－
4. New Approach for Testing Aging of Reactor Structural Materials
		 －Study of Stress Corrosion Cracking (SCC) Mechanisms Using Electron Backscatter
Diffraction (EBSD) in Mesoscale Analysis－
5. Advancement of Nuclear Medicine Dosimetry
		 －Development of a Radionuclide Database Applicable to Microdosimetry in Subcellular		
Dimensions－6. Selective Detection of Ultra-Fine Particles Containing Highly Enriched Uranium
		 －Development of an Environmental Sample Analysis Method for IAEA Safeguards－
7. Improving Performance of Very High Temperature Reactor Fuel		 －Development of ZrC Coated Fuel Particle－
8. For Efficient H2 Production Using IS Process
		 －Success in Measuring SO2 Pressurization Effect on the Bunsen Reaction－

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79
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8 Nuclear Fuel Cycle Technological Development
Promotion of Nuclear Fuel Cycle Development and Technical Co-operation with Industry

81

1. Vitrification Melter Improvement
		 －Development of Long-Life High Level Radioactive Liquid Waste (HLW) Vitrification Melter－
2. Developing to LLW Encapsulation Stage for Disposal
		 －Future Planning of LLW Treatment at the Tokai Reprocessing Center－

82
83

Development of Decommissioning and Radwaste

9 Treatment Technology

Aiming at Establishment of Rational Process for Radwaste Management
1. Determination of 90Sr in Radioactive Waste
		 －Development of Rapid and Simple Method for Evaluation of Radioactivity Inventories in Waste－

84
85

10 Collaboration on Basic Nuclear Engineering
Promotion of R&D to Meet the Demands of Industry and Academia
1. Development of Extra High Purity Stainless Steel (EHP Alloy) with High Corrosion Resistance		 in Severe Environment
		 －Excellent Grain Boundary Corrosion Resistance of EHP Alloy under Boiling Nitric Acid－

86
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11 Photo-Medical Research Cooperation
Aiming at Creating a Photo-Medical Industrial Valley Lasers in Medicine
1. Ion Acceleration from Near-Critical Density Plasma Produced by Lasers
		 －Proposal of a Novel Ion Acceleration Method for Cancer Therapy－
2. One Optical Fiber Can Both Observe and Treat
		 －Development of Extra-Fine Endoscope with Laser Radiation Function－

JAEA R&D Review 2008

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12 Computational Science and E-Systems Research
Atomic Energy Research by Computational Science－Establishment of Advanced Technology for Atomic Energy R&D－

91

1. Computational Nuclear Power Plant Modeling and Structural Integrity Assessment
		 －3-D Vibration Simulator for Predicting Dynamic Behavior of Nuclear Power Plants－
2. Suppressing Bubble-Induced Material Damage by Gas Bubbles
		 －Numerical Study of the Damage Suppression Effect of Bubble Injection－
3. Large-Scale Conformational Change of a Protein Elucidated by Computer Simulation
		 －Molecular Mechanism of Cell Adhesion Protein Integrin－
4. Development of Steel Impurity Behavior Simulation Technology
		 －Estimation of Impurities in Materials Based on Thermal Desorption Gas－

92
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94
95

Scientific & Technical Development for Nuclear
13 Nonproliferation
Development of Technology for Nuclear Nonproliferation, Foundation of Peaceful Useof Nuclear Energy
1. Nuclear Proliferation-Resistance and Safeguards for Future Nuclear Fuel Cycle
		 －Study of the Evaluation Technique and Criteria of Nuclear Proliferation Resistance－

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

96

97

98

1. Evaluation of Radioactive Inventory in Nuclear Facility for Its Decommissioning
102
		 －The Applicability of a Radioactive Inventory Evaluation Technique for “FUGEN” Verified－
2. Achieved Utilization of Short-Lived Radioactive Nuclei Beams
103
		 －Development of Ion Sources for Radioactive Nuclei－
3. Study on Basic Properties for Establishment of a Fast Reactor Cycle System104
		 －Melting Temperatures and Phase Diagram of MOX Fuel－
4. Basic Knowledge on Treating Various Wastes Generated from Pyro-Reprocessing
105
		 －Recovery of Actinides from Solid and Liquid Wastes－
5. Development of Damage Reduction Technique on Mercury Target for the Spallation Neutron Source 106
		 －Visualization of the Bubble Formation Behavior in Mercury－
6. Measuring Displacement and Vibration of High Temperature and High Radiation Components Using 107
		 Optical Fiber
		 －Development of Fast Reactor Integrity Monitoring Technology－
7. Material Research for Transmutation of LLFP in a Fast Reactor
10811
		 －Irradiation Test of B4C as a Neutron Moderator－
8. Contribution to “ITER” by Technical Development of Post Irradiation Examinations
109
		 －Remote-Handling Type Welding / Processing Techniques for Irradiated Materials－
9. Correlation between the Charge-State and Structure of Fragment Ions for Swift C3+ Traversing a
110		 Thin Foil
		 －Study on Interaction between a MeV Cluster Ion and a Matter－
10. Reduction of Water Inflow for Continuation of Deep Shaft Excavation
111
		 －Development of Grouting Technology at the “Mizunami Underground Research Laboratory”－
11. To Accomplish Decommissioning of Uranium Enrichment Plant Safely and Economically112
		 －System Treatment Chemical Decontamination Technology by Fluoride Gas－
12. High Precision Measurement of Dissolved Organic Radiocarbon in Seawater
113
		 －Development of a Pretreatment Method for AMS Measurement of Dissolved Organic Radiocarbon－

JAEA R&D Review 2008

1 Research and Development of Advanced Nuclear System
Steady Implementation of FaCT Project–Assessing Innovative FBR Cycle Technology–
2010

2005

Feasibility Study for commercialized
Fast reactor cycle system
Phase II
（2001〜2005）

Check & Review

Assessment andpolicy of national
government

2015

【JFY】

Fast Reactor Cycle Technology Development
Decision to adopt
judgment of innovativetechnologies(2010)

Assessing feasibility ofInnovative technologies(2015)

Fundamental R&D of innovative technologies

Conceptual design of demonstration / commercial plant

Presentation of Conceptual
design and R&D plan
(2015)

Fast Breeder prototype reactor “MONJU”◆Modification work
◆System function test
◆Cooperation of・Electric utilities
・R&D instituteR&D in MONJU to establish◆Reliability as power generation plant ◆Sodium technology

internal related institute（Central Research Institute of Electric Power Industry etc.）

・Manufacturers
・University◆International cooperation
・GNEP, GEN-Ⅳ, INPRO・Bilateral cooperation(Japan-US, Japan-France, etc)

Fig.1-1 Fast Reactor Cycle Technology Development - FaCT Project -The fast breeder reactor (FBR) cycle is designated as a key
technology of national importance in the The Third Science andTechnology Basic Plan, this being a technology that contributes
to political objectives such as compatibility between theenvironment and economic growth, and the enhancement ofindustrial competitiveness to survive in international markets.
Aiming for the start-up of a demonstration fast reactor (FR)around 2025 and its introduction on a commercial basis before
2050, Japan Atomic Energy Agency (JAEA) is now promoting
its “Fast Reactor Cycle Technology Development (FaCT)”
project, supported by the government’s assessments andpolicy, specifically now carrying out a “Feasibility Study oncommercialized fast breeder cycle system (FS)” (Fig.1-1).
This FaCT project is developing a combination of thesodium-cooled FBR utilizing oxide fuel, advanced aqueous
reprocessing, and the simplified pelletizing fuel fabrication, thetechnologies presently seen to have the best practical prospects
for fostering economic competitiveness, safety, efficient use ofresources, reduction of environment load, and non-proliferation,
based on their performance in existing facilities.
To scales these technologies up to commercial levels, it isnecessary to realize innovative technologies within the above
conceptual framework and to assess their performance. In the
FaCT project, design studies reflecting the R&D results ofthe prototype FBR “Monju” and experimental studies for the
innovating technologies will be implemented in order to decideon which of the innovative technologies to adopt by 2010 andto present conceptual designs of commercial and demonstration
fast reactor cycle facilities along with development plans torealize them by 2015.
FBR System
The commercialization of FBR system still faces challenges;
it needs to improve its economy, safety and reliability.
These challenges are, for example, the development ofhigh-performance materials, structures and components which
will reduce the facility volume and component size, simplify

cooling systems, make the reactor vessel compact, optimize
thermal hydraulic properties, and decisively ensure recriticality
prevention in case of core disruption. The frontline of this R&D
will be introduced in Topic 1-1 to Topic 1-7.
Fuel Cycle System
Major issues facing the fuel cycle system include thesafety and controllability of innovative processes, and thedevelopment of components with outstanding performancein operation, durability, and repairability Issues facing fuelreprocessing include the development of crystallization
technology that enables most uranium in a solution of spent
fuel to be recovered in solid form and a recovering process
for minor actinides. Issues facing fuel fabrication that havejust recently arisen include the development of denitration andconversion techniques, fundamental technology to simplify
the pellets fabricating processes, as well as technologies toremotely treat thermogenic substances. The achievements ofthis R&D will be introduced in Topic 1-8 to Topic 1-11.
Plan for Execution of the Project
A domestic development framework for R&D of thedemonstration reactor, in which responsibility, authority, andengineering functions are concentrated, was established in 2007
to implement effective development. Under this framework,
the R&D and the design study have been performed with thecollaboration of electric utilities and manufacturers. Further, forthe fuel cycle, JAEA has conducted preliminary examinationand investigation a second fuel reprocessing plant from around2010 sponsored by the Japanese government.
To lessen development risks and achieve global
standardization, an R&D collaboration agreement was
concluded by three nations: Japan, France, and the US, forthe sodium-cooled FBR demonstration/prototype reactors.
Furthermore, multilateral international partnerships such as
GNEP are being vigorously pursued.
JAEA R&D Review 2008Research and Development of Advanced Nuclear System

1-1

Coolant Pipes Made Economically Competitive by Shortening–Development of High Chromium Steel for FBR Components Which Has Excellent
High Temperature Strength, Toughness and Microstructural Stability–
Structural material:
High Cr steels

Primary piping:Primary piping: High Cr steels
SUS304

Pump<high Cr steels>
Low thermal expansion
High thermal conductivity
High strength

IHX
Reactor Vessel

Na
Reactor Vessel

Demonstration plant Primary H/L Piping Systemwith a 90°Elbow(〜25m/loop)

Primary sodium temperature529°C

Primary sodium temperature550°C

Fig.1-2 Pipe shortening by employing high Cr steel
SUS304

Normalized to SUS304

SUS316

Mod.9Cr-1Mo

SUS410J3

1.0

0.5

Temperature: 600°C

2.3

Integrated
pump-IHXMONJU Primary H/L Piping Systemwith Eight 90°Elbows(〜100m/loop)

1.5

2.4

Stress σ(log(σ))

Structural material:
SUS304

V and Nb steel(V: 0.12%,Nb: 0.03%)

2.2
2.1
2

V steel(V: 0.18%,Nb: 0.00%)

1.9
1.8
1.0

Nb steel (V: 0.00%,Nb: 0.05%)
2.0

3.0

4.0

5.0

Creep Rupture Time tR(log(tR))

Fig.1-4 Effects of vanadium and/or niobium on creep
strength
Creep strength obtained by V alone is more stable than thatobtained by Nb alone.

Fig.1-3 Material properties of stainless steels and high Cr steels
FBR structural materials require high strength, low thermal expansion, andhigh thermal conductance.
12% Cr steel has high potential as a FBR structural material, because it
has almost the same tensile strength as SUS316 and thermal properties
superior to stainless steels.
Tensile strength
valuesMean thermal
expansion

Thermalconductivity

Applicability of high chromium (Cr) steel as a main
structural material in FBR has been explored to enhance theeconomic competitiveness of the FBR. Since this steel has
excellent high temperature strength and thermal properties, thecoolant pipes can be shortened and a compact plant design canbe achieved (Fig.1-2). Applicability of Mod. 9Cr-1Mo steel
has been investigated for the demonstration plant because thesteel has proven performance in thermal power boilers. For acommercial plant, however, development of innovative highCr steel is required to enhance the economic competitiveness.
Some conventional high Cr steels have properties superior tothose of Mod. 9Cr-1Mo steel, as shown in Fig.1-3. However,such steels cannot be used in the FBR components directly,
because their long term stability at elevated temperature has notdemonstrated yet. In addition, excellent creep-fatigue strengthand fracture toughness are required in the FBR structural
material. Therefore, prior to employing the steels in the FBR
components, it is required that these properties be improved.
Though high temperature strength of the high Cr steel has
been achieved by adding some strengthening elements suchas vanadium (V) and niobium (Nb), long term efficiency andstability of strengthening elements has not been confirmed yet.
Therefore, the long term effectiveness and stability of V and
Nb were investigated.
Generally, both V and Nb contribute to improve the hightemperature strength of high Cr steels by precipitating as finecarbides and/or nitrides (VX and NbX, where X is C or N; allof these termed “MX”). Several trial products in which the V
and Nb contents were varied were made, and mechanical tests
and metallurgical examinations were conducted to investigate
the long term effectiveness and stability of the precipitation
strengthening. As shown in Fig.1-4, both creep strength and longterm stability depend on the V and Nb content. Metallurgicalexaminations revealed that VX was stable after long term creep
and aging. This result shows that precipitation of VX in high Cr
steel is an efficient strengthening method for FBR components.
Based on the above investigations, the stability of precipitation
strengthening can be enhanced by controlling the V and Nb
content.

Reference
Onizawa, T., Ando, M., Wakai, T., Asayama, T. et al., Long Term Efficiency and Stability of MX Precipitation Strengthening of High Chromium Steel(Transformations and Microstructures), Tetsu To Hagane, vol.94, no.3, 2008, p.91-98 (in Japanese).

10

JAEA R&D Review 2008

Research and Development of Advanced Nuclear System

1-2

Evaluation of the Structural Integrity of Reactor Vessels–Measurement of the Thermal Deformation near the Fluid Surface
Using the Laser Speckle Method–

Coil for induction
heatingSpecimen
Movement

Laser (green)

Temperature (°C)

700

① Temperature rising
600 ② Movemet③ Temperature fall
500
400
300
200
100
0

Laser (red)

0

50

100

150

200

250

300

Axial coordinte Z (mm)Observed speckle patternFig.1-5 Laser speckle system and its measurement ofa specimen

Axial component of elastoplastic strain (%)

Fig.1-6 Time history of temperature distribution along
the axis of the specimen

CCD camera

2.0%
Experimental Result
Simple Inelastic Model
Detailed Inelastic Model

1.5%

Elastic Model
Reference:
Detailed Inelastic Model(best -fit analysis)

1.0%

All of these Inelastic model
are not standardized now

0.5%

0.0%
0

5

10

15

20

Number of the temperature cycles

Fig.1-7 Comparison of the strain found in experiments
and analyses (Z=150mm in Fig.1-6)

A main vessel in the FBR is subjected to a severe cyclic
thermal load due to cyclic movement of longitudinaltemperature distribution with a large gradient. This is generated
by the combination of heating of the lower part by hot liquidsodium and cooling of upper part by ambient gas during startup and shut down of the system. As a result, inelastic strain
accumulates near the sodium liquid surface. Since excessive
inelastic strain breaks down the structural integrity, precise
prediction of this strain is important for structural design ofthe reactor vessels. To understand the actual behavior of theinelastic strain and so evaluate an inelastic strain analysis, testsof the strain on a reactor vessel model were carried out.
Fig.1-5 shows the tested cylindrical specimen (material:
316FR steel, height: 300mm, outer diameter: 55mm, thickness
3.5mm). This specimen was given the temperature distribution
shown in Fig.1-6, which imitates an actual reactor vessel. In
this experiment, the following factors were measured: (1)temperature, (2) radial deformation, and (3) axial strain. In
the past it has been difficult to measure local strain on a hightemperature structure with high precision. For the measurementof this strain, we developed a non-contact measurement system

based on the laser speckle method. This system measured the
speckle pattern (Fig.1-5 lower right) of diffuse reflection ofa laser. Through the image processing of the change of thispattern, the strain on the structure surface can be estimated.
The experiments showed that movements of the longitudinaltemperature gradient caused axial strain, especially in theearly stage. With succeeding repetitions of the temperature
movement, less incremental axial strain is caused. As shown in
Fig.1-7, upon 3 or more cycles of the temperature movement,
the traditional elastic analysis overestimates the axial strain.
Although the elastic analysis is on the safe side, the strain
prediction should be based on inelastic analysis to rationalize
the reactor vessel design.
The strain measured here is important and useful for
improvement of inelastic analysis needed to make the structuralmodel. In the future, using this data we will develop a moredetailed model to explain actual behavior such as temperaturedependency, non-linear hardening, and cyclic hardening. Thismodel will improve the prediction of the inelastic strain, andenable design of compact reactor vessels.

Reference
Watanabe, D., Chuman, Y., Otani, T., Shibamoto, H., Inoue, K., Kasahara, N., An Experimental Validation of the Guideline for Inelastic Design Analysis
through Structural Model Tests, Nuclear Engineering and Design, vol.238, issue 2, 2008, p.389-398.
JAEA R&D Review 200811

Research and Development of Advanced Nuclear System

1-3

Determining Transfer / Damping of Temperature
Fluctuation of Fluid to Structure
Non-dimensional Temperature;(T-Tc)/ΔT–Evaluation of High Cycle Thermal Fatigue at Core Outlet in FBR–
Fluid
Structure Surface
Inside Structure

1.0

Cold

Hot

Nusselt number; Nu=h・D/λFig.1-8 Visualized convective mixing ofjets
Only hot jets are colored. The jets oscillate
as they mix together.
1,000
Experimental results
100
Steady State Correlation

10

Pe・D 7/15
Nu =0.638 
x 

1
100

1,000

10,0001.5mm
Stainless
Steel

0.5

0.0

0.0
0.5
1.0

Time (s)Hot

0.5mm
LiquidSodium

1.5
2.0

Fig.1-9 Attenuation of temperature fluctuation during transfer from fluid
to structure
The vertical axis is normalized to the discharged temperature difference
between jets.
(T: fluid temperature, Tc: temperature of cold jet at nozzle exit, ΔT: discharged
temperature difference)
Fig.1-10 Comparison of heat transfer correlation between steady stateand non-stationary conditions
The Vertical axis is the Nusselt number, the ratio of the heat convection
to the heat conduction. The horizontal axis is the Peclet number, theratio of heat transportation based on flow velocity to thermal diffusivity.
(h: heat transfer coefficient, D: representative length, λ: thermal conductivityof fluid, V: velocity of jet discharge, a: thermal diffusivity in fluid, x: distancefrom nozzle exit)

100,000

Peclet number; Pe=V・D/aAt the core outlet of an FBR, hot fluid is mixed with cold
fluid near the reactor structure. Thermal stress is induced dueto the thermal expansion of the structure where the temperaturefluctuation is transferred from the fluid. A large temperaturefluctuation may cause thermal fatigue damage if applied several
million times (high cycle thermal fatigue). In a previous study,
the relation of the frequency of temperature fluctuation to thevulnerability of a structure to damage was evaluated under
special experimental conditions in which this fluctuation could
be controlled (see JAEA R&D Review 2007 Topic 1-8). In
this study, a sodium jet experimental system was constructed
to evaluate the characteristics of the transfer of temperaturefluctuation from fluid to structure.
The experimental test apparatus had three parallel jets passing
along a stainless steel wall. The center was a cold jet, with hotjets on both sides. This experimental setup models a simplified
two dimensional configuration of the reactor core outlet
which has cold flow channels of control rods surrounded by
hot channels of fuel subassemblies. Fig.1-8 shows convective
mixing of the three jets obtained from a water experiment withthe same geometry as the sodium experiment. It was observedthat the jets at either side inclined toward the center jet. The
temperature fluctuation was large at the region where the centerjet contacted with the side jets. In the sodium experiment,
thermocouples were embedded in the wall at this region. The

transfer of temperature fluctuation from fluid to structure was
evaluated by the simultaneous measurement of temperaturesin fluid and structure. Fig.1-9 shows the temperature histories
in fluid near the structure, on the structural surface, and in the
structure. The fluid temperature fluctuated with large amplitude.
On the structural surface, on the other hand, the amplitude oftemperature fluctuation was much smaller than that in fluid.
Furthermore, the fluctuation amplitude in the structure was still
smaller than that on the structural surface.
This quantitative evaluation of the attenuation makes it
possible to efficiently evaluate high cycle thermal fatigue.
This attenuation had dependency on temperature fluctuation.
This was expressed as a theoretical power curve function. We
developed a new method of determining the transfer coefficientof temperature fluctuation from fluid to structure by fitting
measured results to this function. Fig.1-10 shows a comparisonof the calculated heat transfer coefficients and the heat transfer
coefficients under the steady state condition (Poppendiek’scorrelation equation). It was found that the transfer coefficientof temperature fluctuation could be evaluated using a constantheat transfer correlation equation including the flow velocity
dependency (Peclet number dependency). Further experimental
study using a mockup model is planned to develop an evaluation
method of the high cycle thermal fatigue, and to investigate
structural integrity.

Reference
Kimura, N. et al., Experimental Investigation on Transfer Characteristics of Temperature Fluctuation from Liquid Sodium to Wall in Parallel Triple-Jet,
International Journal of Heat and Mass Transfer, vol.50, issues 9-10, 2007, p.2024-2036.

12

JAEA R&D Review 2008

Research and Development of Advanced Nuclear System

1-4

Development of a Fuel Handling Method for a
Compact Reactor Vessel–Fuel Handling Machine Bearing Endurance Test in Sodium Vapor–25
small-scale test(flaking)Deviation (Ǵm)

20
small-scale test(no flaking)15

full-scale test(no flaking)

10
5
0
1200

1400

1600

1800

2000

Contact Pressure (MPa)Fig.1-11 Bearing ring roundness(Over View)(Magnified Surface)Fig.1-12 Full-scale fuel handling machine bearing(Over View)(Magnified Surface)Fig.1-13 Bearing ring after endurance test

Fig.1-14 Bearing ball after endurance test

Our commercial-use sodium cooled reactor design includes
a compact reactor vessel with a new in-vessel fuel exchangesystem. The upper inner core (UIS) structure has a slit for afuel handling machine (FHM). The FHM removes and inserts
each fuel subassembly through the UIS slit without removing
the UIS from above the core.
The FHM has to function in sodium and in the gas inside
the reactor vessel cover including sodium vapor. The usual
lubrication materials such as grease are not usable with sodium,
since sodium is chemically active and the greases could
pollute coolant sodium. Therefore the FHM bearing requires
a special lubrication. Besides, the FHM uses ball bearings,
to avoid interference of the UIS slit with the FHM during anearthquake. Since ball bearings generally are subject to highcontact pressure, endurance of the FHM bearings is one of theissues for the FHM development.
In this study, 1/10-scale and full-scale bearing endurance
tests were conducted. In the 1/10-scale test, several ring and ballmaterials were tested to find a suitable material combination
for the FHM bearing. The relation between endurance andcontact pressure was determined. The 1/10-scale tests wereconducted at 250 ºC, applying 20000 vibration cycles in the
air. The temperature was a little higher than the FHM operation
temperature and the number of vibrations was equivalent to
ten refueling operations. Based on the 1/10-scale test results,

a bearing with stainless steel rings and ceramic balls was
selected. To enhance lubrication, MoS2 coating on the bearing
rings was adopted. The results also showed that surface flaking
would start at a contact pressure of approximately 1800 MPa.
Deviation of the inner ring grew larger as the contact pressure
became larger. When the deviation went beyond 3.5 to 7μm,
the deviation increment rapidly grew larger, as shown in
Fig.1-11.
In the full scale test, a full-scale FHM bearing was
manufactured and an endurance test was conducted inargon gas with sodium vapor simulating the FHM operating
conditions. The full-scale bearing is shown in Fig.1-12. The
contact pressure was controlled to be 1745 MPa, considering
the surface pressure that can be maintained. The other test
conditions are similar to the 1/10-scale tests. Fig.1-13 andFig.1-14 show surface conditions of rings and balls after tests.
There is sign of some MoS2 transfer between ring and ball but
overall surface conditions were maintained well without any
flaking. The deviation of the inner rings was observed to be
2μm, which is lower than the limit suggested by the 1/10-sacle
tests. Those results show that this bearing is suited for the FHM
operating conditions.
In future studies, a full scale FHM test in the air and a gripper
mechanism test in sodium are planned as part of JSFR FHM
development.

Reference
Chikazawa, T. et al., Development of a New Fuel Handling Machine Suitable for an Upper Inner Structure with a Slit-The Performance Test of a Large-Sized
Bearing in Argon Gas Atmosphere-, JAEA-Research 2007-001, 2007, 91p. (in Japanese).
JAEA R&D Review 200813

Research and Development of Advanced Nuclear System

1-5

Improved Analysis of Hypothetical Core Disruption Behavior in FBRs
–3-D Simulation of Complex Thermal-Hydraulic Phenomena under
Very High Temperature–(a)Conventional analysis (R-Z 2-D)

Realistic (3-D)

Up/downward dispersion
Fall due to gravity

Fuel dispersion by fission
gas and steel vapor

Fuel dispersion in CRGT failures
CRGT failures(SIMMER-IV)

Reactivity ($)

2-2

Asymmetric sloshing prevents
fuel compaction(b)

Fuel accumulationin the core

-4
-6
0s

Symmetric fuel compaction at center axis
→Overestimation of inserted reactivity

Inward fuet
compaction

0

SIMMER-III
1s

SIMMER-IV
2s

3s

4s

5s

Fig.1-16 Comparison of reactivity histories
2-D (SIMMER-III) analysis predicts recriticality due to fuelcompaction, while 3-D (SIMMER-IV) analysis predicts
reactivity decrease due to fuel discharge from the core.

Fuel, motion

Vaporization

1s

4s

Clad or wall
Liquid fuel
Fuel particle
Broken pellets

Liquid steel
Steel particle
SodiumGas

Molten fuel/steel pool(more than 3000°C)
Local fuel discharge
CRGT with ring shape is unrealistically treated as massive fuel discharge
path, while actually it prevents fuel compaction by allowing radial sloshing

7s

Discharge

Fig.1-15 Need for a three-dimensional analytical method
3-D analytical method enables realistic simulation.
(a) Reactivity insertion by radial sloshing of whole-core
scale(b) Fuel discharge from the core through control rod guidetubes (CRGTs)

Fig.1-17 Core material dynamic behavior (SIMMER-IV)
3-D core-disruption behavior accompanying melting/freezing,
vaporization/condensation, etc. at around 3000°C.

One of the major issues in FBR development is dealing
with a core disruptive accident (CDA), which might rapidly
release radioactive materials due to its thermal and mechanicalconsequences, although its occurrence frequency is negligibly
low. A high-temperature molten fuel above 3000°C allows
forming molten fuel pool, where molten core materials canmove between fuel assemblies with molten wrapper tube wall.
A safety analysis code, SIMMER-III, has been utilized
for past safety evaluations. This code has been developed inJapan for a multi-phase, multi-component fluid-dynamics
model coupled with structural materials models and neutronics
models. SIMMER-III, however, is a two-dimensional code.
Due to this dimensionality, this code tended to overestimate
the reactivity increase resulting from fuel compaction into
the core center because it could not simulate motion aroundthe axis (Fig.1-15). Further, the reactivity reduction by fueldischarge through control rod guide tubes (CRGTs) could not
be calculated, because the tubes which are dispersed throughout
the core could not be simulated. This deficiency has forced

the past evaluations to be conservative, since fuel discharge
through CRGTs etc. had to be ignored. To reasonably reducethe conservativeness of the evaluation, we have developed the
SIMMER-IV code, extending the SIMMER-III framework to
three-dimensions. This code clarifies the significance of 3-D
simulation by comparing it to the conventional 2-D simulation.
In this study, SIMMER-IV has been applied to 3-D simulationof a CDA in an FBR for the first time in the world.
In the SIMMER-III analysis, after output power exceeds
the limit for 2.4s, the fuel melts and increases in mobility, andthe reactivity reaches prompt criticality in 3.9s due to inward
fuel compaction. In the SIMMER-IV analysis, CRGTs failed
at different times around 2s cooling the fuel so that it did not
become a mobile pool, so that the fuel was not compacted
and reactivity did not rise (Fig.1-16). 3-D motion of complex
phenomena with phase changes, etc. could be simulated(Fig.1-17).
This study has shown the effectiveness of the 3-D simulationin making reasonable evaluation of accidents.

Reference
Yamano, H. et al., Development of a Three-Dimensional CDA Analysis Code: SIMMER-IV and Its First Application to Reactor Case, Nuclear Engineeringand Design, vol.238, no.1, 2008, p.66-73.

14

JAEA R&D Review 2008

Research and Development of Advanced Nuclear System

1-6

Experimental Study on Molten Core Material Behavior–The EAGLE Project as a Collaborative Study with the Republicof Kazakhstan–
Image of molten-corematerial movement toward
the surrounding region

Reflector(graphite)

Reactor vessel

Reactor Hall

IGR

900
TD1
TD2
TD3

600
300
28.7

28.9 29.1
Time (s)

29.3

TT1
TT2

TT3
TT5

Temperature (°C)

1800
1500
1200

TD4
TD5
29.5

TCs
failureDuct failure
initiating fueldischarge

Fuel-Namixing in theplenum region

Fueljacket

900
600
300
0
28.7

Test fuel28.75 28.8
Time (s)28.85

Estimated velocity of
Na/void interface:~27m/s

300
28.9 29.1
Time (s)100 ms

1200

Duct

600

28.7

1500

Downward movement of
hot material

TCs failure

900

Temperature (°C)

Temperature (°C)

100 ms

1200

0
28.5

8kg-scale
UO2 melting

1800

1800

0
28.5

Experimentally observed event sequence
Simulation ofpartial molten
core region andthe surroundings

Fig.1-18 IGR and test channel

1500

Core

Water

Test Channel

Core

Water

Control rods

29.3

29.5

Estimated velocity of the
hot-material leading edge:~31m/s

TD5
TD4

TT5

Trap
TT3
TT1

TD3
TD1
TD2
TT2

Coolant (Na)Duct structure made ofstainless steel (40mmǾ)

Fig.1-19 Concept of molten core accident
simulation

Fig.1-20 Example of data obtained from EAGLEJAEA is conducting an experimental project called EAGLE
in order to clarify molten-core-material behavior in hypotheticalcore-melt accidents of FBR and to confirm that the consequencesof such an accident can be mitigated appropriately, although
the possibility of such accidents is extremely small. This
experimental program is a collaboration between JAEA andNational Nuclear Center of Republic of Kazakhstan using
its IGR (Impulse Graphite Reactor) specifically designed forexperiments.
IGR has a central hole as illustrated in Fig.1-18 toaccommodate the test channel equipped with double-layerpressure vessels enclosing fuel for testing. Three experiments
in which about 8kg of UO2 is melted have been performed up
to now simulating accident conditions. These tests provided
information on accident progress behavior (Fig.1-19). Fig.1-20
shows an example of the data obtained from the experiments.
In this experiment, molten fuel broke through the stainless steel(SS) duct structure and discharged through the duct toward thecold region filled with sodium (downward in this experiment),
and was cooled down through mixing with sodium. Besides

thermocouple (TC) data shown in this figure, various data fromvoid-sensors (detecting bubbles in sodium), pressure gauges,
microphones, etc. were obtained, and they elucidated the seriesof physical processes which occurred, such as the duct-structure
failure, contact with molten fuel pushing sodium downward
from the core (TCs at upper, middle and lower levels detected
temperature increases with time lags with this order) and thefuel discharge through the duct.
In the conventional safety evaluation of FBR core-melt
accidents, it was assumed that a large amount of molten-fuelremained within the core region which could potentially cause
a rapid power excursion with massive molten-fuel movement.
However, through the EAGLE experiments, it was shown thatmolten fuel tends to discharge out of the core region so thatmassive amounts of molten fuel remaining within the coreregion is very unlikely. The new findings obtained throughthe EAGLE project indicate that the conventional safety
evaluations are conservative, and that core design featuresenhancing fuel-discharge characteristics can prevent severe
power excursions.

Reference
Konishi, K., Toyooka, J., Kamiyama, K., Sato, I. et al., The Result of a Wall Failure In-Pile Experiment under the EAGLE Project, Nuclear Engineering and
Design, vol.237, issue 22, 2007, p.2165-2174.
JAEA R&D Review 200815

Research and Development of Advanced Nuclear System

1-7

Analysis of Complex Phenomena upon Tube Failure in SG–Development of a Numerical Simulation Program Evaluating
Sodium-Water Reaction Phenomena for Safety Assessment–Na(liquid)

Water

Gas-liquid interface
H2● Pressurized water(vapor) leaks from afailed heat transfer
tube● The high-temperatureand high-velocity
reaction jet may
cause wastage of theadjacent tubes

Failed tube

H2O

Na(gas)

NaOH

NaOH

H2
H2O

SodiumSurface reaction = reaction at interface
between water vapor and liquid sodium

Heat transfer tubes in SG

Fig.1-21 Sodium-water reaction phenomena upontube failure

Fig.1-22 Two chemical reaction mechanisms(weight averaged)
Cylindrical vessel
Diameter: 0.4m
Height: 1.8m
Simulated tube
bundles (43tubes)
Water vapor leaks
from the lowest tube
and goes upward
Conditions of water
vapor:
17.0MPa, 352°C
Conditions of
sodium:0.2MPa, 470°C(a) Analysis region

Gas phase reaction = reaction
between water vapor and gas sodium(measured)

Water
vapor goes
upward(°C)
1400(kg/m3)3

1

Nozzle0
Void fraction

0400

NaOH concentration

Vertical plane in tube bundle region
White circle: tubes
Scale of NaOH concentration: Mass of
NaOH per unit volume (kg/m3)(b) Numerical results

Calculation

Experiment

Vertical plane in tube bundle region
Black circle: simulated failed tube(c) Calculated and measured temperature

Fig.1-23 Numerical analysis of sodium-water reaction experiment

When pressurized water leaks from a failed heat transfer
tube in a steam generator (SG) of a sodium cooled fast breeder
reactor, the water vapor reacts with the liquid sodium outside
the tube. It is known that in some cases the high-temperatureand high-velocity reaction jet may cause wastage of theadjacent tubes (Fig.1-21). Evaluation of these phenomena
under anticipated conditions is a key issue in safety assessmentof the SG. In this study, a numerical analysis code SERAPHIM
analyzing compressible multi-phase flows with sodium-waterchemical reaction has been developed.
A multi-fluid model was adopted to calculate multiphase flows of water, liquid sodium and multi-component
gas. Advection and diffusion of the gaseous species is also
considered in our method. Two chemical reaction mechanisms

were modeled (Fig.1-22). One is a surface reaction modelfor the reaction at the interface between the water vapor andthe liquid sodium. The surface reaction rate is assumed tobe diffusion-controlled in our model. Another is a gas-phasereaction model for the reaction between the water vapor andthe gas sodium generated by evaporation. In this model thereaction rate constant is estimated based on the Arrhenius law.
Applicability of the SERAPHIM code has been validated
through the analysis of an experiment on the sodium-water
reaction (Fig.1-23(a)). By using the SERAPHIM code, profilesof void fractions, concentrations (Fig.1-23(b)), and temperatureand velocity fields can be obtained. As shown in Fig.1-23(c),the temperature field calculated by the SERAPHIM code shows
good agreement with the experimental result.

Reference
Takata, T., Yamaguchi, A., Uchibori, A., Ohshima, H., Numerical Investigation of Sodium-Water Reaction Phenomenon in a Tube Bundle Configuration,
Proceedings of 2007 International Congress on Advances in Nuclear Power Plants (ICAPP 2007), Nice, France, 2007, Paper 7510, 8p., in CD-ROM.

16

JAEA R&D Review 2008

Research and Development of Advanced Nuclear System

1-8

Development of Uranium Recovery System by
Crystallization Technologies–R&D Status on Uranium Crystallization System for Future FBR Recycle System–

Solid
Separation

WashedCrystalCrystalWashing

Rinsedsolution

Filtration

Fig.1-24 Steps in uranium crystallization from uranium and ruthenium nitrate solution
The solution in the vessel is cooled to lower temperature as e.g. 0°C. The crystal and mother solution are separated by suction
filtration by a glass filter, the crystal is washed to remove mother solution remained.

Fig.1-25 Development ofEngineerring Scale Crystallizer
Crystallizer tank (about
300mm(ID) × 900mm(L),hold-up volume is about 5
liters) has a cooling jacket
made of transparent plastic
to allow internal observation.
The crystal size is about
700 to 800μm.

The crystallization is considered one of the key technologies
for uranium recovery from the spent fuel in the future
reprocessing plant, because it promises much desired safety
and economical efficiency.
The crystallization is a separation technology based on
differences in solubility, executed by changing temperatureand concentration, with no organic reagent needed, so that the
separation system can be simple (Fig.1-24) .
Since 1990s, JAEA has been carrying out small scale hot
experiments at the Chemical Processing Facility (CPF), anddevelopment of an engineering-scale system is being carriedout.
In the hot experiments, the uranium recovery rate, thedecontamination factors of plutonium, and fission products(FPs) have been confirmed, and the crystal purification
technology have been investigated.
In research for practical application, we selected a rotary
driven type crystallizer which has the advantages of criticality
safety, high throughput, and remote maintenance, etc. (Fig.1-25).

(a)

(b)

Fig.1-26 Appearanceof non-steady state
simulation test(a) Crystal blockade isintentionally causedby reducing therotation speed.
(b) The crystal blockage
is removed bymanually.

The uranium crystal is generated in the cooled solution in the
crystallizer, and the crystal is isolated from residual solution(mother solution) by the cylinder rotation and is discharged
from the device.
In the system development of the engineering scale system,
we fabricated an engineering scale crystallizer system, and havebeen investigating the properties of steady and non-steady state
operations. In these investigations, the operation conditions forstable crystallization, the crystal accumulation and blockage
phenomena during crystal discharge, and the outlet of mothersolution are being confirmed. The crystal accumulation canbe monitored by a cylinder torque meter, and the event canbe overcome by the stoppage of feed solution and stepwise
increment of the screw speed (Fig.1-26). Development of the
instrumentation and control system also is progressing.
This work was financed by the Ministry of Economy Tradeand Industry of Japan (METI), and the Ministry of Education,
Culture, Sports, Science and Technology of Japan (MEXT).

Reference
Ohyama, K., Nomura, K. et al., Development of Uranium Crystallization System in“NEXT”Reprocessing Process, Proceedings of International
Conference on Advanced Nuclear Fuel Cycles and Systems, Boise, USA, 2007, p.1461-1466, in CD-ROM.
JAEA R&D Review 200817

Research and Development of Advanced Nuclear System

1-9

Toward Engineering Scale Minor Actinides Recovery–Development of Extraction Chromatography Techniques–
HLLW*(containing MA)①

EluantWashingsol.

②③* High Level Liquid Waste
Conc.

N

P
O

O

Extractant(e.g. CMPO (Octyl (phenyl)-N,
N-diisobutylcarbamoylmethylphosphine oxide))

ColumnPacked intoa column
25μm
SiO2-P (SiO2 support coated withstyrene-divinylbenzene polymer)
adsorbents impregnated with CMPO

①’

②’

③’

HLLW*

MA
product

Washing
waste

Time
①’
②’
③’
FP** MA FP**(nonadsorbable) (adsorbable)** Fission Products

Fig.1-27 MA recovery by extraction chromatography
SiO2-P adsorbent material impregnated with an extractant is packed into a column. Solution containing MA (①) is fed into thiscolumn, and MA is adsorbed. The adsorbed MA is eluted by a suitable eluant (②), and recovered into a solution (②’). Several
columns with different adsorbents (extractants) may be used for better separation of MA from other elements.
1.6

Sensor inlets

Adsorbents
inlets & outlets

Linear Velocity (mm/s)

1.4
1.2
1
0.8
0.6
Flowrate:121 cm3/s (downstream)

0.4

Flowrate:121 cm3/s (upstream)

0.2

Flowrate:181 cm3/s (downstream)0

Water jackets

0

100

200

Distance from column wall (mm)Fig.1-28 Engineering scale column
Stainless steel column with 48cm inner diameter, which has
water jackets for temperature control and sensor inlets forexperiments.

Fig.1-29 Linear velocity in a column
The uniform flow through a column with 48cm diameter canbe obtained with flow of 121cm3/s (downstream).
About 1.7kg of MA is supported to be recovered in thisflowrate condition.

Studies on minor actinide (MA) recovery by solvent extraction
have been carried out in several countries. These processes use
extractants with diluents, which cause various kinds of liquid
waste in large amounts. In the extraction chromatography
technique, in which extractants are impregnated into support
particles, no diluent is used, and higher MA loading canbe achieved, so this is potentially more economical. We
developed a process and system of MA recovery by extraction
chromatography which can be expanded to an engineering
scale.
We selected a SiO2 support coated with styrenedivinylbenzene polymer (SiO2-P) (Fig.1-27), which has
superior safety and handling properties, and started assessing
SiO2-P adsorbent materials impregnated with several kindsof extractants, e.g. CMPO; adsorption/elution of MA, acidand radiation resistance, and ease of after-treatment of these
adsorbents. These properties are now being measured, andseveral process flows for MA recovery will be designed andevaluated based on these data.
In our development of an engineering scale system, wefabricated an engineering scale column system (Fig.1-28)
and have been investigating the fluidics and thermal control
properties in actual operation, and the durability through longterm operation. Through these investigations, we determined
the operational conditions necessary for uniform flow throughthe column, which is indispensable for steady MA recovery(Fig.1-29).
We also investigated remote control and instrumentation forthe column operation, and we will have overall assessmentsof separation performance, safety, instrumentation and control,
and remote handling based on these results and engineering
scale experiments.
This work was financed by the Ministry of Education,
Culture, Sports, Science and Technology of Japan (MEXT)
under the framework of “The Development of Innovative
Nuclear Technologies.”

Reference
Koma, Y., Watanabe, S., Sano, Y. et al., Extraction Chromatography for Am and Cm Recovery in Engineering Scale, Proceedings of 3rd International
ATALANTE Conference (ATALANTE 2008), Montpellier, France, 2008, O1-19, p.8, in CD-ROM.

18

JAEA R&D Review 2008

Research and Development of Advanced Nuclear System

1-10 Toward Innovative MOX Fuel Production by Simplified Process–Developing a Unified Process Combining
Microwave Heating with Granulation–
Present Process

Recovered MOX

UO2 powder

De-NitrationbySimplified Pellet Fuel Process

Pu/U Solution

Pu EAS*M㧴 conversion

M㧴 conversion/Binder-less Granulation

Calcination / Reduction

Calcination / Reduction

MOX powder (50%Pu)

MOX powder

Refill*Pu EAS
Pu Enrichment
Adjustment
in Solution

RefillLot Blending

Weighing

Cross Blending

Ball Milling
Binder Mixing

Reduction of powder
treatment processes etc

Granulation
Dissolution

Lubricant Mix.

Effect of Pu EAS

Pressing
Pre-waxing
Sintering
Crashing / Milling

De-gassing

Effect of Binder-less Granulation

MOX hollow pellet
Pressing

Recovered MOX

Effect of Die-lubrication pressing

Sintering

Grinding

Grinding

Density and Appearance Inspection

Density and Appearance Inspection

Inspection by Gov.

Inspection by Gov.

Product Pellet

Product Pellet

Lot Mixing

ǩ-Auto radio

Fig.1-30 Simplified pellet fuel production process
Simplified pellet fuel making process (right) compared with present process (left)

The key for the industrial production of MOX fuel to beburned in the FBR is the reduction of its manufacturing cost.
The most promising method for meeting this requirement is thesimplified technology shown in Fig.1-30. The key ideas behind
this method are (1) adjustment of mixing rate of PuO2/UO2 inthe liquid state, (2) combining de-nitration with granulation,and (3) fabricating a hollow-pellet with a lubricated die. The
details are as follows.
Apart from its simplicity, the distinctive characteristic of thesimplified technology is its accuracy in adjustment of mixing
rate of PuO2/UO2, since the job is carried out in the liquid state.
By this means, complicated powder mixing processes neededin the conventional solid state mechanical mixing systemwere greatly improved. In simulation experiments employing
a mockup system and typical liquids, a mixing accuracy of±2.5% was attained. The overflow type fixed quantity service
cylinder and an air lift separator which supplies air freeliquid into the service cylinder played important roles in thissuccess. In laboratory experiments employing real fuel liquids(Pu/U=3/7), we observed the mixing state of the powder
products which were de-nitrated by microwave heating, andconfirmed that non uniform distribution of Pu was not appearedin the hollow-pellet. The results clearly show that the uniformityof MOX powder products obtained by the liquid phase mixing
method is better than that by the conventional method.
Another distinctive characteristic of the simplified technology
is the granulation done in process of adjusting the U/Pu mixing
rate. In this method, high speed agitation with rotating blades isefficiently employed, and not an organic binder but rather wateris sprayed, resulting in slippery particles. Thus, the flowabilityof particles is improved (over 60 in Carr coefficient), a great
benefit. The density of MOX hollow-pellets, which were filledwith these particles and compressed by the usual press in a die,reached over 95% TD after sintering.
Further, a hollow-pellet was fabricated utilizing the same
MOX particles as above, and compressed in a die which
was lubricated. These results demonstrated the feasibility ofthis simplified technology for MOX fuel production on thelaboratory scale.
Hereafter, we would like to carry out larger scale experiments
to attain the target production rate by 2010 when the adoptionof innovative technology will be made. The improvement of
production efficiency is the key for this innovative technology,
we believe.

Reference
Suzuki, M. et al., Development of FR Fuel Cycle in Japan(3) -Current State on Unified Technology of Denitration Conversion and Granulation for
Simplified Pellet Fuel Fabrication Based on Microwave Heating- , Proceedings of 2008 International Congress on Advances in Nuclear Power Plants(ICAPP '08), Anaheim, CA, USA, 2008, p.2036-2045, in CD-ROM.
JAEA R&D Review 200819

Research and Development of Advanced Nuclear System

1-11 Heat Removal during TRU Fuel Fabrication–Design Study of Heat Removal for TRU Fuel Fabrication System–
473K

Feed solution withadjusted Pu content
Denitration

Pelletizing
Sintering

Fin

Pellet fabrication
O/M ratio adjustment
Pin manufacturing

273K
Heat generation density: 20W/kgHM
Inner diameter of hopper: 80mm
Cooling method: natural convection

Grinding
Assembling
Inspection
458

New fuel assemblyFig.1-31 Process flow of simplified pelletizing fuelfabrication
The processes in which heat removal is important is shownin red.
Fromsinteringprocess377

Fig.1-32 Powder supply hopper and temperature
distribution
To prevent accumulation of decay heat, hopper is designedwith narrow shape and fins is installed. An inner diameterwith which the temperature of powder doesn't exceed 473K
was obtained by heat flow calculation.

To grinding
process

423K

Fuel
element

373K

20 W/kgHM
Flow velocity
1.5m3/s

298K

Bafflepannel

303K

Cooling air

Heat generation density: 30W/kgHM
Flow velocity: 1m/s

Fig.1-33 O/M ratio adjustment furnace and temperature
distribution
The furnace is designed to be separated into batches tokeep oxygen partial pressure very low. When pellets areunloaded, cooling air is fed to keep pellet temperature below
358K.

Fig.1-34 Concept of fuel assembling equipment andtemperature distribution
Fuel elements are layered horizontally. Cooling air is fed
from under the bundle.

TRU fuel is being considered as a future FBR fuel because
it can decrease the environmental burden by closing in minor
actinides (MA) such as americium and curium in the nuclear
fuel cycle. The TRU fuel generates high heat on the order oftens of W/kgHM (kgHM means weight of heavy metal elementin the fuel material) by the decay of the contained MA. Heatgeneration by the fuel causes oxidation of MOX powder
and pellets, oxidation of the cladding tube, and mechanicalinterference between the cladding tube and wrapper tube.
Therefore, it is necessary to cool the fuel during the fuelfabrication process. In the fuel assembling process, the powder
supply hopper in pelletizing process, the pellet unloading fromthe O/M (atomic ratio of oxygen and the metal in the oxide) ratio
adjustment furnace and the fuel bundle in the fuel assemblingprocess are considered to have overheating problems.
Concepts of a heat removal system for these areas were
examined, and the temperature limit for each was evaluatedthrough heat flow calculations. Fig.1-31 shows the process
flow of a simplified pelletizing method.
For the examination, provisional temperature limits were set
at 358K for powder and pellet in the air, 473K for powder ininert gas, and 573K for fuel elements and the fuel assembly.
The powder supply hopper is designed to substitute theinside contents with the inert gas and has heat radiation finswere installed outside to cool the hopper naturally (Fig.1-32).
The O/M adjustment furnace is designed to cool down pellets
to 358K in a reduction atmosphere, and then to unload pellets
while cooling them with air (Fig.1-33). The fuel assembling
equipment is designed so that cooling air is blown from under
the fuel element bundle (Fig.1-34).
Based on these designs, temperature distribution was
evaluated with heat flow calculations (Fig.1-32, Fig.1-33 andFig.1-34). Especially for fuel element bundle, because there
are complex passages with many cladding tubes and wrapping
wires, evaluation with a model of 20,000 meshes and evaluation
with a detailed model of up to five million meshes werecombined. Temperature distribution was calculated for various
heat generation densities and cooling conditions, and the size ofcontainer and/or cooling conditions with which the maximum
temperature of the fuel doesn't exceed the temperature limit
were confirmed.
From these examinations and evaluations, the prospect of
keeping below the temperature limit by combining natural
cooling and forced cooling was confirmed for fuel with heatgeneration density 20W/kgHM.

Reference
Kawaguchi, K. et al., Conceptual Study of Measures against Heat Generation for TRU Fuel Fabrication System, Proceedings of International Conference on
Advanced Nuclear Fuel Cycles and Systems (GLOBAL 2007), Boise, USA, 2007, p.290-295, in CD-ROM.

20

JAEA R&D Review 2008

2 Research and Development on Geological Disposal of High-Level Radioactive Waste
R&D Supporting the Technology and Reliability ofGeological Disposal in Japan
High-level radioactive waste (hereafter, HLW) is generated
when we utilize nuclear power. In Japan, reprocessing is done
to recover uranium and plutonium remaining in spent fuelfrom power generating reactors, for recycling into useful fuel.
The liquid remaining after such retrieval is finished is vitrified
to produce a chemically and physically stable glass, which
is HLW. International consensus has been reached that HLW
can be disposed of in stable deep geological strata (geologicaldisposal) so as to be isolated from human environments for along term. Waste glass inserted in an over pack (e.g. carbon
steel) with a bentonite buffer around that will be placed at a
depth below 300 m in the Japanese proposal (Fig.2-1).
To initiate repository operation in the late 2030s, theNuclear Waste Management Organization of Japan (NUMO),
which has the responsibility for geological disposal of HLW,
is now calling for municipalities to volunteer as candidate
site areas and be surveyed, which is the first stage in the final
disposal project.
Geological disposal of HLW will be a long term projectof more than one hundred years from siting to construction,
operation and closing. Therefore, continuous strengtheningof the technical foundations of these activities is very
important. To support such a national project technically,
we have conducted research and development in various
fields, e.g. geoscientific research, engineering development,
and performance assessment (PA) of the geological disposal
system.
A particular focus of our research and development
activity at present involves projects at two UndergroundResearch Laboratories (URLs) to establish techniques forcharacterizing the geological environment: one at Mizunami
city in crystalline rock and the other at Horonobe town in
sedimentary rock (Fig.2-2). The investigations from thesurface (phase1) have been completed. The results fromthis phase of both URL projects were reported at an open
symposium in September 2007. The research of the excavation
phase is ongoing currently to verify and refine the results ofphase1 such as methodology of investigation and modelling ofgeological environments. Research on geological phenomena
such as faults and volcanic activities in order to evaluate thelong-term stability of deep geological environments also has
progressed (Topic 2-7, Topic 2-8).
At the same time, we are conducting laboratory
experimental studies at the Tokai research center. Relatively
large-scale and non-radioactive experiments, performance
Tono Geoscience Center(Toki city, Mizunami city,
Gifu pref.)●Mizunami URL(Crystalline rock)

Vitrified Waste

Overpack

Stable Geological
Environment

below 300m

Buffer material
Massive EBS(Engineered Barrier System)

Fig.2-1 Basic Concept of HLW Disposal in Japanassessment of the multibarrier system supported by extensive
computer analysis, and research to obtain basic data
concerning the chemical properties and migration behaviorof radionuclides under geological disposal conditions has
been carried out. These studies are linked with the geologicalenvironment data obtained from URLs. In 2007, the data forlong-term stability of overpack were complied (Topic 2-3) and“FepMatrix” was released on the Web to serve as an important
supporting tool for scenario development in safety assessment(Topic 2-4).
Based on such research and development activities, wehave conducted a project to develop the next generation novel
knowledge management system (KMS) to systematicallymanage multiple lines of evidence and scientific results
relevant to safety in the form of a knowledge base. The designof KMS was completed in March 2008. System integration isunderway at present (Topic 2-1).
In April 2008, the basic policy and plan for implementationof geological disposal of HLW were revised by government.
It was apparent that research and development organizations
should contribute to understanding by the nation of these
efforts through publicity on their research and development
facilities by means such as URL. In this revision, theimplementation schedule was also changed. We will makesteady progress in improving technology and reliability ofgeological disposal in Japan, taking advantage of trends inbusiness and planning our R&D to be optimally effective.
Horonobe UndergroundResearch Center(Horonobe town, Hokkaido)●Horonobe URL(Sedimentary rock)

JAEA Knowledge
Management System

Otherorganizations

Cooperation
andCollaboration

Concept Image

Tokai R&D Center(Tokai vill., Ibaraki pref.)●Engineering-scale Test & Research FacilitY(ENTRY)●Quantitative Assessment Radionuclide

Concept Image

MigrationExperiment FaciLITY(QUALITY)ENTRYQUALITY

Knowledge Base・Data・Documents
・Software・Experience and methodology
・Synthesis
・Guidance
・Presentation

Implementer

Regulator

URL:UndergroundResearch Laboratory

Fig.2-2 R&D Facilities
JAEA R&D Review 200821

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

2-1

For the Utilization of Knowledge on Geological
Disposal Technology–Detailed Design of Knowledge Management System–

Knowledge Base・Confirmation of methodology and applicable examples for sitecharacterization using ‫ޓ‬expert system⇒Backup for planning etc.

Common Working Memory (Blackboard)・Abstraction of weak-point of argument and of newly needed argument⇒Setting of new issues for R&D etc.

General Working Component

Database

Investigation
Planning of R&D programme

Analysis
SoftwareRepository
Design
DomainData Analysis
Modeling, Simulation

Expert System(e.g. Information Synthesis and Interpretation System＊)Geological
Environment
Domain⹏ଔ䊶⸃㉼Performance Assessment
DomainDocument Base・Report・Paperetc.

The figure shows that defense 2 and 3
Release rates of radionuclides
with support from defense 4 and 5
from N.F. is low enough.
attack threat 1 which, in turn, attacks Defense 1
defense 1.
Overpack can containradionuclides during the first
Defense1,000 years.
Threat 1

Defense 2

・Confirmation of the framework and needed knowledge of safety
・Confirmation of impact on safety by new knowledge⇒Setting of safety requirement⇒Ensuring safety⇒Quality assurance etc.

Knowledge Note
Natural analogue
・
・
・
・
・
・
・

Counter-argument (threat)

Prediction of corrosion rate of iron for 1,000 years
based on the observations in the experiments
lasting only several years contains significant
uncertainty.

Argumentationfor Safety Case

SupportAttackDefense 3

There is some archaeological evidence supportingthat iron corosin rate veraged over 1,000 years ismuch less than the short term observations.

Extrapolating corrosion rate estimated byshort-term laboratory experiments to longer term
gives conservative estimates.

Defense 4

Defense 5

It is observed that growth of analtered product layer suppresses ironcorrosion.

It is observed that corosion rates
measured during experimets decreases
with time.

Common Platform Using Argumentation Model

Fig.2-3 Schematic detailed design of knowledge management system(＊: performed under a contract with ANRE)

The unique feature of geological disposal of radioactive
waste is that safety needs to be ensured for an extremely longtime, tens of thousands of years in the future. It is necessary
to have society accept explanations of the safety of geologicaldisposal based on a variety of evidence, because it is impossible
to actually demonstrate safety for such a long term. A reliableknowledge base on geological disposal is indispensable to givea trustworthy explanation. The objective of R&D on geologicaldisposal technology is to accumulate knowledge and to enhance
the reliability of the technology.
Explanation of the safety of geological disposal iscomplicated, drawing on an enormous amount of knowledge in
multiple fields. The main body of a safety report alone runs toseveral thousand pages. In addition, the amount of knowledge
related is anticipated to increase exponentially during the next
few decades as waste disposal is implemented. It is important
to continue to systematically collect such new knowledge inorder to demonstrate safety.
The results of R&D, such as a database, software for analysis
and technical report, are organized into a knowledge base. The
knowledge base is updated with new knowledge provided
through the R&D. The results of R&D are evaluated as to
how they contribute to explanation of safety. Newly gained
knowledge is examined carefully as to whether it supports
the reliability of procedures or creates doubt. The evaluationof knowledge is a very time-consuming task. Therefore,
it is necessary to build a new system in order to share theinformation among those concerned. A framework of knowledge
management that optimizes production, dissemination, anduse of the relevant knowledge has been proposed, based on aconcept that regards development of the safety case as a chainof arguments and counter-arguments (argumentation model)
regarding the design.
Our Knowledge Management System (KMS) enables thoseconcerned to understand the effect of new knowledge uponsafety, identify issues in use of knowledge, and make plans
for the necessary investigations and analyses to solve theidentified issues. In many cases these activities depend on“tacit knowledge” of the experts which has been accumulated
over decades. In our KMS, the tacit knowledge, including
technical know-how on application of complex investigation
techniques and experience gained in the past, is synthesized as
an expert system in order to improve the efficiency and qualityof the safety case and to transfer the knowledge to the next
generations (Fig.2-3). The detailed design of an intelligent
KMS that utilizes state of the art technology in knowledge
engineering has been completed. A prototype of a novel
KMS will be developed by the end of the current five-year
R&D program (2005 - 2010), taking into account both present
requirements and possible future needs of users.

Reference
Osawa, H. et al., Design Concept of a Knowledge Management System of Geological Disposal Technology, Karyoku Genshiryoku Hatsuden, vol.621, no.6,
p.26-33 (in Japanese).

22

JAEA R&D Review 2008

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

2-2

Development of Buffer Material Database–Compilation of Characteristic Data of Buffer Material–(b) Data search(a) Top page of the database website(c) Example of graph-plotting
Fig.2-4 Screen displays of the buffer material database

The basic characteristics of the buffer material such as
hydraulic, swelling, mechanical and thermal properties are
essential in designing and long-term behavior/performanceassessment of a engineered barrier system of the HLW repository.
It is beneficial to systematically compile these basic buffer
data to allow both the implementing company and regulators
to easily access data for their use in future stages. Therefore,
we have developed the Buffer Material Database focusing on
the basic properties of Japanese bentonite. A Japanese versionof the database has been on the JAEA Website since Mar. 2006,
and an English version since Mar. 2007 (http://bufferdb.jaea.
go.jp/bmdb/).
The basic characteristics of the buffer material have beencollected mainly through experiments since 1989. All of thedata are now available at the above database website(Fig.2-4 (a)). Some functions such as data search (Fig.2-4 (b))
and graph-plotting (Fig.2-4 (c)) are built in the database system.
The data is searchable by selecting variable items such as dry

density, sand mixtures ratio, test temperature, water content,
test solution (distilled water, synthetic seawater, Horonobe
groundwater and NaCl solution), and ionic strength of the test
solution. The searched data can be downloaded from the website
in CSV format. In addition, information on the experiments
such as composition of the bentonite material, sand material
and test solution are also available. At present, data of about
1,820 cases are stored in the database, and the data related to
bentonite extrusion behavior into rock fractures was added at
the end of last year. Up to now, the buffer material databasehas been used by research institutes, universities, privatecompanies and the nuclear waste management organizations athome and abroad such as NUMO, SKB, NAGRA and POSIVA.
The number of user registrations has now reached 145, andthere have been 427 data downloads and 625 uses of the graphplotting function. Updating of the database is continuously
conducted and also the system will be improved in response tousers’ opinions and requests.

Reference
Kikuchi, H., Tanai, K. et al., Database Development of Fundamental Properties for the Buffer Material in Japan, Proceedings of International Conference on
Nuclear Energy System for Future Generarion and Global Sustainability (GLOBAL2005), Tsukuba, Japan, paper no.238, 2005, 6p., in CD-ROM.
JAEA R&D Review 200823

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

Estimation of Long-Term Metallic Corrosion Using Archaeological Sample–Application of Analogize Phenomena in Nature on Geological
Isolation Research–

Iron platecross section

mound

Corrosion depth (mm)

2-3

adzes
Laboratory
experiment

Predicted
carbon steel
corrosion after
1000years(31.8mm)
JNC(2000)
Iron plate

Extrapolated line based on
experimental data

Time (year)

Fig.2-5 Condition of iron plate buried for 1500 years andits excavation site

Fig.2-6 Conservatism of prediction of corrosion after
1000 years based on data of excavated artifacts

Carbon steel material being considered for use as anengineered barrier system (overpack) is expected to keep its
strength for more than 1000 years. We estimated maximum
corrosion depth of the overpack as 31.8 mm for 1000 years in
the case of vertical setting in rock. For the purpose to obtainsome data as second evidence for the long stability estimation,
we investigated corrosion phenomena for iron material ofarchaeological sample which buried for more 1000 years andtheir corroded environmental condition.
Two adzes buried in clay for 750 years were dug out fromthe Izumo-Oyashiro-keidai ruins (Taisha-machi, Shimane).
One of candidate metal material of the overpack is carbon steel
which will be surrounded by clay in a proposed concept ofJapanese high-level waste disposal. The environment of these
excavated samples is similar to the proposed Japanese disposal
environment.
Yamato 6th tumulus is famous as a ruin from which 872
iron plate artifacts were excavated. We examined the corrosion
depth of these samples using X-ray CT. Although they wereburied for 1500 years, they had metallic brightness in theircross sections (Fig.2-5).

Corrosion data for these artifacts such as thickness anddensity of the rust were analyzed non-destructively using ahigh-power X-ray computer tomography system, becauseof their historical importance. The density distribution was
obtained from the X-ray CT image, and thickness of the rust
was determined by the density difference between the base ironmaterial (ρ=8.0) and their rust (magnetite:ρ=4.0). In Fig.2-6,
the corrosion data of the artifacts is shown together with datafrom laboratory experiments with corrosion periods of several
years. The red square symbol ■ is carbon steel corrosion
depth, 31.8 mm, after 1000 years predicted by extrapolation of
the laboratory data (red line). Two green lines extrapolations
based on experimental corrosion data under slightly oxidizingor reducing conditions. The upper line corresponds to a 0.001
mm/y corrosion rate. It is clear from these results that thepredicted corrosion data is conservative.
We will continue to obtain corrosion data of iron artifacts
as secondary evidence for the long term stability estimation ofcarbon steel and work to develop a high reliability estimation
method.

Reference
Yoshikawa, H. et al., Application of Archaeological Analogues for a Repository Safety Case : Arguments Supporting the Waste Container Lifetime, Safety
Cases for Deep Geological Diposal of Radioactive Waset : Where Do We Stand ? , Symposium Proceedings, Paris, France, 2007, OECD-NEA, 2008,
p.365-371.

24

JAEA R&D Review 2008

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

2-4

Realistic Application of Safety Assessment
Methodology to Disposal Environments–Arrangement and Development of Basic Information Using FepMatrix Tools–
Revaluation of phenomenas andtreatment of results,
Addition of Safety Functions

DesignSite-Design
Geologiacl
Environment
Modification
Alteration

Using
FepMatrix tool

Systematical arrangement of existing
information on “handling andcharacterization of a phenomenon”
phenomenon1
phenomenon2

THM CG

phenomenon3
SafetyFunction

phenomenon4
phenomenon5

New scenarios, Evaluation of new
phenomenas, Requested for data
Inside of DataBase

OverView

Reaction ofground-waterand buffer

Site-Design

Neutrality alkalinity
condition

Chemistry

Sorption on
the Buffer

Reference

Considerations were made into exchange ofinterlayer cations, alteration of smectite into
illite, and alteration of smectite
intopseudochlorite, and into iron smectite due
tointeractions with iron and alteration of
buffer material into C-H-S gel or zeolite
without high pH environment,and this influence
was estimated to be small. Other mineralogigcal
change estimate qualitatively based onnatural-analog evidence ‥‥Points of revatuation extracted
Corrosion of
the overpack

MicrobeOrganic
matterCompositionofShotconcrete

Composition of
the buffer

Compositionof ground water

Colloid fromthe buffer

Environmental Change
Mineral compositionof host rock

Colloidin the NBS

Site-Design

Time and spacedependency ofalkaline plume?

Using ofShotconcrete

Movement ofhigh alkaline
plume

Chemistry

Data for each
Nuclides?

CollidHigh alkalinecondition

Sorption on
the Buffer

Additional phenomenon:Chemical condition of porewater from concrete material is affected for an
extended period by Ca2+ from portlandite, porewater in EBS changes to high alkaline water. Whensmectite is exposed to a high pH environment,
alteration of smectite into C-S-Hgel or zeorite
may occur. ‥‥Fig.2-7 Example of information flow in safety assessment

In the safety assessment of radioactive waste geologicaldisposal, there are wide-ranging characteristic features andphenomena to deal with, and it is also essential to consider
those features and phenomena in the time scale of hundredsof thousands of years and in the space scale of an enormous
natural host-rock. For the safety assessment, it is very importantto accurately take in these uncertainties, hence, investigation of
the uncertainty comprises several stages, such as the stage ofunderstanding the relevant phenomena classifiable into various
specialized fields and the stage of the model building based on
this understanding. Therefore, a method to easily understandthe process of creating the safety assessment procedure andto improve the tracking of this process are considered in thisproject.
The characteristic features and phenomena to deal with in
the assessment can broadly be classified into four categories,
Thermal (T), Hydrological (H), Mechanical (M), and Chemical(C). It is important to organize them considering geometric(G) time change. (Hereinafter “THMCG”.) This is because
operation of a barrier function (safety function) giving concrete
form to the long-term behavior and the safety assessment
thought to result from the geological disposal system basedon the geological environment as interpreted by THMCG
and a design suited to that environment can be systematicallycharacterized. By organizing the above information, it will
be possible to review scenarios considering the interactionof the phenomena which should be assessed. To confirm theeffectiveness of this concept, we tried organizing the processof making the safety assessment of a reference case which
was used for the H12 report’s second evaluation of geologicaldisposal R&D using FepMatrix which is a scenario analysis
tool (http://www.jaea.go.jp/02/press2007/p07061901/index.
html). FepMartix is a publicly available tool developed fororganizing the mutual relation functions of vast numbers ofcharacteristics and phenomena in chart form in a computer.
With this method, the conditions of geological environment,
design of the disposal, and the phenomena of concern and how
they are dealt with are reflected in the safety assessment of H12in an organized fashion.
For instance, in investigating the impact on the environmentof the alkaline component of cement in the rock which needs
concrete support, this method enables us to extract easily
points of the current safety assessment which have been
modified or added, by following the modification history of
the information (in the balloon boxes of Fig.2-7). Moreover,it was found that when settings which were made according toquantitative assessment of safety functions are too conservativeor oversimplified and were to be removed, by tracing back theflow of information from the safety function, we could extract
easily the phenomena which should be considered in moredetail and the scenario that should be given attention. Thus, theprospect for the extraction of scenarios and the modificationof analysis cases looks brighter, even though the results are
updated and the designs are modified in future geologicalresearch.

Reference
Inagaki, M. et al., A Study of Methodology and It’s Applications for the Evaluation of Total System Performance which Considered the Site-Information andthe Design-Information, JAEA-Research 2008-022, 2008, 37p. (in Japanese).
JAEA R&D Review 200825

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

2-5

Development of Knowledge Base for HLW Disposal Technology–Technical Know-How and Analysis of Decision Criteria for
Characterization of Geological Environment–

Fig.2-8 Start page of expert system to aid in selectionof tracer for drilling fluid

Fig.2-9 Proposed tracers for drilling fluid
In this example, two criteria were chosen by the user. Then,the expert system proposed eight compounds.

The Mizunami Underground Research Laboratory Project(MIU Project) has been carried out since 1996 in Mizunami
city, Gifu prefecture, as a part of R&D for HLW disposal. A
report giving all results of the surface-based investigation phase(Phase I) was published in 2007. In this report, methodology
for the characterization of geological environments was
summarized. On the other hand, the decision making process
leading to the modification of the investigation program andthe setting of input parameters to be used for modeling was
not described. The experiences and knowledge gained during
investigations of geological environment are only preservedin the program specifications. The above decision making
process, experiences, and knowledge are however important
for the implementation of HLW disposal.
In this study, we tried to clarify these decision making
process, experiences and knowledge obtained through the Phase
I investigations. For example, the decision making process canbe expressed as following rules of an “if-then” format. In total,seven types of expert system have been developed. Fig.2-8
shows an expert system for the selection of tracer for drilling
fluid. Suitable tracers are proposed by the expert system basedon the experience obtained through the MIU Project (Fig.2-9).
Additionally, basic information (chemical structure etc.) of
each compound is available.
We found that so-called “expert judgment” can be expressed
as rules. In this way, knowledge and experiences dealing withcharacterization of geological environments can be passed
on to the next generation. This approach is desired for the
implementation of HLW disposal project because the target
duration of this project is several decades
This work was carried out by JAEA under the contract
with the Natural Resources and Energy Agency, Ministry of
Economy, Trade and Industry.

Reference
Saegusa, H. et al., Final Report on the Surface-based Investigation (Phase I) at the Mizunami Underground Laboratory Project, JAEA-Research 2007-043,
2008, 337p. (in Japanese).

26

JAEA R&D Review 2008

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

2-6

Development of Quality Assessment Method for
Hydrochemical Data
–Proposal of Methodology Based on Evidence Support Logic–

1. pH measurement is of adequate quality
2. Drilling fluid contamination sufficiently low

Elevation(m)Elevation(m)3. Degree of contamination by drilling water, as indicated by artificial tracers
4. Degree of contamination by drilling water, as indicated by natural tracers
5. Evidence other than drilling fluid contamination
6. Analytical evidence indicates that pH has not changed significantly‫ ޓ‬from in-situ value
7. pH measurement is of adequately high quality
17. Stability of physico-chemical parameters
18. Chemical preservation of sample adequately preserved in-situ pH
19. CO2 not lost or gained by the sample

25. Indirect evidence indicates that pH has not changed significantly from in-situ value
26. Contamination by untraced formation water affected pH insignificantly
27. Indirect geochemical evidence indicates that pH has not changed significantly
35. Hydrogeological evidence indicates that pH has not changed significantly
43. Evidence that quality of equipment and material is adequate
55. Methodological information indicates that pH has not changed significantly

Fig.2-10 A part of process model for assessing accuracyof measured pH value
Process model is constructed based on classification of all
items needed to decide a quality evaluation proposition (inthis case, on the accuracy of measured pH). Probability
that the proposition is true, false or uncertain is calculated
based on the values input to the end nodes of the processmodel.

Fig.2-11 Result of quality assessment of measured pHvalue
The plot of data obtained by borehole investigations isevaluated to have a certain degree of accuracy (b) fall withina narrower range than the plot of all data (a).

Hydrochemical investigations of deep geologicalenvironments for the study of water resources, high-level
radioactive waste disposal, etc. have become frequent inrecent years. However, the methodology of the hydrochemical
investigation and of the quality control differs with each
investigation. Therefore, there is a possibility that someproblems will occur due to difference of data quality when
we use the hydrochemical data of different investigations
in common. Development of methodology for selectinghydrochemical data that have enough quality to meet userneeds will be useful for solving this problem. This study is
aimed at establishing a methodology for quality assessmentof chemical data obtained from groundwater samples using amethod known as Evidence Support Logic.
The basic operations of ESL are similar to Boolean
operations. However, the element “uncertainty” is also used in
ESL in addition the “true” and “false” in Boolean operations.
Uncertainty in ESL means the probability that a proposition
cannot be judged either true or false. One of the main
characteristics of ESL is its process model (Fig.2-10). This
process model is a hierarchical structure extracting evidence
and classifying related items needed to prove a hypothesis. In

the case of Fig.2-10, the hypothesis is “measured pH value isthe true pH value of groundwater”. The process model searches
and organizes all elements related to quality of the data given in
the proposition. Therefore, it is useful not only for elucidating
the process of quality assessment but also for improvementof investigation method by detection of problems with theinvestigation. The results showed that 23 out of 63 samples
satisfied the applied quality criteria. The plotted data from the23 samples (Fig.2-11 (b)) have a more constant value from the
top to the bottom of investigation area (from 8 to 9) than thedata from all samples (Fig.2-11 (a)). After low quality data wasidentified, it was verified that sufficient quality control was notapplied, and there was contamination during borehole drillingand on groundwater sampling. These results show that the data
which meet the chosen criterion express pH of each depth more
precisely.
This study indicates that the quality assessment method
demonstrated is applicable to hydrochemical studies and thatby thus confirming the sample quality it will be possible todevelop a clearer interpretation of the deep undergroundgeochemical environment.

Reference
Mizuno, T. et. al., Proposal for Quality Assessment Method of Groundwater Chemistry for Hydrochemical Investigation, Nihon Chikasui Gakkaishi, vol.49,
no.2, 2007, p.139-152 (in Japanese).
JAEA R&D Review 200827

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

2-7

Study of Features of Active Faults Based on Measurementof Hydrogen Gas Emission from the Underground–Development and Application of Portable Hydrogen Gas Equipment–(a)Drilling
Surface

Measurement
Plastic syringe 20 cc
Silicone tube
Teflon tube

SamplingClaycharacterization
Download

Incohesive fault

Fig.2-12 Hydrogen measurement in fault zone (a) andprocess flow of hydrogen measurement (b)(a) Data logger with H2 sensor. The key part of the new
equipment is a light, compact hydrogen gas sensor which
can make measurements easily.
(b) A Teflon tube is fixed in the hole using clay recoveredfrom the hole. The equipment is left logging on the site
for approximately three hours or more.

Fig.2-13 Distribution of density of hydrogen gas aroundYamazaki fault zone
The integrated value over one hour (2-3 hours after drilling);○ is less than 10ppm, ● is 10ppm or more.

Research on the influence of fault activity on deep geologicalenvironments contributes to the reliability of geological disposal
systems for high level radioactive wastes. The permeability ofa fault zone tends to be higher than the surrounding bedrock.
Thus, it is important to understand the long-term changes ofregional groundwater flow around the fault zone, where a wide
crushed bedrock zone frequently exists. We are examining theapplicability of an investigation technique using hydrogen gas
emission from fracture zone.
It has been observed that the hydrogen gas emitted fromunderground along the active fault exceeds the atmosphericconcentration (ca.0.5ppm). Various laboratory experiments
have shown that hydrogen gas is generated during fracturing ofrocks in deep underground near the hypocenter of an earthquake.
It is reported that the hydrogen gas migrates to ground surface
along the same fractures through which groundwater flows.
Therefore, we expected that in this way we could estimate
hydraulic properties such as the groundwater pathways and thecontinuity of fractured rock zones around the fault.
To determine hydraulic flow properties around an active
fault, it is necessary to measure the concentration distributionof hydrogen gas over a wide area. However, the conventionalhydrogen gas measurement tool is a large device requiring a
day or more to accurately measure concentration of hydrogengas discharged from underground. Therefore, we developed
a new measuring tool using portable hydrogen gas detectors
which enables measurement of many points in a short time
(Fig.2-12). We investigated the concentration of hydrogen gas
emitted from the Yamasaki fault zone, in southeast Japan. The
Yamasaki fault zone is one of the major active faults in Japan
which caused the A.D. 868 Harima Earthquake. In our survey,
the points where high concentrations of hydrogen gas emission
was detected were unevenly distributed along the active fault,
and in the southeast portion of the study area where small
earthquakes frequently occurred (Fig.2-13).
Based on these results, we will carry out the measurementof hydrogen gas emission in many study areas to obtain basic
information on hydraulic flow properties around the fault
by and also to understand the relationship between emitted
gas concentration distribution and properties of the geologic
structure.

Reference
Shimada, K., Tanaka, H., Saito, T., Rapid and Simple Measurement of H2 Emission from Active Faults Using Compact Sampling Equipments, Resource
Geology, vol.58, no.2, 2008, p.196-202.

28

JAEA R&D Review 2008

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

2-8Strategy for Long-Term Predictions for Geological Disposal R&D–Long Term Prediction of Groundwater Flow Changes Based on
the Geological History–

Description of relevant natural phenomena(geological understanding)

West

East

Present
Construction of conceptual model for geological evolution andthe impact of the natural phenomena on geological environment
Systematic assessment of impact ofnatural phenomena on groundwaterflow (FEPs/PID format)

sea-level

shoreline

Sarabetsu Fm.
Yuchi Fm.
Koetoi Fm.
Wakkanai Fm.

Activefault

Future glacial stage

Permafrost

low activity
fault
groundwaterflow

Construction of hydrogeological model
including past evolution
Associated groundwater flow(GW)
quantitative simulation

Sea-level change

Comparison of model output with paleohydrogeological informationand current hydrological/geochemical dataes.) Does the model give the best-possible representation of GWflow evolution over the time considered?

Climate change

Denudation& sedimentation Weathering

UpliftUplift

Fault

Fault
Best-possible description of future GW flow

Fig.2-14 Basic framework for long-term
prediction of groundwater (GW) flow changes
Understanding of evolution of the groundwater(GW) flow from the past to the present wasused to develop descriptions of likely future
evolution of GW flow.

sea-level

Regional uplift
Fig.2-15 A conceptual model of likely future evolution of GW
flow in the Horonobe area
The figure represents the situation in a future glacial stage, with someexaggeration, some 100,000 years in the future. The conceptual
sketches of the future GW flow are based on an evaluation of the
expected impact of natural phenomena, such as permafrost andshoreline and topographic changes, on the present GW flow.

A representative scenario for a human environment to beaffected by waste disposed in a deep geological formation, isthe groundwater scenario, where radionuclides are transported
to the surface along with flowing groundwater (GW). The
GW flow will be changed by natural phenomena such as
precipitation/shoreline changes due to a climatic/sea-levelchange, and changes of topography/geological structures due to
uplift, subsidence, and denudation on the geological timescale.
Therefore, a critical issue for improvement of confidence inthe long-term safety of the geological disposal is assessmentof the safety taking into account expected long-term evolutionof GW flow. We have been doing R & D of prediction of theGW flow evolution in the Horonobe area, northern Hokkaido
as a case study, taking into consideration the impact of natural
phenomena.
Basic methods of geological prediction are classified into
two groups: (1) prediction by extrapolation of the tendencies
and regularities of natural phenomena in the past, (2) prediction
by analogy based on a study of the effects phenomena which

resemble those expected in the study area. In any method,
studying the past geological history is the key to predicting thefuture. We have developed a methodology for describing future
GW flow using various site investigation technologies (Fig.2-14).
Future likely evolution of the GW flow is described on thebasis of past GW flow evolution determined by the geologicalhistory; it assumes that future natural phenomena will occur
under much the same conditions as they did in the past.
The GW recharge will be reduced in a future glacial stage
due to development of permafrost and precipitation reduction,and the flux and pathway of GW flow also will change due toa change of hydraulic gradient caused by topographic changes
and lowering of sea-level (Fig.2-15).
We have been carrying out GW flow simulation for
estimation of the past GW flow evolution taking account of the
impact of natural phenomena. The results will be confirmed
by comparison with the groundwater chemistry obtained byborehole surveys etc.

Reference
Niizato, T. et al., Paleogeography and Geological Evolution Since the Late Pliocene in and around the Horonobe Area, Northern Hokkaido, Chishitsugaku
Zasshi, vol.113, suppl., 2007, p.119-135 (in Japanese).
JAEA R&D Review 200829

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

2-9

Tracing Water Movement from Surface to Underground–Surface Hydrological Investigation at Horonobe Area, Hokkaido, Japan–
Precipitation

Evapotranspiration
Estimation of evaporationfrom snow surface
Measurement of
evapotranspiration
over forest

Measurement of snow
depth and density

Snow

Rain gauge
Soil moistureobservation

Measurement of
evapotranspirationfrom grassland

Rain gauge

Groundwater level
observation
Groundwater level observation(for estimation of groundwater recharge rate)(for estimationof watershed)

Fig.2-16 Schematic view ofthe surface hydrologicalinvestigation
Groundwater recharge ratefrom surface into undergroundis estimated by subtraction of
evapotranspiration and runoff
from amount of precipitation.

RunoffRiver flux
observing system

Water recharge into the basement rockElevation (m)Topographical watershed
Groundwater divide
estimated by borehole data

Distance from BP-5 borehole (m)After rain and snow fall, there is evapotranspiration fromplants and the ground surface, and runoff into rivers andlakes. The rest seeps underground through the soil. Knowing
this recharge rate is necessary for groundwater flow analysis
which in turn is important to evaluate the safety of geologicaldisposal. Recharge rate is estimated by considering the above
water balance of precipitation, runoff and evapotranspiration
because it is difficult to measure directly.
In the Horonobe Underground Research Laboratory Project,
we carry out surface hydrological investigations which consistof meteorological observation, observation of river flux andwater quality, soil moisture observation and groundwater level
observation in shallow boreholes extending several meters
below the ground surface. These investigations aim to aid indeveloping the methodology to understand water balance andgroundwater flow properties, by application of the series ofSurface of ground
Groundwater level (2nd Dec., 2004)
Groundwater level (2nd May, 2005)
Boundary of weathered andstrong weathered mudstone
Boundary of unweathered andweathered mudstoneFig.2-17 Comparison between topographicalwatershed and groundwater divide
A rise in groundwater level indicates a groundwater
divide. We confirmed that the topographicalwatershed differs from the groundwater divides
found in shallow borehole data. (In this figure,alphabet-and-numbers is a name of a shallow
borehole drilled in this area.)

research techniques used for these observations specifically tothe estimation of recharge rate (Fig.2-16) .
From the current investigation, recharge rate of thiswhole area can be estimated as about 120 mm a year, whichcorresponds to 10% of precipitation (about 1,375mm a year) inthis area. We also confirmed that the topographical watersheddiffers from the groundwater divide found in shallow boreholedata. These results provide significant information for setting
the boundary and initial conditions for groundwater flow
analysis (Fig.2-17) .
We plan to evaluate the validity of these methodologies andthe results of these investigations, including the appropriatenessof the size of the observation areas, by comparison ofobservations up to now with data of the distribution of water
quality and age, etc. and precise observations of precipitationand evapotranspiration taking account of snow accumulation.

Reference
Tomura, G. et al., Surface hydrological Investigation of Horonobe Underground Research Laboratory Project, JAEA-Research 2007-063, 2007, 46p.
(in Japanese).

30

JAEA R&D Review 2008

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

2-10 Tunneling Underground to Investigate Initial Stress Conditions–3-D Initial Underground Stress Measurement by Hydraulic Fracturing–

Borehole ԙ
Borehole Ԛ
Borehole Ԙ
Borehole to measureinitial stress(20m×3)
Niche
Ventilation shaft

5q
30q

Borehole Ԛ

Horizontal drift
depth 140m

West shaft

Fig.2-19 Measured
initial stress(Schmidt net lower
hemisphere of theprincipal stress)

Borehole Ԙԙ
Niche
Ventilation shaft

Horizontal drift
Borehole to measure depth 140m
initial stress(20m×3)

East shaft

BoreholeԘ N60W, Upward 5°
Boreholeԙ NS, Upward 5°
BoreholeԚ N30W, Downward 30°

stress evaluated●，▲，■；Principalfrom borehole breakout

The inside of theparenthesis in a figuremeans strike and dip.

Initial stress measured
140m Niche:Borehole breakout (Max)underground(b) 140m deep level tunnel(niche) side view

140m Niche:Borehole breakout (M i n)140m Niche:Hydraulic fracturing (Max)140m Niche:Hydraulic fracturing (M i n)

Depth from surface(m)(a) 140m deep horizontal tunnel(niche) plane view

Dug part
Undug partstress evaluated by●，▲，■；Principal
hydraulic fracturing technique

Initial stress measured
underground

Overburdenpressure(HDB-6)

Overburdenpressure(HDB-11)

140m Niche:Borehole breakout
140m Niche:Hydraulic fracturing

Max・Min of explanatory notes shows themagnitudes of the maximum and theminimum horizontal principal stress

Magnitudes of the maximum andminimum horizontal principal stess(c) Borehole ① drilling situation(d) Relation between the
maximum/minimum horizontal
principal stress and depth

Directions of the maximum horizontal
principal stress(e) Relation between the directionof the maximum horizontal
principal stress and depth

Fig.2-18 Borehole drilling at initial stress measurement
positionFig.2-20 Comparison with the initial stress estimated inPhase1The Underground Research Laboratory (URL) has been
conducting a research project in Horonobe-cho, Hokkaido. This
project aims at improvement of the reliability of technologies
for geological disposal of high level radioactive wastes. It
started in 2000, and the surface-based investigation phase
was completed in 2005. Now it has shifted to the construction
phase2 (investigation during construction of the undergroundfacilities).
In the surface-based investigation phase1, the area of about
3 km×3 km around URL in the Hokushin district, Horonobecho was selected as the main research area, and geophysicalexploration, geological survey, borehole survey, etc. werecarried out in this area. In general, initial stress due to crustal
movement and overburden pressure has been taking place
underground here. In designing underground structures, itis necessary to measure initial stress around undergroundstructures. Therefore, estimation of initial stress distributedin the URL area has been attempted during surface-based
investigations in order to design URL.
In investigation during construction of the undergroundfacilities, the updating and check of the validity of initial stressestimations made during surface-based investigations is oneof the main issues. Thus, three boreholes were made from theniche (GL-140 m) of ventilation shaft shown in Fig.2-18, anda three-dimensional evaluation of the initial stress in Koetoi
Formation was done by the method called hydraulic fracturing.
Breakout (spalling) of the walls of the boreholes due to thestress concentration was detected on plastic films wrapped on
an impression packer inserted in the borehole. Then, analysisof the borehole breakout was performed to estimate in situ theorientation and magnitude of stress.
The measurement result of initial stress is shown in Fig.2-19.
Initial stress was broken down into principal stress σ1 (major
principal stress), σ2 (intermediate principal stress), and σ3(minor principal stress), and comparatively good coincidence indistribution of the direction of the principal stress evaluated by
the two above methods was observed. Although the magnitudes
of principal stress σ1, σ2, and σ3 found by the surface based
investigation and the impression packer were 2.3, 1.8, 1.2
MPa and 2.4, 2.3, 1.3 MPa, respectively and σ2 values were
somewhat different, general agreement was obtained. Fig.2-20
shows the result of the horizontal initial stress obtained fromthe borehole investigation (HDB-1, 3, 6, 9, 11 holes) carriedout in the past, and the result obtained from the present
underground test. Fig.2-20(d) shows the relation between themagnitudes of the horizontal principal stress, and depth. The
Fig.2-20(e) shows the relation between the direction of the
maximum horizontal principal stress, and depth. From thisresult, although the magnitudes of the maximum horizontal
principal stress obtained here are mostly in agreement withthe overburden pressure, they are slightly smaller than the past
results. Moreover, the direction of the maximum horizontal
principal stress had been found to be east/west in general.
However, the present test showed that the direction was 25° or
43° counterclockwise from east/west. To gain data to explain
this difference, initial stress measurements in the pressure at thedepth 140 m (not counting the niche) in Koetoi Formation, andpressure at the still deeper depth for the Wakkanai Formation
will be carried out.

Reference
Nakamura, T., Sanada, H. et al., Initial Stress Estimation of Diatomaceous Mudstone using Hydraulic Fracturing Test”, The 12th Japan Rock Mechanics
Symposium, 2008, p.297-303, in CD-ROM (in Japanese).
JAEA R&D Review 200831

3 Nuclear Fusion Research and Development
For Practical Use of Fusion Energy
DEMO
Reactor

Demonstrate
Fusion Power
Q=30-50
Power Generation of
500-1000 MW

ITER Project: Demonstrate the Scientificand Technological Feasibilityof Fusion Energy
Construction
Site;FranceExperimental
Reactor

Broader Approach:
Develop Technological Basisfor DEMO with supporting andcomplementing ITER

Achievement of
Fusion Burn with Q≧10
for 300-500 seconds

Hosted by Japan

Q: Energy Multiplication Factor

Contribution as a Semi-Host Country (20% of Equipments and Staff)

AOMORI

International Fusion Energy
Research Center

Fulfillment of ITER Large R&D
Projects:
Development of Central
Solenoid Model Coil,
Vacuum Vessel, etc.

IBARAKI

World Highest Energy Multiplication
Factor: 1.25
World Highest Temperature:
520 million K

Fusion Engineering:
Material, Blanket, etc.

Engineering Validation andEngineering Design Activities for
IFMIF
Hosting international design teamBlanketFusion Plasma:JT-60

DEMO design and R&D coordination center
Fusion computer simulation center
ITER remote experimentation center

Satellite Tokamak
Fusion plasma research for the effective
start of ITER experiment and the DEMO

Fig.3-1 Development Steps Toward the Fusion DEMO Reactor

At the Fusion Research and Development Directorate,crucial R&D for practical use of fusion energy (fusion plasma
and fusion engineering R&D) is being carried out in wide scaleinternational cooperative projects such as ITER project andBroader Approach, aiming at realization of the fusion DEMO
reactor (Fig.3-1).
ITER (International Thermonuclear Experimental
Reactor) Project
The ITER project is an international cooperative project todemonstrate the scientific and technical feasibility of fusionenergy through construction and operation of an experimental
reactor. The participants are Japan, EU, US, Russia, China,
Korea, and India. The participants account for more than halfof the world population. The experimental reactor, ITER, isbeing constructed in Cadarache in France. The ITER agreement
entered into force in October 2007, and JAEA was designatedas the domestic agency of ITER project in Japan. Topic 3-5 andTopic 3-7, “World Record Achievement in Plasma Heating
Device” and “Demonstration of Final Installation to within
0.5mm of a 4-ton Blanket by Remote Handling” are research
results achieved in the course of preparation for ITER
equipment procurement.
Broader Approach (BA) Activity
The BA activity is a joint project by Japan and EU executing
research supporting ITER and the research and development
for the DEMO reactor, the next step of ITER, aiming at theearly realization of fusion energy. It is being implemented inJapan during the period of ITER construction (about 10 years).
It consists of three projects; the activities of the International
Fusion Energy Research Center, the engineering validationand engineering design activities of the international fusionmaterial irradiation facility (IFMIF), and the research activities
of Satellite Tokamak (upgrade of JT-60 to a superconducting
machine). The former two is being conducted at Rokkasho
village in Aomori Prefecture, and the satellite tokamak projectat Naka city in Ibaraki Prefecture. The agreement launching
32

JAEA R&D Review 2008

BA entered into force in June 2007, and JAEA was designatedas the implementing agency of BA activity in Japan. The BA
activity includes a remote experiment of ITER, and Topic 3-3,
“Development of the Plasma Movie Database System for
JT-60”, is a result which contributes to realize it.
Fusion Plasma Research
The critical requirement for the future fusion reactor is
to attain a high level of economical efficiency, namely, tosustain a high fusion power in a reactor core of compact size.
It is necessary to improve the plasma pressure (temperature ×
density) to accomplish that. Topic 3-1, “Establishment of LongPulse Operation for “ITER” Burning Plasmas”, experimentallydemonstrated that a high plasma pressure can be maintained
stably using JT-60. Furthermore, Topic 3-2, “Finding theIntrinsic Rotation by Plasma Pressure”, clarified the guiding
principle to achieve high plasma pressure.
Fusion Engineering Research
The fuel of fusion energy is deuterium and tritium. A large
quantity of deuterium exists in seawater. The tritium hardly
exists naturally at all, but it can be produced from lithium,which is also contained plentifully in seawater. Topic 3-8 andTopic 3-9, “Materials Necessary for Stably Supplying Fuel
Tritium to Fusion Reactors” and “Tritium Extraction Usinga Fuel Cell” yielded results which contribute to the steadyproduction of tritium.
It is necessary to heat plasma to cause the fusion reaction.
Topic 3-4 and Topic 3-6, “Achieving High Power in the JT60U RF System” and “Development of a Large Negative Ion
Source” are results which enable the plasma to be heated more
efficiently.
It is necessary to streamline the maintenance of the fusion
reactor in order to raise its operation rate and thus improve its
economic efficiency. Topic 3-10, “Torus Configuration and
Maintenance Concept of Compact Fusion DEMO Reactor” is a
result which raises the operation rate of the plant by executing
regular maintenance in a short time.

Nuclear Fusion Research and Development

Establishment of Long-Pulse Operation for “ITER” Burning Plasmas–Steady State Sustainment of High Confinement and High β Plasmas–

β3-1

Fig.3-2 Comparison of plasma pressure profile
with different plasma rotation
The plasma rotation along the torus wascontrolled by neutral beam injections providing
torque. When the plasma rotated in the positive
direction, the plasma confinement was improved.
Then, higher plasma pressure was obtained asshown by the red line.

Fig.3-3 Plasma performance and sustained duration in long-pulse
plasmasThe controllability of the plasma rotation and the efficiency of neutralbeam heating were improved by toroidal magnetic field ripple reduction
through use of ferritic steel tiles. As a result, a high product of normalized
plasma pressure and confinement enhancement factor exceeding the
ITER standard was sustained for 28.6s. This new world record is about
1.7times longer than the previous world record of 16.5s achieved in
JT-60U.

One of the major objectives of ITER is to sustain burning
plasmas with fusion gain (Q = fusion power output/auxiliary
heating power input) larger than 10 for a long time (morethan 400s). The fusion power is proportional to the square of
the plasma pressure (temperature × density). To obtain suchhigh Q plasmas, both A) High fusion power output, achievedby increasing the plasma pressure, and B) Reduced auxiliary
heating power, achieved by improving the plasma confinement,
must be satisfied simultaneously. “Normalized plasma pressure(βN)” is used as the index of plasma pressure and “confinement
enhancement factor (HH)” is used as the index of the quality ofthe plasma confinement. The product of βN and HH, βN·HH,
is widely used as a measure of Q, since Q is proportional to
βN·HH,In order to establish long-pulse burning plasma operation in
ITER, steady state plasma operation with high value of βN·HH,
exceeding the value required for ITER, has been a main target of
fusion plasma research. Making use of the good controllabilityof plasma rotation in JT-60, we found that the achievable βN
and HH can be increased when the plasma rotation in the core

becomes positive (Fig.3-2). In addition, the uniformity oftoroidal magnetic field was improved by installing ferritic steel
tiles inside the vacuum vessel. As a result of the reduction in
the loss of fast ions from neutral beam injections for plasma
heating, both the controllability of the plasma rotation andthe efficiency of the neutral beam heating were improved
successfully.
By optimization of the toroidal rotation profile and the edge
plasma density so as to maintain high β and high confinement,
steady-state operation of the high performance plasma has been
established in JT-60 (Fig.3-3). In this study, high performance
plasma with βN·HH of 2.0 was sustained for 28.6s, the world
record. The achieved value of βN·HH is higher than that required
for ITER standard operation of 1.8. The sustained duration islimited by the maximum injection period for the neutral beam
heating system in JT-60. Since the pulse duration of the heating
system for ITER is longer than the target regime (more than400s), the long-pulse operation of burning plasmas for ITER
seems to have been established.

Reference
Oyama, N. et al., Improved Performance in Long-Pulse ELMy H-Mode Plasmas with Internal Transport Barrier in JT-60U, Nuclear Fusion, vol.47, no.7,
2007, p.689-697.
JAEA R&D Review 200833

Nuclear Fusion Research and Development

Intrinsic Rotation Due to Plasma Pressure Discovered–Toward the Understanding of High Pressure / Highly SelfRegulating Plasma–(a) Rotation
Mechanism Pressure?
edge core edge

NB injection

pressure
gradient

Neutral beam (NB)
injection can change
the plasma rotation

Transport
Understoodby the original
experimentedge core edge

Plasmarotation

Pressure
gradient

Plasma rotationedge core edge

Current

(b)

(c)

(d)

Intrinsic rotation (km/s)

3-2Intrinsic rotation

Final rotation profile
Intrinsic rotation grows
with pressure gradient

Pressure gradient (104Pa/m)
Fig.3-4(a) Schematic view of self-regulating high pressureplasmaThe plasma pressure, rotation, and electric current profiles
are strongly linked to each other.
(b) Schematic drawing of JT-60 plasma
JT-60 has various neutral beams (NB) with different injection
angles, which can change the direction of the plasma rotationand the rotation profile flexibly.

Fig.3-5(c) Formation mechanism of the rotation profile
The rotation profile is determined by the external momentum
input by NBs, momentum transport, and the intrinsic rotation
caused by the plasma itself. Using an original experimentand analysis method, we have separately evaluated theexternally induced rotation and the intrinsic rotation.
(d) Relation between the intrinsic rotation and thepressure gradient
The intrinsic rotation increases with increasing pressure
gradient. The difference in color corresponds to the different
NB power and torque.

Burning plasmas for nuclear fusion reactors are highly selfregulating, governed by strong linkage among plasma pressure,electric current, and rotation profiles as shown in Fig.3-4(a).
To construct fusion reactors, it is essential to understand andcontrol this self-regulating system. However, it has been an
open critical issue for nuclear fusion research worldwide tounderstand the formation mechanism of the rotation profileand the relation between the rotation and pressure profiles.
The rotation profiles are determined by various mechanisms:(1) the external torque input by neutral beams (NBs), (2) the
momentum transport, and (3) the intrinsic rotation. However,these terms cannot be evaluated separately by the conventional
steady-state-experiment and analysis.
In order to address this issue, we newly developed andapplied an experimental method utilizing the modulation of apart of the NBs, and as a result we evaluated the momentum

transport correctly. At low plasma pressure, it was found that therotation profiles are determined by the external torque appliedby the NBs and the momentum transport. At higher plasmapressures, the self-regulation appeared. In this case, we foundthat the measured rotation profiles cannot be explained merely
by the momentum transport and the external momentum input
from the NBs, and the intrinsic rotation becomes dominant.
The relation found between the plasma pressure gradient andthe intrinsic rotation in various confinement modes is shownin Fig.3-5(c). It was experimentally found for the first timethat the intrinsic rotation grows with increase in the pressure
gradient in all cases. This study enables us to understand themechanism of the intrinsic rotation in high pressure and highly
self-regulating plasmas, and gives important information about
the prediction and operation of the self-regulation plasmas in
ITER and other future fusion reactors.

Reference
Yoshida, M. et al., Role of Pressure Gradient on Intrinsic Toroidal Rotation in Tokamak Plasmas, Physical Review Letters, vol.100, 2008,
p.105002-1-105002-4.

34

JAEA R&D Review 2008

Nuclear Fusion Research and Development

3-3

Development of the Plasma Movie Database System for JT-60–Movie Data Handling Technology for Remote Experiment in the
ITER-BA Project–
Plasma Movie
Capturing System
TimingSignals
Workstation

RAID

Plasma Movie
Web Server

Fig.3-6 An Example View of Plasma Movie
The plasma movie consists of the plasma video camera
picture, plasma cross section computer graphics (CG), andthe voltage signal from the magnetic field probe input to thesound channel.

Visible Light
Video Camera
Top Viewofthe
Vacuum VesselSoundAmplifier
Speaker
Plasma Shape
Reconstruction
SystemPort

Multiple Views
Composer

JT-60
PlasmaVideo
CameraMagnetic
Sensors

JT-60 Central
Control RoomPC

Fig.3-7 Hardware Configuration of the Plasma Movie
Database System
Plasma Movie Capturing System (PMCS) automaticallyrecords a plasma movie according to the timing signal of
JT-60 discharge sequence. This system executes 2 types ofcompression, MPEG2 and MPEG4, for an effective databasehandling. A user can download the plasma movie data filefrom the database through LAN.

A plasma movie is one of the most efficient methods toconvey what discharge has been made in the JT-60. This movie
on a large TV monitor is able to provide various types of useful
information in real time to the physicists; hot spots due to heatflow from a plasma on the vacuum vessel wall, time-evolutionof plasma position and shape, change of macroscopic plasma
rotation, etc. This innovative technology has contributed tosafe, efficient operation, and has added an important chapter tothe history of tokamak experiment. This real-time monitor has
already be come an indispensable tool in JT-60.
This movie is composed of 4 elements: a video camera
picture (right: Fig.3-6), a plasma cross section CG (left: Fig.3-6)
generated by physics-based calculations from magnetic fieldmeasurements, a sound which is just the magnetic probe
oscillation signal converted to audio, and a display of the
waveforms of the plasma current and of the poloidal field coilcurrent. This plasma monitoring system for JT-60 was the firstdeveloped in the world for tokamaks, and is still a unique featureof JT-60. The sound signal sometimes has a mysterious timbre
originating from amplification of the raw voltage signal of asmall magnetic field pick-up coil sensor in the vacuum vessel.
This sound seems to imply various plasma states; changes inplasma rotation speed appear to cause changes in the tone ofsound. A stable plasma produces smooth high tone as it rotates
at a high constant speed. On the other hand, an unstable plasma
has a noisy low tone as it rotates at an irregular speed.
The plasma movie database system (PMDS) converts analog
to digital movie data using 2 compression methods (MPEG2
and MPEG4). The MPEG2 generates a high-resolution (large
volume) data file which corresponds to the high definition TV,
while the MPEG4, originally developed for the high-speed
network broadcasting, generates a low-resolution, data file.

The hardware configuration of PMDS is illustrated in
Fig.3-7. The plasma movie capturing system (PMCS)
converts a movie to MPEG2 format compressing in real time,and records the resultant MPEG2 movie file automatically.
Immediately after the MPEG2 movie is completed, the systemstarts converting from MPEG2 to MPEG4 format. When all
the conversion is completed, the system transfers an MPEG4
movie file to the plasma movie web server (PMWS). Twoformatted movie files are stored in corresponding databases.
The PMDS checks what new movie files become available
for distribution through the network every 10 minutes. A user
can download a movie file from the database. This system has
been configured using a personal computer with Windows XP
and the existing network, which minimized the development
cost and time. We solved a technical issue in efficient handlingof the huge-volume movie data by employing 2 compression
methods newly available in the market, as mentioned above.
Since this system creates a movie file for every discharge in
the JT-60 experiment sequence, it provides a movie with soundgiving sensory support to the conventional understandingof each discharge using waveforms of time-series data. In
addition, the physicists can use this system also for search of
the latest interesting pulse discharge.
For conducting experiments on ITER and the satellite
tokamak fusion device JT-60SA from remote sites, thephysicists and operators at the remote site have to know what
discharge is going as accurately and immediately as possible.
This system could be also the key technology of such future
remote experiment since it includes advanced real-timeprocessing and movie data handling technology. It has alreadybeen used in a demonstration of JT-60 remote experimentationfrom Europe in Dec., 2007.

Reference
Sueoka, M. et al., The Plasma Movie Database System for JT-60, Fusion Engineering and Design, vol.82, issues 5-14, 2007, p.1008-1014.
JAEA R&D Review 200835

Nuclear Fusion Research and Development

3-4

Achieving High Power in the JT-60U RF System–World Record for Gyrotron Oscillation (1.5MW,1s)–Gyrotron output power (MW)World record 1.5MW, 1s
High power gyrotron
development toward
high performance
plasma control

FY. 2007~ FY. 2006
Original spec. of JT-60U
RF-heating system

Long pulse gyrotron
development toward
continuous oscillation
for “ITER” and “Demo”

Collector

Outputwindow

RF

Electron beam
Ceramic
insulator

CavityPulse length (s)Achieved in JT-60U RF-heating system
Achieved by JAEA for ITER
Achieved elsewhere in the world
Fig.3-8 Performance of the world's gyrotrons (>1MW)
Remarkable power and pulse length were achieved by the
JT-60U RF-heating system.

Fig.3-9 Schematic view of gyrotron
Energy of electron beam is converted to RF energy at thecavity, and the RF radiation is output through the output
window. The cavity is strongly heated due to the high RF
power density of the cavity resonator, and stray-RF radiation
heats the ceramic insulator.

Gyrotron oscillation power of 1.5MW with pulse length of
1s, a world record, was achieved by the gyrotron of the JT-60U
RF-heating system (Original spec.: 1MW for 5s) (Fig.3-8).
RF injection with pulse length over 1s is an important
issue in fusion experiments because the typical time scale ofphenomena in plasma is around 1s. Capability to do experimentson controlling high performance plasma strongly depends onthe maximum RF power. Therefore, it is essential to improve
the oscillation power of each gyrotron and also to develop along pulse gyrotron, striving for continuous oscillation in
future devices such as “ITER” and “DEMO”. In a gyrotron
system, electron beam and RF power heats up the internal
components. That leads to failure of the components, arcing
and sudden oscillation termination due to thermal expansionof a cavity, especially during high power operation. Those
phenomena had limited oscillation power to 1MW in previous
fusion experiments. Oscillation power of over 1.5MW had
been achieved only in short pulse tests (less than around 0.1s)
with gyrotrons without enough heat removal capability.
Important issues are to clarify heat distribution, to improvehigh-temperature durability, and to improve heat removal
capability. It is important to demonstrate stable oscillation withhigh power and pulse length over 1s.
In previous work, we found that a ceramic insulator was
dielectrically heated by stray-RF radiation (Fig.3-9), so we
introduced a silicon-nitride ceramic into one of the JT-60U
gyrotrons in place of an alumina ceramic. Temperature rises ofsome internal components were measured and investigated overa wide range of oscillation power up to 1.3MW by means of the
gyrotron. If a scaling up of those experimental results is possible
there could be higher beam current operation with the improved
cooling performance of the cavity. We have achieved stable
1.5MW oscillation for 1s by precise optimization of oscillation
conditions using the improved gyrotron. Transmission of thishigh-power RF has been also demonstrated. Moreover, thermal
expansion of the cavity had no significant ill effect on stableoscillation and the cavity temperature reached its maximum in
0.5s and saturated. Therefore, further pulse length extension
with high power is possible in the near future when there are
improvements in the power supply and transmission line.

Reference
Kobayashi, T. et al., Achievement of 1.5MW, 1s Oscillation by the JT-60U Gyrotron, Plasma and Fusion Research, vol.3, 2008, p.014-1-014-3.

36

JAEA R&D Review 2008

Nuclear Fusion Research and Development

3-5World Record Achievement in Plasma Heating Device–Demonstrated High-Power and Steady-State Operation in Gyrotron
toward ITER Construction–

Output Power (MW)

Target Performance for ITER(>1MW, >500s, >50%)

~2003

~2008

~2005

Achievement of ITER
requirement(1MW, 800s, 55%)
Quasi-steady state
operation with 0.8MW
outputPulse Duration (s)Fig.3-10 The plasma heating device“gyrotron” for ITERHeight is 3m, and weight is 800kg. The
gyrotron is installed in a superconducting
magnet of 7T field strength. Synthetic
diamond is used as the output window.

Fig.3-11 The progress of ITER gyrotron development at JAEA
This gyrotron is the first in the world to exceed the requirements of ITER
(1MW, 500s, 50% efficiency). (ITER gyrotron is also under development in EU(1.2MW, short pulse) and Russia (1MW, 110s).

A gyrotron is an electron tube which can transduce electron
beam energy into electromagnetic wave energy in the range ofseveral GHz ~ THz. It will be used as a plasma heating device
in the International Thermonuclear Experimental Reactor(ITER) (Fig.3-10). The high temperature plasma needs to be
one hundred million degrees Celsius for the fusion reaction,
which can be realized by the injection of RF beams like a
microwave oven.
The ITER gyrotron is required to have 170 GHz frequency (70times as high as that of microwave oven), 1MW output (2000
times as high as that of microwave oven), 50% efficiency (ratioof DC input power to RF power) and 500s pulse duration, which
is a typical pulse time of ITER. We started the development of
the high power gyrotron for ITER about 15 years ago and havedeveloped advanced technologies of energy recovery, highpower oscillation, and synthetic diamond windows ahead ofother countries. Furthermore, we combined the technologies
of steady state oscillation (to control the beam current) and ofchanging the rotational frequency of the electron beam and therotation ratio (the ratio of the rotation speed to the travel speed)
during the oscillation, which is a source of RF energy. The

gyrotron succeeded in stably-shifting the operating region to thehard-self-excitation region, which achieved higher efficiency
than the normal operating region. As a result, the oscillation
efficiency of 55% was attained with 1MW power and the pulse
duration greatly exceeded the performance required for the
ITER gyrotron. Furthermore, quasi-steady state oscillation of
1 hour, which is necessary for the steady state operation phaseplanned for ITER, was achieved with 0.8MW output power
and 57% efficiency (Fig.3-11). These results show that for, thefirst time in the world, the requirements of the ITER gyrotron
have been met. The achievement is a great step forward for the
ITER project. The physics mechanism by which high efficiency
oscillation in the hard-self-excitation region can occur was
clarified experimentally at the same time. This progress willstabilize and heighten gyrotron performance, and it will greatly
contribute to the improvement of the plasma heating system
in ITER. Application of gyrotrons to other fields, public andprivate, is also expected.
This study was awarded the Technological Prize (Research
Section) in 2008 by the Minister of Education, Culture, Sports,
Science and Technology.

Reference
Sakamoto, K., Kasugai, A. et al., Achievement of Robust High-Efficiency 1MW Oscillation in the Hard-Self-Excitation Region by a 170GHz ContinuousWave Gyrotron, Nature Physics, vol.3, no.6, 2007, p.411-414.
JAEA R&D Review 200837

Nuclear Fusion Research and Development

3-6

Development of a Large Negative Ion Source–R&D for Maintenance Free Ion Source for Uniform Negative Ion Beam–

expansion camber
RF antenna(a)

RF driver

RF antenna

plasmapermanent
magnetexpansion chamber

tent-shaped
magnetic field(b)cusp field

ceramic chamber(Ǟ18cm)RF power (kW)●RF disharge㧗tent-field ●RF discharge

lower pressure
lower powerGas pressure (Pa)Fig.3-12 RF negative ion sourceThis source consists of a ceramic chamber (φ18 cm) for
plasma production and a plasma expansion chamber (48cm (height) × 24 cm (wide) × 20 cm (depth)). Plasma is
produced with the RF antenna wound around the ceramicchamber.

Fig.3-13 (a) Cross section of RF negative ion source
and magnetic field lines, (b) operational region of RF
negative ion source with and without the tent-shaped
magnetic field

In “ITER” and “DEMO”, neutral beam injectors (NBI)
play an essential role for heating of the core plasma up to 100
million °C. In NBI, negative ions of hydrogen or deuterium areaccelerated in a 1MV-class and are transformed into neutral
beams. Such a high energy neutral beam must be stably inthe core plasma for a long period. Therefore, the negative ion
source must operate at high performance with high power, good
uniformity and long lifetime. In NBIs installed in existing large
fusion devices, the spatial profile of the negative ion beam is not
uniform. This non-uniform beam causes local beam divergence,
which results in high heat load on acceleration grids and beamline components. We have discovered that a magnetic field in
the negative ion source induces the non-uniformity of negative
ion beam. With a modified configuration with a tent-shaped
magnetic field, we have achieved uniform negative ion beam
extraction.
In conventional negative ion sources, negative ions are
extracted from source plasma produced by heating filaments.
The filaments often break during high power operation, which
prevents a long-term operation of the negative ion source.

Since the maintenance period is planned to be twice a year
during “ITER” operation, the negative ion source must holdup
under repetitive operation for six months.
For a maintenance-free ion source, we have started
a development of an RF negative ion source (Fig.3-12)
which does not use the filaments for plasma production. An
electromagnetic wave is generated with an RF antenna, which
ionizes a working gas, thus producing RF plasma. In caseof hydrogen gas, plasmas produced with the RF waves of
2MHz could not be sustained under the 4 Pa in the ion source.
However, the negative ion source for “ITER” is required tooperate at 0.3 Pa to reduce the stripping loss of accelerated
negative ions by collisions with residual gases.
Here we have applied a tent-shaped field combined witha cusp field formed by permanent magnets (Fig.3-13). As aresult, plasma loss on the wall was suppressed, enabling RF
plasma production even when gas pressure is under 1 Pa andwith lower RF power. This method is effective for the lowpressure operation of a RF negative ion source needed for the
ITER NBI.

Reference
Tobari, H. et al., Uniform H- Ion Beam Extraction in a Large Negative Ion Source with a Tent-Shaped Magnetic Filter, Review of Scientific Instruments,
vol.79, 2008, p.02C111-1-02C111-4.

38

JAEA R&D Review 2008

Nuclear Fusion Research and Development

3-7

Demonstration of Final Installation to within
0.5mm of a 4-ton Blanket by Remote Handling–Improvement in a Robotic Positioning Control Scheme for ITER–

Vehicle travelingon the rail

BlanketVacuum vessel(VV)
Maintenance port

Rail support

VehiclePinionRail
MaintenancePort(a) Rail deployment in VV(b) Arcing of the rail

Fig.3-15 Installation of robot in VV (Rail deployment)

Rail

Rail support

Fig.3-14 The ITER blanket module maintenance robot
composed of a vehicle and a rail
We have adopted a modular structure of about 400 parts
to enable maintenance of damaged blanket modules only.
Blanket maintenance requires the remote manipulationof 4-ton modules, with final installation to within 0.5mm
between the module and two keys on the vacuum vessel(VV). Designing a remote maintenance robot that can install
modules on 2 keys with a chamfer requires a high degreeof positioning accuracy.

BlanketEnd-effector

Fig.3-16 Blanket handling for high positioning accuracy

The inner lining (termed a “blanket”) of the vacuum vessel(VV) of the International Thermonuclear Experimental Reactor(ITER) will be activated by 14 MeV neutrons generated byfusion reactions. Maintenance of the blanket in the VV thus iscarried out under high gamma ray radiation conditions (~0.5kGy/h). All interventions inside the VV therefore must be
performed remotely. The principal features of this maintenancerobot are as follows:- The vehicle travels on a rail and has telescopic arms that can
extend up to 6m. The rail is supported every 90° around theannular vessel, which provides sufficient stiffness (Fig.3-14).
- The rail is deployed by the vehicle which will be used fortraveling along the rail (Fig.3-15). The rail therefore canbe composed of simple structures that are not sensitive toradiation.
We have demonstrated rail deployment into the VV andthe basic capabilities needed for blanket handling with a highdegree of positioning accuracy.
The most critical issue of sensor-based control of positioning
is designing a measurement scheme to compensate for the
relative error between the location of the mechanism that grips
the blanket and the target position for installing the blanket,

thereby minimizing assembly error. A contact method using aforce sensor that is uninfluenced by assembly error is employedfor positioning. However, a technical problem of measuring via
contact sensors is to allow for sensing time so as to be able
to measure along each of the six degrees of freedom, rotation
about three axes and linear movement along three coordinates.
To solve this problem, we invented a new control scheme which
initially determines the degree of each rotation. This initial
measurement scheme limits the number of possible degrees
of freedom to the three linear coordinates. The proposal wastested using a full-scale robot in order to confirm the feasibilityof the new control scheme (Fig.3-16). The new control scheme
resulted in a positioning accuracy of less than 3mm, within thepositioning requirement of 5mm. Sensing time was reduced aswell.
In conclusion, we found that the new proposal for improving
the control scheme is appropriate for reducing sensing timeand achieving the required positioning accuracy. Based onthese results, we will proceed toward procuring this remote
maintenance robot in order to deliver an efficient, reliable
remote handling system to ITER.

Reference
Kakudate, S. et al., Effect on Positioning Accuracy due to Assembly Error of Large Robot for ITER Blanket Maintenance, Purazuma, Kaku Yugo Gakkaishi, vol.84, no.5, 2008, p.269-278 (in Japanese).
JAEA R&D Review 200839

Nuclear Fusion Research and Development

3-8

Materials Necessary for Stably Supplying Fuel
Tritium to Fusion Reactors

Heating to high temperaturesin hydrogen atmosphereBlackened by reduction
Without Li addition
Li2TiO3

No change
White sample With Li addition
Li2TiO3(Li2TiO3)

Molar fraction of oxygen deficiency(10-3mol)–Development of Tritium Breeding Materials Resistant to Reductionin Hydrogen–

Reduction (Ar-5%H2)1000㷄30h

Additive
None
Ca
Li

Below
detectable
limit
None

Additive
Fig.3-17 Color of Li2TiO3 after heating to high temperaturesin hydrogen atmosphere
The color of Li2TiO3 without added Li changed from white
to black in a hydrogen atmosphere at high temperatures.
This color-change corresponds to reduction of Li2TiO3. In
the case of Li2TiO3 with added Li, the color did not change,
indicating that this sample was not reduced in the hydrogen
atmosphere.

Fig.3-18 Molar fraction of oxygen deficiency of Li2TiO3with different additives
Reduction of Ti in Li2TiO3, namely, the valence change
from Ti4+ to Ti3+, is accompanied by oxygen vacancies in
the crystal structure. The Li/Ti = 3.9 sample exhibited nooxygen vacancies, which indicates clearly that it was highly
resistant to reduction.

D-T fusion reactors need deuterium (D) and tritium (T) as
their fuel. Since tritium does not exist in nature, it is necessary
to produce tritium artificially by neutron irradiation of alithium-filled blanket.
Lithium titanate (Li2TiO3) has been recognized as one of the
primary candidates for tritium breeding materials because of its
good tritium release and its low activation. Addition of H2 to theinert sweep gas at the blanket has been proposed for enhancing
the release of the generated tritium from the breeder material.
However, Li2TiO3 is reduced in a H2 atmosphere at an operating
temperature above 600°C, which will lead to degradationof the tritium release characteristics. The reduction of Ti in
Li2TiO3 from Ti4+ to Ti3+ is accompanied by mass decrease
due to decrease in oxygen content of Li2TiO3. In a previous
study, Li2TiO3 (Li/Ti > 2.0) to which CaO was added exhibited
smaller oxygen deficiency than Li2TiO3 without such addition.
As the next research step, development of Li2TiO3 with excess
Li was attempted to improve its resistance to deoxidization athigh temperatures.
Usually, solid state reaction of Li2CO3 and TiO2 is used in

the synthesis of Li2TiO3. However, Li addition by solid state
reaction was found to be difficult. In the present study, we havedeveloped a new process, in which Li alkoxide and Ti alkoxide
are mixed to add Li2O to Li2TiO3.
Fig.3-17 shows a photograph of Li2TiO3 samples heatedin hydrogen atmosphere at high temperatures. The color of a
Li2TiO3 sample changed black, while that of a Li2TiO3 sample
with added Li remained white. This black discoloration
indicates valence change from Ti4+ to Ti3+, which accompanies
decrease in the oxygen content of the sample.
Fig.3-18 compares the molar fraction of oxygen deficiencyof the Li2TiO3 with added Li to Li2TiO3 of our previous study
with added CaO. The molar fraction of oxygen deficiency
increased as the molecular ratio Li/Ti decreased. The result for
Li/Ti = 3.9 indicated no oxygen vacancies.
The overall results indicated that Li addition is very effective
in suppressing oxygen deficiency. Thus, an advanced tritium
breeding material for high temperature use in fusion reactors
has been developed through successful synthesis of Li2TiO3with added Li.

Reference
Hoshino, T. et al., Non-Stoichiometory and Vaporization Characteristic of Li2.1TiO3.05 in Hydrogen Atmosphere, Fusion Engineering and Design, vol.82,
issues 15-24, 2007, p.2269-2273.

40

JAEA R&D Review 2008

Nuclear Fusion Research and Development

Tritium Extraction Using a Fuel Cell

Proton Conductor
Mass Transfer Rate (x10-7 mol-H2/s)

Porous Electrode873K
0.4Ͱ/minObservation

H2
Current

H2 Conc. at Perm. Side Outlet (ppm)–Tritium Recovery from Blanket Sweep Gas–

Current (mA)

3-9Applied Voitage (mV)

Sweep Gas IN(from Blanket)
Sweep Gas OUT(to Blanket)

HeaterRecoveredGas(to Fuel Proc.
Sys.)
H2,HT

Electrode

Mass Transfer Resistance ob H2 (ǡm)

Fig.3-20 Water decomposition and extraction of hydrogen
He with H2O at 2668 Pa was fed to the anode side, andhydrogen was measured at the cathode (Perm.) side.

Flow Rate: 0.4Ͱ/min
H2 Conc. (Perm. Side Inlet)

Temp.: 873K
Applied Voltage: 0.8V

Ceramic Proton
Conductor
H2 Conc. at Feed Side Inlet (%)

Fig.3-19 The principle of the hydrogen pump (upper),and the schematic of a hydrogen pump using proton
conductor ceramic membrane in test tube shape(down)

Fig.3-21 Correlation of the mass transfer resistance ofhydrogen through the membrane and hydrogen gas
concentration
Hydrogen concentrations at both sides were changed.

A fusion reactor is a system extracting the energy of thenuclear fusion reaction between deuterium and tritium.
However, tritium is scarce in nature. Therefore, in a fusion
reactor, tritium is made by the nuclear reaction between
the neutron and the lithium in the blanket. Tritium bred in
the blanket is purged out by helium (He) sweep gas, and isseparated from He at a recovery system. To operate the fusion
reactor, we have to make more tritium than is burned up in thereactor.
Possible tritium recovery method includes metallic film
permeation and use of hydrogen storing alloy. However, in afusion DEMO reactor, the temperature and flow rate of thesweep gas are expected to increase greatly. A more effective
method is needed. In response, we directed our attention to
the hydrogen pump an apparatus using the principle of thefuel cell in reverse. When electric potential is applied to theproton conductor (fuel cell), hydrogen can move inversely
against a hydrogen concentration gradient and can be extracted
selectively (Fig.3-19).
We selected a ceramic proton conductor since we would
use it at a high temperature. The proton current which canbe passed through the ceramic conductor is quite small incomparison with that of a polymer electrolyte, but we could
extract hydrogen from water vapor using the ceramic conductor(Fig.3-20). In addition, we found out that the rate of decreaseof the mass transfer resistance is unchanged with the increaseof hydrogen gas concentration (Fig.3-21).
We thus established the fundamental feasibility of thehydrogen pump. Hereafter, our effort shifts to the practicalapplication of the hydrogen pump, requiring suchaccomplishments as the enhancement of pumping performance.

Reference
Kawamura, Y. et al., Mass Transfer Process of Hydrogen via Ceramic Proton Conductor Membrane of Electrochemical Hydrogen Pump, Fusion Engineeringand Design, vol.82, issue 2, 2007, p.113-121.
JAEA R&D Review 200841

Nuclear Fusion Research and Development

3-10 Torus Configuration and Maintenance Concept ofCompact Fusion DEMO Reactor–Torus Configuration for High Operation Rate–cryostatcryostatsectorcryostat door

blanketdivertor

toroidal coilFig.3-22 Concept of torus configuration and maintenancefor fusion DEMO reactor

cask

turn-table

Fig.3-23 Sector transport for maintenanceA cask containing a sector moves on rails placed on thefloor and turn-tables.

saddle shaped shell
replaceable blanket
120qassembly of
saddle shaped shells

Fig.3-24 Structure of sector and saddle shaped shell
Eddy current on the shell assembly stabilizes the burning
plasma.

permanent blanket

eddy current
apparent
current
shieldA conceptual study of compact fusion DEMO reactor design
resulted in a torus configuration and maintenance scheme which
shortens the time required for periodic replacement of corecomponents such as blanket and divertor. Such shortening ofmaintenance time can lead to an improvement of plant operation
rate, and thus to a reduction of the cost of electricity.
To quicken maintenance, we adopted a “sector maintenancescheme” in which the core components are divided into
twelve toroidal sectors and each sector is en bloc removed
and transported to a neighboring hot cell for maintenance andreplacement of blanket modules (Fig.3-22). Each sector loads
about sixty replaceable blanket modules that must be replaced
every two or three years. In the replacement, six steel tubes
connected to each blanket module (four for cooling and two
for tritium fuel recovery) must be cut, re-welded and tested byremote control. When these replacement procedures are carriedout in situ in the cryostat, the replacement time is expected
to exceed half a year, unacceptable from the point of view ofoperation rate. With sector maintenance, the required time will
be reduced to about three months because, by replacing used
sectors with spare sectors, time-consuming procedures (pipecutting re-welding and testing) can be carried out in the hot cell
during tokamak operation. On the other hand, this maintenancescheme raises concern about transport of heavy sectors. Thisissue is resolved with the use of a wheel and bearing systemand turn-tables (Fig.3-23). A saddle shaped conducting shellplaced in between replaceable and permanent blankets is one of
the key components of DEMO. The assembly of the shells also
plays the role of passively stabilizing the plasma by inducing
eddy currents, as shown in Fig.3-24.

Reference
Tobita, K. et al., SlimCS－Compact Low Aspect Ratio DEMO Reactor with Reduced-Size Central Solenoid, Nuclear Fusion, vol.47, no.8, 2007, p.892-899.

42

JAEA R&D Review 2008

4 Quantum Beam Science Research
The Future Explored by Quantum Beam Technology
functions, allowing us “to create” more. High quality beams
of atoms or ions enables us to observe and create with nm
level precision. Through the observation and fabrication of
atomic arrangement, electronic state, and various elements,
we can explore the mystery why a substance becomessuperconductive, or the various functions and reactions
the human body has. We also are contributing to various
important tasks such as developing new medicines, catalysts
of exhaust gas from automobiles using less noble metal, andpost-oil fuels made from plants, which are carbon-neutral.

What are quantum beams?
Quantum beams are high quality beams including
electromagnetic waves (laser, X-ray, γ-ray, etc.), leptons(electron, positron, etc.) and hadrons (proton, neutron, ion,etc.), which possess both wave characteristics and particle
characteristics. At JAEA we are utilizing these quantum
beams (neutrons, ions, electrons, γ-rays, high intense ultrashort pulsed lasers and synchrotron radiation) in innovative
R&D (Fig.4-1). We thereby aim to contribute to the progressin science and technology and promotion of industry.
Specifically, we are dedicated to R&D in the “Four Priority
Fields to be Promoted” contained in the “Science andTechnology Basic Plan”.
These fields are: development of highly durable fuel cell
membrane (nanotechnology and materials area), fine structure
analysis of proteins aiding drug discovery, analyzing DNA
damage and restoration mechanism and the creating new
species (life sciences and biotechnology area), developmentof materials and technologies for environmental protection(environmental sciences and energy area), and developmentof radiation-resistant semiconductors (information andcommunication technology area). Further, in the advanced
medical treatment area, techniques for miniaturization ofparticle beam radiotherapy equipment by generating proton
beams with high intensity lasers are being developed.

Recent Accomplishment
As specific examples of our accomplishments, we haveexperimentally demonstrated the principle of the flying plasma
mirror with a high intensity laser, and this result suggests thepossibility of generating ato (10-18) second quantum beams
from tunable lasers. In Topic 4-1 and Topic 4-5 we show
our steady progress in generating high energy particles withlasers, for medical applications. In Topic 4-3 and Topic 4-6,
we show the possibility of using quantum beams in basic
science by analyzing the mechanism by which a material
has a huge negative coefficient of thermal expansion, or aheavy ion that has a special effect on living cells. In Topic
4-4 and Topic 4-9, we applied radiation-induced cross-linking
and graft techniques utilizing electron beams and γ-rays todevelop a post oil material made from carbon-neutral plants,
and to develop an electrolyte film that operates at the hightemperatures of a fuel cell. We promote collaboration inside
JAEA so as to make basic science contributions to solving
problems facing the Fast Breeder Reactor project, for instanceresidual stress measurement with neutrons and synchrotron
radiation.

The characteristics of quantum beams
In this R&D, the characteristics of each quantum beam are
utilized effectively and organically, leading to various uniqueachievements only possible for JAEA.
Quantum beams are excellent at revealing structure,
allowing us “to see” better, and have new fabricating

Quantum Beam Technology
nanotech material

life science biotechnology

environment energy

information
communication

advanced therapy

Quantum beam explores the Advanced Science and Technology
J-PARC neutron

high intensity laser

SPring-8
synchrotron
radiation

Wide application
High quality

JRR-3‫ޓ‬neutron

TIARA
Ion beam

From Radiation
To Quantum Beam
Progress ofapplication technology

Discovery of
new phenomena

ionizing radiation = photon or particle with ability to ionize
air directly or indirectlyǿ, ǩ, Ǫ, ǫ, neutron, proton, electron, etc.

radioactive technologythat is put to practical use‫ޓ‬Semiconductor production,
radial fire production, foodirradiation

Electron beam
60Co

ǫ-ray

Fig.4-1 The Quantum Beam Facilities at JAEA
JAEA R&D Review 200843

Quantum Beam Science Research

4-1

Making a Mirror Moving Nearly at the Speed of Light by Lasers–Demonstration of Light Reflection and Frequency-Upshifting by
Electron Density Modulation in Plasmas–
Top-view monitor

Probe laser
monitors

MagnetElectrons
XUV Spectrograph
Source laser pulse

Gas-valve

Signal intensity (a.u.)

150
Reflected light

100

50

0

Probe laser pulse-50

Driver laser pulse

5

10
Wavelength (nm)15

Fig.4-2 A schematic of the proof-of-principle experimentof the flying mirror
The Driver pulse was focused onto a gas-jet to generate
flying mirrors. The Source pulse was aimed at the generated
mirrors with a crossing angle of 45°. The reflected pulses
were in almost the same direction of that of the Driverpulse.

Fig.4-3 The spectrum of the light (X-ray) reflected byflying mirrors generated in the plasma
The signal was observed at the wavelength of 13.4nm.

Recently, focused laser intensity has been increasing
dramatically, and such intense lasers enable us to explore
various fields of science such as nonlinear optics, quantum
beam generation, and so on. However, conventional highpower lasers must be made very large to attain intensity higher
than 1022 W/cm2. Addressing this issue, a revolutionary idea
was proposed at JAEA in 2003 making large-scale lasers
unnecessary. In the proposed method, a moderate intensity laser
is focused onto a plasma, then some of the plasma electrons
start to group together. This agglomeration of electrons acts
as a mirror (relativistic-speed flying mirror) which reflects
and focuses incoming light partially. The characteristics ofthis flying mirror are summarized as follows: (1) the mirror
is stable against intense lasers because it consists of electrons
instead of common optical components such as metals; (2) highfocused intensity is possible because the reflected light can befocused to a smaller spot due to the wavelength reduction, or
frequency upshifting, caused by the Doppler effect; (3) the
pulse duration of the reflected light also is compressed byfrequency upshifting, which increases the intensity.
Up to now, a proof of this principle has not been made becauseof the experimental difficulty of colliding two laser pulses withmicrometer accuracy. We succeeded in demonstrating thisconcept by careful design of the experiment and the fine tuningof the colliding of two laser pulses using crossing Ti:Sapphirelasers. As shown in Fig. 4-2, Ti:Sa laser pulses with a peak powerof 2 TW and a pulse duration of 80fs were focused onto heliumgas. In order to conduct the experiment, we accomplished thefollowing: (i) we improved the pointing stability of lasers by
eliminating vibration sources; (ii) we invented a new method toalign two laser pulses to collide in a small designated region. A
small portion of the laser energy split from the main laser wasused as a exposure light and the colliding region in the plasma
was thereby monitored with a device similar to a microscope. In
addition, viewing the scattered light from the region through a
top-view monitor was also implemented to achieve colliding.
When two laser pulses collided properly, we observed
extreme ultraviolet frequency photons as seen in Fig.4-3. The
incident source laser whose wavelength (frequency) was 780
nm (380 THz) was successfully compressed (upshifted) to 13.4
nm(22,300 THz). The number of the reflected photons per unit
solid angle is estimated to be 3 × 107/sr, which is 100 times
higher than that expected from incoherent Thomson scatteringof plasma electrons. In addition to this, we carefully analyzed
experimental data, and we conclude that these photons wereobtained by reflection from the electron agglomerate formedin the plasma.
This is the first demonstration of the flying mirror concept,
confirmed by observing the frequency upshift of the reflected
photons. This method will lead to an ultra-high intensity of 1029
W/cm2, where electron-positron pairs would be produced fromvacuum. Moreover, because this method also compresses pulse
duration, it may help us achieve attosecond X-rays, a field
which is just now opening up.

Reference
Kando, M. et al., Demonstration of Laser-Frequency Upshift by Electron-Density Modulations in a Plasma Wakefield, Physical Review Letter, vol.99, 2007,
p.135001-1-135001-4.

44

JAEA R&D Review 2008

Quantum Beam Science Research

4-2

Observation of a Key Hydrogen Atom in an Enzymatic Reaction–Neutron Structure Analysis of a Compound Modeling theReaction Center of [NiFe]Hydrogenase–S
Ni

Purple:Carbon atoms

N

H?

N

Fe

S

S

N

CO

S

H

N

Fig.4-4 Deduced structure of the active site of [NiFe]hydrogenase
The activation center of [NiFe]hydrogenase is thought to have sucha Ni-H-Fe structure. S’s are sulfur atoms.

O

Ni

S

Ru

S

N
N
Ni

O

H2

S
Ru
S

,

N

H+
N
O

Ni

Fig.4-6 The hydrogen atom between the Ni and
the Ru atoms observed by neutron diffraction
The red mesh in this figure shows the distributionof negative nuclear scattering length density
observed by neutron diffraction (the water-bluemesh shows the positive distribution). In neutron
structure analysis, hydrogen atoms are observed
as negative distribution. Consequently, a large
red mesh between Ni and Ru clearly shows theexistence of a H atom. This result shows that thiscompound has Ni-H-Ru structure.

H
S

Ru

S

Fig.4-5 Reaction of the model compound of [NiFe]hydrogenase
The reaction from (a) to (b) is similar to the enzymatic reactionof [NiFe]hydrogenase, and (b) is similar to the reaction center of[NiFe]hydrogenase shown in Fig. 4-4 (only Fe was replaced by Ru).
Thus, (b) can be considered a compound modeling the reaction
center of [NiFe]hydrogenase.

Hydrogenases are bacterial enzymes that catalyze theactivation of H2 into two protons (H+) and two electrons (e-).
Their structure when they are active in this enzymatic reaction
has been attracting much attention since this would elucidate
the reaction mechanism. Though a Ni-H-Fe structure, a nickel
atom (Ni) and an iron atom (Fe) bridged by a hydrogen atom
as shown in Fig.4-4, has been proposed as a possible structure,such a structure has not been observed.
Recently, Prof. Ogo’s group at Kyushu University successfully
synthesized a model compound with reactivity similar to thereaction center of a [NiFe]hydrogenase, [NiFe]Hase, by using
ruthenium (Ru) instead of Fe as shown in Fig.4-5. Whether
this compound has a Ni-H-Ru structure or not is important for
determining the activation structure of the [NiFe]Hase and to
elucidate the mechanism of the enzymatic reaction. However,observation of the hydrogen atom is difficult by the usual
structure analysis method, X-ray diffraction, and consequently

the Ni-H-Ru structure had not been observed.
Thus, JAEA carried out single crystal neutron structure
analysis of this model compound using the “BIX-3”
diffractometer at “JRR-3” reactor. A neutron is well scatteredby a hydrogen atom and thus a neutron beam is suited forobservation of hydrogen atoms. We measured the strengthof 10161 diffraction spots in 9 days of measurement andsuccessfully observed the hydrogen atom between the Ni and
the Ru atoms as shown in Fig.4-6. Consequently, this modelcompound was confirmed to be the first compound which hasa Ni-H-Ru structure, which indicates that the activation site of
the natural [NiFe]Hase has a similar Ni-H-Fe structure.
This result may lead to development of new hydrogen
activation catalysts to produce hydrogen energy resources. This
research was carried out by Kyushu Univ., JST, Osaka Univ.,
Univ. of Hyogo and JAEA and published in Science in 2007.
JAEA carried out the structure analysis by neutron diffraction.

Reference
Ogo, S., Ohhara, T. et al., A Dinuclear Ni(μ-H) Ru Complex Derived from H2, Science, vol.316, 2007, p.585-587.
JAEA R&D Review 200845

Quantum Beam Science Research

4-3

Why Do Heavy Ions Have Great Effect on a Cell?
Chemistry Elucidates the Reason

Distribution of the radicals along with the trajectory is
decided by the mass and energy of the incident ion.

Yield decreased with mass

–Quantitative Analysis of Reactive Species Which Induce DNA Damage–
Yield increased with specific energy
Yield（radicals/100eV）

3

Heavy ion

He ion2

C ion
1

Ne ion0

Aqueous solution

0

10

20

Specific energy (MeV/n)

Fig.4-7 Radicals generated along heavy ion trajectoryin waterA heavy ion deposits its kinetic energy densely along its
trajectory to produce reactive species. Their distributions are
decided by the mass and specific energy of the incident ion.
Moreover, reactive species react and diffuse with time.

Fig.4-8 Dependence of yield of hydroxyl radicals on
specific energy and mass of incident ion
Yields of hydroxyl radicals increased with the specific energy
for each ion, but decreased with the mass of each ion at
the same specific energy.

Yield（radicals/100eV）

4

He ionRadical yield edcreased
by diffusion and reaction

3

Fig.4-9 Yield of hydroxyl radicals decreased with elapsed time
just after helium ion irradiation
Yield of hydroxyl radicals decreased in ns time scale, with diffusion
and reaction of the radicals. The same tendency was observed inother energies and ions.

2

1

0

1

10

100

Elapsed time（ns）

1000

High energy heavy ions are used for plant breeding andcancer treatment. The heavy ion can produce highly reactive
species (e.g. hydroxyl (OH) radical) with a different distribution
from that of conventionally used radiation, such as γ-rays, in acellular tissue, i.e. an aqueous solution, as shown in Fig.4-7.
Although the characteristics of the heavy ion irradiation effects
can be explained in terms of LET (linear energy transfer) inmany cases, one cannot explain the spatial distribution of thereactive species using LET. Moreover, the reactive species
diffuse and react with time. Since the OH is considered themost important basic radical for radiation chemistry research,
we estimated the yield of the OH radicals based on the massand energy of the incident ion, and the elapsed time.
We irradiated aqueous phenol solution with heavy ions,changing the mass and energy systematically, and estimatedthe OH radical yield from the yield of reaction products ofphenol. As a result, the yield of the OH radicals increased
with the specific energy in the case of each ion, but decreased
with the mass of the ion at the same specific energy as shownin Fig.4-8. The yield decreased with elapsed time (Fig.4-9).
These results can be derived from the initial distribution of thereactive species around the heavy ion trajectory, and subsequentdiffusion and reaction of the radicals.
Presently, we are constructing a time resolved opticalmeasurement system. We will thereby increase the reliabilityof data about the radical yield, and observe directly reactionsof radials with biomolecules.

Reference
Taguchi, M. et al., Yield of OH Radicals in Water under Heavy Ion Irradiation. Dependence on Mass, Specific Energy and Elapsed Time, Nuclear Scienceand Techniques, vol.18, issue 1, 2007, p.35-38.

46

JAEA R&D Review 2008

Quantum Beam Science Research

4-4

Biodegradable Elastic Gel from Cellulose Derivatives–Combination of Radiation-Induced Crosslinking and Aggregation–
Gel

Carboxymethyl
Cellulose(CMC)
Paste

Gel

Elastic Gel

ǫ-ray
or EB

Crosslinking
Point

Irradiation

Acid
Treatment
Aggregation byHydrogen
Bonding

Insolubilization
by Induction of
Crosslinking
Structure

Crosslinking
Point
Hydrogen
Bond

Broken
50% Compression

Removal of Load

Elastic Gel

Shape
RecoveryFig.4-10 Preparation of biodegradable elastic gel
10wt% or higher aqueous solution of water-soluble
carboxymethylcellulose is formed into a gel by γ-ray orelectron beam irradiation. Rubbery elasticity can be given to
the radiation crosslinked gel by immersion in acid solution.

Fig.4-11 Mechanical property of biodegradable elastic gel
At 50% compression, the carboxymethylcellulose gel crosslinked by radiation disintegrates into small pieces. On thecontrary, the elastic gel given acid treatment does not fallapart, even at 50% compression, and returns to its original
shape when the load is removed.

Development of cellulose, the chief material of plants which
is produced through growth by photosynthesis, has received
considerable attention as a carbon neutral high polymer.
Though polysaccharide gel, manufactured from cellulose withcrosslinking agents, absorbs hundreds of times its volume ofwater, its production is discouraged because the crosslinking
agents are toxic in general.
To solve this problem, the development of a method toprepare the gel from water-soluble carboxymethylcellulose(CMC) without using a crosslinking agent was attempted. As
a result, kneading CMC powder in water at a concentration of
more the 10wt% yielded a homogeneous paste, and when the
CMC in this condition was irradiated by γ-rays or electron
beams, the crosslinking occurred and biodegradable gel wasformed. Though the gel prepared by this technique had highwater absorbency, the area of its application was very limited
as the gel was fragile and brittle when swollen.
To solve this problem, the control of molecular structure
is important, and we utilized the molecular aggregation
caused by acid treatment to improve gel strength. We wereable to make the rubbery elastic gel by immersing radiationcrosslinked CMC in an acid solution (Fig.4-10). The reason forthis transformation is that a Na ion, which exists as a counterion of a carboxyl group in the radiation crosslinked CMC gel,is replaced by hydrogen during immersion in the acid solution.
As a result, the electrostatic repulsion power within the CMC
molecular chain decreases, and the carboxymethylcellulose
molecular chain agglomerates, pulled together by the hydrogenbond. Strong interaction between the molecules ensues.

Measuring the thermal decomposition characteristics of theelastic CMC gel, it was found that a broadening and shiftingof the peak to a higher temperature due to the decompositionof the CMC occurred only after immersion in the acid solution.
This indicates that there is strong intermolecular interaction
between CMC molecules.
Though the CMC gel made by the irradiation breaks up at
50% compression as shown in Fig.4-11, the elastic CMC geldoes not break and returns to the same shape before it wascompressed once the load is removed. The breaking strengthof the acid treated elastic CMC gel was 150 times that of theuntreated gel (3N/mm2). When the pH is lowered and theprocessing time of immersion is lengthened, this elastic CMC
gel has increased strength and becomes harder, regardlessof the kind of the acid. For instance, when the immersion
processing is done for four days by using 0.5M hydrochloric
acid, it becomes a gel that is about 100 times harder than that of
CMC gel formed by radiation-induced crosslinking alone.
Thus, we succeeded in making a novel biodegradable gelwith elasticity like rubber by the combination of radiationinduced crosslinking and acid treatment. Therefore, this elastic
gel promises to be a new carbon neutral material of plant origin
with a wide range of applications for cosmetics, environmental
protection, and medical treatment.
This work was supported in part by Japan Science andTechnology Agency, and is a product of the Gunma Prefecture
Collaboration of Regional Entities for the Advancement ofTechnological Excellence.

Reference
Takigami, M., Amada, H., Nagasawa, N. et al., Preparation and Properties of CMC Gel, Transactions of the Materials Research Society of Japan, vol.32,
no.3, 2007, p.713-716.
JAEA R&D Review 200847

Quantum Beam Science Research

4-5

Medical Applications for the Laser-Driven Proton Beam–High Quality Beam Is Produced with High Conversion Efficiency–

Laser

TargetL=2.2mX-raypinholecameramagnet 0°
TOF-0°
3.0

Our experimental results

Proton spectrum (1/MeV/msr)

Interferometer(wavelength = 800nm)

22.5°
TOF-22.5°1010

Conversion Efficiency

2.5

magnet

2.0
1.5
1.0

109

0.5

108

0.0
0.01

10

7

Sweden Lund Univ.
Neely et al
Appl.Phys.Lett.89,215002（2006）

0.1

1

10

Max-Planck-Institute
fur Quantenoptik
Kaluza et al
Phys.Rev.Lett.
（2004）
100 93,45003

Polyimide Target Thickness (Ǵm)

106
105

1 ‫ޓޓޓޓ‬2 ‫ޓޓޓޓ‬3 ‫ޓޓޓޓ‬4

Energy (MeV)

Fig.4-12 Experimental set up and the energy spectrumof the proton beam
The magnet in front of the time-of-flight spectrometer is set
there to eliminate background noise from electrons.

Fig.4-13 The experimental results obtained by the~1 J class laser system
Our efficiency is higher than those obtained by othergroups.

A high energy proton beam is produced by the interaction
between a short-pulse high intensity laser and a target material
such as metal or plastic. We can down size the accelerator using
this laser-driven proton beam, because the laser-driven proton
beam has very small (~μm) acceleration field compared withthat of the conventional accelerator. However, for the practicaluse of this laser-driven accelerator in industry or medicine,
we have to obtain the necessary proton dose within a limited
time period with compact laser system. We have to increase
the energy conversion efficiency from laser to proton beam inorder to obtain a higher proton number per laser shot. In orderto obtain conversion efficiency improvement of more than a
few %, a large laser system whose energy is more than a few
tens of J is needed. The energy conversion efficiency with asmall laser system (~1 J class) is limited to less than 1%.
We have carried out an experiment in collaboration with
GIST in Korea, CRIEPI, and ILE. We produced a proton beamwith maximum energy up to 4 MeV with good conversion
efficiency, by irradiating a polyimide 7.5μm target with laser
pulses whose peak intensity and pulse width were 50 TW
and 34 fs, respectively, from a compact 1J class laser system(Fig.4-12 and Fig.4-13). The conversion efficiency from laser
to proton was more than 3%. The peak current was 0.3 MA,
which is more than 2 orders of magnitude higher than thoseobtained by a conventional accelerator with energy on theorder of a few MeV.
This realization of a laser-driven proton beam with highconversion efficiency, will surely contribute to the applicationof the laser-driven proton beam to medical as well as industrial
fields. This increase in the conversion efficiency will lower
the quantitative demands made upon the laser (presently laser
energy of 10 J, repetition rate of 100 Hz). This makes it easierto reach our goal of realizing a laser-driven proton accelerator.
We have shown that a compact laser system can accelerate
the proton beam very efficiently. This paves the way toward
the application of the laser-driven proton beam in industrial aswell as medical fields. However, problems still remain. First ofall, we have to increase the energy of the proton at least to 80
MeV in order to apply the laser-driven proton beam to radiation
therapy, if no additional accelerator is attached. In the next step,
we have to develop the transport and irradiation system of the
proton beam, where the spatial distribution and number of the
protons as well as the energy is optimized, in order to supply
the necessary and sufficient dose to the diseased part of thepatient. We are now developing not only the source of the laserdriven proton but also the transport and irradiation systems, inorder to realize the laser-driven proton accelerator.

Reference
Nishiuchi, M. et al., Efficient Production of a Collimated MeV Proton Beam from a Polyimide Target Driven by an Intense Femtosecond Laser Pulse,
Physics of Plasmas, vol.15, issue 5, 2008, p.053104-1-053104-10.

48

JAEA R&D Review 2008

Quantum Beam Science Research

Local Structure in a Material Shrinking with Heat

–Broadening of Magneto-Volume Effect (Invar Effect) Caused by
Local Lattice Distortion–（Cu,Ge）

N
Mn

Intensity㧔a.u.㧕

4000

Mn3Cu1-xGexN

Peak

3000

x=0.7

G
（r）
（a.u.）

4-6

2000

0.15

Peak

x=0.5
x=0.15

1000

0.5

0.7

T=300K

00

Fig.4-14 Crystal and magnetic
structures of Mn3Cu1-xGexN
Cu or Ge atoms form a cube,
and Mn atoms locate at thecenter of each face, forming
an octahedron. A N atom
locates at the center of theoctahedron. Red arrows show
the direction of Mn moments.

100

200

300

400

500

600

700

Fig.4-15 Temperature (T) dependencesof intensities of magnetic scattering(main panel) and lattice constants(inset) of Mn3Cu1-xGexN(x=0.15,0.5,0.7)
Intensities of magnetic scatterings ofsamples, whose lattice constants gradually
increase in T-region shown by red circle,
gradually increase.

Generally, a material expands with increasing temperature(T). However, some rare materials mysteriously contract withincrease in T. One of the origins of such phenomenon is themagneto-volume effect (MVE), in which volume expansion
accompanies magnetic ordering with decreasing T. There is a
special case of this MVE in which the contraction is gradualwith rise in T, the so-called Invar effect. The mechanism forthis is not yet clear although it has been studied for about
100 years. According to the study by Takenaka and Takagi of
RIKEN concerning the material Mn3Cu1-xGexN (Fig.4-14), thesystem with Ge-concentration of x~0.15 exhibits sharp volume
contraction with rise in T by MVE, and the volume-change withrise in T becomes gradual with increase in x. At around room
temperature, the rate of volume contraction is the largest in thematerial with x~0.5. In order to investigate the mechanism ofmoderation of the MVE in this material, we performed neutron
powder diffraction measurements.
To study the crystal and magnetic structures of this system
including the material with x<0.15 which does not exhibit
MVE, we performed neutron diffraction measurement at
JRR-3. Only the material with the cubic structure and thearrangement of Mn-moment shown in Fig.4-14 was found
to exhibit large MVE. As shown in Fig.4-15, a material with
gradually increasing Mn-moment (corresponding to magnetic

0.15

0.20

0.25

0.30

0.35

Fig.4-16 Atomic pair distribution function
G(r) of Mn3Cu1-xGexN(x=0.15,0.5,0.7) obtained
at 300 K (main panel) and the rotation of
Mn-octahedron expected from G(r) (inset)
Arrows of peak (2) show atomic correlation
between Mn and Cu (Ge). The peak splitting
increases with increasing x.

scattering intensity) exhibits gradual volume expansion.
In this material, crystal and magnetic structures do notchange and only T-dependence of magnitude of orderedmagnetic moment changes against x. We focused on a possible
local lattice distortion caused by Cu and Ge having different
atomic sizes. We derived the atomic pair distribution function
G(r) from the total scattering measurement made at Los Alamos
National Laboratory, to study the local structure. G(r) is shownin Fig.4-16. The peak (1) around 0.19 nm shows the relation
between Mn and N atoms, and the peak (2) around 0.28 nm
superposes the relation between Mn atoms and the relation
between Mn and Cu (Ge) atoms. Peak (1) is a sharp single peak.
In contrast, peak (2) splits in to two peaks shown by arrows.
The former suggests that the shape of the Mn-octahedron does
not change (see inset), and the latter indicates that the rotationof the octahedron causes both short and long distances between
Mn and Cu (Ge) atoms. This rotation of the octahedron is notobserved in the average structure, because it is not periodic
and is due to local lattice distortion. A system with large x andmoderate volume-contraction with increase in T exhibits large
rotation of the Mn-octahedron, resulting in large splitting ofpeak (2), as shown by arrows. From these studies, we foundthat the local lattice distortion strongly correlates with themoderation of MVE, i.e., the Invar effect.

Reference
Iikubo, S., Kodama, K. et al., Magnetovolume Effect in Mn3Cu1-xGexN Related to the Magnetic Structure: Neutron Powder Diffraction Measurements,
Physical Review B, vol.77, no.2, 2008, p.020409-1-020409-4.
JAEA R&D Review 200849

Quantum Beam Science Research

4-7

Observation of Electron Dynamics Utilizing Synchrotron
Radiation X-ray–Charge Excitations in Ladder Cuprates by Resonant Inelastic X-ray Scattering–

3
2

5

1
0.8
0.6
0.4
0.2
0

4
3
2

（h）
Intensity（a.u.）

Momentum

Energy
（eV）
q=
（0,0）
q=
（0,）
q=
（ ,0）
q=
（ , ）

1

（i）

Momentum

5

1
0.8
0.6
0.4
0.2
0

4
3
2
1

Momentum（j）
Intensity（a.u.）（d）

Energy（eV）（c）

Energy
（eV）

Cu O

q=
（0,0）
q=
（0,）
q=
（ ,0）
q=
（ , ）

1

（g）

Cu O

Intensity（a.u.）

4

Intensity（a.u.）

La/Sr/Ca

Nd/Ce

1
0.8
0.6
0.4
0.2
0

Intensity（a.u.）

（b）

（f）

5

Intensity（a.u.）（a）Energy（eV）（e）

Energy
（eV）
q=
（0,0）
q=
（0,）
q=
（ ,0）
q=
（ , ）

Energy（eV）

Fig.4-17 Crystal structure of two representative copperoxides and their relevant Cu-O components(a)Nd2-xCexCuO4 (b)(La,Sr,Ca)14Cu24O41 (c)Two-dimensionalsquare lattice (d)Two-leg ladderFig.4-18 Resonant inelastic x-ray scattering spectra of
La5Sr9Cu24O41 [(e), (f)], Sr14Cu24O41 [ (g), (h)], Sr2.5Ca11.5Cu24O41[(i), (j)]Momentum q = (qrung, qleg) is reduced in the Brillouin zoneof the ladder. Elastic scattering and high-energy excitation
above 5 eV are removed in (e), (g), (i). The right column
gives representative spectra corresponding to the contour
plots in the left column.

One of the important roles of materials science is tounderstand electronic properties of a material from theinvestigation of its electronic structure, that is, energy andmomentum of the electrons. For this purpose, the photon is agood probe because it is mainly scattered by the charge of theelectron. Brilliant x-rays from a large synchrotron radiation
facility such as SPring-8 now enable us to measure electronic
excitations by so-called inelastic x-ray scattering (IXS). A
great advantage of IXS over conventional optical methods is
momentum resolution. Furthermore, resonant inelastic x-ray
scattering (RIXS), where the incident photon energy is tuned
to the electron level of a specific element, has an additional
merit of element selectivity.
A main target of our research is strongly correlated copperoxides such as high-Tc superconductors. Fig.4-17(a) shows a
representative superconducting copper oxide, Nd2-xCexCuO4,
and the playground of the superconductivity is the twodimensional square lattice shown in Fig.4-17(c). We havereported on Nd2-xCexCuO4 two years ago. Here we present a
RIXS study of another copper oxide system, (La,Sr,Ca)14Cu24O41,
whose crystal structure is shown in Fig.4-17(b). The relevant
component of the system is the two-leg ladder in Fig.4-17(d).
Superconductivity was also found in the system under highpressure and it is considered to belong to the second class ofsuperconducting copper oxides which has a different structuralcomponent.
Fig.4-18 shows RIXS spectra of three concentrations of(La,Sr,Ca)14Cu24O41. The carrier number increases from the
top to bottom graphs. The dispersive feature at 2-4 eV is an
excitation between the Cu 3d orbital split by the strong Coulomb
repulsion and the O 2p orbital. Energy-momentum dependenceof the excitation is almost independent of the carrier number,which is a characteristic of the two-leg ladder. In contrast, it
is dependent on the carrier number in the square lattice. The
continuum excitation at 1-1.5 eV is related to the dynamics ofthe mobile carriers in the ladder and its intensity is proportional
to the carrier number in the ladder.
These detailed studies give fundamental information of theelectronic structure which is necessary to construct a model ofthe superconductivity.
In both the square lattice and two-leg ladder, a chargeordered state where mobile carriers stop their motion
and align periodically was found at some concentrations.
Whether the charge order is cooperative or competitive withsuperconductivity is an important issue. The next aim of our
RIXS studies is to detect the electron dynamics in the chargeordered state.

Reference
Ishii, K. et al., Momentum Dependent Charge Excitations of Two-leg Ladder: Resonant Inelastic X-ray Scattering of (La,Sr,Ca)14Cu24O41, Physical Review B,
vol.76, no.4, 2007, p.045124-1-045124-7.

50

JAEA R&D Review 2008

Quantum Beam Science Research

4-8

Observation of Curious Long-Period Structures in
Metal Hydride–A Clue to Understand the Hydrogen-Metal Interactions and Their Role–

hexagonal

Intensity (a.u.)

experimentintermediate

cubic

simulation

Rietveld refinement°
Fig.4-19 X-ray diffraction patterns of YH3(a) X-ray diffraction pattern of YH3 measured at 14.0 GPa.
(b) Diffraction pattern simulated with a 27-layered longperiod structural model.
(c) Result of pattern fitting by Rietveld method.

Fig.4-20 Pressure variation of metal lattices of YH3
From right, low-pressure hexagonal structure, long-period
stacking structures at 14.0 and 17.9 GPa, and high-pressure
cubic structure.

Hydrogen (H) is the smallest, lightest and chemically active
element. Hydrogen reacts with almost all metals and form
hydrides, in which hydrogen atoms occupy the interstices ofmetal lattices. The interstitial H atoms form chemical bonds
such as ionic or covalent-like bonds with the surrounding metal
atoms and change the lattice structure and electronic state.
Crystal structure and electronic state are changed by applying
pressure as well. We have investigated the structure of yttrium
trihydride (YH3) under high pressure. YH3 is an insulator, but
was found to become a metal again upon compression beyond
20 GPa. The high-pressure structural study of YH3 would
provide a clue to elucidate the insulator-metal transition interms of the metal-H and/or H-H interactions.
The process of the pressure-induced hexagonal-cubictransition of YH3 was investigated by synchrotron radiation
X-ray diffraction experiments at “SPring-8”. The hexagonal
structure gradually transformed into a cubic structure in awide pressure span from 11 to 20 GPa, while the electronic
state changed from insulating to metal. The intermediate state
while in the above pressure range has been interpreted as thecoexistence of the low-pressure hexagonal and high-pressure
cubic phases. However, we have found that the intermediate
state was a single phase and their diffraction patterns were well
reproduced by models with long-period stacking structures ofY metal layers. Fig.4-19 shows (a) a measured high-resolution
diffraction pattern, (b) a pattern simulated with a long-period
model, and (c) the result of profile fitting analysis. The longperiod structures consisted of Y-metal layers periodicallyarranged in ABA-type (H-type) and ABC-type (K-type) stacking
sequence. The 14.0-GPa structure of a unit cell containing
27 layers was characterized as consisting of three sub unit
cells, each having a 4-H and 5-K layer sequence, which werestacked up along the 3-fold axis. The17.9-GPa structure wasdescribed with the 27 layers as well but here the sub unit cells
consisted of a 2-H and 7-K layer sequence (Fig.4-20). The
K-type component increased in number gradually withincreasing pressure, and eventually became the only type. Suchstructural transition does not occur in Y metal containing no H
atoms. The interstitial H atoms likely played a dominant role in
the formation of the long-period structures.

Reference
Machida, A. et al., Long-period Stacking Structures in Yttrium Trihydride at High Pressure, Physical Review B, vol.76, no.5, 2007, p.052101-1-052101-4.
JAEA R&D Review 200851

Quantum Beam Science Research

4-9

A Breakthrough in Development of High-Temperature
Polymer Electrolyte Membrane Fuel Cells–Successfully Making a Highly Heat-Resistant Proton-Conductive
Aromatic Polymer Electrolyte Membrane–

BranchScaffolding
molecular chains

Trunk

Super-engineering plastic
macromolecular chains

FlowerProton-conductive
molecular chains

High performancefuel cell membrane（structural model）

Cell voltage㧔V㧕

Temperature: 95℃
Relative humidity: 40％

New membrane
Nafion membrane

Run time㧔h㧕Fig.4-21 Schematic diagram for the development of highperformance fuel cell membranes by radiation grafting
methodScaffolding molecular chains (branches) were grafted to ahighly heat-resistant super engineering plastic macromolecular
chain (trunk), followed by radiation grafting of the functional
molecular chains (flower), so that a number of proton
conductive molecular chains were introduced into themembrane.

Fig.4-22 Fuel cell performance of new polymer electrolyte
membranes operated under high temperature and low
humidification
Operating the fuel cell under high temperature and low
humidification, the prepared radiation-grafted polymer
electrolyte membranes exhibited fuel cell performance better
than that of commercial Nafion membrane.

The polymer electrolyte membrane fuel cells arepromising electric power sources for home cogeneration andautomobiles. To enhance the fuel cell performance and catalyst
utilization efficiency, and to simplify the cooling system, it
is desired that the fuel cell be operable at high temperatureand low humification. However, although the commercial
perfluorinated polymer electrolyte membranes, such as Nafion
membranes, have excellent performance below 80°C andwith full humidification, performance deteriorates remarkably
above 90°C or at relative humidity below 50%. Therefore,
development of a new polymer electrolyte membrane that canbe used for thousands of hours under high temperatures andlow humidification has been widely pursued recently.
To meet the demand for high temperature fuel cells, we used athermally stable engineering aromatic polymer as the base film
to develop a polymer electrolyte fuel cell membrane. In thisstudy, the typical super-engineering film polyetheretherketone(PEEK) was chosen as the base film, and a styrenesulfonic ethyl
ester (ETSS) monomer was radiation-grafted to this. The ETSSgrafted PEEK film was hydrolyzed to obtain the desired highperformance polymer electrolyte membrane. By means of thisnew method, the PEEK-based polymer electrolyte membrane
was directly prepared without a severe sulfonation reaction
using a strong acid. Consequently, the intrinsic properties of
PEEK were maintained and the resulting polymer electrolyte
membrane was very tough, suited for application in fuel cells.
However, because of the low grafting activity of the PEEK
chains (trunks), the ETSS monomer (flowers) was difficult to
attach to the PEEK trunks in the desired amounts. Therefore,
we designed a branch which could be easily attached to thetrunks and easily graft with ETSS flowers. We found that the
divinylbenzene (DVB) can serve as this branch well, allowing
many flowers to be radiation grafted to the PEEK trunks. As aresult, new polymer electrolyte membranes with high proton
conductivity and highly thermal durability were successfully
developed (Fig.4-21).
The new polymer electrolyte membrane was tested ina fuel cell under high temperature and low humidity (celltemperature, 95°C; 40% relative humidity). For comparison,
the Nafion membrane was also tested under the same
conditions. It was confirmed that the new membranes exhibited
higher cell performance than the Nafion membrane. The
membrane durability was evaluated in a fuel cell operated ata fixed current density of 0.3 A/cm2. The Nafion membranecannot withstand such a high temperature and low humidity,
its voltage drastically dropping in a short operation time dueto the deterioration-induced crossover of the fuel across themembrane. In contrast, the developed new polymer electrolyte
membranes had significantly higher durability and stability.
The voltage of the fuel cell was almost unchanged after the
initial 250 h of operation, and was estimated to be stable formore than thousands of hours under this high temperature andlow relative humidity (Fig.4-22).
In conclusion, it was possible to develop a polymer electrolyte
membrane for a fuel cell which has high performance underhigh temperature and low humidification, by radiation grafting
techniques. Development of such a highly stable and durable
polymer electrolyte membrane is a big breakthrough for thepractical application of fuel cells using hydrogen as the fuel.

Reference
Chen, J. et al., Fuel Cell Performance of Polyetheretherketone-based Polymer Electrolyte Membranes Prepared by a Two-step Grafting Method, Jounal of
Membrane Science, vol.319, issues 1-2, 2008, p.1-4.

52

JAEA R&D Review 2008

Quantum Beam Science Research

4-10 Strain Measurement on Rebars in Reinforced Concrete–Application of Neutron Diffraction to Nondestructive Strain
Measurement–

Peinforced Concrete
SampleTensile stressron

Peak shift

Neut

Atom
Diffraction

d0

d0

d

Crystal structure

Detector（Stress-free）

Fig.4-23 Principle of neutron strain measurement
Lattice strain can be measured by observing change in diffraction
angle, 2θ, which is due to change in lattice spacing.

Fig.4-25 Layout of neutron strain measurement
systemNeutrons irradiate the rebar through the concrete,
and then the diffractions from the rebar aredetected by the neutron detector.

Stress (MPa)Applied stress 309MPs
Applied stress 155MPs
stress-free

Fig.4-24 Stress distributions on the rebar
Stress distributions in axial direction (loading direction) on
a rebar with diameter of 16mm embedded in concrete ofdimensions 50mm × 50mm × 800mm.

Position (mm)It is already known that the performance of a reinforced
concrete structure is affected by the bond condition between therebar and concrete, and this bond condition has been evaluatedby evaluating the strain on the rebar. In general, strains on
the rebar have been measured using strain gauges; however,waterproofing treatment and wiring of the strain gauges affectthe bond condition, and strain gauges discretely set on therebar cannot measure strain distributions in detail. On the other
hand, the neutron diffraction method, which is nondestructive
and non-contact, can measure the lattice spacing of crystalline
materials (Fig.4-23), and residual stresses and strains insidematerials at a depth on the order of centimeters can be evaluatedfrom the change in lattice spacing with the spatial resolution on
the order of a few millimeters. Although the water in hydrated
cement and the evaporated water in the concrete structures
would decrease the neutron intensity due to high absorption
by hydrogen atoms, drying the water in the concrete makes it
possible to measure the strains on the rebar in concrete with thesize of 50mm to 100mm square.
Fig.4-24 shows the lattice stress distribution on a rebar inconcrete, derived from measurements by the neutron diffraction
method. The residual stress distributions in this figure were
measured along the rebar under uniaxial tensile stress of 155
MPa and 309 MPa, as well as the stress-free condition, as
shown in Fig.4-25. Stress transferred to the rebar increased
with an increase in the applied stress, though it was less than
the applied stress because stress was shared with the concrete.
Moreover, dispersion of stress around cracks was observed due
to loss of adhesion there and also due to stress concentration
around rebars. These results clearly indicate that the neutron
diffraction method enables detailed measurement of stress
distribution on the rebar.
As indicated above, the neutron diffraction method is
expected to make possible measurement of the detailed residual
stresses on the rebar nondestructively and also to measure thestress distributions around cracks in the concrete which cannot
be measured by strain gauges. There is a possibility that theneutron diffraction method will gives us new knowledge in the
structural engineering field.

Reference
Kanematsu, M., Noguchi, T., Yasuda, M., Suzuki H., Nondestructive Stress Measurement of Rebar using Neutron Diffractometer “RESA”, Proceedings ofAnnual Meeting of Japan Concrete Institute, Fukuoka, Japan, 2008, p.775-780, in CD-ROM (in Japanese).
JAEA R&D Review 200853

Quantum Beam Science Research

4-11 Compact and Low Noise Ultra-High Intensity Laser

Temporal profile of high intensity laser pulses

intensitylaser pulse

noise leveltime
starting time of light-matter interaction

normalized intensity (a.u.)–Good Prospects for a Laser Driven Proton Accelerator for
Cancer Therapy–

Laser
pulse
noise level ofconventionallasersnoise level of laser
developed in this study

time (10-12s)

Fig.4-26 Temporal contrast of laser pulse
The output laser pulse has a noise component. The temporal contrast is defined to be the ratio of the intensityof the noise to the peak intensity of the laser pulse and is plotted as “normalized intensity”. The contrast is
improved by over 2 orders in magnitude with our laser system (red curve) compared to conventional systems(blue curve).

Pump beam

Amplified signal
Depleted pump

Signal beam

Fig.4-27 Conceptual diagram of Optical Parametric ChirpedPulse Amplification (OPCPA)The OPCPA scheme amplifies a frequency stretched “signal”
pulse when a pump beam and a signal beam are present in anonlinear crystal.

Nonlinear crystal

Ti:sapphire crystal is widely used as the lasing medium of
choice for high intensity lasers. However, this material also
causes critical technical problems such as unwieldy equipment
size and unacceptably large noise levels which means low
temporal contrast (Fig.4-26). Here, the contrast is timedependent and defined as the ratio of the intensity of the noise
to the peak intensity of the main pulse. Typically, terawatt(TW, 1012 W) Ti:sapphire laser amplifiers themselves requirea space of a few meters square and have a contrast of ~10-6.
Therefore, at high intensity, the noise level can exceed
the ionization threshold of a target, which can destroy orsignificantly modify the laser-plasma interaction processes. For
the laser driven accelerator, the accelerated energy is limited by
the noise level.
In order to solve these problems, we have improved theoptical parametric chirped-pulse amplification (OPCPA)
technique (Fig.4-27). The OPCPA affords significant reductionin the noise level compared with conventional systems because

the amplification occurs only within the time of the pump pulse.
For shorter pump pulse durations, the amplification time widthis shorter, minimizing the noise amplification, and allowing
high contrast. For the OPCPA we have (1) improved the pumpto-signal efficiency by increasing the spatial overlap between
pump and signal beams and (2) added more precise control ofamplification gain and spectrum by the use of multiple stages.
By using these techniques, it is possible to increase the contrast
and reduce the size of the high intensity laser.
This improved OPCPA laser facilitates a decrease in opticalcomponent size by at least an order of magnitude and at least
a two order of magnitude improvement in contrast. With thiscompact system, a peak power of 3 TW was realized. Thisis the world’s highest power compact OPCPA laser. These
improvements in the high intensity laser technology are
expected to make possible medical applications that use laser
driven accelerators for therapy, and industrial technology.

Reference
Kiriyama, H. et al., High-energy, High-contrast, Multiterawatt Laser Pulses by Optical Parametric Chirped-pulse Amplification, Optics Letters, vol.32, no.16,
2007, p.2315-2317.

54

JAEA R&D Review 2008

Quantum Beam Science Research

4-12 Mechanism for Biological Sensing of Ionizing Radiation–The Role of γ Subunit of G Protein in Caenorhabditis elegans–

Avoidance

Salt chemotaxis learning
NaCl

Attraction
NaCl

NaCl

C.elegansfood

Ionizing
radiation（Starvation）

Strong avoidance
NaCl

Fig.4-28 Salt chemotaxis learning in C. elegans and radiation effects
C. elegans is normally attracted to NaCl (salt), but the presentation of NaCl in the absence offood decreases its attraction to NaCl. The behavioral change from attraction to avoidance is
the “salt chemotaxis learning”, which is enhanced by ionizing irradiation.

Salt chemotaxis Iearning

Response to NaCl

Attraction

NaCI

gpc-1 mutant

Enhancement

Enhance
Wild-type

Suppress

Ionizing
radiation

Starvation

Singnal transduction in
the nervous system

GPC-1
Sensory
neurons

Avoidance
Dose
（Gy）

Avoidance of NaCI

Immediately after the discovery of ionizing radiation (IR),it was employed for medical uses as a roentgenography since
it has permeability. At present, IR is an indispensable tool fora medical treatment, and has other uses. On the other hand,the biological effects of IR on humans have been studied andaccumulated. Recently, effects of IR on the nervous system
are drawing attention in the medical and space research fields.
An important issue for medicine is the IR-induced suppressionof adult neurogenesis in the hippocampus, where neurons arecontinuously born throughout life and which is an integral
site for learning and memory. In space science, acute IRinduced learning impairment and abnormalities in the nervous
system are seen as important potential risks associated withinterplanetary space missions.
The research background made us to examine the effects of
IR on learning using the nematode Caenorhabditis elegans.
C. elegans, which is used for study of Alzheimer disease, is
an attractive multi-cellular model organism for neurosciences

Fig.4-29 Radiation responses andthe mechanism
Acute exposure to ionizing radiationinduced an additional decrease in
attraction to salt in the transition stageof the salt chemotaxis learning. The
ionizing radiation induced additional
decrease in attraction is significantly
suppressed in the gpc-1 mutant, which
was defective in GPC-1 (one of thetwo γ subunits of the heterotrimeric
G-protein). This suggests that ionizing
radiation behaves as a modulator in
the salt chemotaxis learning via C.
elegans GPC-1.

because the connections in its neural network are completely
known and has many genes similar to human genes. Fromexperiments with salt chemotaxis learning, we found that onlyin the transition stage of learning, exposure to an acute dose of
IR immediately induced an additional decrease (avoidance) of
salt-attraction (Fig.4-28). To investigate the cause of radiation
effects, we used a mutant with a deletion of gene. Finally,
we found that IR-induced additional avoidance of NaCl was
significantly suppressed in the gpc-1 mutant (Fig.4-29), which
was defective in GPC-1 (one of the two γ subunits of theheterotrimeric G-protein). It suggests that IR modulates thesignal transduction in the nervous system of C. elegans via
specific sensory neurons, and so qualitatively and quantitatively
alters the salt chemotaxis learning.
In future, we will study the molecular mechanism of IR
sensing in the sensory neurons and the effect of signals inducedby IR in the nervous system of C. elegans.

Reference
Sakashita, T. et al., Modulatory Effect of Ionizing Radiation on Food-NaCl Associative Learning: the Role of γ Subunit of G Protein in Caenorhabditis
Elegans, FASEB Journal, vol.22, no.3, 2008, p.713-720.
JAEA R&D Review 200855

5 Nuclear Safety Research
To Support Safety Regulations and to Ensure Nuclear
Safety and Confidence
Research programs conducted inaccordance with the policy determinedby the Japanese Nuclear Safety Commission
・Probabilistic safety assessment (including
study on emergency preparedness andanalysis of operating experience)・Safety of high burnup fuel・Thermohydraulic safety of advanced
utilizations of LWRs
・Safety assessment of plant aging
・Safety of nuclear fuel cycle facilities
・Safety of waste disposal and decommissioning

Provide technical datafor regulatory judgment

・Ensure safety
・Increase public
Confidence

Fig.5-1 Major Subjects and Tasks of Safety Research

The latest scientific and technical knowledge is essential forthe development and improvement of the safety guidelines andcriteria with which the regulatory authorities perform safety
regulation to assure the safety of nuclear installations.
The Nuclear Safety Commission (NSC) proposed a“Prioritized Plan for Nuclear Safety Research” in July 2004, tobe carried out in order to meet the future regulatory needs. The
main tasks expected of us are shown in Fig.5-1.
The results of nuclear safety research contribute to themaintenance and improvement of safety of the nuclear facilities
and also to fostering public confidence in nuclear safety.
Furthermore, we are supporting the regulatory body Nuclear
and Industrial Safety Agency (NISA) and the Japan Nuclear
Safety Organization (JNES) by conducting research on various
technical issues in safety regulation which they have contracted
us to do.
The programs specified in the NSC’s prioritized researchplan and these contracts with the regulatory bodies are steadily
providing a wide spectrum of useful results. The following
paragraphs briefly describe the current status of research andnew results from the programs shown in Fig.5-1.
The program on probabilistic safety assessment (PSA)
is developing PSA methodologies for nuclear fuel cycle
facilities and studying effective uses of risk information forsafety regulation and management. The research program on
emergency preparedness is applying the methodologies of PSA
and environmental consequence analysis to the development ofguides for decision making in various emergencies, including
56

JAEA R&D Review 2008

action levels, area, and timing of protective actions. Analysis
of operating experience aims to obtain lessons from incidents
that have occurred in nuclear facilities. An analysis of reported
events of primary water stress corrosion cracking (PWSCC)
provided information useful for early detection of PWSCC(Topic 5-1).
For the study of high burnup fuel safety, the FEMAXI-6
code for simulation of fuel behavior under normal operation
was improved for application to MOX fuel and was tested bycomparison with an experiment (Topic 5-2). A test at the nuclear
safety research reactor (NSRR) simulating a reactivity initiatedaccident (RIA) clarified the fission gas release behavior in fuelpellets of high-burnup fuel (Topic 5-3)
In the study of safety assessment on advanced uses of lightwater reactors, tests at the Large Scale Transient test Facility(LSTF) under the international collaboration with the OECD/
NEA are in progress and have produced test data useful for
diagnosis of accident conditions and validation of analysis
codes (Topic 5-4). A new measurement system for rapid boilingbehavior in coolant channels in fuel bundle was developed
and obtained test data for validation of RIA analysis codes(Topic 5-5).
For research on structural integrity assessment of reactor
components, data on the neutron irradiation embrittlement
mechanism such as intergranular embrittlement were obtained.
The data suggested that the effect of phosphorus (P) impurities
is not significant even with high values of neutron fluence.
Moreover, we proposed a modified “Master Curve” method
for fracture toughness evaluation of reactor pressure vessel
steel, that enables a conservative fracture toughness estimate
considering intergranular embrittlement (Topic 5-6).
For research on safety evaluation of nuclear fuel cycle
facilities, we performed tests of the degree to which fission
products in spent fuel reduce reactivity (burn up credit), andusing this data examined the accuracy of the computation
analysis used in criticality safety assessment of spent fuelstorage facilities (Topic 5-7).
In the field of radioactive waste disposal and facility
decommissioning safety research, we developed procedures
to derive the upper bound of radioactivity concentration
for three disposal methods of transuranium (TRU) wastes:
trench disposal, concrete vault disposal and intermediate
depth disposal. Calculation results of these procedures wereincorporated in a report by the nuclear safety commission andin the ordinance of Ministry of Economy, Trade and Industry in2008 that prescribed requirements on disposal of TRU wastes(Topic 5-8).
For the safety assessment of geological disposal of high-level
radioactive waste (HLW), we developed a computer code forprediction of long-term performance alteration of compacted
bentonite that covers the metal container encapsulating the
HLW (Topic 5-9).

Nuclear Safety Research

5-1

To Learn from Operating Experience–Analysis of Primary Water Stress Corrosion Cracking Events at PWRs–

Level Instrument
Nozzle (1)Pressurizer
CRDM Nozzle (20)
Vent Pipe (1)SteamGenerator
Heater Bundle
Diaphragm(1)Thermocouple
Nozzle (5)Hot Leg
Instrument
Nozzle (7)Heater Sleeve(12)
Reactor
Hot Leg Drain
Pressure Hot Leg Nozzle (1) Pipe (1) SG Drain
Nozzle (3)VesselBottom Mounted
Instrument Nozzle (1)Note) The number in parentheses means thenumber of applicable LERs. There are some
LERs that include degradations of two or
more RCPB components.

It is worldwide recognized that it is important to learn fromoperating experience in all technologies. In the field of nuclear
technology, nowadays, regulatory authorities and industries
have been actively carrying out analysis of incidents to identify
their causes and to feed back the lessons learned to the design,operation and maintenance of installations on an international
basis.
Since a severe accident at a nuclear installation, in particulara nuclear power plant, could cause serious damage to theenvironment as well as the surrounding populace, it is generally
considered essential to eliminate potential or latent causes that
could lead to such an accident in advance. Therefore, it isnecessary to collect and analyze event information to consider
corrective actions for preventing recurrence of the event.
Further, it is important to continuously conduct such activities.
At JAEA, such activities have been carried out for some time,
and the results from these activities have been provided to theregulatory authorities and utilities in Japan.
Fig.5-2 illustrates an example of results from analysis ofrecent events which occurred at nuclear power plants. This
analysis was carried out on the events involving primary water
stress corrosion cracking (PWSCC) observed in the reactor

Fig.5-2 Summary of Reactor Coolant Pressure
Boundary Components with PWSCC Observed
Although primary water stress corrosion cracking(PWSCC) has been observed since early 1990’s,
it was not then recognized as an urgent issue
for reactor safety. However, recently several
significant degradations originating from PWSCC
in the reactor coolant pressure boundary (RCPB)
components have been observed at U.S. PWR
plants. We analyzed the U.S. experience withalloy 600 degradation by reviewing licensee event
reports (LERs) from 1999 to 2005 to examine
trends of such events mainly focusing on affectedcomponents, characteristics of cracking, andinspection approaches for detecting the PWSCC.
This figure provides a summary of locations
where PWSCC was observed. As shown in thisfigure, PWSCC was found to have occurredin the RCPB components exposed to the hightemperature environment, such as the control
rod drive mechanism (CRDM) nozzles on reactor
vessel head, and thus high temperature conditionis one of the factors causing high susceptibility
to PWSCC.

coolant pressure boundary components such as the reactor
pressure vessel (RPV) and reactor cooling system (RCS)
piping.
Should the RPV or RCS piping break, resulting in a loss-ofcoolant-accident (LOCA), the capabilities of core cooling andconfinement of radioactive materials would be degraded. To
obtain insights useful for preventing such a serious condition
from occurring, we analyzed PWSCC events focusing on where
PWSCC was observed, a method which could detect PWSCC,
and so on. The results from this analysis revealed that PWSCC
tends to occur at alloy 600 components exposed to relatively
high temperature, such as control rod drive mechanism(CRDM) nozzles and pressurizer heater sleeves. In addition,
it was shown that depending on the component affected, thenon-destructive examination such as ultrasonic testing and/or eddy current testing is generally needed to detect and/or
confirm the PWSCC , as well as visual inspection and different
repair techniques should be applied. These results have been
published to provide the relevant regulatory authorities andindustries in Japan with the lessons learned and insights gainedthat seem useful for the early detection of PWSCC at nuclear
power plants.

Reference
Takahara, S., Watanabe, N., Trending Analysis of Incidents Involving Primary Water Stress Corrosion Cracking on Alloy 600 Components at U.S. PWRs,
Nippon Genshiryoku Gakkai Wabun Ronbunshi, vol.5, no.4, 2006, p.282-291 (in Japanese).
JAEA R&D Review 200857

Nuclear Safety Research

5-2

Simulation of MOX Fuel Irradiation Behavior for LWRs

BurnupEnhanceEnhance

Cladding

Solid FP
accumulationin pellet
Enhance?

Pelletswelling
MOX fuelpelletCoolant

FP gasaccumulationin pellet

Possibly Enhanceenhances
FP gas release
into P/C gap

Suppress
EnhanceEnhance

Pellet temperature
increase(b)

Rod internal pressure (MPa)(a)
Gap between
pellet and cladding(P/C gap)

Pellet center temperature (K)–Analysis of MOX Fuel Behavior in Halden Reactor by FEMAXI-6
Code–

1600
Conv.
model

1400

GS model1200
1000
Measured(a)800

Conv. model3

Steep pressure
increase

Measured2

GS model

1

(b)
0

5

10

15

20

25

30

35

MOX fuel rod burnup (GWd/tUO2)

Fig.5-3 Burnup-induced phenomena in a fuel rod
FEMAXI-6 simulates behavior of one fuel rod. Fuel burnup
induces pellet swelling, fission product (FP) gas accumulation,and its release to P/C gap. These phenomena interact witheach other as shown above. This indicates that FP gas at
the grain boundary can promote pellet swelling, which isusually induced by solid FP accumulation.

Fig.5-4 Comparison
betweenmeasurements
andcalculations(a) Calculated pellet temperature agrees with the measurement
especially in the GS model, supporting the gas-induced
swelling theory. (b) Calculation cannot reproduce the steeppressure increase satisfactorily. More accurate treatment of
FP gas behavior is needed.

In nuclear reactors, energy is produced via fission
reactions of Uranium (U) or Plutonium (Pu) in fuel pellets.
This is the so-called “burnup” process. Because a variety ofphenomena take place along with burnup, and these interact
in a complicated manner as shown in Fig.5-3, computer
simulation is indispensable to predict such fuel behavior. We
have developed a high burnup fuel behavior analysis code,
FEMAXI-6, in order to understand the behavior of high burnup
fuels for reliable safety evaluation. As extensive use of U-Pu
Mixed Oxide (MOX) fuels in LWRs is planned now, we have
improved FEMAXI-6 to achieve more accurate prediction ofhigh burnup MOX fuel behavior.
Some thermal and mechanical property models for
MOX pellets were implemented to properly calculate pellet
temperature, which is the most important factor influencing
total fuel behavior. The next factor to be treated carefully is
the effect of fission product (FP) gas which is accumulated
inside the pellet grain or at the grain boundary. The latter,intergranular gas bubble growth, can enhance pellet swelling,
and for this a new gas-induced swelling model (GS model) was
implemented. With the improved version of FEMAXI-6, thebehaviors of the MOX fuels irradiated in the Halden (Norway)
reactor were analyzed.
Fig.5-4(a) shows changes in pellet center temperatureduring irradiation. Calculations successfully reproduced themeasurement, especially in the GS model calculations. Aspellet swelling is enhanced in the GS model, pellets moreeasily contact the cladding than in the conventional model(Conv. model) which does not consider the gas effect. This P/C
contact enhances heat transfer and lowers pellet temperature.
The agreement between the GS model and measurement
supports the gas-induced swelling theory.
Fig.5-4(b) shows changes in the fuel-rod internal pressure.
The steep increase indicates an appreciable FP gas release,which also implies that there was a large accumulation of FP
gas at the pellet grain boundary before the sudden release.
The deviation of the calculation results from the measurement
is, therefore, mainly attributable to underestimation ofintergranular FP gas inventory. As such deviation could affectprediction of pellet temperature in higher burnup cases, moreaccurate treatment of FP gas behavior is needed.
The present improvement enabled more accurate predictionof MOX fuel pellet temperature. A new model to calculate grainboundary FP gas inventory is under development. FEMAXI-6
is an open code and is extensively used for research and safety
evaluation. The next release is planned at the end of FY2008(to be shown at http://www.rist.or.jp/nucis/).

Reference
Udagawa, Y. et al., Analysis of MOX Fuel Behavior in Halden Reactor by FEMAXI-6 Code, Journal of Nuclear Science and Technology, vol.44, no.8, 2007,
p.1070-1080.

58

JAEA R&D Review 2008

Nuclear Safety Research

5-3

Does Pellet Microstructure Change Affect Fuel
Behavior during a RIA ?
–Fission Gas Release in High Burnup BWR Fuel under RIA Conditions–

Microstructure at the pellet periphery

Cladding
50Ǵm

1mm

A cross section of fuel rod

Fuel rod

2.0

1.5

2.0
Before power burst test
After power burst test

1.5

Rim structure area
1.0

1.0

0.5

0.5

Relative burnup(-)

Relative concentration of Xe(-)

Pellet

0.0
0.0
0.0
0.2
0.4
0.6
0.8
1.0
(Center)(Periphery)
Relative distance from center, r/r0(-)

Fig.5-5 Crystal structure of fuel pellet before power
burst test
A cross section of high burnup fuel pellet was observedby an optical microscope and scanning electron microscope(SEM). Fine fission gas bubbles of high density precipitatedat the pellet periphery than in the center and middle, andthe fabricated microstructure was lost in the periphery.

Fig.5-6 Comparison of radial Xe concentration profiles
Pulse operation of the Nuclear Safety Research Reactor(NSRR) can simulate safely a Reactivity Initiated Accident(RIA) which is envisioned in light water reactors. Comparing
Xe concentration profiles which were measured before andafter a power burst test, decrease of Xe concentration isseen in the rim structure, but this was negligible. Since
the Xe concentration substantially decreased in the region
between the center and mid-radius of pellet, it is consideredthat most fission gas was released from this region.

Extending the utilization period of reactor fuel, i.e. burnup
extension, is progressing in stages, for the efficient use ofnatural resources and the reduction of fuel cycle cost. With
increase in fuel burnup, more fission products accumulate in
the fuel pellet, and cladding corrosion worsens. Accordingly,
fuel safety with high burnup must be confirmed under accident
conditions as well as under normal operation conditions.
At the periphery of a high burnup pellet, the microstructure
is different from that at the time of pellet fabrication (Fig.5-5).
This microstructure is called rim structure, and high pressure
fission gas is accumulated in the bubbles generated in this region.
In the case that the accumulated fission gas is released during areactivity initiated accident (RIA), pellet temperature increase
and the cladding deformation which is caused by additional
fission gas release would occur due to the degradation of theheat conduction between pellet and cladding. These phenomena
may affect the fuel rod safety. Consequently, the investigationof the effect of rim structure formation on fission gas release
behavior under an RIA is an important subject of study.
A test rod was prepared from a fuel rod with high burnup after
use in a commercial reactor, and power burst tests simulating
a RIA were carried out at the NSRR with the test rod. In oneof the post irradiation examinations (PIEs), the amount andcomposition of the fission gas released during the power burst
test were investigated. Fig.5-6 shows the radial concentration
profile of fission gas (Xe) which was measured by electronprobe micro analysis (EPMA). Comparing the data obtained
after power burst test with those before power burst test, it isseen that the Xe concentration decreased in the region between
the center and mid radius region, while it hardly changed at
all at the pellet periphery where the rim structure formed. The
amount of the fission gas released during the power burst test
can be evaluated from the difference of the Xe concentration
before and after the test, shown in Fig.5-6, and this evaluatedvalue is comparable with the value obtained from the PIE.
In consideration of the grain boundary separation which wasobserved in the mid-radius region, it is found that most fission
gas was released from the grain boundary in the mid-radiusregion, and that fission gas in the rim structure was hardly
released at all during the power burst test.
The result of this study shows that the amount of fission gas
released during a RIA is barely affected at all by rim structureformation, important information for analyzing fuel behavior
during a RIA.

Reference
Amaya, M. et al., Fission Gas Release in BWR Fuel with a Burnup of 56GWd/t during Simulated Reactivity Initiated Accident (RIA) Condition, Journal of
Nuclear Science and Technology, vol.45, no.5, 2008, p.423-431.
JAEA R&D Review 200859

Nuclear Safety Research

5-4

Reactor Instruments to Prevent PWR Severe Accident

Depressurization byrelief valve (RV)
Pressurizer

Level
L meter

Normalized
Superheat (-)–LSTF Test to Estimate Usefulness of Core Exit Thermocouple (CET)–

0.4(a) Superheat of CET

0.2

1074 s,Start ofAM action

910 s,Start ofsuper-heat

0
700

900

Min.
1000

1100

1.8

Normalized Rod
Temperature (-)LT(b) Rod surface temperature
at several elevations

1.4

Gas

1.2

840 s,Start ofheat-up

0.8
700

Low pressure injection(LPI) (P<1.2MPa)
High pressure injection(HPI)

Fig.5-7 CET validation in a PWR top break LOCA test
LSTF is the largest PWR simulator in the world, with fullheight and 1/48 volume scaling. A top break test was
performed, motivated by an incident of significant wall thinningaround the control rod penetration nozzle at pressure vessel
head of Davis Besse reactor (see Topic 5-1). Capability of
the CET (“T” in Fig. 5-7) to detect core overheating in caseof an HPI failure was examined. The test revealed that the
CET capability was influenced by three-dimensional steam
flow from the core toward the CRGT inlets as well as theirlocation around the CRGT.

1500

1600

1215 s,Power limited
to 50%

P.5(Middle)

800

Tsat

P.3

900

1000

1100

1200

1300

1400

1500

1600

Time after Break (s)Fig.5-8 Maximum steam superheat detected by CET (a)was later and lower than superheat of rods in core (b)4.0Normalized Core Height (-)

Accumulator injection
system (AIS) (P<4.5MPa)

1400

P.7
P.9(Top)

1.0

L

1300

1.6

CRGT

PWR
vessel1200

Time after Break (s)Steamgenerator(SG)

Vessel top
break

800

Max.
Ave.

3.5
3.0
2.5

(Top)
P.9
t=1000 s
P.8
P.7
t=1100 s

P.6

2.0
1.5
1.0
0.5

t=1200 s

P.5(Middle)P.4

Rod bundle with CRGT (B20)
Rod bundle w/o CRGT (B15)

P.3
Locationof T/Cs
0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

Normalized Superheat (-)

Fig.5-9 Comparison of axial steam temperatures in two
rod bundles with and without CRGT at the core exit
Temperature profiles significantly changed with CRGT.

We conducted a Large Scale Test Facility (LSTF) test
simulating a pressurized water reactor (PWR) loss-of-coolant
accident (LOCA) caused by a small break at the vessel top
under assumption of the total failure of high pressure injection(HPI) system (Fig.5-7). This test was part of the OECD/NEA
ROSA Project which is an international cooperative study
on reactor safety and advanced code development, one goal
being the validation of a core exit thermocouple (CET) to
detect core overheating and give a cue to start steam generator
depressurization, an accident management (AM) action.
The CET detected superheated steam temperature, but thiswas 70 s later than the start of core superheating, and there
was a further delay of 160 s to detect the temperature atwhich AM action is started, 623K. Thus, a large temperature
difference was found between the CET and the hottest fuel rod(Fig.5-8). We found two reasons for this. One was the cooling
effect of structural metals on uprising steam. The second one,

specific for this test, was that steam flow convergence into thecontrol rod guide tube (CRGT) inlet (Fig.5-7) induced threedimensional steam flow in the core. In fact, steam temperatures
in upper part of a high power bundle without a CRGT (B15)
were significantly lower than those in B20 bundle with CRGT,
suggesting colder steam inflow from the periphery following
after the hot steam that went to a neighboring bundle with a
CRGT (Fig.5-9). The CET installation outside the CRGTs may
have caused further discrepancy from the hot core.
We showed that such CET response was possible in actual
PWR operation, and furthermore demonstrated the usefulnessof alternative indexes measured by water level meters (3 kinds;“L” in Fig.5-7).
OECD/NEA has just started study of CET effectiveness
in PWRs. (This paper contains findings of the OECD/ROSA
Project with its consent for publication.)

Reference
Suzuki, M. et al., Performance of Core Exit Thermocouple for PWR Accident Management Action in Vessel Top Break LOCA Simulation Experimentat OECD/NEA ROSA Project, Proceedings of 16th International Conference on Nuclear Engineering (ICONE16), Orlando, Florida, USA, 2008,
ICONE16-48754, 11p., in CD-ROM.

60

JAEA R&D Review 2008

Nuclear Safety Research

5-5Measurement of Rapid Boiling and Vapor Void
Behavior in Coolant among Fuel Rods(b)
Simulated
Heater Rods

Plate-type
ElectrodeCentralLine-type
Electrode

Void Fraction

Line-type
Electrodes
between Rods

Coil-type
Electrode

Top View
Coolant Water

Lateral View

Fig.5-10 Layout and measurement area of impedance
type void fraction meter
Measured impedance between inserted electrodes andsimulated fuel rods also serving as electrodes is converted
into void fraction. The void fractions of central area (C),inter-rod area (P) and near-wall area (W) are measured
by measuring the impedance between coil and central line
electrodes, between inter-rod line electrodes and rods, andbetween plate electrode and rods, respectively.

Recently, high burn-up fuels are being introduced in lightwater reactors (LWRs), making it possible to produce a larger
amount of energy and utilize resources efficiently. However, it
may cause the degradation of fuels due to the greater changeof material composition with higher burn-up. To ensure thereactor integrity in an accident situation, LWRs are designed
to satisfy regulatory criteria with a certain safety margin. Our
research aims at contributing to more realistic evaluation of
the safety margin of boiling water reactors (BWRs) during
reactivity initiated accidents (RIAs), in which nuclear reaction
is abruptly enhanced due to sudden extraction of control rods
resulting in a rapid increase of core power.
In the BWR core, fast neutrons created by the nuclear reaction
are decelerated by surrounding water to become thermal
neutrons which sustain the chain reaction. Thus, the mostfavorable condition for nuclear reactions is where the fuels are
fully surrounded by single-phase water. In the current safetyassessment of RIAs, the power increase of the core is evaluatedunder such a condition. But in reality, the water around the fuelsboils to generate vapor voids in the water. As the voids lower
the amount of water around the fuels, the nuclear reaction
becomes less active; the core power decreases by the feedbackof negative void-reactivity. Therefore, the safety margin canbe more realistically evaluated by taking the void-reactivity
feedback into account. However, because of difficulties in thevoid behavior measurement under such rapid power increase
conditions during RIAs, technical knowledge necessary for theevaluation of the void-reactivity feedback has been limited. We
thus developed a fast-response measurement technique of thevolume fraction of voids (void fraction), installing electrodes

Void Fraction(a)Void Fraction–Toward Improved Prediction of Power Increase during RIAs–
Upper region

Middle region

Lower region

Time (s)Fig.5-11 Time histories of void fraction in eachmeasurement area
Simulated fuel rods were heated during the period of theunshaded area. Abrupt void fraction increase was delayed in
the near-wall area. Location where the abrupt increase in void
fraction occurs first varies depending on the conditions.

in a test flow channel having four simulated fuel rods as shownin Fig.5-10 and conducting a series of experiments simulating
RIA conditions.
The void fraction meter utilizes the property of waterthat electrical impedance increases with the void fraction.
The electrical impedance measured between several pairsof electrodes is converted into a local void fraction in theflow channel. Since the geometry of the test flow channel iscomplex, several types of electrodes including coil, line andplate types were installed in various spatial arrangements, andthe combination of the electrodes was rapidly switched. The
void fraction meters were placed in three regions along the flow
direction (upper, middle and lower regions). In each region, thevoid fraction was measured in the central area (C), an inter-rod
area (P) and a near-wall area (W). The changes in void fraction
under conditions simulating the rapid heat generation of a RIA
with a heating duration of 0.55 s are shown in Fig.5-11 as an
example. It is clearly indicated that the variation of the void
fraction strongly depends on the location in the channel. We
conducted a number of experiments varying the heat generation
rate and the temperature and velocity of water. The measured
results form a valuable database to improve the prediction
accuracy of the void fraction under RIA conditions. We madeefforts further to improve the predictive capability of numericalanalysis codes using the database, aiming at technical supportof the national safety regulations.
This study was performed under a contract between Nuclear
and Industrial Safety Agency (NISA) of the Ministry of
Economy, Trade and Industry (METI) of Japan and JAEA.

Reference
Satou, A. et al., Study on Transient Void Behavior During Reactivity Initiated Accidents Under Low Pressure Condition－Development and Application of
Measurement Technique for Void Fraction in Bundle Geometry－, Journal of Power and Energy Systems, vol.1, no.2, 2007, p.154-165.
JAEA R&D Review 200861

Nuclear Safety Research

5-6

Estimating Fracture Resistance of Reactor Pressure Vessel Steels–Estimating Toughness against Combination of Cleavage andIntergranular Fracture–(MPa√m)Ͱn (Ͱn(1/(1-Pf))

Pf 㧦fracture probability
Modified MC
methodStandard MC
method for cleavage
fracture only

Low fracture toughness
due to grain-boundary
decohesion

Fracture toughness (KJc)

Intergranular fractureIntergranular fracture

Mean curve:Standard MC
Lower tolerance
bound: Standard MCMean curve:
Modified MC
Lower tolerance
bound: Modified MC

Test temp.㧦 㧙10͠Ͱn (KJc-20)

Test temperature (͠)

Fig.5-12 A Weibull plot of fracture toughness ofintergranularly brittle material
The distribution of fracture toughness (KJc) values indicating
grain boundary decohesion does not follow the linear
regression defined in the standard method for cleavage
fracture. A modified method is needed to estimate lower
KJc values.

Fig.5-13 Temperature dependence of KJc values for theintergranularly brittle material together with the MasterCurve
The KJc values are beyond the lower tolerance bound. The
modified method enabled us to estimate lower KJc values
due to intergranular fracture.

The reactor pressure vessel (RPV) is one of the mostimportant safety-related structural components, because it has
no redundancy and its failure is not considered in the designof the reactor. There is concern is that because the portionof the RPV surrounding the core region called “beltline”
is exposed to neutron irradiation, the RPV material will bedegraded (irradiation embrittlement). The effect is manifested
by an increase in yield strength, a decrease in ductility, anddegradation of toughness which are monitored by surveillance
tests during plant operation. In the technical evaluation of every
nuclear power plant whose operating period reaches 30 years,
the irradiation embrittlement upon 60-years operation at highneutron fluence is estimated to ensure the structural integrityof RPV. One of possible types of embrittlement by such highneutron fluence is intergranular embrittlement caused bygrain-boundary phosphorus (P) segregation. The P segregation
is promoted by neutron irradiation and the presence of P
weakens the cohesive strength of grain-boundaries, leading toembrittlement through intergranular fracture. How to estimate
fracture toughness for the materials exhibiting intergranularfracture is one of the issues.
For fracture toughness evaluation of RPV steels, extensive
efforts have been made to achieve direct determination of
fracture toughness (KJc) in the ductile-to-brittle transition
temperature region, using the so-called the “Master Curve”(MC) method. This estimates the fracture toughness versus
temperature curve from a test at one temperature. Since

fracture initiation is assumed to be cleavage in the MC method,
it is questionable whether the method is applicable to materials
subject to intergranular fracture.
The KJc values of the material exhibiting a mixed mode fractureof intergranular and cleavage due to intergranular embrittlement
exhibit a large scatter. A Weibull plot of KJc values is shown in
Fig.5-12, together with the data fitting denoted as a blue line by
the standard MC method. The figure clearly indicates that the
distribution of KJc values does not follow the linear regression
defined in the standard method. It seems that the distribution
of KJc values is from an inhomogeneous material creating
two distributions of fracture toughness. This feature becomes
more pronounced with increase in P concentration at the grainboundaries. Since fracture toughness depends on the type ofbrittle fracture which initiates at the crack front, the obtained
KJc values may follow the dual fracture toughness distributionsof cleavage and intergranular fracture. We applied a modified
MC analysis to the estimation of intergranular fracture bycensoring the upper KJc data in Fig.5-12. The test temperaturedependence of KJc is shown in Fig.5-13. The solid and dashedblue lines represent the mean curve and lower bound for
fracture, respectively. The dependence of material subject tointergranular fracture as well exhibits low KJc values below thelower tolerance bound of brittle fracture. The modified method
enables a more conservative fracture toughness estimate, as
shown by the solid and dashed red lines in Fig.5-13.

Reference
Nishiyama, Y., Onizawa, K. et al., Phosphorus Segregation and Intergranular Embrittlement in Thermally Aged and Neutron Irradiated Reactor Pressure
Vessel Steels, Journal of ASTM International, vol.4, issue 8, 2007, p.12, Paper ID JAI100690.

62

JAEA R&D Review 2008

Nuclear Safety Research

5-7

Taking Advantage of Spent Fuel Being Harder toMake Critical than Fresh Fuel–Acquisition of Verification Data for Burn Up Credit Analysis Methods–

Core
Tank

In. Dia.
59cm

150cm

Fuel
Rod

Sol. Level

Sol.
Fuel

Critical Solution Level (cm)

Rh conc.
㨪1.3g/L
Cs conc.
㨪3.4g/L
Sm conc.
㨪0.14g/L

Rh sol.
Add

Cs sol.
Add

Eu sol.
Add
Eu conc.
㨪0.02g/L

Final U sol. composition(Experiment No.14)‫ޓ‬U: 323g/L Sm: 0.12g/L
Sm sol.‫ޓ‬Cs: 2.8g/L Rh: 1.2g/L
Add ‫ޓ‬Eu: 0.02g/L Free Acidity: 2.0N

No FP (only U sol. fuel)U conc.㨪330g/L

Sol. FuelNozzleFig.5-14 Heterogeneous system consisting of fuel rods
and solution fuel
There were 333 uranium dioxide fuel rods arrayed at 1.5-cm
intervals in a cylindrical core tank. Uranyl nitrate solution
was fed from the bottom, and the level of the solution when
it became critical was measured.

In a nuclear reactor, nuclear burn of fuel proceeds withfission chain reaction, and thus fissile material such as uranium
is depleted and fission products (FP) accumulate. Because
both such changes hinder the chain reaction, spent fuel is by
its nature harder to make critical than fresh fuel. The so-called“introduction of burn up credit” quantitatively evaluates thisnature which enables handling of larger quantities of spent fuelin transportation, storage, reprocessing, etc. without danger ofcritical accidents
Up to now in Japan, there has been no example of burnup credit introduction considering of FP accumulation
although there is one considering fissile material depletion.
To quantitatively evaluate the FP accumulation effect, it isnecessary to accurately estimate both the amount of FP in spent
fuel and its negative effect on criticality.
Precise measurement of the negative effect of FP was
conducted at the Static Critical Experiment Facility (STACY) in
the Nuclear Science Research Center. Simulating the dissolving
process of a reprocessing facility, where the advantage ofburn up credit introduction is large, uranium solution fuel(235U 6wt%) was fed through a bottom nozzle to a coretank in which uranium fuel rods (235U 5wt%) were arrayed(Fig.5-14) and the critical level of the solution was measured.
Natural elements of samarium (Sm), cesium (Cs), rhodium(Rh) and europium (Eu) were added in that order to thesolution so that their final concentrations became equivalent
to that produced by the power generation of 30 GWday per 1-t
uranium. As a result, criticality was harder to reach, i.e., critical

Experiment No.

Fig.5-15 FP elements addition and critical level increase
As Sm, Cs and Rh were added in sequence to the uranium
solution fuel, neutron absorption increased and critical solution
level became higher. Although Eu was also added, theeffect of reduced concentration of Sm etc. due to increaseof total solution volume overcame the effect of increased Eu
concentration, and critical solution level decreased.

solution level was raised as shown in Fig.5-15.
Effect on criticality is expressed as change of neutron effective
multiplication factor keff (keff is the change rate of the number ofneutrons mediating chain reactions. keff is 1.0 at criticality). The
effects of each element per unit concentration were estimatedto be －3293±40 ¢/(g/L) for Sm, －26.5±1.4 ¢/(g/L) for Cs,
etc. (“¢” is a unit of keff change; 7.6×10-5 for this experiment.)
Multi-group cross sections based on the nuclear data library
JENDL-3.3 were produced by the SRAC code, and keffs were
calculated by the TWODANT code. Computation analysis of
the effect of each element based on the keff s agreed well withthe measurement for Eu but gave slightly larger values than
measurement for Sm, Cs, and Rh. Although the calculation
was significantly larger than the measurement uncertainty
only in the case of Sm, it should be noted that estimation bycomputation only may result in a larger negative effect thanin reality.
Thanks to recent developments, the perturbation method now
can be used with Monte-Carlo codes such as MCNP and MVP,so that the effect on criticality can be calculated, which should
be validated with this measurement results. It is also necessary
to improve accuracy of the post irradiation examination (PIE)
to measure the amount of FP in actual spent fuel.
This report includes results of experimental andcomputational studies performed under a contract with the
Ministry of Education, Culture, Science and Technology ofJapan.

Reference
Tonoike, K. et al., Benchmark Critical Experiments and FP Worth Evaluation for a Heterogeneous System of Uranium Fuel Rods and Pseudo FP Doped
Uranium Solution, Proceedings of 8th International Conference on Nuclear Criticality Safety(ICNC 2007), Russia, 2007, p.222-227.
JAEA R&D Review 2008

63

Nuclear Safety Research

5-8Evaluating the Radioactivity Concentration of Radioactive
Wastes below Which They Can Be Disposed–Dose Calculation to Derive the Upper Bound of Radioactivity
Concentration in Disposal of Transuranium Wastes–

Groundwater migration scenario

Site-reuse scenario
Disposal at construction andresidence site࡮External exposure, Inhalation
and Ingestion (farm products)

Waste
TrenchDisposal

River
Waste
Concrete
Vault
Disposal

Depth50m㨪100m
Waste

Discharge of contaminatedgroundwater into a river࡮Ingestion (water, fish, farmand livestock products) etc.

Table5-1 Recommended values of upper bound ofradioactivity concentration for 3 disposal methods
We calculated exposure doses resulting from threedisposal methods for low-level radioactive wastes.
Based on the results, the NSC decided the upperbounds, considering both easiness of measurement ofradioactivity concentration and concentration distributionof radioactive waste to be disposed of.

Intermediate
Depth Disposal

Radionuclide
Development of model and code forderiving the upper bound of theradioactivity concentration, according
to scenarios and barrier features ofthree disposal methods

Fig.5-16 Scenarios and exposure pathways used to derive
upper bound of radioactivity concentration in disposal
We developed an assessment code (GSA-GCL ver. 2) toderive the upper bound of radioactivity concentration for threedisposal methods: trench disposal, concrete vault disposal
and intermediate depth disposal. This code supports thedose estimation for the exposure pathways of a site-reusescenario and of a groundwater migration scenario.

The upper bound of radioactivity concentration during
disposal means the maximum concentration of radionuclides
of waste repository allowable in a license application for that
repository. In Japan, there are three concepts for methods ofdisposing of low-level waste categories: near surface disposal
without an artificial barrier (trench disposal), near surface
disposal with an artificial barrier (concrete vault disposal), andintermediate depth disposal. The Nuclear Reactor Regulation
Law has not been amended to set the upper bound of theradioactivity concentration for each of these disposal methods oftransuranium (TRU) wastes. We developed an assessment code
(GSA-GCL ver. 2) to derive the upper bound of radioactivity
concentration for TRU wastes, according to geological andartificial barrier features in three disposal concepts. This code
supports the dose estimation for the exposure pathways in two
typical scenarios: (a) the site is a residential area and (b) the

C-14

Recommended Values of Upper bound ofRadioactivity Concentration(Bq/ton)
TrenchConcrete Vault Intermediate
Disposal
Disposal
Depth Disposal
─

1E+11

1E+16

Cl-36

─

─

1E+13

Co-60

1E+10

1E+15

─

Ni-63

─

1E+13

─

Sr-90

1E+07

1E+13

─

Tc-99

─

1E+09

1E+14

I-129

─

─

1E+12

Cs-137

1E+08

1E+14

─

α-radionuclide＊

─

1E+10

1E+11

(＊)The values of upper bound for radionuclide emitting α
radiation are represented as those for Am-241 for trenchdisposal and Np-237 for intermediate depth disposal,
respectively.

groundwater from the site is discharged into a river (Fig.5-16).
We specified important radionuclides and their dose levels
based on calculations using our code, setting appropriate
geological and artificial barrier conditions for each disposal
method. These dose estimations were adopted in the reportpublished by the Nuclear Safety Commission (NSC); “Upper
Bounds of Radioactive Concentration in Burial of Low-Level
Radioactive Solid Waste (in Japanese), May 2007”. In thisreport, the NSC recommended upper bound values consideringboth ease of measurement of the radioactivity concentrationand the variation in concentration of radioactive waste to bedisposed of (Table5-1). Their recommended values wereprescribed by ordinance of Ministry of Economy, Trade andIndustry in 2008. This work was performed by JAEA undercontract with Nuclear and Industrial Safety Agency in Ministryof Economy, Trade and Industry.

References
Takeda, S. et al., Estimation of Radioactivity Concentration Limit for Trench Disposal of Transuranium and Uranium Wastes (Contract Research), JAEAResearch 2008-044, 2008, 64p. (in Japanese).
Takeda, S. et al., Estimation of Radioactivity Concentration Limit for Intermediate Depth Disposal of Transuranium and Uranium Wastes (Contract
Research), JAEA-Research 2008-045, 2008, 60p. (in Japanese).
Sawaguchi, T. et al., Estimation of Radioactivity Concentration Limit for Concrete Vault Disposal of Transuranium and Uranium Wastes (Contract Research),
JAEA-Research 2008-046, 2008, 62p. (in Japanese).

64

JAEA R&D Review 2008

Nuclear Safety Research

5-9

How Long Does a Repository Confine Radionuclides?

–Modeling of Long-Term Alteration of Bentonite Buffer Material–

1710͠
50 90 30
͠ ͠ ͠

170͠
130͠

st
u

dy

90͠

is

50͠

Th

Overpack

Dissolution rate

CementBentonitekg/m3/s)

BedrockCalculated values using
a previous model

Vitrified waste

Activity of OH- (mol/dm3)

Fig.5-17 A conceptual view of high-level radioactive
waste (HLW) disposal system
Vitrified HLW is encapsulated in an iron container (overpack),surrounded by engineered buffer material (bentonite) andplaced in a shaft contained in stable bedrock. Cement isused for mechanical support of the shafts, but it causes
surrounding water to become highly alkaline, which is likely
to deteriorate the properties of the bentonite buffer over along period.

Fig.5-18 Dissolution rate of montmorillonite in highly
alkaline solutions
The dissolution rate of montmorillonite, the main constituent
of bentonite, was investigated using compacted bentonite.
The dependence of the rate on activity of hydroxide ions,aOH-, and on the temperature is shown here. There is higher
dependence on aOH- than a model based on results ofexperiments using pulverized bentonite. The rate is sensitive
to the alkalinity.

In the Japanese program, high-level radioactive waste (HLW)
is vitrified, encapsulated in a metal container called overpack,surrounded by engineered bentonite buffer material, andplaced in a repository cut out from stable bedrock as shown in
Fig.5-17. Cement is expected to be applied for the mechanicalsupport of the repository.
Besides its mechanical buffer function, bentonite prevents
groundwater from intruding into the waste and retards
radionuclide movement outwards by adsorbing them. It isnecessary to estimate the long-term changes of such propertiesof bentonite confining long-lived radionuclides in HLW.
Radioactive decay of radionuclides in HLW increases thetemperature of the surrounding bentonite up to 90ºC right after
the emplacement, and an alkaline environment is likely to be
induced by the cement used in the repository. The temperature
and alkaline condition possibly accelerate alteration of
bentonite.
Because the radionuclide confining property of thebentonite arises from its main constituent montmorillonite,dissolution of montmorillonite in contacting groundwater
results in deterioration of the confinement. Several researchers
have obtained the dissolution rates of pure pulverized
montmorillonite in free water with high water/solid-ratio,
which are far from the repository condition where bentonite isin a compacted state. The property of water filling the narrow
pore spaces (porewater) in the compacted bentonite may bedifferent from that of free water. We performed the dissolution
experiments of montmorillonite in a compacted state simulating
the repository condition.
The dissolution rate of montmorillonite we obtained wasa function of activity of hydroxide ions, aOH- (mol/dm3), andtemperature, T (K), as shown in Fig.5-18. The rate, RA (kg/m3/s),
could be formulated as:
RA = 3.5 × 103(aOH-)1.4exp(-51000/RT)[kg/m3/s]
where R is the gas constant. The dissolution rate obtained for acompacted state has a higher dependence on the aOH- than thatof previous models that were based on high water/solid-ratio
dissolution experiments using pulverized materials.
The aOH- of porewater is an important factor determining
the dissolution rate, and this is determined by both chemical
reactions with minerals and mass transport. We developed acoupled mass-transport/ chemical-reaction code for predicting
the aOH- and the long-term alteration of the bentonite.
Further investigation is necessary to understand themechanisms of the dissolution of compacted montmorillonite
and to verify the model by comparing results of the model
calculation with laboratory experiments or field observations.

Reference
Yamaguchi, T. et al., Experimental and Modeling Study on Long-Term Alteration of Compacted Bentonite with Alkaline Groundwater, Physics andChemistry of the Earth, vol.32, no.1-7, 2007, p.298-310.
JAEA R&D Review 2008

65

6 Advanced Science Research
Advanced Basic Research to Create the Future
and transuranium compounds and measurement of theirmacroscopic quantities and electronic structure, clarification ofmagnetic structure, magnetic excitation, and the mechanism of
superconductivity using NMR, μSR, neutron scattering, andtheoretical methods; design of novel materials using megagravitation and nano-particle deposition, topmost surface
studies using bright and coherent positron beams; elucidationof the interaction mechanism of heavy elements and ionizing
radiation within living cells by spectroscopy, and studying
the primary and fundamental processes in the interactions ofionizing radiation with matter.
In order to promote this research, we are collaborating withother research sections in JAEA, and several international
collaborations are ongoing. In addition, we accept new researchsubjects based on public suggestions within the framework of
the Reimei Research Promotion project of JAEA.

In the Advanced Science Research Center, new frontierresearch of nuclear energy and ionizing radiation expected tobear fruit in the future is conducted to discover new principles
and phenomena, and furthermore to create new materials andtechnologies. In order to achieve these aims, we have fourbasic policies; (1) to pursue research for which the high levelresearch capability (researchers and facilities) in JAEA is
effectively used and which is difficult to do in other researchorganizations, (2) to achieve results before the rest of the world
does, (3) to nurture a new basic research area until it becomes
fruitful, (4) to explain and apply the research, thus fulfilling our
responsibility to society, in conformity with the Third Scienceand Technology Basic Plan.
The following research is going on: nuclear physics andnuclear chemistry of superheavy elements, the nuclear shellstructure, reaction dynamics and electrochemistry using
heavy-ion beams of accelerators; synthesis of uranium

Extreme Environment and Substance Science
Magnetoresistance ratio (%)

Superheavy Element Nuclear Science

Recoil mass separator for experimental research
on fusion reaction mechanisms to produce
super-heavy elements by bombarding deformed
nuclei with heavy-ion beams from JAEA tandembooster accelerator.

experimental data
connectionof nano-Co
particles

C60-Co compound with
Co nanoparticles (nano-Co)

Research of new
hybrid materials oforganic molecules
and transition metals
for highly functional
spintronics.

Co content, x (C60Cox)
Finding of anomalously large tunnel magnetoresistanceeffect (~90%) not predicted by existing theory

Actinide Material Science

Material and Life Sciences
Discovery of thefirst Np-based
heavy fermion
superconductor
with large upper
critical magnetic
field (500,000
times of thegeomagnetism)

Reference electrode Counter electrodeProteinWorking electrode
Incident light

Uranium
solutionTransmitted
light

Apparatus for
examining electrontransfer between
bacteria and heavy
elements.

Slab optical guide
Incident light

Transmitted light

NpPd5Al2 high quality single crystal

Advanced Science Research
Collaborations with Universities, Reimei
Research Promotion Project

Other Research Sections in JAEA

Fig.6-1 Four Research Fields and Collaboration in Advanced Science Research Center

66

JAEA R&D Review 2008

International Collaborations

Advanced Science Research

Chemical Characterization of Superheavy Elements–Fluoride Complexation of Element 104, Rutherfordium–

Distribution Coefficients (mL/g)

6-1

Superheavy Elements
Lanthanide[MF6]2(M=Rf,Zr,Hf)

Slope -2

Slope -2

Concentration of Nitrate Ion (mol/L)

ActinideFig.6-2 Periodic table of the elements
Superheavy elements are represented by blue panels. Group-4 elements
studied in the present work are indicated by red panels. (Superheavy
Elements beyond element 112 have not been formally recognized yet.)Superheavy elements with atomic numbers ≥ 104 are currently
placed on the uppermost end of the periodic table as shown in
Fig.6-2. Chemical properties of lighter elements arranged in
the same columns of the periodic table show gradual variation,and those of superheavy elements are theoretically expected
to have greater deviations among homologues due to strong
relativistic effects on their electronic structure.
In the present study, we studied anion-exchange behavior ofelement 104, rutherfordium (Rf) in hydrofluoric acid and nitricacid (HF/HNO3) mixed solutions. The Rf atoms were producedin the nuclear reaction between a 248Cm target and 18O beams
delivered from the JAEA tandem accelerator. It was clarified
for the first time that the properties of fluoro complexes of Rf
are significantly different from those of its lighter homologues
Zr and Hf.
In Fig.6-3, the variations of distribution coefficients (Kd) of
Rf, Zr and Hf are shown as a function of the concentration of
the nitrate ion. Kd is the concentration ratio of the element in
the resin phase to that in aqueous solution; larger Kd shows
stronger adsorption on the resin. The Kd values of Rf, Zr and
Hf decreased with increase in the concentration of the nitrateion. These decreasing trends indicate that the nitrate ionsentering in the binding sites of the anion-exchange resin drive
the anionic fluoride-complexes of Rf, Zr, and Hf out of theFig.6-3 Variations of distribution coefficients(Kd) of rutherfordium (Rf), zirconium (Zr) andhafnium (Hf) as a function of the concentrationof the nitrate ion
The Kd values of Rf, Zr and Hf are indicated byred, green, and blue circles, respectively.

resin into the aqueous solution. The slopes in the logarithmicplot of Kd vs. the concentration of the nitrate ion are -2 for all
the elements. These -2 slopes show that the anionic fluoridecomplexes have the same -2 charges, clearly demonstratingthat Rf forms a hexafluoro complex, [RfF6]2-, as do Zr and Hf.
It was also found that the Kd values of Rf are smaller than thoseof the homologues by about 2 orders of magnitude. We further
investigated the variation of the Kd values depending on theconcentration of the fluoride ion to understand the differencein the Kd values. We then found that about 2 orders of higher
concentration of the fluoride ion is necessary to form [RfF6]2than that to form [ZrF6]2- and [HfF6]2-. It was thus understood thatthe difference in Kd between Rf and its homologues in Fig.6-3
results from the fact that Zr and Hf have completely formed
hexafluoro complexes while the Rf complex is incomplete at
the concentration of fluoride ions used.
This suggests that Rf 4+ has a larger ionic radius than Zr4+
and Hf 4+ because formation of fluoro complexes of the group4 elements depends on the ionic radii of the tetravalent ions.
An ionic radius of Rf would be influenced by the relativistic
effects that change the electronic structure of Rf. Although it
has not so far known how relativistic effects change the ionicradii, the present result is a key to solve the question.

ReferenceToyoshima, A. et al., Hexafluoro Complex of Rutherfordium in Mixed HF/HNO3 Solutions, Radiochimica Acta, vol.96, no.3, 2008, p.125-134.
JAEA R&D Review 200867

Advanced Science Research

Heavy Fermion Superconductivity in Neptunium Compound–A New Material Advances Actinide Science–

Resistivity (Ǵǡcm)6-2

Temperature (K)Fig.6-4

Photograph of NpPd5Al2 single crystal

Fig.6-5
NpPd5Al2 crystallizes into a body-centered tetragonal lattice.
Np is surrounded by Pd atoms. Electrical resistivity goes tozero at 5 K, demonstrating superconductivity.

Actinide elements such as uranium (U), neptunium (Np),plutonium (Pu), which are used as nuclear fuels, are very
important materials in atomic energy science and technology.
At the same time, these elements and their compounds have also
attracted attention for their novel physical properties, because
5f electrons in actinides sometimes exhibit novel magnetic and/or superconducting behavior. While uranium compounds withless radioactivity have widely been investigated, the solid-statescience of transuranium compounds including neptunium andplutonium began rather recently. The discovery of plutoniumbased superconductor PuCoGa5 in 2002 has demonstrated
that many new compounds as well as unknown scientificphenomena might be hidden in actinide compounds.
In 2007, we found a new neptunium-based superconductor(Fig.6-4). The occurrence of superconductivity in neptuniumcompounds is very surprising because they often exhibit strong
magnetism, which usually destroys superconductivity. The
detailed characterization of the new compound has shownthat it has the composition NpPd5Al2 and crystallizes into atetragonal crystal structure (Fig.6-5).
The discovery of NpPd5Al2 stimulated scientists in this field,
leading to the discovery of several new rare-earth and actinide
compounds with the same structure.
5f electrons in actinide elements change their behavior
depending on the chemical or physical environment
surrounding them. In particular, plutonium exhibits multiple
structural phase transitions among six different crystal
structures with increase in temperature from room temperature
to its melting temperature. No other elements behave in such away. The discovery of NpPd5Al2 demonstrates that discovering“new materials” by providing a new environment for actinide
elements can lead to discovery of new phenomena. We hopethat the present actinide material research will aid in deeper
understanding of physical behavior of actinides.
This work has been performed in collaboration with Tohoku
University and Osaka University.

References
Aoki, D., Haga, Y. et al., Unconventional Heavy-Fermion Superconductivity of a New Transuranium Compound NpPd5Al2, Journal of the Physical Societyof Japan, vol.76, no.6, 2007, p.063701-1-063701-4.
Haga, Y. et al., Crystal Structure and Magnetic Properties of the New Ternary Actinide Compounds AnPd5Al2 (An = U,Np), Journal of Alloys andCompounds, vol.464, 2008, p.47-50.

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Advanced Science Research

6-3

Microscopic Investigations of Uranium and TransUranium Oxides

signal Intensity (a.u.)–Nuclear Magnetic Resonance Studies of the Electronic States–(a) UO2

T= 28 KInternal Field (Oe)(b) NpO2

T= 17 KInternal Field (Oe)(c) PuO2

T= 6 KInternal Field (Oe)Fig.6-6 17O-NMR study of UO2, NpO2 and PuO2
The transverse axis shows an internal field at oxygen sites induced byf-electron spins. The vertical axis shows signal intensity, which is proportional
to the number of oxygen ions responding to that internal field.

Fig.6-7 Magnetic octupolar ordering in NpO2
Colors indicate the weights of spin up (red) and down (blue) states.
The spin is distributed anisotropically according to the anisotropic charge
distribution, and the electronic state of each Np ion has a finite magnetic
octupole moment.

To understand macroscopic properties of materials, it isessential to understand their electronic properties from a
microscopic viewpoint. Recently, we have carried out a seriesof nuclear magnetic resonance (NMR) studies for actinideoxides ( AnO2 : An= U, Np, Pu, Am, etc ). The microscopic
NMR data make a clear distinction between their electronic
properties at low temperatures.
Fig.6-6 shows the 17O-NMR spectra observed in UO2, NpO2
and PuO2, respectively. There is a marked difference in the widthand shape of the spectra. For UO2 (a), we have obtained a broad
spectrum with a rectangular line shape. This characteristic line
shape indicates the existence of a large internal filed at oxygensites due to the ordering of magnetic dipoles. On the other
hand, the spectrum for NpO2 (b) broadens rather moderately,
as compared with that for UO2, while it exhibits a complex
structure. The electronic ground state of NpO2 had remained
a mystery for many years. Our recent NMR study on a single
crystal has revealed the occurrence of a novel magnetic
octupolar ordering in the ground state (Fig.6-7). Finally, we
have observed a narrow spectrum for PuO2 (c). This confirmsthat a nonmagnetic electronic ground state is realized even at

the low temperature of 6K.
The actinide dioxides discussed here are all insulators withthe same cubic crystal structure. Even though, these materials
exhibit a variety of electronic states at low temperatures. Why
they are so different?
All the actinide ions in AnO2 have the same tetravalentstate. Therefore, the number of localized f-electrons per ion
is two for U4+, three for Np4+ and four for Pu4+. In addition,
due to the strong spin-orbit interaction, the f-electrons carry
multipole degrees of freedom: dipole, quadrupole, octupole,etc. The available multipoles on each material are dependent
on the number of the f-electrons as well as the symmetry of
crystal. A variety of physical phenomena result from a variety
of available multipoles in AnO2.
AnO2 represents perhaps the most studied series ofany actinide compounds. From a chemical and industrial
perspectives, this interest has stemmed from their use as nuclear
fuels. Nowadays, however, AnO2 is also an important series ofmaterials to promote a better understanding of the fundamental
physics of multipoles. We are now preparing to advance our
NMR study to AmO2 as a joint work with Tohoku University.

Reference
Tokunaga, Y. et al., NMR Studies of Actinide Dioxides, Journal of Alloys and Compounds, vol.444-445, 2007, p.241-245.
JAEA R&D Review 200869

Advanced Science Research

6-4

Atomic Scale Observation of Metal Degradation–Development of a Positron Scanning Microscope–

AnalyzingmagnetCold head

Magnetic lensSmall positronsourcePositron gun
Density ofvacancies
High

Objective lensTwo-dimensionalscanning system
DriverLow
Controller

Detector

Fig.6-8 Schematic drawing of the positron microbeam
apparatus
Positron beam from the source is focused by the objective
lens. Synchronizing the measurement of γ-rays and stage
scanning, distribution of atom vacancies beneath the surface
is obtained.

Fig.6-9(a) Optical image of the stress corrosion cracking of stainlesssteel.
(b) The spatial distribution of atom vacancies.
(c) Overlapped image of (a) and (b). Atom vacancies appear
before the crack grows.

Positron annihilation method is a powerful tool to detect
atom vacancies in solids. This technique is based on theunique property of positrons. That is, positrons get trappedby atom vacancies and annihilate with electrons to emit
γ-rays. By measuring the annihilation γ-rays, vacancy defects
are characterized. To implant positrons into a specimen at acontrolled depth, positron beams are used. The diameter ofa conventional positron source is, however, rather large andhence it is difficult to focus the beam down to a few microns.
This means that measurement of vacancies in a specific area
is impossible. In this study, we newly developed a positronsource which is suitable for the formation of well-converged
beam. Employing an objective lens, we succeeded in focusing
a positron microbeam to a diameter of 1.9 microns. Using
this microbeam, we can inject positrons in a certain small
area only and evaluate vacancy type defects in that specificregion. Moreover, combining a stage controller which can aim
the beam at an arbitrary place on a sample, we constructed ascanning positron microscope which can obtain distributionsof vacancy type defects beneath the surface (Fig.6-8).
Using this positron microscope, degradation of materials canbe diagnosed through the observation of defects on the levelof atom vacancies. Stainless steel, which is hard to corrode
and tough is widely used as a structural material of nuclear
power plants. The interior of the nuclear reactor is a severe
environment, with high temperature, high pressure, and highradiation dose. In such extreme conditions, stress corrosion
cracking (SCC) may occur. In order to secure the safety of
nuclear reactors, understanding of detailed mechanism of
crack progress is required. Using the positron microscope, we
observed the tip of a crack formed in stainless steel by stresscorrosion in an environment simulating a nuclear reactor(Fig.6-9). We then found that the density of atom vacancy typedefects increased at the region beyond the tip of the crack. Thisis the world’s first observation of atom vacancy type defects
being formed before a crack grows.
Thus, the positron microscope can offer original information
about atom vacancy type defects which are undetectable byconventional devices, such as the optical microscope and theelectron microscope.
Present study is the result of “Innovative Nuclear Researchand Development Program” entrusted to “Japan Atomic Energy
Agency (JAEA)” by the Ministry of Education, Culture, Sports,
Science and Technology of Japan (MEXT).

Reference
Maekawa, M. et al., Development and Application of Positron Microprobe, 17th Iketani Conference “Doyama Symposium on Advanced Materials”,
Transactions of the Materials Research Society of Japan, Tokyo, Japan, vol.33, issue 2, 2008, p.287-290.

70

JAEA R&D Review 2008

Advanced Science Research

Biological Effects of Clustered DNA Damage–Mutation Induction by DNA Lesions in Close Proximity–DNAradiation
track

ionization
excitation
clustered DNA
damage

3.4nm

Mutation frequency (㧑)6-5isolated damage

nodamage8-oxoG
isolated

DHT 8-oxoG+DHT
isolated clusteredFig.6-10 DNA damage by ionizing radiation
DNA is damaged due to ionization or excitation causedby radiation. Clustered DNA damage would be produced
where the density of ionization/excitation is high, whereas
the isolated damage would be generated where it is low.

Fig.6-11 Induced mutation frequency by cluster DNA
damageUsing a newly developed mutation detection system toobserve effects of synthetic DNA lesions, it was foundthat the mutation frequency was higher at the cluster DNA
damage than that at isolated lesions.

Exposing cells to ionizing radiation leads to lethality,
mutation induction, and carcinogenesis. These biologicaleffects of radiation are considered to result from various
chemical alterations (DNA damage) induced in DNA. Various
types of DNA damage are induced by radiation, such as “basedamage” where the base of DNA is damaged, “DNA strand
break” where DNA helix is broken, and so on. On the other
hand, it has been known for a long time that certain chemical
agents generate same kind of DNA damage with those inducedby radiation. However, it was pointed out that the biologicaleffects of chemical agents are not as significant as those ofionizing radiation even when an equivalent number of lesions
are induced. In order to explain this seemingly contradictory
outcome, an idea of “clustered DNA damage” was proposed.
Clustered DNA damage is a type of damage in which multiple
DNA lesions are induced within a region of a few nm (Fig.6-10).
It is assumed that dense ionization and excitation in or aroundthe DNA caused by radiation leads to clustered DNA damage,and that, as a result of its low reparability, the biological effectof radiation becomes substantial. Although results indicating
the presence of clustered DNA damage have been accumulated,

there was no experimental evidence on whether clustered DNA
damage is relevant to biological consequences. This was because
it is difficult to relate the effect of clustering of damage withbiological effects, as ionizing radiation induces various kinds
and numbers of damage in DNA at random positions. In orderto overcome this difficulty, we have devised and developed anexperimental system using model cluster damage composed ofseveral artificially-synthesized base lesions positioned at close
proximity. With this method, it has become possible to examine
the biological effects of defined types, numbers, and positionsof damage. Our experiment reveals that the clustered DNA
damage including two base lesions (8-oxo-7,8-dihydroguanine(8-oxoG) and dihydrothymine (DHT)) is more mutagenic thanisolated lesions (Fig.6-11). It is experimentally demonstrated,
for the first time, that base lesions at close proximity enhance
the mutagenic potential. This result confirms the biologicalsignificance of clustered DNA damage and will provide
important implications for the fields of biological applicationof radiation, such as risk assessment, radiotherapy of cancer,
and mutation breeding.

Reference
Shikazono, N. et al., The Roles of Specific Glycosylases in Determining the Mutagenic Consequences of Clustered DNA Base Damage, Nucleic Acids
Research, vol.34, no.13, 2006, p.3722-3730.
JAEA R&D Review 200871

7 Nuclear Science and Engineering Research
Formation of Basis for Nuclear Energy R&D, and Creationof Innovative Nuclear Energy Utilization Technology
The nuclear science and engineering R&D activities at
the Japan Atomic Energy Agency have the four roles shownin Fig.7-1. In order to fulfill these roles, research in nuclear
data and reactor engineering, fuels and materials engineering,
environment and radiation science, and nuclear applied heattechnology is being conducted.
(1)Forming sound basis forinnovation(high-temperature gas-cooled reactor,
partitioning and transmutation, etc.)(2)Supporting various nuclear-related
activities of government and industries(foundations of nuclear safety, technology
for non-proliferation regime, etc.)

Light Water Reactor
Advanced Reactor

Nuclear Fusion Reactor

Innovative andsustainable developmentof Nuclear technology

Fast Breeder Reactor

Quantum BeamTechnology

Fuel Cycle Technology

Collaboration with
Industries and academia(3)Providing technological platform
to industries and universities(Nuclear Energy Research Collaboration
Center, etc.)(4)Preparing basic codes andatabases(Nuclear data, Reactor engineering, Material
engineering, environment technology)

Fig.7-1 Roles of Nuclear Science and Engineering Research

Nuclear data and reactor engineering
Various types of research are being performed to investigate
the feasibility of advanced nuclear systems and to establish thebasic technology for these systems. Development of nuclear
data evaluation codes in Topic 7-1 and measurements ofnuclear data are continuing, and the JENDL Actinoid file 2008,
which can predict FBR neutronic characteristics with higher
precision, was completed. The extended bias factor method
was established to evaluate nuclear design accuracy based on
multiple criticality experiments and to identify new experiments
needed for achievement of the required accuracy. Detailed
two-phase analysis codes processed with supercomputing
technology as well as model experiments for code validation
are being developed to establish new thermal design methods
for advanced nuclear systems. Topic 7-2 describes a threedimensional thermal-hydraulic measurement to build a detailed
database for code validation. Basic studies on transmutationof long-lived nuclides are being conducted so as to reduce theburden of radioactive waste management.
Fuels and materials engineering
Basic studies on advanced nuclear fuel and cycle technology
and the degradation of nuclear power plant materials are being
made. In the advanced nuclear fuel cycles, minor actinides(MA), which are classified as high level radioactive wastes
in the current nuclear fuel cycle, are expected to be recycled
to reduce the burden of waste disposal. To develop thistechnology, it is required to understand and control behaviorof MA in the system. We are, therefore, performing R&D on
separation technology of actinides and fission products, suchas high decontamination (high purity) uranium separation by
72

JAEA R&D Review 2008

mono-amide extractants, together with fabrication technology
and property measurements of MA-bearing fuels (Topic 7-3).
Moreover, R&D is being carried out on irradiation effects ofmaterials for nuclear reactors, the corrosion damage mechanismof stress corrosion cracking of reactor structural materials,
and the corrosion mechanism of reprocessing plant materials.
Topic 7-4 describes the result of recent research on stresscorrosion cracking mechanisms.
Environment and radiation science
R&D on environmental behavior of radionuclides (Fig.7-2),
utilization and detoxication techniques of wastes, and advanced
analytical and monitoring techniques for ultra-trace amounts ofsubstances in the environment are being carried out, together
with R&D on fundamental mechanisms of biological radiation
effects, dose assessment, and a shielding calculation techniquefor various radiations. Topic 7-5 describes development of aradionuclide database which enables accurate dose assessment
on a gene level. Topic 7-6 describes the only environmental
sample analysis method which can determine 235U enrichment
in micro-particles, a safeguard technology without mass
spectrometry.
Carbon stock in soil and CO2 flux
in beech forests of Appi, Iwate
Mean residence time

Assessed based on
14
C values
Carbon stock in soil
11.3 kgC m-2(Top 20 cm depth)

Fig.7-2 Carbon Cycle inForests
Flux of CO2 released fromsoil was assessed bydetecting 14C from nuclear
weapons testing. These
results will contribute toresearch for elucidating
the role of forests asrelease and absorption
sources of CO2.

CO2 flux

Nuclear applied heat technology
To expand nuclear energy application to heat utilizing
industries, we are continuing extensive R&D for hightemperature gas-cooled reactor (HTGR) technology and for
an HTGR-heated hydrogen production system. Through thefirst 30 day full power operation of the High Temperature
Engineering Test Reactor (HTTR), the HTTR fuel exhibited
excellent fission product retention. Moreover, we are performing
R&D for an advanced core characteristic evaluation techniquebased on the HTTR data, for higher fuel burn-up and longlife graphite structure. Topic 7-7 describes the development ofZrC coated fuel particles. For hydrogen production, research
for a higher efficiency thermo-chemical IS process (Topic 7-8)
is being carried out, together with research into concentrationof hydrogen iodine using a polymer electrolyte demarcation
membrane, corrosion resistant materials, use of glass lining
for hot section piping, and an indirect measuring method ofcomposition of the sulfuric acid and hydrogen iodide solution.

Nuclear Science and Engineering Research

7-1Compilation of Reliable Nuclear Reaction Data

–Development of Comprehensive Code for Nuclear Data Evaluation,
CCONE–

Proton(n,p)

Elastic Scattering

Fission
Neutron
Nucleus

Neutron Spectrum (mb/MeV)En

14 MeV

Measured Data
Total
DirectPre-equilibrium
FissionInelastic Scattering
CaptureSpallation

Emitted Neutron Energy (MeV)

Fig.7-3 Types of nuclear reactions

Fig.7-4 Comparison of calculated and measured neutron
spectraReactor designs, shielding analyses, and safety evaluationsof criticality etc. need precise prediction of neutron behavior.
The neutron behavior is determined by nuclear reactions(Fig.7-3) of the nuclei which make up the materials used. The
probability of a nuclear reaction is determined by nuclear crosssection. The data related to nuclear reactions like cross section
are called nuclear reaction data. Reliable nuclear reaction
data are indispensable for accurate reactor design and otherapplications. The nuclear reaction data and other data such as
half-lives of radioactive nuclides are broadly called nuclear
data, and they are the basis for planning nuclear applications.
Neutron reaction data are especially important for reactor
applications. Reliable reactor designs and safety analyses
depend on the reliability of the basic nuclear data used.
Activities to provide reliable nuclear reaction data havecontinued from the early days of nuclear development. The
nuclear reaction data are based on theoretical models andmeasurements, and they are required to cover neutrons in theenergy range from thermal (0.025 eV) to high energy producedby fission (~ 10 MeV) in their effects on all used materials tounderstand the neutron behavior in a reactor. The number ofnaturally occurring stable isotopes which make up materials
amounts to over 220 and the measurement of these over the fullenergy region is difficult. The measured data and the theoreticalmodels are appropriately combined to provide all the needed
nuclear reaction data. As there is no single theoretical model
applicable to this wide energy region, various kinds of nuclear
models have to be used to predict the nuclear reactions. This
procedure requires much effort to keep consistency among themodel calculations.
We have developed a comprehensive code for nuclear data
evaluation, CCONE, which maintains this consistency. As
the code uses an object-oriented programming language, itis easy to calculate the neutron and γ-ray spectra emitted bycomplicated reactions producing many nuclei by dynamicallytreating them as objects. The code has been tested for the
nuclear data of thorium, uranium and plutonium isotopes andits accuracy has been confirmed. Fig.7-4 shows the energy
spectrum of neutrons emitted after impinging on uranium-238.
Although the spectrum consists of many nuclear reaction
processes (direct, pre-equilibrium, fission processes; the total
spectrum is their composite), the calculation by CCONE agrees
well with the measured data. This code is to be used for futurenuclear data evaluation to provide reliable nuclear reaction
data. The development of CCONE was honored with a Special
Award and Award for Distinguished Technology by the Atomic
Energy Society of Japan in March, 2008.

Reference
Iwamoto, O., Development of a Comprehensive Code for Nuclear Data Evaluation, CCONE, and Validation Using Neutron-Induced Cross Sections for
Uranium Isotopes, Journal of Nuclear Science and Technology, vol.44, no.5, 2007, p.687-697.
JAEA R&D Review 200873

Nuclear Science and Engineering Research

7-2

Detailed Measurement of Boiling Flow–Monitoring 3D Change of Vapor / Water Distribution by Neutron
Beam–

Penetration
direction ofneutron beam

Crosssection

Vapor
Ratio ofvapor inboiling water
Water

Inside view of thetest section

Vapor

Growth of vapor
Neutron
beam

Water

Outside view of
the test section

Time (ms)Power 60kW
Time-averaged 3D distribution
Time-change of vapor and waterof vapor and water(Time resolution 1ms)
(Spatial resolution 0.2mm)

Fig.7-5 Measuring 3D changes of vapor/water distribution in the heated rod-bundle by neutron beam
Changes in spatial distribution of boiling flow in the fuel assembly of a water-cooled breeder reactor affect the cooling
performance of the nuclear reactor. In this study, we developed a technique by which we can measure the changes in spatial
distribution of vapor and water, using a neutron beam.

One of the important considerations in the design of thewater-cooled breeder reactor is the estimation of cooling limit ofa fuel assembly. Since the cooling performance depends on the
distribution of water and vapor, we need high definition data inorder to understand the flow behavior between rods and thereby
estimate the cooling limit. However, it is difficult by traditional
measurement techniques to measure the flow distribution in aheated bundle. Thus, we developed a new technique which
can measure in detail the distribution of boiling flow, using aneutron beam which has a high sensitivity to water. We carriedout tight-lattice 14-rod bundle experiments using “JRR-3”
as the neutron source (Fig.7-5). The aim of the experiments
was to measure the changes in spatial distribution of vapor
and water in the heated rod bundle. This is a new technique

combining neutron radiography and computed tomography.
We can understand the high definition 3D distribution of vapor
in boiling flow. We also measured the flow behavior. It was
found from the visualization with the time resolution of 1ms
that the vapor accumulation point which was observed in 3D
data was washed away by cooling water periodically. These
results were of use not only for advancing our understandingof various phenomena but also for the verification of advanced
thermal hydraulic codes.
This page contains some results obtained by the project“Development of Fuel Assembly for Very High Burn-up
Water-cooled Breeding Reactor” entrusted from the Ministry of
Education, Culture, Sports, Science and Technology of Japan.

Reference
Kureta, M., Experimental Study of Three-Dimensional Void Fraction Distribution in Heated Tight-Lattice Rod Bundles Using Three-Dimensional Neutron
Tomography, Journal of Power and Energy Systems, vol.1, no.3, 2007, p.225-238.

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7-3

Demonstration of Excellent Thermal Properties of
Minor Actinide Nitride Fuel–High-Quality Thermal Conductivity Data of MA Nitride Obtained
from Small Samples–
Glove box

SampleHeaterVacuumpump
AmplifierSample holder

Laser
PC

Fig.7-6 Schematic diagram of the thermal
diffusivity measurement apparatus
In the glove box, the oxygen and moisture contents
were controlled to less than two and three ppm,
respectively.

Thermal conductivity (Wm-1K-1)

Infrared detector

Temperature (K)Fig.7-8 Thermal conductivity of AmN corrected totheoretical density, together with those for UN, NpN,
PuN and UO2Thermal conductivity of AmN was smaller than UN, NpN and
PuN, but larger than UO2.
Fig.7-7 Sintered AmN disk sample
The sample color was black. The diameter and thickness
were about 3 mm and 0.6 mm, respectively.

The spent fuels arising from nuclear power plant containminor actinides (MAs:Np,Am,Cm) which are long-lived
radioactive nuclides, and thus, the management of MA is a key
issue for sustainable nuclear energy use. To solve this problem,
the technology of partitioning and transmutation of MA, which
can significantly reduce the heat generation and radiotoxicity
of high level radioactive waste, has been examined. Among
various fuel types, nitride fuel is one of the candidates for MAbearing fuels for transmutation because of its high melting
point, good thermal conductivity and mutual solubility.
Facing this task, it is important to obtain thermal property
data such as heat capacity and thermal conductivity to design
the MA-bearing fuels. However, the thermal properties of MA
nitrides are little known due to their high specific radioactivity
and the high reactivity with water and oxygen. Thus, weinstalled a laser flash apparatus in a glove box with a highlypurified argon gas atmosphere (Fig.7-6) for thermal diffusivity
measurement and a drop calorimeter for heat capacity
measurement. These apparatuses enabled us to obtain highquality thermal property data of MA nitrides.
The diameter of the MA nitride disk sample prepared by the
carbothermic reduction of MA oxide was about 3 mm (Fig.7-7).

With this small disk and its fragments, we measured the thermal
diffusivity and heat capacity, respectively.
We have measured the thermal properties of various MA
nitrides. As an example, the results of thermal conductivity
measurement of AmN is here described. It was found that thethermal diffusivity of AmN tended to slightly decrease withincreasing temperature and the heat capacity of AmN was nearly
the same as other actinide nitrides. The thermal conductivityof AmN, which was determined from the thermal diffusivity,
heat capacity and bulk density, tended to slightly increase
with temperature in the temperature range from 473 to 1473
K (Fig.7-8). It was also found that the thermal conductivityof AmN was smaller than UN, NpN and PuN, but larger than
UO2. Thus, in this study, the advantage of MA nitride fuel was
confirmed in view of its thermal properties. In addition, we
have been measuring the thermal expansion of MA nitrides byhigh temperature X-ray diffraction analysis.
The present study was carried out within the task
“Technological development of a nuclear fuel cycle basedon nitride fuel and pyrochemical reprocessing” entrusted
from the Ministry of Education, Culture, Sports, Science andTechnology of Japan.

Reference
Nishi, T. et al., Thermal Diffusivity of Americium Mononitride from 373 to 1473 K, Journal of Nuclear Materials, vol.355, issues 1-3, 2006, p.114-118.
JAEA R&D Review 200875

Nuclear Science and Engineering Research

7-4

New Approach for Testing Aging of Reactor Structural Materials–Study of Stress Corrosion Cracking (SCC) Mechanisms Using
Electron Backscatter Diffraction (EBSD) in Mesoscale Analysis–(a) SCC cracks at heataffected zone(b) Plastic strain distribution
at area with no cracks (Background)

Color bar of plastic
deformation level

Tensile load

20%

Ǵ
10%

(c) Plastic strain distribution at SCC

Compacttension
specimen0%ǴFig.7-9 (a) Intergranular type SCC in compact-tension specimen, (b) plastic strain distribution at no crack area, and(c) at SCC area
In area (c), background strain was subtracted; blue lines indicate SCC cracks.

Austenitic stainless steels and nickel base alloys are used as
structural materials of primary loop recirculation (PLR) pipes
and shrouds/core barrels in light water reactors (LWRs). Thesesteels and alloys are susceptible to SCC which is one of thequasi-brittle fracture modes. SCC was observed in the structural
materials of LWRs after long term operation. Estimation of
SCC initiation and growth behavior is important for preventive
maintenance of LWRs.
The mechanisms of SCC are not clear yet. Recently, thefollowing results on macro- and microscopic scale SCC havebeen obtained, but the mechanism relating the microscopic
phenomena to macroscopic phenomena is still unknown. On
the macroscopic scale, crack growth rate of SCC increases
with increase in the materials’ hardness and the stress intensity
factor at the crack tip. The effect on hardness induced byirradiation damage was equivalent to that induced by cold work
and welding processes. On the microscopic scale, segregationand precipitation are induced by heating during the welding
process, and the effects of neutron irradiation are observed at
grain boundaries (GBs). The segregation and precipitation areaffected by the grain boundary character. SCC occurred also in
low-carbon stainless steel which had no Cr depletion at GBs.

Change of deformation mode from uniform to localized occurs
in the grain matrix of irradiated materials.
We developed a new experimental technique using an EBSD
in order to determine the properties of grain boundaries with andwithout crack propagation and plastic strain distribution nearcracks. Fig.7-9 shows plastic strain near SCC cracks measured
by this new technique in a compact-tension heat-hardened zoneof a real size mock-up PLR pipe joint which is made of low
carbon stainless steel. It was revealed in the heat affected zonethat the plastic strain near grain boundaries reached about 15%
and was much higher than that in the grain matrix (Fig.7-9(b)).
When SCC occurred in such hardened stainless steel, additional
plastic strain (about 10%) was observed along the SCC cracks(Fig.7-9(c)). The plastic strain gradient became very steep at
the crack tip. It is likely that this steep strain gradient is one of
the causes of SCC in low carbon stainless steel.
This pioneering work enables us to study the mechanismof SCC on a mesoscale and to model the SCC initiation andpropagation behavior. We will improve prediction of aging
degradation of reactor structural materials by means of multiscale modeling.

Reference
Kaji, Y., Miwa, Y. et al., Multi-scale Analysis of Deformation Behavior at SCC Crack Tip (II) (Contract Research), JAEA-Research 2007-008, 2007, 69p.
(in Japanese).

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JAEA R&D Review 2008

Nuclear Science and Engineering Research

Advancement of Nuclear Medicine Dosimetry–Development of a Radionuclide Database
Microdosimetry in Subcellular Dimensions–ApplicabletoNumber of electrons per nuclear transformation

7-5

Auger electron energy (keV)

Fig.7-10 Radionuclide data book
“MIRD: Radionuclide Data and Decay Schemes, 2nd Edition”
published by the Society of Nuclear Medicine (Reprinted bypermission of the Society of Nuclear Medicine)Fig.7-11 Improvement of resolution of Auger electron
spectrum of 125I(a) First Edition (b) Second Edition

Nuclear medicine has been widely used as a safe and painlesstechnique to diagnose and treat disease using small amounts ofradiopharmaceuticals labeled with 99mTc, 123I, 67Ga, and so on.
In diagnostic and therapeutic nuclear medicine, radiation doseto patients from radiopharmaceuticals is evaluated to perform
appropriate examination while avoiding unsafe radiation
exposure. We have updated a radionuclide data book entitled“MIRD: Radionuclide Data and Decay Schemes” (published in1989) used for nuclear medicine dosimetry and have completed
its Second Edition (Fig.7-10) with the cooperation of theMedical Internal Radiation Dose (MIRD) Committee of theSociety of Nuclear Medicine (SNM), a leading organization in
nuclear medicine research.
In the Second Edition, the data of Auger electrons emitted
from radiopharmaceuticals have been enhanced so thatmicrodoses can be calculated. Several radionuclides used asradiopharmaceuticals decay by electron capture and/or internalconversion and emit a number of low-energy Auger electrons

as a consequence of these decay processes. These lowenergy Auger electrons deposit their energy over ranges with
subcellular dimensions (nm ~μm) in the human body and cause
severe biological damage by DNA strand break. To evaluate
DNA damage, we have developed a computer code EDISTR04
which calculates detailed spectra of Auger electrons emitted
from all radiopharmaceuticals by electron capture and internalconversion (Fig.7-11). Using this code, we have developed a
database concerning 333 radionuclides for radiation dosimetry
calculation. These radionuclides include the 242 nuclides
addressed in the First Edition and an additional 91 nuclides for
future use in nuclear medicine.
Radiation dosimetry methods and related data developedby SNM comprise the standard methods used in the field of
nuclear medicine. This new radionuclide database also will be
used for radiation dosimetry to improve and develop diagnostic
and therapeutic methods in nuclear medicine.

Reference
Eckerman, K. F., Endo, A., MIRD: Radionuclide Data and Decay Schemes, 2nd Edition, The Society of Nuclear Medicine, 2008, 671p.
JAEA R&D Review 200877

Nuclear Science and Engineering Research

7-6

Selective Detection of Ultra-Fine Particles Containing
Highly Enriched Uranium–Development of an Environmental Sample Analysis Method for IAEA Safeguards–
(a)

(b)

(c)

Enrichment of uranium in particle (㧑)Fig.7-12 Microscopic images of FTs created by uranium particles: (a) natural,(b) 10% enriched, (c) 85% enriched
(d)

(e)

(f)

(g)

・Etching condition
・Particle diameter

Etching time (min)Fig.7-13 Correlation between the time until a fission
track appears by etching and the enrichment of uranium
in particle

Fig.7-14 Microscopic images of FT morphologies (d, f)
and of uranium particles (e, g) corresponding to the
FT

Environmental sample analysis technique was adopted by
the International Atomic Energy Agency (IAEA) as a newtechnique for the strengthened safeguards system to detect
undeclared nuclear activities. In the technique, particle analysis,
which measures isotope ratios of uranium and/or plutonium inindividual particles, is an effective tool for safeguards because
the particle analysis can surmise the history of the nuclear
activities in the past.
The particle analysis by a fission track (FT) technique
combined with thermal ionization mass spectrometry (TIMS)
can measure the isotope ratios of uranium particles smaller
than one micrometer in diameter (equivalent to about 4 pg for
uranium). Fig.7-12 shows the microscopic images of FTs ofuranium particles with different enrichment irradiated withthermal neutrons. By observing the FTs, a particle containing
fissile materials can be detected with high sensitivity from anenvironmental sample taken from nuclear facilities.
The number of FTs is dependent on the size as well as the235
U enrichment of uranium particle. The etching rate of FTs
by the uranium particle increases with the 235U enrichment of
uranium particles in the case of irradiation with a fixed neutron
fluence. This implies that the detection of uranium particles
with a specific enrichment is possible by controlling theetching time. Fig.7-13 shows the correlation between the time

until a fission track appears by etching and the enrichment of
uranium in particle. The etching time to detect the FT becomesshorter with increase in 235U enrichment. This result indicates,for example, that the uranium particles with an enrichment of
more than 25% and 10% will be detected mainly by the etchingtime is set to 60 min and 100 min, respectively.
In the case of a mixture of highly enriched uranium particles
with small size and lowly enriched particles with large size, it
is difficult to distinguish between the two solely by controllingthe etching time. The size of both the particles is compared inthis case. When the numbers of FTs are comparable, it is clearthat the smaller particle has the higher enrichment (Fig.7-14).
In the real sample containing particles of various sizes, weanalyze the particles which made size distribution uniform
roughly by a suction filter.
Comparing the FT morphology and the particle size enables
the screening of the uranium particles according to theirenrichment. Particularly, even when uranium particles witha higher 235U enrichment are scarce in a myriad of uranium
particles, it is possible to detect them reliably. The uranium
particles detected are transferred onto a TIMS filament, and theisotope ratios are measured precisely. The FT-TIMS method
developed has attracted the attention of IAEA as a strong tool
for safeguards.

Reference
Lee, C. G. et al., Development in Fission Track-Thermal Ionization Mass Spectrometry for Particle Analysis of Safeguards Environmental Samples, Journalof Radioanalytical and Nuclear Chemistry, vol.272, no.2, 2007, p.299-302.

78

JAEA R&D Review 2008

Nuclear Science and Engineering Research

7-7

Improving Performance of Very High Temperature
Reactor Fuel
–Development of ZrC Coated Fuel Particle–
UO2 kernel
Low-density pyrolytic
carbonFluidized
particles

The 3rd layer：SiC→ZrC
High-density pyrolytic
carbonInjection ofsource gas

Approx. 0.9 mm
Fig.7-15 Structure of the coated fuel particle
The tiny coated fuel particle is used as HTGR fuel, andconsists of a uranium dioxide (UO2) kernel and four thinceramic coatings.

Fig.7-16 ZrC coater
Inside the coater, ZrC is applied by chemical vapor deposition
on particles fluidized by the coating material gases, zirconium
bromide and methane, at over 1300°C.

Zirconia kernel
Pyrolytic
carbonFig.7-17 Photos of ZrC coated particles, (a) appearance and (b)(c) cross sections
A ZrC coating layer with C/Zr atom ratio of 1.00 was successfully deposited after coating tests using
dummy zirconia particles with weight close to UO2 fuel particles and after developing inspection techniques
for ZrC stoichiometry etc. Advanced techniques to coat pyrolytic carbon and ZrC successively are being
developed.

The High Temperature Gas-cooled Reactor (HTGR) has
inherent safety features with heat output of nearly 1000°C
which has prospects to be utilized for hydrogen production andhigh efficiency power generation. The Very High Temperature
Reactor (VHTR) is evaluated highly worldwide, and is aprincipal candidate for the Generation-IV reactor systems.
The HTGR fuel uses coated fuel particles made with refractory
ceramics as shown in Fig.7-15. Silicon carbide (SiC) coated fuelparticles were installed in the High Temperature Engineering
Test Reactor “HTTR”. We are developing zirconium carbide(ZrC) coated fuel particle as an alternative to SiC coated fuel totake advantage of ZrC’s excellent properties. ZrC has a higher
melting point than the temperature of SiC decomposition andhigher chemical stability against metallic fission products suchas palladium than SiC. ZrC is expected to be applied as the
VHTR fuel, which will be used under high temperature andextended burnup conditions. On the other hand, for ZrC coatedfuel particle to have optimum performance, stoichiometric ZrC,
whose quantity ratio of Zr to C is 1, should be deposited. We
have constructed a ZrC coater (Fig.7-16) applying an original
fabrication technique using zirconium bromide and methanegases, and also developed ZrC layer inspection techniques.

At first, non-uniform stripes were observed in the deposited
ZrC. It was found by transmission electron microscopy
observation that segregated excess free carbon made thenon-uniformity. Uniform ZrC was successfully deposited bycontrolling ZrC depositing temperature to be 1300-1400°C inorder to prevent excessive pyrolysis of methane. In parallel, amethod to measure ZrC stoichiometry was newly developedby conducting both inductively coupled plasma - atomic
emission spectrometry and infrared absorption spectrometry,enabling measurement of C/Zr atom ratio with high accuracy
on the order of 0.01. Finally, the fabrication conditions forhigh quality stoichiometric ZrC as shown in Fig.7-17 weresuccessfully determined. Now we are accumulating data fordesigning a future commercial ZrC coater (kg-batch scale), andare irradiating ZrC samples in a US research reactor to study
the performance of ZrC under irradiation. The present study
is the result of “Research and development for advanced hightemperature gas cooled reactor fuels and graphite components”
entrusted to the Japan Atomic Energy Agency by the Ministryof Education, Culture, Sports, Science and Technology ofJapan.

Reference
Ueta, S. et al., Development on Fabrication and Inspection Techniques for the ZrC-coated Fuel Particle as an Advanced High Temperature Gas Cooled
Reactor Fuel, Hyomen, vol.46 ,no.4, 2008, p.222-232 (in Japanese).
JAEA R&D Review 2008

79

Nuclear Science and Engineering Research

7-8

For Efficient H2 Production Using IS Process–Success in Measuring SO2 Pressurization Effect on the Bunsen
Reaction–

SO2

Pressurization

New analysis procedure toacquire composition information
important to evaluate H2
production thermal efficiency.

Water
Preventing
evaporation

New method to sample
the product solution while
Pressurized.

Composition
ChangeSampling

Gas Phase
SO2

Light Phase
H2SO4+H2O+SO2

Heavy Phase
HI+I2+H2O+SO2

Bunsen Reaction

Fig.7-18 Outline of the method to measure SO2 pressurization effect on the product of the Bunsen reaction
The Bunsen reaction is the most complicated reaction in the IS process and a key reaction for improving the efficiency of
the H2 production process. In this reaction, SO2 gas contacts with a mixture of water and iodine, generating sulfuric acid andhydrogen iodide. The product solution separates into a light phase and heavy phase spontaneously.

We are investigating a large scale hydrogen (H2) production
method, the IS process, using a High Temperature Gas-cooled
Reactor (HTGR) as the heat resource. By using the IS process
which is a thermo-chemical reaction cycle using Iodine (I) and
Sulfur (S) as recycling agents, H2 can be obtained from H2O
without emitting CO2.
The following reaction is the Bunsen reaction which is oneof the major reactions in the IS process.
SO2+I2+2H2O→H2SO4+2HI
By obtaining a highly concentrated solution of sulfuric acid
(H2SO4) and hydrogen iodide (HI), we can reduce the separationenergy of H2SO4 or HI from H2O and other impurities, whichis essential for high thermal efficiency of the IS process.
From Le Châtelier’s principle, by increasing the iodine (I2)
concentration and the partial sulfur dioxide (SO2) pressure, thereaction equilibrium shifts to the product side. We have foundthat increasing the I2 concentration is effective, but it has beena problem how to measure the SO2 pressurization effect.
There was no indirect measurement method suited to thispressurized system, because of I2 dissolving in the productsolution (the heavy phase, Fig.7-18). Even in the case of
direct composition analysis using sampling method, it was the
problem that the components dissolved in the product solution
dissipated into the atmosphere during sampling and thecomposition change due to the reaction between the productsolution and the water to prevent evaporation of the productsolution.
We developed a new method of direct composition analysis
by which we could sample the product solution while
maintaining the reaction conditions including the high pressure.
To deal with the composition change caused by water, we
devised an analysis procedure of the post-reaction composition
to acquire the composition information needed to evaluate theH2 production thermal efficiency.
With this method, we can easily search for favorable
reaction conditions for obtaining more concentrated productsolution and thus accelerate the development of an efficient H2
production process.

Reference
Nakajima, H., Imai, Y. et al., Effect of Sulfur Dioxide Partial Pressure on the Reaction of Iodine, Sulfur Dioxide and Water, Kagaku Kogaku Ronbunshu,
vol.33, no.3, 2007, p.257-260 (in Japanese).

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JAEA R&D Review 2008

8 Nuclear Fuel Cycle Technological Development
Promotion of Nuclear Fuel Cycle Development andTechnical Co-operation with Industry
We are carrying out research and development of the nuclear
fuel cycle, specifically such activities as examination of mixed
oxide (MOX) spent fuel for reprocessing and development oftechnology for vitrification of high-level radioactive waste, tosupport and promote reprocessing and plutonium recycling inJapan.
Moreover, in our technical co-operation with the nuclear
fuel cycle private business of Japan Nuclear Fuel Ltd. (JNFL)
at Rokkasho village, Aomori Prefecture, we are providing
technology transfer of results of our research and development,
dispatching our engineers and training their engineers, etc. as
requested by JNFL, to ensure that their business progresses
satisfactorily.
1. Reprocess technology development
We began examination of reprocessing of the “FUGEN”
reactor MOX type B spent fuel which has higher plutonium
content and higher burn-up than the fuel which has been
reprocessed since February 2007 at Tokai Reprocessing Plant(TRP). We obtained various data on dissolving processes,
undissolved residue, solvent degradation, etc. of MOX spent fuel(Fig.8-1). We will be organizing these data, and reconsidering
our plans for examinations after FY 2009 based on evaluationof the earthquake resistance improvement measures carried out
at TRP.
To develop technology for vitrification of high-level
radioactive waste, we obtained data on stable operation of themelter during the manufacturing of vitrified canisters. We are
executing research and development concerning an advanced
vitrification melter which is to have long life. We made asmall test melter in FY 2007 and will test it in order to set
specifications for a commercial melter in the future.
To develop technology for reduction and stabilization of lowlevel radioactive waste, we examined cement solidification ofsimulated low-level liquid waste, and clarified the solidification
conditions suited to various kinds of waste fluid. Moreover,
we carried out tests of nitrate decomposition of low-level
radioactive liquid waste containing nitrates, and gathered data

concerning the decomposition performance etc. of reductants,
confirming their applicability.
In addition, we are making arrangements with relatedorganizations for examination of light-water nuclear reactor
high burn-up spent fuel reprocessing to upgrade reprocess
technology for high burn-up of fuel.
2. Technical co-operation(1) Technical co-operation for uranium enrichment
business
We are carrying out technical co-operation with JNFL for
their R&D of a super-efficient new material centrifuge, sharing
our uranium enrichment technology.
We will continue to execute technical co-operation with
JNFL necessary for the introduction of this super-efficient newmaterial centrifuge into the Rokkasho Enrichment Plant.
(2) Technicalco-operationforreprocessing
business
We are carrying out technical co-operation, such as dispatchof our engineers who have operation experience at TRP, andextra funding for operation and management of the vitrification
facility etc. to support smooth execution of the tests for
activation of the Rokkasho Reprocessing Plant JNFL whichhave been going on since March 2006.
(3) Technical co-operation to MOX fuel fabrication
business
We are adjusting our engineer dispatch plan with JNFL for
operation of a MOX fabrication facility, a permit for which
JNFL has applied.
(4) Other technical co-operation
We dispatch our engineers upon request of the Nuclear
Material Control Center, and provide technical co-operation
concerning nuclear material control in Rokkasho district.
Moreover, we are offering our technical information to
JNFL, based on the “Agreement of technical co-operation
concerning technological proposals for the NFRC construction
as part of the GNEP plan of the United States” concluded with
JNFL in June, 2007.

Undissolved residue
collected from dissolution
step of reprocessing

Fig.8-1 “FUGEN” MOX spent fuel reprocessing test
Componential analysis and grain degree distribution measurementof the undissolved residue collected from the dissolution step ofreprocessing are to be carried out for data on MOX spent fuelreprocessing.
JAEA R&D Review 200881

Nuclear Fuel Cycle Technological Development

8-1

Vitrification Melter Improvement–Development of Long-Life High
Waste (HLW) Vitrification Melter–

Level

Radioactive

LiquidGlass & HLW
Electrodes(2) Movable and remotely
exchangeable

Off gas(1) Skull layer to extend
life-time

Skull layer
Cooling jacket(Air or Water)
Molten glass

Fig.8-2 Outline of advanced melter and development objectives
To extend vitrification melter life-time from 5 years of the current design base to 20 years, corrosion rate must be decreased
drastically. (1) Skull layer formation technology and (2) Movable and remotely exchangeable electrodes are introduced for thispurpose.

In a vitrification melter, borosilicate glass cartridges filledwith high level radioactive liquid waste (HLW) are fed directly
into a melter and melted at high temperature around 1150°C.
Thus, the refractory melter wall is gradually corroded by thehigh temperature glass.
The Tokai Vitrification Facility (TVF) melter is designed
to be exchanged every five years because of corrosion of its
electrodes and refractory wall.
The objective of this development is to extend vitrification
melter life-time from the 5 year standard upon which current
design is based to 20 years to reduce the economic burdenof high level radioactive waste management, by applying
advanced technologies (Fig.8-2).
Development of a new long-life vitrification melter for thispurpose is focused on establishing measures for preventing
corrosion of the refractory and electrode, and preventing
accumulation of noble metal particles at the bottom of themelter.
The new melter wall structure, which forms a skull layer(a solid glass layer or low-temperature high viscosity melted
glass layer) on the wall by cooling the wall, is being considered
as a measure against refractory wall corrosion. To prevent
electrode corrosion, a “commutative electrode structure” isbeing considered. This structure avoids fixing the current
electrodes within the refractory wall by inserting them fromthe top of the melter in an exchangeable configuration.
Moreover, basic tests of the corrosion rate are being carriedout for data on candidate materials for the refractory and theelectrode.
In the advanced melter, good drainage of noble metal will beattained by special temperature control of melted glass in thelower part of the melter. Viscosity of glass changes according
to glass temperature and concentration of noble metal. Thus weare continuing to gather basic data in our laboratory. In a normal
condition, glass temperature of a conical chamber decreases
as the radius decreases. To drain noble metal smoothly, higher
glass temperature along the wall of conical part than inside isnecessary. Glass viscosity along wall then becomes lower thanthat of glass near the center, resulting in flow of accumulated
noble metal rich glass along the wall and out of the melter. We
call this temperature distribution mode the “reverse viscosity
operation mode”.
We started advanced melter development in 2005 and will
promote melter design and related R&D activity until 2008,
aiming to determine the most feasible melter concept design.

Reference
Aoshima, A. et al., Vitrification Experience and New Technology Development in Tokai Vitrification Facility, Proceedings of European Nuclear Conference2007(ENC2007), Brussels, Belgium, 2007, 5p., in CD-ROM.

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Nuclear Fuel Cycle Technological Development

8-2

Developing to LLW Encapsulation Stage for Disposal–Future Planning of LLW Treatment at the Tokai Reprocessing
Center–【Nuclide Separation】【Nitrate Decomposition】

Reagent【Cement Encapsulation】
Evaporator

Catalyst,
Reductant

S C

Nitrate effluent

Mixer

UF

ReceivervesselReaction
vesselIon
exchanger

Reaction vessel

StorageSlurry effluent

Evaporator【Nitrate Decomposition Technology】

S C
Mixer

(Cement hardening)(Recycle)
Storage

Present
use

Fig.8-3 The low level radioactive waste treatment
facility (LWTF)
2 underground stories and 5 above-ground stories
of 2,400m2 (40m × 60m); Reinforced concrete
structure. Solid waste is treated above ground andeffluent is treated in the basement.

Futureplan【Cement Encapsulation】

Fig.8-4 Present and future procedure of LLW treatment

The Low Level Radioactive Waste Treatment Facility(LWTF) was completed on September 2006 after about
4.5 years of construction (Fig.8-3). The role of LWTF is todemonstrate treatment technology for solid and effluent LLW.
This section describes the effluent treatment, comprising
Nuclide Separation, Hardening with Cement, and Nitrate
Decomposition, the last two technologies to be installed in the
future (Fig.8-4) .
1. Development of Nuclide Separation Technology
In the Nuclide Separation Process, the nuclide and salt in
the effluent are separated prior to the cement hardening to
reduce the cost of disposal. Following the Nuclide Separation,
almost all the waste is converted into a nitrate effluent witha low radioactivity. The amount of slurry effluent with highradioactivity that will require disposal deep underground isreduced to about 10% of the original waste. In consideration of
the chemical behavior of each nuclide, the following processes
were incorporated in the Nuclide Separation.
(1) Radio-iodine is precipitated as Silver Iodide (AgI) by
adding Silver Nitrate (AgNO3) and separated from theeffluent with an Ultra-Filtration Membrane (UF) whose
pore diameter is several tens of Å.
(2) α nuclides such as Pu and βγ nuclides such as Ruthenium
are co-precipitated with Iron Hydroxide (FeO(OH)) by
adding Ferric Nitrate (Fe(NO3)3), and separated from theeffluent with a UF membrane.
(3) Highly soluble materials like Cesium are adsorbed after
ion exchange performed with [TiO(CO3) / TiO(ONa)2]n andK2Co[Fe(CN)6].
2. Development of Cement Hardening Technology
Each of the liquids produced during separation of eachnuclide is evaporated to adjust the water volume and mixed
with cement by In-drum Mixing. The cement used is “SC,”slag based cement. SC has ability to encapsulate the waste athigh loading in aptly short time. SC has proved to be capable ofuniform solidification when nitrate effluent is mixed in up to 50
wt% in a beaker. In these cases, adequate compressive strength
over 10MPa was achieved. Further 200ℓ trials have been
performed, measuring the strength and density. It was found inthese tests that the nitrate effluent was uniformly solidified.
3. Development of Nitrate Decomposition Technology
A high concentration of NaNO3 is contained in the effluent.
Density of nitrate in the disposal environment has to be kept
appropriately low. To solve this problem, we plan to decompose
the sodium nitrate by a Catalyzed Reduction Process using
a metal catalyst and reductant (hydrazine (N2H4) and formic
acid (HCOOH)). Sodium Hydroxide (NaOH) is generated
with decomposition of NaNO3. The NaOH is then recycled
as a reagent, which is expected to reduce the encapsulated
waste. The process has been demonstrated to be capable ofdecomposing over 90% of the nitrate in a NaNO3 solution
whose concentration is about 4.7mol/ℓ. Further study isnecessary to optimize the operating conditions and to improvethe catalyst life-time, enabling practical applications.

Reference
Saito, Y. et al., Development of New Treatment Process for Low Level Liquid Waste at Tokai Reprocessing Plant, Proceedings of International Symposiumon Radiation Safety Management 2007 (ISRSM2007), Daejeon, Korea, 2007, p.275-280.
JAEA R&D Review 200883

9 Development of Decommissioning and Radwaste Treatment Technology
Aiming at Establishment of Rational Process for Radwaste
Management
A well-planned, rational, and safe execution of
decommissioning and radwaste treatment/disposal is one ofour major missions, and these are goals of our midterm plan.
In addition, promotion of the disposal business of radwaste
arising from our research activities and those of universities,
institutes and industrial facilities etc. is specified as another ofour missions in the amendment to the law governing us on June2008.
Because great cost is required to accomplish these missions,research and development (R&D) for cost effective measures
is called for. We therefore are pursing R&D aiming at
establishment of a rational process for decommissioning andtreatment/disposal of radwaste as shown in Fig.9-1. In addition,
specific technologies have been developed at each site suited tothose nuclear facilities.
The major progress in 2007 was as follows.
R&D for Decommissioning
Prototypes of an engineering system to plan decommissioningof nuclear facilities and of an evaluation system to provide
safety clearance for wastes were constructed.
At the Fugen reactor site, an evaluation method was verified
for residual radioactive inventories of core components(Topic 14-1). The technique of uranium removal using IF7 wasdeveloped at the Ningyo-toge site (Topic 14-11). An all-atonce removal technique was applied to the waste storage tanks

at the Tokai site.
R&D for Waste Treatment
A decontamination technique using supercritical carbon
dioxide fluid to remove plutonium from radwaste has beenstudied. Furthermore, we are developing a calcination techniqueto remove the vinyl bag packing waste and a reduction method
using a precious metal catalyst for denitration of TRU waste.
A waste data management system using input such as waste
identification was studied, and a model database for Nuclear
Science Research Institute was developed.
R&D for Waste Disposal
For simple and rapid analysis of radionuclides in radwaste,a method was developed for rapid determination of 89Sr and90
Sr in liquid waste using a Sr extraction disk and β-ray
spectrometer (Topic 9-1).
Evaluation of radioactive characteristics of waste from anuclear reactor and safety evaluation of the uranium waste
disposal process are in progress. For the long term safety
evaluation of disposal of uranium waste, exposure doses were
evaluated in the cases of topography or sea level change caused
by geologic upheavals or climate change. It was found that the
exposure dose would be relatively high if there were upheaval
or erosion.

Below Clearance
LevelsGeneration of RadWaste

Verification

TrenchRepository

Incombustible
Separation

Operation of Nuclear
Facilities

Noncombustible

Decommissioning /Dismantlement

Combustibles

Incineration

LiquidConcentration

Technologies for
Decommissioning
Nuclear facilities

Compression

Technologies for
RadWaste Processing

Free-release

Store in
Storage
Facilities

Processfor Final
DisposalConcrete-pit
Repository
Deep depth
repository
Geological
DisposalTechnologies for Final
Disposal of RadWaste

Fig.9-1 R&D Subjects in Decommissioning and Radwaste Processes from Generation to Disposal

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JAEA R&D Review 2008

Development of Decommissioning and Radwaste Treatment Technology

Determination of 90Sr in Radioactive Waste–Development of Rapid and Simple Method for Evaluation ofRadioactivity Inventories in Waste–

Sample solution

Shielding

PhotomultiplierOutput

Plastic
scintillator
Solid phaseextraction disk

Coincidence
moduleGM tubeOutputSampleMulti Channel
Analyzer

Suction(a) Solid phase extraction
Sr is extracted by the disk.

(c) β-ray spectrometryRadioactivity of 90Sr is determinedthrough analysis of the β-ray spectrumof sample by comparing with standard
libraries.

Conventional method Bq/ml)
Fig.9-3 Comparison between 90Sr
values found by conventional andrapid methods
The amount of 90Sr in actual radioactive
liquid waste determined by the two
methods were in agreement with eachother.

CountsCountsSpectrumas measured

β-ray energy

Rapid method (Bq/ml)

9-1

β-ray energy(b) β-ray measurement
When the scintillation pulses are sorted by a multi-channel analyzer
using a coincidence module, a β-ray spectrum without influence of
γ-rays is obtained.
Fig.9-2 Rapid and simple 90Sr determination method

For proper disposal of radioactive waste packages generated
from various research facilities, an evaluation of radioactivity
inventories in the waste packages is indispensable. In orderto establish a practical and reliable verification method,
we are currently collecting data concerning radioactivity
concentrations in radioactive wastes before processing.
90
Sr is one of the most important nuclides from the viewpointof safety assessment of the radioactive waste disposal process.
Since 90Sr emits only β-rays when it decays, nondestructive
measurement cannot be done. In a conventional method, Sr isseparated by ion-exchange chromatography or a precipitation
method. After the Sr fraction reaches radioactive equilibriumwith 90Y, the daughter nuclide of 90Sr, the 90Sr is separated
and quantified. The tedious and time-consuming chemical
separation in this method is a major problem.
Therefore, we developed a new rapid and simple

determination method for 90Sr in radioactive wastes. In thismethod, 90Sr in the waste sample solution is quickly separated bya solid phase extraction disk (Fig.9-2(a)). After the extraction,
radioactivity of 90Sr is also quickly determined throughanalysis of the β-ray spectrum of the sample by comparing
with predetermined standard libraries (Fig.9-2(b),(c)). Thismethod was successfully applied to the determination of 90Sr in
liquid waste from nuclear research facilities.
The values of 90Sr determined by the rapid method werecompared with the results of the conventional method, andthey were in agreement with each other (Fig.9-3). As a resultof our development work, separation of 90Y after reachingradioactive equilibrium was no longer needed. The time for the90
Sr determination was shortened to be less than one-third to
one-fifth of the conventional method.

Reference
Kameo, Y. et al., Rapid Determination of 89Sr and 90Sr in Radioactive Waste Using Sr Extraction Disk and Beta-ray Spectrometer, Journal of Radioanalyticaland Nuclear Chemistry, vol. 274, no.1, 2007, p.71-78.
JAEA R&D Review 200885

10 Collaboration on Basic Nuclear Engineering
Promotion of R&D to Meet the Demands of Industry and Academia
Nuclear Engineering Research
Collaboration Center
Director General of
Collaboration CenterIndustry
Academia

Nuclear Science andEngineering Directorate

President

Director General

Concurrent
post

Joint research
Advanced Reprocessing
Materials Development Group
Human
resources

Industrial Collaboration
Promotion Department

Protection of intellectual property
and confidential information
Research and Development
Center○ Tokai Research andDevelopment Center○ Oarai Research andDevelopment Center○ Tsuruga Head Office

SupportSpecial Groupfor Advanced Technology
on LWR Thermal Hydraulics
Special Groupof Developing Hypersensitive
NDA Method for U &Pu

Deputy Director General
Research Co-ordination andPromotion Office

Collaboration

Division of Nuclear Data andReactor Engineering
Division of Fuels andMaterials EngineeringHuman
resources

Graphite & Carbon Materials
Characterization Special Group

Nuclear Applied Heat
Technology Division
Division of Environment andRadiation Sciences

These groups are presently
located in the Tokai Research andDevelopment Center.

Fig.10-1 Nuclear Engineering Research Collaboration Center and Related Organizations

JAEA has been promoting joint research and developmentwith industry and academia and has made its nuclear facilities
and experimental apparatuses available to meet their needs.
This collaboration is done under the auspices of the Nuclear
Engineering Research Collaboration Center (NERCC),
established in 2005 and supported mainly by the Nuclear Scienceand Engineering Directorate, as shown in Fig.10-1. NERCC
provides a platform for joint research and development ofinnovative nuclear technology. The platform manages specifictechnology groups established in collaboration agreements with
university and industry partners interested in collaboration,
cooperation, technology transfer, etc. Four technology groups,
organized by researchers and experts of universities, industries,
and JAEA, are presently active.
The Advanced Reprocessing Materials Development Group
has succeeded through cooperation with Kobe Steel, Ltd.
in making original extra high purity (EHP) alloys by cold
crucible induction melting (CCIM) for Ca/CaF reduction andthen electron beam cold hearth refining (EB-CHR). The EHP
alloys have demonstrated excellent resistance against grainboundary attack in a severe reprocessing environment andgood weldability, as reported in Topic 10-1. The trial meltingof several hundred kg has determined the necessary conditions
to limit harmful impurities entering the EHP alloys to less than
100 ppm, even if nuclear facility grade scrap is used as raw
materials.
The Special Group for Advanced Technologies for
LWR Thermal-hydraulics is conducting mock-up tests incollaboration with an industry partner to attain improved
86

JAEA R&D Review 2008

information management. The Special Group for Developing
Hypersensitive NDA Method for U&Pu has invented a fastneutron direct interrogation method, by which distributionsof U and Pu in the interior of the waste matrix can bemeasured with more than 100 times higher sensitivity thanby conventional methods. The group is cooperating with IHI
Ltd. in R&D to make this method practicable by improving
detection accuracy and sensitivity. Additional ongoing R&D
collaboration with IHI Ltd. and Tokyo University aims atapplying the method to detect nuclear material hidden inpersonal luggage.
The Graphite & Carbon Materials Characterization Special
Group conducts material development for HTGRs and isdeveloping an irradiation material database and evaluation
method in cooperation with Toyo Tanso Co. A goal is to extend
the lifetime of in-core materials for the VHTR. 3D- X-ray
CT of fine-grained graphite of 20 μm grain size with highmagnification (Fig.10-2) was achieved. This image technique
can be used for irradiated graphite evaluation.

Fig.10-2
X-ray CT image for IG-110
graphiteBlack:pore

Gray:crystal

Collaboration on Basic Nuclear Engineering

10-1 Development of Extra High Purity Stainless Steel (EHP Alloy)
with High Corrosion Resistance in Severe Environment

To remove non-volatile
impurities

Melting metal
Cold crucible

Raw materials(Fe,Ni,Cr, etc㧕Induction coilSkull

1st refining
CCIM㨺Ca/CaF

To remove volatile
impurities

Corrosion rate (g/m2/h)–Excellent Grain Boundary Corrosion Resistance of EHP Alloy under Boiling Nitric Acid–

SUS310ULCSUS310 EHP alloy

Coolant

Electron beam

Corrosion time (h)SUS310ULC
CCIM
ingot

2nd refining
EB-CHRSUS310 EHP alloy

water-cooled
copper hearth
EHP ingot

water-cooled
copper rectangularmold

Grain boundary attack

Fig.10-3 Multiple refining technology for EHP alloy
EHP alloy with harmful impurities limited to less than 100ppm
was produced by a cold crucible induction melting (CCIM)
for Ca/CaF reduction and an electron beam cold hearth
refining (EB-CHR).

Fig.10-4 Results of the corrosion test in the boiling
nitric acid
SUS310 EHP alloy showed excellent resistance against
grain boundary corrosion compared with conventional
SUS310ULC.

The development of high performance next generation
nuclear energy systems is expected to place priority on global
warming and effective utilization of resources. In these systems,
fuels of high burnup light water reactors and of fast breeder
reactors must be stored. Therefore, it is necessary to develop
the stainless steels with higher corrosion resistance more thanthe current materials for reprocessing plants.
The life of stainless steels for reprocessing plants mainly
depends on resistance to grain boundary corrosion. Grainboundary corrosion is local corrosion which starts at inclusions
or at grain boundaries. We carried out the fundamental researchof an Extra High Purity alloy (EHP alloy) in which impurities
are reduced to the lowest practical level. For practical use of
EHP alloy, the development of commercial refining technology
with high quality and low cost is necessary. Thus, we aredeveloping new EHP alloy multiple refining technology andevaluation jointly with Kobe Steel, Ltd. The technology
comprises cold crucible induction melting (CCIM) for Ca/CaF
reduction and electron beam cold hearth refining (EB-CHR)
as shown in Fig.10-3. Non-volatile impurities are removed by
CCIM and then volatile impurities are removed by EB-CHR.
Moreover, an intermediate product with rectangular crosssection can be directly produced using a water-cooled copper

rectangular mold. In trial melting of several hundred kg, EHP
alloy whose total harmful impurities were less than 100 ppm
was produced even if nuclear facility grade scrap was used as
raw materials.
Fig.10-4 shows results of a corrosion test in boiling nitricacid which simulated the environment of a reprocessing plant.
The specimens tested were the conventional SUS310ULC(Ultra Low Carbon) and the present SUS310 EHP alloy.
Grain boundary corrosion was observed on the surface of
SUS310ULC, causing the corrosion rate to be accelerated
greatly. On the other hand, the corrosion rate of SUS310
EHP alloy was approximately constant and showed excellent
resistance against grain boundary corrosion. It was confirmedthat EHP alloy is unaffected by boiling nitric acid. As the weld
cracking resistance of EHP alloy was improved by removalof harmful impurities, EHP alloy can be used as a welding
material.
The study was carried out as part of the “Research andDevelopment of Nitric Acid Resistant Material Technology
Applicable to the Next Generation of Reprocessing Equipment”
entrusted by the Ministry of Education, Culture Sports, Scienceand Technology of Japan.

Reference
Ioka, I. et al., Susceptibility of Intergranular Corrosion for Extra High Purity Austenitic Stainless Steel in Nitric Acid, Proceedings of 16th International
Conference on Nuclear Engineering (ICONE16), Orlando, Florida, USA, 2008, ICONE16-48776, 5p., in CD-ROM.
JAEA R&D Review 200887

11 Photo-Medical Research Cooperation
Aiming at Creating a Photo-Medical Industrial Valley
Lasers in Medicine
Roles of 10 Partner
Steering committee
Compact cancer
diagnosis &therapeutic instrument

Clinical trial of cancer
therapeutic instrument
Hyogo Ion Beam Medical Center (HIBMC)

Clinical
Irradiation
plan

Development of newirradiation monitor method
for medical treatment
application
Shimadzu Corporation

Systemdevelopment

Development of minimallyinvasive laser treatment
deviceHOYA Corporation

Development of laser
driven irradiation device
Toshiba Corporation

Industrial
application

Development of laser
system for medicine
Hamamatsu Photonics
Ushio Inc.

Fujikura Ltd.

Industrial
application

Diagnosis
technology
development

Industrial application &development of Photo
medical treatment system
Toyota Central R&D Labs.,Inc.

Laser
development

This project aims to develop a much more compact cancer
diagnosis and treatment instrument using ion beams driven andcontrolled by high intensity laser technology.
JAEA has successfully proposed this novel project to the
Ministry of Education, Culture, Sports, Science and Technology(MEXT), seeking the creation of a “photo-medical industrial
valley” base in FY 2007.
In October 2007 we established the Photo-Medical ResearchCenter (PMRC) as a chief directorate and central control “hub”
for developing a network for cooperation with industry anduniversities, and to promote this new endeavour.
With the possibility of compact laser-driven treatment
facilities being built within the next 10-15 years, we envisionthat the world will have much greater access to hadron cancer
therapy, for which the global demand is continuing to grow.
Promotion of PMRC is fundamentally based on thedevelopment of industrial laser technologies for ion accelerationand companion development of related applications such as
endoscopy.
The cooperative organization with ten partners and theirroles are shown in Fig.11-1. The innovations for which thecreation of “Photo-medical industrial valley” is intended areshown in Fig.11-2.

Construction andoperation of innovative
management system
Nippon Advanced
Supporting Laser
Technology Co.,Ltd.
Technology

We are collaboratingwith industry indeveloping laser
acceleration of ionsand other applications
such as endoscopes.

HOC

Fig.11-1 PMRC Partners and Their Roles

Innovations after 10〜15 years“Laser acceleration”

・Laser-driven particle
acceleration(JAEA)

Ⅰ ．Size Reduction of Accelerator: Compact
Laser-Driven Particle Therapy
Ⅰ
Ⅰ ．Treatment during diagnosis
Ⅰ
Ⅰ
Ⅰ．Spot-Scanning Therapy Technology
ⅠⅤ ．Personnel Training
Ⅴ ．Create New High-technology Industries

Compact Laser
Acceleration for
Therapy“Ion beam therapy”

・Cancer treatment(HIBMC㧕‫ޓޓ‬・PET diagnosis(Shimadzu Corporation)

Treatment during
diagnosis

Ion beam therapy
anytime and anywhere

ProtonsourceLaser
PETtransport

combined〜5m
Laser system

Patient

ControlsystemTreatment plan
Compact Laser-Driven Particle Therapeutic Instrument

Fig.11-2 Concept of the Desired “Photo-Medical Industrial Valley”

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JAEA R&D Review 2008

Photo-Medical Research Cooperation

11-1 Ion Acceleration from Near-Critical Density Plasma
Produced by Lasers–Proposal of a Novel Ion Acceleration Method for Cancer Therapy–

Electrons

10cm

Magnetic
field

Ion Beam
Magnetic
field

Main pulse

di
re
ct

io
n

Solid-foil target

Electrons

La

se

r

ASE

Laser

Near-critical
Density plasma

TargetNumber of protons (1/MeV/msr)

Fig.11-3 Laser ion accelerator
A laser pulse is focused onto a thin-foiltarget with intensity exceeding 1018 W/cm2,
which generates high-density plasma at
the focal spot. This plasma produces a
MeV-scale proton beam.

Fig.11-4 A novel mechanism of ion
acceleration
A laser pulse can invade into a nearcritical density plasma target, and thelaser energy is efficiently converted into
electron kinetic energy.

Fig.11-5 Near-critical density
plasma production method
A solid foil target is converted
into near-critical density plasma
by the amplified spontaneousemission (ASE) light preceding
the main laser pulse.

Maximum energy: 3.8MeV
Number of protons: 1010(E>1MeV)

Fig.11-6 A proton energy spectrum
We have successfully obtained protons of
3.8 MeV maximum energy.

Energy (MeV)

When a laser pulse is focused onto a thin-foil target withintensity exceeding 1018 W/cm2, high-intensity plasma is
induced at the laser focal spot, generating several kinds ofradiation including high-energy ions. By using this phenomenonof laser-ion acceleration, conventional ion accelerators, thathave lengths of tens of meters, can be replaced by a compact
apparatus as shown in Fig.11-3. The Photo-Medical ResearchCentre (PMRC) aims to develop a compact and novel laserdriven ion accelerator for cancer therapy.
Recently, we proposed and demonstrated experimentally a
new method of laser-ion acceleration by which laser pulse energy
can be converted more efficiently into ion kinetic energy. The
method is based on creation of a near-critical density plasma
target. In the case of a usual solid target, the target surface canreflect a large part of the laser pulse energy. By contrast, a laser
pulse can penetrate into a near-critical density plasma target(Fig.11-4), where laser energy is efficiently converted into
electron kinetic energy. The vortex of moving electrons thus
created generates a magnetic field on the target rear surface
and sustains an electrostatic ion-accelerating field. A solid foil
target is converted into near-critical density plasma by the partof the light pulse that precedes the main high intensity laser,
amplified spontaneous emission (ASE) (Fig.11-5).
In experiments we have generated protons with maximum
energy of 3.8 MeV. Fig.11-6 shows the proton energy spectrum,
which also shows that approximately 1010 protons are generated
by a single laser shot. Higher-energy protons will be achievedby optimizing the target density through control of the ASE
intensity and duration.
This work is supported by Special Coordination Funds
for Promoting Science (SCF) provided by the Ministry of
Education, Culture, Sports, Science and Technology of Japan.

Reference
Yogo, A. et al., Laser Ion Acceleration via Control of the Near-Critical Density Target, Physical Review E, vol.77, no.1-2, 2008, p.016401-1-016401-6.

JAEA R&D Review 200889

Photo-Medical Research Cooperation

11-2 One Optical Fiber Can Both Observe and Treat–Development of Extra-Fine Endoscope with Laser Radiation
Function–
Umbilical
cord B

Umbilical
cord A

Amniocentesis(symptomatic treatment)

Laser source Coupling device
TV monitor

Fetal endoscope

Excess blood

Placenta
Vascularcauterization bylaser irradiation(radical treatment)

Vascularanastomoses

PC monitor

Umbilical
cord

Mono-villus,
diamniotic twins
Approximately 45 to 50%
of all Japanese twins areof the mono-villus,
biamniotic type, of whom
10 to 15% may develop
TTTS as a complication.

UterusPlacenta
Excessamniotic
fluid

Insufficient blood,
Insufficientamniotic fluid

Treatment of TTTS: Increases the likelihood of both fetuses
being delivered normally

Light source
for illumination

Anastomotic vessels
in the placenta cut
using a laser

Fig.11-7 Example of congenital disease: Twin-to-Twin Transfusion
Syndrome (TTTS)
In TTTS, the blood vessels to twins become connected inside theplacenta, resulting in one twin receiving less blood than the other. If
left untreated, there is a 70% or greater probability that one or both
fetuses will die. A radical treatment is vascular cauterization method bylaser irradiation (FLP).
The laser can be targetedat the center of the image

Abdominal side
Placenta

Blood vessel (simulation)

UterusBack sidePC

Composite-type
optical fiberscope

Fig.11-8 Extra-fine endoscope having a laser
radiation function
We produced a trial system which performs laser
radiation and observation at the same time by
one optical fiberscope with an outside diameterof 2 mm.

Fig.11-9 Technology that enables placenta
treatment at all positions
This device enables treatment even if theplacenta is in front.

Bent-type external pipe
(containing composite-type optical fiberscope)

Twin to twin transfusion syndrome (popular name TTTS) is
one of the congenital diseases that twins develop in utero. Thisdisease that only occurs in cases consisting of two amnions andone placenta (mono-villus diamniotic twins) requires highly
advanced medical treatment. In the case of normal monovillus diamniotic-twins, each blood vessel is connected to theplacenta and the balance of blood circulating through each isgood. When blood flow between twins becomes unbalanced,
and blood continues flowing to only one fetus, TTTS develops.
A laser treatment method while observing with an endoscope(popular name FLP) recently is attracting attention as a radical
therapy of TTTS, (Fig.11-7).
However, experience in FLP treatment is superficial
nationwide, and the reality is that there are only a few hospitals
nationwide that can perform FLP treatment. In addition, there
are the following problems because this treatment is intended
for a fetus floating in the uterus (in amniotic fluid) which isa small space; (1) the definite vascular configuration (a map)of connection to the placenta is not clear, (2) the distance
between the tip of the optical fiber and a blood vessel is notclear, (3) quantitative determination whether the bloodstream
is stopped or not is not possible, (4) the endoscope image isnarrow, and (5) it is hard if the placenta is in front of the fetuses(placenta previa). Advanced technical progress and a highly

skilled doctor are necessary to do such an operation with theconventional endoscope.
Therefore to make a contribution to FLP treatment, byutilizing a composite-type optical fiber by which observationand laser irradiation are possible in one optical fiber at thesame time, we produced a device which promises to solve theproblems with the conventional endoscope (Fig.11-8). We notethat this device was created as a spin-off from development ofa robot for inspecting and maintaining nuclear fusion reactors.
In this system, the middle of an image can be laser-irradiated,
and there are functions to measure the distance to an object
and blood quantity/flow speed. This allows laser radiation to
be applied precisely, safely and comparatively easily. This
instrument has potential for prenatal surgical treatment of
the congenital disease mentioned above. In addition, thismakes laser radiation easy even in the case of placenta previa(Fig.11-9). Furthermore, because the diameter of the hole(possibly through the abdomen) can be reduced by reducing
the endoscopic external diameter to as little as 2 mm, we can
expect that the burden to a mother's body will be reduced
greatly (Minimally Invasive Treatment). The efficacy of thissystem has already been confirmed in animal experiments.
We are proceeding with development for clinical applications
now.

Reference
Oka, K., Development of Extra-fine Endoscope Having a Laser Radiation Function -We Apply It to Prenatal Treatment-, Energy Review, vol.27, no.7, 2007,
p.7-10 (in Japanese).

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JAEA R&D Review 2008

12 Computational Science and E-Systems Research
Atomic Energy Research by Computational Science–Establishment of Advanced Technology for Atomic Energy R&D–

Nuclear Reactor
Nuclear Fusion

High Intensity Proton Accelerator

Atomic Energy R&D

R&D forComputer Science
Seismic Analysis

R&D forAdvanced Simulation
Technologies

3-in-1 Function

Radiation Influence

Operation & Maintenance
Grid Platform R&D

Nano-Device

Fig.12-1 Computational Science in Atomic Energy Related Fields
The CCSE is carrying out three-fold coordinated R&D in advanced computational science: for computer science, for advanced
simulations, and for better operation/maintenance of computer systems.

For the increased safety of atomic energy, we are carryingout R&D of computational science which could be used withtheoretical and experimental advances to predict the safety
state of an entire nuclear facility. Computational science is athird method along with “theory” and “experiment” to pursue
R&D, and is currently expanding to new fields such as global
warming predictions. Further development of computational
science is expected also in the field of atomic energy.
The Center for Computational Science and e-Systems(CCSE) promotes three missions, “R&D for computer science
to allow computational scientific research in the atomic energy
field and to promote cooperative research”, “R&D for advanced
simulation technologies to make predictions with higher
accuracy” and “operation & maintenance of computer systems
allowing active research with the guaranty of information
security” as depicted in Fig.12-1.
Presently, we are tackling the following seven topics in a
mid-term plan to promote innovative atomic energy researchand to drive domestic R&D in the computational science field;“distributed parallel computing through grid technology”,“vibration simulator of an entire nuclear facility”, “mechanismsof crack-propagation in nuclear reactor materials and of rimstructure formation in UO2 nuclear fuel”, “nano-scale device
development for atomic energy”, “multi-scale modelingincorporating micro to macro level simulations”, “development
of databases for genome analysis” and “tools for studying
DNA/RNA repairing proteins”.
CCSE has actively participated in various national projects. For
instance, we participated in an ITBL (Information Technology
Based Laboratory) project from FY2001 to FY2005, which was
one of the e-Japan priority policy programs. In this project, we
collaboratively constructed a virtual supercomputer consistingof distributed supercomputers. ITBL shares computer capacityof 60 TFLOPS (from 17 supercomputers) with 11 researchcommunities in 64 research institutes and is continuously
supporting their research activities. We also contributed to the“Development & Application of Advanced High-PerformanceSupercomputer” project through computational science
technologies cultivated in the atomic energy field.
We are also promoting international cooperation with theUnited States (2institutes), Germany (3) and France (5) in orderto share in worldwide leading edge research on computational
science for the atomic energy. Simulation and modelingtechnology has been selected as a one area of the Global
Nuclear Energy Partnership (GNEP) activities of Japan and theUnited States, and we are cooperating with this partnership.
CCSE will keep accelerating computational science R&D in
the field of atomic energy.
JAEA R&D Review 200891

Computational Science and E-Systems Research

12-1 Computational Nuclear Power Plant Modeling andStructural Integrity Assessment–3-D Vibration Simulator for Predicting Dynamic Behavior of Nuclear Power Plants–
Level 2
Piping system 1

Piping systems
Boundary
conditions

Equipment system 3
Piping system 2

Equipment systems

Equipment
system 1

Responses

Seismic wavestLevel 1

Grid-computing environment

Fig.12-2 Hierarchization in assembly structural analysis
by finite element method
Analysis using the conventional finite element method has acomputational limit in terms of data capacity when modelingan object because the method treats the object as acontinuum structure. We propose a procedure that mitigatesthis problem by dividing the object into several parts. This
figure shows an example of how hierarchization is used todecrease the scale of analysis by dividing the entire system
into two levels. Specifically vibrations of the equipment
system and piping system are handled separately, and theinteraction between these data sets is considered.

Fig.12-3 Vibration analysis of entire nuclear power plantby hierarchization
Vibration responses of three level-1 equipment systems(equipment systems 1, 2, 3) were calculated and then, theresponses at the boundary between the equipment andpiping systems were input to two level-2 piping systems(piping systems 1, 2) as boundary conditions. The vibration
responses of the two piping systems were calculated torealize vibration analysis of the entire nuclear plant.

There is much concern with validation of the structural
integrity of nuclear power plants, especially in earthquakeprone nations, due to the recent occurrence of several large
earthquakes. In conventional structural design, the equipment
and building constituting a plant have been modeled
individually in comparatively simple approximations based on
past experience and testing. It is, however, required to treat theplant in its entirety and in more detail in order to analyze thestructural integrity of the plant consisting of many parts.
The CCSE of JAEA has been conducting studies using athree-dimensional vibration simulation technique to evaluate
the structural integrity of nuclear power plants using the latest
computational techniques. In the Hanshin-Awaji Earthquake,a recent large-scale earthquake, physical phenomena such as
local deformations caused by the interaction between parts wereobserved which cannot be easily understood by conventionalpart analysis techniques. We are trying to understand theseinteractions between parts by using a three-dimensional
vibration simulation technique to model and study the entire
structure.
Thus far, we have developed a data processing techniquecalled assembly structural analysis method for plants,
comprising more than 10 million parts. We have also developed
a procedure to model an entire structure using a grid computing
approach, because a single supercomputer cannot process
the huge data required for modeling. An assembly structural
analysis method can be used to generate model data for eachpart and assemble them in a way similar to how real structures
are manufactured into a finished structure; this method is
expected to solve the problem presented by huge data. We havedeveloped a prototype of a vibration simulation system bycombining these techniques, and began numerical experiments
to analyze the structure of the main facilities of the “High
Temperature Engineering Test Reactor (HTTR)” at the Oarai
Research and Development Center of JAEA. As a first step, we
succeeded in performing a static analysis of HTTR.
It is necessary to perform iterative calculations that consider
sequential state changes of extremely large components
in order to realize vibration analysis of an entire nuclear
power plant; this has proved difficult using conventional
analytical techniques due to the scale of the problem andthe computational time required. Therefore, we proposed ahierarchized assembling structural analysis method to mitigate
the problems. In hierarchization, components are classified
suitably into units, for example into those serving a common
function, and these units are first processed separately. Fig.12-2
shows an example of a nuclear plant system that is classified
into two levels, equipment systems and piping systems; theplant is analyzed by considering the interaction between these
two levels.
The work of calculation of the structural model withapproximately 180 million degrees of freedom was distributed
optimally among several supercomputers and was simplified
by hierarchizing the input data, as shown in Fig.12-3. We
succeeded in conducting a vibration analysis of HTTR, although
the model did not include inner structures, using a prototype ofa three-dimensional vibration simulation system. Comparedwith conventional analysis, the data size and processing time
were reduced to less than one-third and one-tenth respectively.
This result increases the possibility of performing detailed
vibration analysis of an entire nuclear power plant having morethan 1 billion degrees of freedom.

Equipment system 2

Reference
Nishida, A. et al., Numerical Simulation System “Three-Dimensional Virtual Plant Vibration Simulator” for Nuclear Plants by Using Assembled Structural
Analysis, Nippon Genshiryoku Gakkai Wabun Ronbunshi, vol.6, no.3, 2007, p.376-382 (in Japanese).

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JAEA R&D Review 2008

Computational Science and E-Systems Research

12-2 Suppressing Bubble-Induced Material Damage by Gas Bubbles

without Injection
with Injection

Pressure(MPa)

Bubble radii(Ǵm)–Numerical Study of the Damage Suppression Effect of Bubble
Injection–

Explosive expansion

Gentle oscillation

Time(ms)

Fig.12-4 Suppressing bubbles’ destructiveness by injecting
gas bubbles
When subjected to a sudden pressure change, liquid mercury
is ruptured and tiny bubbles emerge (the green sphere in(a)). These bubbles will expand (b) and collapse violently,
causing significant material damage. Bubble injection is atechnique to suppress this violent motion by injecting gas
bubbles (the blue spheres in (c)) into liquid mercury.

Fig.12-5 Bubble radii (d) and pressure (e) in mercury
without (blue) and with (red) injected bubbles asfunctions of time
This numerical result reveals that the injected bubbles
suppress the violent motion of the emerged bubble andreduce the negative pressure in liquid mercury. This is closely
consistent with our experimental findings, and elucidates thefundamental mechanism of damage suppression.

We are now developing an intense pulsed neutron source
using a MW-class proton accelerator and liquid mercury as
part of the J-PARC project. Using this facility, leading-edge
research on, e.g., high-T superconductors and DNA will be
performed in the near future.
In this “Big Science” facility, tiny bubbles, smaller than 1 mm,
are posing a serious problem. The facility produces neutron
beams by breaking mercury nuclei with energetic protons,
at which moment a large amount of energy is released anda sudden pressure change occurs in mercury. The pressurechange ruptures the mercury and there many bubbles emerge.
The bubbles are so violent, causing significant material damageon the metal vessel. To suppress this damage that reduces the
lifetime of the facility and limits the output energy, we are now
developing several techniques.
One of them is “Bubble Injection,” in which gas bubbles
are injected into flowing liquid mercury (Fig.12-4). To realize
this technique, several issues must be resolved such as thedevelopment of an injection method (Topic 14-5) and clarifying
the physical effect of the injected bubbles (this article). Our
recent investigations have verified the effectiveness of thistechnique. Fig.12-5 shows the dynamics of a bubble which has
emerged, calculated numerically using a theoretical model thattakes the interaction with the injected bubbles into account.
With no bubble injection (the blue curve), the emerged bubble
expands explosively and then collapses very violently: such an
aggressive bubble is certain to cause material damage. With
bubble injection, on the other hand, the bubble becomes very
gentle, as shown by the red curve.
From this numerical result we found that the change in bubble
dynamics was caused by positive pressure waves emitted by
the injected bubbles, which reduce the negative pressure inmercury (the lower panel of Fig.12-5) that drives the bubble
expansion. This is the hidden function of bubble injection.
This work was partly supported by a Grant-in-Aid for Young
Scientists (B) (No. 17760151).

References
Ida, M. et al., Direct Observation and Theoretical Study of Cavitation Bubbles in Liquid Mercury, Physical Review E, vol.75, no.4, 2007,
p.046304-1-046304-7.
Ida, M. et al., Suppression of Cavitation Inception by Gas Bubble Injection: A Numerical Study Focusing on Bubble-bubble Interaction, Physical Review E,
vol.76, no.4, 2007, p.046309-1-046309-10.
JAEA R&D Review 200893

Computational Science and E-Systems Research

12-3 Large-Scale Conformational Change of a Protein
Elucidated by Computer Simulation–Molecular Mechanism of Cell Adhesion Protein Integrin–

Asp550

Fig.12-6 (a) and (b) X-ray crystal structure of integrin αVβ3; (c) computer model of the extended conformation(a) 3D structure of integrin solved by X-ray crystal structure analysis: ball model of the molecule when “bowing down”.
(b) A Magnified tube model of part of the integrin structure. Important amino-acid residues are shown by ball model andtheir names are shown.
(c) Extended conformation of integrin obtained by computer simulation of large-scale conformational change.

Cells in multicellular organisms such as humans are usually
not in isolation, but they are bound to each other directly or
indirectly through an extracellular matrix. In this binding, celladhesion proteins, which exist on the cell surface, are involved.
Integrin is a type of cell adhesion protein, and a variety ofintegrins with different functions and structures have been
identified.
Some kinds of integrins are known to be normally in an
inactive state changing to an active state only when necessary.
Integrin αIIbβ3, which exists abundantly on the cell surface
of a blood platelet, is one of them. When we bleed, this integrinchanges to an active state to promote the platelet aggregationand stop the bleeding. Large-scale conformational changes ofintegrin (Fig.12-6 (a),(c)) are considered to be involved in thisactivation.
We made computer simulations to study the mechanism of
the large-scale conformational change of integrin αVβ3,
which is similar to integrin αIIbβ3 and whose structure wasdetermined by X-ray crystal structure analysis at the atomic

level. Among ~1500 amino-acid residues which constitute theprotein, we found that two amino-acid residues play importantroles in the conformational change (Arg633 and Arg404 in
Fig.12-6 (b)). That is, we found that the conformational change
occurs more easily if the interactions of these amino-acidresidues with nearby residues are eliminated in the simulations.
This computational result suggests that these amino-acidresidues work as switches for the conformational change. The
residues had not been recognized as important, but experiments
performed by our co-worker demonstrated that the mutationof these residues activates integrin, thereby verifying oursimulation results.
Heart attacks and strokes, which are the major causes of deathfor Japanese people, are caused by the formation of the bloodclot at undesirable locations in the body. For the developmentof a cure for these deceases, many researchers are focusing on
integrin, which is involved in the formation of the blood clot.
Our discovery about this protein will be helpful in advancing
this development.

Reference
Matsumoto, A. et al., Key Interactions in Integrin Ectodomain Responsible for Global Conformational Change Detected by Elastic Network Normal Mode
Analysis, Biophysical Journal, vol.95, no.6, 2008, p.2895-2908.

94

JAEA R&D Review 2008

Computational Science and E-Systems Research

12-4 Development of Steel Impurity Behavior Simulation
Technology–Estimation of Impurities in Materials Based on Thermal Desorption Gas–
Simulations

Reduction cross section

Specimen temperature (K)Hydrogen evolution rate (mppm/min)Hydrogen evolution rate (mppm/min)Experiments

Reduction cross section

Specimen temperature (K)Fig.12-7 Comparison of the hydrogen desorption profile of deformed eutectoid steel: (a) Experiment results (b)Simulation results
The numerical model simulates one peak at the low temperature side, which may relate to hydrogen desorption from defects ofvacancy or dislocation, and the other peak at the high temperature side, which may be due to an increase in phase interfaces
caused by the cold drawing.

Steel is a basic structural material which is used for various
constructions. Steel material inevitably includes impurities anddefects disrupting the crystal structure. which are vacancies,
dislocations, grain boundaries, phase interfaces, and inclusions.
Hydrogen, the smallest and lightest element, also enters intomaterials from the environment as impurities. Furthermorehydrogen seriously changes the mechanical properties ofmaterials by migrating and combining with defects. The changeof mechanical properties causes the degradation of material
strength. Therefore, the estimation of the hydrogen and defects
within materials is an important issue.
Defects in the steel lattice structure are called trapping sites
because they capture hydrogen migrating from the defect-freecrystalline regions. Type of defects can be distinguished by theirdiffering energy for capture of hydrogen. Defects with deepenergy valleys trap hydrogen strongly and defects with shallow
energy valleys trap hydrogen weakly. Thus, the former defects
release hydrogen at higher temperature, and the latter defects
release at lower temperature. Inferring from this property, if
there is a peak of desorbed hydrogen from a specimen heatedat a constant rate, one can regard the hydrogen of the peak as
being released from the defects with a specific trapping energy.
Therefore, we can determine the type and amount of defectin the specimen, the trapping energy of the types of defects,
and the amount of trapped hydrogen by analyzing the profile
representing the relation between specimen temperatures anddesorbed hydrogen. This analysis is called thermal desorption

analysis. In order to estimate these quantities and the relation
between them, we developed a numerical model of thehydrogen desorption process and tried to simulate the hydrogen
desorption profile obtained by the thermal desorption analysis.
Our model computes the equilibrium state between the hydrogenin trapping sites and the hydrogen in surrounding regions at
each temperature, and regards hydrogen in the surrounding
regions as desorbed hydrogen that produces the profile, though
in practice hydrogen in the surrounding regions diffuses withinmaterials. This model of the hydrogen desorption process isbased the assumption that hydrogen migration is very fast in
the defect-free crystalline regions.
As shown in Fig.12-7, our numerical model simulates very
well the actual hydrogen desorption profile of eutectoid steel,which indicates several types of defects distinguished bydifferent trapping energy. This successfully simulated profile
has two peaks, which was difficult for previous models to
reproduce, and allow us to estimate the amount of defects
and trapped hydrogen and the trapping energy of the defectsof each peak. This simulation also may suggest experimental
conditions which will isolate the peak of a specific defect.
This idea and technique can also be applied to the estimationof helium in steel materials. The estimation of the amountof defects and trapped impurities, and the trapping energyof defects should be of use for research into the mechanicalproperty change that causes embrittlement or hardening of
nuclear steel materials.

Reference
Ebihara, K. et al., Modeling of Hydrogen Thermal Desorption Profile of Pure Iron and Eutectoid Steel, ISIJ International, vol.47, no.8, 2007, p.1131-1140.

JAEA R&D Review 200895

13 Scientific & Technical Development for Nuclear Nonproliferation
Development of Technology for Nuclear Nonproliferation,
Foundation of Peaceful Use of Nuclear Energy
We have two primary missions regarding nuclear
nonproliferation. One mission is to support the government
in developing nonproliferation-related policies throughresearch and study. The other mission is to support government
and international organizations by developing nuclear
nonproliferation technology. Other important missions of
JAEA are to support denuclearization, nuclear material control
by the facility using it, and human resource development.
Policy Research and Study“Modelling the nonproliferation process in Japan”, and“Improving Trustworthiness and Transparency of Peaceful
Use of Nuclear Energy in Asia”, based on in-house technicalknowledge.
In 2007 we comprehensively reviewed Japan's current
nuclear nonproliferation efforts with regard to items suchas guarantee of peaceful nuclear use by domestic laws andregulations, ensuring transparency of plutonium use, export
control, sensitive technology control, physical protection /nuclear
security, and safeguards measures, and also considered
concrete cooperative arrangements with the countries inSoutheast Asia where introduction of nuclear power generation
is likely. Moreover, in support of the Japanese government, weinvestigated fuel supply guarantees and nuclear nonproliferation
strengthening, which are international issues.
Development of technology for nuclear nonproliferation
Development of an advanced safeguards system which will
provide effective and efficient safeguards for the future Fast
Breeder Reactor fuel cycle is underway. In collaboration with
the U.S. as part of GNEP, cooperative research plans for next
generation safeguards measures/physical protection have been
discussed at a working group level.
Regarding environmental sampling for safeguards, we are
performing further development of the fission track-thermal
ionization mass spectroscopy method with which a uranium
particle less than 1μm can be detected.
We have started demonstration tests of a new color surveillancecamera with high resolution, as a technical development for
improvement of the reliability of, and confidence in, peacefuluse of nuclear energy.
We also contribute to international activities such the
Generation IV Atomic Energy International Forum, and weare performing research on evaluation methodology for the
proliferation resistant features of future nuclear cycle systems.
Support of Denuclearization
We have been developing technology with contributes todisarmament and denuclearization of the world, in particular
cooperative research with Russia to use surplus nuclear weapons
plutonium as MOX vipac fuels in fast breeder reactors. The
effectiveness of this method was recognized by both the U.S.
and the Russian Federation, and both countries issued a jointstatement that this application of plutonium in the fast breeder
reactor BN600 has been approved.
In support of the CTBT (Comprehensive Nuclear-Test-Ban
Treaty), at Tokai laboratory we continue to perform precise
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JAEA R&D Review 2008

analysis of environmental samples collected at radionuclide
monitoring stations throughout the world.
We have started provisional operation of the Takasaki
and Okinawa monitoring stations after certification by the
CTBT Organization. Operation of those stations is treatedas provisional operation until CTBT enters into force, but
it actually has reached the stage of full-scale operation withcompletion of all planned R&D.
At the National Data Center located in the Nuclear Science
Research Institute in Tokai, we continued development
and upgrade of software for γ-ray spectrometry, β-γcoincidence data analysis for radioxenon, a database system,and source location estimation of radionuclide release using an
atmospheric transport model.
Nuclear Material Management
We have been making technical contributions to IAEA
safeguards implementation by conducting several technicaldevelopment projects in cooperation with the U.S. Departmentof Energy. We have also cooperated in domestic andinternational personnel training.
From the results of our long-term environmental durability
tests of intruder automatic detection systems performed
in order to strengthen physical protection, we identified
causes of problems arising upon long term outdoor use andcountermeasures therefore. We have verified that this is aneffective system which can quickly detect and automaticallymonitor suspicious persons, if there is proper installation
position of surveillance cameras and setup of effective detection
rules, etc.
We have examined the medium-to-long term nuclear
materials transportation plans of each project of JAEA for
possible problems. Also, we have started negotiation with U.S.
DOE on a contract for spent fuel transportation from research
reactors. Moreover, in our development of a MOX powdertransportation container, we carried out handling tests at a
facility and heat transfer tests using a prototype container.
Policy Research and Study“Modelling the nonproliferation process in Japan”“Improving of Transparency and Confidence Building for Peaceful
Use of Nuclear Energy in Asia”

Technical Development relating to Nuclear Nonproliferation◆Safeguards Concept◆Analytical Technology of Environmental Samples◆Technology for Higher Nuclear Transparency◆Development of Nuclear Proliferation Resistance Assessment Method

Support of Denuclearization

Nuclear Material Management

Support of Use of surplus
nuclear weapons plutonium
Operation of CTBT Monitoring
Stations, Certificated Laboratory,
National Data Center

Safeguards Tech.
Physical Protection Tech.
Design & R&D of Transportation
Container

Fig.13-1 Research Activities of Nuclear Nonproliferation
Science Technology

Scientific & Technical Development for Nuclear Nonproliferation

13-1 Nuclear Proliferation-Resistance and Safeguards
for Future Nuclear Fuel Cycle–Study of the Evaluation Technique and Criteria of Nuclear Proliferation Resistance–
Clandestine Nuclear Weapon Program
Detection
Program

Fail

Detection
Activities

Fail

Clandestine Facility
Fail

Fail

Detection
Activities

Diversionof NM

Civilian Reprocessing Plant
Detection by
DIV

Fail

Detection by
Process Monitoring

Fail

Detection by
NRTA,C/S

Fail

Detection by
SNRI

Fail

Detection by
PIV, IIV

Fail

Intrinsic Barriers(Technical Difficulty)

Fail

Virtual Design

Acquisition of Nuclear Weapon

Frequent IIV(PIV-level IIV)

Incorporation into Design
Laboratory,
Engineering
Scale

Qualitative Process
Control based onMaterial Accountancy
Small Process Inventory,
Semi-metric Vessel

Real Time
Process Monitoring

Demonstration
Scale Plant

Unattended Inspections
by Verification System,
Verification of SSAC
Analytical Lab.
MUF Simulation

Advanced Material
Accountancy System

Evaluation

Intrinsic
Barriers(Technical Difficulty)

Advanced Safeguards System(High Proliferation Resistance)

Advanced Nuclear
Fuel Cycle Facilities

Incorporation ofIntrinsic Barriers(Technical Difficulty)

Transparency by
Remote andReal Time Process
Monitoring(Solution Monitoring)

Extrinsic Barrier(AP)Extrinsic Barriers(CSA)

Commercial
Plant

Evaluation Methods(Evaluation of Extrinsic & Intrinsic Barriers)

Intrinsic BarriersFig.13-2 Proliferation resistance of future reprocessing
plant (Extrinsic barriers, intrinsic barriers)

Fig.13-3 Incorporation of advanced safeguards system
into design of NFC (Future reprocessing plant)

The demand for peaceful use of nuclear power is remarkably
increasing, in particular in order to ensure global energy
security and prevent global warming. A future nuclear fuelcycle (NFC) (i.e. fast neutron reactor fuel cycle) must recycle amuch larger amount of plutonium than at present, which raises
nuclear security issues (prevention of nuclear proliferation).
In this context, for the future NFC, very robust measures fornuclear security, namely nuclear proliferation-resistance (PR),have to be taken in advance. A proliferation resistant NFC
will impede diversion of nuclear materials by host states into
nuclear explosive devices or other weapons. It is said thatPR should be supported both by intrinsic features of nuclear
energy system and extrinsic measures. International safeguards
are very reliable and effective for early and timely detectionof diversions of nuclear materials, but intrinsic features ofthe future NFC also have to be investigated at the same timeto establish very robust nuclear security. We are developing
reasonable PR measures (intrinsic features and extrinsic
measures) for the future NFC. Here we explain briefly an
example of our study results for a next generation reprocessing
plant, including ways for improvement of technologies for
detection of diversion (safeguards) and deterrence of diversion.
Also, we summarize requirements for the future reprocessing
technologies to have PR.
As proliferation threats, acquisition (diversion) of nuclear
material by states, misuse of NFC facilities, withdrawal from
NPT and theft of nuclear material by terrorist groups are
mentioned. The effects of the extrinsic measures (extrinsic
barriers) and intrinsic features (intrinsic barrier) on thesethreats are shown in Fig.13-2.
Extrinsic measures (IAEA safeguards), based oncomprehensive safeguards agreement (CSA) and additional
protocol (AP), are very effective for timely detection ofdiversion of nuclear material for clandestine weapon programs,
because they establish many checkpoints. This is proved by
the fact that so far no state which has entered into CSA and AP
has attempted a clandestine weapon program. For realizationof the future NFC with very heavy plutonium recycling, it isessential that extrinsic measures maintain high level detection
capability.
To deal with the threat of a state’s withdrawal from NPT ortheft of nuclear material by terrorist groups, intrinsic features(technical & material barriers) are useful. The next generation
reprocessing (advanced aqueous reprocessing) technology
which is proposed in the FaCT project, features the requirementof technically difficult plutonium separation, due to minor
actinide (MA) and fission product (FP) content in plutonium,
making diversion less attractive. These intrinsic barriers arevery effective for preventing a state from withdrawing fromNPT.
We are proposing following requirements those make thefuture NFC technologies have higher proliferation-resistance,
based on the results of our study. These requirements should beconsidered in designing the next generation reprocessing plant(Fig.13-3).
1. Frequent PIV (Physical Inventory Verification)-level IIV(Interim Inventory Verification); Implementation of frequent
IIV close to PIV level in addition to PIV (once a year) in orderto make MUF (Materials Unaccounted For) value low
2. Advanced nuclear material accountancy; Introduction ofsmall process inventory based on semi-metric vessel procedures/ Systemized material measurement / Creating quantitativemodes for process control based on material accountancy
3. Real time process monitoring (Improvement of detection
capability); Real time material accountancy with advanced
solution monitoring / Detection of change of operation
conditions (including acid content)
4. Incorporation of intrinsic barriers; Application of technology
not readily capable of plutonium separation / Treatment
resulting in less attractive plutonium mixed with MA and FP.
In order to establish a system concept of the future NFC which
will be internationally accepted and has economic rationality,
it is very important to combine extrinsic measures (safeguards)
and intrinsic features properly. Also, explaining the concept
to the world is important. For this purpose, it is necessary toparticipate in international non-proliferation activities suchas INPRO and GEN-IV to set up internationally-accepted PR
evaluation methodologies and PR norms.
In our study, so far we have investigated minimum conceptual
requirements of proliferation-resistance. Hereafter, we need toinvestigate them in further detail.

Reference
Senzaki, M., Kuno, Y. et al., Proliferation Resistance of Next Generation Nuclear Energy Systems, Nippon Genshiryoku Gakkaishi, vol.50, no.6, 2008,
p.368-373 (in Japanese).
JAEA R&D Review 200897

14 Development of Experimental Techniques / Facilities at JAEA R&D Centers
We have been promoting a wide range of R&D activities systematically by coordinating between R&D sectors, including the R&D
Directorates whose activities have been described in previous chapters and 12 R&D Centers located in various locations in Japan.
The R&D Directorates of JAEA have been promoting R&D for their respective purposes with the experimental equipment/facilities
at each R&D Center. The R&D Centers have not only been operating and managing various types of equipment/facilities but have also
been working on innovations and improvements for them, and in addition have themselves been developing experimental techniques,
management techniques and equipment/facilities necessary for the various R&D projects of JAEA.
This chapter introduces the developments which have been made over recent years at each R&D center.
Tsuruga Head Office
In “MONJU”, a “Function Confirmation Test” to confirm theperformance and function of the facility which was modified
to prevent sodium leakage, was completed in August, 2007,
and a “Plant Confirmation Test” to check the entire function of
the plant is being performed. In March, 2008, a report on thereevaluation of the seismic safety of “MONJU” was submitted
to the national authority in light of the revised Regulatory
Guide for Renewing Seismic Design. As a result of a false
sodium leakage alarm which occurred in March, 2008, all of
the sodium leakage detectors have been inspected.
In “FUGEN,” the spent fuel has been shipped out, andheavy water has been collected/shipped out however, after thedecommissioning project was approved in February, 2008, the“Fugen Decommissioning Engineering Center” was reorganized
to play a leading role in decommissioning projects. During
the modification construction, the facilities and equipmentswith relatively low levels of radiation, and non-contaminatedfacilities and equipment are being dismantled.
Tokai Research and Development Center, Nuclear
Science Research Institute (NSRI)
In research reactor JRR-3, a high performance vessel is being
developed to increase the intensity of a cold neutron beam. In
JRR-4, the technology to increase the efficiency of dosimetry
is being developed, in response to the demand for the increased
usage of medical irradiation at the boron neutron capture
therapy (BNCT). The technology for a Tandem Accelerator
is being developed to achieve a higher performance of an ion
source. Also, in the Facility of Radiation Standards, the facility
for monoenergetic neutron calibration is being developed, toperform energetic experiments of neutron monitors.
In addition, based on an external research fund which wasstarted in 2007, irradiation technology of a 12 inch siliconingot is being developed in JRR-4 and JRR-3.
Tokai Research and Development Center, Nuclear
Fuel Cycle Engineering Laboratories
In the Tokai Reprocessing Technology Development Center,a reprocessing test of spent MOX fuel was conducted byusing about 3.1 t of fuel from “FUGEN”. The research anddevelopment of the vitrification technology to decrease andstabilize the generation of waste which occurs in the tests
was performed. In the Plutonium Fuel Development Center,a fuel production technology development test, and a physicalcharacteristics measurement for the melting point and thermal
conductivity of MOX pellet was performed for future FBR
utilization.
In April, 2008, based on the cooperation agreement between
the U.S. Department of Energy (DOE) and JAEA, a cooperationagreement for the purpose of personnel training in the nuclear
fuel cycle field was concluded at the Nuclear Fuel Cycle
Engineering Laboratories and the Oarai R&D Center with the
Idaho National Laboratory (INL). Hereafter, an assignmentof personnel for training and education, and the exchange of
information will be carried out based on this agreement.
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JAEA R&D Review 2008

“Fugen Decommissioning Engineering Center”
Reorganized February, 2008

Silicon Ingots and Silicon Wafers (upper)
Equipment for Neutron Transmutation
Doping of Silicon Ingot in JRR-3 (left)
Semiconductors with uniform resistivity can be made byadding 31P evenly by nuclear transmutation of 31Si, andis widely used for thyristors to control high-voltages andhigh currents, CCD for video cameras and other devices.

Conclusion of Cooperation Agreement Ceremony with INL
Shigeo Nomura (right) Director General, Nuclear Fuel Cycle
Engineering Laboratories (JAEA)David J. Hill (center)
Deputy Laboratory Director for
Science & Technology, INL
Hiroshi Hiroi (left) Director General, Oarai Research andDevelopment Center (JAEA)

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

J-PARC Center
In Japan Proton Accelerator Research Complex (J-PARC), thebuilding for three facilities for accelerators and a Material andLife Science experimental facility (MLF) were completed andtrial operations are progressing for proton beam transportation.
Particularly, in October, 2007, the acceleration of a beam bya Linac was successfully accelerated to 3 GeV by a 3-GeV
synchrotron. Also, on May 22, 2008, the revolution control of
an incidence beam was succeeded by a 50-GeV synchrotron,and on May 30, 2008, the generation of a neutron by a nuclear
spallation reaction was successful in the MLF. The beam
intensity will be increased further by the end of December,
2008, to start using MLF.
Neutron beam lines now underMLF (J-PARC) Photo May 30, 2008

Oarai Research and Development Center
Regarding the R&D for the Fast Reactor Cycle Technology
Development Project “FaCT”, the irradiation tests for FBR
high burn-up fuel and minor actinide containing fuel were
performed in the experimental fast reactor “JOYO” and thepost irradiation examination facilities. The experimental
researches have been also carried out in the water and sodium
test facilities.
“JOYO” completed a highly advanced project (MK-III)
which improved the irradiation capability by about four times
more than in the past, and the results of the FBR cycle researchusing “JOYO” were accepted. As a result, “JOYO” received
the “2008 Minister of Education, Culture, Sports, Scienceand Technology Award” in the science and technology field(Research Dept.).
The refurbishment of the Japan Materials Testing Reactor(JMTR) was started for the restart in 2011. In May, 2007, theHigh Temperature Engineering Test Reactor (HTTR) achieved
the first 30 days continuous operation with a nuclear reactor
exit coolant temperature of 850 degrees C.
Also the development of the hydrogen production
technologies with an iodine-sulfur process and with athermochemical-electrolysis hybrid cycle has progressed.

Naka Fusion InstituteNaka Fusion Institute in close cooperation with Fusion R&D
Directorate, are aiming for the utilization of fusion energy.
As well as fusion plasma research and fusion engineeringresearch, the project to upgrade JT-60 is ongoing as a partof the Broader Approach Activities, to support and perform
supplemental research for the International Thermonuclear
Experimental Reactor (ITER)(to be built in France). In order tobroadly introduce the forefront of nuclear fusion R&D, facility
tours have been implemented often for local junior high school
students and high school students from across the country(Super Science High Schools). Also, the facility is open to the
public in October every year.

construction

in

HTTR Central Control Room

Junior high-school students marveling at the characteristics
of a synthetic diamond, which is used for a plasma
heating device at ITER
JAEA R&D Review 200899

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

Takasaki Advanced Radiation Research Institute
The Takasaki Advanced Radiation Research Institute of
JAEA opens the Takasaki Ion Accelerator for Advanced
Radiation Application (TIARA), which consists of four ion
accelerators, an electron accelerator, and gamma irradiation
facilities, to be used by JAEA and any organization to performR & D on biotechnology, new functional and environmentfriendly materials, radiation effects of materials and quantum
beam analysis. In 2007, the supply of a heavy ion micro beam ofa several hundred MeV class was started in the experiment for
the evaluation testing of radiation resistance of semiconductor
and radiation effect study for living cells. The development of
a high speed single ion hit technology at 10 hits per second or
more is also progressing.

Divergence defining slit for formation of a heavy ion
micro beam of a several hundred MeV classKansai Photon Science InstituteIn the Kizu district, research on advanced lasers with a subpetawatt laser oscillation output, and the oscillation of an
X-ray laser with a output wavelength of about 10 nm have been
performed and are progressing satisfactorily. In the research
for laser utilization, the generation of semi-monochromaticelectrons, high energy ion generation and basic research for
medical applications are being researched.
The new project, the creation of a “Photo Medical Industrial
Valley”, which had been proposed in the "Formation of creative
base for innovations integrating frontier fields" program
funded by the Special Coordination Funds for Promoting
Science and Technology, commissioned by the Ministry of
Education, Culture, Sports, Science and Technology (MEXT)
of Japan, was commissioned in 2007. The Photo-Medical
JAEA Kansai Advanced Relativistic Engineering LASER
Research Center was established for this project in October,
System (J-KAREN)
and activities for the creation of a “Photo Medical Industrial
Valley” were started in collaboration with industries and universities in the Keihanna region, which extends over Kyoto, Osaka, andNara prefectures, and is a new base of culture, science, and research in the 21st century.
In the Harima district, various researches such as the treatment of spent-nuclear fuel using synchrotron radiation in SPring-8, andthe superconductivity of a uranium compound are being researched. In addition, four dedicated beam-lines commonly used in thefacility are available for use by any organization.
Horonobe Underground Research Center
As a part of the research and development program on
the geological disposal of high-level radioactive waste, theHoronobe Underground Research Center, a division of theJapan Atomic Energy Agency, is implementing the Horonobe
Underground Research Laboratory Project (Horonobe URL
Project) with the aim of investigating sedimentary rockformations.
Regarding the construction of URL in 2007, the derrick
facilities used to excavate the shafts and pithead heater
facilities for the entrances of tunnel were constructed, and the
excavation of ventilation shafts reached 161 m, and 110 m for
the East shaft.
Winter view of the Horonobe Underground Research
In addition, a pilot boring survey was conducted for the
Laboratory from the viewing platform of the Public
purpose of estimating the amount of groundwater that would
Information House
flow into the tunnel during excavation, and to predict the pointof flow.
Regarding the above ground facilities, the construction of the Public Information House (Yume Chisoukan) of the Horonobe URL
center was finished in May, and was opened in June 30. In December, the number of visitors reached 10,000.
Regarding the construction of URL in 2008, the ventilation shaft, excavation of the east shaft and the drilling of a portion of thehorizontal shaft continued to be constructed, and the waste water treatment facility will be extended. A pilot boring survey near theventilation shaft, and joint researches with the other research institutes will also be continued.
Regarding the above ground facilities, the construction of the International Communication House as the base of international
exchange activities for domestic and international researchers will be started.
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JAEA R&D Review 2008

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

Tono Geoscience Center (TGC)
In the R&D to provide the scientific and technical basis forsafe geological disposal of high-level radioactive waste, theR&D for investigation technology to characterize the deep
geological environment mainly crystalline rock (Granite), a
wide of range of engineering for deep underground applicationand the long term stability of the geological environmentare being researched. In FY 2007, a drift was completed at
GL-200m level, and the shaft has reached GL-231.2m inthe construction of the Mizunami Underground ResearchLaboratory(MIU). This research is being performed in close
collaboration with universities and scientific institutions, both inJapan and abroad. In order to promote a mutual understanding,
TGC holds project explanation meetings for the local residents
and municipalities, and openly accepts site tours of the facility.
More than 3,000 visitors have visited MIU in the past year, andhave experienced underground research first-hand.
Ningyo-toge Environmental Engineering Center
The Ningyo-toge Environmental Engineering Center has
been developing a dry decontamination technology using an
Iodine heptafluoride (IF7) gas as a part of the decontamination
technology R&D for uranium enrichment facilities.
In 2007, to promote the optimization of the decontamination
processing condition, such as flux velocity, and flow rate, thedecontamination time was reduced to more than half the time
when compared to the start of operation.
The preparation for the decontamination of a single type
centrifugal separator cascade will continue to be promoted.
In addition, the dismantling of a Refinement and Conversion
facility was started and is in full scale, and is the first nuclear
fuel facility in Japan. The facility dismantling data will be
provided for the “Decommissioning Engineering System
Database”.
The Ningyo-toge Mine continues to perform technology
development for the mining complex
Aomori Research and Development Center
The Aomori office in the Rokkasho area and the Mutsu
office were unified and reorganized as the Aomori research anddevelopment center in April, 2007.
In the Rokkasho area, the construction of research facilities
for Broader Approach (BA) activities was started in March,
2008. The BA program is a joint implementation program
between Europe and Japan which will be advanced in parallelwith the ITER project which just started in France.
In the Mutsu area, a radioactivity related database for aclearance evaluation system, a practical decommission method
for the nuclear ship “Mutsu” related facilities, and an ultra low
level measurement and related techniques about C-14 and I-129
by accelerator mass spectrometer (AMS) are being developed.
“The 3rd symposium on marine environmental sciences” washeld at Mutsu to disclose information to the local residents,
and “The 1st JAEA Tandetron AMS utilization workshop” washeld to exchange information between AMS users.

Geological mapping at the bottom of main shaft

Decommissioning status
Conversion facility

of

the

Refinement

andIn order to propagate and to deepen the community’sunderstanding of the Broader Approach activities, abillboard was put up on the site in June, 2007.

JAEA R&D Review 2008

101

Tsuruga Head Office ■Development of Experimental Techniques / Facilities at JAEA R&D Centers14-1 Evaluation of Radioactive Inventory in Nuclear
Facility for Its Decommissioning–The Applicability of a Radioactive Inventory Evaluation Technique for “FUGEN” Verified–
Control rod(a)Measurement data of Activation foil

Extension portionof Pressure tube

Evaluation at the Extension portionof Pressure tube(▲position)(Bq/g)

Iron-water shield

Neutrontransport
calculation(ANISN, DOT-3.5)
Neutron flux
evaluationRadio
activation
calculation(ORIGEN)

Biologicalshield(Plant investigation)
Neutron flux
measurement

Takingsamples・Activation foil
・Spectrometer, etc

Comparative
estimation

Water weightmeasurement in
the Biological shield

Material
composition
analysismeasurement
data
calculation
data

Heavy
water

measurement
data
calculation
data

Radio
activation
measurement

Control rod

Extension portionof Pressure tube

Radio activation
evaluation

Fig.14-1 The estimation technique of radioactivation in “FUGEN”
The neutron flux is estimated from measurements
and calculations, and then, the radio-activation
is calculated with the ORIGEN code using it.
We had been acquiring data on neutron flux,
material composition, and water content in
concrete during the reactor operation period
in order to calculate the radio-activation. We
also had been measuring radio-activation.
We verified the validity of our technique bycomparing those data with calculation data.

BiologicalshieldFuelPressure
tube
Calandria
tube
Heavy
water

3E+04

1E+03

4E+04

Co-58

Co-60

Au-198

6E+01

1E+04

3E+05

2E+02

1E+04

7E+05

(a)Measurement
data of
Activation
foil
Co-58
Co-60
Au-198Iron-water shield
Comparative
estimation

Au-198

Evaluation at the biological shield’s surface(◎position)(Bq/g)

Input data
Input data

Co-60
8E+02

Evaluation at the Iron-water shield’s surface(●position)(Bq/g)

Cross-section of reactor core of “FUGEN”
Note; ▲●◎ indicate the measurement
positions of activation foil, andindicates the sampling point of
concrete specimens.

measurement
Evaluation
Extension1E+04
portion
8E+01
data at theof Pressure
calculation tube(▲position)
2E+02
1E+04
data
(Bq/g)

Co-60

tube

Calandria
tube

When we carry out the decommissioning of the “FUGEN”
reactor, it is important that the safety assessment of the public
exposure dose, the evaluation of radioactive waste amountand the planning of an appropriate method and procedure fordismantlement are based on an accurate radioactive inventory.
The radioactive inventory evaluation consists of (1)evaluation of the radio-activation on the structural materials
such as reactor cores, caused by the neutron irradiation during
reactor operation and (2) evaluation of the corrosion layer
adhering to the equipment and piping, etc. in which there aresubstances which were in the coolant water and were radioactivated in the reactor core.
The radio-activation of many parts can be evaluated byanalysis following the flow of “Evaluation by analysis” inFig.14-1. However, “FUGEN” is a complicated pressure tube
type reactor different from a light water reactor, and its neutron
behavior near the reactor can be complicated. Thus, it isnecessary to understand the neutron flux distribution accurately
before analyzing the radio-activation, and to evaluate theappropriateness of the result. We acquired as much data as
possible of the element composition, neutron flux, and the radio
activation in investigation of “FUGEN”, and we comparedthose with the evaluations made by analysis (Fig.14-1).

measurement
data
calculation
data

JAEA R&D Review 2008

concrete

Co-58

Co-60

Au-198

6E+01

1E+04

3E+05

2E+02

1E+04

7E+05

Evaluation at the biological shield’s surface(◎position)(Bq/g)
Co-58

Co-60

Au-198

8E+01

1E+04

3E+05

2E+02
1E+04
7E+05
Calculation
data by ORIGEN

Measurement data
Measurement data(less than ND)

Distance from inside of the
Biological shield (cm)In “FUGEN”, the radio-activation calculation was analyzed
by dividing the reactor into the core structure area and theshielding area surrounding reactor core.
The radio-activation measurement was carried out as follows.
A high pressure tube test piece irradiated for a certain period
in the core structure area during the reactor operation wasmeasured. Also, many Activation foils set around the reactor
core in the shielding area and irradiated for a certain period
during the reactor operation were gathered and measured.
Moreover, after reactor shutdown, pieces of concrete weretaken from the Biological shield and were measured.
The analytical data values were almost equal or moreconservative than the measurement data in all cases. Thus, thevalidity of our evaluation method was confirmed (Fig.14-2).
This technique was applied to the radioactive inventory
evaluation in the “Decommissioning plan of FUGEN”(authorized on February 12, 2008), and the public exposureand the amount of generation of waste were evaluatedappropriately from these results. Moreover, this technique will
also be applicable to the safety clearance evaluation of thewaste surrounding the nuclear reactor after dismantling which
had been exposed to neutrons.

Reference
Kitamura, K. et al., Evaluation of Radioactive Inventory for Fugen’s Decommissioning, Hoshasen, vol.34, no.1, 2008, p.53-63 (in Japanese).

102

7E+05

specimenin the BiologicalshieldEvaluation
at the Iron-water
shield’s surface(●position)(Bq/g)

measurement
data
calculation
data

FuelComparison between calculated
Fig.14-2
Pressure radio-activation values
and measured

3E+05

Au-198measurement
3E+04
8E+02
data
calculation
4E+04
1E+03
(b)Measurement
data ofdata

Radio activation：Co-60 (Bq/g)(Evaluation by
analysis)■ Nuclear Science Research Institute

Development of Experimental Techniques / Facilities at JAEA R&D Centers14-2 Achieved Utilization of Short-Lived Radioactive
Nuclei Beams–Development of Ion Sources for Radioactive Nuclei–
Filament

Proton beam

Proton number(a)Radioactive Nuclei
Beam

La

Ba
Xe

Cs
I

Te
Sn

Sb
In

Cd

Ag

Pd

Neutron number
Ionizer

Uranium Carbide
TargetRadioactive Nuclei
Beam

Y

Sr

Proton number

(b)

Rb

Kr
Se

Br
As

Ge
Zn

Cu

Ni

Neutron number
PlasmaHeaterYield of ion beam(ions/s)
1g/cm2 of 238U@1ǴA

Ga

Stable nuclide
1x107~9x107
1x106~9x106
1x105~9x105
1x104~9x104
1x103~9x103
1x102~9x102

Fig.14-3 Schematic view of Ion sources for Radioactive Nuclei
Two types of integrated uranium target-ion source systems
were developed that are considered to be chemical properties
of elements.
(a) Surface ionization type ion source (operated at ~2300°C):
Ionized by the interchange of electron at the heatedmetal surface on the inner wall of the ionizer.
(b) Low-pressure arc-discharge ion source (operated at~1500°C): Ionized in plasma created in an ionizer by anarc-discharge.

Fig.14-4 Beam intensities of the nuclei successfully
ionizedWe succeeded in the ionization of neutron rich nuclei with
19 elements and 105 nuclei, that were produced in a proton
induced uranium fission.

We are developing ion sources to ionize radioactive nuclei,
which do not occur naturally, for acceleration by a radioactive
nuclei beam accelerator in the facility of the JAEA tandem
accelerator. Radioactive nuclei of about 40 elements and 400
nuclei are produced at a time when uranium is bombarded bya proton beam. For an effective acceleration of these nuclei,
the diffusion/evaporation should be fast from the uranium
target and ionization at once, because most of these nuclei arein minute amounts and short-lived, as half-life is less than afew seconds. Therefore, we developed a unique uranium target
and two types of ion sources for radioactive nuclei that areconsidered to be chemical properties of these elements.
The more the weight of the uranium is increased, moreradioactive nuclei in the target was obtained. However the
diffusion/evaporation of the fission products becomes slower if
a highly-condensed target is made. In addition, the target shouldbe raised to a high temperature to accelerate the speed of the
diffusion/evaporation. Therefore, we made a uranium carbide
target, which has a chemical stability at a high-temperature on
a graphite fiber, so that the amount of uranium (~800 mg/cm2)
was maintained. This uranium target has a large surface areaand fast diffusion/evaporation speed of the fission products.
High ionization efficiency is required for the ion source forradioactive nuclei that are produced in minute amounts. It is
impossible to ionize every element with one type of ion source
produced in the reaction of proton induced uranium fission.
Therefore, two types of integrated uranium target-ion source
systems were developed that are considered to be chemical
properties of elements. It is suitable that the surface ionization
ion source (Fig.14-3 (a)) to ionize alkali, alkaline-earth andrare-earth elements that have a low ionization potential. We
developed an ion source that could operate at 2300°C to obtainhigh ionization efficiency.
The low-pressure arc-discharge ion source (Fig.14-3(b)) issuitable for the ionization of gaseous elements, Kr and Xe, andvolatile elements such as Sn, In, Ag and Cu. We built a heater
to maintain a high-temperature of the uranium target on theion source. This modification made it possible to efficiently
transfer radioactive nuclei from the target to the ionizationchamber.
Using these ion sources, 105 isotopes in 19 elements producedin the proton induced fission of uranium have been provided to
the radioactive nuclei beam accelerator with maximum yield of
107 ions/s as shown in Fig.14-4. With these radioactive nuclei
beams, advances are expected in the studies of synthesis ofelements in the universe and structural analysis of materials.

Reference
Osa, A. et al., Ion Source Development for the On-line Isotope Separator at JAEA, Nuclear Instruments and Methods in Physics Research B, vol.266, issues
19-20, 2008, p.4394-4397, DOI: 10.1016/j.nimb.2008.05.063.
JAEA R&D Review 2008

103

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

Nuclear Fuel Cycle Engineering Laboratories ■14-3 Study on Basic Properties for Establishment of a
Fast Reactor Cycle System–Melting Temperatures and Phase Diagram of MOX Fuel–(b)

Temperature (K)(a)

Fig.14-5 Microstructures of heated MOX with 40% Pu content(a) The sample was encapsulated in W and heated to 2963K.
Precipitates of W metal and Pu oxide were observed by electronprobe micro analyses in the grain boundary.
(b) The sample was heated to 2978K using a W capsule with a
Re inner. Grain growth was observed, and there was no traceof melting.

It is essential that nuclear fuels are operated in an undamaged
condition during their entire lifetime. Therefore, the maximum
temperature of the fuel is limited by the melting temperatureof MOX.
JAEA has developed a fast reactor recycle technology. As apart of the development, the basic properties of uranium andplutonium mixed oxide (MOX) fuels that are used as fuel inthe fast reactor have been studied. The melting temperature of
MOX fuels was measured during the period of 1960-1970’s asfunctions of Pu content and O/M ratio, and their phase diagram
was evaluated. Unfortunately, analyses of the melted samples
were not performed in those studies. As a result, the measured
melting temperature of the actual compounds is unreliable.
In this study, the melting temperature was investigated by a
new measurement technique using a W capsule with a Re inner.
It was found that the previously obtained data was not correct,
because a reaction between the sample and the W capsulematerial occurred and the compound which formed affected
the melting. The melting temperature of MOX fuels wasdetermined to be a function of the Pu content and O/M ratio,
and the phase diagram of the UO2-PuO2 system was revised forthe first time in about 40 years.
Fig.14-5 (a) and (b) show the microstructures of the heated
MOX fuel with about 40% Pu content. The sample shown inExperimentThis work Solidus
Liquidus
Kato et al. Solidus
Liquidus
Aitken et al. Solidus
Liquidus
Lyon and Baily Solidus
LiquidusCalculation
This work
Epstein
AdamsonPuO2 (wt%)

Fig.14-6 Solidus and liquidus temperatures in the
UO2-PuO2 system

Fig.14-5 (a) was encapsulated in W and heated to 2963K, which
was the method used in the conventional melting temperature
measurements. The sample was transformed into a block, andthe precipitates of metallic W and Pu oxide were observed in
the grain boundary. A reaction occurred when the MOX fuelscontained more than 30% Pu.
The sample shown in Fig.14-5 (b) was heated to 2973K
inside a W capsule with a Re inner. Grain growth was observed,
but no traces of melting, such as deformation was seen. Theseobservations show that the previously obtained data is notcorrect, and a reaction between the MOX and the W capsulematerial occurred and affected the melting measurements.
The melting temperature was determined by the thermalarrest method which was observed in the heating curve whena sample melted. Fig.14-6 shows the measurement results ofsolidus and liquidus temperatures which were measured byusing the W capsule with a Re inner. The measurement results
were evaluated by the ideal solution model, and the phase
diagram of the UO2-PuO2 system was revised. The calculated
results represented the experimental data with an accuracy of±20K. It was found that the solidus temperatures of fast reactor
MOX fuels were 50-100K higher than the existing data had
shown.

Reference
Kato, M. et al., Solidus and Liquidus Temperatures in the UO2-PuO2 System, Journal of Nuclear Materials, vol.373, issues 1-3, 2008, p.237-245.

104

JAEA R&D Review 2008

■ Nuclear Fuel Cycle Engineering Laboratories

Development of Experimental Techniques / Facilities at JAEA R&D Centers14-4 Basic Knowledge on Treating Various Wastes
Generated from Pyro-Reprocessing–Recovery of Actinides from Solid and Liquid Wastes–
UO2
PuO2

High Purity Ar Atmosphere Glovebox

Fig.14-7 Outline of main pyro-reprocessing
steps
UElectrochemical reduction:+Zr
Oxide spent fuel is reduced to a metal- ++form by the electrochemical procedurein the lithium chloride (LiCl) molten salt.
Electrorefining:
Uranium (U), and plutonium (Pu) were
salt・Cd
anode Solid
Cd
recoveredbytheelectrochemicalcathode cathodeprocedure in lithium chloride-potassium
3+Pu
chloride (LiCl-KCl) eutectic molten salt
O2
for separating from fission products.
U3+
LiCl-KCl
Distillation:
Removal of adhered salt and cadmium
Electrochemical reduction
Electrorefining
Salt・Cd distillation
Injection castingfrom a recovered U, Pu, and U-Pu metal(650°C)(500°C)(500-1500°C)(1500°C)
ingot was obtained.
23+
UO2 + 4e → U +2O
U → U→ U㧔with salt㧕
→U㧔ingot㧕Injection casting:→U-Pu-Zr㧔alloy㧕
→ U,Pu㧔with Cd㧕 →U㧙Pu㧔alloy㧕
PuO2+ 4e- → Pu +2O2Pu → Pu3+
Injection casting of U, Pu to form a metal
fuel.
U,Pu

U,Pu

Solid wastes

Liquid wastes

Convert into
chloride form

Actinide and chlorine
are precipitated

Recyclein themain steps

Evaporation

U+Pu

silver nitratesodium hydrate

Fig.14-8 Flow chart of waste processing
Procedure for recycling of actinide from solidand liquid wastes in pyro-reprocessing.

water

The pyro-reprocessing (Fig.14-7) is one of the advanced
technologies for the separation of actinide elements from spent
nuclear fuel. The elements are electrochemically separated
in a molten salt medium unlike the conventional aqueousreprocessing, which is basically performed in nitric acidand an organic solvent medium. Since actinide elements arenaturally recovered together in the pyro-reprocessing, thisprocess has a superior resistance to the nuclear proliferation.
Moreover, this process is expected to enhance the economical
potential of nuclear fuel reprocessing, because the processitself is quite simple when compared to the conventionalaqueous-reprocessing. Concerning the wastes from the pyroreprocessing, some of the waste might contain significant
amounts of actinide elements. It is believed that the recoveryof these actinide elements enhances the potential of the pyroreprocessing in regards to the efficient utilization of nuclear
resources and reduction of environmental impact.
The actinide in the chemical formation such as oxide, which
can not be treated in a normal process (Fig.14-7), is includedin the solid wastes of the residue from the electrodes andsediments. On the other hand, the process operation of thepyro-reprocessing requires a chemical analysis in each step inorder to control the process condition. Various liquid wastes
containing actinide elements are generated from the analysis. A
procedure is being developed to recycle the actinide elementsfrom these wastes.

By conversion of the solid wastes to a chloride form, theactinide elements in the solid wastes can be put back into thepyro-reprocessing. The solid wastes was poured into a LiCl-KCl
molten salt, which is the same mixture as the agents currently
used in the electrolytic refining process, and a chlorination
reagent (zirconium chloride) was gradually added to the
LiCl-KCl molten salt. Most of actinide elements weresuccessfully converted to a chloride form after sufficient
reaction time (about 10 hours). As for the liquid wastes, theseparation method by evaporation was investigated. The actinide
elements were recovered as a residue. Since the liquid wastes
contain chlorine which causes corrosion of equipments andharmful cadmium, it is necessary to prevent these substancesfrom converting to water which is separated from the residue.
These elements are removed by precipitation by using silver
nitrate and sodium hydrate before the evaporation. The actinide
elements in the residue are also converted to a chloride form as
mentioned above.
We have not only been studying the main steps of pyroreprocessing but also, the incidental technology such as waste
treatment (Fig.14-8), in order to realize the pyro-reprocessing
for the advanced nuclear reactor fuel. This study is being
performed in the Chemical Processing Facility (CPF) in
collaboration with Central Research Institute of the ElectricPower Industry.

Reference
Nakayoshi, A. et al., Basic Knowledge on Treating Various Wastes Generated from Practical Operation of Metal Pyro-reprocessing, International Symposium
on EcoTopia Science (ISETS07), Nagoya, Japan, 2007, p.1062-1066, in CD-ROM.
JAEA R&D Review 2008

105

J-PARC Center ■Development of Experimental Techniques / Facilities at JAEA R&D Centers14-5 Development of Damage Reduction Technique on
Mercury Target for the Spallation Neutron Source–Visualization of the Bubble Formation Behavior in Mercury–

H2O

Hg
Hg

H2O
（a）
Measurement of contact angle
Heliumbubble
ozzle
OutN
er:20
0 m
Inne
r:100
m

Bubble formation
in mercury
visualized by using X-rays

Hg

HeliumbubbleNozzleOuter:200Ǵm
Inner:100Ǵm

H2O
HeliumbubbleNozzleOuter:200Ǵm
Inner:100Ǵm

HeliumbubbleBubble formation in water（b）
Results of numerical analyses

Fig.14-9 Visualization of Bubble Formation Behavior
The bubble formation behavior in opaque mercury wasvisualized by using high intensity X-rays. The bubble formed
around the tip of the nozzle in mercury because of the poor
wettability of the mercury while the bubble formed from thetip of the nozzle in water.

Fig.14-10 Measurement of Wettability and Prediction of
Bubble Formation Behavior by Numerical analyses
Contact angles, θc, which are a parameter to express
the wettability between the liquid and nozzle material were
measured. The results of the numerical analyses clearly
represented the visualizing test results by using the measured
contact angles in the analyses.

In the materials and life science experimental facility of“J-PARC”, innovative sciences in the materials and life sciencefields will be developed by using high intensity neutrons froma MW-class spallation neutron source. Liquid mercury is usedas the target material to produce neutrons by the spallation
reaction from a viewpoint of heat removal. However, pitting
damage degrades the target structural material, which is caused
by pressure waves generated in rapid heating while the proton
beam bombards the mercury. To improve the endurance of themercury target and to use it under a high power proton beam,
it is essential to reduce the pressure waves which cause thepitting damage. We aimed for a technique that would reduce
the pressure wave by injecting helium gas bubbles into themercury, because the gas bubbles act as a cushion.
It was predicted that the pressure wave would be reducedby half by injecting 0.1 Vol.% of tiny bubbles of 100 μm
in diameter (microbubbles). The reduction mechanism of
the pitting damage by the injection of microbubbles is being
considered in corporation with CCSE.
To understand the bubble formation behavior in mercury
and to develop the technique for injecting the microbubbles
into mercury, we tried visualizing bubble formation behavior

in opaque mercury by using X-rays. Although it was difficult
to visualize the behavior clearly because the absorption rateof the X-rays is high in mercury, the visualization of the
bubble formation behavior was successful by increasing thetransmission of X-rays using high intensity X-rays in “SPring-8”,
and optimizing the thickness of the mercury. Fig.14-9 shows
the bubble formation behavior while helium gas was injected
into stagnant mercury from a narrow nozzle (inner diameter:
100 μm, outer diameter: 200 μm) when compared with thecase in water. In water, the bubble formed from the tip of thenozzle, but in mercury, the helium gas flows to the boundary
between the mercury and the nozzle wall because of the poor
wettability of the mercury as shown in Fig.14-10 (a), and the
bubble formed around the tip of the nozzle.
The numerical analysis became a useful tool to predict the
bubble formation behavior in opaque mercury. It was confirmedthat the numerical analysis represented the visualized resultsby considering contact angle, which is a parameter to express
the wettability as shown in Fig.14-10.
Hereafter, a bubble injecting device to apply the practicalmercury target will be considered to improve the endurance ofthe mercury target.

Reference
Kogawa, H. et al., Effect of Wettabitlity on Bubble Formation at Gas Nozzle under Stagnant Condition, Journal of Nuclear Materials, vol.377, issue 1, 2008,
p.189-194.

106

JAEA R&D Review 2008

■ Oarai Research and Development Center

Development of Experimental Techniques / Facilities at JAEA R&D Centers14-6 Measuring Displacement and Vibration of High Temperatureand High Radiation Components Using Optical Fiber–Development of Fast Reactor Integrity Monitoring Technology–
Calculation
Measurement

FBG

Reactor power (MWt)

Spectrum analyzer

Displacement (%)

SLD light source

Optical fiber

Reactor power

FBG
Primarypump

Measurement date

ReactorPrimary cooling system area
Fig.14-11 Experimental setup of measurement systemusing optical fiber in the “JOYO” primary cooling area
A few FBGs were combined into a single fiber, and mounted
on the piping supports. In the measurement, a wideband light
was injected into the optical fiber, and then the wavelengthof reflected light from FBGs was scanned with a spectrum
analyzer.

In the sodium-cooled fast reactor (SFR), circulating pumps
and several piping support systems were installed aroundthe primary cooling system piping. How to efficiently andaccurately measure the displacement and vibration in hightemperature and high radiation components is important,in order to monitor the SFR structural integrity. In theexperimental fast reactor “JOYO”, displacement and vibration
measurements of the primary cooling system were carried out
using Fiber Bragg Grating (FBG) to resolve this challenging
issue (Fig.14-11).
FBG is an optical fiber sensor, which reflects light at a
specific wavelength. Temperature variations or displacement
applied to the FBG causes the reflected wavelength to shift.
The FBG is a simple instrument, because several FBGs can be
combined in a single fiber.
Displacement can be evaluated from a wavelength shift Δλ1of the FBG with both-ends fixed. However, Δλ1 includes
a wavelength shift Δλ2 due to temperature variation. This
becomes to be a source of displacement measurement error.
Therefore, Δλ2 of an FBG with one end fixed for detecting
only temperature variation was eliminated fromΔλ1. The

Fig.14-12 Measured displacement of the piping support
by FBG
Displacement following reactor power was successfullymeasured.

Signal voltage (mV)IHX

At the primary pump operation(flow rate:1538m3/h)
At the primary pump stop

Time (ms)Fig.14-13 Measured vibration of the piping support by
FBG
Vibration changes with the primary pump operation weresuccessfully measured.

measured displacement matched the calculation based on thethermal expansion. The measured result indicates that thermal
displacement of a piping support was accurately measured(Fig.14-12).
Vibration can be measured as a high frequency displacement.
A faster vibration, however, cannot be measured by aspectrum analyzer, because its sampling time is in the orderof seconds. In the measurement, the reflected wavelength shift
was converted into light magnitude using an optical filter,
and then the voltage signal corresponding to the displacement
was obtained using a photo diode. This enables a high speed
measurement of the vibration. Signal fluctuation at the primary
pump operation was larger than when the pump was stopped.
The measured result indicates that vibration due to the pump
operation was successfully measured (Fig.14-13).
The measurements were successfully conducted with nodegradation of the optical fiber up to an accumulated γ-ray
dose of 4×104Gy, corresponding to 120 effective full powerdays of operation. The measured results demonstrated that
the FBG is suitable for the monitoring of the SFR structural
integrity.

Reference
Matsuba, K., Ito, C. et al., Development of Fast Reactor Structural Integrity Monitoring Technology Using Optical Fiber Sensors, Journal of Power andEnergy Systems, vol.2, no.2, 2008, p.545-556.
JAEA R&D Review 2008107

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

Oarai Research and Development Center ■14-7 Material Research for Transmutation of LLFP in a
Fast Reactor–Irradiation Test of 11B4C as a Neutron Moderator–

(a)

(d)

(f)

Grain

(c)

5mm

(g)

Grain Boundary

(b)

(e)

11

B4C

5mm

Fig.14-14 Appearance of (a) Irradiated 11B4C and(b) 10B4C at around 500°C
An irradiated 10B4C pellet was broken, and the11
B4C pellet kept the same shape as before being
irradiated.

10

B4C
25nm

Fig.14-15 Microstructure of irradiated 11B4C and 10B4C
Irradiation defects in 11B4C microstructures. Micro dislocation loops(c), bubbles (d). Small helium bubbles (e) were gathering in the grainboundary of 11B4C. Helium (He) bubbles (f) were gathering in highdensity on the grain in 10B4C. Large bubbles (g) grew around grainboundaries in 10B4C.

It is difficult to store high level waste that contains longlived fission products (LLFP). Transmutation of LLFP, whichchanges them into short lived or stable nuclides, is a solution.
Thermal reactors are efficient for LLFP transmutation, but
fast reactors also have possibilities for it by modifying a fastneutron into a thermal neutron. 11B4C was fabricated fromconcentrating 11B in natural boron carbide. Both hydride and11
B4C are candidates to be moderators. It is possible to raise
the efficiency by loading LLFP as a target with the moderator,
in order to transmute LLFP in the fast reactor. The slow-down
power of 11B4C is lower than hydride. 11B4C has an advantage
in the thermal design, because the boron carbide has beenstabilized to the melting point chemically, while the hydridedissociates the hydrogen when it reaches a high temperature.
Hydride and 11B4C were irradiated in Phenix, a fast reactor in
France. We have many irradiation experiences with 10B4C as anabsorber material. 10B4C was fabricated from concentrating 10B.
He was produced by a 10B (n,α) 7Li reaction. A B4C pellet was
sintered for irradiation. The He stored in it and changed into He
bubbles. These bubbles caused the pellets to crack, which is abad effect for materials stability. The volume of He generated

in 11B4C was expected to be about 1/100 in comparison withthat in 10B4C.
In this study, 11B4C was irradiated at 530°C to 1.9×1026 n/
m2 (E>0.1 MeV), and 10B4C was irradiated at 800-900°C to
3.1×1026 n/m2 (E>0.1 MeV). Afterwards, a post irradiation
experiment was performed. The 10B4C pellet was difficult tobe divided around 700-800°C. Then, the 11B4C was specially
irradiated at 530°C. The irradiated sample became very brittle.
Therefore, a test piece of irradiated 11B4C was prepared by ionmilling method in full detail.
In Fig.14-14, the appearance of 11B4C and 11B4C after
irradiation is shown. In Fig.14-15, the microstructure of 11B4C
and 10B4C after irradiation is shown. In the 11B4C grain, large
bubbles of a high-density were observed compared to 10B4C.
In the grain boundary, additional large bubbles were observedin the grain. It was considered that in the 11B4C, the growth of
the bubble was slow, and the nucleation frequency is also low.
From the observation results, it was confirmed that 11B4C wasstable under the moderator service conditions. In the future, amore detailed microstructure observation will be performed.

Reference
Donomae, T. et al., Neutron Irradiation Effects on 11B4C and Recovery by Annealing, Nippon Seramikkusu Kyokai Gakujutsu Ronbunshi, vol.115, no.1345,
2007, p.551-555 (in Japanese).

108

JAEA R&D Review 2008

■ Oarai Research and Development Center

Development of Experimental Techniques / Facilities at JAEA R&D Centers14-8 Contribution to “ITER” by Technical Developmentof Post Irradiation Examinations–Remote-Handling Type Welding / Processing Techniques for Irradiated Materials–

Arc distancecontrol motor

Welding torch

Operation
unit

Workinghead

Clamping jig(a) TIG-welding machine
By using a manipulator for remote-handling, it is easy toclamp the specimens onto the welding machine, to adjust thewelding position and to adjust the gap distance between thetorch and the specimen.
A narrow heat affect zone (HAZ) in the welded specimen canbe achieved by a clamping jig with high thermal conductivity.

Workingtable(b) Processing machine
By using a manipulator for remote-handling, it is easy to
fix the specimen onto the remote-controlled NC lathe. Thismachine warrants the high accuracy of processing, the sameas the out of the hot cell processing.

Fig.14-16 Technical development of a remote-handling type welding/processing machines(Non-irradiated)Torch
Ar gaswelding
direction

Sample(Irr.or non-irradiated SUS316LN)
He gas
Cross section of welding
Welding(Remote-controlled TIG welding) SS316 specimen(irr./non-irr.)

Before re-irradiation
After re-irradiation(Irr.Temp.: 150°C)
Tensile strength (MPa)(Irradiated)(Test Temp.: 20°C)(Test Temp.: 150°C)
Test Temp.

Specimen

20°C

Irr./non-irr. weldment(He generation:3appm)

150°C
20°C
150°C
20°C
150°C

Processing(Remote-controlled NC lathe)

Irr.

Irr./non-irr. weldment(He generation:10appm)
Base materials

Non-irr.
Irradiation damage (dpa)

Fig.14-17 Welding/processing procedures for fabricationof test specimens
The weldment specimen was fabricated from irradiatedand non-irradiated specimens by remotely handled welding/
processing machines in a hot cell. The plate type specimen
with a tab was chosen, and the conditions for welding andprocessing were determined. By applying this technique, data
can be obtained from the specimens after re-irradiation.

Fig.14-18 Neutron irradiation test results of welding
material before and after irradiation
Neutron re-irradiation tests on SS316LN weldment specimens,
joined with irradiated and non-irradiated materials, were
performed. From the results of a mechanical property test,
it was clarified that the tensile strength of the weldment
specimens is the same as that of the base materials.

For maintenance and/or replacement of the fusion blanket,
new cooling pipes for the blanket will be joined by welding to
the existing cooling pipes irradiated by high energy neutrons.
Therefore, it is necessary to evaluate the effect of helium
generation by neutron irradiation on the mechanical propertiesof the weldment. In this study, all the work from the welding
and processing of test specimens, to the installation of thespecimens into the irradiation capsule, were performed in the
hot cell by remote handling. Therefore, the remote-handlingwelding/processing machines, and the assembling procedureof the irradiation capsules were developed based on theexperiences of the post irradiation examinations in the Japan
Materials Testing Reactor (JMTR) hot laboratory.
The remote-handling type Tungsten inert gas (TIG) welding
machine (Fig.14-16 (a)) was developed. In order to evaluate
the mechanical properties of the weldment correctly, a platewith a tab where the crater part is able to be removed, wasadopted as the shape of the test specimens. On the other hand,the processing machine (Fig.14-16 (b)) with the end-millingtype numerical control (NC) lathe was developed. It is possible
to fabricate the weldment specimens with high accuracy indimensions to within 50 μm. These developments for remotehandling techniques enabled a systematic evaluation on themechanical properties of weldment specimens with irradiatedand non-irradiated materials.
Using austenitic stainless steel SS316-LN-IG (IG means“ITER” grade) irradiated up to the “ITER” irradiation
conditions in “JMTR”, the TIG welding between the irradiatedand non-irradiated materials was tested in a hot cell (Fig.14-17).
From the test results, it was obvious that the sealed gas and thewelding heat affected the welding, as there were defects suchas, a crack on the surface and the cross section of weldment,
an undercut and insufficient welding. Especially, it was foundthat when the welding heat input was about 1 to 2 kJ/cm, agood weldment specimen without cracks could be obtained.
Furthermore, their welding test specimens were processed as
tensile-type weldment specimens, and they were installed onto
the irradiation capsule in the hot cell. The neutron re-irradiation
test was successful in ”JMTR”, and the mechanical propertiesof a re-irradiated weldment was obtained for the first time in
the world (Fig.14-18).

Reference
Tsuchiya, K. et al., Effect of Re-irradiation by Neutrons on Mechanical Properties of Un-irradiated/Irradiated SS316LN Weldments, Journal of Nuclear
Materials, vol.373, 2008, p.212-216.
JAEA R&D Review 2008

109

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

Takasaki Advanced Radiation Research Institute ■and Structure of
14-9 Correlation between the Charge-State+
Fragment Ions for Swift C3 Traversing a Thin Foil

–Study on Interaction between a MeV Cluster Ion and a Matter–
Deflection plate

Micro channel plateNon-deflection
pointsCarbon thin foil

CCD camera
Deflected
pointsDeviation

Fig.14-19 Apparatus for simultaneous measurement ofcharge-state and spatial pattern of a cluster ion
At impingement of a swift cluster ion on a thin foil, itsvalance electrons are stripped away in a short time, anda rapid explosion process is initiated subsequently by a
Coulomb repulsion, which provides a snap-shot of theposition of the nuclei in the incident cluster ion. The chargestate of each fragment ion is determined by the position on
the MCP, on the basis of the deflected angle relevant tothe charge depending mainly on an electrostatic field of thedeflection plates.

Triangular patternLinear patternFig.14-20 Two patterns of correlation of the deflected
points and the non-deflection points
The charge-state of each deflected point is identified by the
relation between its position and charge distribution on the
MCP, as shown in (a). The instance of pattern (b) and (c)shows an almost equilateral triangular pattern and a linearchain pattern with almost regular intervals, respectively.
The blue round symbol shows the original pattern before
deflection.

Table14-1 Average charges of the fragment ions arising from the dissociation of
a C3+ cluster ion passing through a thin carbon foil, estimated experimentallyand theoreticallyComparison of atomic-position in a
Comparison of cluster structure
liner-chain patternLinearTriangle

Center

Edge

Expt.

1.96±0.03

1.89±0.02

Expt.

1.86±0.04

2.01±0.03

Calc.

1.91

1.89

Calc.

1.88

1.93

Studies on material science and biotechnology for various
ion beams with an energy ranging from 30 keV to hundredsof MeV are in progress in the Takasaki advanced radiation
research institute. Understanding the interaction between
an incident ion and a target is indispensable for selection ofsuitable experimental conditions, i.e., an incident ion species,
its energy, and a target, and analyses of experimental results.
The interaction for irradiation with mono-atomic ion has
been investigated from the early days, and some models
have succeeded in explaining most of experimental results.
However, various strange phenomena out of such models havebeen observed for irradiation with a swift cluster ion consistingof a number of atoms for example, secondary-ion or secondaryelectron emission.
When a swift cluster ion impinges on a matter, a lot of atoms
are brought into a very small volume simultaneously therefore,
vicinal fragment ions interact with each other even after the
dissociation of the cluster. In this process, a spatial pattern of thecluster constituents could affect their interactions with a matter.
The effect of the spatial pattern on charge-state distribution offragment ions arising from the dissociation of a cluster passing
through a thin foil was investigated, to make a new model forthe interaction between a cluster ion and a matter.
A new apparatus as shown in Fig.14-19, was developed
for the investigation. Ions emerging from a thin carbon foil
deflected depending on their charge-state by an electrostatic
deflector placed between the foil and a micro channel plate(MCP). A two-dimensional pattern of the deflected ions wasobserved as luminance points on a fluorescent screen equippedwith the MCP. The apparatus enables us to measure the chargestate of each fragment ion of a cluster ion and their spatial
pattern, just after emerging from the foil simultaneously, as
shown in Fig.14-20. The charge-state distribution of fragment
ions of a 3-MeV C3+ was measured for different patterns, i.e.,a linear-chain pattern and an equilateral triangular pattern after
emerging from the foil.
The results are summarized in Table14-1. The cluster
average charge obtained for the liner-chain pattern is larger
than that for the triangular pattern, and the average charge of
two edge-position ions in a liner-chain pattern is larger thanthat of the middle-position ion. The theoretical average chargeof the fragment ions of a C3+ ion was calculated considering theeffect of the nuclear charge of the other neighboring fragmentson the binding energy of the valence electrons, and the spatial
pattern of the fragments, a repulsive Coulomb force among thefragment ions, a wake force induced by each fragment and etc.,
were taken into account in the calculation. The obtained averagecharges are qualitatively consistent with the experimental
results as shown in Table14-1.
Although the measurement of the charge-state distribution offragment ions have been done by other researchers so far, this is
the first time measurements were performed by classifying thespecial pattern of the fragment ions. The interaction between aswift cluster ion and a matter correlates with the spatial patternof the constituents in a cluster, which should be demonstrated as
one of the most characteristic of the cluster-matter interaction.

Reference
Chiba, A. et al., Average Charge and Its Structure Dependence of Fragment Ions under Irradiation of a Thin Carbon Foil with a 1-MeV/atom C3+ Cluster Ion,
Physical Review A, vol.76, 2007, p.063201-1-063201-6, DOI:10.1103/PhysRevA.76.063201.

110

JAEA R&D Review 2008

■ Tono Geoscience Center

Development of Experimental Techniques / Facilities at JAEA R&D Centers14-10 Reduction of Water Inflow for Continuation of Deep
Shaft Excavation–Development of Grouting Technology at the “Mizunami Underground Research Laboratory”–
Niche (Boring area)
Ventilation Shaft

Quantity of injected cement(kg/m)

Injection-holeGL-100m

Location of the preexcavation grouting
Niche(Boring area)
Ventilation Shaft:
200.2m (ø4.5m)

Horizontal tunnel
Niche(Boring area)
GL-200m

Fig.14-22 Quantity of injected cement per unit length of the preexcavation grouting at the Ventilation Shaft niche (boring area)
At the end of the Ventilation Shaft niche (boring area), the excavation
confirmed a large volume of injected cement.

Main Shaft:231.2m (ø6.5m)

Fig.14-21 Image of the “Mizunami UndergroundResearch Laboratory”
Depth of shafts as of March 15 2008.

The “Mizunami Underground Research Laboratory”(MIU), one of the main facilities in Japan for research anddevelopment of the technology for high-level radioactive waste
disposal, is being constructed in Mizunami City. Excavation ofshafts and galleries is being done to confirm the results of thepredictions made in the surface-based investigation phase. Asof March, 2008, the excavation of the Main and Ventilation
Shafts had reached GL-231.2m and -200.2m (below groundlevel) (Fig.14-21). Optimization of the costs for treatmentof the groundwater as well as reduction of the inflow water
volume was required for continuing the MIU excavations. In
planning the excavation of the shafts and tunnels below about
GL-180m, it was necessary to get more reliable informationon the conditions in terms of the rock mass stability andhydrogeology. Therefore, pilot borehole investigations wereconducted from the bottom of the shafts. The results indicatedthat large water inflow could be expected during the excavation
around the Ventilation Shaft and the Ventilation Shaft niche(boring area) at GL-200m. In order to reduce the drilling time,procedures for grout injection-hole drilling incorporated the useof a shaft jumbo (blast hole drilling rig) instead of a conventional
borehole drill rig. This is one of the engineering techniques
tried in the development of the construction procedures. As aresult, it was possible for grouting to be accomplished withinFig.14-23 Photograph of cement
filled fractures
Red arrow shows cement filled
fractures.

the planned excavation cycle. The two shafts were extended toGL-200m followed by pre-excavation grouting of the rock massaround the Ventilation Shaft and the planned Ventilation Shaft
niche (Fig.14-21). Then, horizontal tunnels were excavated atthat level. The pre-excavation grouting around the Ventilation
Shaft niche is shown in Fig.14-22.
After the pre-excavation grouting and tunnel excavation,
geological mapping was performed to determine the locationand extent of the grout penetration. The geological mappingof the tunnels and walls confirmed the successful injectionof a large volume of cement into the fractures. An example
of a cement filled fracture in the Ventilation Shaft niche isshown in Fig.14-23. The inflow of water through the groutedrock mass was quite small. In areas with a large volume of
injected cement, the inflow of water during the excavation
was minimal. The grouting procedure using a shaft jumbofor drilling also resulted in a 6% decrease in the excavation
cycle time, compared with the use of a conventional boringmachine. Constant injection pressure during the last step in the
grouting injection cycle prevented cement backflow from thefractures. Pre-excavation grouting will be the standard method
for reducing the inflow of water. In this way, constructionof the MIU can proceed, while minimizing the cost of water
treatment by reducing the inflow of water.

Reference
Kuji, M., Hara, M., Minamide, K., Takeuchi, S., Mikake, S. et al., Pre-excavation Grouting for Reduction of Water Inflow into a Deep Shaft, Proceedings of the 37th
Symposium on Rock Mechanics, Dai-37-Kai Ganban Rikigaku Ni Kansuru Shimpojiumu Koen Rombunshu, Tokyo, Japan, 2008, p.251-256, in CD-ROM (in Japanese).
JAEA R&D Review 2008

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

Ningyo-toge Environmental Engineering Center ■14-11 To Accomplish Decommissioning of Uranium Enrichment
Plant Safely and Economically–System Treatment Chemical Decontamination Technology by Fluoride Gas–
UF6+IF5 gas+Non-reaction IF7 gas

Exhaustsystem

UF6+IF5Gas
Trap
Cold-Trap

Non-reaction
IF7 gasIF7 gasUraniumenrichment
plant
Compressor

IF7 gas tank(a) Uranium enrichment plant※IF5 is extracted from UF6+ IF5 gas
in the next process(b) Decontamination method

Decontamination time (day)(c) Uranium recovery progress

Quantity of adhered uranium (kgU/unit)

Uranium recovery rate (%)

Fig.14-24 Outline of system treatment chemical decontamination technology
Before system chemical decontamination
After system chemical decontamination

Measurement position(d) Quantity of adhered uranium

Fig.14-25 Results of system treatment chemical decontamination

The Ningyo-toge Environmental Engineering Center is
now entering into the decommissioning stage of the plant.
This center has a gas centrifuge uranium enrichment plant,
which produced the enriched uranium by supplying Uranium
Hexafluoride (UF6) gas into the cascade for about 10 years.
As a result, the uranium compounds have adhered in theequipments of the plant.
Therefore, we have developed the decontamination
technology for uranium enrichment plants. The basic policy forthis development is to decontaminate equipments in the plant
to a level below the clearance level (assumed clearance level:
0.1 Bq/g ).
Our decontamination method is performed by the combinationof two technologies as follows:• System treatment chemical decontamination technology• Sulfuric acid immersing decontamination technology
These are new technologies, and the knowledge has not yet
been reported. However, we could reveal that these technologies
have a very high decontamination performance through ourdevelopment.
In this paper, the system treatment chemical decontamination
technology is explained.
The main uranium compounds in the uranium enrichment
plant are estimated to UFX (4 ≤X≤ 5). We decontaminate theuranium compounds by supplying Iodine Heptafluoride (IF7)gas into the plant because the UFX (4 ≤X≤ 5) is changed to UF6
gas and Iodine Pentafluoride (IF5) gas, by the chemical reaction
such as in equation (1). (Fig.14-24)
UFX(s)+IF7(g) → UF6(g)+IF5(g) .............(1)After the basic experiment, the decontamination of the actual
uranium enrichment plant in our center was performed. The
following results were obtained.
• Decontamination time was about 60 days.
• About 99% of the uranium compounds in the plant could
recovered. (Fig.14-25 (c))• Average of decontamination level was about 1.0 Bq/g.
(Fig.14-25 (d))
As a result, we confirmed that the system treatment chemical
decontamination has a very high performance. Therefore,
we could obtain the prospect that equipments of the plant
are decontaminated below the clearance level by using a
combination of the system treatment chemical decontamination
technology, and sulfuric acid immersing decontamination
technology.
In a few years, the system treatment chemical decontamination
technology we have developed will be moved to the uranium
enrichment plant of private company’s.

Reference
Ema, A. et al., Technology Development of System Chemical Decontamination Using Iodine Heptafluoride Gas -Evaluation of IF7 Treatment Condition and
Uranium Decontamination Result-, JAEA-Technology 2008-037, 2008, 50p. (in Japanese).

112

JAEA R&D Review 2008

■ Aomori Research and Development Center

Development of Experimental Techniques / Facilities at JAEA R&D Centers14-12 High Precision Measurement of Dissolved Organic
Radiocarbon in Seawater–Development of a Pretreatment Method for AMS Measurement of Dissolved Organic Radiocarbon–
Fig.14-27 CO2 purification glass
line
This is the glass line used to collectand purify the CO2 emitted from theultraviolet irradiation step. Liquid
nitrogen is used to collect the CO2.

UV radiation emission source

Liquid nitrogen

Fig.14-26 DOM oxidation reaction
vesselFour liters of seawater was
irradiated with UV radiation, andDOM was oxidized to produce
CO2 in this reaction vessel. The
UV lamp is located in the center.
During the irradiation, the apparatus
is placed in a stainless steel box
for ultraviolet shielding.

Recovery efficiency

pM

(%)

(%)

1

98

79.3

2

96

78.5

3

98

78.0

4

99

77.4

Mean

98

78.3±0.8

An Accelerator Mass Spectrometry (AMS) can measureextremely small amounts of long half-life radionuclides withhigh sensitivity and high precision. An AMS system with theability to measure radiocarbon (14C) and radioiodine (129I) has
been installed in the Mutsu Office of the Aomori Research andDevelopment Center. The characteristics of AMS allow it to be
used not only in fields closely related to nuclear power, such as
for environmental radiation measurements, but also in various
fields such as earth sciences and archaeology.
In recent years, global warming has become a serious problem
and a more accurate understanding of the earth’s carbon cycle
is being sought. Because 14C (half-life 5,730 years) can serve
as a time-scale index for this carbon cycle, it can be said that
14
C measurements using AMS and building a global scale 14C
database is the key to resolving the carbon cycle mechanism.
The organic matter fixed by photosynthesis from atmosphericcarbon dioxide (CO2) plays an important role in the carbon cycle.
However, 14C research on the portion of organic matter that isin the form of dissolved organic matter (DOM) in seawater has
been extremely small compared to 14C research on terrestrialorganic matter or particulate organic matter in seawater. Thisis because the experimental techniques required to measure 14C
in DOM in seawater are difficult, as great effort is required
for shipboard filtering of large amounts of seawater (4ℓ), andfrom samples that contain a salinity more than 10,000 times theconcentration of DOM, it is difficult to extract the carbon only
for the 14C measurements. Particularly for the ocean around
Japan, there are no examples of research. DOM in seawateraccounts for about 20% of the organic matter on the earth’s
surface and more than 90% of the organic matter in the ocean.
In terms of carbon, this amount rivals that of atmospheric CO2.

Table14-2 Experimental precision testing using
surface seawater
These experiments used surface seawater with a
DOM carbon concentration of 64μM. The recovery
efficiency was calculated by the DOM concentration
recovered using the method established by thisstudy and 64μM. pM(%) refers to the proportionof 14C in the sample, if 14C of 1950 is used as thestandard (100%).

Therefore, it is extremely important when considering thecarbon cycle on the surface of the earth. Therefore, there is agreat interest from this perspective in the development of AMS
measurement methods for 14C in seawater DOM.
To resolve these issues, we have established a method todecompose DOM using ultraviolet (UV) irradiation that canbe used with high salinity samples and have combined thatwith AMS. The two-step decomposition method involves first
irradiating DOM samples with UV to produce CO2 (Fig.14-26)
that is then collected and purified (Fig.14-27). Our study of the
UV irradiation time enabled the recovery efficiency to exceed
96%. Improvements to the reaction vessel and glass lines
allowed exclusion of all CO2 other than that from the sample,
and we were successful in reducing contamination to less than
1% of the sample (world’s highest standard). When standard
samples of organic matter were decomposed using this method,
the results for 14C measurement using AMS had a precision of
0.2% (n = 10). These results mean that it is possible to measure
the time scale for DOM with a small measurement error of adecade or so. Experiments with actual seawater showed thatan adequate recovery efficiency (98%) was obtained (Table 14-2).
Furthermore, for DOM in seawater (oldest samples being
about 6,000 years old) that has a low 14C compared to otherenvironmental samples, it was verified that a highly replicable
result (0.8%: a period of about 80 years) was obtained.
As DOM affects (and is affected by) all biological activities
on the earth, there is interest in this method from researchers inmany fields. Measurement of 14C in DOM using this system in
addition to contributing to the correct understanding of global
warming, may also contribute to fundamental research on next
generation resource utilization and environmental forecasting.

Reference
Tanaka, T. et al., Development of Analysis System of Dissolved Organic Radiocarbon and Study on Its Circulation in Sea Water, JAEA-Conf 2008-003,
2008, p.71-74 (in Japanese).
JAEA R&D Review 2008

113

114

JAEA R&D Review 2008

FUGEN

Tokai Research and Development Center

R&D of Radiation
Applications

Fusion R&D, ITER Support

Naka Fusion InstituteExperimental Reactors;
JMTR, JOYO, and HTTR,
R&D of FBR Cycle System
Technology,
Irradiation Test Technology,
High Temperature Gas
Reactor and Hydrogen
Production Technology

Oarai Research and Development Center

Japan Proton Accelerators
Research Complex,
Accelerators and Science

Nuclear Fuel Cycle
Engineering Laboratories

Photon & Synchrotron
Radiation Science

Nuclear Science
Research Institute

Basic Research, Safety Studies, Neutron Science, Nuclear FuelCycle Technologies, Rad-Waste Management and Disposal, etc.

J-PARC Center

Takasaki Advanced Radiation
Research InstituteR&D of Geoscientific Research
Related to Geological Disposal
for HLW

Tono Geoscience Center

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

Aomori Research andDevelopment Center

Kansai Photon
Science InstituteDecommissioningof Uranium
Enrichment Plants

Ningyo-toge Environmental
Engineering Center

MONJU

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

Tsuruga Head Office

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

Horonobe UndergroundResearch Center

R&D Centers of JAEA

JAEA R&D Review 2008115

Photo-Medical Research Center

Nuclear Engineering Research Collaboration Center

Nuclear Cycle Backend Directorate

Geological Isolation Research and Development Directorate

Nuclear Fuel Cycle Technology Development Directorate

Advanced Nuclear System Research and Development Directorate

Fusion Research and Development Directorate

Quantum Beam Science Directorate

Nuclear Science and Engineering Directorate

Advanced Science Research Center

Nuclear Safety Research Center

Research and Development Directorates

Intellectual Resources Department

Takasaki Advanced Radiation Research InstituteTokyo Office

Nuclear Emergency Assistance and Training Center

Ningyo-toge Environmental Engineering Center
Aomori Research and Development Center

Nuclear Technology and Education Center

Tono Geoscience Center‫ޓ‬

Nuclear Nonproliferation Science and Technology Center

Construction Department

Naka Fusion Institute‫ޓޓ‬

Horonobe Underground Research Center‫ޓ‬International Affairs Department

Oarai Research and Development Center‫ޓޓ‬Center for Computational Science & e-Systems

Industrial Collaboration Promotion Department

J-PARC Center‫ޓ‬

Kansai Photon Science Institute‫ޓ‬

Public Relations Department

Tokai Research and Development Center

Safety Administration Department

Tsuruga Head OfficeContract Department

Financial Affairs Department

Labor Relations Department

Personnel Department

Legal Office

Auditor Office

General Affairs Department

Policy Planning and Administration Department

Management Sector

Project Promotion Sector

AuditorResearch and Development Institutes/Centers

President
Executive Vice President
Executive Director

Japan Atomic Energy Agency 㧙Outline of Organization㧙About the Design of the Cover:
The cover is designed to envisage a hopeful future shining in the sky with a blue color derived from the color of the JAEA logo.
This is accompanied with white colored hexagons similar to the pattern in a tortoise shell which symbolizes the wish of people
for longer lives from ancient times in Japan. Coincidentally, this shape is the same as that of the core fuel assemblies for both theprototype “MONJU” Fast Breeder Reactor, and the High Temperature Test Reactor “HTTR”.
The picture in the upper left shows a Compact Laser-driven ion accelerator which has the ability to produce a MeV-energy protonbeam. The picture in the lower right circle shows crystals of uranyl nitrate which were extracted by a rotary driven crystallizer.

JAEA R&D Review 2008

Editorial Board
Chief editor: Hisao Ojima
Vice editor: Shunichi Miyakawa
Editors: Shimpei Matsuhashi, Hideo Kaburaki, Yu Hashimoto, Ken Muramatsu,

Norito Ishikawa, Kaoru Onuki, Takamasa Mori, Yasuhiko Kobayashi,
Atsushi Yokoyama, Hirotaka Kubo, Makoto Ishikawa, Ippei Amamoto,Shinji Yoshikawa, Yoshihiko Nagasato, Yoshifumi Yamaguchi, Hiromichi Ogawa,Michio Yoshizawa, Shintaro Ishiyama, Koji Kawata, Masahiro Ishihara,
Kazunori Hirao, Takuji Kojima, Kaoru Koide, Mitsuhiro Fujiwara
Published by Japan Atomic Energy Agency (JAEA)
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