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Shinohara, Masanori; Ishitsuka, Etsuo; Shimazaki, Yosuke; Sawahata, Hiroaki
JAEA-Technology 2016-033, 65 Pages, 2017/01
JAEA-Technology-2016-033.pdf:11.14MB
To reduce the neutron exposure dose for workers during the replacement works of the startup neutron sources of the High Temperature Engineering Test Reactor, calculations of the exposure dose in case of temporary neutron shielding at the bottom of fuels handling machine were carried out by the PHITS code. As a result, it is clear that the dose equivalent rate due to neutron radiation can be reduced to about an order of magnitude by setting a temporary neutron shielding at the bottom of shielding cask for the fuel handling machine. In the actual replacement works, by setting temporary neutron shielding, it was achieved that the cumulative equivalent dose of the workers was reduced to 0.3 man mSv which is less than half of cumulative equivalent dose for the previous replacement works; 0.7 man mSv.
Information Technology Systems' Management and Operating Office
JAEA-Review 2016-024, 259 Pages, 2017/01
JAEA-Review-2016-024.pdf:31.96MB
Japan Atomic Energy Agency (JAEA) conducts research and development (R&D) in various fields related to nuclear power as a comprehensive institution of nuclear energy R&Ds, and utilizes computational science and technology in many activities. As shown in the fact that about 20% of papers published by JAEA are concerned with R&D using computational science, the supercomputer system of JAEA has become an important infrastructure to support computational science and technology. In FY2015, the system was used for R&D aiming to restore Fukushima (nuclear plant decommissioning and environmental restoration) as a priority issue, as well as for JAEA's major projects such as Fast Reactor Cycle System, Fusion R&D and Quantum Beam Science. This report presents a great number of R&D results accomplished by using the system in FY2015, as well as user support, operational records and overviews of the system, and so on.
Hosoma, Takashi
JAEA-Research 2016-019, 53 Pages, 2017/01
JAEA-Research-2016-019.pdf:5.71MB
Application of probability generating function for nondestructive nuclear materials assay system was studied. First, high-order neutron correlations were derived algebraically up to septuplet and basic characteristics of the correlations were investigated. It was found that higher-order correlation increases rapidly in response to the increase of leakage multiplication, crosses and leaves lower-order correlations behind, when leakage multiplication is 
1.3 that depends on detector efficiency and counter setting. Next, fission rates and doubles count rates by fast neutron and by thermal neutron in their coexisting system were derived algebraically. It was found that the number of induced fissions per unit time by fast neutron and by thermal neutron, the number of induced fissions (
1) by one source neutron, and individual doubles count rates were possible to be estimated from Rossi-alpha combined distribution and measured ratio of each area obtained by differential die-away self-interrogation and conventional assay data.
Tanaka, Masaru*; Aoyama, Isao*; Ishizaka, Kaoru*; Ohata, Yuki*; Fukuike, Iori*; Miyagawa, Hiroshi*; Ishimori, Yuu
JAEA-Research 2016-017, 76 Pages, 2017/01
JAEA-Research-2016-017.pdf:10.57MB
From 1955 to 2001, Japan Atomic Energy Agency (JAEA) carried out research and development projects related to uranium exploration, mining refining, conversion and enrichment at/around Ningyo-toge in Japan. Subsequently, JAEA has been conducting projects related to decommissioning of nuclear fuel cycle facilities and remediation of closed mine sites. JAEA had opportunities of communication with local stakeholders through the projects. Consensus building with local stakeholders and maintain it for over decades are important challenges for JAEA Ningyo-toge Environmental Engineering Center. For this aim, similar domestic case were investigated and compared, and evaluated about required measures for long term relationship with local stakeholder.
Center for Computational Science & e-Systems
JAEA-Evaluation 2016-003, 38 Pages, 2017/01
JAEA-Evaluation-2016-003.pdf:4.56MB
Research on advanced computational science for nuclear applications, based on "Plan to Achieve Medium to Long-term Objectives of the Japan Atomic Energy Agency (Medium to Long-term Plan)", has been performed at Center for Computational Science & e-Systems (CCSE), Japan Atomic Energy Agency. CCSE established the committee consisting outside experts and authorities which does research evaluation and advices for the assistance of the research and development. This report summarizes the followings. (1) Results of the R&D performed at CCSE in FY 2015 (April 1st, 2015 - March 31st, 2016), (2) Results of the evaluation on the R&D by the committee in FY 2015 (April 1st, 2015 - March 31st, 2016)
Katata, Genki; Ota, Masakazu
JAEA-Data/Code 2016-014, 35 Pages, 2017/01
JAEA-Data-Code-2016-014.pdf:1.64MB
In order to predict the impact of atmospheric pollutants (gases and aerosols) to the terrestrial ecosystem, new schemes for calculating the processes of dry deposition of gases and aerosols, and water and carbon cycles in terrestrial ecosystems were implemented in the one-dimensional atmosphere-SOiL-VEGetation model, SOLVEG. We made performance tests at various vegetation areas to validate the newly developed schemes. In this report, the detail in each modeled process is described with an instruction how to use the modified SOLVEG. The framework of "terrestrial ecosystem model" was developed for investigation of a change in water, energy, and carbon cycles associated with global warming and air pollution and its impact on terrestrial ecosystems.
Saito, Hiroshi; Yamaguchi, Masaaki; Kitamura, Akihiro
JAEA-Testing 2016-003, 68 Pages, 2016/12
JAEA-Testing-2016-003.pdf:6.4MB
JAEA has developed a simple and fast simulation program "SACT" (Soil and Cesium Transport) to predict a long-term distribution of Cs deposited on the land surface due to the Fukushima Daiichi Nuclear Power Station accident. It calculates soil movement (erosion, transportation, deposition) and Cs migration, and predicts its future distribution with the assumption that it is adhered to soil. SACT uses USLE (Universal Soil Loss Equation) for potential soil loss and simple equations for soil transportation and deposition. The Cs amount is predicted by the amount of soil movement and Cs concentration ratio for each grain-size of soil. SACT is characterized by its simplicity which enables fast calculation for wide area for long-term duration using existing equations. Data for parameters are widely available and site-specific calculations are possible using data of the targeted area. This manual provides useful and necessary information to users and facilitates the use of SACT widely.
Umeda, Ryota; Kurihara, Akikazu; Shimoyama, Kazuhito
JAEA-Technology 2016-030, 50 Pages, 2016/12
JAEA-Technology-2016-030.pdf:5.22MB
In case of tube failure of a steam generator in sodium-cooled fast reactors, the reaction jet with high temperature and high velocity under highly alkaline environment is formed by cited exothermic reaction (sodium-water reaction). When the high temperature reaction jet covers the adjacent tubes, the material strength of tube decreases in the high temperature condition, and the adjacent tube may be swollen and failed by inner pressure (overheating tube rupture). For evaluation of the overheating tube rupture, tube failure is judged by comparison the hoop stress loaded by inner pressure with stress strength standard defined as creep strength depending on tube temperature. Thus, it is important to confirm the validation of this failure criterion based on the findings obtained in the simulated experiment of overheating tube rupture. In this report, for consideration on the validation of the failure criteria and elucidation on the failure mode and strength characteristics of failure, the authors carried out the rapid heating rupture experiment for the thin single and double-walled 9Cr steel tubes at high temperature up to 1500 K by using TRUST-2 rig in the Japan Atomic Energy Agency.
Tazawa, Yujiro; Nishihara, Kenji; Sugawara, Takanori; Tsujimoto, Kazufumi; Sasa, Toshinobu; Eguchi, Yuta; Kikuchi, Masashi*; Inoue, Akira*
JAEA-Technology 2016-029, 52 Pages, 2016/12
JAEA-Technology-2016-029.pdf:5.34MB
Transmutation Physics Experimental Facility (TEF-P) planned in the J-PARC project uses minor actinide (MA) fuels in the experiments. These MA fuels are highly-radioactive, so the fuel handling equipment in TEF-P is necessary to be designed as remote-handling system. This report summarizes fabrication and test results of the testing equipment for fuel loading that is one of components of the testing equipment for remote-handling of MA fuels. The testing equipment which had a remote-handling system for fuel loading was fabricated. And the test in combination with the mock-up core was performed. Through the test, it was confirmed to load/take the dummy fuel pin to/from the mock-up core without failure. It was shown that the concept design of the fuel loading equipment of TEF-P was reasonable.
Furuse, Takahiro*; Taguchi, Shigeo; Kuno, Takehiko; Surugaya, Naoki
JAEA-Technology 2016-028, 19 Pages, 2016/12
JAEA-Technology-2016-028.pdf:1.79MB
Metal impurities in MOX powder obtained from uranium and plutonium recovered from reprocessing process of spent nuclear fuel are needed to be determined for its characterization. Direct current arc atomic emission spectroscopy (DCA-AES) is one of the useful methods for direct analysis of powder sample without dissolving the analyte into aqueous solution. However, the selection of standard material, which can overcome concerns such as matrix matching, is quite important to create adequate calibration curves for DCA-AES. In this study, we apply standard addition method using the certified U
O
containing known amounts of metal impurities to avoid the matrix problems. The proposed method provides good results for determination of Fe, Cr and Ni at a significant quantity level contained in MOX samples.
Suzuki, Yoshitaka; Ishida, Takuya*; Suzuki, Yumi*; Matsukura, Minoru*; Kurosaki, Fumio*; Nishikata, Kaori; Mimura, Hitoshi*; Tsuchiya, Kunihiko
JAEA-Technology 2016-027, 24 Pages, 2016/12
JAEA-Technology-2016-027.pdf:4.15MB
The research and development (R&D) on the production of 
Mo/
Tc by (n,
) method has been carried out in the Neutron Irradiation and Testing Reactor Center. The Mo production by (n,) reaction is a simple and easy method, and it also is advantageous from viewpoints of nuclear proliferation resistance and waste management. However, it is difficult to produce the Tc solution with high radioactive concentration because the specific radioactivity of Mo by this method is extremely low. Up to now, various Mo absorbents such as Polyzirconium Compound (PZC) and Polytitanium Compound (PTC) have been developed with high Mo adsorption efficiency. It is necessary for utilization to the generator of these absorbents to evaluate the effect of elements containing these absorbents and to assure the quality of Tc solution. In this report, the status of R&D of the Mo adsorbents was investigated. The alumina as Mo adsorbent, which uses in medical Mo/Tc generator, was focused and Mo adsorption/desorption properties of three kinds of alumina was evaluated by different properties such as crystal structure and specific surface.
2015.12Horigome, Kazushi; Suzuki, Hisanori; Suzuki, Yoshimasa; Ishibashi, Atsushi; Taguchi, Shigeo; Inada, Satoshi; Kuno, Takehiko; Surugaya, Naoki
JAEA-Technology 2016-026, 21 Pages, 2016/12
JAEA-Technology-2016-026.pdf:1.14MB
In order to mitigate potential hazards of storage plutonium in solution such as hydrogen generation, conversion of plutonium solution into MOX powder has been carried out since 2014 in the Plutonium Conversion Development Facility. With respect to the samples taken from the conversion process, about 3500 items of plutonium/uranium solutions and MOX powders have been analyzed for the operation control in the related analytical laboratories at the Tokai Reprocessing Plant. This paper describes the reports on analytical activities and related maintenance works in the analytical laboratories conducted from April 2014 to December 2015.
Ishihara, Keisuke; Yokota, Akira; Kanazawa, Shingo; Iketani, Shotaro; Sudo, Tomoyuki; Myodo, Masato; Irie, Hirobumi; Kato, Mitsugu; Iseda, Hirokatsu; Kishimoto, Katsumi; et al.
JAEA-Technology 2016-024, 108 Pages, 2016/12
JAEA-Technology-2016-024.pdf:29.74MB
Radioactive isotope, nuclear fuel material and radiation generators are utilized in research institutes, universities, hospitals, private enterprises, etc. As a result, various low-level radioactive wastes (hereinafter referred to as non-nuclear radioactive wastes) are produced. Disposal site for non-nuclear radioactive wastes have not been settled yet and those wastes are stored in storage facilities of each operator for a long period. The Advanced Volume Reduction Facilities (AVRF) are built to produce waste packages so that they satisfy requirements for shallow underground disposal. In the AVRF, low-level beta-gamma solid radioactive wastes produced in the Nuclear Science Research Institute are mainly treated. To produce waste packages meeting requirements for disposal safely and efficiently, it is necessary to cut large radioactive wastes into pieces of suitable size and segregate those depending on their types of material. This report summarizes activities of pretreatment to dispose of non-nuclear radioactive wastes in the AVRF.
Osaka, Masahiko; Miwa, Shuhei; Nakajima, Kunihisa; Di Lemma, F. G.*; Suzuki, Chikashi; Miyahara, Naoya; Kobata, Masaaki; Okane, Tetsuo; Suzuki, Eriko
JAEA-Review 2016-026, 32 Pages, 2016/12
JAEA-Review-2016-026.pdf:6.18MB
A fundamental research program on fission product (FP) chemistry has started since 2012 for the purpose of establishment of a FP chemistry database in each region of LWR under severe accident and improvement of FP chemical models based on the database. Research outputs are reflected as fundamental knowledge to both the research and development of decommissioning of Fukushima Daiichi Nuclear Power Station (1F) and enhancement of LWR safety. Four research items have thus been established considering the specific issues of 1F and the priority in the source term research area, as follows: effects of boron (B) release kinetics and thermal-hydraulic conditions on FP behavior, cesium (Cs) chemisorption and reactions with structural materials, enlargement of a thermodynamic and thermophysical properties database for FP compounds and development of experimental and analytical techniques for the reproduction of FP behavior and for direct measurement methods of chemical form of FP compounds. In this report, the research results and progress for the year 2015 are described. The main accomplishment was the installation of a reproductive test facility for FP release and transport behavior. Moreover, basic knowledge about the Cs chemisorption behavior was also obtained. In addition to the four research items, a further research item is being considered for deeper interpretation of FP behavior by the analysis of samples outside of the 1F units.
Department of Research Reactor and Tandem Accelerator
JAEA-Review 2016-025, 101 Pages, 2016/12
JAEA-Review-2016-025.pdf:7.86MB
The JAEA-Tokai tandem accelerator complex has been used in various research fields such as nuclear science and material science by researchers not only of JAEA but also from universities, research institutes and industrial companies. This annual report covers developments of accelerators and research activities carried out using the tandem accelerator and superconducting booster from April 1, 2013 to March 31, 2014. Thirty-one summary reports were categorized into seven research/development fields:(1) accelerator operation and development, (2) nuclear structure, (3) nuclear reaction, (4) nuclear chemistry, (5) nuclear theory, (6) atomic physics and solid state physics (7) radiation effects in materials. This report also lists publications, meetings, personnel, committee members, cooperative researches and common use programs.
Hama, Katsuhiro; Iwatsuki, Teruki; Matsui, Hiroya; Mikake, Shinichiro; Ishibashi, Masayuki; Onoe, Hironori; Takeuchi, Ryuji; Nohara, Tsuyoshi; Sasao, Eiji; Ikeda, Koki; et al.
JAEA-Review 2016-023, 65 Pages, 2016/12
JAEA-Review-2016-023.pdf:47.32MB
The Mizunami Underground Research Laboratory (MIU) project is being pursued by the Japan Atomic Energy Agency (JAEA) to enhance the reliability of geological disposal technologies through investigations of the deep geological environment in the crystalline host rock (granite) at Mizunami City in Gifu Prefecture, central Japan. On the occasion of the reform of the entire JAEA organization in 2014, JAEA identified three important issues on the geoscientific research program: "Development of countermeasure technologies for reducing groundwater inflow", "Development of modelling technologies for mass transport" and "Development of drift backfilling technologies", based on the latest results of the synthesizing research and development (R&D). These R&D on three important issues have been carrying out on the MIU project. In this report, the current status of R&D activities and construction in 2015 is summarized.
Ichikawa, Yasuaki*; Kimoto, Kazushi*; Matsui, Hiroya; Kuwabara, Kazumichi; Ozaki, Yusuke
JAEA-Research 2016-018, 23 Pages, 2016/12
JAEA-Research-2016-018.pdf:4.41MB
It is important to evaluate the stability of a repository for high-level radioactive waste not only during the design, construction and operation phases, but also during the post-closure period, for time frames likely exceeding several millennia or longer. The rock mass around the tunnels could be deformed through time in response to time dependent behavior. On the other hand, it was revealed that the chemical reaction of groundwater in a rock had an influence on the long-term behavior. An evaluation of the microcracks to have an influence on this mechanical and chemical coupled phenomena should be worked on chiefly. In fiscal year 2015, using a laser Doppler vibrometer that extends a frequency band up to 20 MHz, and measuring the surface wave transmitted through the granite specimens were estimated group velocity. As a result, group velocity until 100 kHz 500 kHz, revealed that tends to decrease while vibrating. The group speed estimate from a group delay was shown to be easier than the estimate by wave number - frequency spectrum. This is because in order to improve reliability, the estimated frequency band is by using a spatially averaged waveform. As a result obtained, in the case of the modeling by the viscoelastic theory of the granite and a microcrack nondestructiveness evaluation, it is thought that it is useful information in the future. In order to use the knowledge of this study, there is a need to clarify the correspondence between the microscopic properties of the medium such as a crack and crystal grain and the change of the group velocity.
Manabe, Kentaro; Sato, Kaoru; Takahashi, Fumiaki
JAEA-Data/Code 2016-013, 48 Pages, 2016/12
JAEA-Data-Code-2016-013.pdf:1.3MB
JAEA-Data-Code-2016-013-appendix(CD-ROM).zip:0.47MB
In the 2007 Recommendations of the International Commission on Radiological Protection (ICRP), an effective dose is defined as a sum of equivalent doses which are calculated by using male and female reference phantoms based on Caucasian physiological data and averaged over the sexes by tissue weighting factors. Specific absorbed fractions (SAFs), which are essential for internal dosimetry, depend on the body weight and organ masses of phantoms. Then, the dose coefficients, which are committed effective doses per unit intake of radionuclides, developed by ICRP on the basis of the 2007 Recommendations reflect the physical characteristics of Caucasians and are averaged over the sexes. Meanwhile, the physiques of adult Japanese are generally smaller than those of adult Caucasians, and organ masses are also different from each other. Knowledge of the influence of race differences on dose coefficients is important to apply the sex averaged dose coefficients of ICRP to the Japanese system of radiation protection. In this study, SAFs for 25 kinds of mono-energetic electrons and photons ranging from 10 keV to 10 MeV were calculated about the combinations of 67 source regions and 42 target organs using the average adult Japanese female phantom, JF-103, incorporated with a general purpose radiation transport code, MCNPX 2.6.0. The data of this report and the previously published data of JM-103 are applicable to evaluate sex-specific and sex-averaged dose coefficients reflecting the physical characteristics of the average adult Japanese for intakes of all radionuclides not to emit other than photons and electrons.
Kobayashi, Fuyumi; Sumiya, Masato; Kida, Takashi; Kokusen, Junya; Uchida, Shoji; Kaminaga, Jota; Oki, Keiichi; Fukaya, Hiroyuki; Sono, Hiroki
JAEA-Technology 2016-025, 42 Pages, 2016/11
JAEA-Technology-2016-025.pdf:17.88MB
A preliminary test on MOX fuel dissolution for the STACY critical experiments had been conducted in 2000 through 2003 at Nuclear Science Research Institute of JAEA. Accordingly, the uranyl / plutonium nitrate solution should be reconverted into oxide powder to store the fuel for a long period. For this storage, the moisture content in the oxide powder should be controlled from the viewpoint of criticality safety. The stabilization of uranium / plutonium solution was carried out under a precipitation process using ammonia or oxalic acid solution, and a calcination process using a sintering furnace. As a result of the stabilization operation, recovery rate was 95.6% for uranium and 95.0% for plutonium. Further, the recovered oxide powder was calcined again in nitrogen atmosphere and sealed immediately with a plastic bag to keep its moisture content low and to prevent from reabsorbing atmospheric moisture.
Seno, Yasuhiro*; Nakayama, Masashi; Sugita, Yutaka; Tanai, Kenji; Fujita, Tomoo
JAEA-Data/Code 2016-011, 164 Pages, 2016/11
JAEA-Data-Code-2016-011.pdf:8.45MBJAEA-Data-Code-2016-011-appendix(CD-ROM).zip:0.1MB
The cementitious materials are used as candidate materials for the tunnel support of the deep geological repository of high-level radioactive wastes (HLW).Generally the pH of leachate from concrete mixed Ordinary Portland Cement (OPC) shows a range of 12 to 13. The barrier function of bentonite used as a buffer material and that of host rock might be damaged by the high alkaline leachate from cementitious materials. Therefore, low alkalinity that does not damage each barrier function is necessary for cementitious materials used for the tunnel support system of the HLW geological repository. JAEA has developed a low alkaline cement named as HFSC (Highly Fly-ash contained Silicafume Cement) which the pH of the cement leachate could lower approximately 11. We have confirmed the applicability of HFSC for the tunnel support materials, by using experimentally as the shotcreting materials to the part of gallery wall at 140m, 250m and 350m depth in Horonobe Underground Research Laboratory. And moreover, HFSC has been used as the cast-in-place concrete for the shaft lining concrete at the depth of 374m-380m. This Data/Code summarized the past HFSC mix proportion test results about the fresh concrete properties and hardened concrete properties, in order to offer the information as a reference for selecting the mix proportion of HFSC concrete adopted to the disposal galleries et al. in the future.