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Journal Articles

Model updates and performance evaluations on fuel performance code FEMAXI-8 for light water reactor fuel analysis

Udagawa, Yutaka; Amaya, Masaki

Journal of Nuclear Science and Technology, 56(6), p.461 - 470, 2019/06

no abstracts in English

Journal Articles

Development of granulation system for simplified MOX pellet fabrication process

Ishii, Katsunori; Segawa, Tomoomi; Kawaguchi, Koichi; Suzuki, Masahiro

Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 5 Pages, 2019/05

Japan Atomic Energy Agency (JAEA) is developing a simplified pelletizing process for MOX fuel fabrication. In this process, the flowability of MOX powder produced by de-nitration conversion based on microwave heating, calcination, and reduction is improved using the wet granulation method. In a previous paper, to produce MOX granules of appropriate sizes for pelletizing them effectively, we proposed a granulation system composed of a wet granulator and a sizing machine. In the present work, we modernized the wet granulator, completed the granulation system by adding auxiliary equipment, and conducted performance tests of the granulation system with WO$$_{3}$$ powder. The results of a performance test indicated that it is possible to convert raw powder into granules characterized by appropriate size and excellent flowability. The time required to process 5 kg of WO$$_{3}$$ powder was about 70 min, which almost satisfies the target time.

Journal Articles

The Effects of plutonium content and self-irradiation on thermal conductivity of mixed oxide fuel

Ikusawa, Yoshihisa; Morimoto, Kyoichi; Kato, Masato; Saito, Kosuke; Uno, Masayoshi*

Nuclear Technology, 205(3), p.474 - 485, 2019/03

 Percentile:100(Nuclear Science & Technology)

This study evaluated the effects of plutonium content and self-irradiation on the thermal conductivity of mixed-oxide (MOX) fuel. Samples of UO$$_{2}$$ fuel and various MOX fuels were tested. The MOX fuels had a range of plutonium contents, and some samples were stored for 20 years. The thermal conductivity of these samples was determined from thermal diffusivity measurements taken via laser flash analysis. Although the thermal conductivity decreased with increasing plutonium content, this effect was slight. The effect of self-irradiation was investigated using the stored samples. The reduction in thermal conductivity caused by self-irradiation depended on the plutonium content, its isotopic composition, and storage time. The reduction in thermal conductivity over 20 years' storage can be predicted from the change of lattice parameter. In addition, the decrease in thermal conductivity caused by self-irradiation was recovered with heat treatment, and recovered almost completely at temperatures over 1200 K. From these evaluation results, we formulated an equation for thermal conductivity that is based on the classical phonon-transport model. This equation can predict the thermal conductivity of MOX fuel thermal conductivity by accounting for the influences of plutonium content and self-irradiation.

JAEA Reports

Development of fuel performance code FEMAXI-8; Model improvements for light water reactor fuel analysis and systematic validation

Udagawa, Yutaka; Yamauchi, Akihiro*; Kitano, Koji*; Amaya, Masaki

JAEA-Data/Code 2018-016, 79 Pages, 2019/01

JAEA-Data-Code-2018-016.pdf:2.75MB

FEMAXI-8 is the latest version of the fuel performance code FEMAXI developed by JAEA. A systematic validation work has been achieved against 144 irradiation test cases, after many efforts have been made, in development of new models, improvements in existing models and the code structure, bug-fixes, construction of irradiation-tests database and other infrastructures.

Journal Articles

Modeling and simulation of redistribution of oxygen-to-metal ratio in MOX

Hirooka, Shun; Kato, Masato; Watanabe, Masashi

Transactions of the American Nuclear Society, 118, p.1624 - 1626, 2018/06

This study suggested the time development of oxygen-to-metal ratio (O/M) redistribution model with oxygen-related properties in MOX. Irradiation simulation including the suggested O/M redistribution and pore migration with vaporization-condensation model which bares density redistribution was demonstrated. The simulation results showed that O/M redistribution proceeded at lower temperature than density redistribution, which indicated that oxygen diffusion got influential at lower temperature than vaporization-condensation of MOX. Another find was that O/M redistribution was very slow at the surface because temperature kept low. However, near the surface (inside from the surface) where the temperature exceeded 1000 K, O/M redistribution was rather recognizable with oxygen flown from inner region to the near-surface. The results will be evaluated by comparison with post-irradiation examination data.

Journal Articles

Sound speeds in and mechanical properties of (U,Pu)O$$_{2-x}$$

Hirooka, Shun; Kato, Masato

Journal of Nuclear Science and Technology, 55(3), p.356 - 362, 2018/03

 Times Cited Count:1 Percentile:38.14(Nuclear Science & Technology)

The sound speeds of longitudinal and transverse waves in the uranium-plutonium mixed oxide (MOX) pellets were measured as functions of porosity, oxygen-to-metal ratio (O/M) and plutonium content. The effect of each parameter was well fitted by a linear function and the equations were obtained to calculate the sound speeds. Mechanical properties were evaluated with the sound speeds and the result of Young's modulus showed that porosity was the most important factor to decrease Young's modulus. Temperature dependence on Young's modulus was also evaluated with previously reported thermal expansion. Decrease of Young's modules with increasing temperature was in good agreement with available literature.

Journal Articles

Radiative neutron capture on $$^{242}$$Pu in the resonance region at the CERN n_TOF-EAR1 facility

Lerendegui-Marco, J.*; Guerrero, C.*; Mendoza, E.*; Quesada, J. M.*; Eberhardt, K.*; Junghans, A. R.*; Kimura, Atsushi; n_TOF Collaboration*; 126 of others*

Physical Review C, 97(2), p.024605_1 - 024605_21, 2018/02

Journal Articles

Fabrication and short-term irradiation behaviour of Am-bearing MOX fuels

Kihara, Yoshiyuki; Tanaka, Kosuke; Koyama, Shinichi; Yoshimochi, Hiroshi; Seki, Takayuki; Katsuyama, Kozo

NEA/NSC/R(2017)3, p.341 - 350, 2017/11

In order to investigate the effect of the addition of americium to MOX fuels on the irradiation behaviour, the "Am-1" program is being conducted at the JAEA. The Am-1 program consists of two short-term irradiation tests of 10-min and 24-h irradiation periods, and a steady-state irradiation test. The short-term irradiation tests and their post irradiation examinations (PIEs) have been successfully completed. To date, the data for PIE of the Am-MOX fuels focused on the microstructural evolution and redistribution behaviour of Am at the initial stage of irradiation have been obtained and reported. In this paper, the results obtained from the Am-1 program are reviewed and detailed descriptions of the fabrication and inspection techniques for the Am-MOX fuels prepared for the program are provided. PIE data for the Am-MOX fuels at the initial stage of irradiation have been accumulated. In this paper, unpublished PIE data for the Am-MOX fuels are also presented.

JAEA Reports

Preparation of uranium and plutonium mixed spike optimized for MOX analysis by isotope dilution mass spectrometry

Horigome, Kazushi; Taguchi, Shigeo; Yamamoto, Masahiko; Kuno, Takehiko; Surugaya, Naoki

JAEA-Technology 2017-016, 20 Pages, 2017/07

JAEA-Technology-2017-016.pdf:1.68MB

Mixed spikes of uranium and plutonium have been prepared for the determination of uranium and plutonium in dissolved MOX solution by isotope dilution mass spectrometry. Enriched uranium metal NBL CRM116 and plutonium metal NBL CRM126 were accurately weighed and then dissolved in nitric acid, respectively. Their dissolved solutions were mixed in a mass ratio of 1 to 2. The preparation values of uranium and plutonium were 1.0530 $$pm$$ 0.0008 mg/g (k=2) of uranium with a $$^{235}$$U relative mass fraction of 93.114 wt% and 2.0046 $$pm$$ 0.0019 mg/g (k=2) of plutonium with a $$^{239}$$Pu relative mass fraction of 97.934 wt%, respectively. The concentrations of uranium and plutonium in spike were confirmed by reverse isotope dilution mass spectrometry using tracer of $$^{233}$$U and $$^{242}$$Pu. Finally, the prepared spike was validated by parallel analysis of simulated sample of dissolved MOX solution. This spike was applied to measure the uranium and plutonium amount content of dissolved MOX solutions using isotope dilution mass spectrometry.

Journal Articles

Oxide-metal ratio dependence of central void formation of mixed oxide fuel irradiated in fast reactors

Ikusawa, Yoshihisa; Maeda, Koji; Kato, Masato; Uno, Masayoshi*

Nuclear Technology, 199(1), p.83 - 95, 2017/07

 Times Cited Count:1 Percentile:64.68(Nuclear Science & Technology)

Based on thermal computation results obtained using an irradiation behavior analysis code, we have evaluated the effect of O/M ratio on fuel restructuring from the results of PIEs for the B14 irradiation test fuel, which was a mixed oxide fuel and was irradiated in the experimental reactor Joyo. The thermal computation results showed that fuel restructuring in the stoichiometric oxide fuel was accelerated, though the fuel temperature in the stoichiometric oxide fuel was evaluated as lower than that of the hypo-stoichiometric one. We explained this behavior as follows: first, the fuel temperature decreased due to the high thermal conductivity at stoichiometry; second, the pore migration velocity increased due to the increase in vapor pressure caused by the high vapor pressure of UO$$_{3}$$, which was derived from the high oxygen potential at stoichiometry. In addition, our results indicated that the central void diameter strongly depended on not only fuel temperature, but also vapor pressure.

Journal Articles

Flushing phenomena and flow structure by microwave heating

Fujita, Shunya*; Abe, Yutaka*; Kaneko, Akiko*; Chonan, Fuminori*; Yuasa, Tomohisa*; Yamaki, Tatsunori*; Segawa, Tomoomi; Yamada, Yoshikazu

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 8 Pages, 2017/07

From the observation results, in the process of flushing, the behaviors leading to flushing were classified divided into three types. First type is that first generation bubble from heating leads to flushing. Second type is that nucleate boiling continues during heating and stop, finally single bubble generates and leads to flushing. Third type is defined that gradual evaporation occurs without bubbles. It was revealed that the total quantities of heat released by flushing are approximately equal when assuming the flushing mechanism, it can be triggered that a large amount of micro bubbles are instantaneously generated and grew.

Journal Articles

Mechanical and thermal properties of (U,Pu)O$$_{2-x}$$

Hirooka, Shun; Kato, Masato

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 6 Pages, 2017/06

Young's modulus of MOX pellets was evaluated by measuring the sound velocities of longitudinal and transverse waves in the pellets as functions of porosity, oxygen-to-metal ratio (O/M) and plutonium content. The results showed that porosity was the most important factor that 20% of the porosity decreased Young's modulus by neatly 100 GPa while O/M and plutonium content could change the Young's modulus by ~20 GPa. From the measured sound velocities, temperature dependence on Young's modulus and specific heat capacity were calculated on the Debye model by leveraging the thermal expansion data. The temperature dependence that Young's modulus decreases with increasing temperature is in good agreement with literature data. The specific heat capacity also agrees with that of calculated value by Kopp's method, taken the Schottky term and the excited term into account.

Journal Articles

Current status of the next generation fast reactor core & fuel design and related R&Ds in Japan

Maeda, Seiichiro; Oki, Shigeo; Otsuka, Satoshi; Morimoto, Kyoichi; Ozawa, Takayuki; Kamide, Hideki

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 10 Pages, 2017/06

The next generation fast reactor is being investigated in Japan, aiming at several targets such as "safety", "reduction of environmental burden" and "economic competitiveness". As for the safety aspect, FAIDUS concept is adopted to avoid re-criticality in core destructive accidents. The uranium-plutonium mixed oxide fuel, in which minor actinide elements are included, will be applied to reduce the amount and potential radio-toxicity of radioactive wastes. The high burn-up fuel is pursued to reduce fuel cycle cost. The candidate concept of the core and fuel design, which could satisfy various design criteria by design devisals, has been established. In addition, JAEA is investigating material properties and irradiation behavior of MA-MOX fuel. JAEA is developing the fuel design code especially for the fuel pin with annular pellets of MA-bearing MOX. Furthermore, JAEA is developing oxide dispersion strengthened (ODS) ferritic steel cladding for the high burnup fuel.

Journal Articles

Oxygen chemical diffusion coefficients of (U, Pu)O$$_{2-x}$$

Watanabe, Masashi; Sunaoshi, Takeo*; Kato, Masato

Defect and Diffusion Forum, 375, p.84 - 90, 2017/05

The oxygen chemical diffusion coefficient in (U, Pu)O$$_{2-x}$$ was determined by thermo-gravimetry as functions of the Pu content, oxygen-to-metal ratio and temperature. The surface reaction was considered in the diffusion coefficient determination. The activation energy for the chemical diffusion coefficient was 60 kJ/mol and 65 kJ/mol, respectively, in (U$$_{0.8}$$Pu$$_{0.2}$$)O$$_{2-x}$$ and (U$$_{0.7}$$Pu$$_{0.3}$$)O$$_{2-x}$$.

Journal Articles

Oxygen potentials, oxygen diffusion coefficients and defect equilibria of nonstoichiometric (U,Pu)O$$_{2pm x}$$

Kato, Masato; Watanabe, Masashi; Matsumoto, Taku; Hirooka, Shun; Akashi, Masatoshi

Journal of Nuclear Materials, 487, p.424 - 432, 2017/04

 Times Cited Count:1 Percentile:64.68(Materials Science, Multidisciplinary)

Oxygen potential of (U,Pu)O$$_{2pm x}$$ was evaluated based on defect chemistry using an updated experimental data set. The relationship between oxygen partial pressure and deviation $$x$$ in (U,Pu)O$$_{2pm x}$$ was analyzed, and equilibrium constants of defect formation were determined as functions of Pu content and temperature. Brouwer's diagrams were constructed using the determined equilibrium constants, and a relational equation to determine O/M ratio was derived as functions of O/M ratio, Pu content and temperature. In addition, relationship between oxygen potential and oxygen diffusion coefficients were described.

JAEA Reports

Determination of metal impurities in MOX powder by direct current arc atomic emission spectroscopy; Application of standard addition method for direct analysis of powder sample

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$$_{3}$$O$$_{8}$$ 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.

Journal Articles

Development of Pu standard material preparation and characterization technique in Japan

Okazaki, Hiro; Shibano, Koya; Abe, Katsuo; Sumi, Mika; Kayano, Masashi; Kageyama, Tomio; Mason, P.*

Proceedings of INMM 57th Annual Meeting (Internet), 7 Pages, 2016/07

The reference materials used for accountancy analysis by IDMS, referred to as LSD spikes allows to obtain accurate measurement results for a variety of nuclear materials under potentially difficult handling or measurement conditions. Because of expected difficulties in a stably long term supply of plutonium reference materials, which are the main source of plutonium for LSD spikes preparation, the JAEA's PFDC decided to evaluate the possibility of using domestically available MOX powders as a plutonium source material for LSD spikes. Therefore, PFDC started to separate, purify and characterize plutonium in MOX powders for plutonium reference materials (MOX-Pu) preparation that are suitable for source of LSD spikes in cooperation with the US DOE's NBL. The detailed preparation procedure of the latest MOX-Pu standard material and results of intercomparison for verification are presented.

Journal Articles

The Research of MOX fuels in Japan

Kato, Masato

Transactions of the American Nuclear Society, 114, p.987 - 988, 2016/06

In Japan, uranium and plutonium mixed oxide (MOX) has been developed as fuels of sodium-cooled fast reactors. The developing MOX fuels come in variety of O/M ratio, Pu content, minor actinide (MA) content and density. We have studied a science based fuel technology to evaluate fuel behaviors in fabrication process and irradiation condition of such various fuels. The technologies which are constructed based on experimental database can apply to mechanistic evaluation of fuel behaviors. To develop the science based fuel technology, many different varieties of basic properties have been investigated, and experimental database was constructed. And a mechanistic physical property model has been studied. The models contribute to describe various behaviors in fuel fabrication process and irradiation condition.

Journal Articles

Thermal expansion measurement of (U,Pu)O$$_{2-x}$$ in oxygen partial pressure-controlled atmosphere

Kato, Masato; Ikusawa, Yoshihisa; Sunaoshi, Takeo*; Nelson, A. T.*; McClellan, K. J.*

Journal of Nuclear Materials, 469, p.223 - 227, 2016/02

 Times Cited Count:4 Percentile:31.55(Materials Science, Multidisciplinary)

Thermal expansion of (U$$_{0.7}$$Pu$$_{0.3}$$)O$$_{2-x}$$ (x = 0, 0.01, 0.02, 0.03) and (U$$_{0.52}$$Pu$$_{0.48}$$)O$$_{2.00}$$ was measured with a dilatometer in an oxygen partial pressure-controlled atmosphere. The oxygen partial pressure was controlled to hold a constant oxygen-to-metal ratio in the (U,Pu)O$$_{2-x}$$ during the measurement. Thermal expansion slightly increased with the decrease in oxygen-to-metal ratio. The relationship was derived to describe thermal expansion.

JAEA Reports

Report on the evaluation of research and development activities in FY2014 issue; "Research and Development on Reprocessing of Nuclear Fuel Materials" (Ex-post evaluation)

Tokai Reprocessing Technology Development Center

JAEA-Evaluation 2015-012, 83 Pages, 2015/12

JAEA-Evaluation-2015-012.pdf:6.67MB

Japan Atomic Energy Agency (hereafter referred as "JAEA") consulted the "Evaluation Committee of Research and Development Activities for Fast Reactor Cycle" to assess the issue on "Research and Development on Reprocessing of Nuclear Fuel Materials" conducted by JAEA during the period from FY2010 to FY2014. In response to the JAEA's request, the committee assessed the R&D programs and the activities of JAEA related to the issue and concluded the mission was accomplished. This evaluation was performed based on the "General guideline for the evaluation of government R&D activities", the "Guideline for evaluation of R&D in Ministry of Education, Culture, Sports, Science and Technology (MEXT)" and the "Operational rule for evaluation of R&D activities" by JAEA.

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