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Mori, Tetsuya; Ohgama, Kazuya; Hazama, Taira
Nuclear Technology, 209(7), p.1008 - 1023, 2023/07
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)In this study, the sodium radioactivity of 
Na and 
Na in the primary system measured in the prototype fast breeder reactor Monju was evaluated, and the reliability of measurements and calculations was examined. The calculated-to-experiment (C/E) values and their uncertainties for Na and Na radioactivities were 0.97-1.07 and 8.1%-11.0% and 1.03-1.16 and 23.3%-24.1%, respectively, using JENDL-4.0 nuclear data library. The Na radioactivity calculated with ENDF/B-VIII.0 was larger by 40% than those calculated with JENDL-4.0 and JEFF-3.3 due to the 
Na(n,2n) cross-section discrepancy. The importance of the Na neutron capture effect was also confirmed herein for the accurate evaluation of the Na radioactivity. The experimental data was judged to be useful for validating the calculation method for improving the reliability of the future designs of sodium-cooled fast reactors.
Mori, Tetsuya; Naganuma, Masayuki; Oki, Shigeo
Nuclear Technology, 209(4), p.532 - 548, 2023/04
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)This paper deals with a conceptual study on a plutonium (Pu) and minor actinide (MA) burning fast reactor core for the distant future phaseout of a fast-reactor fuel cycle after it is commercialized and used for a long time. This burning core aims to reduce the Pu and MA inventories contained in the fuel cycle through multiple recycling. A key point for the core design is the degradation of Pu and MA during multiple recycling. This degradation affects the core feasibility by increasing the sodium void reactivity and decreasing the absolute value of the Doppler constant. A feasible core concept was found by incorporating the following three factors to improve the reactivity coefficients: core flattening, fuel burnup reduction, and the use of silicon carbide (SiC) in the cladding and wrapper tubes. Notably, softening the neutron spectrum using the SiC structural material not only improved the reactivity coefficients but also indirectly mitigated the degradation of Pu and MA. Consequently, the designed core allowed for multiple recycling to continue until the Pu and MA reduced significantly, particularly by about 99% in a phaseout scenario starting from a fast-reactor fleet of 30-GWe nuclear power capacity. Fast reactors were found to have the potential to become self-contained energy systems that can minimize the inventories of Pu they produced themselves, as well as long-lived MA. Fast reactors can be among the important options for environmental burden reduction in the future.
Taninaka, Hiroshi; Takegoshi, Atsushi; Kishimoto, Yasufumi*; Mori, Tetsuya; Usami, Shin
Progress in Nuclear Energy, 101(Part C), p.329 - 337, 2017/11
Times Cited Count：2 Percentile：19.65(Nuclear Science & Technology)The present paper describes the evaluation of the power reactivity loss data obtained in the Japanese prototype fast breeder reactor Monju. The most recent analysis on the power reactivity loss measurement (Takano, et al., 2008) is updated considering the following findings: (a) in-core temperature distribution effect, (b) crystalline binding effect, (c) logarithmic averaging of the fuel temperature, (d) localized fuel thermal elongation effect, (e) updated Japanese Evaluated Nuclear Data Library, JENDL-4.0, and (f) refined corrections on the measured value. The influences of the updates are quantitatively identified and the most precise and probable C/E value is derived together with a thorough uncertainty evaluation. As a result, it is revealed that the analysis overestimates the measurement by 4.6% for the measurement uncertainty of 2.0%. The discrepancy is reduced to as small as 1.1% when the core bowing effect is considered, which implies the importance of the core bowing effect in the calculation of the power reactivity loss.
Taninaka, Hiroshi; Kishimoto, Yasufumi; Mori, Tetsuya; Usami, Shin
Proceedings of International Conference on the Physics of Reactors; Unifying Theory and Experiments in the 21st Century (PHYSOR 2016) (USB Flash Drive), p.2610 - 2621, 2016/05
Reactivity loss due to power ascension (power reactivity loss or power coefficient of reactivity) is thus an important design parameter for determining the number of CRs and plutonium content or inventory in the SFR core design, along with the burnup reactivity loss. Measurements on these reactivity losses were therefore performed during the system startup tests in the Japanese prototype SFR Monju in 1995 and analyses have been carried out for several times. The most recent analysis on the power coefficient measurement in Monju was presented by Takano (Takano, et al., 2008). The following latest findings, which have not been taken into account in the past analyses, are available at present and may affect the existing results: (a) in-core temperature distribution effect, (b) crystalline binding effect, (c) logarithmic averaging of the fuel temperature, (d) localized fuel thermal elongation effect, (e) updated Japanese Evaluated Nuclear Data Library, JENDL-4.0, and (f) refined corrections on the measured value. The influences of refining the calculational models and measured value corrections were therefore quantitatively identified in this study by considering all of these new findings. As a result, it was revealed that the analysis overestimates the experiment by 8.1% for the total uncertainty of 5.9%. Therefore, an additional effect, that is the core bowing effect, was considered in the calculation, and the discrepancy was reduced to 2.9%. The possibility of a significant contribution from the core bowing or deformation effect was thus suggested.
Takano, Kazuya; Mori, Tetsuya; Kishimoto, Yasufumi; Hazama, Taira
Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 13 Pages, 2014/09
This paper describes details of the IAEA/CRP benchmark calculation by JAEA on the control rod withdrawal test in the Phenix End-of-Life Experiments. The power distribution deviation by the control rod insertion/withdrawal, which is the major target of the benchmark, is well simulated by calculation. In addition to the CRP activities, neutron and photon transport effect is evaluated in the nuclear heating calculation of the benchmark analysis. It is confirmed that the neutron and photon transport effect contributes to the improvement of the absolute power calculation results in the breeder blanket region.
Mori, Tetsuya; Maruyama, Shuhei; Hazama, Taira; Suzuki, Takayuki
Nuclear Technology, 179(2), p.286 - 307, 2012/08
Times Cited Count：11 Percentile：63.41(Nuclear Science & Technology)The present paper describes the evaluation of the isothermal temperature coefficient data obtained in the Monju restart core. As in the preceding evaluations on the criticality and the control rod worth, the best-estimate value and its uncertainty are evaluated as accurately as possible. Data obtained in the previous test is evaluated in the same level of detail. The measured data shows that the fuel composition change from the previous test decreases the magnitude of the temperature coefficient by 
8%. Through a sensitivity analysis, it is confirmed that the decrease is mainly brought by the composition of 
Pu and Am. The best accuracy within the experimental uncertainty of 2% is attained for the previous core by a calculation with JENDL-4.0. Results for the restart core show inconsistent behavior and require a further investigation.
Jo, Takahisa; Goto, Takehiro; Yabuki, Kentaro; Ikegami, Kazunori; Miyagawa, Takayuki; Mori, Tetsuya; Kubo, Atsuhiko; Kitano, Akihiro; Nakagawa, Hiroki; Kawamura, Yoshiaki; et al.
JAEA-Technology 2010-052, 84 Pages, 2011/03
JAEA-Technology-2010-052.pdf:17.14MB
The prototype fast breeder reactor MONJU resumed the System Startup Test (SST) on May 6th 2010 after five months and fourteen years shutdown since the sodium leakage of the secondary heat transport system on December 1995. Core Confirmation Test (CCT) is the first step of SST, which consists of three steps. CCT was finished on July 22nd after 78 days tests. CCT is composed 20 test items including control rods' worth evaluation, radiation dose measurement etc..
Takano, Kazuya; Sugino, Kazuteru; Mori, Tetsuya; Kishimoto, Yasufumi*; Usami, Shin
Proceedings of International Conference on the Physics of Reactors, Nuclear Power; A Sustainable Resource (PHYSOR 2008) (CD-ROM), 8 Pages, 2008/09
Monju core physics test analysis was performed using JAEA's neutronics calculation system with various nuclear data libraries (JENDL-3.2, JENDL-3.3, JEFF-3.1, ENDF/B-VII) for the purpose to validate the JAEA's neutronics calculation system, which utilizes JENDL-3.3. Subsequent sensitivity analysis was carried out to clarify the cause of differences in calculation results among nuclear data libraries. It is found that the calculation results obtained by JENDL-3.3 and JAEA's neutronics analysis system showed good agreement with the measured values and its accuracy is identical or better than JEFF-3.1, ENDF/B-VII in most core characteristics. Thus, the validity of JAEA's neutronics analysis system with JENDL-3.3 was confirmed. From the sensitivity analysis, it was identified that Monju can be quite valuable for the verification of the cross sections of such high-order Pu isotopes as 
Pu and Pu and also for the validity of temperature dependency of the self-shielding using its property as a power reactor.
Mizoguchi, Tadanori*; *; Fujisawa, Noboru; Abe, Tetsuya; Hirayama, Toshio; Hitoki, Shigehisa*; *; Koide, Yoshihiko; *; *; et al.
JAERI-M 88-045, 126 Pages, 1988/03
no abstracts in English
Mori, Tetsuya; Sato, Wakaei; Uematsu, Mari; Hazama, Taira; Suzuki, Takayuki
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The previous Monju core physics test is analyzed using a JAEA's neutronics calculation system with various nuclear data libraries, JENDL-3.2, JENDL-3.3, JENDL/AC-2008, JEFF-3.1, and ENDF/B-VII. Analyzed the core characteristics are fixed absorber reactivity worth, fuel sub-assembly reactivity worth, coolant reactivity worth, burn-up coefficient, and reaction rate. It is found that the C/E (calculation over measurement) values are within experimental errors for the fixed absorber reactivity worth and the fuel sub-assembly reactivity worth. But that for the burn-up reactivity coefficient is around the experimental error and shows a tendency of overestimation. Further investigation is ongoing to improve the accuracy of the system. Obtained knowledge and experience will be reflected in the next physics test analyzes.
Kitano, Akihiro; Mori, Tetsuya; Nagata, Akito*; Saito, Kosuke; Misawa, Tsuyoshi*; Tamagawa, Yoichi*
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Mori, Tetsuya; Morohashi, Yuko; Maruyama, Shuhei; Kasahara, Hideyuki; Yabuki, Kentaro; Okawachi, Yasushi
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Morohashi, Yuko; Mori, Tetsuya; Maruyama, Shuhei; Kasahara, Hideyuki; Yabuki, Kentaro; Okawachi, Yasushi
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Mori, Tetsuya
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Mori, Tetsuya; Takano, Kazuya; Kitano, Akihiro; Morohashi, Yuko; Kato, Yuko; Yabuki, Kentaro; Miyagawa, Takayuki; Okawachi, Yasushi; Hazama, Taira
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Monju restarted safely on May 6, 2010 after 14 years and 5 months suspension. Core Confirmation Test was performed until July 22. The core fuel contains Am-241 because Pu-241 (half-life 14 years) decayed during 14.4 years suspension. Therefore, physics data of the core containing Am-241 are obtained. The mainly test items are criticality, control rod worth and isothermal temperature coefficient. In the criticality, the measured CR position at the criticality was confirmed to be within the predicted CR position range. Criticality was predicted in good accuracy. In the control rod worth measurement, CR worth of CCR1 was measured by the period method. CR worth of other CR was measured by the balancing method. In the isothermal temperature coefficient measurement, the measured value was a little bit smaller than that of the previous test due to the accumulation of Am-241, the decay of Pu-241, and other composition change by refuelling.
Mori, Tetsuya; Hazama, Taira; Nishi, Hiroshi
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Yamada, Fumiaki; Kurisaka, Kenichi; Takano, Kazuya; Mori, Tetsuya
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Present status of the safety evaluation for the CDA prevention on the Fast Reactor based on new nuclear safety regulation.
Mori, Tetsuya; Mitsumoto, Rika; Futagami, Satoshi; Suse, Iwao; Enuma, Yasuhiro
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Mori, Tetsuya; Sugino, Kazuteru; Oki, Shigeo
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Mori, Tetsuya; Sugino, Kazuteru; Oki, Shigeo
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