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JAEA Reports

Evaluation of decay heat used for effectiveness evaluations of countermeasures against severe accidents in the prototype FBR Monju

Usami, Shin; Kishimoto, Yasufumi*; Taninaka, Hiroshi; Maeda, Shigetaka

JAEA-Technology 2018-003, 97 Pages, 2018/07

JAEA-Technology-2018-003.pdf:12.54MB

The decay heat used for effectiveness evaluation of the prevention measures against severe accidents in the prototype fast breeder reactor Monju was evaluated by applying the updated nuclear data libraries based on JENDL-4.0, reflecting the realistic core operation pattern, and setting the rational extent of uncertainty. The decay heats of fission products, the actinide nuclides such as Cm-242, and radioactive structural materials were calculated by FPGS code. The decay heat of U-239 and Np-239 was evaluated based on ANSI/ANS-5.1-1994. The calculation uncertainty of each decay heat was evaluated based on summation of uncertainty factors, C/E values of reaction rates obtained in Monju system startup test, and so on. Furthermore, the decay heat evaluation method based on the FPGS90 was verified by the comparison of the results of the decay heat measurement of the two spent MOX fuel subassemblies in the experimental fast reactor Joyo MK-II core.

Journal Articles

A Refined analysis on the power reactivity loss measurement in Monju

Taninaka, Hiroshi; Takegoshi, Atsushi; Kishimoto, Yasufumi*; Mori, Tetsuya; Usami, Shin

Progress in Nuclear Energy, 101(Part C), p.329 - 337, 2017/11

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.

Journal Articles

Validation of decay heat evaluation method based on FPGS cord for fast reactor spent MOX fuels

Usami, Shin; Kishimoto, Yasufumi; Taninaka, Hiroshi; Maeda, Shigetaka

Proceedings of International Conference on the Physics of Reactors; Unifying Theory and Experiments in the 21st Century (PHYSOR 2016) (USB Flash Drive), p.3263 - 3274, 2016/05

The present paper describes the validation of the new decay heat evaluation method using FPGS90 code with both the updated nuclear data library and the rational extent of uncertainty, by comparing the results of the decay heat measurement of the spent fuel subassemblies in Joyo MK-II core and by comparing with the calculation results of ORIGEN2.2 code. The calculated values of decay heat (C) by FPGS90 based on the JENDL-4.0 library were coincident with the measured ones (E) within the calculation uncertainties, and the C/E ranged from 1.01 to 0.93. FPGS90 evaluated the decay heat almost 3% larger than ORIGEN2.2, and it improved the C/E in comparison with the ORIGEN2.2 code. Furthermore, The C/E by FPGS90 based on the JENDL-4.0 library was improved than that based on the JENDL-3.2 library, and the contribution of the revision of reaction cross section library to the improvement was dominant rather than that of the decay data and fission yield data libraries.

Journal Articles

A Scrutinized analysis on the power reactivity loss measurement in Monju

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.

Journal Articles

Basic experiment for kinetics analysis in sub-critical sate

Kitano, Akihiro; Kishimoto, Yasufumi; Misawa, Tsuyoshi*; Hazama, Taira

KURRI Progress Report 2013, 1 Pages, 2014/10

The approach to criticality is conventionally performed by the inverse multiplication method. The method uses neutron count rate at a steady state attained in a certain waiting time after a reactivity insertion; thus it requires long time (for example, several hours from the startup in Monju reactor). We have developed a more efficient method based on Critical Index (CI) featuring the time behavior of delayed neutrons.

Journal Articles

IAEA benchmark calculations on control rod withdrawal test performed during Phenix End-of-Life experiments; JAEA's calculation results

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.

Journal Articles

Criticality evaluation for the Monju restart core

Hazama, Taira; Kitano, Akihiro; Kishimoto, Yasufumi*

Nuclear Technology, 179(2), p.250 - 265, 2012/08

 Times Cited Count:6 Percentile:42.4(Nuclear Science & Technology)

The Japanese prototype fast breeder reactor Monju restarted its system startup test in May 2010 after a 14-year interruption. In the first stage of the test, reactor physics parameters have been measured at a zero power level. The present paper describes the evaluation of the criticality data. The best-estimate value and its uncertainty are evaluated as accurately as possible. The restart core contains 1.5 wt% of $$^{241}$$Am which is three times larger than the previous test. To extract an influence of the $$^{241}$$Am accumulation on calculation accuracy, criticality data obtained in the previous test is evaluated in the same level of detail. The calculation accuracy is investigated with four major nuclear data libraries. It is confirmed that the accuracy is within 0.3%, 2$$sigma$$ value of experimental uncertainty, with JENDL-3.3, JENDL-4.0, and ENDF/B-VII.0. The reactivity change due to the $$^{241}$$Pu decay can be simulated within an accuracy of 1% with JENDL-4.0 and JEFF-3.1.

Journal Articles

Monju reactor physics experiments in the restart core

Kitano, Akihiro; Okawachi, Yasushi; Kishimoto, Yasufumi*; Hazama, Taira

Transactions of the American Nuclear Society, 103(1), p.785 - 786, 2010/11

The Japanese prototype fast breeder reactor Monju has restarted its operation in May, 2010 after 14-year interruption. This paper summarizes reactor physics experiments in the restart core, for criticality, control rod worth, and isothermal temperature coefficient. The largest change from the previous core, a core before the interruption, is in the contents of $$^{241}$$Pu and $$^{241}$$Am. The content of $$^{241}$$Pu has halved and that of $$^{241}$$Am doubled through the $$^{241}$$Pu decay during the interruption. The calculation accuracy on the transition from the previous core to the restart core is investigated. The transition is best simulated with JENDL-4 among three nuclear data; JENDL-3.3, JENDL-4, and ENDF/B-VII. The difference mainly appears in $$^{241}$$Pu fission and $$^{241}$$Am capture cross sections. It is confirmed that the reactor physics data measured in the Monju restart core is valuable to verify nuclear data of the two nuclides.

Journal Articles

Monju core physics test analysis with JAEA's calculation system

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 $$^{240}$$Pu and $$^{241}$$Pu and also for the validity of temperature dependency of the self-shielding using its property as a power reactor.

Oral presentation

Oral presentation

Analyses of Monju core by European reactor analysis system "ERANOS"

Usami, Shin; Kageyama, Takeshi*; Morohashi, Yuko; Kitano, Akihiro; Kishimoto, Yasufumi*; Teruyama, Hidehiko; Nishi, Hiroshi

no journal, , 

no abstracts in English

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