Ohgama, Kazuya; Takegoshi, Atsushi; Katagiri, Hiroki*; Hazama, Taira
Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 8 Pages, 2019/05
Takano, Kazuya; Maruyama, Shuhei; Hazama, Taira; Usami, Shin
Proceedings of Reactor Physics Paving the Way Towards More Efficient Systems (PHYSOR 2018) (USB Flash Drive), p.1725 - 1735, 2018/04
Irradiation dependence of the core excess reactivity was investigated for the Monju system startup tests at zero-power carried out in 2010. The excess reactivity basically decreases with the decay of Pu in zero-power operation. However, the excess reactivity little changed in the two month period of the startup tests, which suggests a positive reactivity insertion during the period. The investigated irradiation dependence shows that the positive reactivity increases with reactor operation and mostly saturates by the fission-dose attained during the Monju zero-power operation in a month (10 fissions/cm). The saturated positive reactivity is equivalent to approximately 47% of the initially accumulated self-irradiation damage recovery assuming the defects were recovered by the fission-fragment irradiation in the reactor operation.
Kitano, Akihiro; Takegoshi, Atsushi*; Hazama, Taira
Journal of Nuclear Science and Technology, 53(7), p.992 - 1008, 2016/07
A feedback reactivity measurement technique was developed based on a reactivity model featuring components that depend on the reactivity coefficients, denoted as reactor power (K) and reactor vessel inlet temperature (K). This technique was applied to the feedback reactivity experiment conducted in the Monju system start-up test in May 2010. A thorough evaluation considering all possible biases and uncertainties revealed that the reactivity coefficients can be evaluated with a measurement uncertainty smaller than 3%. The evaluated reactivity coefficients were simulated considering the temperature distribution in the core. The C/E value of K showed good agreement between calculated and measured values within the established uncertainty, and the value of K was consistent with that reported in a previous isothermal temperature coefficient experiment. The measured and calculated fuel subassembly outlet temperatures also agreed well within 0.2C.
Yokoyama, Kenji; Jin, Tomoyuki; Hirai, Yasushi*; Hazama, Taira
JAEA-Data/Code 2015-009, 120 Pages, 2015/07
The second version of the versatile reactor analysis code system, MARBLE2, has been developed. A lot of new functions have been added inMARBLE2 by using the base technology developed in the first version (MARBLE1). Introducing the remaining functions of the conventional code system (JOINT-FR and SAGEP-FR), MARBLE2 enables one to execute almost all analysis functions of the conventional code system with the unified user interfaces of its subsystem, SCHEME. In particular, the sensitivity analysis functionality is available in MARBLE2. On the other hand, new built-in solvers have been developed, and existing ones have been upgraded. Furthermore, some other analysis codes and libraries developed in JAEA have been consolidated and prepared in SCHEME. In addition, several analysis codes developed in the other institutes have been additionally introduced as plug-in solvers. Consequently, -ray transport calculation and heating evaluation become available. As for another subsystem, ORPHEUS, various functionality updates and speed-up techniques have been applied based on user experience of MARBLE1 to enhance its usability.
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.
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.
Takeda, Toshikazu*; Hazama, Taira; Fujimura, Koji*; Sawada, Shusaku*
Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 15 Pages, 2014/09
A national project started in 2013 in Japan entitled "technology development for the environmental burden reduction". The present study is one of the studies adopted as the national project. We are aiming to develop MA transmutation core concepts harmonizing MA transmutation performance with core safety and to improve design accuracy related to MA transmutation performance. To validate and improve design accuracy of the high safety and high MA transmutation performance of SFR cores, we develop methods for calculating the transmutation rate of individual MA nuclides and estimating uncertainty of MA transmutation by using burnup sensitivity. Also we develop reliable reactor physics database to reduce the uncertainty of MA transmutation calculations. The overall consistency of the measured data is investigated by evaluating the usefulness of conventional static data as well as those related to MA transmutation obtained from various facilities like Monju, Joyo, FCA, BFS and PFR.
Yokoyama, Kenji; Hazama, Taira; Numata, Kazuyuki*; Jin, Tomoyuki*
Annals of Nuclear Energy, 66, p.51 - 60, 2014/04
A comprehensive and versatile reactor analysis code system, MARBLE, has been developed. MARBLE is designed as a software development frame-work for reactor analysis, which offers reusable and extendible functions and data models based on physical concepts, rather than a reactor analysis code system. From a viewpoint of the code system, it provides a set of functionalities utilized in a detailed reactor analysis scheme for fast criticality assemblies and power reactors, and nuclear data related uncertainty quantication such as cross-section adjustment. MARBLE covers all phases required in fast reactor core design prediction and improvement procedures, i.e. integral experiment database management, nuclear data processing, fast criticality assembly analysis, uncertainty quantication, and power reactor analysis. In the present paper, these functionalities are summarized and system validation results are described.
Hazama, Taira; Yokoyama, Kenji
Journal of Nuclear Science and Technology, 50(5), p.525 - 533, 2013/05
A versatile nuclear heating calculation system is developed in MARBLE, a neutronics calculation platform based on the object-oriented language. The system provides a variety of heating calculation methods from the conventional simple methods based on energy-independent Q-values, to a detailed method based on NJOY. A particular feature is in hybrid methods that consider energy dependence employed in NJOY in the system. The hybrid methods provide tools to assess errors in the simple methods and to evaluate the heat transport property by neutrons and photons precisely. The accuracy of the hybrid methods and the errors in the simple methods are investigated in a Monju design calculation. It is confirmed that the hybrid methods realize the nuclear heating by the detailed method within an accuracy of 0.1%. The neglect of incident energy dependence of the fission Q-value in the simple methods causes clear overestimation in the Q-value but its influence on the total heating is less than 1%.
Kitano, Akihiro; Miyagawa, Takayuki*; Okawachi, Yasushi; Hazama, Taira
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Safe Technologies and Sustainable Scenarios (FR-13) (USB Flash Drive), 9 Pages, 2013/03
The feedback reactivity was measured in Monju start-up test conducted in 2010. The two reactivity components related either to power or to the core inlet coolant temperature were evaluated by fitting to a reactivity balance equation as a function of neutron count rate and coolant temperature. The measured feedback reactivity and the two components were compared with calculation taking account of the temperature distribution in the core. The calculated and the measured values of the feedback reactivity showed a reasonable agreement.
Hazama, Taira; Kitano, Akihiro; Kishimoto, Yasufumi*
Nuclear Technology, 179(2), p.250 - 265, 2012/08
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 Am which is three times larger than the previous test. To extract an influence of the 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 value of experimental uncertainty, with JENDL-3.3, JENDL-4.0, and ENDF/B-VII.0. The reactivity change due to the Pu decay can be simulated within an accuracy of 1% with JENDL-4.0 and JEFF-3.1.
Mori, Tetsuya; Maruyama, Shuhei; Hazama, Taira; Suzuki, Takayuki
Nuclear Technology, 179(2), p.286 - 307, 2012/08
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.
Takano, Kazuya; Fukushima, Masahiro; Hazama, Taira; Suzuki, Takayuki
Nuclear Technology, 179(2), p.266 - 285, 2012/08
The present paper describes the evaluation of the control rod worth data obtained in the Monju restart core. The best-estimate value and its uncertainty are evaluated in detail. As in the criticality evaluation, data obtained in the previous test is evaluated in the same level of detail. The correlation in the uncertainties is also evaluated among different control rods and tests of the previous and the restart cores. Based on the evaluated data, calculation accuracy is investigated with JENDL-3.3 and JENDL-4.0. It is confirmed that the calculation accuracy is within the experimental uncertainty of 2% for each layer and B content. A reduction in the uncertainty related to the delayed neutron fraction is effective for a further improvement in the calculation accuracy.
Sugino, Kazuteru; Ishikawa, Makoto; Numata, Kazuyuki*; Iwai, Takehiko*; Jin, Tomoyuki*; Nagaya, Yasunobu; Hazama, Taira; Chiba, Go*; Yokoyama, Kenji; Kugo, Teruhiko
JAEA-Research 2012-013, 411 Pages, 2012/07
Aiming at evaluating the core design prediction accuracy of fast reactors, various kinds of fast reactor core experiments/tests have been analyzed with the Japan's latest evaluated nuclear data library JENDL-4.0. Totally 643 characteristics of reactor physics experiments/tests and irradiation tests performed using the critical facilities: ZPPR, FCA, ZEBRA, BFS, MASURCA, ultra-small cores of LANL and power plants: SEFOR, Joyo, Monju were dealt. In analyses, a standard scheme/method for fast reactor cores was applied including detailed or precise calculations for best estimation. In addition, results of analyses were investigated from the viewpoints of uncertainties caused by experiment/test, analytical modeling and cross-section data in order to synthetically evaluate the consistency among different cores and characteristics. Further, by utilizing these evaluations, prediction accuracy of core characteristics were evaluated for fast power reactor cores that are under designing in the fast reactor cycle technology development (FaCT) project.
Sugino, Kazuteru; Jin, Tomoyuki*; Hazama, Taira; Numata, Kazuyuki*
JAEA-Data/Code 2011-017, 44 Pages, 2012/01
Fast reactor group constant sets UFLIB.J40 and JFS-3-J4.0 were prepared, which are based on the latest Japanese evaluated nuclear data library JENDL-4.0. Concerning UFLIB.J40, several fine group constant sets, which covered 70-group, 73-group, 175-group and 900-group structures, and the ultra fine group constant set were prepared. The number of nuclides for cross-sections of lumped fission products was extended so as to follow the extension of the number of fissile species for fission yield data.
Chiba, Go*; Hazama, Taira; Kinjo, Hidehito*; Nishi, Hiroshi; Suzuki, Takayuki
JAEA-Research 2011-034, 42 Pages, 2011/12
Uncertainty of Doppler coefficient is quantified for a Monju core, reflecting present knowledge. Various uncertainty sources are evaluated: (1) Nuclear data and numerical method, (2) Fission product nuclear data, (3) Control rod position, (4) Approximated treatment of temperature dependence, (5) Averaged fuel temperature, (6) Approximated treatment of temperature spatial distribution, etc. Resulting uncertainty for Doppler coefficient is estimated at 11.7% for the 2 reliability.
Sugino, Kazuteru; Ishikawa, Makoto; Yokoyama, Kenji; Nagaya, Yasunobu; Chiba, Go; Hazama, Taira; Kugo, Teruhiko; Numata, Kazuyuki*; Iwai, Takehiko*; Jin, Tomoyuki*
Journal of the Korean Physical Society, 59(2), p.1357 - 1360, 2011/08
In order to improve the prediction accuracy of core performances in the fast reactor core design study, the unified cross-section set has been developed in Japan. The unified cross-section set, which combines a wide range of integral experimental information with differential nuclear data, is produced by using the cross-section adjustment technique based on the Bayesian parameter-estimation theory. A new set ADJ2010 is currently under development. The present paper describes the results of the cross-section adjustment for ADJ2010 which is based on the JENDL-4.0 data. The evaluation of the core design accuracy for a commercial power fast reactor core is also discussed. ADJ2010 will be released soon and will be expected to be utilized for core design of future fast reactors.
Hazama, Taira; Takano, Kazuya; Kitano, Akihiro
Proceedings of 2011 International Congress on Advances in Nuclear Power Plants (ICAPP '11) (CD-ROM), p.1527 - 1535, 2011/05
The Japanese prototype fast breeder reactor Monju restarted its reactor physics test in May, 2010 after a 14-year interruption. The accumulation of Am due to the Pu decay during the interruption reaches 1.5wt% in average. An impact of the reactor physics data obtained in the restart core is investigated by the cross section adjustment technique with JENDL-3.3 and JENDL-4.0. Criticality data obtained before and after the interruption are applied. It is confirmed that Monju reactor physics data, when the two data are used together, effectively adjust Am capture cross sections. Consistent results are obtained among JENDL-3.3 after adjustment and JENDL-4.0 before and after the adjustment.
Yokoyama, Kenji; Tatsumi, Masahiro*; Hirai, Yasushi*; Hyodo, Hideaki*; Numata, Kazuyuki*; Iwai, Takehiko*; Jin, Tomoyuki*; Hazama, Taira; Nagaya, Yasunobu; Chiba, Go; et al.
JAEA-Data/Code 2010-030, 148 Pages, 2011/03
A next generation reactor analysis code system, MARBLE, has been developed. MARBLE is a successor of the fast reactor neutronics analysis code systems, JOINT-FR and SAGEP-FR (conventional system), which were developed for so-called JUPITER standard analysis methods. MARBLE has the equivalent analysis capability to the conventional system. On the other hand, burnup analysis functionality for power reactors as improved compared with the conventional system. In the development of MARBLE, the object oriented technology was adopted. As a result, MARBLE became an assembly of components for building an analysis code (i.e. framework) but not an independent analysis code system which simply receives input and returns output. Furthermore, MARBLE provides some pre-built analysis code systems such as the fast reactor neutronics analysis code system, SCHEME, which corresponds to the conventional code and the fast reactor burnup analysis code system, ORPHEUS.
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 Pu and Am. The content of Pu has halved and that of Am doubled through the 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 Pu fission and 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.