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.

Chapter 5, Transmutation system, 5.1 Transmutation system by fast reactor, 5.1.2 Experience in experimental reactor and prototype reactor and future research plan, 5.1.2(2) Prototype fast breeder reactor Monju

Kitano, Akihiro

Bunri Henkan Gijutsu Soron, p.215 - 222, 2016/09

Prototype fast breeder reactor Monju has restarted in May 2010 after the long shutdown since the sodium leak accident. Before the restart the safety authority has examined the influence of the long shutdown, such as the characteristics of Am accumulated in MOX fuel. The integral data of the core with about 1.5% was acquired in the core confirmation test conducted in 2010 and the newest analysis method shows the enough accuracy. The irradiation test for the MA transmutation is planned, called GACID project by Japan, France and United States. The research of MA transmutation is one of the most important issues in Monju research plan included in the basic research plan decided by the cabinet council. It is expected that Monju would be utilized to demonstrate MA transmutation in the fast reactor.

Development of three-dimensional diffusion and burn-up code HIZER for Monju core management

Kato, Shinya; Shimomoto, Yoshihiko; Kato, Yuko; Kitano, Akihiro

JAEA-Technology 2014-043, 36 Pages, 2015/02

JAEA-Technology-2014-043.pdf:8.94MB

The core management and operation code system aims to perform core management task efficiently by systematic management of data, analyses and edits, which are needed in the reactor core management and operation. The system consists of the five calculation modules: the reactor constant generation module, the neutronic-thermal calculation module, the radiation analysis module, the core structural integrity estimation module, and the core operation analysis module. In these modules, the neutronic-thermal calculation module is based on the dedicated three-dimensional diffusion and burn-up code HIZER. HIZER can execute core calculations easily for specific design specification and operation patterns of Monju, enabling efficient and accurate evaluation of the Monju core characteristics. This report describes its calculation method and validation results.

Monju system start-up test report evaluation of the feedback reactivity

Miyagawa, Takayuki*; Kitano, Akihiro; Okawachi, Yasushi

JAEA-Technology 2014-008, 60 Pages, 2014/05

JAEA-Technology-2014-008.pdf:29.75MB

The prototype fast breeder reactor Monju resumed the system startup test (SST) on May 6, 2010 after fourteen year and five month shutdown since the sodium leakage of the secondary heat transport system in December 1995 and reached criticality on May 8, 2010. Core Confirmation Test (CCT) is the first step of SST which consists of three steps, and finished on July 22 after 78 days test. In the evaluation of the feedback reactivity at the part of the CCT, the "self-stability" of Monju was observed when the positive reactivity was inserted with the control rod withdrawal, due to the negative feedback property of the reactor, and due to the control properties of the auxiliary cooling system. Parameters represented with reactor power, sodium temperature of the primary loops became to be stable after transient without any operations. Additionally, the quantitative feedback reactivity was evaluated using the results of this test tentatively.

Criticality evaluation for the Monju restart core

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

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

https://doi.org/10.13182/NT12-A14097

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.

Adjustment of Am cross section with Monju reactor physics data

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.

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 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.

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Kinjo; Fukuhara; Oshita, Hironori; Takayama; Kitano, Akihiro; Takao; Yamazaki, Osamu*

JNC TN4410 2005-004, 243 Pages, 2005/09

JOYO MK-III Performance Test; Criticality Test, Excess Reactivity Measurement and Burn-up Coefficient Measurement

Maeda, Shigetaka; Sekine, Takashi; Kitano, Akihiro; Nagasaki, Hideaki*

JNC TN9400 2005-022, 31 Pages, 2005/03

JNC-TN9400-2005-022.pdf:2.41MB

The MK-III performance test began in June 2003 to fully characterize the upgraded core and heat transfer system. This paper describes the results of the approach to criticality, excess reactivity evaluation and burn-up coefficient measurement. In the approach to criticality test, the MK-III core achieved the initial criticality at control rod bank position of 412.8 mm on 14:03 July 2nd, 2003. Because the replacement of the outer two rows of reflector subassemblies with shielding subassemblies reduced source range monitor signals by a factor of 3 at the same reactor power, we measured the change of detector response and decide the count rate to define zero power criticality as 2 10 cps. In the excess reactivity evaluation, Based on the measured critical rod bank position and the measured control rod worths, the zero power excess reactivity at 250C was 2.990.10%k/kk'. The prediction by JOYO core management code system HESTIA, which is 3.130.16% k/kk', agree with the measured value well. On the other hand, the measured excess reactivity was within a safety requirement limit. In the burn-up coefficient measurement, the reactivity change against the reactor burn-up. The measurement method was adopted to measure control rod position at rated power operation. Finally, -2.1210 k/kk'/Mwd was obtained as a measured of burn-up coefficient. The calculated value by HESTIA became -2.1210 k/kk'/Mwd and it agree with the measured value well. The measurements are emphasized but comparisons with calculated predictions are included. The design predictions are consistent with the performance test results, and all technical safety specifications are satisfied.

International Sodium Handling Technology Training Course

Aoki, Tadao; Sawada, Makoto; Oshita, Hironori; Matsuno, Yoshiaki*; Kitano, Akihiro; KAMEI, Michiru*

JNC TN4520 2004-001, 350 Pages, 2004/09

JNC-TN4520-2004-001.pdf:36.76MB

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