Udagawa, Yutaka; Yamauchi, Akihiro*; Kitano, Koji*; Amaya, Masaki
JAEA-Data/Code 2018-016, 79 Pages, 2019/01
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.
Kitano, Akihiro; Nakajima, Ken*
Proceedings of 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) (CD-ROM), p.1205 - 1210, 2018/04
The feedback reactivity is taken into account in fast reactor core design especially in order to make the power coefficient negative, which is required to be confirmed in the operation. In the feedback reactivity experiment, the positive reactivity was inserted in the critical state at zero power, and the thermal data, such as reactor power and the R/V inlet temperature, was acquired until the power got stable by the feedback reactivity. In the conventional study, only two critical points in an experiment are available for evaluation of the feedback reactivity coefficients. This method needs three days for evaluation. The advanced method based on the inverse kinetics is newly applied in this work using the more extensive data. It is confirmed that this approach can evaluate the feedback reactivity coefficients in one experiment.
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.
Kato, Shinya; Shimomoto, Yoshihiko; Kato, Yuko; Kitano, Akihiro
JAEA-Technology 2014-043, 36 Pages, 2015/02
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.
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.
Miyagawa, Takayuki*; Kitano, Akihiro; Okawachi, Yasushi
JAEA-Technology 2014-008, 60 Pages, 2014/05
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.
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.
Kitano, Akihiro; Nishi, Hiroshi; Suzuki, Takayuki; Okajima, Shigeaki; Kanemoto, Shigeru*
Proceedings of International Conference on Physics of Reactors; Advances in Reactor Physics; Linking Research, Industry, and Education (PHYSOR 2012) (CD-ROM), 14 Pages, 2012/04
The "Synthesis Method", a systematic and sophisticated method of sub-criticality measurement, is proposed in this work to ensure the safety margin before operation. The "Synthesis Method" is based on the modified source multiplication method (MSM) combined with the noise analysis method to measure the reference sub-criticality level for MSM. As a result of numerical simulation, it was suggested that a neutron detector located above the core center and three or more neutron detectors located above the radial blanket region enable the measurement of sub-criticality within 10% uncertainty from -0.5 to -2 and within 15% uncertainty for the deeper sub-criticality.
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.
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
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..
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.
Kinjo, Hidehito*; Kageyama, Takeshi*; Kitano, Akihiro; Usami, Shin
Nuclear Technology, 167(2), p.254 - 267, 2009/08
A conceptual design study has been performed on upgrading the core performance of the Japanese FBR Monju. The main aim of this study is to investigate and demonstrate the feasibility of an upgraded core with an extended refueling interval of 365 EFPD and increased average fuel burnup of 150 GWd/t, which are expected in future commercial FBRs. Two design measures have been taken to accommodate the largely increased burnup reactivity for the longer cycle: (1) A modified fuel pin with increased pin diameter, pellet density and active core height has been introduced to improve the burnup reactivity, (2) The control rod specification has been modified to enhance the reactivity worth by increasing the B content to assure sufficient shutdown margin. The evaluation results show that even a medium sized core of about 2.5 m could achieve the target, without causing significant drawbacks to the core characteristics. The feasibility is thus demonstrated.
Kinjo; Fukuhara; Oshita; Takayama; Kitano, Akihiro; Takao; Yamazaki, Osamu*
JNC-TN4410 2005-004, 243 Pages, 2005/09
Takeda, Toshikazu*; Imai, Hideki*; Kitada, Takanori*; Nishi, Hiroshi; Ishibashi, Junichi; Kitano, Akihiro
Proceedings of International Topical Meeting on Mathematics and Computation, Supercomputing, Reactor Physics and Nuclear and Biological Applications (M&C 2005) (CD-ROM), 12 Pages, 2005/09
A new detailed 3-D transport calculation method taking into account the heterogeneity of fuel assemblies has been developed in hexagonal-z geometry by combining the method of characteristics and the nodal transport method. From the nodal transport calculation which uses assembly homogenized cross sections, the axial leakage is calculated, and it is used for the MOC calculation which treats the heterogeneity of fuel assemblies. Series of homogeneous MOC calculations which use assembly homogeneous cross sections are carried out of obtain effective cross sections, which preserve assembly reaction rates. This effective cross sections are again used in the 3-D nodal transport calculations. The numerical calculations have been performed to verify 3-D radial calculations of FBR assemblies and partial core calculations. Results are compared with the reference Monte-Carlo calculations. A good agreement has been achieved. It is shown that the present method has an advantage in calculating reaction tates in small region.
Maeda, Shigetaka; Sekine, Takashi; Kitano, Akihiro; Nagasaki, Hideaki*
JNC-TN9400 2005-022, 31 Pages, 2005/03
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.
Chiba, Go; Kitano, Akihiro; Maeda, Shigetaka; Sekine, Takashi
JNC-TN9400 2004-057, 90 Pages, 2004/10
The control rod worth was measured in a series of MK-III start-up test of the experimental fast reactor "JOYO". In the test, the control rod worth of all the rods were measured by the inverse kinetics method. Additinal measurements were carried out with a four-rod-juggling method and the neutron multiplication method. The control rod shadowing effect was also measured. Core characteristics related to control rod worth satisfies a safety criteria and agreement between calculated and measured values was observed.
Aoki, Tadao; Sawada, Makoto; ; Matsuno, Yoshiaki*; Kitano, Akihiro; KAMEI, Michiru*
JNC-TN4520 2004-001, 350 Pages, 2004/09
Mitsudo, Seitaro*; Hoshizuki, Hisanori*; Matsuura, Kazunari*; Saji, T.*; Idehara, Toshitaka*; Glyavin, M.*; Eremeev, A.*; Zapevalov, V.*; Kitano, Akihiro; Nishi, Hiroshi; et al.
Proceedings of 29th International Conference on Infrared and Millimeter Waves (IRMMW 2004)/12th International Conference on Terahertz Electronics (THz 2004), p.727 - 728 , 2004/09
Boron carbide(B4C) is one of advanced materials and is being used in a wide rage of applications. The unique feature of this material is its large neutron-absorbing cross-section. Some of its most prominent applications are controlling rods in nuclear reactors and radiation protection. 24 GHz microwave processing for B4C ceramics were performed under flowing argon gas using the sintering system. The sintered samples were characterized by the density, the shrinkage and SEM micrographs of fracture surface. Above the temperature of 2000C, the shrinkage and the grain grows were observed.
Idehara, Toshitaka*; Mitsudo, Seitaro*; Hoshizuki, Hisanori*; Ogawa, Isamu*; Shibahara, Itaru; Nishi, Hiroshi; Kitano, Akihiro; Ishibashi, Junichi
JNC-TY4400 2003-005, 106 Pages, 2003/03
Boron Carbide B4C pellet is an important part of control rod used to control the reactivity of nuclear reactors. B4C pellet put in a nuclear reactor suffers heavy radiation damage and deformation, which result in a partial destruction and shorten the lifetime of B4C. It is important to improve the characteristics of the B4C pellets for extension of its lifetime. As the results, if the control rod without the shroud will be available, we can realize much simpler structure. In order to improve, the B4C pellet, which was sintered by the hot-press methods, we have re-sintered it by high power millimeter wave ceramics nano-indentation test. The increase of the plasticity is observed. The same improvement of plasticity was observed in alumina pellets that were sintered by millimeter wave sintering methods. Such results imply that the further improvement is expected, if the B4C pellet is sintered from powder specimen by the high power millimeter-wave sintering method. To simulate a partial destruction of B4C pellet under the thermal stress, preliminary internal heating experiments of B4C pellet are performed by using high power millimeter-wave. At the difference between internal and surface temperatures of 1000C, the partial destructions and small cracks are observed in B4C pellet. Thes may be a kind of model experiment for destruction of B4C pellet irradiated by neutrons.