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.
Kato, Yuko; Yabuki, Kentaro*; Okawachi, Yasushi
JAEA-Technology 2013-018, 118 Pages, 2013/07
The prototype fast breeder reactor MONJU resumed the system startup test (SST) on May 6th 2010 after fourteen years and five months shutdown since the sodium leakage of the secondary heat transport system in December 1995 and reached criticality on May 8th. Core confirmation test (CCT) is the first step of SST which consists of three steps, and finished on July 22nd after 78 days test. Control rod reactivity worth measurements were carried out in order to calibrate the reactivity worth of control rods and back-up rods. In addition, we also aimed at a basic data acquisition for the control rod reactivity worth calibration.
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.
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.
Okawachi, Yasushi; Maeda, Shigetaka; Ito, Chikara; Kawahara, Hirotaka; Aoyama, Takafumi; Ishida, Koichi
JAEA-Technology 2009-047, 130 Pages, 2009/09
This report summarizes the contents about "Reactor physics and plant dynamics experiments using the Joyo simulator" which is one of the training themes. Training is performed using the full scope nuclear reactor simulator for Joyo operation training. While pushing from starting of a nuclear reactor in each experiment of criticality, a control rod proofreading examination, measurement of the temperature of a nuclear reactor, or the reactivity coefficient accompanying output change, feedback reactivity measurement of a fast reactor, etc. and understanding self-regulating characteristics peculiar to a nuclear reactor, the operation of a nuclear reactor can be experienced.
Maeda, Shigetaka; Ito, Chikara; Okawachi, Yasushi; Sekine, Takashi; Aoyama, Takafumi
Reactor Dosimetry State of the Art 2008, p.474 - 482, 2009/00
In 2003 the Joyo reactor upgrade to the MK-III core was completed to increase the irradiation testing capability. This study describes the detail distributions of neutron flux and reaction rate in the MK-III core were experimentally obtained by characterization test during the first two operating cycles. Accuracy of the calculated methods in fast reactor was evaluated by comparing results of DORT, TORT and MCNP. The all calculated reaction rates of U(n,f) agreed well with the measured values about 5% in the fuel region and less than 10% in the reflector region and BC shielding subassembly. However, a large discrepancy more than 10% was observed in the central non-fuel irradiation test subassembly and radial reflector region by DORT and TORT. The MCNP can reduce this discrepancy to 6%. Specific areas of difference, such as uranium fission reaction in non-fuel subassemblies, are understood and correction methods were identified.
Okawachi, Yasushi; Sekine, Takashi; Aoyama, Takafumi
JAEA-Conf 2006-003, p.126 - 139, 2006/05
Twenty eight years of operations at the experimental fast reactor JOYO provide a wealth of experience with core management and characterization of fast neutron field. This experience has been applied to several core modifications to upgrade JOYO's irradiation capability. Reactor physics tests and neutron flux measurement results have been used to confirm the accuracy of neutron diffusion theory calculations. These experiences and accumulated data will be useful for the core design in future fast reactors.
Okawachi, Yasushi; Oki, Shigeo; Wakabayashi, Toshio; Yamaguchi, Kenji*; Yamawaki, Michio*
JNC TY9400 2004-004, 37 Pages, 2004/05
In the framework of the study on transmutation of minor actinide (MA) nuclides with fast reactors, the fission cross section ratios of MA (241Am, 243Am) relative to 235U have been measured by using a back-to-back (BTB) fission chamber at YAYOI fast neutron source reactor. The compact BTB fission chamber was prepared for the measurements in the main experimental hole (Glory hole) passing though the core of YAYOI. Dependence of fission cross section ratios on neutron spectrum was investigated by changing the measurement position in Glory hole from core center to depleted uranium blanket.The measured values of fission cross section ratios were compared with those calculated with the following nuclear data libraries: JENDL-3.2, ENDF/B-VI, and JEF-2.2. It was found that the calculated values for both 241Am and 243Am in the center of the core systematically underestimated the measurements by 10 to 20%. Dispersion of the calculated values among the nuclear data libraries was smaller than the above difference from the measured value. We also see the dependence of C/E value on the measurement position.The present result remained some issues in terms of the measurement accuracy. If we can get rid of those causes, for example, by means of the unfolding technique on pulse height distribution, the result could be utilized as expletive information in nuclear data validation not only for 241Am and 243Am but also for 235U and 238U which characterize the neutron spectrum.
Okawachi, Yasushi; Maeda, Shigetaka; Nagasaki, Hideaki*; Sekine, Takashi
JNC TN9400 2003-029, 96 Pages, 2003/04
The (JOYO) MK-II core characteristics database was compiled and published in 1998. Comments and requests from many users led to the creation of a revised edition in 2001. The revisions include changes to the MAGI calculation code system to use the 70 group JFS-3-J3.2 constant set processed from the JENDL-3.2 library. However, after the database was published, it was recently found that there were errors in the process of making the group constant set JFS-3-J3.2, and lt was revised at JFS-3-J3.2R. Then, the group constant set was updated at JFS-3-J3.2R in this database. The MK-II core management data and core characteristics data were recorded on CD-ROM for user convenience. The structure of the database is the same as in the first edition. The (configuration Data) include the core arrangement and refueling record for each operational cycle. The (Subassembly Library Data) include the atomic number density, neutron fluence, burn-up, integral power of 362 driver fuel subassemblies and 69 irradiation test subassemblies. The (Output Data) contain the calculated neutron flux, gamma flux, power density, linear heat rate, coolant and fuel temperature distribution of all the fuel subassemblies at the beginning and end of each operational cycle. The (Core Characteristics Data) include the measured excess reactivity, control rod worth calibration curve, and reactivity coefficients of temperature, power and burn-up. The effect of updating the group constant set, the calculation results excess reactivity decreased by about 0.15 delta-k/kk' , and the effects to other core characteristics were negligible.
Okawachi, Yasushi; Maeda, Shigetaka; Sekine, Takashi; Nagasaki, Hideaki*
JNC TN9400 2002-070, 49 Pages, 2003/01
As part of the JOYO upgrading program (MK-III program), the JOYO MK-III core management code system "HESTIA" was developed in order to improve the calculation accuracy concerning the core and fuel management and irradiation condition evaluation in the MK-III core. The neutronic calculation of HESTIA was modified to improve the power and neutron flux distribution. The calculation geometry was changed to Tri-Z geometry from Hex-Z geometry which was used in the MK-II core management code system "MAGI". The number of calculation mesh per subassembly was increased to 24 meshes in the radial direction, and the fuel region was divided into 20 meshes in the axial direction. The number of neutron energy group was increased from 7 to 18, and that of gamma energy group was increased from 3 to 7 groups respectively. As a result, HESTIA can accurately calculate the local neutron flux distribution and spectrum change within the fuel subassembly at the boundary between the fuel and reflector regions, which can not be fully simulated by MAGI. It was also confirmed that HESTIA can improve the power distribution in the driver fuel subassembly adjacent to radial reflector. As to thermo-hydraulic calculation, the porous body model was adopted to improve the calculation accuracy of coolant temperature. This model can take into account the detailed power distribution and the turbulent heat transfer in a fuel subassembly. It was found that the calculated value by HESTIA agreed well with that of the subchannel model. In order to verify the calculation accuracy of HESTIA, JOYO MK-II core characteristics calculation was conducted using HESTIA, and calculation results were compared with those of MAGI. The MAGI calculation results were already confirmed by the core performance test and post irradiation examination data. The comparison of both calculation code systems showed that the excess reactivity agreed within 0.01 % k/kk', the maximum neutron flux agreed within 3%, the ...
Okawachi, Yasushi; Maeda, Shigetaka; ; *
JNC TN9410 2001-019, 96 Pages, 2001/12
The JOYO MK-II core characteristics database was compiled and published in 1998. Comments and requests from many users led to the creation of a revised edition. The revisions include changes to the MAGI calculation code system to use the 70 group JFS-3-J3.2 constant set processed from the JENDL-3.2 library. Total control rod worths, reactor kinetic parameters and the MK-II core performance test results were included per user's request. The core characteristics obtained from the 32 to 35th operational cycles, which were conducted in the MK-III transition core, were newly added in this revised version. The MK-II core management data and core characteristics data were recorded on CD-ROM for user convenience. The structure of database is the same as the first edition. The "Configuration Data" include the core arrangement and refueling record for each operational cycle. The "Subassembly Library Data" include the atomic namber density, neutron fluence, burn-up, integral power of 362 driver fuel subassemblies and 69 irradiation test subassemblies. The "Output Data" contain the calculated neutron flux, gamma flux, power density, linear heat rate, coolant and fuel temperature distribution of all the fuel subassemblies at the beginning and end of each operational cycle. The "Core Characteristics Data" include the measured excess reactivities, control rod worth calibration curve, and reactivity coefficients of temperature, power and burn-up.
Okawachi, Yasushi; ; Koshizuka, Seiichi*
JNC TY9400 2001-017, 117 Pages, 2001/05
no abstracts in English
Kai, Tetsuya; Kobayashi, Katsuhei*; Yamamoto, Shuji*; Cho, H.*; Fujita, Yoshiaki*; Kimura, Itsuro*; Okawachi, Yasushi*; Wakabayashi, Toshio*
Annals of Nuclear Energy, 28(8), p.723 - 739, 2001/05
no abstracts in English
Okawachi, Yasushi; Shono, Akira
JAERI-Conf 2001-006, p.121 - 124, 2001/03
JNC TN9400 2001-001, 100 Pages, 2000/08
Gamma-ray decay heat released from fission products has been measured for fast neutron fissions of U-235 and Np-237 using the radiation spectrometry method. The samples were irradiated at fast neutron source reactor "YAYOI" of the University of Tokyo. Gamma-ray energy spectra were measured using a NaI(TI) scintillation detector. And, the number of fission was evaluated from measured gamma spectra by Ge detector. For the measured gamma-ray, the background count was subtracted from the pulse height distribution of 1024 channels measured. The results were grouped by 340 channels to match the response matrix of the detector. This distribution was converted to energy spectra using the FERDO code and the response matrix of the detector. Normalized decay heat by the number of fission was obtained by integration of the energy spectra for each time step. The finite irradiation decay heat that is directly obtained by experiments can not be compared with experimental results and calculational results obtained under various irradiation conditions. So, the finite irradiation decay heat was converted to the fission burst decay heat. These results were compared with summation calculations using JNDC-V2 dacy data file. The present results on U-235 were compared with other experimental data using the same method. The present results agreed with other experimental data using the same method within 10%, suggesting the repeatability of experimental method. The present results on Np-237 were compared with the results of summation calculations using JNDC-V2 decay data file. As the result, the present results agreed with summation calculations within 8%. Probelms to be solved for the future are estimation of the experimental error, re-evaluation of the number of fission using updated nuclear data. To improve accuracy of decay heat data in shorter cooling time range, less irradiation experiment will be useful. Furthermore, to improve accuracy of decay heat data in longer cooloing ...
Kiyanagi, Yoshiaki*; Kamiyama, Takashi*; Hiraga, Fujio*; ; Okawachi, Yasushi
JNC TY9400 2000-015, 238 Pages, 2000/05
no abstracts in English
Suyama, Kenya; Katakura, Junichi; Okawachi, Yasushi*; Ishikawa, Makoto*
Proceedings of the ANS International Topical Meeting on Advances in Reactor Physics and Mathematics and Computation into the Next Millennium (PHYSOR2000) (CD-ROM), p.20 - 0, 2000/00
no abstracts in English
Okawachi, Yasushi; Fukushima, Manabu*
JNC TN9400 99-051, 100 Pages, 1999/05
ORIGEN2 is one of the most widely-used burnup analysis code in the world. This code has one-grouped cross section libraries compiled for various types of reactors. However, these libraries have some problems. One is that these libraries were developed from old nuclear data libraries (ENDF/B-IV,V) and the other is that core and fule designs from which these libraries are generated do not match the current analysis. In order to solve the probrems, analysis tool is developed for generating ORIGEN2 library from JENDL-3.2 considering multi-energy neutron spectrum. And eight new libraries are prepared using this tool for analysis of sodium-cooled FBR. These new libralies are prepared for eight kinds of cores in total. Seven of them are made by changing core size (small core large core), fuel type (oxide, nitride, metal) and Pu vector as a parameter. The eighth one is a Pu burner core. Burnup calculation using both new and original libraries, shows large difference in buildup or depletion numbers of nuclides among the libraries. It is estimated that the analysis result is greatly influenced by the neutron spectrum which is used in collapse of cross section. By using this tool or new libraries, it seems to improve evaluation accuracy of buildup or depletion numbers of nuclides in transmutation research on FBR fuel cycle.