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Fukuda, Kodai; Obara, Toru*; Suyama, Kenya
Nuclear Technology, 11 Pages, 2024/00
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Takezawa, Hiroki*; Tuya, D.; Obara, Toru*
Nuclear Science and Engineering, 195(11), p.1236 - 1246, 2021/11
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)This study introduces new methodologies for integrating fission reactions induced by delayed neutrons into the Multi-Region Integral Kinetic (MIK) code by using a Monte Carlo neutron transport calculation. First, it was confirmed that it is feasible to solve the Integral Kinetic Model (IKM) with delayed neutrons by the forward Euler discretization method in terms of the number of time steps. This can be done with the help of the law of radioactive decay to reflect the delay in the emission of delayed neutrons in the discretized IKM. Second, a new Monte Carlo-based methodology was introduced for calculating the cumulative distribution functions of secondary fission induced by prompt and delayed neutrons. These functions are necessary for the discretized IKM. The results of preliminary verification using the Godiva reactor confirmed the applicability of the new Monte Carlo-based methodology.
Komeda, Masao; Obara, Toru*
Annals of Nuclear Energy, 65, p.338 - 344, 2014/03
Times Cited Count：5 Percentile：34.81(Nuclear Science & Technology)We studied the core design of neutron transmutation doping (NTD) silicon reactors using MTR (material testing reactor) fuel elements. Silicon ingots are irradiated in the reflector area, outside of the core. A larger core has smaller leakage of neutrons from the core and is able to burn fuel efficiently. However, since a larger core has less neutron leakage from the core, the irradiation efficiency of neutrons in the reflector area is lower. We discuss the most efficient design of a reactor core based on our calculations of fuel consumption and amount of silicon doping.
Komeda, Masao; Kawasaki, Kozo*; Obara, Toru*
Applied Radiation and Isotopes, 74, p.70 - 77, 2013/04
Times Cited Count：8 Percentile：50.81(Chemistry, Inorganic & Nuclear)A new silicon irradiation holder with a neutron filter designed to make the vertical neutron flux profile uniform was studied. Irradiation methods to achieve uniform flux with a filter were discussed using Monte-Carlo calculation code MVP. Validation of the use of the MVP code for the holder's analyses was also discussed via characteristic experiments. It was found that a uniform profile of the vertical flux could be achieved by using the new type holder designed in this study. The vertical uniformity was 
3% when using the new type holder, while it was 19.5% when using the normal holder. By using the new type holder, NTD-Si can be increased to 7 tons per year from 4 tons per year.
C particles for NTD-Si irradiationKomeda, Masao; Obara, Toru*
Annals of Nuclear Energy, 53, p.35 - 39, 2013/03
Times Cited Count：1 Percentile：10.07(Nuclear Science & Technology)We have investigated an alternative irradiation method for silicon doping using a thermal neutron filter to increase the irradiation efficiency in the Japan Research Reactor No. 3 (JRR-3). Because of the restrictions caused by engineering design and radioactivity, aluminum mixed BC particles was adopted as a filter material. The burn-up characteristics of a thermal neutron filter containing a zillion BC particles were described. A method for evaluating neutron transmissivity as boron burning in the neutron filter was developed. The applicable scope of the method via experiments using thermal neutron filters was clarified. A method by which we could calculate the suitable diameter and density of BC particles was derived.
Obara, Toru*; Yamazawa, Yu*; Sasa, Toshinobu
Progress in Nuclear Energy, 53(7), p.1056 - 1060, 2011/09
Times Cited Count：9 Percentile：55.27(Nuclear Science & Technology)Lead-Bismuth eutectic (LBE) has many good characteristics as a coolant for fast reactors. One of the issues remeining to be solved, however, is the polonium issue. The purpose of the present study was to estimate the decontamination performance of a polonium filter by experiment in the penetration condition. Two types of stainless steel wire meshes, fine wire mesh and loose wire mesh, were tested in the experiments. The results show that polonium filters made of stainless steel wire mesh can be very useful device for the removal of polonium in the gas phase. These filters can be used for the decontamination of primary loops by the baking method.
Irwanto, D.*; Chiba, Go; Nagaya, Yasunobu; Obara, Toru*
JAEA-Research 2010-061, 28 Pages, 2011/03
JAEA-Research-2010-061.pdf:1.1MB
It is rather difficult to measure the effective delayed neutron fraction with experiments, so it becomes important to obtain their accurate calculation values. We study calculation methods for the effective delayed neutron fraction by analyzing various fast neutron systems including the bare spherical systems (Godiva, Jezebel, Skidoo, Jezebel-240), the reflective spherical systems (Popsy, Topsy, Flattop-23), MASURCA-R2 and MASURCA-ZONA2, and FCA XIX-1, XIX-2 and XIX-3. These analyses are performed by using SLAROM-UF and CBG for the deterministic method and MVP-II for the Monte Carlo method. We calculate the effective delayed neutron fraction with various definitions such as the fundamental value, the standard definition, Nauchi's definition and Meulekamp's definition, and compare these results with each other.
Obara, Satoshi; Takaya, Shigeru; Wakai, Takashi; Asayama, Tai; Suzuki, Hiroshi; Saito, Toru; Martin, L.*
Kensa Gijutsu, 16(3), p.24 - 30, 2011/03
For the commercialization of fast breeder reactors (FBR), it is essential to enhance the economic competitiveness by reduction of total cost by elongation of plant service period. In this point of view, it is important to establish the assessment method of integrity of aged weld joint and repair welding for the components of future long life FBR. In the present study, evolution of microstructure is evaluated for the 304SS-304SS weld joint which was used for 88,000h at 526-545
C in the French proto-type fast reactor Phenix (as secondary pipes), and for the repair weld joint made from the 304SS of Phenix and new 316LSS plate. In addition, residual stress of the joints were measured by means of RESA and RESA-II. As the results, the microstructure and the residual stress of the joints had changed in the high temperature-long service environment and by the repair welding, and those of the repair weld joint were correlated with its hardness.
Obara, Satoshi; Takaya, Shigeru; Wakai, Takashi; Asayama, Tai; Suzuki, Hiroshi; Saito, Toru*; Martin, L.*
Hozengaku, 9(1), p.32 - 38, 2010/04
For the commercialization of fast breeder reactors (FBR), it is essential to enhance the economic competitiveness by reduction of total cost by elongation of plant service period. In this point of view, it is important to establish the assessment method of integrity of aged weld joint and repair welding for the components of future long life FBR. In the present study, evolution of microstructure is evaluated for the 304SS-304SS weld joint which was used for 88,000h at 526-545C in the French proto-type fast reactor Phenix (secondary pipes), and for the repair weld joint made from the 304SS of Phenix and new 316LSS plate. In addition, residual stress of the joints were measured by means of RESA and RESA-II. As the results, the microstructure and the residual stress of the joints had changed in the high temperature-long service environment and by the repair welding, and those of the repair weld joint were correlated with its hardness.
Sekimoto, Hiroshi*; Igashira, Masayuki*; Yano, Toyohiko*; Obara, Toru*; Osaki, Toshiro*
JAERI-Tech 2002-008, 58 Pages, 2002/03
no abstracts in English
Obara, Toru*; Nakajima, Ken; Miyoshi, Yoshinori; Sekimoto, Hiroshi*
JAERI-Research 2001-037, 60 Pages, 2001/06
JAERI-Research-2001-037.pdf:2.7MB
no abstracts in English
Obara, Toru*; Horiki, Oichiro*; Nakajima, Teruo; Watanabe, Shukichi; Ishijima, Kiyomi; Katanishi, Shoji
JAERI-Review 95-010, 39 Pages, 1995/06
JAERI-Review-95-010.pdf:1.14MB
no abstracts in English
Obara, Toru*; Nakajima, Ken; ; Sekimoto, Hiroshi*; Suzaki, Takenori
JAERI-M 94-004, 40 Pages, 1994/02
no abstracts in English
Irwanto, D.*; Chiba, Go; Nagaya, Yasunobu; Obara, Toru*
no journal, ,
To calculate the effective delayed neutron fraction 
with a Monte Carlo code, Nauchi's method and van der Marck's method have been proposed. The errors due to these methods are quantified by using a deterministic code. It is found that van der Marck's method always results in larger values of than Nauchi's method and that a multi-generation effect, which is neglected in both the methods, has a significant impact on for reflected systems.
Yamazawa, Yu*; Obara, Toru*; Sasa, Toshinobu
no journal, ,
The treatment of polonium, reaction product of bismuth, is one of the important issues for the nuclear systems using lead-bismuth alloy such as spallation target of transmutation experimental facility planned in J-PARC project. To collect polonium evaporated from liquid lead-bismuth, filtering experiment with stainless steel mesh was performed. Using vacuum vessel prepared by JAEA, small amount of lead-bismuth which was irradiated at JRR-4 reactor was heated and evaporated polonium can be collected by the stainless steel mesh located at the entrance of exhaust line. The transmission of polonium was observed as a function of lead-bismuth pot temperature, filter temperature, and filter configuration (mesh size and number). From the experimental results, high decontamination factor is obtained at low filter temperature condition.
Fukuda, Kodai; Suyama, Kenya; Obara, Toru*
no journal, ,
In recent years, efforts have been made in Japan to develop an offshore nuclear power plant with BWR. The effects of the marine environment, such as an inclination, to the reactor behavior are considered to be clarified. In this presentation, the result of the preliminary neutronics - thermal hydraulics coupled analysis, which aims for the clarification of the reactor behavior when BWR is inclined, is reported. The reactor system code TRACE and neutronics code PARCS are used for the analysis. The Peach Bottom 2 is modeled and analyzed for the preliminary analysis.
Obara, Satoshi; Takaya, Shigeru; Wakai, Takashi; Asayama, Tai; Suzuki, Hiroshi; Saito, Toru*; Lauernt, M.*
no journal, ,
For the commercialization of fast breeder reactors, it is considered reduction of total cost by elongation of plant service period. For this reason, it is needed to establish the aged weld joint integrity assessment method for future FBR components. Additionally, since it is expected to employ the repair weld for aged structural components, it is essential to investigate the effect of repair weld. To investigate the influence of high temperature-long service on weld joint, it is examined using microscopy, residual stress analysis and hardness tester. In addition, similar examinations are also conducted for repair weld joint to investigate the effect of repair weld.
