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Kimura, Akihiro; Awaludin, R.*; Shiina, Takayuki*; Tanase, Masakazu*; Kawauchi, Yukimasa*; Gunawan, A. H.*; Lubis, H.*; Sriyono*; Ota, Akio*; Genka, Tsuguo; et al.
Proceedings of 3rd Asian Symposium on Material Testing Reactors (ASMTR 2013), p.109 - 115, 2013/11
Tc is generated by decay of Mo. Production of Mo is carried out by (n,f) method with high enriched uranium targets, and the production are currently producing to meet about 95% of global supply. Recently, it is difficult to carry out a stable supply for some problems such as aging of reactors etc. Furthermore, the production has difficulties in nuclear proliferation resistance etc. Thus, (n,) method has lately attracted considerable attention. The (n,) method has several advantages, but the extremely low specific activity makes its uses less convenient than (n,f) method. We proposed a method based on the solvent extraction, followed by adsorption of Tc with alumina column. In this paper, a practical production of Tc was tried by the method with 1Ci of Mo produced in MPR-30. The recovery yields were approximately 70%. Impurity of Mo was less than 4.010% and the radiochemical purity was over 99.2%.
Hirai, Yasushi*; Hyodo, Hideaki*; Tatsumi, Masahiro*; Jin, Tomoyuki*; Yokoyama, Kenji
JAEA-Data/Code 2008-020, 188 Pages, 2008/10
Japan Atomic Energy Agency promotes development of innovative analysis methods and models in fundamental studies for next-generation nuclear reactor systems. In order to efficiently and effectively reflect the latest analysis methods and models to primary design of prototype reactor and/or in-core fuel management for power reactors, a next-generation analysis system MARBLE has been developed. In this study, detailed design of a framework, its implementation and tests are conducted so that a Python system layer can drive calculation codes written in C++ and/or Fortran. It is confirmed that various type of calculation codes such as diffusion, transport and burnup codes can be treated in the same manner on the platform for unified management system for calculation codes with a data exchange mechanism for abstracted data model between the Python and the calculation code layers.
Oyama, Kazuhiro; Kuroha, Takaya*; Takamatsu, Misao; Sekine, Takashi
JNC TN9410 2005-010, 57 Pages, 2005/03
A study of passive safety test using JOYO has been carried out to demonstrate the inherent safety of sodium cooled fast reactors. In this study, emphasis was placed on the improvement of the accuracy of plant kinetics calculations. The Mimir-N2 analysis code, developed to analyze JOYO plant kinetics, was selected as the standard code for predicting plant behavior during transients.Mimir-N2 was previously modified for MK-III core in 2001, 2002. Then, it implemented MK-III performance testing estimate analysis. The MK-III performance test included manual reactor shutdown test and loss of power supply test etc. as transient tests. In order to further improve the accuracy of the calculation, the Mimir-N2 heat transport system models of the reactor vessel upper plenum, the hot leg of secondary heat transport system and the dump heat exchanger were modified based on the results of the MK-III performance test in 2003.In this year, it stores up to get the prospect to have paid to the implementation of the UTOP, the ULOF test which is planned as the passive safety test, it evaluated about the plant structure and UTOP, the ULOF analysis for the parameter of which was the investing reactivity and so on by the Mimir-N2 analysis code. As a result, we could work out the testing condition which has prospect.
Sakai, Takaaki; Kawamura, Takumi*; Miura, Akihiko; Iwasaki, Takashi*
JNC TN9400 2004-053, 77 Pages, 2004/08
Numerical analysis was performed to investigate the electric current concentration to the contact area between the electrode and the noble metal sediments in the glass melter of Tokai reprocessing center. The maximum temperature was evaluated by the simplified model for the current path. In addition, three dimensional temperature profile inside the electrode was calculated numerically, in order to evaluated the possibility to detect the temperature increase by the thermo-couple which settled in the electrode. As a result, current density in the melter showed high concentration at the contact area between the electrode and the sediments. It was found that the maximum temperature at the current path had a possibility to increase more than melting temperature of 1360 Centi-grade. In addition, the thermo-couple inside the electrode scarcely showed sensitivity to the melting temperature at the edge of the electrode header where the failure might occur. In conclusion, the possibility of the electrode melting by the temperature increase can not be excluded as a cause of the failure.
Sakai, Takaaki; Iwasaki, Takashi*; Eguchi, Yuzuru*; Ohshima, Hiroyuki
JNC TN9400 2004-017, 79 Pages, 2004/04
Numerical analysis was performed to validate a design method for a gas entrainment occurrence from a steady free surface vortex. In conclusion, it was clarified that the design method predicted the gas core lengths sufficiently conservative to the experimental data. If the gas entrainment criterion was supposed to be 50mm of the gas core length, it showed good agreement with experimental data.
Hazama, Taira; Chiba, Go; Sato, Wakaei*; Numata, Kazuyuki*
JNC TN9520 2004-001, 97 Pages, 2004/03
A cell calculation code SLAROM-UF was developed to improve calculation accuracy of effective cross sections for various fast reactor types. SLAROM-UF has a capability to calculate effective cross sections in ultra fine groups of about 100,000 below 50keV and in fine groups above the energy (maximum 900 groups), Resonance interaction among the fuel, the coolant, and the structure materials can be treated accurately even in a heterogeneous cell structure. Temperature can be set up freely in a cell by the ultra fine group calculation. Improvement in nuclear characteristics was observed in the analysis of JUPITER critical experiment, as O.1% for criticality, 4% for sodium void reactivity, several % for radial reaction rate distribution, when SLAROM-UF was used insead of the typical cell calculation code. The effect of the ultra fine group calculation is remarkable in the non-leakage term of sodium void reactivity, and that of the fine group calculation is in the case that neutron spectrum in a core can not be represented by the cell calculation. When it is compared with a calculation by continuous energy Monte Carlo code MVP in a homogeneous lattice system, agreement lies within 1% for criticality, a few % for sodium void reactivity, and several % for radial reaction rate distribution of ZPPR-13A whose non-homogeneity is significant. The differences are reduced by about half, from those with the typical cell calculation code. SLRAOM-UF is easily available in the JOINT system currently being used in JNC, including all the functions available in the existent cell calculation code.
Jin, Tomoyuki*; Oki, Shigeo
JNC TN9410 2004-001, 79 Pages, 2004/03
ORLIBJ32 is a set of ORIGEN2 cross-section libraries for light water reactors and fast reactors based on Japanese evaluated nuclear data library JENDL-3.2. It has been opened since 1999. Following the latest revision of JENDL-3.2 to JENDL-3.3 in 2002, we have prepared new ORIGRN2 libraries for fast reactors. By using the same tool that made the fast reactor libraries in ORLIBJ32, the 73-group infinitely-diluted cross sections for 327 nuclides generated from JENDL-3.3 were collapsed into 1-group cross sections with an arbitrary weighting spectrum. For main nuclides, we used the 1-group shielded cross sections obtained by the fast reactor group constant set JFS-3-J3.3. As an isomeric ratio (g/(g+m)) for 241Am capture reaction, the value 0.85 was used instead of the conventional value of 0.80 in consequence of the latest research development of nuclear data. The new libraries were prepared for the following sodium-cooled fast reactors just like ORLIBJ32: JOYO (MK-I), MONJU, several kinds of a prototype reactor (600 MWe) parameterized by both fuel type (MOX, Metal, Nitride) and Pu isotopic composition, a commercial-size reactor (1300 MWe), and a Pu burning reactor. We performed burnup calculations with the new ORIGEN2 libraries in order to investigate the effect caused by the revision of the library.
Igarashi, Minoru; Tanaka, Masaaki; Kimura, Nobuyuki; Nakane, Shigeru*; Kawashima, Shigeyo*; Hayashi, Kenji; Tobita, Akira; Kamide, Hideki
JNC TN9400 2003-092, 100 Pages, 2003/11
A water experiment for thermal hydraulics in a mixing tee was performed to investigate thermal striping phenomena. Measurement of flow velocity using particle image velocimetry and temperature measurement were carried out. Normalized power spectrum density of temperature fluctuation had same profile, when the momentum ratio of the main and branch pips is the same. From the velocity measurement test, when the momentum ratio is the same, flow pattern at mixing region shows the alomost same tendency. Temperature transfer characteristics from fluid to structure can be estimated by a constant heat transfer coefficient in time.
Shono, Akira; Sato, Wakaei*; Hazama, Taira; Iwai, Takehiko*; Ishikawa, Makoto
JNC TN9400 2003-074, 401 Pages, 2003/08
Nuclear design accuracy on the BN-600 hybrid core has been evaluated using the JNC's nuclear analysis system for FBR cores, by utilizing the critical experiment analysis results on BFS-62 configuration that had been obtained under JNC's efforts for Russian surplus weapons plutonium disposition. In the BN-600 hybrid core design, a part of the current UO2 fuel region is replaced by MOX fue1, and the Peripheral blanket region by stainless steel reflectors, respectively. These changes were simulated in a series of critical experiment configurations (BFS-62-1 to 4). Based on the analysis results on both BFS-62 configurations and other fast reactor cores, nuclear design accuracy on the BN-600 hybrid core has been evaluated by applying both the group constant adjustment method and the bias method. Evaluated nuclear parameters include, the criticality, fission rate distribution, sodium void reactivity, control rod worth, burn-up reactivity loss, etc. It is concluded, by applying the group constant adjustment method, that the evaluated accuracy (uncertainty) of most of the nuclear parameters can be decreased to less than half of those based on the basic nuclear constant without reflecting any experimental data. The improvement was mainly achieved by reducing the covariance of the iron elastjc cross section. This significant effect results from the feature of the BN-600 hybrid core, which has relatively larger power density, adopts U235 as the main fissile nucljde, and has the stainless steel reflector surrounding the fuel region. In addition, good consistency of analysis results between the BFS and other fast reactor cores is confirmed. Information obtained by BFS-62 experiment show significant contribution to the accuracy improvement. It is also found that the bias method shows less significant effects on the accuracy improvement than the group constant adjustment method. Furthermore, the bias method may degrade the accuracy for certain nuclear parameters that have large e
Yamano, Hidemasa; Fujita, Satoshi; Tobita, Yoshiharu; Kamiyama, Kenji; Kondo, Satoru; Morita, Koji*; Fischer, E. A.; Brear, D. J.; Shirakawa, Noriyuki*; Cao, X.; et al.
JNC TN9400 2003-071, 340 Pages, 2003/08
An advanced safety analysis computer code, SIMMER-III, has been developed to investigate postulated core disruptive accidents in liquid-metal fast reactors (LMFRs). SIMMER-III is a two-dimensional, three-velocity-field, multiphase, multicomponent, Eulerian, fluid-dynamics code coupled with a space-dependent neutron kinetics model. By completing and integrating all the physical models originally intended at the beginning of this code development project, SIMMER-III is now applicable to integral reactor calculations and other complex multiphase flow problems. A systematic code assessment program, conducted in collaboration with European research organizations, has shown that the advanced features of the code have resolved many of the limitations and problem areas in the previous SIMMER-II code. In this report, the models, numerical algorithms and code features of SIMMER-III Version 3.A are described along with detailed program description. Areas which require future model refinement are also discussed. SIMMER-III Version 3.A, a coupled fluid-dynamics and neutronics code system, is expected to significantly improve the flexibility and reliability of LMFR safety analyses.
Yamano, Hidemasa; Fujita, Satoshi; Tobita, Yoshiharu; Kondo, Satoru; Morita, Koji*; Sugaya, Masaaki*; Mizuno, Masahiro*; Hosono, Seigo*; Kondo, Teppei*
JNC TN9400 2003-070, 333 Pages, 2003/08
An advanced safety analysis computer code, SIMMER-III, has been developed at Japan Nuclear Cycle Development Institute (JNC) to more realistically investigate postulated core disruptive accidents in liquid-metal fast reactors. The two-dimensional framework of SIMMER-III fluid dynamics has been extended to three dimensions to a new code, SIMMER-IV, which is currently (in Version 2) coupled with the three-dimensional neutronics model. With the completion of the SIMMER-IV version, the applicability of the code is further enhanced and the many of the known limitations in SIMMER-III are eliminated. The sample calculations demonstrated the general validity of SIMMER-IV. This report describes SIMMER-IV Version 2.A, by documenting the models, numerical algorithms and code features, along with the program description and input and output information to aid the users.
Kimura, Nobuyuki; Miyake, Yasuhiro*; Miyakoshi, Hiroyuki; Nagasawa, Kazuyoshi*; Igarashi, Minoru; Kamide, Hideki
JNC TN9400 2003-077, 96 Pages, 2003/06
A quantitative evaluation on thermal striping, in which temperature fluctuation due to convective mixing causes high cycle thermal fatigue in structural components, is of importance for structural integrity and reactor safety.Thermal conductivity of sodium is approximately 100 times larger than that of water. Thus, temperature fluctuation characteristics will be different between sodium, which is used as a coolant of a fast reactor, and water, which is used in general industries. In this study, a comparison of convective mixing among jets was performed in parallel triple wall jets with the same geometries between sodium and water. The discharged velocity in the sodium experiment was experimental parameter and set at the same velocity and the same Reynolds number in comparison with the water experiment. And also, the velocity ratio among the triple jets was varied to change flow pattern. It was seen that the water jets were mixed in slightly closer region to the nozzle than in sodium jets. As for the power spectrum densities (PSD) of temperature fluctuation, the PSD of sodium was similar to the PSD of water under the same discharged velocity condition. At the neighborhood of the wall, the lower frequency component in the PSD of sodium decreased in comparison with the PSD of water. It was shown that the amplitude and frequency characteristics obtained by rain-flow method, which was important to evaluate structural damage by the thermal fatigue, were identical between sodium and water overall. These experimental results show that water experiment could simulate the frequency and the amplitude in temperature fluctuation characteristics in the sodium cooled reactor.
Numata, Kazuyuki*; Yokoyama, Kenji; Ishikawa, Makoto
no journal, ,
no abstracts in English
Muramatsu, Toshiharu; Yamada, Tomonori; Nguyen, P. L.; Yoshiuji, Takahiro*; Kondo, Atsuya*; Furutani, Akira*
no journal, ,
no abstracts in English
Sato, Yuji; Shirahama, Takuma*; Ishibashi, Junichi*; Muramatsu, Toshiharu
no journal, ,
no abstracts in English
Suzuki, Risa; Nomi, Takayoshi; Nagatani, Taketeru; Shiromo, Hideo; Shiba, Tomooki; Kaburagi, Masaaki; Okumura, Keisuke; Kosuge, Yoshihiro*; Takada, Akira*; Nauchi, Yasushi*
no journal, ,
no abstracts in English
Kaburagi, Masaaki; Shiba, Tomooki; Okumura, Keisuke; Nauchi, Yasushi*; Suzuki, Risa; Nomi, Takayoshi; Nagatani, Taketeru; Takada, Akira*; Kosuge, Yoshihiro*
no journal, ,
The passive gamma-ray spectroscopy for plutonium nuclear fuels was performed using a small volume CeBr detector specific to high dose-rate measurements. This presentation reports that the wide energy range of gamma-ray spectra was obtained by the spectroscopy, and the energy range covered the principal gamma-decay lines of 59.5 keV emitted from Am and 2615 keV emitted from Tl, which is a descendant nuclide of Pu, comparing the measurements using HPGe.