Fujimoto, Nozomu*; Fukuda, Kodai*; Honda, Yuki*; Tochio, Daisuke; Ho, H. Q.; Nagasumi, Satoru; Ishii, Toshiaki; Hamamoto, Shimpei; Nakano, Yumi*; Ishitsuka, Etsuo
JAEA-Technology 2021-008, 23 Pages, 2021/06
The effect of mesh division around the burnable poison rod on the burnup calculation of the HTTR core was investigated using the SRAC code system. As a result, the mesh division inside the burnable poison rod does not have a large effect on the burnup calculation, and the effective multiplication factor is closer to the measured value than the conventional calculation by dividing the graphite region around the burnable poison rod into a mesh. It became clear that the mesh division of the graphite region around the burnable poison rod is important for more appropriately evaluating the burnup behavior of the HTTR core..
Tada, Kenichi; Nagaya, Yasunobu; Kunieda, Satoshi; Suyama, Kenya; Fukahori, Tokio
Journal of Nuclear Science and Technology, 54(7), p.806 - 817, 2017/07
JAEA has developed an evaluated nuclear data library JENDL and several nuclear analysis codes such as MARBLE2, SRAC, MVP and PHITS. Though JENDL and these computer codes have been widely used in many countries, the nuclear data processing system to generate the data library for application programs had not been developed in Japan and foreign nuclear data processing systems, e.g., NJOY and PREPRO are used. To process the new library for JAEA's computer codes immediately and independently, JAEA started to develop the new nuclear data processing system FRENDY in 2013. In this paper, outline, function, and verification of FRENDY are described.
Proceedings of International Conference on the Physics of Reactors; Unifying Theory and Experiments in the 21st Century (PHYSOR 2016) (USB Flash Drive), p.3283 - 3292, 2016/05
The MOSRA system has been developing to improve the applicability of the neutronic characteristic analyses. The cell calculation module MOSRA-SRAC is a core module of MOSRA, and applicability tests for realistic problems are required. As a test, MOSRA-SRAC is validated by comparison with measured values. As the measurement, the post irradiation examination SFCOMPO 99-5 is chosen. In the examination, the compositions of major heavy metal and fission product nuclides in a UO-GdO fuel rod pulled from the 88 BWR fuel assembly used in TEPCO's Fukushima-Daini-2 were measured. The result shows good agreement between calculated and measured value. For uranium and plutonium nuclides, calculated values agree within 5% except for Pu. Pu composition is overestimated by 30%, and the overestimation is caused by the unclearness of the void faction history of the fuel rod. For fission products, calculated values agree within approximately 10%.
JAEA-Data/Code 2015-015, 162 Pages, 2015/10
MOSRA-SRAC is a lattice calculation module of the Modular code System for nuclear Reactor Analyses (MOSRA). This module performs the neutron transport calculation for various types of fuel elements including existing light water reactors, research reactors, etc. based on the collision probability method with a set of the 200-group cross-sections generated from the Japanese Evaluated Nuclear Data Library JENDL-4.0. It has also a function of the isotope generation and depletion calculation for up to 234 nuclides in each fuel material in the lattice. In these ways, MOSRA-SRAC prepares the burn-up dependent effective microscopic and macroscopic cross-section data to be used in core calculations.
JAEA-Review 2014-058, 103 Pages, 2015/03
The following issues were investigated using experimental data of HTTR, which is a Japan's HTGR with 30 MW thermal power. (1)Applicability of nuclear data libraries to nuclear analysis for HTGR, (2) Applicability of the improved nuclear analysis method for HTGR, (3) Effectiveness of a rod-type burnable poison on HTGR reactivity control. Using these investigation results, a nuclear design of a small-sized HTGR with 50 MW thermal power (HTR50S) was performed. In the nuclear design of HTR50S, we challenged to decrease the number of the fuel enrichments and to increase the power density compared with HTTR. As a result, the nuclear design was completed successfully by reducing the number of the fuel enrichment to only three from twelve of HTTR and increasing the power density by 1.4 times of HTTR.
Kashima, Takao; Suyama, Kenya; Takada, Tomoyuki*
JAEA-Data/Code 2014-028, 152 Pages, 2015/03
There have been two versions of SWAT depending on details of its development history: the revised SWAT that uses the deterministic calculation code SRAC as a neutron transportation solver, and the SWAT3.1 that uses the continuous energy Monte Carlo code MVP or MCNP5 for the same purpose. It takes several hours, however, to execute one calculation by the continuous energy Monte Carlo code even on the super computer of the Japan Atomic Energy Agency. Moreover, two-dimensional burnup calculation is not practical using the revised SWAT because it has problems on production of effective cross section data and applying them to arbitrary fuel geometry when a calculation model has multiple burnup zones. Therefore, SWAT4.0 has been developed by adding, to SWAT3.1, a function to utilize the deterministic code SARC2006, which has shorter calculation time, as an outer module of neutron transportation solver for burnup calculation. SWAT4.0 has been enabled to execute two-dimensional burnup calculation by providing an input data template of SRAC2006 to SWAT4.0 input data, and updating atomic number densities of burnup zones in each burnup step. This report describes outline, input data instruction, and examples of calculations of SWAT4.0.
Nojiri, Naoki; Handa, Yuichi*; Shimakawa, Satoshi; Goto, Minoru; Kaneko, Yoshihiko*
Nihon Genshiryoku Gakkai Wabun Rombunshi, 5(3), p.241 - 250, 2006/09
It was shown from the annular core experiment of the HTTR that the discrepancy of excess reactivity between experiment and analysis reached about 3 % Dk/k at maximum. Sensitivity analysis for the annular core of the HTTR was performed to improve the discrepancy. The SRAC code system was used for the core analysis. As the results of the analysis, it was found clearly that the multiplication factor of the annular core is affected by (1) mesh interval in the core diffusion calculation, (2) mesh structure of graphite region in fuel lattice cell and (3) the Benoist's anisotropic diffusion coefficients. The significantly large discrepancy previously reported was reduced down to about 1 % Dk/k by the revised annular core model.
Ando, Masaki; Kawasaki, Kenji*; Okajima, Shigeaki; Fukushima, Masahiro; Matsuura, Yutaka*; Kaneko, Yuji*
JAERI-Research 2005-026, 39 Pages, 2005/09
U Doppler effect measurements in moderated neutron spectra (uranium fuel and MOX simulated fuel) were carried out using FCA for the purpose of contributing to the improvement in prediction accuracy for Doppler coefficient in LWR. In the mockup cores for MOX fuel, the measurements were performed in different neutron spectra, where the voidage of moderator material was varied systematically. The experimental data were obtained using cylindrical uranium samples with different outer diameter up to 800C. Analyses were performed using a standard code system designed to analyze fast reactor mock-up experiments at FCA with the use of the JENDL-3.2 library. The results of the analyses showed that the calculation accuracy did not depend on the types of the core fuel or the Doppler samples. The calculated values agreed with the experimental ones within the experimental error. Any dependency of the prediction accuracy on the neutron spectra was not observed in the MOX simulated fuel cores.
Okuno, Hiroshi; Takada, Tomoyuki
Journal of Nuclear Science and Technology, 41(4), p.481 - 492, 2004/04
Nuclear characteristic parameters were calculated and subcriticality judgement graphs were drawn for revision purposes of the Data Collection for the Nuclear Criticality Safety Handbook. The nuclear characteristic parameters were the neutron multiplication factor in infinite media, migration area and diffusion constants for 11 kinds of typical fuels encountered in criticality safety evaluation of nuclear fuel cycle facilities. These fuels included ADU-HO, UF6-HF and Pu(NO)-UO(NO) solution, of which data were not cited in the Data Collection. The calculation was made with the Japanese evaluated nuclear data library JENDL-3.2 and a sequence of criticality calculation codes, SRAC, POST and SIMCRI. The subcriticality judgement graphs that depict the region satisfying the inequality relation of the neutron multiplication factor less than 0.98 between the two variables (a) uranium enrichment, 239Pu/Pu ratio or plutonium enrichment and (b) H/(Pu+U) ratio were drawn for the same kinds of fuels except UF6-HF in infinite media.
Ando, Masaki; Nakano, Yoshihiro; Okajima, Shigeaki; Kawasaki, Kenji
JAERI-Research 2003-029, 72 Pages, 2003/12
The objectives of this study is to clarify calculation accuracy for the Doppler effect of the resonance materials; erbium (Er), tungsten (W) and thorium (ThO). Doppler effect measurements were carried out in a fast neutron spectrum (XX-2 core) and in an intermediate neutron spectrum (XXI-1D2 core) by the sample-heated and reactivity worth measurement method up to 800C using FCA. The experiment was analyzed with the standard analysis method for fast reactor cores at FCA with the use of the JENDL-3.2. The SRAC system was also used to investigate the calculation accuracy of the system and to compare it with that of the FCA standard analysis method. The standard analysis method underestimated for the XX-2 core and agreed the experiments within the experimental errors for the XXI-1D2 core. The analysis with the SRAC system gave smaller values by 3%10% for the Er sample and bigger values by 2%5% for the W sample than the standard analysis method.
Cao, X.; Suzaki, Takenori; Kugo, Teruhiko; Mori, Takamasa
JAERI-Tech 2003-069, 36 Pages, 2003/08
From the viewpoint of nuclear criticality safety of fuel rod storage and transport, a series of critical experiments concerning effects of water hole size, water gap width, water-to-fuel volume ratio and non-uniform arrangement of water moderator have been performed at the Tank-type Critical Assembly (TCA) of Japan Atomic Energy Research Institute. In the present study, the effects of volume fraction and non-uniform arrangement of water moderator on reactivity are evaluated by the water level worth method and analyzed by the SRAC code. Error sources of experiments and calculations are discussed, especially for an energy group model. The calculation results of diffusion model with 17-group model show good agreement with the experiment results within a few dozen cents.
Sono, Hiroki; Fukaya, Yuji; Yanagisawa, Hiroshi; Miyoshi, Yoshinori
JAERI-Tech 2003-065, 61 Pages, 2003/07
A series of critical experiments using a heterogeneous core of the Static Experiment Critical Facility (STACY) in the Japan Atomic Energy Research Institute is planned in F.Y. 2003. In the experiment, the core is composed of uranyl nitrate solution (U enrichment 6 wt%) and 333 pins of uranium dioxide (U enrichment 5 wt%) loaded in lattice-pitch of 1.5 cm. Prior to the experiment, neutronic characteristics are analyzed to evaluate neutronic safety and criticality limitations of the core. The analyzed items are the parameters on criticality, reactivity and reactor shutdown margins. In the analyses, a Monte Carlo code, MVP, and a neutronics code system, SRAC, have been used with an evaluated nuclear data library, JENDL-3.3. By using the calculated characteristics, simplified equations to interpolate these values and criticality limitations of the core are evaluated. It has been also confirmed that the reactor shutdown margins will comply with safety criteria under all fuel conditions in the experiments.
Ando, Masaki; Nakano, Yoshihiro; Okajima, Shigeaki; Kawasaki, Kenji
Journal of Nuclear Materials, 319, p.126 - 130, 2003/06
Doppler effect experiments on resonance materials for ROX fuels were carried out to examine the calculation accuracy in the intermediate neutron spectrum using Fast Critical Assembly (FCA) at Japan Atomic Energy Research Institute. This study is the second phase of a series of the Doppler effect experiments on the resonance materials, which is following the measurements in the fast neutron spectrum. The Doppler effect was measured as the sample reactivity change between the heated and unheated samples. The cylindrical samples of the resonance materials such as erbium (Er), tungsten (W) and thorium (ThO) were used. The sample was heated up to 800C at the center of the FCA core. The Doppler effect measurements were analyzed using the SRAC 95 code system with the use of JENDL 3.2. The calculated values agreed with the experiment within the experimental error for the W and ThO samples, while the calculation overestimated the experiment for the Er sample about 10 %.
Tonoike, Kotaro; Miyoshi, Yoshinori; Kikuchi, Tsukasa*; Yamamoto, Toshihiro
Journal of Nuclear Science and Technology, 39(11), p.1227 - 1236, 2002/11
Kinetic parameter of low enriched uranyl nitrate solution was measured by the pulsed neutron source method in the STACY. This measurement was repeated systematically over several uranium concentrations from 193.7 gU/ to 432.1 gU/. Used core tanks were two cylindrical tanks whose diameters are 600 mm and 800 mm and one slab tank which has 280 mm thickness and 700 mm width. In this report, experimental data such as solution conditions, critical solution level for each solution condition, subcritical solution levels where measurements were conducted, measured decay time constants of prompt neutron and extrapolated values are described as well as basic principle of the pulsed neutron source method. values were evaluated also by computation with the diffusion code CITATION in SRAC and the nuclear data library JENDL 3.2. Both experimental and computational values show good agreement.
Komukai, Bunsaku; Naka, Michihiro; Tabata, Toshio; Nagao, Yoshiharu; Takeda, Takashi*; Fujiki, Kazuo
JAERI-Tech 2002-067, 75 Pages, 2002/08
no abstracts in English
Hayashi, Takafumi*; Suyama, Kenya; Mochizuki, Hiroki*; Nomura, Yasushi
JAERI-Tech 2001-041, 158 Pages, 2001/06
no abstracts in English
JAERI-Research 2001-003, 29 Pages, 2001/03
no abstracts in English
Nakamura, Takehiko; Takahashi, Masato*; Yoshinaga, Makio
JAERI-Research 2000-048, 77 Pages, 2000/11
no abstracts in English
Suyama, Kenya; Kiyosumi, Takehide*; Mochizuki, Hiroki*
JAERI-Data/Code 2000-027, 88 Pages, 2000/07
no abstracts in English
; Sato, Wakaei*;
JNC TN9400 2000-037, 87 Pages, 2000/03
ln order to compare the nuclear characteristics of water-cooled bleeder cores with that of LMFBR, MOX fuel cell models are established for boiling and non-boiling LWR, non-boiling HWR and sodium-cooled reactor. Frst, the comarison is made between the heterogeneous cell calculation results by SRAC and those by SLAROM. The results show some differences as for neutron energy spectrum, one-grouped cross section and conversion ratio due to the different grouped cross section library (both are based on JENDL-3.2, though) used for each code, however, the difference is acceptably small for grasping the basic characteristics of the above-mentioned cores. Second, using the SLAROM code, main core parameters such as mean neutron energy, ratio of fast neutron and -value, are analyzed. The comparison between the cores show that softened neutron spectrum by the scattering effect of hydrogen or heavy hydrogen increase the contribution of nuclear reaction (especially for neutron capture reaction rather than fission reaction) in lower energy region comparing with LMFBR. ln order to overcome the effect, tighter lattice than LMFBR is necessary for water-cooled cores to realize the breeding of fissile nuclides. Third, effects of Pu isotopic composition on the breeding ratio are evaluated using SRAC burnup calculation. From the results, it is confirmed that degraded Pu (larger ratio of Pu-240) show the larger breeding ratio. At last, sensitivity analyses are made for k-effective and main reaction ratios. As for k-effective, using a temporary covariance data of JENDL-3.2, uncertainty resulting from the cross sections' error is analyzed for a boiling LWR and a sodium-cooled reactor. The boiling LWR core shows larger sensitivity in lower energy region than the sodium-cooled reactor (especially for the energy region lower than 1kev), And, 18-group analysis that is considered acceptably good for LMFBR analysis, should not be enough for accurate sensitivity estimation of ...