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Malins, A.; Lemoine, T.*
Journal of Open Source Software (Internet), 7(71), p.3318_1 - 3318_6, 2022/03
Riyana, E. S.; Okumura, Keisuke; Terashima, Kenichi; Matsumura, Taichi; Sakamoto, Masahiro
Mechanical Engineering Journal (Internet), 7(3), p.19-00543_1 - 19-00543_8, 2020/06
Riyana, E. S.; Okumura, Keisuke; Terashima, Kenichi
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 4 Pages, 2019/05
Nagaya, Yasunobu; Okumura, Keisuke; Sakurai, Takeshi; Mori, Takamasa
JAEA-Data/Code 2016-018, 421 Pages, 2017/03
JAEA-Data-Code-2016-018.pdf:3.89MB
JAEA-Data-Code-2016-018-appendix(CD-ROM).zip:4.02MBJAEA-Data-Code-2016-018-hyperlink.zip:1.94MB
In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two Monte Carlo codes MVP (continuous-energy method) and GMVP (multigroup method) have been developed at Japan Atomic Energy Agency. The codes have adopted a vectorized algorithm and have been developed for vector-type supercomputers. They also support parallel processing with a standard parallelization library MPI and thus a speed-up of Monte Carlo calculations can be achieved on general computing platforms. The first and second versions of the codes were released in 1994 and 2005, respectively. They have been extensively improved and new capabilities have been implemented. The major improvements and new capabilities are as follows: (1) perturbation calculation for effective multiplication factor, (2) exact resonant elastic scattering model, (3) calculation of reactor kinetics parameters, (4) photo-nuclear model, (5) simulation of delayed neutrons, (6) generation of group constants, etc. This report describes the physical model, geometry description method used in the codes, new capabilities and input instructions.
Inaba, Yoshitomo; Nishihara, Tetsuo
Annals of Nuclear Energy, 101, p.383 - 389, 2017/03
Times Cited Count：7 Percentile：56.46(Nuclear Science & Technology)In order to ensure the thermal integrity of fuel in High Temperature Gas-cooled Reactors (HTGRs), it is necessary that the maximum fuel temperature in normal operation is to be lower than a thermal design target. In the core thermal-hydraulic design of block-type HTGRs, the maximum fuel temperature should be evaluated considering data such as thermal power, core geometry, power density and neutron fluence distributions, and core coolant flow distribution. The fuel temperature calculation code used in the design stage of the High Temperature engineering Test Reactor (HTTR) presupposes to run on UNIX systems, and its operation and execution procedure are complicated and are not user-friendly. Therefore, a new fuel temperature calculation code named FTCC which has a user-friendly system such as a simple and easy operation and execution procedure, was developed. This paper describes calculation objects and models, basic equations, improvement points from the HTTR design code in FTCC, and the result of a validation calculation with FTCC. The calculation result obtained by FTCC provides good agreement with that of the HTTR design code, and then FTCC will be used as one of the design codes for HTGRs. In addition, the effect of cooling forms on the maximum fuel temperature is investigated by using FTCC. As a result, it was found that the effect of center hole cooling for hollow fuel compacts and gapless cooling with monolithic type fuel rods on reducing the temperature is very high.
Shinohara, Masanori; Ishitsuka, Etsuo; Shimazaki, Yosuke; Sawahata, Hiroaki
JAEA-Technology 2016-033, 65 Pages, 2017/01
JAEA-Technology-2016-033.pdf:11.14MB
To reduce the neutron exposure dose for workers during the replacement works of the startup neutron sources of the High Temperature Engineering Test Reactor, calculations of the exposure dose in case of temporary neutron shielding at the bottom of fuels handling machine were carried out by the PHITS code. As a result, it is clear that the dose equivalent rate due to neutron radiation can be reduced to about an order of magnitude by setting a temporary neutron shielding at the bottom of shielding cask for the fuel handling machine. In the actual replacement works, by setting temporary neutron shielding, it was achieved that the cumulative equivalent dose of the workers was reduced to 0.3 man mSv which is less than half of cumulative equivalent dose for the previous replacement works; 0.7 man mSv.
Ishitsuka, Etsuo; Kenzhina, I. E.*; Okumura, Keisuke; Takemoto, Noriyuki; Chikhray, Y.*
JAEA-Technology 2016-022, 35 Pages, 2016/10
JAEA-Technology-2016-022.pdf:3.73MB
As a part of study on the mechanism of tritium release to the primary coolant in research and testing reactors, the calculation methods by PHITS code is studied to evaluate the recoil tritium release rate from beryllium core components. Calculations using neutron and triton sources were compared, and it is clear that the tritium release rates in both cases show similar values. However, the calculation speed for the triton source cases is two orders faster than that for the neutron source case. It is also clear that the calculation up to history number per unit volume of 2
10
(cm
) is necessary to determine the recoil tritium release rate of two effective digits precision. Furthermore, the relationship between the beryllium shape and recoil tritium release rate using the triton sources was studied. Recoil tritium release rate showed linear relation to the surface area per volume of beryllium, and the recoil tritium release rate showed about half of the conventional equation value.
Kashima, Takao; Suyama, Kenya; Takada, Tomoyuki*
JAEA-Data/Code 2014-028, 152 Pages, 2015/03
JAEA-Data-Code-2014-028.pdf:13.39MB
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.
Takahashi, Fumiaki; Endo, Akira; Yamaguchi, Yasuhiro
Journal of Nuclear Science and Technology, 42(4), p.378 - 383, 2005/04
Times Cited Count：3 Percentile：24.17(Nuclear Science & Technology)Experiments were made to verify a dose assessment method from activated sodium in body in criticality accidents. A phantom containing sodium chloride solution was irradiated in the Transient Experiment Critical Facility to simulate activation of sodium. Monte Carlo calculations were performed to obtain quantitative relation between the activity of induced Na-24 and neutron dose in the phantom. In the previous work, conversion coefficients from specific activity of induced Na-24 to neutron dose had been analyzed with the MCNP-4B code concerning neutron spectra at some hypothesized configurations. One of the prepared coefficients was applied to evaluate neutron dose from the measured activity. The estimated dose agreed with the dose analyzed by the Monte Carlo calculation in the present study within an acceptable uncertainty, which is indicated by the IAEA. In addition, the dose calculated with the prepared coefficient was close to the result measured with dosimeters. These results suggest that the prepared coefficients can be applied to dose assessments from induced Na-24 in body.
Iijima, Susumu*; Kato, Yuichi*; Takasaki, Kenichi*; Okajima, Shigeaki
JAERI-Data/Code 2004-016, 91 Pages, 2004/12
JAERI-Data-Code-2004-016.pdf:7.45MB
The calculation code system "EXPARAM" was designed to analyze the experimental results systematically measured at the fast critical assembly (FCA). Some calculation codes developed independently in JAERI and in US research institutes were collected and arranged as the fast reactor physics calculation code system. The multi-group core calculation code and the perturbation calculation code based on the diffusion theory and the transport theory calculate the reactor physics parameters such as eigenvalue, reaction rate, Doppler reactivity worth and sodium void worth. The dynamic physics parameters such as prompt neutron lifetime and effective delayed neutron fraction are also calculated. Input and Output data of calculation codes are transferred to each other using a direct access file on UNIX computer system.
Suyama, Kenya; Mochizuki, Hiroki*; Okuno, Hiroshi; Miyoshi, Yoshinori
Proceedings of International Conference on Physics of Fuel Cycles and Advanced Nuclear Systems; Global Developments (PHYSOR 2004) (CD-ROM), 10 Pages, 2004/04
This paper provides validation results of SWAT2, the revised version of SWAT, which is a code system combining point burnup code ORIGEN2 and continuous energy Monte Carlo code MVP, by the analysis of post irradiation examinations (PIEs). Some isotopes show differences of calculation results between SWAT and SWAT2. However, generally, the differences are smaller than the error of PIE analysis that was reported in previous SWAT validation activity, and improved results are obtained for several important fission product nuclides. This study also includes comparison between an assembly and a single pin cell geometry models.
Watanabe, Shoichi; Yamane, Yuichi; Miyoshi, Yoshinori
JAERI-Tech 2003-045, 73 Pages, 2003/03
JAERI-Tech-2003-045.pdf:4.96MB
Since exact information is not always acquired in the criticality accident of fuel-solution, parametric survey calculations are required for grasping behaviors of the thermal-hydraulics. On the other hand, the practical methods of the calculation which can reduce the computation time with allowable accuracy will be also required, since the conventional method takes a long calculation time. In order to fulfill the requirement, a three-dimensional nuclear-kinetics analysis code considering thermal-hydraulic based on the multi-region kinetic equations with one-group neutron energy was created by incorporating the thermal-hydraulics analysis code PHOENICS as a subroutine. The computation time of the code was shortened by separating time mesh intervals of the nuclear- and heat-calculations from that of the hydraulics calculation, and by regulating automatically the time mesh intervals in proportion to power change rate. A series of analysis were performed for the natural-cooling characteristic test using TRACY in which the power changed slowly for 5 hours after the transient power resulting from the reactivity insertion of a 0.5 dollar. It was found that the code system was able to calculate within the limit of practical time, and acquired the prospect of reproducing the experimental values considerably for the power and temperature change.
Tobita, Masahiro*; Matsui, Yoshinori
JAERI-Tech 2003-042, 132 Pages, 2003/03
JAERI-Tech-2003-042.pdf:7.19MB
Prediction of irradiation temperature is one of the important issues in the design of the capsule for irradiation test. Many kinds of capsules with complex structure have been designed for recent irradiation requests, and three-dimensional (3D) temperature calculation becomes inevitable for the evaluation of irradiation temperature. For such 3D calculation, however, many works are usually needed for input data preparation, and a lot of time and resources are necessary for parametric studies in the design. To improve such situation, JAERI introduced 3D-FEM (finite element method) code NISA (Numerically Integrated elements for System Analysis) and developed several subprograms, which enabled to support input preparation works in the capsule design. The 3D temperature calculation of the capsule are able to carried out in much easier way by the help of the subprograms, and specific features in the irradiation tests such as non-uniform gamma heating in the capsule, becomes to be considered.
Takahashi, Fumiaki; Yamaguchi, Yasuhiro
Radioisotopes, 52(2), p.94 - 97, 2003/02
Effect of phantom material on backscattered radiations was studied for photon irradiation. Monte Carlo calculations using MCNP-4B code were performed to analyze scattered radiation on the surface of 30x30x15cm3 slab phantoms with different material. Dose on the surface of a human body was also estimated with a modified MIRD-5 type phantom. No significant difference of dose due to scattered radiation was observed between a soft tissue slab and phantom the water-filled slab phantom recommended by the International Organization for Standardization. On the other hand, dose on the surface of the PMMA phantom was found to be larger than doses on the phantom with water or soft tissue. The results also showed that response of dosimeter on the ISO phantom would be near to that on the trunk of a human body.
Takahashi, Fumiaki; Yamaguchi, Yasuhiro; Iwasaki, Midori*; Miyazawa, Chuzo*; Hamada, Tatsuji*; Funabiki, Jun*; Saito, Kimiaki
Radiation Protection Dosimetry, 103(2), p.125 - 130, 2003/01
Times Cited Count：4 Percentile：31.59(Environmental Sciences)Absorbed dose to tooth enamels against external photon exposure was examined by the Electron Spin Resonance (ESR) dosimetry using tooth samples placed in a realistic physical phantom. Dose to teeth region was also measured with thermo-luminescence dosimeters (TLDs). A voxel-type phantom was constructed from CT images of the physical phantom. Monte Carlo calculations with this voxel-type phantom were performed to analyse the results of the experiments. The obtained data in this study were compared to the enamel doses, which were calculated with a modified MIRD-type and already given in a previous paper. The results suggested that the conversion factors from enamel dose to organ doses obtained by the modified MIRD-type phantom are to be applicable for retrospective individual dose assessments by the ESR dosimetry. The analysis, however, indicated that the size and figure of the head can affect the enamel dose for low photon energy region below 100keV.
Yokoyama, Sumi; Noguchi, Hiroshi; Ryufuku, Susumu*; Sasaki, Toshihisa*; Kurosawa, Naohiro*
JAERI-Data/Code 2002-022, 87 Pages, 2002/11
JAERI-Data-Code-2002-022.pdf:4.26MB
Tritium, which is used as a fuel of a D-T burning fusion reactor, is the most important radionuclide for the safety assessment of a nuclear fusion experimental reactor such as ITER. Thus, a computer code, ACUTRI, which calculates the radiological impact of tritium released accidentally to the atmosphere, has been developed, aiming to be of use in a discussion on licensing of a fusion experimental reactor and an environmental safety evaluation method in Japan. ACUTRI calculates an individual tritium dose based on transfer models specific to tritium in the environment. A Gaussian plume model is used for calculating the atmospheric dispersion of tritium gas (HT) and/or tritiated water (HTO). The environmental pathway model in ACUTRI considers the following internal exposures: inhalation from a primary plume (HT and/or HTO) released from the facilities and inhalation from a secondary plume (HTO) reemitted from the ground following deposition of HT and HTO. This report describes an outline of the ACUTRI code, a user guide and the results of test calculation.
Inaba, Yoshitomo; Nishihara, Tetsuo; Inagaki, Yoshiyuki
Proceedings of 14th Hydrogen Energy Conference (WHEC 2002) (CD-ROM), 9 Pages, 2002/06
The Japan Atomic Energy Research Institute (JAERI) has the demonstration test plan to connect a hydrogen production system by steam reforming of methane with the High Temperature engineering Test Reactor (HTTR). One of the most important safety design issues for the HTTR hydrogen production system is to ensure reactor safety against fire and explosion accidents. Therefore, we developed the P2A code system to analyze event sequences and consequences in detail, on assumed fire and explosion accidents in the HTTR hydrogen production system. It is possible that the P2A analyzes the process of leakage, dispersion and combustion including deflagration and detonation of the combustible fluid in the internal and external area of the reactor building. This paper describes the outline of the P2A and the results of preliminary calculations.
measurement and analysisKato, Yuichi*; Okajima, Shigeaki; Sakurai, Takeshi
JAERI-Data/Code 99-006, 71 Pages, 1999/03
JAERI-Data-Code-99-006.pdf:3.14MB
no abstracts in English
T.V.Hung*; Sakamoto, Yukio; Yasuda, Hideshi
JAERI-Research 98-057, 25 Pages, 1998/10
JAERI-Research-98-057.pdf:1.04MB
no abstracts in English
Okumura, Keisuke
JAERI-Data/Code 98-025, 243 Pages, 1998/10
JAERI-Data-Code-98-025.pdf:10.15MB
no abstracts in English
