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Akimoto, Hajime; Sugawara, Takanori
JAEA-Data/Code 2016-008, 87 Pages, 2016/09
JAEA-Data-Code-2016-008.pdf:15.62MB
Thermal hydraulic behavior in a lead-bismuth cooled accelerator driven system (ADS) is analyzed under normal operation condition. Input data for the ADS version of J-TRAC code have been constructed to integrate the conceptual design. The core part of the ADS is modeled in detail to evaluate the core radial power profile effect on the core cooling. As the result of the analyses, the followings are found; (1) Both maximum clad temperature and fuel temperature are below the design limits. (2) The radial power profile has little effect on the coolant flow distribution among fuel assemblies. (3) The radial power profile has little effect on the heat transfer coefficients along fuel rods. (4) The thermal hydraulic behaviors along four steam generators are identical. The thermal hydraulic behaviors along two pumps are also identical. A fast running input data is developed by the simplification of the detailed input data based on the findings mentioned above.
Yamamoto, Masanobu
Proceedings of 57th ICFA Advanced Beam Dynamics Workshop on High-Intensity and High-Brightness Hadron Beams (HB 2016) (Internet), p.110 - 114, 2016/08
We have been developing a longitudinal particle tracking code for a high intensity proton synchrotron, especially for the J-PARC Synchrotron. Although some longitudinal particle tracking codes exist, our code can track the particles with a wake voltage and a space charge effect, and also can calculate a beam emittance and a momentum filling factor under a multi-harmonics to evaluate the margin of a rf bucket. Furthermore, we originally have developed the calculation method of a synchronous particle, which realizes the simulation in the case that the revolution frequency of the synchronous particle is not proportional to an acceleration frequency pattern. This is useful to check an adiabaticity. We have achieved 1 MW-eq. beam acceleration at J-PARC RCS by using the code because we can calculate the optimum acceleration conditions for the high intensity beam. We will describe the basic design of the code and the simulation results for the J-PARC RCS and MR.
Akimoto, Hajime
JAEA-Data/Code 2014-031, 75 Pages, 2015/03
JAEA-Data-Code-2014-031.pdf:37.23MB
A thermal-hydraulic analysis code for transmutation system with lead-bismuth cooled accelerator-driven system (ADS) has been developed using the Japanese-version of Transient Reactor Analysis Code (J-TRAC) as the framework to apply the design studies of ADS. To identify the required capabilities of the thermal-hydraulic analysis code for ADS, previous thermal-hydraulic analyses of light water reactors, sodium-cooled fast reactor and ADS have been surveyed. To make up for insufficient capabilities of the J-TRAC code as a thermal-hydraulic analysis code of ADS, physical properties of lead-bismuth eutectic (LBE), argon gas and nitride nuclear fuel were implemented to the J-TRAC code. It was confirmed that the implemented capabilities worked as expected through verification calculations on (1) single-phase LBE flow, (2) heat transfer in a fuel assembly, and (3) heat transfer in a steam generator.
Sono, Hiroki; Ono, Akio*; Kojima, Takuji; Takahashi, Fumiaki; Yamane, Yoshihiro*
Journal of Nuclear Science and Technology, 43(3), p.276 - 284, 2006/03
Times Cited Count：1 Percentile：9.98(Nuclear Science & Technology)For a study on the applicability of a personal dosimetry method to criticality accident dosimetry, an assessment of the human body surface and internal dose estimations was performed by experimental and computational simulations. The experimental simulation was carried out in a criticality accident situation at the TRACY facility. The neutron and 
-ray absorbed doses in muscle tissue were separately estimated by a dosimeter set of an alanine dosimeter and a thermoluminescence dosimeter made of enriched lithium tetra borate with a phantom. The computational simulation was conducted with a Monte Carlo code taking account of dose components of neutrons, prompt -rays and delayed -rays. The computational simulation was ascertained to be valid by comparison between the calculated dose distributions in the phantom and the measured ones. The assessment based on the experimental and computational simulations confirmed that the personal dosimetry using the dosimeter set provided a first estimation of the body surface and internal doses with precision.
Asakura, Toshihide; Hotoku, Shinobu; Ban, Yasutoshi; Matsumura, Masakazu; Morita, Yasuji
Journal of Nuclear and Radiochemical Sciences, 6(3), p.271 - 274, 2005/12
Tc extraction and separation experiments were performed basing on PUREX technique with using spent UO
fuel with burn-up of 44 GWd/t. The experimental results were examined with performing calculations by a simulation code ESSCAR (Extraction System Simulation Code for Advanced Reprocessing). It was demonstrated that Tc can be almost quantitatively extracted from a dissolver solution and that Tc can also be almost quantitatively recovered by scrubbing. Further, it was clearly presented from the calculation results of ESSCAR that the extraction mechanism of Tc is dominated by the synergistic effect of Zr and U.
Asakura, Toshihide; Sato, Makoto; Matsumura, Masakazu; Morita, Yasuji
JAERI-Conf 2005-007, p.345 - 347, 2005/08
This paper reviews the succeeding development and utilization of Extraction System Simulation Code for Advanced Reprocessing (ESSCAR). From the viewpoint of development, more tests with spent fuel and calculations should be performed with better understanding of the physico-chemical phenomena in a separation process. From the viewpoint of process safety research on fuel cycle facilities, it is important to know the process behavior of a key substance; being highly reactive but existing only trace amount.
-ray dose components in criticality accident situationsSono, Hiroki; Yanagisawa, Hiroshi*; Ono, Akio*; Kojima, Takuji; Soramasu, Noboru*
Journal of Nuclear Science and Technology, 42(8), p.678 - 687, 2005/08
Times Cited Count：4 Percentile：30.51(Nuclear Science & Technology)Component analysis of -ray doses in criticality accident situations is indispensable for further understanding on emission behavior of -rays and accurate evaluation of external exposure to human bodies. Such dose components were evaluated, categorizing -rays into four components: prompt, delayed, pseudo components in the period of criticality, and a residual component in the period after the termination of criticality. This evaluation was performed by the combination of dosimetry experiments at the TRACY facility using a thermoluminescent dosimeter (TLD) made of lithium tetra borate and computational analyses using a Monte Carlo code. The evaluation confirmed that the dose proportions of the above components varied with the distance from the TRACY core tank. This variation was due to the difference in attenuation of the individual components with the distance from the core tank. The evaluated dose proportions quantitatively clarified the contribution of the pseudo and the residual components to be excluded for accurate evaluation of -ray exposure.
Takahashi, Fumiaki; Endo, Akira; Yamaguchi, Yasuhiro
Journal of Nuclear Science and Technology, 42(4), p.378 - 383, 2005/04
Times Cited Count：3 Percentile：24.22(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.
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.
Yoshizawa, Michio; Endo, Akira
JAERI-Conf 2003-002, 166 Pages, 2003/03
JAERI-Conf-2003-002.pdf:9.79MB
The present report is Proceedings of the Third Workshop on Dosimetry for External Radiations, held at the Tokai Research Establishment, Japan Atomic Energy Research Institute (JAERI), in November 28-29, 2002. The proceedings comprises 16 papers and a summary of general discussion. The Third Workshop, subtitled "On an opportunity of the completion of the facility of calibration standards for neutron at JAERI", focused on neutron dosimetry and included presentations on the status of international neutron standards, the development of calibration techniques of neutron dosimeters using accelerator neutron sources, and dosimetry for high-energy neutrons. The workshop identified the directions for the future research and development in this field.
Hotoku, Shinobu; Asakura, Toshihide; Mineo, Hideaki; Uchiyama, Gunzo
Journal of Nuclear Science and Technology, 39(Suppl.3), p.313 - 316, 2002/11
no abstracts in English
Takase, Kazuyuki; Akimoto, Hajime
Applied Electromagnetics in Materials, p.177 - 178, 2001/00
no abstracts in English
*; *; *; *; *; Adachi, Masaaki*; Ogasawara, Shinobu*; *; Kume, Etsuo
JAERI-Data/Code 99-027, 39 Pages, 1999/05
JAERI-Data-Code-99-027.pdf:1.21MB
no abstracts in English
; Murao, Yoshio
Journal of Nuclear Science and Technology, 33(4), p.290 - 297, 1996/04
Times Cited Count：3 Percentile：32.72(Nuclear Science & Technology)no abstracts in English
Onuki, Akira; ; Murao, Yoshio
Journal of Nuclear Science and Technology, 32(3), p.245 - 256, 1995/03
Times Cited Count：1 Percentile：17.53(Nuclear Science & Technology)no abstracts in English
; Onuki, Akira; Murao, Yoshio
Proc. of the 2nd Int. Conf. on Multiphase Flow 95-Kyoto, 0, p.P2_37 - P2_44, 1995/00
no abstracts in English
; Onuki, Akira; Murao, Yoshio
Validation of Systems Transients Analysis Codes (FED-Vol. 223), 0, 8 Pages, 1995/00
no abstracts in English
; Abe, Yutaka*; Onuki, Akira; Murao, Yoshio
JAERI-Data/Code 94-006, 40 Pages, 1994/07
JAERI-Data-Code-94-006.pdf:1.26MB
no abstracts in English
Onuki, Akira; ; Murao, Yoshio
JAERI-M 94-026, 60 Pages, 1994/03
no abstracts in English
; Onuki, Akira; Murao, Yoshio
JAERI-M 93-240, 83 Pages, 1993/12
no abstracts in English
