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Kobe, Mitsuru*; Tsunoda, Hirokazu*; Mishima, Kaichiro*; Kawasaki, Akira*; Iwamura, Takamichi
JAERI-Tech 2003-016, 68 Pages, 2003/03
JAERI-Tech-2003-016.pdf:4.37MB
The 200 kWe uranium nitride fueled lithium cooled fast reactor "RAPID-L" combined with thermoelectric power conversion system that can be operated unmanned without refueling for up to ten years has been demonstrated. The RAPID refueling concept enables quick and simplified refueling, and achieves plant design lifetime over 20 years. A significant advantage of the RAPID-L design, which does not require the use of control rods - is the introduction of the innovative reactivity control systems: lithium expansion module (LEM), lithium injection module (LIM) and lithium release module (LRM). LEM is the most promisiong candidate for improving inherent reactivity feedback. LEMs could realize burnup compensation. LIMs assure sufficient negative reactivity feedback in unprotected transients. LRMs enable an automated reactor startup by detecting the hot standby temperature of the primary coolant. All these systems use 
Li as liquid poison and are actuated by highly reliable physical properties (volume expansion of Li for LEM, and freeze seal melting for LIM and LRM).
*; *
JNC TJ9410 2002-001, 96 Pages, 2002/03
JNC-TJ9410-2002-001.pdf:2.94MB
The following codes were developed to enhance the applicability of the three-dimensional Hex-Z geometry codes that can model the BN and BFS reactor core accurately. The neutron flux reconstruction method in the three-dimensional transport-based perturbation code SNPERT-HEXZ was modified to improve the calculation accuracy. New functions to calculate design parameters, such as peaking coefficients, maximum linear power etc., based on the reconstructed neutron flux were added to the three dimensional Hex-Z geometry transport burn-up code NSHEX-BURN. The three-dimensional Hex-Z finite difference transport code MINIHEX was modified to improve the computational performance. First, cash-tuning and a flux extrapolation method were applied to MINIHEX. This modification reduced the computational time by a factor of 3.8 at most. Second, the modified MINIHEX was parallelized with MPI by an angular space decomposition method. A series of performance check calculations was performed on GP7000 and DEC/alpha EWS cluster. Speed-up obtained on GP7000 was at most 5.3 for 9 PE and that on DEC/alpha cluster was at most 1.5 for 4 PE. The parallelization of MINIHEX also enhanced the applicability of the code to larger scale computational problems and more detailed reactor core modeling. An interface to MINIHEX was added to the three-dimensional Hex-Z geometry reaction rate calculation code LAGOON-HEXZ. The added interface enables the code to calculate reaction rates using neutron flux obtained by MINIHEX. A three-dimensional transport-based perturbation code for MINIHEX was developed based on the three-dimensional tansport-based perturbation code NSHEX-HEXZ. This code makes it possible to evaluate reactivity cbange from the neutron flux and eigenvalue obtained by MINIHEX.
Kobe, Mitsuru*; Tsunoda, Hirokazu*; Mishima, Kaichiro*; Kawasaki, Akira*; Iwamura, Takamichi
Genshiryoku eye, 48(1), p.23 - 28, 2002/01
no abstracts in English
Tsunoda, Hirokazu*; Sato, Osamu*; Okajima, Shigeaki; Yamane, Tsuyoshi; Iijima, Susumu; Kobe, Mitsuru*
Proceedings of International Congress on Advanced Nuclear Power Plants (ICAPP) (CD-ROM), 6 Pages, 2002/00
no abstracts in English
*; *
JNC TJ9400 2001-008, 256 Pages, 2001/03
JNC-TJ9400-2001-008.pdf:6.54MB
The following codes were developed to enhance the applicability of the three-dimensional Hex-Z geometry codes that can model the BN and BFS reactor core accurately. A burn-up calculation code named NSHEX-BURN was developed. This code utilizes the three-dimensional transport code NSHEX as a neutron flux solver. Although the geometry of this code is restricted to the whole core configuration, the code is equipped with the function of fuel management and burn-up analysis as the diffusion-based code MOSES. Test calculations showed that the spatial distribution of physical quantities, such as power density, is slightly different from the results that calculated by the diffusion-based burn-up code due to the transport effect. 0n the other hand, the changes of the effective multiplication factor caused by burn-up were well agreed with both codes. A three-dimensional, transport-based perturbation code was developed. The code uses forward and adjoint angular fluxes obtained by the NSHEX code. The results from test calculation indicate that the accuracy of the reconstruction method of neutron angular flux in calculation node plays an essential role in attaining the accurate evaluation of reactivity change. Various acceleration methods were applied to the NSHEX code to improve the computational performance. Performance check calculations showed that the extrapolation of neutron flux moments in outer iterations is most effective and that the extrapolation of higher order flux moment is very important. This method reduced the computational time by a factor of three at most. The reaction rate analysis code LAGOON was modified to determine flux at any point of hexagonal assembly. This modification enhanced the applicability of LAGOON code to Hex-Z and Tri-Z geometry, in addition to currently available XYZ, XY, and RZ geometry. The two-dimensional, transport-based perturbation code SNPERT was extended to use microscopic cross section libraries. New functions to calculate ...
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PNC TJ1222 96-003, 82 Pages, 1996/03
no abstracts in English
; *; *; *
Proceedings of 1996 Radiation Protection and Shielding Division Topical Meeting, 0 Pages, 1996/00
None
Shono, Akira; Tsunoda, Hirokazu; Takemura, Morio; Handa, Hiroyuki
PNC TN9410 95-171, 280 Pages, 1995/06
All experiments planned in the JASPER (Japanese-American Shielding Program for Experimental Research) project have been conducted from '85 to '92. Results obtained from the post-experimental analyses are described in the annual reports ('86
'94). This report is intended to review and summarize the enormous information of the annual reports. For the evaluation, the following topics were chosen. (1)Bulk Shielding attenuation characteristics and analysis accuracy (2)Design-dependent shielding characteristics and analysis accuracy (3)Evaluation of the shielding cross section libraries (4)Improvements in shielding analysis methods. Major conclusions are briefly as follows. (a)Both bulk shielding attenuation characteristics and streaming characteristics of configurations consisting of several materials including boron-carbide(B
C), graphite, stainless steel, sodium and etc. were clarified and the analysis accuracy confirmed. (b)JSDJ2 (based on the JENDL-2) was demonstrated to be a better cross section library for shielding analysis by reviewing experimental analyses results. (c)The shielding analysis system for fast reactors using a 2-dimensional discrete ordinates code as the standard has been improved and verified. (d)Useful experiences have been gained in verifying both a Monte Carlo code and a 3-dimensional discrete ordinates code, as well as in optimizing various parameters including those for mesh spacing techniques. Another objective of this report is to specify key information which will be useful to review JASPER. For this purpose, information on experiment and analysis is presented in a table format by each experimental item. All technical reports related to JASPER are listed.
Otani, Nobuo*; *; *; Haga, Kazuo*
PNC TN9410 89-145, 98 Pages, 1989/10
Core physics of small reactor was examined as a part of conceptual design study of space reactor which is an application of transportable reactor. The design requirements were a fast spectrum reactor using nitride fuel and lithium coolant. Firstly, characteristics of typical uranium core and plutonium core was compared by means of one-dimensional calculation using simple sphere model. Followings were revealed from the comparison of both calculation results. (1)Reactivity loss of uranium core in ten years is smaller than that of plutonium core. (2)Shorter lifetime of plutonium core is due to 
decay of Pu-241. Hence, plutonium core must be designed to compensate higher fuel degradation. (3)Installation of thermal neutron absorber between core and reflector region is effective to extend lifetime of plutonium core. Secondly, study to optimize the design of enriched uranium core was performed as a parameter of core configuration, fuel composition and core size from the point of reactivity adjustment. Attention was focused on criticality of fresh core, reactivity degradation, sub-criticality and reactor shut down margin. The results showed that some safety margin can be obtained although it was less than that objected. A 3-dimensional Monte Carlo code was partly used in the analysis. It was revealed that a more superior code in simulating core configuration is necessary.
Komuro, Yuichi; *; *
JAERI-M 87-092, 115 Pages, 1987/07
no abstracts in English
