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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.
Yamaguchi, Isoo*; Morita, Yasuji; Fujiwara, Takeshi; Yamagishi, Isao
JAERI-Tech 2005-054, 61 Pages, 2005/09
JAERI-Tech-2005-054.pdf:12.38MB
The HLW-79Y-4T type transportation cask for liquid radioactive fuel material (commonly called "Cendrillon") was imported from France and modified for Japanese regulation in order to obtain high-level radioactive liquid waste (HLW) for partitioning tests in JAERI by transportation from Tokai Establishment of Japan Nuclear Fuel Cycle Development Institute. The cask was used for the HLW transportation five times from 1982 to 1990. After that, it was kept and maintained for next transportation of HLW from facilities outside JAERI. Finally, we decided to decompose the cask because HLW can be obtained in JAERI Tokai. For the decomposition, radiation dose and contamination by radioactivity was first measured and then the methods to reduce those levels were determined. The cask was decomposed after the decontamination to separate the part that has high radiation level. The separated part was put in a vessel specially prepared. The present report describes those procedures for the decomposition of the transportation cask.
Review Group on the Structure of the Spent Nuclear Fuel Transportation Casks for
JAERI-Review 2005-023, 133 Pages, 2005/07
JAERI-Review-2005-023.pdf:18.88MB
The Japan Atomic Energy Research Institute (JAERI) constructed two stainless steel transportation casks, JRC-80Y-20T, for spent nuclear fuels of research reactors and had utilized them for transportation since 1981. A modification of the design was applied to the USA for transportation of silicide fuels. Additional analyses employing the impact analysis code LS-DYNA that was often used for safety analysis were submitted by the JAERI to the USA to show integrity of the packages; the casks were still not approved, because inelastic deformation was occurred on the surface of the lid touching to the body. To resolve this problem on design approval of transportation casks, a review group was formed at the end of this June. The group examined the impact analyses by reviewing the input data and performing the sensitivity analyses. As the drop impact analyses were found to be practically reasonable, it was concluded that the approval of the USA for the transportation casks could not be obtained just by revising the analyses; therefore, remodelling the casks is required.
Kuroishi, Takeshi; Nomura, Yasushi
Journal of Nuclear Science and Technology, 40(6), p.433 - 440, 2003/06
Times Cited Count：2 Percentile：19(Nuclear Science & Technology)Effective source acceleration method is studied in criticality safety analysis for realistic spent fuel transport cask. Various axial burnup profiles based on in-core flux measurements are proposed in the OECD/NEA/BUC benchmark Phase II-C. In some cases, calculations by ordinary Monte Carlo method show very slow convergence of fission source distribution, and unacceptably large skipped cycles are needed. The matrix eigenvector calculation that has been developed and incorporated in the ordinary Monte Carlo calculation to improve the slow convergence is applied to the benchmark. The efficiency of this method depends on the precision of matrix elements. In a certain stage of insufficient convergence of fission source distribution, especially for this benchmark of very slow convergence, more acceleration procedure causes anomalous results because of large statistical fluctuations of matrix elements corresponding to low source levels. Therefore, we propose effective source acceleration method with less calculation time than increasing histories for the estimation of matrix elements.
Kuroishi, Takeshi; Hoang, A.; Nomura, Yasushi; Okuno, Hiroshi
JAERI-Tech 2003-021, 60 Pages, 2003/03
JAERI-Tech-2003-021.pdf:4.56MB
The reactivity effect of the asymmetry of axial burnup profile is studied for PWR spent fuel transport cask proposed in OECD/NEA Phase II-C benchmark. The axial burnup profiles are based on in-core flux measurements. Criticality calculations are performed with the continuous energy Monte Carlo code MCNP-4B2 and the nuclear data library JENDL-3.2. Calculations are carried out not only for cases in the benchmark but also for symmetric burnup cases. Both actinide-only approach and actinide plus fission product approach is considered. The end effect is more sensitive to higher burnup asymmetry. The axial fission distribution becomes strongly asymmetric as its peak shifts toward the fuel top end. The peak of fission distribution gets higher with the increase of either the burnup asymmetry or the assembly-averaged burnup. The conservatism of uniform axial burnup assumption for the actinide-only approach is estimated quantitatively in comparison with the keff result calculated with experiment-based strongest asymmetric axial burnup profile for the actinide plus fission product approach.
JAERI-Data/Code 98-018, 109 Pages, 1998/05
JAERI-Data-Code-98-018.pdf:2.65MB
no abstracts in English
JAERI-Data/Code 98-009, 136 Pages, 1998/03
JAERI-Data-Code-98-009.pdf:5.04MB
no abstracts in English
JAERI-Data/Code 97-049, 101 Pages, 1997/12
JAERI-Data-Code-97-049.pdf:2.36MB
no abstracts in English
JAERI-Data/Code 97-018, 61 Pages, 1997/05
JAERI-Data-Code-97-018.pdf:1.31MB
no abstracts in English
JAERI-Data/Code 97-001, 85 Pages, 1997/02
JAERI-Data-Code-97-001.pdf:1.76MB
no abstracts in English
JAERI-Data/Code 96-025, 71 Pages, 1996/07
JAERI-Data-Code-96-025.pdf:1.56MB
no abstracts in English
; *; *
PATRAM 95: 11th Int. Conf. on the Packaging and Transportation of Radioactive Materials, 3, p.1174 - 1181, 1996/00
no abstracts in English
JAERI-Data/Code 95-017, 64 Pages, 1995/11
JAERI-Data-Code-95-017.pdf:1.24MB
no abstracts in English
*; Naito, Yoshitaka; Okuno, Hiroshi
Proc. of the 10th Int. Symp. on the Packaging and Transportation of Radioactive Materials,Vol. 1, p.177 - 184, 1993/00
no abstracts in English
JAERI-M 91-061, 119 Pages, 1991/04
no abstracts in English
JAERI-M 91-060, 164 Pages, 1991/04
no abstracts in English
; *
Nihon Genshiryoku Gakkai-Shi, 33(4), p.381 - 390, 1991/04
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)no abstracts in English
; ; ;
JAERI-M 9957, 94 Pages, 1982/03
no abstracts in English
; ; Shimooke, Takanori
JAERI-M 9562, 44 Pages, 1981/07
no abstracts in English
