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Department of Hot Laboratories
JAERI-Review 2005-047, 95 Pages, 2005/09
This is an annual report in 2004 fiscal year that describes activities of the Reactor Fuel Examination Facility (RFEF), the Waste Safety Testing Facility (WASTEF), and the Research Hot Laboratory (RHL) in the Department of Hot laboratories. In RFEF, BWR fuel rods were withdrawn from a fuel assembly irradiated for 5 cycles in the Fukushima-2 Nuclear Power Station Unit-1 and PIEs including nondestructive examination of those rods were carried out. In WASTEF, Slow Strain Rate Tests for detecting the susceptibility to IASCC, the corrosion test of reprocessing plant materials, tests for evaluating barrier performance in terms of waste disposal were performed. A secondary system pipe from the Mihama Nuclear Power Station Unit-3 was accepted to inspect the ageing fracture of it. In RHL, 15 lead cells are dismantled under the decommissioning plan at JAERI Tokai. And an arrangement of the RHL facility was started to use the storage of unirradiated nuclear materials.
Sugiyama, Tomoyuki; Fuketa, Toyoshi
Journal of Nuclear Science and Technology, 41(11), p.1083 - 1090, 2004/11
Times Cited Count:10 Percentile:55.55(Nuclear Science & Technology)The effect of cladding surface pre-oxidation on the rod coolability under reactivity initiated accidents was investigated. NSRR tests on irradiated fuel rods have shown higher rod coolability than that of fresh rods, which arose from suppressed DNB and early quench at the surface. To identify the dominant factor, possible factors such as pellet cracking and so on, were assessed. The most probable factor, the cladding pre-oxidation, was examined by pulse irradiation tests on fresh rods with three cladding surface conditions, no oxide layer, 1m and 10m-thick oxide layers. Temperature measurements showed increased thresholds for DNB and quench at the pre-oxidized surface, leading to a reduced film boiling duration. The shifts of the critical and minimum heat flux points could be caused by the surface wettability increase. In the present tests, the wettability change was probably dominated by the chemical potential change at the surface due to pre-oxidation. The test results indicate the effects do not depend on the oxide layer thickness, but on the presence of the oxide layer.
Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Kureta, Masatoshi; Tamai, Hidesada; Akimoto, Hajime
Proceedings of 5th International Conference on Multiphase Flow (ICMF 2004) (CD-ROM), 14 Pages, 2004/06
no abstracts in English
Katanishi, Shoji; Takei, Masanobu; Nakata, Tetsuo*; Kunitomi, Kazuhiko
Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(1), p.67 - 75, 2004/03
no abstracts in English
*
JNC TN9440 2000-008, 79 Pages, 2000/08
This report summarizes the operating and irradiatlon data of the experimental reactor "JOYO" 35th cycle. Irradiation tests in the 35th cycle are as follows: (1)C-type irradiation rig (C4F) (a)High burnup performance test of advanced austenitic stainless steel cladding fuel pins (in collaboration with France) (2)C-type irradiation rig (C6D) (a)Large diameter fuel pins irradiation tests (3)Core Materials Irradiation Rig (CMIR-5) (a)Cladding tube materials irradiation tests for "MONJU" (4)Structure Materials Irradiation Rigs (SMIR) (a)Decision of material design base standard of structure materials for prototype reactor and large scale reactor (5)Upper core structure irradiation Plug Rig (UPR-1-5) (a)Upper core neutron spectrum effect and accelerated irradiation effect (6)SurVeillance un-instrument Irradiation Rig (SVIR) (a)Confimation of surveillance irradiation condition for "JOYO" (b)Material irradiation tests (based on a contract with universities) The maximum burnup driver assembly "PFD253" reached 67,600 MWd/t (pin average).
; Sato, Wakaei*; Iwai, Takehiko*
JNC TN9400 2000-096, 113 Pages, 2000/06
This report describes the updated analyses results on the BFS-58-1-I1 core. The experiment was conducted at BFS-2 of Russian Institute of Physics & Power Engineering (IPPE). The central region is "non-Uranium fuel zone", where only Pu can induce fission reaction. The non-U zone is surrounded by MOx fuel zone, which is surrounded by U0 fuel zone. Sodium is used for simulating the coolant material. As it was found that the lattice pitch had been incorrectly understood in the past analyses, all items have been re-calculated using the corrected number densities. Furthermore, significantly softened neutron spectrum in the central region caused problems in applying the plate-stretch model that has been established for fast reactor cores through JUPITER experimental analyses. Both keeping the pellet density and using SRAC library for the elastic cross section for lighter nuclides allow us to obtain reasonable analysis accuracy on the spectral indices that were measured at the center of the core. Application of such a cell model was justified through comparison among various cell models using continuous energy Monte-Carlo code MVP. It is confirmed that both the MOX zone and the U0 zone can be correctly evaluated by the plate-stretch model. Based on the updated cell calculation, both the effective multiplication factor (k-eff)and the spectral indexes agree well with the measured values. The transport and mesh-size correction is made for the k-eff evaluation. Those results also agree well within reasonable difference between those obtained by IPPE and CEA, which were obtained by using sub-group method or continuous-energy Monte Carlo code. Evaluation by the nuclear data library adjustment confirmed that the analyses results of the BFS-58-1-I1 core have no significant inconsistency with JUPITER experimental analyses results. Those results are quite important for starting BFS-62 cores, which will be analyzed in the framework of supporting program for Russian ...
Sasajima, Hideo; Fuketa, Toyoshi; Nakamura, Takehiko; Nakamura, Jinichi; Kikuchi, Keiichi*
Journal of Nuclear Science and Technology, 37(5), p.455 - 464, 2000/05
no abstracts in English
; ;
JNC TN9400 2000-041, 29 Pages, 2000/03
Irradiation behavior and performance models were investigated in order to apply for nitride fuel options in feasibility study on fast breeder reactor and related recycle systems. (1)MechanicaI design of nitride fuel pin: The behaviors of fission gas release (increase of internal Pressure) and fuel-to-cladding chemical interaction (decrease of cladding thickness) are needed to evaluate cumulative damage fraction in case of fuel pin mechanical design. The behaviors of fission gas release and fuel-to-cladding chemical interaction were investigated from the past studies up to high burnuP, since the lower fission gas release in nitride fuel than in oxide fuel could contribute to reduce the plenum volume and result in the shortening of fuel Pin length. (2)Fuel pin smear density: The higher fuel smear density is preferred for the higher fissile density to improve the core characteristic. The behaviors of fuel pellet swelling were investigated from the past studies up to higher burnup, since the larger fuel pellet swelling in nitride fuel than in oxide fuel would restrict high burunp capability due to fuel-cladding mechanical interaction. (3)Compatibility of nitride fuel with high Temperature water: Compatibility of nitride fuel with high temperature water were investigated from the past studies to contribute water cooled fast breeder reactor options.
Sasajima, Hideo; Fuketa, Toyoshi; Nakamura, Takehiko; Nakamura, Jinichi; Uetsuka, Hiroshi; Kikuchi, Keiichi*; Abe, Tomoyuki*
JAERI-Research 99-060, p.62 - 0, 2000/03
no abstracts in English
Saito, Hioraki*; Iriya, Yoshikazu*
JNC TJ8440 99-003, 156 Pages, 1999/03
no abstracts in English
;
PNC TN9410 98-058, 12 Pages, 1998/06
Based on the RB1 test result in the CABRI-RAFT Program, it was agreed between the partners to perform the RB2 test which aims at observation of molten fuel ejection into the coolant channel at further fuel melting and at confirmation of coolability of ejected fuel. In this study, a preliminary post-test calculation for the RB1 test is performed first to reflect the fuel thermal condition expected for the pins with the special artificial defect preparation. Pre-test calculations for the RB2 test are then performed based on the results of this RB1 calculation. Power and coolant flow histories as well as the axial location of defect were selected as parameters in this study and a set of test condition is proposed which is believed to be most suitable to fulfill the test objectives.
; Ohshima, Hiroyuki
PNC TN9410 97-104, 69 Pages, 1997/12
As the part of the improvement of single-phase subchannel analysis code ASFRE-III, verification study about fuel-pin heat transfer model and flow resistance model of the code was carried out. Temperature distributions in a fuel pin predicted by the fuel-pin heat transfer model of ASFRE-III were compared with those calculated by the structure analysis code FINAS, which has been well validated and applied to various structure analyses, using the same boundary conditions. The comparison showed that the results by these two codes agreed with maximum difference of 1 %. and therefore the validity of the model was confirmed. With respect to the flow resistance model, distributed resistance model (DRM), which can enhance the consistent description of the fluid flow and wire-spacer interaction, was examined through analyses of two hydraulic simulation tests using the fifth mock-up fuel subassembly for the prototype LMFBR and the second mock-up fuel subassembly for the experimental rector. The calculated pressure difference between pressure measurement taps whose positions were near the top and the bottom of the fuel-pin bundle agreed with the measured data of both tests. The predicted pressure distribution in a horizontal cross section was also compared with the calculational result by the finite element analysis code SPIRAL and agreement was good.
Kajitani, Yukio; ; Abe, Kazuyuki; Osaka, Masahiko; ; Hirosawa, Takashi; Koyama, Shinichi
PNC TN9440 97-004, 186 Pages, 1997/02
The 25th anniversary meeting of the Alpha-Gamma Facility (AGF) at O-arai Engineering Center of PNC was held on February 7. The AGF started to examine irradiated materials on october 1 and fuel pins irradiated in the Dounreay Fast Reactor, DFR332/2 on December 1, 1971. The contents in this paper of the anniversary meeting are as follows. (1)25 years history and challenging plan for 2000 year. (2)Maintenance logbook of the facility, apparatus and manipulators for 25 years. (3)Recent results of melting temperature, thermal conductivity and lattice constants in irradiated MOX fuels. (4)Development on fission products release measuring apparatus and results of cold run tests. (5)Post irradiated examination results operated at the metallography cell in the Fuels Monitoring Facility (FMF). (6)Development on chemical analysis method for minor actinides (MA) in irradiated MOX fuels. (7)Refurbishment for MA containing MOX fuels, status and specifications for the fabrication and quality control apparatus.
Takase, Kazuyuki
Nucl. Eng. Des., 165, p.225 - 237, 1996/00
Times Cited Count:9 Percentile:61.99(Nuclear Science & Technology)no abstracts in English
Yoshida, Mamoru; ; Shobu, Nobuhiro; Aihara, Nagafumi;
PNC TN9700 95-001, 804 Pages, 1995/08
no abstracts in English
; Saito, Junichi; ; Endo, Yasuichi; ; Nakagawa, Tetsuya; ; Kawamata, Kazuo; ; Kawamura, Hiroshi; et al.
JAERI-Tech 95-037, 87 Pages, 1995/07
no abstracts in English
Takase, Kazuyuki; Akino, Norio
The 3rd JSME/ASME Joint Int. Conf. on Nuclear Engineering (ICONE),Vol. 1, 0, p.535 - 540, 1995/00
no abstracts in English
JAERI-Tech 94-021, 79 Pages, 1994/09
no abstracts in English
Nakajima, Ken; ; Suzaki, Takenori
Nuclear Science and Engineering, 116, p.138 - 146, 1994/00
Times Cited Count:16 Percentile:78.44(Nuclear Science & Technology)no abstracts in English