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Yano, Yasuhide; Uwaba, Tomoyuki; Tanno, Takashi; Yoshitake, Tsunemitsu; Otsuka, Satoshi; Kaito, Takeji
Journal of Nuclear Science and Technology, 9 Pages, 2023/00
Times Cited Count：1 Percentile：68.31(Nuclear Science & Technology)The effects of fast neutron irradiation on tensile properties of modified 316 stainless steel (PNC316) claddings and wrappers for fast reactors were investigated. PNC316 claddings and wrappers were irradiated in the experimental fast reactor Joyo at irradiation temperatures between 400 and 735 
C to fast neutron doses ranging from 21 to 125 dpa. The post-irradiation tensile tests were carried out at room and irradiation temperatures. Elongations of PNC316 measured by the tensile tests were maintained at an engineering level, although the material incurred significant irradiation hardening and softening. The maximum swelling of PNC316 wrappers was about 2.5 vol.% at irradiation temperature between 400 and 500C up to 110 dpa. Japanese 20% cold-worked austenitic steels, PNC316 and 15Cr-20Ni, had sufficient ductility and work-hardenability even after above 10 vol.% swelling, while they had very weak plastic instabilities.
; 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 ...
Brear, D. J.
PNC TN9410 98-005, 53 Pages, 1998/01
When liquid fuel makes contact with steel structure the liquid can freeze as a crust and the structure can melt at the surface. The melting and freezing processes that occur can influence the mode of fuel freezing and hence fuel relocation. Furthermore the temperature gradients established in the fuel and steel phases determine the rate at which heat is transferred from fuel to steel. In this memo the 1-D transient heat conduction equations are applied to the case of initially liquid UO brought into contact with solid steel using up-to-date materials properties. The solutions predict criteria for fuel crust formation and steel melting and provide a simple algorithm to determine the interface temperature when one or both of the materials is undergoing phase change. The predicted steel melting criterion is compared with available experimental results.
Sawada, Shusaku*; *; *
PNC TJ9124 91-002, 331 Pages, 1991/03
"HIBEACON", which is a three dimensional core deformation analysis code for the experimental fast reactor "JOYO", had been modified in order to expand the code's functions for planning a long-term operation and managing the operation of "JOYO", and to analyse the core deformation characteristics speedily and accurately. In this study, "HITETRAS", which is a code for calculating temperatures and fast neutron fluxes on wrapper tubes, has been modified in order for it to correspond with the modification of "HIBEACON". And, the core deformation analysis on "JOYO" MK-III has been performed with those modified codes. Results of this study are as follows: (1) Improvement of core deformation analysis code. (a) Modification of function of "HITETRAS" to output temperatures and neutron fluxes. The method to output wrapper tubes' temperatures and neutron fluxes has been modified corresponding to the modification of "HIBEACON". (b) Addition of ability to alter program size of "HITETRAS". An ability to alter the program size has been added to "HITETRAS" by using INCLUDE statement and PARAMETER statement of FORTRAN language. (c) Modification of "HIBEACON". Following modifications, which are required for analysis on "JOYO" MK-III, have been perfomed. (1) alteration of calculation method for gap clearance between wrapper tubes. (2) addition of function for outputting elements of free bowing deformation (3) addition of function for selecting items to be output on list (2) Core deformation analysis on "JOYO" MK-III. "JOYO" MK-III will have different core characteristics from MK-II because of its higer neutron flux feature, two core regions, wider irradiation space, and so on. So, the core deformation behavior of "JOYO" MK-III has been analysed with the modified codes above-mentioned, and it has been clarified that there is no problem on the core integrity of MK-III from the view point of the core deformation.
