Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Miyazaki Noriyuki*; Hagiwara Seiya*; Chino, Eiichi*; Maruyama, Yu*; Hashimoto, Kazuichiro*; Maeda, Akio*
JAERI-Research 2001-047, 35 Pages, 2001/10
JAERI-Research-2001-047.pdf:1.63MB
no abstracts in English
; Ohno, Shuji;
JNC TN2400 2000-006, 56 Pages, 2000/12
JNC-TN2400-2000-006.pdf:1.22MB
Sodium combustion analyses were performed using ASSCOPS version 2.1 in order to obtain background data for evaluating the validity of the mitigation system against secondary sodium leak of MONJU. The calculated results are summarized as follows. (1)Peak atmospheric pressure 
4.3 kPa[gage] (2)Peak floor liner temperature 870
C, Maximum thinning of liner 2.6mm (3)Peak hydrogen concentration <2% (4)Peak floor liner temperature in the spilt sodium storage eell 400C , Peak floor concrete temperature in the spilt sodium storage cell 140C.
; ; Ueno, Fumiyoshi; ; ; ;
JNC TN2400 2000-005, 103 Pages, 2000/12
JNC-TN2400-2000-005.pdf:3.98MB
Inelastic analyses of the floor liner subjected to thermal loading due to sodium leakage and combustion were carried out, considering thinning of the liner plate due to molten salt type corrosion. Because the inelastic strain obtained by the analyses stayed below the ductility limit of the material, mechanical integrity, i.e., there exist no through-wall crack on the floor liner, was confirmed. Partial structural model tests were conducted, with a band of local thinning of the liner plate. Displacements were controlled to give specimens much larger strains than those obtained by the inelastic analyses above. No through-wall crack was observed by these tests. Mechanical integrity of the floor liner was confirmed by these results of the inelastic analyses and the partial structural model tests.
; ; *; Yamaguchi, Akira
JNC TN9400 2000-109, 96 Pages, 2000/11
JNC-TN9400-2000-109.pdf:9.56MB
Numerical calculations were carried out for a free surface sloshing, a thermal stratification, a thermal striping, and a natural convection as key phenomena of in-vessel thermohydraulics in future fast reactor systems with various fluids as coolants. This numerical work was initiaied based on a recognition that the fundamental characteristics of the phenomena have been unsolved quantitatively in the use of various coolants. From the analysis for the phenomena, the following results were obtained. [Free Surface Sloshing phenomena] (1)Ther is no remarkable difference betweeen liquid sodium and luquid Pb-Bi in characteristics of internal flows and free surface charatristics based on Fr number. (2)the AQUA-VOF code has a potentiall enough to evaluate gas entrainment behavior from the free surface including the internal flow characteristics. [thermal Stratification Phenomena] (1)On-set position of thermal entainment process due to dynamic vortex flows was moved to downstream direction with decreasing of Ri number. 0n the other hand, the position in the case of C0
gas was shifted to upstream side with decreasing of Ri number. (2)Destruction speed of the thermal stratyification interface was dependent on thermal diffusivity as fluid properties. therefor it was concluded that an elimination method is necessary for the interface generated in C0 gas. [thermal Striping Phenomena] (1)Large amplitudes of fluid temperature fluctuations was reached to down stream area in the use of CO gas, due to larger fluid viscosity and smaller thermal diffusivity, compared with liquid sodium and liquid Pb-Bi cases. (2)To simulate thermal striping conditions such as amplitude and frequency of the fluid temperature fluctuations, it isnecessary for coincidences of Re number for the amplitude and of velocity value for the frequency, in various coolants. [Natural Convection Phenomena] (1)Fundamental behavior of the natural convection in various coolant follows buoyant jet ....
Harada, Yuhei*; Maruyama, Yu; Maeda, Akio*; Chino, Eiichi; Shibazaki, Hiroaki*; Kudo, Tamotsu; Hidaka, Akihide; Hashimoto, Kazuichiro; Sugimoto, Jun
JAERI-Conf 2000-015, p.309 - 314, 2000/11
no abstracts in English
Shibazaki, Hiroaki*; Maruyama, Yu; Kudo, Tamotsu; Yuchi, Yoko; Chino, Eiichi; Nakamura, Hideo; Hidaka, Akihide; Hashimoto, Kazuichiro
JAERI-Conf 2000-015, p.225 - 230, 2000/11
no abstracts in English
Okamoto, Koji*; *
JNC TY9400 2000-016, 90 Pages, 2000/06
JNC-TY9400-2000-016.pdf:2.53MB
no abstracts in English
; 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 ...
JNC TN9410 2000-010, 72 Pages, 2000/03
JNC-TN9410-2000-010.pdf:2.14MB
The experimental fast reactor JOYO served as the MK-II irradiation bed core for testing fuel and material for FBR development for 16 years from 1982 to 1997. During the MK-II core operation, extensive data were accumulated from the plant characteristic tests. Tests conducted at JOYO included operating characteristic tests for confirming operational safety, performance tests for confirming design performance of the MK-II core, and special tests for research and development ofthe plant. In this report, the outline and the results of each test item are shown. These test data can be provided by the magnet-optical disk.
; ; Saikawa, Takuya*; Sukegawa, Kazuya*
JNC TN9410 2000-008, 66 Pages, 2000/03
JNC-TN9410-2000-008.pdf:1.39MB
The experimental fast reactor "JOYO" served as the MK-II irradiation bed core for testing fuel and material for FBR development for 15 years from 1982 to 1997. During the MK-II operation, impurities concentrations in the sodium and the argon gas were determined by 67 samples of primary sodium, 81 samples of secondary sodium, 75 samples of primary argon gas, 89 samples of secondary argon gas (the overflow tank) and 89 samples of secondary argon gas (the dump tank). The sodium and the argon gas purity control data were accumulated from in thirty-one duty operations, thirteen special test operations and eight annual inspections. These purity control results and related plant data were compiled into database, which were recorded on CD-ROM for user convenience. Purity control data include concentration of oxygen, carbon, hydrogen, nitrogen, chlorine, iron, nickel and chromium in sodium, concentration of oxygen, hydrogen, nitrogen, carbon monoxide, carbon dioxide, methane and helium in argon gas with the reactor condition.
Tsukimori, Kazuyuki; Furuhashi, Ichiro*
JNC TN9400 2000-049, 93 Pages, 2000/03
JNC-TN9400-2000-049.pdf:2.82MB
lt is one of the important key points to reduce thermal stress of the primary piping system in the design of sodium coolant loop-type FBR plants. The objectives of this study are to understand the characteristics of the thermal stresses in the simple S-shaped hot leg piping systems which run from the outlet nozzle of the reactor vessel (R/V) to the inlet nozzle of the intermediate heat exchanger (IHX), and to propose some recommendable routings of piping systems. Results are summarized as follows. (1)Generally, the thermal stresses in elbows are severer than those at nozzles. The tendency was observed that the stress in elbow decreases with the increase of the distance between the outlet nozzle of R/V and the inlet nozzle of IHX and also the distance between the outlet nozzle of R/V and the liquid surface level. (2)lt is expected to reduce thermal stresses in elbow to big extent by adopting super 90 degree elbows. Therefore, in these cases the dimension region which satisfies the allowable stress is broad compared with that in the case of the conventional 90 degree elbow. (3)The stress estimations in elbow based on 'MITl notice No.501' become excessively large compared with the results by FEA using shell elements, when the maximum stress occurs at the end of elbow. ln these cases, the estimation can be rationalized by replacing the maximum stress by the mean of stresses at the end and at the middle of the elbow. (4)Two routings with 105 degree elbows are recommended. 0ne has the advantage from the view point of reduction of length of pipe and the other does from the view point of reduction of thermal stresses, compared with the routing with 90 degree elbows.
gas cooled reactor; ; Mizuta, Shunji
JNC TN9400 2000-040, 41 Pages, 2000/03
JNC-TN9400-2000-040.pdf:0.85MB
The corrosion behavior of ferritic stainless steels applied to core components under C0 gas environment was investigated in order to be helpful to fuel design in C0 gas cooled reactor as the feasibility study for fast breeder reactor. The dependence of the corrosion behavior, before a breakaway occurs, on C0 gas temperature, Si and Cr contents of ferritic steels was determined quantitatively. The following correlations to calculate the metal loss thickness was established. X = 4.4w w = √(k
t) k = 
exp( - 5.45[Si]) exp( - 1.09[Cr]) exp( - 11253/T) = 1.65 10
4.40 10
X : metal loss thickness[
ml, w : corrosion weight gain [mg/cm
] k : parabola constant [(mg/cm)/hr], t : time [hr], : constant [Si] : Si content[wt.%], [Cr] : Cr content [wt.%], T : temperature [K]
Harada, Yuhei; Maruyama, Yu; Maeda, Akio; Shibazaki, Hiroaki*; Kudo, Tamotsu; Hidaka, Akihide; Hashimoto, Kazuichiro; Sugimoto, Jun
Journal of Nuclear Science and Technology, 36(10), p.923 - 933, 1999/10
Times Cited Count：3 Percentile：28.69(Nuclear Science & Technology)no abstracts in English
Hashimoto, Kazuichiro; Harada, Yuhei; Maeda, Akio; Maruyama, Yu; Shibazaki, Hiroaki*; Kudo, Tamotsu; Hidaka, Akihide; Sugimoto, Jun
JAERI-Conf 99-005, p.161 - 164, 1999/07
no abstracts in English
Harada, Yuhei; Maruyama, Yu; Maeda, Akio; Shibazaki, Hiroaki*; Kudo, Tamotsu; Hidaka, Akihide; Hashimoto, Kazuichiro; Sugimoto, Jun
JAERI-Conf 99-005, p.171 - 175, 1999/07
no abstracts in English
Maeda, Akio; Maruyama, Yu; Hashimoto, Kazuichiro; Harada, Yuhei; Shibazaki, Hiroaki*; Kudo, Tamotsu; Nakamura, Naohiko; Hidaka, Akihide; Sugimoto, Jun
JAERI-Conf 99-005, p.165 - 170, 1999/07
no abstracts in English
Shibazaki, Hiroaki*; Maruyama, Yu; Kudo, Tamotsu; Hashimoto, Kazuichiro; Maeda, Akio; Harada, Yuhei; Hidaka, Akihide; Sugimoto, Jun
JAERI-Conf 99-005, p.191 - 196, 1999/07
no abstracts in English
JNC TN9400 99-041, 187 Pages, 1999/05
lt is widely recognized that the current seismic design methods for piping involve a large amount of safety margin. From this viewpoint, a series of seismic analyses and evaluations with various design codes were made on typical LMFBR main sodium piping systems. Actual capability against seismic loads were also estimated on the piping systems. Margins contained in the current codes were quantified based on these results, and potential benefits and impacts to the piping seismic design were assessed on possible mitigation of the current code allowables. From the study, the following points were clarified; (1)A combination of inelastic time history analysis and true(without margin) strength capability allows several to twenty times as large seismic load compared with the allowable load with the current methods. (2)The new rule of the ASME is relatively compatible with the results of inelastic analysis evaluation. Hence, this new rule might be a goal for the mitigation of seismic design rule. (3)With this mitigation, seismic design accommodation such as equipping with a large number of seismic supports may become unnecessary.
Ohshima, Hiroyuki; Sakai, Takaaki; ; Yamaguchi, Akira; Nishi, Yoshihisa*; Ueda, Nobuyuki*; *
JNC TN9400 2000-077, 223 Pages, 1999/05
JNC-TN9400-2000-077.pdf:6.24MB
The feasibility study (Phase l) is being carried out at JNC to build up new design concepts of practical fast reactors (FRs) from the viewpoint of economy, safety, effective use of resources, reduction of environmental burden and non-proliferation. This report describes the results of the investigation, related to decay heat removal, core/fuel-assembly thermal-hydraulics and thermal-hydraulic correlations, that was performed in fiscal l999 as a part of the feasibility study. ln the study of the decay heat removal, the effects of several design parameters on the performance of the reactor vessel auxiliary cooling system (RVACS) in a middle-scale sodium-cooled FR were clarified by using a plant dynamic analysis code. The upper limit of RVACS performance was preliminarily estimated at approximately 0.50.6 MWe. Numerical methods for the plant dynamic analysis of gas-and heavy-metal-cooled FRs were also developed. They were applied to the preliminary calculations of the transition from scram to natural circulation and the transient characteristics in tentative plant design concepts were clarified. ln addition, a dimensionless number indicating natural circulation performance was deduced for the comparison of several plant design concepts. With respect to the core/fuel-assembly thermal-hydraulics, numerical analysis methods were improved for the pin-type fuel assembly of gas-and heavy-metal-cooled FRs, the coated-particle- type fuel assembly of helium-gas-cooled FR, and the ductless core of sodium-and heavy-metal-cooled FRs. As preliminary evaluations, thermal-hydraulics in the heavy-metal-cooled FR fuel assembly was compared with sodium-cooled one and thermal-hydraulic analyses of carbon-dioxide- and helium-gas-cooled FR fuel assemblies were performed. The analysis for the fuel assembly with inside duct of sodium-cooled FR was also carried out. The correlations of pressure loss and heat transfer coefficient were investigated for the thermal-hydraulic ...
