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Laird, J. S.; Hirao, Toshio; Onoda, Shinobu; Ito, Hisayoshi
Journal of Applied Physics, 98(1), p.013530_1 - 013530_14, 2005/07
Times Cited Count：45 Percentile：80.59(Physics, Applied)no abstracts in English
Laird, J. S.; Hirao, Toshio; Onoda, Shinobu; Wakasa, Takeshi; Yamakawa, Takeshi; Abe, Hiroshi; Kamiya, Tomihiro; Ito, Hisayoshi
Proceedings of the 6th International Workshop on Radiation Effects on Semiconductor Devices for Space Application (RASEDA-6), p.125 - 129, 2004/10
no abstracts in English
Kusunoki, Tsuyoshi; Odano, Naoteru; Yoritsune, Tsutomu; Fukuhara, Yoshifumi*; Nakajima, Nobuya; Ochiai, Masaaki
Proceedings of the 4th JSME-KSME Thermal Engineering Conference, p.1_61 - 1_66, 2000/00
no abstracts in English
Nakamura, Hiroo; Ladd, P.*; Federici, G.*; Janeschitz, G.*; Schaubel, K. M.*; Sugihara, Masayoshi; Busigin, A.*; Gierszewski, P.*; Hiroki, Seiji; Hurztmeirer, H. S.*; et al.
Fusion Engineering and Design, 39-40, p.883 - 891, 1998/09
Times Cited Count：6 Percentile：44.27(Nuclear Science & Technology)no abstracts in English
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PNC TN9410 98-057, 55 Pages, 1998/05
Existing data of in-pile ramp-type transient-overpower tests (slow TOPs hereafter), such as those of the CABRI-2 and CABRI-FAST tests, the EBR-II TOPI-1E test and the former TREAT tests, were extensively surveyed and this led to a global interpretation which provided a consistency among the tests. Through this study, a basic fuel pin failure mechanism was comprehended and it was confirmed that fuel pins with low to intermediate smear density have a very high failure threshold with significant mitigation effects against fuel-cladding mechanical interactions. Such high failure threshold of low to intermediate smear density fuel is considered to be attributed to the following three effects: (1)absorption of fuel thermal expansion and fuel swelling by void space (porosity or cracks) within the fuel, (2)mitigation of fuel swelling by an early gas escape into the free volume, and (3)mitigation of molten cavity pressurization upon fuel melting. These effects were refrected to the analytical model of the transient fuel behavior code PAPAS-2S. Application of this improved PAPAS-2S model to representative slow TOP tests provided results consistent with the test data, suggesting that the above-mentioned consideration is valid.
Sagawa, Norihiko*
PNC TJ9613 97-002, 95 Pages, 1997/10
The diffusion coefficient of cesium iodide vapor in rare gases was determined by a modified Stefan's method. The rare gas in a diffusion column was saturated with vapor of the cesium iodide, crystals of which were heated to melt at the bottom of the column. By opening a valve united at a top of the column, the vapor diffusing through the column was transferred with the carrier rare gas to an ionization sensor. The concentration of cesium iodide in the carrier gas was continuously monitored with the sensor. The diffusion coefficient was determined by analyzing the transient response of the concentration. Increasing tendency with temperature is observed in the coefficients obtained in argon, kripton and xenon at temperatures between 631 and 691 
C and no significant difference among the coefficients in argon, krypton and xenon.
Takeda, Tetsuaki; A.Ying*; M.A.Abdou*
Fusion Engineering and Design, 28, p.278 - 285, 1995/00
Times Cited Count：5 Percentile：49.89(Nuclear Science & Technology)no abstracts in English
; Yamaguchi, Akira
PNC TN9410 93-213, 28 Pages, 1993/10
Loop-version of Super System Code (SSC-L) has been applied to the analysis of the secondary loop natural circulation test (heated up by the pumps: 4.3MWt) in Monju. The purposes of this study are to validate the computer program and to point out the additional plant data necessary for the analysis of the proposed tests with better accuracy. From the test results, generated heat in the pumps is 4.3 MWt while the removed heat at ACS is 3.4 MWt in the initial steady state. The difference is caused by heat losses from the heat transport system and it is taken into account in the SSC model. The tansient thermohydraulic performance in the secondary heat tansport system simulated using SSC-L is in agreement with the test data. Hence, the pressure loss model in SSC-L is validated and the code is applicable to the natural circulation conditions. Validation of other component models in SSC-L is in progress using Monju data towards a whole plant natural circulation test.
; *; ; Yamaguchi, Akira; ; Sugawara, Satoru
PNC TN9410 91-089, 130 Pages, 1991/03
In-vessel thermohydraulic analysis with multi-dimensional code AQUA was conducted for transient simulating a pump coast down and reactor scram (manual reactor trip event) to confirm efficiency of outer barrel equipments on a large-scale fast breeder reactor. Through the analysis using the AQUA code and the discussion based on their results, the following results have been obtained: [Main Loop Temperature Transient] The transient rate with the outer barrel equipments are approximately equal to the result when an inner barrel was adopted. [Thermal stratification] Axial temperature distributions are approximately equal to the result in the case without an inner barrel. Therefore appearance of an axial temperature distribution can be neglected from a structural design. [Circumferential Temperature Distribution] Maximum temperature gradient 104C/m was confirmed. The value is equivalent to three times of that when an inner barrel was not adopted. Further investigation on a thermal stress at reactor vessel is necessary. [Sodium Surface Velocity] Maximum velocity is the same as that described for the case without an inner barrel. From the above results, it is concluded that the outer barrel considered here is an efficient equipment to relax the main loop temperature transient.
*
PNC TN9410 90-103, 126 Pages, 1990/07
Thermal transient strength tests of a Welded Vessel Model is carring out in Therma1 Transient Strength Test Facility for Structures. The objective of this test is to grasp the thermal transient strength of fast breeder reactor's components and to develop a life estimation method for creep fatigue failure. The model is designed and fabricated in order to grasp the thermal creep-fatigue life of SUS304 and modified SUS316 under the noteworthy typical stress distribution encountered in the structures design of Fast Breeder Reactors. This report shows the results of thermal stress analysis and creep-fatigue damage evaluation of the model. The analytical model was divided the test model into five portions, namely a inlet nozzle, an upper vessel, a lower vessel, an outlet nozzle and an inner shell ring. Thermal transient condition used in the analysis is the same that used in design of the model.
*
PNC TN9410 90-096, 162 Pages, 1990/06
Thermal transient strength tests in using the Thermal Transient Strength Test Facility for Structures (TTS) have been carried out to develop the design method for fast breeder reactor components under thermal loadings. The design and the fabrication of the Welded Vessel Model which is to be tested at TTS are described in this report. In this model, the noteworthy typical shape and stress distribution as seen in the structural design of Large Fast Breeder Reactor (LFBR) and the modified SUS316 which is a hopeful material for LFBR are incorporated. The Welded vessel Model is a cocoon-like model which has 25mm thickness, 2210mm hight and 850mm inner diameter, and it has seven test portions of interest, namely inlet/outlet nozzle, upper/skirt Y-janctions, vessel ring, inner shell ring and inner shell support flange. In design of this model, thermal transient conditions were established by performing Hydraulic-thermal analyses, and then, heat transfer analyses and thermal stress analyses were performed. Test portions of interest were evaluated using the special design guide for TTS tests. Materials and welding method adopted are basically the ones applicable to LFBR's reactor structures.
Nabeshima, Kunihiko; Kusunoki, Tsuyoshi; Shimazaki, Junya; Shinohara, Yoshikuni
JAERI-M 90-040, 117 Pages, 1990/03
no abstracts in English
Kawamura, Hiroshi; Ando, Hiroei
Nihon Genshiryoku Gakkai-Shi, 31(7), p.852 - 860, 1989/07
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)no abstracts in English
*; *; *; *; Nakanishi, Seiji; *
PNC TN9410 88-131, 75 Pages, 1988/08
As a series of the Study on the Main Design Parameters of Large Scale Fast Breeder Reactor (II) in 1987, the transient behavior at the loss of primary piping integrity accident of the loop-type plant of the Key Technological Design Study (II) in 1985 was analyzed by the FBR system code SSC-L, and then the effects of the coolant leakage on the core coolability were evaluated. (1)In case of the leakage from a crack opening area of 1cm
, whieh was rationalized by fracture mechanics, at the cold leg piping near the inlet nozzle of the reactor vessel, the maximum leak mass flow rate was immediately reached to 3.6kg/sec after pipe break, and the saturated leak mass flow rate was reached to 0.9kg/sec at 300sec in the pump pony motor driving condition. (2)In case of the leakage from 1cm area as the originated event, with failure of the succeed of pony motor driving in two-loops due to failure of the starting of a emergency diesel generator as the single-failure criteria, the maximum cladding temperature was reached to 758C, therefore the reactor core was not damaged seriously, and the core coolability was secured sufficiently. (3)In order to compare the effects of the rationalization of crack opening area, in case of the enlargement of the leakage area from 1cm to 0.25D
t(25cm in this analysis), which was assumed in prototype reactor "MONJU", the maximum cladding temperature was increased only about 5C compared with that of the 1cm area. (4)Taking aim to get the setting ground of the source terms on the located evaluation, as a superposition of the obstruct condition on the core coolability, in case of the failure of the succeed of pony motor driving in three-loops except the accident loop, the maximum cladding temperature was reached to 847C (1cm area), and reached to 854C (25cm area) respectively, so both results were exceeded 830C, which was set up as the restriction temperat
*; *; Nakanishi, Seiji; *
PNC TN9410 88-103, 115 Pages, 1988/08
Current emphasis in the development of liquid-metal cooled fast breeder reactors (LMFBRs) is placed on the reduction of the plant construction cost without spoiling the safety in view of its practical application. So much effort is paid for the cost reduction. The reduction of piping length of heat transport system piping is considered as an effective measure for the cost reduction. An application of component floating support piping which brings good results in LWR leads to reduction of piping length and it is an effective measure as well as bellows expansion joint piping and high-chrome-steel piping in LMFBR. A design study for the application of the IHX floating support piping system to primary main heat transport system piping of LMFBR was conducted to demonstrate the adequacy of the piping system by using the improved structural design method considering the characteristics of LMFBR. The stress analysis of piping due to dead weight, thermal expansion at steady and transient conditions, and earthquake was performed, while the nozzles stress due to internal pressure, dead weight, earthquake and thermal expansion reaction force, and thermal transients was analyzed. It was confirmed that the analytical results were satisfied the allowable values and the piping support equipments were highly put into practical use. Therefore it was concluded that the adequacy of the IHX floating support piping system was demonstrated.
Takeda, Toshikazu*; Unesaki, Hironobu*; Kurisaka, Kenichi*; Sakuma, Hiroomi*; Shimoda, Masayuki*; Ito, Noboru*; Kugo, Teruhiko*; Aoki, Shigeaki*; Uto, Nariaki*; Tanaka, Motonari*
PNC TJ2605 88-001, 230 Pages, 1988/03
no abstracts in English
*; *;
JAERI-M 83-081, 96 Pages, 1983/05
no abstracts in English
*; *;
JAERI-M 83-048, 120 Pages, 1983/03
no abstracts in English
