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JNC TN9400 2001-048, 21 Pages, 2001/03
JNC-TN9400-2001-048.pdf:0.55MB
In the CABRI-FAST LT1 test, simulating a ULOF (Unprotected Loss of Flow) accident of LMFBR, pin failure took place rather early during the transient. No fuel melting is expected at this failure because the energy injection was too low and a rapid gas-release-like response leading to coolant-channel voiding was observed. This channel voiding was followed by a gradual fuel breakup and axial relocation. With an aid of SAS4A analysis, interpretation of this test was performed. Although the original SAS4A model was not well fitted to this type of early pin failure, the global behavior after the pin failure was reasonably simulated with temporary modifications. Through this study, gas release behavior from the failed fuel pin and its effect on further transient were well understood. It was also demonstrated that the SAS4A code has a potential to simulate the post-failure behavior initiated by a very early pin failure provided that necessary model modification is given.
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JNC TN9400 2001-047, 42 Pages, 2001/03
JNC-TN9400-2001-047.pdf:1.05MB
The LT4 test was performed in the CABRI-FAST in-pile experiment program carried out in 1992
1995. The objectives of this test were to study the fuel pin failure mechanism and to observe the transient fuel motion within the pin and in the coolant channel. The objectives of the present study are to clarify phenomena taking place in the experiment through data evaluation and SAS4A code analysis. Various experimental data have been analyzed with a help of SAS4A code calculation to interpret fuel pin failure mechanism and post-failure material relocation behavior. Through this study, the rapid fissile elongation up to the fuel pin failure was recognized to have potential to delay the failure by about 50 ms, and Probable effect of plenum gas to enhance dispersive fuel relocation has been recognized. And it was confirmed that SAS4A can reasonably simulate rapid molten-fuel ejection from failed fuel pin, rapid fuel relocation within the coolant channel assisted by the Plenum-gas and the fuel freezing in the last Part of transient.
Yamano, Hidemasa; Tobita, Yoshiharu
JNC TN9400 2001-039, 90 Pages, 2001/03
We carried out the analyses of a core-disruptive-accident (CDA) sequence using a SIMMER-IIl code for a pin-type MOX-fueled, carbon-dioxide-cooled fast breeder reactor, which is a one of candidate concepts in Phase I of Feasibility Study on Commercialized Fast Reactor Cycle System. The objectives of this study were: to draw a rough sketch of CDA event sequences, to understand the characteristics of CDA, and to judge whether the measures to eliminate energetic recriticality are necessary or not. In the present analysis, an unprotected loss-of flow(ULOF) accident was chosen as a representative CDA. The analysis showed that a power burst occurred when the reactivity reached prompt criticality as a result of molten cladding relocation out of the core. The negative reactivity feedback due to Doppler effect determined the power peak and the energy release. High pressure generation as a result of rapid heating of the channel gas by the ejected molten fuel and fission gas discharge allowed the molten fuel to disperse mainly toward a downward direction, which has the small flow resistance, then the reactivity settled down under the sub-critical condition. This sequence did not vary much qualitatively, even if a smaller reactivity ramp rate was applied to take into account the incoherency of the molten cladding motion. The gas-cooled reactor has no measures to mitigate the positive reactivity effect caused by the molten cladding motion. However, the mechanical loading to the reactor vessel boundary is negligibly small due to the absence of liquid coolant as a working fluid for the slug impact even if the recriticality occurs. On the other hand, the countermeasure to prevent the upper core structures (UCS) from being accelerated as a solid missile by the high pressure in the core just after the fuel disruption might be necessary. In the present design with the short lower axial blanket (LAB) of 10cm length, the measure for avoiding the energetic recriticality is not ...
kondo, Satoru; Yamano, Hidemasa; Tobita, Yoshiharu; Fujita, Satoshi; Morita, Koji*; Mizuno, Masahiro*; *
JNC TN9400 2001-003, 307 Pages, 2000/11
JNC-TN9400-2001-003.pdf:8.33MB
An advanced safety analysis computer code, SIMMER-III, has been developed at Japan Nuclear Cycle Development Institute (JNC) to more realistically investigate postulated core disruptive accidents in liquid-metal fast reactors. The two-dimensional framework of SIMMER-III fluid dynamics has been extended to three dimensions to a new code, SIMMER-IV, which is currently (in Version 1) coupled with the existing two-dimensional neutronics model. With the completion of the first SIMMER-IV version, the applicability of the code is further enhanced and the many of the known limitations in SIMMER-III are eliminated. The sample calculations demonstrated the general validity of SIMMER-IV. This report describes SIMMER-IV version 1.B, by documenting the models, numerical algorithms and code features, along with the program description and input and output information to aid the users. Further extension of the code is planned to couple the three-dimensional neutronics in the future.
kondo, Satoru; Yamano, Hidemasa; Suzuki, Toru; Tobita, Yoshiharu; Fujita, Satoshi; ; Kamiyama, Kenji
JNC TN9400 2001-002, 318 Pages, 2000/11
JNC-TN9400-2001-002.pdf:8.66MB
An advanced safety analysis computer eode, SIMMER-III, has been developed to investigate postulated core disruptive accidents in liquid-metal fast reactors (LMFRs). SIMMER-III is a two-dimensional, three-velocity-field, multiphase, multicomponent, Eulerian, fluid--dynamics code coupled with a space-dependent neutron kinetics model. By completing and integrating all the physical models originally intended at the beginning of this code development project, SIMMER-III is now applicable to integral reactor calculations and other complex multiphase flow problems. A systematic code assessment program, conducted in collaboration with European research organizations, has shown that the advanced features of the code have resolved many of the limitations and problem areas in the previous SIMMER-II code. In this report, the models, numerical algorithms and code features of SIMMER-III version 2.H are described along with detailed program description. Areas which require future model refinement are also discussed. SIMMER-III Version 2.H, a coupled fluid-dynamics and neutronics code system, is expected to significantly improve the flexibility and reliablity of LMFR safety analyses.
kondo, Satoru; Yamano, Hidemasa; Tobita, Yoshiharu; Fujita, Satoshi; Kamiyama, Kenji; W.Maschek*; P.Coste*
JNC TN9400 2000-105, 777 Pages, 2000/09
JNC-TN9400-2000-105.pdf:33.07MB
The liquid-metal fast reactor (LMFR) safety analysis computer code SIMMER-III successfully reached its first milestone with development of Version 2, a two-dimensional, three-velocity-field, multiphase, multicomponent, Eulerian, fluid-dynamics code coupled with a spase-dependent neutron kinetics model. The development and assessment of SIMMER-III has been participated internationally by Forschungszentrum Karlsruhe, Germany and Commissariat 
l'Energie Atomique, France. To advance the code as a next-generation standard tool for LMFR safety analysis, it was agreed among the partners that a joint code assessment program should be conducted comprehensively and systematically. This program consists of a two-step effort: Phase 1 for fundamental or separate-effect assessment of individual code models; and Phase 2 for integral assessment of key physical phenomena relevant to LMFR safety. THis report describes the achivement of the SIMMER-III Phase 2 assessment program. The report details the results of each of the 34 test problems analyzed, conducted in five major categories of key LMFR safety phenomena, and synthesizes the outcome of the analyses. Through this extensive study, effectively utilizing the international collaboration scheme and world experimental database available, the SIMMER-III code has proved to be basically valid both numerically and physically, with significantly enhanced applicability and flexibility over its predecessor, SIMMER-II. Thus the code is now applicable to integral reactor safety calculations. The study has also identified limitations and problem areas on which future code development and assessment should focus.