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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.
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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.
*; *; *; *; 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 758
C, 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 88-111, 134 Pages, 1988/08
In order to rationalize the fuel handling system of Large Scale-up Fast Breeder Reactor, the equipment design of the Ex-vessel Storage (EVS) type fuel handling system was conducted in 1986. It can be considered that EVS type is coordinate with in-Vessel Storage (IVS) type from a viewpoint of materials and advantageous to Large Scale-up Fast Breeder Reactor for (1) reduction of reactor vessel diameter, (2) removal of the fuel handling process from sodium environment to water, and (3) realizationof ex-vessel NIS. As concerns sodium pot EVS type, the heat removal analysis at normal and transient conditions was performed to confirm the adequacy of the results gained from the equipment design in 1986 and to investigate requirements for the system. It was confirmed from the analysis results that the design was adequate and the requirements were extracted.
