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JNC TN9400 2000-038, 98 Pages, 2000/04
JNC-TN9400-2000-038.pdf:7.49MB
As an effort in the feasibility study on commercialized Fast Breeder Reactor cycle systems, an evaluation of the measures to prevent the energetic re-criticality in sodium-cooled large MOX core, which is one of the candidates for the commercialized reactor, has been performed. The core disruptive accident analysis of Demonstration FBR showed that the fuel compaction of the molten fuel by radial motion in a large molten core pool had a potential to drive the severe super-prompt re-criticality phenomena in ULOF sequence. ln order to prevent occurrence of the energetic re-criticality, a subassembly with an inner duct and the removal of a part of LAB are suggested based on CMR (Controlled Material Relocation) concept. The objective of this study is the comparison of the effectiveness of CMR among these measures by the analysis using SIMMER-III. The molten fuel in the subassembly with inner duct flows out faster than that from other measures. The subassembly with inner duct will work effectively in preventing energetic re-criticality. Though the molten fuel in the subassembly without a part of LAB flows out a little slower, it is still one of the promising measures. However, the UAB should be also removed from the same pin to prevent the fuel re-entries into the core region due to the pressurization by FCl below the core, unless it disturbs the core performance. The effect of the axial fuel length of the center pin to CMR behavior is small, compared to the effect of the existence of UAB.
; ; *;
JNC TN9400 2000-043, 23 Pages, 2000/03
ln the feasibility study on commercialized fast breeder reactor (FBR) cycle systems conducted in JNC, it is required for candidate FBR plants that the level of safety should be enhanced so as to assure: (1)Comparative or superior safety level to that of light water reactors (LWRs), and (2)releaf of the public from anxiety about potential nuclear hazard. Adopting Passive safety characteristics is one of the measures. To attain the above safety objective, we considered implication of the basic safety principles for nuclear power plants that were created by the international nuclear safety advisory group of IAEA. The way to relieve from the anxiety was also taken into account. Then a definite safety objective was set from the standpoint of prevention of core disruptive accident (CDA). Furthermore, as a definite safety goal relating to reactor coresafety, elimination of re-criticality issues under CDA was set by considering characteristics of FBR in comparison with those of LWR. To examine measures for elimination of re-criticality issues, we developed a quick method to estimate possibility of re-criticality under CDA, by drawing a map about criticality characteristics under CDA in various degraded cores. Then hopeful measures were proposed for elimination of re-criticality issues in sodium-cooled FBR with mixed-oxide fuel. Molten fuel discharge behavior of their measures was preliminarily analyzed. We concluded that discharge capability of "a subassembly with an internal duct" was effective, and that "partial removal of axial blanket" was also effective as one of the measures though it has small effect on core performance.
*; *
JNC TN9200 2000-001, 133 Pages, 2000/02
The 11th Meeting for Reporting Safety Research on FBR and ATR was held at the exhibition hall (TECHNO O-ARAI) in OEC on the 15th of December in 1999. The reports of each subject in FY1996-1998 were presented before discussion at this meeting. The 11 subjects had been selected from the subjects (34 in total) on power reactor in fast breeder reactor, earthquake-proof and probabilistic safety assessment according to the decisions of sub-meetings in Sectional Meeting of Safety Research. This meeting was open to the public, and large attendance outside of JNC was invited for the purpose of getting some advice from related specialists. This report contains presentation papers, questions and answers, list of attendance, etc. Refer to the JNC open report for detailed results of safety research in FY1996-1998.
Kikuchi, Shin
JNC TN9450 98-001, 87 Pages, 1998/10
This paper is arranged PIE data in order to provide data sheet for evaluation of core distortion. In this paper, there are two kinds of PIE data; one is face to face distance data of the spacer pad part of lower guide tube of control lod. The is the axial profile of fuel assemblies bowing.
Mishima, Kaichiro*; Hibiki, Takashi*; *; Tobita, Yoshiharu
PNC TY9604 96-003, 10 Pages, 1996/05
no abstracts in English
; ; Uto, Nariaki; Yamaguchi, Akira; Kamide, Hideki; Ohshima, Hiroyuki; Hayashi, Kenji
PNC TN9410 94-235, 135 Pages, 1994/08
In this report passive prevention and mitigation measures with regard to core disruptive accident in future large scale liquid metal cooled fast breeder reactors are discussed and assessed. First the criteria for the assessment of passive safety measures are proposed, and the commonly proposed passive prevention and mitigation measures are briefly reviewed. Then innovative prevention and mitigation measures are newly proposed to provide additional mechanisms to limit the core damage or to prevent a recriticality event during the core disruption process. After assessing these passive measures based on the proposed criteria, appropriate combinations of the measures are recommended. Further, required R&D programs to confirm their effectiveness are described including necessity of a new in-pile experimental program.
Hioki, Kazumasa
PNC TN9410 94-188, 160 Pages, 1994/05
The Systems Analysis Section has been performing a probabilistic Safety Assessment (PSA) on a large fast breeder reactor (FBR) since JFY 1992. The objective of the study is to apply the PSA method to a plant in a conceptual design stage, develop system models, perform quantitative analyses and systematic evaluation, supply valuable insights to enhance reliability and safety, and reflect them to the basic design. The plant analyzed is a 600MWe class large FBR designed by the Plant Engineering Section in the "Large FBR design study" that has been performed since JFY 1990. The failure probability of the Decay Heat Removal System (DHRS) can be reduced approximately two orders if the Water Steam System (WSS) can remove the decay heat for the first 24 hours. The frequency of PLOHS, however, is not reduced to less than one third because the WSS cannot be used for some initiating events and the PLOHS frequency is dominated by the failure probability of DHRS without the WSS. The failure probability of DHRS is dominated by the common cause failures (CCFs) of vanes, dampers and valves around the air-coolers in the Auxiliary Cooling System (ACS). Therefore it is most important to eliminate the CCFs. Assuming that the CCFs have been eliminated by diversifying the components, the frequencies of PLOHS were evaluated. An analysis has shown that if the WSS can remove the decay heat alone, the PLOHS frequency is reduced approximately two orders. In this case the PLOHS frequency is dominated by the failure probability of the DHRS right after the reactor shutdown. The most effective way to reduce the PLOHS frequency is to increasc the redundancy of the DHRS for the first few hours after reactor shutdown. It is known through the experience of preceding plants that the success criteria can be relaxed to one loop natural circulation instead of forced circulation in the best estimate evaluation. It was shown that under such condition, the PLOHS frequency can be as low as 10
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; ; ; Miyake, Osamu
PNC TN9410 92-068, 73 Pages, 1992/03
ln order to be useful for selecting specifications about the safety of the large scale fast breeder rector on and after Monju, following items were studied. (1)Design conditions of the reactor containment, (2)scenarios as to evaluation of core disruptive accident, and (3)applicability of the method of PSA. Technical documents provided for these studies are su㎜arized in this report.
*; *; *; *; 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
*; Ohshima, Hiroyuki
PNC TN9410 88-006, 71 Pages, 1988/01
Reactivity Feedback Modeling in Super System Code (SSC) has been improved to analyze the whole plant thermal hydraulic response to an anticipated transient without scram (ATWS) in a liquid metal fast breeder reactor (LMFBR). First of all, two-dimensional (2D) fluid flow and heat transfer modeling of reactor upper plenum (UP) has been modified. The heat transfer between the coolant and the control rod driveline (CRD) can be evaluated based on the 2D, not one-point, temperature distribution calculated by the UP model. The CRD is included as a part of in-plenum structure, and the thermal expansion of it is evaluated assuming the elongation is proportional to the temperature rise of the CRD. The reactivity feedback effect is evaluated using the elongation and the control rod worth. SSC-L is now capable of treating the following reactivity feedback effects caused by; (1)fuel doppler, (2)sodium density, (3)fuel axial expansion, (4)thermal expansion of the in-core structure, (5)thermal expansion of the core support structure, and (6)thermal expansion of the CRD. Whole plant thermal hydraulics during the ATWS accident can be analyzed taking the reactivity feedback effect into consideration more realistically than ever. An ATWS accident, i.e. unprotected loss-of-heat-sink (ULOHS), has been analyzed using SSC-L as an example. It is impossible to mitigate the ATWS consequence without core damage if no design change is made. However, it is found that more than 7 minutes of grace time is available for the remedial action at least if the above mentioned reactivity model is used. The accident progression is not so rapid in general in the ULOHS accident. The relatively slow response implies reactor shutdown can be achieved by a manual scram or a shutdown system actuated at slower speed can be utilized for mitigating the ATWS consequence.In the present analysis of CRD thermal expansion, it is assumed that the deformation of the CRD is one-dimensional, linear and elastic. It ...
*; *; Aoi, Sadanori*
PNC TN941 82-74VOL1, 151 Pages, 1982/03
PNC-TN941-82-74VOL1.pdf:7.53MB
A study of hypothetical core disruptive accidents (HCDAs) in the prototype fast breeder reactor Monju (714 MWt) has been conducted by using the SAS3D
accident analysis code. A loss-of-flow (LOF) due to the loss of off-site power and a transient overpower (TOP) due to control assembly withdrawal, both at rated power, are considered as the HCDA initiators with a postulated total failure of the reactor shutdown system. The accident scenarios of each postulated anticipated transient without scram are studied for the three burnup stages of Monju: the beginning-of-initial cycle (BOIC) ; a beginning-of-equilibrium cycle (BOEC); and an end-of-equilibrium cycle (EOEC). The neutronics data used in this study has been obtained by a 3-dimensional HEX-Z diffusion code and the first order perturbation calculations. The reactivity coefficients used in this study are the design nominal values without taking into account their uncertainties. The nominal design value of the maximum positive sodium void worth in Monju is a relatively small value of 2.5$ in the EOEC core. In the 2 cents/sec TOP, the reactor power shows a sudden increase following the onset of FCIs (Molten-Fuel/Coolant Interactions) in high-powered fuel assemblies but the maximum power level reached is less than 5 times the rated power and due to the fuel sweepout negative reactivity in the FCI fuel assemblies, the reactor is shutdown within 0.1 sec at the latest after the first FCI onset. The extent of damaged fuel assemblies is largest in the clean (FP-gas free) BOIC core in which the radial power peaking is smaller than in BOEC and EOEC cores, and about 17% of the fuel assemblies are damaged in the central region of the core. In the equilibrium cycle cores the damage extents are limited to about 5% core-center assemblies and this is smaller than in the BOIC core because of the larger radial power peaking and the rapid fuel sweepout reactivity insertion accelerated by the FP-gas pressure in the ...
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PNC TN941 82-55, 284 Pages, 1982/03
The postdisassembly expansion phase of the Hypothetical Core Disruptive Accident (HCDA) in the MONJU reactor was analyzed by using the SIMMER-II code. Hitherto, the isentropic expansion of fuel vapor has been assumed after the core disassembly phase to estimate the system work energy following a postulated energetic disassembly. Recently, the SIMMER code was applied to analyze che postdisassembly expansion phase for the Clinch River Breeder Reactor (CRBR), and it was shown that the system work energy as a result of an HCDA was remarkably reduced compared with the isentropic expansion. The SIMMER code has attracted attension in the field of postdisassembly expansion analysis because of this possibility of work energy reduction. SIMMER-II was installed at PNC, the O-arai Engineering Center, in May 1980, and has been operational since November 1980. This report is divided into two parts. The first deals with the parameter survey based on the study of the MONJU postdisassembly expansion under simplified initial conditions by Kondo and Aizawa at Los Alamos National Laboratory. The other is based on the results of the initiating phase and of the core disassembly phase analyses by the SAS 3D and VENUS-PM codes, performed at PNC. In the latter, we adopted two cases which yielded largest system kinetic energy in the MONJU system, and we estimated the maximum energy released in the MONJU HCDA by using the SIMMER-II code. The main results obtained are shown below. (1)The maximum system kinetic energy released during the postdisassembly phase of the MONJU HCDA is at most 10 MJ when the active core, upper axial blanket and fission gas plenum are all voided at the initial state. (The maximum system work energy associated with isentropic expansion of the fuel vapor to 1 atm is 992 MJ.) (2)Under the same initial average core fuel temperature, a higher peaking factor of temperature distribution causes a larger system kinetic energy. For example, for temperature ...