Core disruptive accident (CDA) In-vessel retention (IVR) Sodium-cooled fast reactor (SFR) Unprotected loss of flow (ULOF) SAS4A SIMMER-III/IV a b s t r a c t In the original licensing application for the prototype fast-breeder reactor, MONJU, the event progression during an unprotected loss of flow (ULOF), which is one of the technically inconceivable events postulated beyond design basis, was evaluated. Through this evaluation, it was confirmed that radiological consequences could be suitably limited even if mechanical energy was released. Following the Fukushima-Daiichi accident, a new nuclear safety regulation has become effective in Japan. The conformity of MONJU to this new regulation should hence be investigated. The objectives of the present study are to conduct a preliminary evaluation of ULOF for MONJU, reflecting the knowledge obtained after the original licensing application through CABRI experiments and EAGLE projects, and to gain the prospect of in-vessel retention for the conformity of MONJU to the new regulation. The preliminary evaluation in the present study showed that no significant mechanical energy release would take place, and that thermal failure of the reactor vessel could be avoided by the stable cooling of disrupted-core materials. This result suggests that the prospect of invessel retention against ULOF, which lies within the bounds of the original licensing evaluation and conforms to the new nuclear safety regulation, will be gained.
Sodium-cooled Fast Reactor (SFR) core is not in the most reactive configuration and relocation of core materials may lead recriticality and generate energetics. Therefore, Core Disruptive Accident (CDA) has been considered as one of the important safety issues from the beginning of SFR development history. In the evaluation of CDA, the whole sequence is categorized into some phases (Suzuki et al., 2015), e.g., Initiating Phase (IP), Transition Phase (TP), and Post-Accident Material Relocation/Post-Accident Heat Removal (PAMR/PAHR) phase, and the most appropriate method for each phase has been developed to precisely analyze event progressions. In IP which is the earliest stage of CDA, damaged core regions are inside wrapper tubes and materials relocate mainly in axial direction. SAS4A code
SAS-SFR (derived from SAS4A) is presently the most advanced computer code for simulation of the primary phase of the Core Disruptive Accident (CDA) of MOX-fueled Sodium-cooled Fast Reactors (SFR). In the past two decades, intensive model improvement works have been conducted for SAS-SFR utilizing the experimental data from the CABRI programs. The main target of the present work is to confirm validity of these improved models through a systematic and comprehensive set of test analyses to demonstrate that the improved models has a sufficient quality assurance level for applications to reactor conditions. In order to reach these objectives, an approach of PIRT (Phenomena Identification and Ranking Table) on a set of accident scenarios has been applied. Based on the fact that there have been a significant amount of validation studies for decades, development of the code validation matrix concentrated on key issues. Different accident scenarios have been chosen for the PIRT considering typical SFR accident transients that address a large range of phenomena. As the most important and typical Core Disruptive Accident scenarios leading to generalized core melting and to be addressed with SAS-SFR in the present study, ULOF (Unprotected Loss Of Flow), UTOP (Unprotected Transient OverPower) and ULOHS (Unprotected Loss Of Heat Sink) are selected. The PIRT process applied to a given accident scenario consists in an identification of the phenomena involved during the accident, the evaluation of the importance of the phenomena regarding to the evolution and consequences, and the evaluation of the status of knowledge based on the review of available experimental results. The identified phenomena involved in ULOF are explained as follows for the primary phase. Starting from initiating events, a loss of grid power leading to flow coast down without scram is assumed. The scenario up to coolant boiling is the main point within the first part of the ULOF phenomenological chart. Those elements related to reactivity feedback, such as heat up of coolant, fuel and various structures and their deformation due to the thermal transient are picked up. Depending on the time scale before boiling starts, primary, secondary and tertiary loop heat transfer including the DHR (Decay Heat Removal) system response is concerned since it defines the core inlet coolant temperature. Core inlet coolant temperature gives direct impact on the thermal condition of the core. It also affects reactivity through thermal expansion of the grid plate. In the second part of the ULOF phenomenological chart, elements such as coolant boiling, mechanical response of the fuel pin leading to cladding failure, FCI (Fuel-Coolant Interaction) and post-failure material relocation are picked up. This part of the chart is basically common to the ULOHS. Respective identified phenomena are to be simulated in the SAS-SFR code. To validate the function of the models in the code, ten high priority CABRI experiments are selected. Validation studies on these tests are underway. With the present study, important phenomena involved in ULOF, UTOP and ULOHS were identified and an evaluation matrix for the selected CABRI experiments was developed.
In the original licensing application for the prototype fast-breeder reactor, MONJU, the event progression during an unprotected loss-of-flow (ULOF), which is one of the technically inconceivable events postulated beyond design basis, was evaluated. Through this evaluation, it was confirmed that radiological consequences could be suitably limited even if mechanical energy was released. Following the Fukushima-Daiichi accident, a new nuclear safety regulation has become effective in Japan. The conformity of MONJU to this new regulation, hence, should be investigated. The objectives of the present study are to conduct a preliminary evaluation of ULOF for MONJU reflecting the knowledge newly obtained after the original licensing application, and to gain the prospect of In-Vessel Retention (IVR) for the conformity of MONJU to the new regulation. In the evaluation of event progressions during ULOF, the whole sequence was categorized into 1) initiating, 2) transition, and 3) post-accident-material-relocation/post-accident-heat-removal (PAMR/PAHR) phases. In the initiating phase, fuel pin disruption caused by coolant boiling would result in axial fuel dispersion in subassembly (SA). In the transition phase, molten-core pool would be formed due to the failure of SA walls, and the molten fuel would be discharged through the control-rod guide tubes (CRGTs). In the PAMR/PAHR phase, molten fuels discharged through CRGTs would be relocated and be stably cooled in the lower plenum by decay-heat removal. The methodology of the present study and its results can be summarized as below: 1) The initiating phase was evaluated by SAS4A code reflecting the models and parameters for fuel-pin disruption and fuel dispersions based on the CABRI experiments. Contrary to the original licensing evaluation showing 380 MJ in mechanical energy release under conservative conditions, the present evaluation showed that no significant energy release would take place. 2) The transition phase was evaluated by 3-dimensional SIMMER-IV code reflecting the models and parameters for CRGT failure and molten-fuel discharge based on the EAGLE experiments. Contrary to the past 2-dimensional evaluation showing 150 MJ in mechanical energy release under conservative conditions, the present evaluation showed that the released mechanical energy would be remarkably reduced because the non-physical axisymmetric/coherent fuel compaction peculiar to 2-dimensional evaluation was appropriately mitigated in 3-dimensional evaluation. 3) The PAMR/PAHR phase was evaluated by S-COPD, FLUENT codes and heat-balance calculations reflecting the present evaluation of the precedent phases. Contrary to the past evaluation involving the uncertainties in molten-fuel fragmentation and debris-bed formation, the present evaluation showed that stable cooling of discharged core materials could be achieved even if fragmentation was incomplete. The preliminary evaluation in the present study showed that no significant mechanical energy release would take place, and that thermal failure of the reactor vessel could be avoided by the stable cooling of disrupted-core materials. This result suggests that the prospect of IVR against ULOF, which lies within the bounds of the original licensing evaluation and conforms to the new nuclear safety regulation, will be gained.
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