Ikken Sato

Safety Strategy of JSFR Eliminating Severe Recriticality Events and Establishing In-Vessel Retention in the Core Disruptive Accident

In the Japan Sodium Cooled Fast Reactor (JSFR) design, elimination of severe power burst events in the Core Disruptive Accident (CDA) is intended as an effective measure to ensure retention of the core materials within the reactor vessel. The design strategy is to control the potential of excessive void reactivity insertion in the initiating phase by selecting appropriate design parameters such as maximum void reactivity on one hand, and to exclude core-wide molten-fuel-pool formation, which has been the main issue of CDA, by introducing an inner duct on the other hand. The effectiveness of these measures is evaluated based on existing experimental data and computer simulation with validated analytical tools. It is judged that the present JSFR design can exclude severe power burst events. Phenomenological consideration of general characteristics and preliminary evaluations for the long-term material relocation and cooling phases gave the perspective that in-vessel retention would be attained with appropriate design measures.

Experimental study on fuel-discharge behaviour through in-core coolant channels

In core-disruptive accidents of sodium-cooled fast reactors, fuel discharge from the core region reduces the possibility of severe re-criticality events. In-core coolant channels with large hydraulic diameters, such as the control-rod guide tube and a concept of the Fuel Assembly with Inner Duct Structure have a potential to provide effective fuel-discharge paths if effects of sodium in these paths on molten fuel discharge are limited. Two series of experiments were conducted to investigate fuel-discharge behaviour through the sodium-filled channels. In the first series of experiments, an alloy with low melting temperature was ejected into a water channel to clarify dominant phenomena for melt discharge through the coolant-filled channel and to develop methodologies for evaluating the effects of coolant on melt discharge. In the second series of experiments, a molten alumina was discharged through the sodium-filled channel in order to verify the applicability of the knowledge and evaluation methodologies obtained in the first series of experiments to the sodium-filled channel. These series of experiments showed that the discharge path can be entirely voided by the vaporisation of a part of the coolant at the initial melt discharge phase that this is followed by coolant vapour expansion and that melt penetrates significantly into the voided channel. Preliminary extrapolation of the present results to the in-core coolant channel suggests that the effects of the sodium on fuel discharge are limited and, therefore, in-core coolant channels will provide effective fuel-discharge paths for reducing neutronic activity.

Transient Heat Transfer Characteristics Between Molten Fuel and Steel with Steel Boiling in the CABRI-TPA2 Test

Abstract In the TPA2 test of the CABRI-RAFT program, which is part of a fast reactor safety study, fuel-to-steel heat transfer characteristics within a molten fuel/steel mixture system have been investigated. This test was performed in the French CABRI reactor and used a test capsule that contained fresh 12.3%-enriched UO2 pellets with embedded stainless steel balls. Following a preheating phase, the capsule was subjected to a transient overpower that resulted in fuel melting and steel vaporization. The observed steel vapor pressure buildup was quite low, which suggested the presence of a mechanism that significantly reduced the fuel-to-steel heat transfer. A detailed experimental data evaluation by SIMMER-III led to one possible interpretation that the steel vaporization at the surface of the steel ball blanketed the steel from the molten fuel.

Fuel Pin Behavior under Slow-Ramp-type Transient-Overpower Conditions in the CABRI-FAST Experiments

In the CABRI-FAST experimental program, four in-pile tests were performed with slow-power-ramptype transient-overpower conditions (called hereafter as “slow TOP”) to study transient fuel pin behavior under inadvertent control-rod-withdrawal-type events in liquid-metal-cooled fast breeder reactors. The slow TOP test with a preirradiated solid-pellet fuel pin under a power ramp rate of approximately 3%Po/s was realized as a comparatory test against an existing test in the CABRI-2 program where approximately 1%Po/s was adopted with the same type of fuel pin. In spite of the different power ramp rates, the evaluated fuel thermal conditions at the observed failure time are quite similar. Three slow TOP tests with the preirradiated annular fuel resulted in no pin failure showing a high failure threshold. Based on posttest examination data and a theoretical evaluation, it was concluded that intrapin free spaces, such as central hole, macroscopic cracks, and fuel-cladding gap, effectively mitigated the fuel cladding mechanical interaction. It was also clarified that cavity pressurization became effective only in the case of a very large amount of fuel melting. These CABRI-FAST slow TOP tests, in combination with the existing CABRI and TREAT tests, provided an extended slow TOP test database under various fuel and transient conditions.

Three-Pin Cluster CABRI Tests Simulating the Unprotected Loss-of-Flow Accident in Sodium-Cooled Fast Reactors

Two three-pin cluster tests simulating the Unprotected Loss-of-Flow (ULOF) accident of Sodiumcooled Fast Reactors (SFRs) were conducted focusing on postfailure fuel relocation and freezing behavior. These tests supplied complementary information to the existing CABRI tests with a single-pin geometry. Based on detailed data evaluation and theoretical interpretation for the three-pin cluster tests, it is concluded that axial fuel relocation and freezing are dominated by local fuel enthalpy, and the relation between penetration length and local fuel enthalpy observed in these CABRI tests is basically applicable to the large-bundle condition. It is also clarified that a fuel/steel mixture tends to create tight blockages near the axial ends of the relocating fuel. Part of the fission gas released from the heating-up and melting fuel is expected to be trapped within the bottled-up region between the upper and lower blockages and will keep this region pressurized for a relatively long period.

Fuel Pin Behavior up to Cladding Failure under Pulse-Type Transient Overpower in the CABRI-FAST and CABRI-RAFT Experiments

In the CABRI-FAST and CABRI-RAFT programs within a collaboration with the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) and Forschungszentrum Karlsruhe (FZK), five pulse-type transient overpower tests were performed in order to study fuel pin behavior and failure condition in the Unprotected Loss-of-Flow (ULOF) accident. In these tests, two types of low-smear-density fuels irradiated in the French Phénix reactor at different burn-up levels were used so that an experimental database extension from the former CABRI-1 and CABRI-2 programs can be obtained. Pin failure took place in three of these tests giving information on the failure threshold. In two tests, no pin failure took place and useful information related to the transient fuel behavior up to failure and failure mechanism was obtained. These test results were interpreted through detailed analysis of experimental data and PAPAS-2S code calculations. In these calculations, pretransient fuel characteristics obtained from the sibling fuels were reflected, such that the uncertainty of the boundary condition can be minimized. Through the comparison among these tests and formerly existing CABRI tests, generalized understanding on the transient fuel behavior was obtained. It was concluded that the low-smear-density fuel mitigates cavity pressurization, thereby enhancing the margin-to-failure. It was also understood that this failure-thresholdenhancing capability is dependent on the type of transient.

Experimental studies on the upward fuel discharge for elimination of severe recriticality during core-disruptive accidents in sodium-cooled fast reactors

In order to eliminate the energetic potential in the case of postulated core-disruptive accidents (CDAs) of sodium-cooled fast reactors, introduction of a fuel subassembly with an inner-duct structure (FAIDUS) has been considered. Recently, a design option of FAIDUS which leads molten fuel to upward discharge has been considered as the reference core design of the Japan Sodium-Cooled Fast Reactor (JSFR). In this study, a series of experiments which consisted of three out-of-pile tests and one in-pile test were conducted to obtain experimental knowledge of the upward discharge of molten fuel. Experimental data which showed a sequence of upward fuel discharge and effects of initial pressure conditions on upward discharge were obtained through the out-of-pile and in-pile test. Preliminary extrapolation of the present results to the supposed condition in the early phase of the CDA in the JSFR design suggests that the sufficient upward flow rate of molten fuel is expected to prevent the core melting from progressing beyond the fuel subassembly scale and that the upward discharge option will be effective in eliminating the energetic potential.

Development of PIRT (Phenomena Identification and Ranking Table) for SAS-SFR (SAS4A) Validation

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.Copyright