Accidents In Sodium-cooled Fast Reactors Research Articles

Detailed analyses of key phenomena in core disruptive accidents of sodium-cooled fast reactors by the COMPASS code

A five-year research project has been initiated in 2005 to develop a code based on the MPS (Moving Particle Semi-implicit) method for detailed analysis of key phenomena in core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). The code is named COMPASS (Computer Code with Moving Particle Semi-implicit for Reactor Safety Analysis). The key phenomena include (1) fuel pin failure and disruption, (2) molten pool boiling, (3) melt freezing and blockage formation, (4) duct wall failure, (5) low-energy disruptive core motion, (6) debris-bed coolability, and (7) metal–fuel pin failure. Validation study of COMPASS is progressing for these key phenomena. In this paper, recent COMPASS results of detailed analyses for the several key phenomena are summarized. Simulations of GEYSER and THEFIS experiments were performed for dispersion and freezing behaviors of molten materials in narrow flow channels. In particular, the latter experiment using melt–solid mixture is also related to fundamental behavior of low energy disruptive core. CABRI-TPA2 experiment was simulated for boiling behavior of molten core pool. Expected mechanism of heat transfer between molten fuel and steel mixture was reproduced by the simulation. Analyses of structural dynamics using elastoplastic mechanics and fracture criteria were performed for SCARABEE BE+3 and CABRI E7 experiments. These two analyses are especially focused on thermal and mechanical failure of steel duct wall and fuel pin, respectively. The present results demonstrate COMPASS will be useful to understand and clarify the key phenomena of CDAs in SFRs in details.

Uncertainty Analysis for Unprotected Accidents in Sodium-Cooled Fast Reactors

Reactor safety analyses often utilize a deterministic approach where in addition to performing best estimate calculations, uncertainty is accommodated by performing calculations with pessimistic values for input parameters that are important to safety. Here, a stochastic approach is considered for explicitly including uncertainty in safety parameters by applying Monte Carlo sampling coupled with established deterministic reactor safety analysis tools. The Monte Carlo approach yields frequency distributions for reactor safety metrics (e.g., peak temperatures) that can be compared to performance limits, allowing for an improved determination of the safety margin and a clear determination of which safety parameters are most important to the transient response. Because the approach enables the estimation of probabilities for violating safety boundaries, it should be useful in a risk-based regulatory environment. It has the advantage of not requiring any substantial rewriting of existing safety analysis computer codes.

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.

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.

Uncertainty Analysis for Unprotected Accidents in Sodium-Cooled Fast Reactors

Uncertainty Analysis for Unprotected Accidents in Sodium-Cooled Fast Reactors

Unanticipated accidents in sodium-cooled fast reactors

The definitions and requirements of normative documents for unanticipated accidents at nuclear power plants with fast reactors are analyzed. Definitions are constructed between one another and with a collection of scenarios which can lead to unanticipated accidents, likewise determined by normative documents independently of the probability of these accidents actually happening. It is concluded that the normative approaches to fast-reactor safety must be refined with respect to strengthening the probabilistic criteria as a tool limiting the list of required unanticipated accidents for validating reactor safety. Special attention is devoted to the need to strengthen the motivation of designers to make the maximum possible use of passively triggered safety systems.

Development of a Computational Framework on Fluid-Solid Mixture Flow Simulations for the COMPASS Code

The COMPASS code is designed based on the moving particle semi-implicit method to simulate various complex mesoscale phenomena relevant to core disruptive accidents of sodium-cooled fast reactors. In this study, a computational framework for fluid-solid mixture flow simulations was developed for the COMPASS code. The passively moving solid model was used to simulate hydrodynamic interactions between fluid and solids. Mechanical interactions between solids were modeled by the distinct element method. A multi-time-step algorithm was introduced to couple these two calculations. The proposed computational framework for fluid-solid mixture flow simulations was verified by the comparison between experimental and numerical studies on the water-dam break with multiple solid rods.

A three-dimensional neutronics-thermohydraulics simulation of core disruptive accident in sodium-cooled fast reactor

The SIMMER-III code is a two-dimensional, multi-velocity-field, multi-phase, multi-component, Eulerian, fluid-dynamics code coupled with a fuel-pin model and a space- and energy-dependent neutron transport kinetics model. Since the three-dimensional representation of the core enables realistic distribution of the materials constituting the core, including control rods, SIMMER-IV has been developed as a direct extension of SIMMER-III to three dimensions with retaining exactly the same physical models as SIMMER-III. Recently, the parallelization of SIMMER-IV has been achieved, allowing application to reactor calculations within available computational resources. A three-dimensional simulation using SIMMER-IV has drawn more realistic accident scenario including a late stage during the transition phase. Additional static neutronic calculations identified major factors significantly influencing the reactivity change shown in the SIMMER-IV simulation.

Improvement of Basic Fluid Dynamics Models for the COMPASS Code

The COMPASS code is a new next generation safety analysis code to provide local information for various key phenomena in core disruptive accidents of sodium-cooled fast reactors, which is based on the moving particle semi-implicit (MPS) method. In this study, improvement of basic fluid dynamics models for the COMPASS code was carried out and verified with fundamental verification calculations. A fully implicit pressure solution algorithm was introduced to improve the numerical stability of MPS simulations. With a newly developed free surface model, numerical difficulty caused by poor pressure solutions is overcome by involving free surface particles in the pressure Poisson equation. In addition, applicability of the MPS method to interactions between fluid and multi-solid bodies was investigated in comparison with dam-break experiments with solid balls. It was found that the PISO algorithm and free surface model makes simulation with the passively moving solid model stable numerically. The characteristic behavior of solid balls was successfully reproduced by the present numerical simulations.