Masataka Hidaka

Development of severe accident analysis code SAMPSON in IMPACT project

IMPACT is the name of a program and of specific simulation software, which will perform full-scope and detailed calculations of various phenomena in a nuclear power plant for a wide range of event scenarios. The four years of the IMPACT project Phase 1 have been completed, and each analysis module of the prototype version of the severe accident analysis code SAMPSON has been developed and verified by comparison with separate-effect test data. Verification of the integrated code with combinations of up to 11 analysis modules has been conducted, with the Analysis Control Module, to demonstrate the code capability and integrity. A 10-inch cold leg failure Loss of Coolant Accident in the Surry Plant was the assumed initiating event. The system analysis was divided into two cases; one was an in-vessel retention analysis when gap cooling was effective, the other was an analysis of phenomena when the event was extended to ex-vessel due to the reactor pressure vessel failure when gap cooling was not sufficient. U...

Verification for Flow Analysis Capability in the Model of Three-Dimensional Natural Convection with Simultaneous Spreading, Melting and Solidification for the Debris Coolability Analysis Module in the Severe Accident Analysis Code 'SAMPSON', (I)

The debris coolability analysis module in the severe accident analysis code ‘SAMPSON’ has been enhanced to predict more mechanistically the safety margin of present reactor pressure vessels in a severe accident. The module calculates debris three-dimensional natural convection with simultaneous spreading, melting and solidification using the ‘debris spreading-cooling model’ in combination with the temperature distribution of the vessel wall and it evaluates the wall failure. Debris spreading is solved by the free surface calculation method in which the height function is applied. The model makes possible a multiplex heat and mass transfer analysis with flow spearhead and melt front transportation for a single-phase flow analysis code through the resetting of two types of mesh attributions and re-arrangement of the pressure matrix at each time step. The results calculated with the present model are compared with the results from a water spreading experiment. The comparisons verify the model capability for predictions of debris flow in the spreading process. The module provides a good tool for prediction of the reactor safety margin in a severe accident through the three-dimensional natural convection analysis of debris with simultaneous spreading, melting and solidification.

Development of Debris Coolability Analysis Module in Severe Accident Analysis Code SAMPSON for IMPACT Project

Debris coolability in the lower plenum of the reactor pressure vessel is an important factor for the evaluation of in-vessel debris retention. The debris coolability analysis module has been developed to predict more mechanistically the safety margin of the present reactor vessels in a severe accident. The module calculates debris spreading and cooling through melting and solidification in combination with the temperature distribution of the vessel wall and it evaluates the wall failure. Debris spreading is solved by the explicit method on a quasi-three-dimensional scheme and debris coolability is solved on the basis of natural convection analysis with melting and solidification. The calculated results for spreading were compared with the results from a water spreading experiment on the floor and the results for coolability were compared with those from an n-octadecane melting experiment in the rectangular vessel. The comparisons showed the capability for predictions of the spearhead transportation in the...

Model Verification of the Debris Coolability Analysis Module in the Severe Accident Analysis Code ‘SAMPSON’

The debris coolability analysis module in the severe accident analysis code ‘SAMPSON’ has been enhanced to predict more mechanistically the safety margin of present reactor pressure vessels in a severe accident. The module calculates debris spreading and cooling through melting and solidification in combination with the temperature distribution of the vessel wall and it evaluates the wall failure. Debris cooling after spreading is solved on the basis of natural convection analysis with melting and solidification on three-dimensional Cartesian co-ordinates. The calculated results for the cooling model are compared with the results from a three-dimensional natural convection experiment. The comparisons show the module capability for predictions of the debris temperature in the cooling process. Furthermore, it is seen that the prediction capability in the thermal load of the vessel wall is improved, since the penetration nozzles melting is modeled and combined with the cooling model. The module provides a good tool for the prediction of the reactor safety margin in a severe accident through the three-dimensional analysis of debris cooling.

Improvement of molten core–concrete interaction model in debris spreading analysis module with consideration of concrete degradation by heat

ABSTRACT Understanding the situation inside of the reactors at TEPCOs Fukushima Daiichi Nuclear Power Plant and planning of the methods for debris removal are important for decommissioning the reactors. A debris spreading analysis (DSA) module in the severe accident analysis code SAMPSON has been improved and verified to analyze composite phenomena of molten core (debris) spreading on a reactor containment floor and concrete erosion to the inside of the floor by molten core–concrete interaction (MCCI). The primary models in the DSA module were three-dimensional natural convection with simultaneous spreading, melting and solidification in an open space. In addition to these, the analysis capability has been improved to treat phenomena in a closed space, such as debris eroding laterally under concrete floors at the bottom of the sump pit which is done by an advanced method for boundary processing. A buffer cell for flow analysis, which is defined by a different array variable, is arranged in the same coordinates of the concrete cell (structure cell). Mass, momentum, and the advection term of energy between the debris melt cells and the buffer cells are solved. At the same instant, the heat transfer is calculated between the debris melt cells and the structure cells coexisting side by side with the buffer cells. In this study, technical knowledge regarding changes in physical properties due to thermal degradation of concrete was considered for the prediction of erosion rate, and the DSA module with the models noted above was verified by comparison with erosion data of the core–concrete interaction tests in the OECD/MCCI program. The calculated erosion depth, width, and erosion rate under the concrete floor showed good agreement with the test data and the analysis capability of the module was confirmed.

Development of the models for advection-diffusion of eroded concrete into debris and concrete volume reduction in molten core-concrete interactions

ABSTRACT Models for the three-dimensional (3D) advection, diffusion, and volume reduction of eroded concrete into molten core are being developed. As part of the assessment of the reactor interior at TEPCOs Fukushima Daiichi Nuclear Power Plant, analytical models of molten core–concrete interaction (MCCI) to predict locations and condition of molten core (debris) have been improved in the debris spreading analysis (DSA) module of the severe accident analysis code SAMPSON. In addition to the primary model for 3D natural convection with simultaneous spreading, melting, and solidification in an open space, the analysis model to treat phenomena in a closed space, such as debris eroding laterally under concrete floors at the bottom of the sump pits, had been improved. This modeling with practical applicability is referred to as the full-3D MCCI model. This paper presents modeling of the advection and diffusion of eroded concrete into debris melt and calculation processes that were installed for simulating volume reduction when concrete decomposed. They were developed and incorporated into the full-3D MCCI model. The advanced DSA module with the models noted above was validated using MCCI test data. The calculated erosion rates agreed with the test data within a margin of about 16%.

Experimental Study on Gas Carry-Under in Liquid Down Flow from a Two-Phase Mixture

The gas carry-under characteristics in liquid down flow from a two-phase mixture flow have been studied for various flow parameters, based on experiments with a small scale air- water system simulating the concept of a natural circulation BWR with no separators. For high void fraction in the riser, as the liquid superficial velocity jf increased to 0.17 m/s, the void fraction in the lower part of the downcomer αd increased sharply due to the descent of comparatively large bubbles (diameter: about 4–6mm). In the region of jf> 0.17m/s, on increasing jf, the void fraction αd increased until it reached a maximum value at jf.3. For liquid descending velocities higher than 0.3 m/s, αd became almost constant and the level of the mixture above the riser had little effect on the void fraction ad due to the phase separation of the large bubbles formed by bubble coalescence in the upper part of the downcomer. The void fraction αd increased as the void fraction αr increased until bubble coalescence occurred in the up...