Influence of RCS Depressurization Strategy on Hydrogen Risk

For advanced passive PWR, reactor coolant system (RCS) depressurization through automatic depressurization system (ADS) is an important measurement to avoid high-pressure melt ejection and direct containment heating. It allows injection from passive core cooling system and the implement of in-vessel retention. However, it has negative impact that hydrogen in the RCS can be released to the containment together with coolant, which may lead to hydrogen burning or even explosion in the containment. Therefore, this paper analyzes the RCS depressurization strategy during severe accident, and evaluates its negative impact. Severe accident sequences induced by station black out (SBO) was selected and analyzed with integral severe accident analysis code as a typical high pressure core melt accident scenario. Different depressurization strategies with ADS system were discussed based on Severe Accident Management Guideline (SAMG.) ADS valves were manually opened at a core exit temperature of 923 K with 20min delay for operator reaction. Both depressurization effect and hydrogen risk were evaluated for different strategies. Hydrogen distribution was calculated, which was used to determine the combustion mode in different compartments. Result shows all three strategies analyzed in this paper can depressurize the RCS effectively. And opening the ADS stage 1–3 valves causes rapidly increase of the hydrogen concentration in the in-containment refueling water storage tank (IRWST) compartment and may lead to hydrogen denotation. However, hydrogen can be well dispersed in the loop compartment with intentional open of ADS stage 4 valves to RCS depressurization. Therefore, suggestions are proposed for SAMG: implement RCS depressurization strategy with stage 4 ADS instead of ADS stage 1–3.

On the Analysis and Evaluation of Direct Containment Heating with the Multidimensional Multiphase Flow Code MC3D

In the course of a postulated severe accident in an NPP, Direct Containment Heating (DCH) may occur after an eventual failure of the vessel. DCH is related to dynamical, thermal, and chemical phenomena involved by the eventual fine fragmentation and dispersal of the corium melt out of the vessel pit. It may threaten the integrity of the containment by pressurization of its atmosphere. Several simplified modellings have been proposed in the past but they require a very strong fitting which renders any extrapolation regarding geometry, material, and scales rather doubtful. With the development of multidimensional multiphase flow computer codes, it is now possible to investigate the phenomenon numerically with more details. We present an analysis of the potential of the MC3D code to support the analysis of this phenomenon, restricting our discussion to the dynamical processes. The analysis is applied to the case of French 1300 MWe PWR reactors for which we derive a correlation for the corium dispersal rate for application in a Probabilistic Safety Analysis (PSA) level 2 study.

Preliminary Analysis of Effect of the Intentional Depressurization on Fission Product Behavior During TMLB’ Severe Accident

It has been identified that the pressure in the reactor coolant system (RCS) remains high in some severe accident sequences at the time of reactor vessel failure, with the risk of causing direct containment heating (DCH). Intentional depressurization is an effective accident management strategy to prevent DCH or to mitigate its effects. Fission product behavior is affected by intentional depressurization, especially for inert gas and volatile fission product. Because the pressurizer power-operated relief valves (PORVs) are latched open, fission product will transport into the containment directly. This may cause larger radiological consequences in containment before reactor vessel failure. Four cases are selected, including the TMLB’ base case and opening one, two and three pressurizer PORVs. The results show that inert gas transports into containment more quickly when opening one and two PORVs, but more slowly when opening three PORVs; more volatile fission product deposit in containment and less in reactor coolant system (RCS) for intentional depressurization cases. When opening one PORV, the phenomena of revaporization is strong in the RCS.

Onset of Gas Blowthrough During Melt Expulsion From a Hole in a RPV: Numerical Analysis Using the SIMMER Code

In the case of a severe accident in a nuclear power plant there is a residual risk, that the Reactor Pressure Vessel (RPV) does not withstand the thermal attack of the molten core material, of which the temperature can be about 3000 K. For the analysis of the processes governing melt dispersal and heating up of the containment atmosphere of a nuclear power plant in the case of such an event, it is important to know the time of the onset of gas blowthrough during the melt expulsion through the hole in the bottom of the RPV. In the test facility DISCO-C (Dispersion of Simulant Corium-Cold) at the FZK /6/, experiments were performed to furnish data for modeling Direct Containment Heating (DCH) processes in computer codes that will be used to extrapolate these results to the reactor case. DISCO-C models the RPV, the Reactor Coolant System (RCS), cavity and the annular subcompartments of a large European reactor in a scale 1:18. The liquid type, the initial liquid mass, the type of the driving gas and the size of the hole were varied in these experiments. We present results for the onset of the gas blowthrough that were reached by numerical analysis with the Multiphase-Code SIMMER. We compare the results with the experimental results from the DISCO-C experiments and with analytical correlations, given by other authors.

Direct Containment Heating: A Solved Risk Issue

In case of a High Pressure Melt Ejection (HPME) heated gas and corium may be expelled from the bottom head of a reactor vessel reaching the containment atmosphere, leading to a Direct Containment Heating (DCH). In addition, released gases might burn (e.g. hydrogen) causing a high load of the reactor containment building. Corium dispersal phenomena also strongly affect consequences of Molten Core-Concrete Interaction (MCCI) since the corium mass, which remains within the cavity, may remarkably diminish. Several computer codes are able to simulate the response of nuclear plants during hypothetical severe accidents: MELCOR, MAAP, ESCADRE and ASTEC have the capability to describe corium slump into the reactor cavity, vessel gases blow-down and possible corium entrainment. In this paper the various steps of model development, validation, plant-specific applications, etc., are described in the attempt of establishing a risk-oriented methodology with the target of solving this particular risk-issue. ENEA mature expertise in level-2 PSA analyses shows that the DCH phenomenology can be considered a solved risk issue. The applied methodology gives also hints and/or guidelines for solving similar risk issues in current PSA (level 2) analysis.