The generic severe accident management guidelines (SAMG) were developed as a response to TMI-2 accident to improve the defense-in-depth (DiD) concept of light water reactors (LWRs). Various SAMGs were developed for plant specific application considering different objectives of severe accident management (SAM) and design features of nuclear reactors. To verify and validate the effectiveness of SAMG and SAM actions for mitigating accident consequences and terminating accident progression, the analytical simulation through best estimate codes were performed extensively to provide quantitative details for the assessment of an SAMG and its actions. The present study is carried out to review the representative works concerned with the assessment of SAMG actions in the pressurized water reactors (PWR) and European VVERs using analytical simulation. The outcomes would be valid to realize the improvement and development of assessment methodology in future studies.

Analytical simulation using best-estimate codes were suggested to be extended for the elaboration and improvement of SAMG in current PWRs. This work is about to perform an assessment work using MELCOR for the effectiveness of safety injection to achieve the PBF strategy in SAM following a BDBA scenario, that is LOCA with concurrent SBO. In the simulations, the safety injection is assumed to be retrieved with the postulated power recovery at different timing during core relocation. The simulation results illustrates that the grace period of preventing vessel failure varies with LOCA break size and locations. The safety injection implemented in grace period is capable of retarding or ceasing the core relocation, sequentially avoiding the massive core relocation into lower plenum, mitigating the hydrogen generation and fission product release from core. Meanwhile, the injection later than grace period would be failed to prevent RPV failure, and it negatively affects hydrogen generation in some scenarios. The results also indicate that the smallest injection capacity of HPSI system in Swedish PWR is sufficient to achieve the effective mitigation.

During severe accidents in a light water reactor, the core melt (corium) may relocate to the lower head and fail the reactor pressure vessel (RPV). The corium will be ejected to the reactor cavity upon the RPV failure and undergo melt coolant interactions (FCI) if the cavity is flooded with water. The FCI process does not only de-termines the characteristics of the resulting debris bed which are important to coolability, but also induces a steam explosion risk which may threaten containment integrity. The present study is concerned with charac-terization of debris bed formed from FCI of metal-rich corium failing into a deep water pool in the reactor cavity. Low melting-point metals Tin and Tin-Bismuth (20 kg) were employed as the simulant materials of metal-rich corium melt. Ten tests were carried out on the DEFOR-M test facility at KTH to investigate the effects of various parameters on debris bed formation, such as melt superheat, coolant subcooling, material. The melt jet fragmentation and fragments movement in the water pool as well as debris deposition on the pool floor were recorded by high-speed cameras. Melt sensors and weight sensors were installed to detect the period of melt jet discharge and the mass of forming debris bed. The porosity of debris bed was obtained through the debris bed volume measured by a three-dimensional laser scanner and the pore volume measured by water absorption. The final configuration of debris bed was also reconstructed through the laser scanner data, and the debris particles were sieved for their size distribution. The experimental results revealed the FCI phenomena and debris bed characteristics including configuration and porosity of debris bed as well as morphology and size distribution of debris particles under different melt superheats, coolant subcooling, materials.

Severe accident management guidelines (SAMGs) have been developed and applied in nuclear power plants (NPPs) to provide the fourth level of defense worldwide since the TMI-2 accident which occurred in 1979. The primary objective of the SAMGs is to protect remaining safety boundaries and to limit release of fission products into the environment by specific severe accident management (SAM) actions with available safety equipment. This paper is concerned with the assessment of the primary side bleed and feed (PBF) actions in the SAMGs of a Swedish pressurized water reactor (PWR). The PBF actions are realized by operations of emergency injection pumps and power-operated relief valves (PORVs), following a severe accident induced by an event of total loss of feed water (TLOFW) and temporary station blackout. In the present study, MELCOR simulations are performed to investigate the effects of the SAM actions on severe accident progression and consequences. Different sets of available equipment and actuation time due to recovery of AC power are simulated. The interest of the MELCOR analysis is focused on how these uncertain parameters of SAM actions affect thermal–hydraulic response, hydrogen generation, core relocation, reactor pressure vessel failure and release of fission products. The results demonstrate the effectiveness of the PBF strategy to prevent RPV failure in the TLOFW accident and the effects of injection timing on accident consequences. The obtained insights are instrumental for validation and improvement of the strategies and actions in the plant specific SAMGs. 

Severe accident management guidelines (SAMGs) play an important role in the hierarchical structure of the defense-in-depth (DiD) principle of reactor safety. Among different methods to verify and validate the effectiveness of SAMG on mitigating severe accident consequences, the approach of numerical simulations using best-estimate computer codes was extensively applied to evaluate the SAMG and SAM actions. 

In addition to a review on the previous works assessing SAMGs through numerical simulations, the present study is intended to examine and inform the effectiveness of SAMG and its actions for a Swedish pressurized water reactor (PWR) through numerical simulations of the MELCOR code. The research work is composed of i) development and qualification of MELCOR model for the PWR chosen; ii) evaluation of SAMG entry condition; and iii) assessment of operator actions in the SAMG (so-called SAM actions) under different accident scenarios. The SAM actions include depressurization (individual action) and primary-side bleed and feed (PBF) actions which are among the most important SAM actions. The risk-important accident scenarios selected in this study are station blackout (SBO), total loss of feed water (TLOFW), loss of coolant accident (LOCA), and their variations. 

The development and qualification of the MELCOR model for the Swedish PWR is conducted through nodal sensitivity studies which provide the impacts of the COR nodalization and CVH nodalization in the MELCOR model on simulation results. The qualified MELCOR model with achievable accuracy and computational cost is then adopted in the evaluation of SAMG and its actions through numerical simulations.

The interests of the numerical simulations for evaluating the SAMG entry condition and SAM actions are focused on the timing of events, accident consequences, negative/positive effects of SAM actions, etc. Based on the evaluation outcomes, the main points are concluded as follows:

-          The setpoint 650oC of the average core exit temperature (CET) is an effective entry condition of SAMGs (i.e., performing transition from EOPs to SAMGs at the onset of core damage), given the representative accident sequences as the main contributors to the core damage frequency (CDF) of the reactor chosen.

-          The PBF strategy is effective to cease the core relocation and prevent the RPV failure in both TLOFW and LOCA scenarios if the PBF actions are operated within respective grace periods which can be determined through the numerical simulations. 

-          The grace periods of PBF actions are not only dependent on the accident scenarios, but also affected by the timing of bleed/feed actions, RCS depressurization rate (opening of PORVs), injection flowrate, and their combinations.

-          The earlier RCS injection in the grace period can mitigate the hydrogen generation and radioactive release from the core, but a later RCS injection beyond the grace period will produce more hydrogen.

-          The RCS injection in the later stage of core degradation may also mitigate the release of fission products from primary circuits to the containment, since the injected water can scrub the aerosols generated from the core.

The entry conditions of severe accident management guidelines (SAMGs) in pressurized water reactors (PWRs) rely on the indication of core exit temperature (CET). Yet, the setpoints for the CET may be different from plant to plant. Most Westinghouse PWR designs adopt the setpoint of CET at 650℃ as the entry condition of the SAMGs, since this setpoint is an effective indicator of core damage in a wide spectrum of accident sequences. Motivated by the interest in the verification & validation of SAMS after the Fukushima accidents, the present study is conducted numerically to verify the effectiveness of the CET setpoint for the transition from emergency operation procedures (EOPs) to SAMGs in a Swedish nuclear power plant. For this purpose, six representative severe accident sequences covering the main contributors to the core damage frequency (CDF) are analyzed using the MELCOR code. Moreover, different CET readings and alternative entry conditions are also investigated. The simulation results show that the average CET = 650 °C is the effective setpoint as the entry condition of SAMGs, i.e., given this setpoint the transition from EOPs to SAMGs will take place slightly before the occurrence of core degradation, which secures the intended mitigation of SAMGs while keeping EOPs active as long as possible. On the other hand, it is too conservative if the maximum CET = 650 °C is used as the setpoint of entry condition of SAMGs, i.e., it will result in an excessive realization of SAMGs over EOPs. The coolant temperature in the primary circuits, the water level in the RPV and the hydrogen concentration in the containment can also be applied as reference indications of core damage states in the accident management. 

The CVH package in the MELCOR code is responsible for modelling the thermal-hydraulic behavior dur -ing severe accident. This work presents a sensitivity study of thermal-hydraulic nodalization in the core region for MELCOR simulations of postulated severe accidents in a pressurized water reactor (PWR). Three nodal schemes are developed with a good agreement of steady-state parameters. Two accident sce-narios: loss of coolant accident (LOCA) and station blackout (SBO) are simulated. The analysis is focused on the effect of the control volumes (CVs) in the CVH nodalization on the simulation results of in-vessel accident progression, including the core degradation, hydrogen generation and fission products release, etc. It is found that compared with the coarse CVH nodalization (1 CV), the radial refinement of CVH nodalization (7 CVs) leads to different impacts on accident progression in the two scenarios: faster core relocation and more hydrogen generation is predicted in the LOCA scenario, but it is opposite in the SBO scenario. The finest nodal scheme (49 CVs) with refinement in both radial and axial direction tends to predict an earlier occurrence of cladding rupture, RPV failure and faster core relocation, as well as more hydrogen generation. The CVH refinement makes little difference on the radioactive release.

This paper presents a sensitivity study of nodal scheme in MELCOR simulation of severe accidents in a pressurized water reactor, with the objective to estimate the nodal effects on some in-vessel and exvessel processes and phenomena, including thermal-hydraulic response, core degradation and relocation, hydrogen generation, vessel failure, containment pressurization and venting, source term. For this purpose, three nodal schemes (i.e., coarse, medium and fine meshes) of the COR package of the MELCOR code are chosen to analyze two severe accident scenarios: small break loss of coolant accident (SBLOCA) and large break loss-of-coolant accident (LBLOCA), both combined with station blackout. The results show that the nodal schemes mainly affect the calculations of heat transfers from the core to coolant and heat structures, relatively affecting the core degradation and relocation to the lower head of the reactor pressure vessel. As for the consequences, the coarse mesh tends to predict slower core relocation progressions and a later failure of RPV lower head. Moreover, more hydrogen generation by cladding oxidation can be observed in the coarse mesh case. The nodal schemes have little impact on the estimation of in-containment source term. Meanwhile, the simulations with fine mesh may also provide more detailed distributions of corium masses and temperatures, as well as heat fluxes, affecting thermal and mechanic behavior of RPV lower head.

In the framework of this research work, the principle focus is to assess the applicability & reliability of the Phase change Effective Convectivity Model (PECM) as a numerical analysis tool to investigate natural convective heat transfer in single and two-fluid density PCM molten pools. The model is applied in ANSYS FLUENT as User Defined Function (UDF) to predict convective melt pool thermal hydraulics in a volumetrically heated PCM (Phase Change Material) melt pool. As a part of this work, PECM is tested first by a benchmark case against CFD to gain confidence in its applicability as an analysis tool. Two commercial PCMs: RT50 and C58, are introduced in a 3D semicircular vessel slice with their thermo-physical properties as input for modelling. The sidewalls made of quartz glass are used for direct visualization of convective heat transfer phenomena. It is ensured that the conditions of nearly constant density of power deposition over the entire volume of the PCM melt pool throughout the series of simulation cases. The values of characteristic numbers ranged within the following limits with different pool height corresponding modified Rayleigh number Ra=1012-1013 and for Prandtl number Pr=5-7. The selected modelling approach is validated against SIGMA experiment with respect to the angular distribution of heat flux that qualify our model to run in the proceeding calculation using PECM. Following benchmark test results of PECM compared with that of conventional enthalpy porosity method embedded in ANSYS FLUENT, PECM is applied in 1-layer and 2-layer PCM configuration to study in details of the influence of different boundary conditions, internal heat sources (QV) and heat transfer fluid (HTF) cooling condition to quantify the thermal loads. Finally, the comparison is made between two PCM configurations in terms of the quantification of the thermal load to justify PECM as an efficient numerical analysis tool for investigating convective heat transfer phenomena during PCM melting.