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.

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.

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. 

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.