Application of ASTEC, MELCOR, and MAAP Computer Codes for Thermal Hydraulic Analysis of a PWR Containment Equipped with the PCFV and PAR Systems

Abstract

The integrity of the containment will be challenged during a severe accident due to pressurization caused by the accumulation of steam and other gases and possible ignition of hydrogen and carbon monoxide. Installation of a passive filtered venting system and passive autocatalytic recombiners allows control of the pressure, radioactive releases, and concentration of flammable gases. Thermal hydraulic analysis of the containment equipped with dedicated passive safety systems after a hypothetical station blackout event is performed for a two-loop pressurized water reactor NPP with three integral severe accident codes: ASTEC, MELCOR, and MAAP. MELCOR and MAAP are two major US codes for severe accident analyses, and the ASTEC code is the European code, joint property of Institut de Radioprotection et de Sûreté Nucléaire (IRSN, France) and Gesellschaft für Anlagen und Reaktorsicherheit (GRS, Germany). Codes' overall characteristics, physics models, and the analysis results are compared herein. Despite considerable differences between the codes' modelling features, the general trends of the NPP behaviour are found to be similar, although discrepancies related to simulation of the processes in the containment cavity are also observed and discussed in the paper.

Mathematical model of the NPP Krško PCFV system for the RELAP5 computer code

“…induce the reaction between oxygen and hydrogen until all oxygen has reacted. An integral analysis of the whole NEK containment was performed with three system codes: ASTEC, MELCOR and MAAP with emphasis on the containment thermal hydraulic behaviour when the PCFV system is in the operation [5]. Figure 12 shows the gas mixture composition throughout the transient and it is visible that conditions for possible hydrogen ignitions were not achieved in any time period during the calculation.…”

“…The steady state calculation was performed to verify the nodalization by comparing the results with the design operating parameters (pressure, temperature, gas velocity). The transient calculation, in the absence of the measured data, was run with the boundary conditions extracted from the calculations performed with the code MELCOR in scope of the supporting calculations for the plant reported in relevant documents [3] and papers [4], [5]. The knowledge gained in this analysis will be later used to develop a model for the integral plant simulation with explicit models of all power plant systems, including the model of the PCFV system.…”

3D containment modeling of PWR-KWU Trillo NPP with the GOTHIC code

“…Besides these assumptions, LP models proved to be successful in calculating average values for pressure and temperature during a transient, as has been shown in several benchmarks and qualification exercises, (OECD/NEA, 2012). Different codes such as FATHOMS, MELCOR, or MAAP appeared over time, all of them using the LP framework to model the containment and performing accurately in calculating the average pressure and temperature in containment during a transient, (Chiang et al, 2017;Fernandez-Moguel and Birchley, 2015;Šadek et al, 2017). Recent codes such as ATHROC, (Chen et al, 2018) or CAP (Hong et al, 2015) still follow this same trend in the LP approach.…”

Verification of the Efficacy of Passive Autocatalytic Recombiners in a Typical Pressurized Water Reactor under a Station Blackout Condition

“…Generally, for such class of problems, the thermal hydraulic response of the plant can be analyzed using lumped parameter (LP) codes or computational fluid dynamics (CFD) codes. A number of codes have been specially developed by the nuclear industry for multidimensional containment analyses, such as GOTHIC [1,2], GASFLOW [3], and TONUS [4] and notably the severe accident analysis codes, for example, RELAP/SCDAP [5,6], MAAP [7,8], MELCOR [9][10][11], and ASTEC [12][13][14]. Other general-purpose commercially available CFD codes have been also explored for nuclear safety simulations including containment analyses [15], for example, by Heitsch et al [16], Prabhudhardwadkar et al [17], and Filippov et al [18].…”