ASTEC–MAAP Comparison of a 2 Inch Cold Leg LOCA until RPV Failure

A 2 inch, cold-leg loss-of-coolant accident (LOCA) in a 900 MWe generic Western PWR was simulated using ASTEC 2.1.1 and MAAP 5.02. The progression of the accident predicted by the two codes up to the time of vessel failure is compared. It includes the primary system depressurization, accumulator discharge, core heat-up, hydrogen generation, core relocation to lower plenum, and lower head breach. The purpose of the code comparison exercise is to identify modelling differences between the two codes and the user choices affecting the results. The two codes predict similar primary system depressurization behaviour until the accumulation injection, confirming similar break flow and primary system thermal-hydraulic response calculations between the two codes. The choice of the accumulator gas expansion model, either isentropic or isothermal, affects the rate and total amount of coolant injected and thereby determines whether the core is quenched or overheated and attains a noncoolable geometry during reflooding. A sensitivity case was additionally simulated by each code to allow comparisons to be made with either accumulator gas expansion models. The two codes predict similar amount of in-vessel hydrogen generated and core quench status for a given accumulator gas expansion model. ASTEC predicts much larger initial core relocation to lower plenum leading to an earlier vessel failure time. MAAP predicts more gradual core relocation to lower plenum, prolonging the lower plenum debris bed heat-up and time to vessel failure. Beside the effect of the code user in conducting severe accident simulations, some discrepancies are found in the modelling approaches in each code. The biggest differences are found in the in-vessel melt progression and relocation into the lower plenum, which deserve further research to reduce the uncertainties.

[1]  F. Mascari,et al.  Scaling issues for the experimental characterization of reactor coolant system in integral test facilities and role of system code as extrapolation tool , 2015 .

[2]  Fabio Giannetti,et al.  Analysis of unmitigated large break loss of coolant accidents using MELCOR code , 2017 .

[3]  Hwan-Yeol Kim,et al.  Detailed evaluation of safety injection tank effects on in-vessel severe accident progression in a small break LOCA without safety injection , 2013 .

[4]  Mohsen Amini Salehi,et al.  The accumulator effects on in-vessel severe accident progression of a three loop PWR nuclear power plant in a SBLOCA without safety injection system , 2017 .

[5]  Longze Li,et al.  MAAP5 Simulation of the PWR Severe Accident Induced by Pressurizer Safety Valve Stuck-Open Accident , 2014 .

[6]  César Queral,et al.  Accident management actions in an upper-head small-break loss-of-coolant accident injection failed , 2011 .

[7]  Jun Sugimoto,et al.  Severe Accident Analysis in Nuclear Power Plants , 2012 .

[8]  S. Güntay,et al.  Conclusions on severe accident research priorities , 2014 .

[9]  Francesco Saverio D'Auria,et al.  Scaling in nuclear reactor system thermal-hydraulics , 2010 .

[10]  F. D'Auria,et al.  User effects on the thermal-hydraulic transient system code calculations , 1993 .

[11]  Luis E. Herranz,et al.  CESAM – Code for European severe accident management, EURATOM project on ASTEC improvement , 2018, Annals of Nuclear Energy.

[12]  César Queral,et al.  Accident Management Actions in an Upper-Head Small-Break Loss-of-Coolant Accident with High-Pressure Safety Injection Failed , 2011 .

[13]  N. Aksan International Standard Problems and Small Break Loss-of-Coolant Accident (SBLOCA) , 2008 .

[14]  Longze Li,et al.  MAAP5 simulation of the PWR severe accident induced by pressurizer safety valve stuck-open accident , 2014 .

[15]  Francesco Saverio D'Auria,et al.  Thermal hydraulics in water-cooled nuclear reactors , 2017 .

[16]  R.Y. Lee,et al.  Accident source terms for Light-Water Nuclear Power Plants. Final report , 1995 .

[17]  Giacomino Bandini,et al.  Analyses of an unmitigated station blackout transient with ASTEC, MAAP and MELCOR Codes , 2017 .

[18]  L. Carénini,et al.  Modelling issues related to molten pool behaviour in case of In-Vessel Retention strategy , 2018, Annals of Nuclear Energy.

[19]  A. Miassoedov,et al.  Recent severe accident research synthesis of the major outcomes from the SARNET network , 2015 .

[20]  P. Chatelard,et al.  Main modelling features of the ASTEC V2.1 major version , 2016 .

[21]  A. Schaffrath,et al.  ETSON views on R&D priorities for implementation of the 2014 Euratom Directive on safety of nuclear installations , 2016 .