Abstract
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.
Highlights
After the nuclear events unfolded in Fukushima, the interest in severe accidents in nuclear power plants has increased greatly [1,2,3], both in severe accident management and in severe accident phenomena research
Based on the fact that the primary pressure drops below the secondary pressure relatively quickly, a 2 inch break loss-of-coolant accident (LOCA) will probably be located at the high end of the SBLOCA break range; i.e., it behaves similar to a MBLOCA as discussed above
The lower plenum debris bed response predicted by the two code, including vessel failure, should be compared in qualitative and relative terms considering the timing and quantity of corium relocated from the core predicted by each code
Summary
After the nuclear events unfolded in Fukushima, the interest in severe accidents in nuclear power plants has increased greatly [1,2,3], both in severe accident management and in severe accident phenomena research. Because of high level of complexity and interplay between different phenomena, state-of-the-art integrated severe accident codes such as ASTEC [4], MAAP [5], and MELCOR [6] are often used to study and design accident management strategies [7]. These simulation codes, which capture theoretical and experimental knowledge produced in the last four decades, predict progression of a postulated accident leading to fuel melting, core relocation to lower plenum, and vessel failure. Conclusions will be drawn from the following three perspectives:
Sign-in/Register to view highlights & summary Talk to us
Join us for a 30 min session where you can share your feedback and ask us any queries you have
Schedule a callDisclaimer: All third-party content on this website/platform is and will remain the property of their respective owners and is provided on "as is" basis without any warranties, express or implied. Use of third-party content does not indicate any affiliation, sponsorship with or endorsement by them. Any references to third-party content is to identify the corresponding services and shall be considered fair use under The CopyrightLaw.
