Comparative Study of Loss-of-Coolant Accident Using MAAP4.03 and RELAP5/MOD3.2.2

Abstract

Analysis was performed for a large-break loss-of-coolant accident (LOCA) in the APR1400 (Advanced Power Reactor 1400 MWe) with the thermal-hydraulic analysis code RELAP5/ MOD3.2.2 and the severe accident analysis code MAAP4.03. The two codes predicted different sequences for essentially the same initiating condition. As for the break flow and the emergency core cooling (ECC) flow rates, MAAP4.03 predicted considerably higher values in the initial stage than RELAP5/ MOD3.2.2. It was considered that the differing break flow and ECC flow rates would cause the LOCA sequences to deviate from one another between the two codes. Hence, the break flow model in MAAP4.03 was modified with partly implementing the two-phase homogeneous critical flow model and adopting a correction term. The ECC flow model in MAAP4.03 was also varied by changing the hardwired friction factor through the sensitivity study. The modified break flow and ECC flow models yielded more consistent calculational results between RELAP5/MOD3.2.2 and MAAP4.03. It was, however, found that the resultant effect is rather limited unless more mechanistic treatments are done for the primary system in MAAP4.03.

Extended Station Blackout Coping Capabilities of APR1400

“…However, it produces reasonable prediction capabilities comparable to MELCOR and SCDAP [16]. MAAP4 is widely used in Korea in severe accident analysis and level 2 PSA for APR1400 design [17,18]. A noding diagram for APR1400 is shown in Figure 3.…”

RELAP5/MOD3.3 Best Estimate Analyses for Human Reliability Analysis

“…However, information from these is often too conservative to perform a realistic PSA for a riskinformed application [3]. In the last years a few comparative studies were performed to justify the use of MAAP4 for the PSA Level 1 analysis of advanced reactors [4,5]. In the comparison between MAAP4.07 and S-RELAP5 for U.S. EPR reactor [4] MAAP4 has demonstrated that it is a rather good simulator of nuclear plant transient trends.…”

“…Also, there are certain nuclear plant scenarios for which MAAP4 is clearly not applicable, such as early transient of large-break LOCA (at break sizes beyond the area of the largest attached pipe). In the study for APR1400 (Advanced Power Reactor) [5] comparison between MAAP4.03 and RELAP5/MOD3.2.2 was done for large break loss-of-coolant accident (LOCA). It was concluded that for a more mechanistic simulation of the initial stage of the LOCA using MAAP4.03, more detailed calculations of the primary system are required.…”