United States Nuclear Regulatory Commission - Protecting People and the Environment

Assessment of RELAP5/MOD3.3Beta Code for the LOFT Experiment L9-1/L3-3 (NUREG/IA-0228)

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Publication Information

Manuscript Completed: November 2009
Date Published: May 2010

Prepared by:
Kyu Bok Lee, Heedo Lee, Kab Seok Ko /KOPEC
Young Seok Bang, In Goo Kim, Seung Hoon Ahn /KINS

Korea Power Engineering Company Inc. (KOPEC)
257 Yonggudaero, Giheung-gu, Yongin-si
Gyeonggi-do, Korea, 446-713

Korea Institute of Nuclear Safety (KINS)
19 Gusung-Dong, Yusong-Gu
Daejeon, 305-338, Korea

Antony Calvo, NRC Project Manager

Prepared for:
Division of Systems Analysis
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

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Abstract

The purpose of this report is to assess the capability of the RELAP5/MOD3.3Beta computer code to simulate thermal-hydraulic behavior associated with the LOFT Experiment L9-1/L3-3. The Experiment L9-1/L3-3 was a simulation of the total loss-of-feedwater accident and its recovery modes. Experiment L9-1 simulated a loss-of-feedwater accident with delayed reactor scram and no auxiliary feedwater injection. The loss-of-feedwater accident led to a loss-of-coolant accident through the PORV cycling operation.

Generally, the RELAP5/MOD3.3Beta calculation results were in good agreement with the L9-1 experimental data. The discrepancies between the calculation and the experiment were also identified in the temperature behaviors of the SG secondary side after dryout of the SG. However, these discrepancies were not considered to be significant, since the SG was in a de-coupled state from the PCS after the dryout of the SG secondary side.

Experiment L3-3 simulated two recovery modes from the loss-of-feedwater accident L9-1 without the aid of the emergency core coolant system. The first recovery mode consisted of turning off the primary coolant pumps and latching open the PORV to depressurize the primary system. The second mode consisted of refilling the SG to restore the secondary heat sink and removing decay heat through the feed-and-bleed operation using the secondary side of the SG.

The general trends observed in Experiment L3-3 were similar to those of the RELAP5/MOD3.3Beta code calculations. In addition, some differences between the code calculations and the L3-3 experimental data were observed. The code predicted excessive swelling of the PCS fluid during the first recovery mode, and predicted excessive cooling during the second recovery mode. The differences during the first recovery mode were due to the code's under-estimation of total discharged energy through the PORV. For the discrepancies during the second recovery mode, the code modeling deficiency of the SG secondary side was presumed to be one of the reasons. However, it is concluded that further investigation into the deviation sources are needed to enhance the code's predictability, since the specific reasons for the deviation were not clearly identified in this assessment.

Sensitivity studies show that several parameters have significant effect on the predicted thermal-hydraulic behaviors. These parameters include the pressurizer spray system loss coefficient, nodalization of the SG secondary side, SG U-tube heat transfer area, heat transfer coefficient from the LOFT main components to the environment, and the PORV discharge coefficient. The first three parameters are significant during the short-term transient phase, while the last two are significant during the long-term transient phase. Therefore, the five parameters listed above should be carefully modeled to obtain appropriate calculation results for transient types such as the total-loss-of feedwater accident and its recovery modes.

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