A peer-reviewed journal published by K. N. Toosi University of Technology

Document Type : Research Article


1 Young Researchers and Elites club, Science and Research Branch, Islamic Azad University, Tehran, Iran

2 Physics Department, Imam Hossein University, Tehran, Iran


‎The analysis deals with the assessment of best estimate code RELAP5/SCDAP mod3.4 in the simulation of double-ended loss coolant accident as a LBOCA‎, ‎4 in break as a SBLOCA an SBO accident with considering except accumulator water where no core cooling water systems are available‎. ‎The reference plant is SURRY nuclear power plant as a Westinghouse three-loop nuclear power plant‎. ‎In order to mitigation accident‎, ‎the in-vessel retention strategy was investigated for the prevention of lower plenum failure‎. ‎It has been concluded that during the SBLOCA‎, ‎LBLOCA conditions bottom of active fuel is uncovered at 6340 s and 2160 s‎, ‎respectively‎. It occurred for two times at 11650 s and 15608 s in SBO‎. ‎At 6792 s and 57002 s in the LBLOCA and SBO due to reaching melting point and in the SBLOCA at 15215 s due to lower plenum creep rupture‎, ‎failure of the reactor pressure vessel occurred‎. ‎The results show that hydrogen production in the SBO is more than the other two cases‎. ‎For the prevention of the lower plenum failure‎, ‎the in-vessel molten material retention strategy is investigated as a passive system‎. ‎The results show that lower plenum heat flux can be kept below the critical heat flux and its integrity is preserved in two cases of this analysis‎.


  • The RELAP5/SCDAP is modeled for three-loop PWR NPP plant.
  • SBLOCA, LBLOCA, and SBO are analyzed as three significant severe accidents.
  • The main parameters such as creep rupture of the surge line of PRZ and lower plenum failure time are reported.
  • The IVR strategy is investigated for SBO, and small and large LOCA.


Allison, C. and Beers, G. (1984). Scdap/mod1/v0: A computer code for the analysis of LWR vessel behavior during severe accident transients. Technical report, IS-SAAM-83-002.
Chan, S. L. (2006). Assessment of S/R5/MOD3. 3 models of stratified molten pool and debris-to vessel contact resistance by TMI-2 lower-head creep rupture analyses. Nuclear technology, 156(2):191–212.
Commission, N. R. et al. (1990). Severe accident risks: An assessment for five us nuclear power plants: Appendices A, B, and C. Technical report, Nuclear Regulatory Commission.
Gu, Z. (2018). History review of nuclear reactor safety. Annals of Nuclear Energy, 120:682–690.
Hessheimer, M. F. and Dameron, R. (2006). Containment Integrity Research at Sandia National Laboratories. Division of Fuel, Engineering & Radiological Research, Office of Nuclear.
Jahanfarnia, G. and Salehi, M. (2016). The in-vessel melt retention strategy using the RELAP5/SCDAP code for VVER-1000 reactor. The 8th international conference on sustainable development through nuclear research and education.
Jian, D. and Xuewu, C. (2007). Analysis of hot leg natural circulation under station blackout severe accident. Nuclear Science and Techniques, 18(2):123–128.
Jiang, N., Cong, T., and Peng, M. (2019). Margin evaluation of in-vessel melt retention for small IPWR. Progress in Nuclear Energy, 110:224–235.
Kalchev, B., Dimov, D., Tusheva, P., et al. (2005). Comparative severe accident analysis of WWER 1000/B 320 LOCA DN100 computed by computer codes ASTEC V1. 1 and SCDAP/RELAP5.
Larson, F. R. (1952). A time-temperature relationship for rupture and creep stresses. Trans. ASME, 74:765–775.
Lemmon, E. (1980). Couple/fluid: A two-dimensional finite element thermal conduction and advection code. Technical report, EGG-ISD-SCD-80-1.
Ma, W., Yuan, Y., and Sehgal, B. R. (2016). In-vessel melt retention of pressurized water reactors: historical review and future research needs. Engineering, 2(1):103–111.
Manson, S. and Haferd, A. M. (1953). A linear time temperature relation for extrapolation of creep and stressrupture data.
Mladin, M., Dupleac, D., Prisecaru, I., and Mladin, D. (2010). Adapting and applying SCDAP/RELAP5 to CANDU in vessel retention studies. Annals of Nuclear Energy, 37(6):845–852.
NRC (2018). Surry power station, Unit 1. Technical report, https://www.nrc.gov/info-finder/reactors/sur1.html.
Nunez-Carrera, A., Camargo-Camargo, R., Espinosa-Paredes, G., et al. (2012). Simulation of the lower head boiling water reactor vessel in a severe accident. Science and Technology of Nuclear Installations, 2012.
Park, R.-J., Ha, K.-S., Kim, H.-Y., et al. (2013). Detailed evaluation of safety injection tank effects on in-vessel severe accident progression in a small break LOCA without safety injection. Annals of Nuclear Energy, 58:54–59.
Petruzzi, A. and D’Auria, F. (2008). Thermal-hydraulic system codes in nuclear reactor safety and qualification procedures. Science and Technology of Nuclear Installations, 2008.
Pilch, M., Allen, M., Bergeron, K., et al. (1995). The probability of containment failure by direct containment heating in surry. Technical report, Nuclear Regulatory Commission, Washington, DC (United States).
RELAP5 (1995). Relap5/m0d3 code manual, the RELAP5 development team. Technical report, RELAP5/M0D3 Code Manual. NUREG/CR- 5535, INEL-95j0174.
Salehi, M. and Jahanfarnia, G. (2016). Small break LOCA analysis without emergency core cooling systems using the RELAP5/SCDAP code in VVER-1000 reactor. Annals of Nuclear Energy, 87:299–307.
Salehi, M. and Shayesteh, M. (2017). The accumulator effects on in-vessel severe accident progression of a three loop pwr nuclear power plant in a SBLOCA without safety injection system. Annals of Nuclear Energy, 107:1–16.
Seghal, B. R. (2012). Nuclear safety in light water reactors: Severe accident phenomenology, ed.