RELAP5 code posttest analyses were performed on ROSA/LSTF experiments that simulated PWR 0.2% vessel bottom small-break loss-of-coolant accidents with different accident management (AM) measures under assumptions of noncondensable gas inflow and total failure of high-pressure injection system. Depressurization of and auxiliary feedwater (AFW) injection into the secondary-side of both steam generators (SGs) as the AM measures were taken 10？min after a safety injection signal. The primary depressurization rate of 55？K/h caused rather slow primary depressurization being obstructed by the gas accumulation in the SG U-tubes after the completion of accumulator coolant injection. Core temperature excursion thus took place by core boil-off before the actuation of low-pressure injection (LPI) system. The fast primary depressurization by fully opening the relief valves in both SGs with continuous AFW injection led to long-term core cooling by the LPI actuation even under the gas accumulation in the SG U-tubes. The code indicated remaining problems in the predictions of break flow rate during two-phase flow discharge period and primary pressure after the gas inflow. Influences of the primary depressurization rate with continuous AFW injection onto the long-term core cooling were clarified by the sensitivity analyses. 1. Introduction A small amount of residue including boron around the circumference of two instrument-tube penetration nozzles of pressure vessel lower-head was found at the South Texas Project Unit-1 of pressurized water reactor (PWR) in the US in 2003 [1]. This raised a safety issue concerning vessel structural integrity. Ejection of multiple instrument-tubes due to circumferential cracks at the vessel lower-head nozzles would have caused a small-break loss-of-coolant accident (SBLOCA) at pressure vessel bottom. Liquid-phase break is characterized by fast loss of primary coolant inventory with low primary depressurization. Fletcher and Bolander [2] showed core cooling would continue if break at the vessel lower-head is extremely small through RELAP5 code analyses on instrument-tube ruptures in the Zion-1 PWR in the US. A PWR vessel bottom break simulation test was conducted in 1986 [3] among a series of SBLOCA tests with different location breaks [4] using the large scale test facility (LSTF) [5] under full-pressure conditions in the rig of safety assessment (ROSA) program at Japan Atomic Energy Agency. The LSTF simulates a Westinghouse-type four-loop 3423？MW (thermal) PWR by a full-height and 1/48 volumetrically-scaled two-loop system. Common
References
[1]
USNRC, “Leakage from Reactor Pressure Vessel Lower Head Penetrations and Reactor Coolant Pressure Boundary Integrity,” Bulletin 2003-02, OMB Control 3150-0012, 2003.
[2]
C. D. Fletcher and M. A. Bolander, “Analysis of instrument tube ruptures in the Zion-1 pressurized water reactor,” Nuclear Technology, vol. 81, no. 1, pp. 52–62, 1988.
[3]
Y. Kukita, H. Asaka, H. Nakamura, and K. Tasaka, “Pressurized water reactor core instrument tube ruptures: experimental simulation at the ROSA-IV LSTF,” Nuclear Engineering and Design, vol. 120, no. 2-3, pp. 249–257, 1990.
[4]
Y. Kukita, K. Tasaka, H. Asaka, T. Yonomoto, and H. Kumamaru, “The effects of break location on PWR small break LOCA: experimental study at the ROSA-IV LSTF,” Nuclear Engineering and Design, vol. 122, no. 1–3, pp. 255–262, 1990.
[5]
The ROSA-V Group, “ROSA-V large scale test facility (LSTF) system description for the third and fourth simulated fuel assemblies,” JAERI-Tech 2003-037, Japan Atomic Energy Research Institute, Ibaraki, Japan, 2003.
[6]
H. Asaka and Y. Kukita, “Intentional depressurization of steam generator secondary side during a PWR small-break loss-of-coolant accident,” Journal of Nuclear Science and Technology, vol. 32, no. 2, pp. 101–110, 1995.
[7]
G. G. Loomis, “Summary of the Semiscale program (1965–1986),” USNRC Report NUREG/CR-4945, EGG-2509, 1987.
[8]
C. Addabbo and A. Annunziato, “The LOBI integral system test facility experimental programme,” Science and Technology of Nuclear Installations, vol. 2012, Article ID 238019, 16 pages, 2012.
[9]
K. Umminger, L. Dennhardt, S. Schollenberger, and B. Schoen, “Integral test facility PKL: experimental PWR accident investigation,” Science and Technology of Nuclear Installations, vol. 2012, Article ID 891056, 16 pages, 2012.
[10]
H. Asaka, Y. Anoda, Y. Kukita, and I. Ohtsu, “Secondary-side depressurization during PWR cold-leg small break LOCAs based on ROSA-V/LSTF experiments and analyses,” Journal of Nuclear Science and Technology, vol. 35, no. 12, pp. 905–915, 1998.
[11]
T. Takeda, A. Ohwada, and H. Nakamura, “Measurement of non-condensable gas in a PWR small-break LOCA simulation test with LSTF for OECD/NEA ROSA Project and RELAP5 post-test analysis,” Experimental Thermal and Fluid Science, vol. 51, pp. 112–121, 2013.
[12]
H. Nakamura, T. Watanabe, T. Takeda et al., “RELAP5/MOD3 code verification through PWR pressure vessel small break loca tests in OECD/NEA ROSA project,” in Proceedings of the 16th International Conference on Nuclear Engineering (ICONE-16 '08), pp. 659–668, Orlando, Fla, USA, May 2008.
[13]
H. Nakamura, T. Watanabe, T. Takeda, Y. Maruyama, and M. Suzuki, “Overview of recent efforts through ROSA/LSTF experiments,” Nuclear Engineering and Technology, vol. 41, no. 6, pp. 753–764, 2009.
[14]
H. Asaka and Y. Anoda, “Study of accident management measures for prevention of severe core damage in ROSA-V program,” Japanese Journal of Multiphase Flow, vol. 17, no. 2, pp. 116–125, 2003 (Japanese).
[15]
M. Suzuki, T. Takeda, H. Asaka, and H. Nakamura, “Experimental study on secondary depressurization action for PWR vessel bottom small break LOCA with HPI failure and gas inflow (ROSA-V/LSTF Test SB-PV-03),” JAERI-Research 2005-014, Japan Atomic Energy Research Institute, Ibaraki, Japan, 2005.
[16]
M. Suzuki, T. Takeda, H. Asaka, and H. Nakamura, “An experimental study on effective depressurization actions for PWR vessel bottom break LOCA with HPI failure and gas inflow (ROSA-V/LSTF Test SB-PV-04),” JAEA-Research 2006-018, Japan Atomic Energy Agency, Ibaraki, Japan, 2006.
[17]
M. Suzuki, T. Takeda, H. Asaka, and H. Nakamura, “A study on timing of rapid depressurization action during PWR vessel bottom break LOCA with HPI failure and AIS-gas inflow (ROSA-V/LSTF test SB-PV-06),” JAEA-Research-2007-037, Japan Atomic Energy Agency, Ibaraki, Japan, 2007.
[18]
RELAP5 Code Development Team, “RELAP5/MOD3 code manual,” Tech. Rep. NUREG/CR-5535 (INEL-95/0174), Idaho National Engineering Laboratory, 1995.
[19]
M. Suzuki, T. Takeda, H. Asaka, and H. Nakamura, “Effects of secondary depressurization on core cooling in PWR vessel bottom small break LOCA experiments with HPI failure and gas inflow,” Journal of Nuclear Science and Technology, vol. 43, no. 1, pp. 55–64, 2006.
[20]
T. Takeda, T. Watanabe, and H. Nakamura, “Effectiveness of AM measures for long-term core cooling during PWR vessel bottom small-break LOCA based on RELAP5 analyses of ROSA/LSTF experiment,” in Proceedings of the 9th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-9 '12), Kaohsiung, Taiwan, September 2012.
[21]
N. Zuber, “Problems in modeling small break LOCA,” USNRC Report NUREG-0724, 1980.
[22]
H. Kumamaru and K. Tasaka, “Recalculation of simulated post-scram core power decay curve for use in ROSA-IV/LSTF experiments on PWR small-break LOCAs and transients,” Tech. Rep. JAERI-M 90-142, Japan Atomic Energy Research Institute, Ibaraki, Japan, 1990.
[23]
H. Asaka, Y. Kukita, T. Yonomoto, Y. Koizumi, and K. Tasaka, “Results of 0.5% cold-leg small-break LOCA experiments at ROSA-IV/LSTF: effect of break orientation,” Experimental Thermal and Fluid Science, vol. 3, no. 6, pp. 588–596, 1990.
[24]
H. K. Fauske, “The discharge of saturated water through tubes,” Chemical Engineering and Progress Symposium Series, vol. 61, no. 59, pp. 210–216, 1965.
[25]
K. H. Ardron and R. A. Furness, “A study of the critical flow models used in reactor blowdown analysis,” Nuclear Engineering and Design, vol. 39, no. 2-3, pp. 257–266, 1976.
[26]
D. W. Sallet, “Thermal hydraulics of valves for nuclear applications,” Nuclear Science and Engineering, vol. 88, no. 3, pp. 220–244, 1984.
[27]
Susyadi and T. Yonomoto, “Analysis on non uniform flow in steam generator during steady state natural circulation cooling,” JAERI- Research 2005-011, Japan Atomic Energy Research Institute, Ibaraki, Japan, 2005.
[28]
T. Takeda, H. Asaka, and H. Nakamura, “RELAP5 analysis of OECD/NEA ROSA project experiment simulating a PWR loss-of-feedwater transient with high-power natural circulation,” Science and Technology of Nuclear Installations, vol. 2012, Article ID 957285, 15 pages, 2012.
[29]
M. M. Shah, “A general correlation for heat transfer during film condensation inside pipes,” International Journal of Heat and Mass Transfer, vol. 22, no. 4, pp. 547–556, 1979.
[30]
W. A. Nusselt, “The surface condensation of water vapor,” Zeitschrift des Vereines Deutscher Ingenieure, vol. 60, pp. 541–546, 1916.
[31]
K. M. Vierow and V. E Schrock, “Condensation in a natural circulation loop with non-condensable gases part I—heat transfer,” in Proceedings of International Conference on Multiphase Flows, Ibaraki, Japan, September 1991.
[32]
G. E. Wilson and B. E. Boyack, “The role of the PIRT process in experiments, code development and code applications associated with reactor safety analysis,” Nuclear Engineering and Design, vol. 186, no. 1-2, pp. 23–37, 1998.
[33]
B. E. Boyack and L. W. Ward, “Validation test matrix for the consolidated TRAC (TRAC-M) code,” in Proceedings of the International Meeting on Best Estimate Methods in Nuclear Installation Safety Analysis (BE '00), Washington, Wash, USA, November 2000.
[34]
A. Yamaguchi, S. Mizokami, Y. Kudo, and A. Hotta, “Uncertainty and conservatism in safety evaluations based on a BEPU approach,” in Proceedings of the 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13 '09), Ishikawa, Japan, September-October 2009.
