Modeling condensation steam quenching effects in an MSIV closure transient using RETRAN-02-MOD002 with and without nonequilibrium pressurizer model option and TRAC-BF1

Journal Article · Sun Aug 01 04:00:00 EDT 1993 · Nuclear Technology; (United States)

OSTI ID:6150653

Boyer, B D; Hartzell, J W; Lider, S; Robinson, G E; Baratta, A J ^[1]; Roscioli, A J ^[2]

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1. The Pennsylvania State Univ., University Park, PA (United States). Dept. of Nuclear Engineering.
2. Pennsylvania Power and Light Co., Allentown, PA (United States)

The effects of condensation steam quenching in modeling two-phase flow phenomena during a nuclear reactor transient are studied. The RETRAN-02-MOD002 code, with three field equations and a nonequilibrium pressurizer model option, and the TRAC-BF1 code, with six field equations and a nonequilibrium pressurizer model option, and the TRAC-BF1 code, with six field equations, predicted plant response to a boiling water reactor plant test of a main steam isolation valve closure without safety relief valve opening. The basic RETRAN-02-MOD002 field equations cannot model steam quenching by condensation. However, by activating the nonequilibrium modeling option of the basic RETRAN-02-MOD002 code and by inputting appropriate interfacial heat transfer coefficients, steam quenching by condensation was calculated. This approach gave results closer to those obtained with the test data. The two TRAC-BF1 models used two different methods of tracking water level to approximate the condensation quenching effect. Because the void fraction changes too gradually, the calculation without the TRAC two-phase water level tracking option overquenched the pressure and filled the vessel with too much water. However, because the void fraction changes virtually instantaneously (as it does in the plant), the TRAC two-phase water level tracking option's prediction of the quenching of the pressure was 50% closer to the data than was any RETRAN-02-MOD002 calculation, and it followed the water level almost as well as the RETRAN-02-MOD002 best-estimate case. Both codes overpredicted the pressure spike.

• OSTI ID: 6150653
• Journal Information: Nuclear Technology; (United States), Journal Name: Nuclear Technology; (United States) Vol. 103:2; ISSN 0029-5450; ISSN NUTYBB
• Country of Publication: United States
• Language: English

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