Summary on the depressurization from supercritical pressure conditions
- Univ. of Wisconsin Madison, 1500 Engineering Dr., Madison, WI 53706 (United States)
- Dept. of Reactor Engineering, Research and Design, Reactor Thermal-Hydraulic Lab., China Inst. of Atomic Energy, P.O.Box 275 59, 102413 Beijing (China)
- JRC/Inst. for Energy and Transport (Netherlands)
- Dept. of Engineering Physics, McMaster Univ., 1280 Main Street, ON (Canada)
- International Atomic Energy Agency, Vienna International Centre, P.O. Box 100, 1400 Vienna (Austria)
When a fluid discharges from a high pressure and temperature system, a 'choking' or critical condition occurs, and the flow rate becomes independent of the downstream pressure. During a postulated loss of coolant accident (LOCA) of a water reactor the break flow will be subject to this condition. An accurate estimation of the critical flow rate is important for the evaluation of the reactor safety, because this flow rate controls the loss of coolant inventory and energy from the system, and thus has a significant effect on the accident consequences[1]. In the design of safety systems for a super critical water reactor (SCWR), postulated LOCA transients are particularly important due to the lower coolant inventory compared to a typical PWR for the same power output. This lower coolant inventory would result in a faster transient response of the SCWR, and hence accurate prediction of the critical discharge is mandatory. Under potential two-phase conditions critical flow is dominated by the vapor content or quality of the vapor, which is closely related with the onset of vaporization and the interfacial interaction between phases [2]. This presents a major challenge for the estimation of the flow rate due to the lack of the knowledge of those processes, especially under the conditions of interest for the SCWR. According to the limited data of supercritical fluids, the critical flows at conditions above the pseudo-critical point seem to be fairly stable and consistent with the subcritical homogeneous equilibrium model (HEM) model predictions, while having a lower flow rate than those in the two-phase region. Thus the major difficulty in the prediction of the depressurization flow rates remains in the region where two phases co-exist at the top of the vapor dome. In this region, the flow rate is strongly affected by the nozzle geometry and tends to be unstable. Various models for this region have been developed with different assumptions, e.g. the HEM and Moody model [3], and the Henry-Fauske non-equilibrium model [4], and are currently used in subcritical pressure reactor safety design[5]. It appears that some of these models could be reasonably extended to above the thermodynamic pseudo-critical point. The more stable and lower discharge flow rates observed in conditions above the pseudo-critical point suggests that even though SCWR's have a smaller coolant inventory, the safety implications of a LOCA and the subsequent depressurization may not be as severe as expected, this however needs to be confirmed by a rigorous evaluation of the particular event and further evaluation of the critical flow rate. This paper will summarize activities on critical flow models, experimental data and numerical modeling during blowdown from supercritical pressure conditions under the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) on 'Heat Transfer Behaviour and Thermo-hydraulics Code testing for SCWRs'. (authors)
- Research Organization:
- American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
- OSTI ID:
- 22105937
- Resource Relation:
- Conference: ICAPP '12: 2012 International Congress on Advances in Nuclear Power Plants, Chicago, IL (United States), 24-28 Jun 2012; Other Information: Country of input: France; 33 refs.; Related Information: In: Proceedings of the 2012 International Congress on Advances in Nuclear Power Plants - ICAPP '12| 2799 p.
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
BLOWDOWN
COMPUTERIZED SIMULATION
COOLANTS
COORDINATED RESEARCH PROGRAMS
CRITICAL FLOW
DEPRESSURIZATION
EXPERIMENTAL DATA
FLOW RATE
HEAT TRANSFER
IAEA
LOSS OF COOLANT
PRESSURE DEPENDENCE
PRESSURE RANGE MEGA PA 10-100
PWR TYPE REACTORS
REACTOR SAFETY
THERMAL HYDRAULICS
TURBULENT FLOW
TWO-PHASE FLOW
VAPORS