Loss of Coolant Accident Analyses

Bauer, E. G., Houdayer, G. R. Sureau, H. M. 1976. A nonequilibrium axial flow model and application to loss-of-coolant accident analysis The CLYSTERE system code. Proceedings of CSNI Meeting on Transient Two-Phase Flow, Toronto, Canada. 

WCAP-12945-P-A, Volumes 1-5, Westinghouse Code Qualification Doeiunent for Best Estimate Loss of Coolant Accident Analysis, Revision 1, March 1998. 

Describe three key conservatisms inherent in traditional design basis loss of coolant accident analysis with respect to long-term core coolability. 

S. rkejrri, Y. Ishiwatari, Y. Oka, Loss of Coolant Accident Analysis of a Supercritical-Pressure Water-Cooled Fast Reactor with Downward Flow Channels, Proc. ICAPP09, Tokyo, Japan, May 10-15, 2009, Paper 9257 (2009)... 

Further studies concerning the stability limit of water at higher pressure and temperature (where this method is expected to be less accurate) with more realistic pressure-profile are in progress. Those data would be particularly important in the safety analysis of power plants (including nuclear ones) where during a so-called LOCA (Loss of Coolant Accident) part of the cooling liquid can reach some degree of metastability due to sudden pressure loss (see e.g. [ ]). 

An accident at a nuclear power plant can be caused by many combinations of anomalous initiating event, malfunction and human error. The types of possible accidental situations are studied in the specific safety analysis of each plant and the safety systems described above are designed to prevent, or mitigate the effects of all the accidents chosen as DBAs. Table 3-1 provides an approximate indication of the effectiveness of various safety systems in limiting external releases in a typical loss of coolant accident (the break of a large primary circuit pipe). The figures are for the release of iodine-131 (often assumed as the reference isotope in indicative evaluations of source terms and for a 1000 MWe reactor). As can be seen, the reduction of the releases caused by the safety systems is very significant and corresponds to a factor of the order of one million. 

The criteria applied in the design of the Reactor Coolant System supports are that the specific function of the supported equipment be achieved during all normal, earthquake, safety valve actuation and Branch Line Pipe Break (BLPB) conditions. (BLPB includes feedwater line breaks and all loss-of-coolant-accident conditions resulting from breaks not eliminated by leak-before-break analysis in piping to branch nozzles of the reactor coolant system.) Specifically, the supports are designed to support and restrain the Reactor Coolant System components under the combined Safe Shutdown Earthquake and Branch Line Pipe Break loadings in accordance with the stress and deflection limits of Section III, ASME Code. 

Since 1974 the NRC requirements for performing a loss-of-coolant-accident (LOCA) licensing analyses (ECCS analyses) have been specified in 10 CFR 50.46 Appendix K (Reference 2). During the years since 1974, extensive research has been conducted on the various aspects of a LOCA. Because of this research, 10 CFR 50 now states that "...It is now confirmed that the methods specified in Appendix K, combined with other analysis methods currently in use, are conservative and that the actual cladding temperature would be much lower than that calculated using Appendix K methods". 

NUREG-1463, "Regulatory analysis for the resolution of Generic Safety Issue 105 Interfacing system loss of coolant accident in light water reactors", July 1993. 

The important steel reactor pressure vessel analysis is to check them against loss-of-coolant accident (LOCA), especially in PWR. A loss-of-coolant accident (LOCA) occurs as a result of a penetration to the main coolant boundary such that the primary circuit water is released through the break to the containment area, causing a rapid decrease in the pressure and temperature of the primary coolant Fig. 4.19. This will give an impact thermal shock load. The streamline break accident (SLBA) occurs as a result of a complete and partial rupture of a steam line inside the containment vessel. A rapid cool-down and depressurization of the primary circuit normally take place. In order to restore... 

The scope, extent and detail of the safety analysis for low power research reactors may be significantly less than is required for high power research reactors because certain accident scenarios may not apply or may need only a limited analysis. For example, the treatment of loss of coolant accidents may differ significantly, depending on the power and design of the reactor. Paragraphs 6.72-6.78 establish requirements for the scope, factors and process to be considered in the safety analysis. 

III-5. As an alternative to such an analysis of loss of coolant accidents, it is the practice in some States to specify the fractions of core inventory of fission products that are assumed to reach the containment atmosphere after the accident. This fraction is specified differently for different categories of chemical elements, but will usually be independent of the design measures taken against accidents of such types. Thus, these fractions are set as an assumed upper limit irrespective of the performance characteristics of the emergency core cooling system [III-4 to III-6]. 

III-8. If the location of the steam line break is within the containment, the sequence of events is similar to that for loss of coolant accidents, but with a different fraction of the fuel cladding failing. The equilibrium concentration of fission products for full power operating conditions has to be assumed. The design analysis for the potential radioactive release has to consider the time needed for containment isolation to take place and the effectiveness of the coolant purification system. 

The design basis accidents for the ECS are a loss of coolant accident from a double-ended guillotine break (DEGB) of a large process water line (D20) and a loss of pumping accident from a cooling water line break. Section 6.2.3.3.1.1 of the Safety Analysis Report (SAR) states that the ECS addition rate for a... 

The design basis events (DBFs) considered in establishing the DBA-power limits represent the full group of accidents that comprise the transient and accident analysis section (Chapter 15) of the SAR. Historically, in the operation of the SRS reactors, a number of power limits that originate from the different DBAs have controlled the maximum power level at which operation was permitted. Typically, the DBAs that have been considered are the loss of coolant accident, the loss of pumping accident, the gang rod withdrawal accident, and the pump shaft break accident. 

The main pipe rupture accident is the most serious one for the loss of coolant accident of pool type sodium cooled Fast Breeder Reactor (FBR). To simulate this accident, a model is developed based on the OASIS code, which is a French fast reactor system safety analysis code. To abide by the strict accident analysis principles, the main pipe rupture accident is calculated for various position of the pipe. Accident sequence and key parameters, including the fuel cladding temperature of reactor, are obtained for each case. The calculation results show that the fuel cladding temperature is below the safety limitation and the coolant temperature is lower then the saturation temperature of sodium in all cases. 

Abnormal transients are defined as events that will lead to the situation in which the nuclear plant cannot maintain the normal operation due to an external disturbing factor that may occur during the life span of the nuclear plant under the operational conditions including single failure or malfunction of the devices or single operational errors by operators, and to the abnormal situation in which the nuclear plant is not planned to operate and that may occur with the same probability as the former. A set of abnormal transients and accidents as standard safety analysis of the current LWRs is studied for the Super LWR, including the loss of coolant accident (LOCA) and the anticipated transients without scram (ATWS) (see Chap. 6 for details). The requirements for the Super LWR are same as those of LWRs ... 

Pressurization events test the capability of the LRVs to overcome the pressure transient. The accident analysis also includes failure of one of these valves to reclose, thereby testing the ability in the long term to stop any unrecoverable loss of coolant. Depressurization events are similar to a small LOCA. Since in the depressurization sequences there may be no immediate discharge to containment, appropriate signals must be identified for reactor trip and/or operator alarms and, if necessary, for ECC. 

---