LOCA analysis

RPV, and with helps of EDRS and CWCS. Details will be described in later (in LOCA analysis). Thus, the core flooding in a LOCA can be attained passively without ECC pumps or an accumulator if the containment initial water level is appropriate. Experiment [6] on the thermal hydraulic behaviour of water-filled containment was conducted and the relationships between the initial water level of containment and the balance pressure, etc., were obtained, and the principal function of core flooding was confirmed experimentally. 

The functions of the engineered passive safety systems of the MRX which are significantly simplified are evaluated with the safety analyses. The LOCA analysis as an example presented in the previous chapter shows that core flooding is kept and decay heat removal is performed successively. Corresponding to the Japanese governmental PWR licensing safety review guideline , the analyses of the accidents and the anticipated transient events are conducted. The... 

The new design concepts under feasibility studies are considering variable speed in the recirculation pump. This concept requires evaluation of different transients scenarios that those analyzed in the original design, because of the consequences of re-circulation flow transient having different impact in the core performance during transients and also in LOCA analysis. 

Since the conservative ECCS evaluation model identified in 10 CFR 50.46 Appendix K, remains valid for LOCA analysis, and since this method is approved by the NRC, this issue is resolved for the System 80+ Standard Design. 

Generic Safety Issue (GSI) C-05 in NUREG-0933 (Reference 1), addresses the need for a designer to select a specific decay heat function for the nuclear power plant LOCA analysis. There are two permissible decay heat functions, with associated uncertainties, which can be used in ECCS evaluation models. Each of these decay heat functions is a part of a particular LOCA evaluation model. 

The System 80+ standard Design addresses loss-of-coolant (LOCA) events by incorporating a conservative LOCA analysis with a conservative system design. 

With respect to conservative LOCA analysis, the System 80+ Standard Design meets the criteria of 10 CFR 50.46, by using the highly conservative fission product decay heat function identified in ANS 5.0, (Proposed), in lieu of the new and more realistic decay heat model described in ANS 5.1. 

In addition from a design standpoint, the Safety Injection System has been improved by including two additional high pressure safety injection pumps (for a total of four pumps). Because of this upgrade, the safety margin for the LOCA analysis for the System 80+ Standard Design has been increased. 

Detailed reactor physics assessments have been performed for the DUPIC fuel, including lattice studies, detailed time-dependent fuel management simulations, and LOCA analysis (Choi et al. 1997). These studies confirm that DUPIC fuel can be accommodated within existing CANDU reactors. Some of the key results of these studies are summarized below. 

The results of the analysis show that the overtemperature AT reactor protection system signal provides adequate protection against the reactor coolant system depressurisation events. The calculated DNBR remains above the design limit. The long-term plant response due to a stuck-open ADS valve or pressuriser safety valve, which cannot be isolated, is bounded by the small-break LOCA analysis. 

TABLE IX-5. MAJOR ANALYTIC CONDITIONS FOR LOCA ANALYSIS... 

Richards, D.J., Hanna, B.N., Hobson, N., Ardron, K.H., 1985. ATHENA, a two-fluid code for CANDO LOCA analysis. In Third International Topical Meeting on Reactor Thermal Hydraulics, Newport, Rhode Island, USA, October 15—18, 1985, 7E-1 thru 7E-14 (CATHENA Formerly Named ATHENA). 

DOE requires that the assumptions, input data, and analyses following the Upgraded Thermal-Hydraulic Limits Action Plan and the Updated SAR Chapter 15 Action Plan should be consistent. Important parameters for thermal-hydraulic limits analysis are mostly flow related whereas, for non-LOCA analysis, the important parameters are temperature related. Inputs to transient analyses may, therefore, require different sets of boundary conditions. The consistency requirement will be satisfied if any differences are specifically noted and can be satisfactorily explained in the SAR. Compliance with this requirement is an open item. 

One-dimensional bum-up and two-dimensional R-Z core calculation procedures for the fast reactor are found in refs. [108,109]. The LOCA analysis code, SCRELA was described in ref. [110]. The procedure for a simplified PSA is also described there. The single channel thermal-hydrauUc calculation code SPROD, the two-dimensional coupled core calculation scheme of the thermal spectram reactor with water rods and transient and accident analysis code at supercritical-pressure, SPRAT are described in ref. [111]. 

Safety Transient and accident analysis at supercritical-and subcritrical pressure (SPRAT-F, SPRAT-DOWN), ATWS analysis (SPRAT-DOWN), LOCA analysis (SCRELA,SPRAT-DOWN-DP), Time-dependent subchannel analysis Start-up (sliding pressure and constant pressure)... 

Safety analysis of the Super LWR is described in ref. [121]. The SPRAT-DOWN code for the analysis of downward flowing water rods and the SPRAT-DOWN-DP code for depressurization in an LOCA were prepared. The LOCA analysis of the Super LWR was performed in combination with SPRAT-DOWN-DP and the reflooding module of SCRELA. ATWS analysis is also described in ref. [121]. The momentum equation is included in the SPRAT-DOWN code from the ATWS analysis. The design of the two-pass core of the Super LWR and the safety analysis at subcritical pressure during startup are described in ref. [122]. 

J.H. Lee, S. Koshizuka and Y. Oka, Development of a LOCA Analysis Code for the Supercritical-Pressure Light Water Cooled Reactors, Annals of Nuclear Energy, Vol. 25 (16), 1341-1361 (1998)... 

J.H. Lee, LOCA Analysis and Safety System Consideration for the Supereritical-Water Cooled Reactor, Doctoral Thesis, the University of Tokyo (1996)... 

Y. Ishiwatari, Y. Oka and S. Koshizuka, LOCA Analysis of High Temperature Reactor Cooled and Moderated by Supercritical Water, Proc. GENES4/ANP2003, Kyoto, Japan, September 15-19, 2003, Paper 1060 (2003)... 

Since the pressure drop in the main coolant lines and the RPV is negligibly compared with the pressure drop at the break or in the ADS lines, the pressure is assumed to be uniform in the main coolant lines and the RPV. Decrease in the pressure is governed by the break flow rate. Since the break flow rate cannot be calculated by SPRAT-DOWN-DP, it is calculated using correlations and given as the boundary conditimis. The correlations selected and used in the LOCA analysis... 

US NUCLEAR REGULATORY COMMISSION, Compendium of ECCS Research for Realistic LOCA Analysis, NUREG 1230, USNRC, Washington (DC), 1988. 

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