Tests simulating Pressurized Water Reactor pressurizers under inflow and outflow conditions have been performed at MIT. Prediction of pressurizer pressure requires accurate models of wall heat transfer as well as interfacial liquid-steam heat transfer. The US NRC has two computer programs used for predicting thermal hydraulic behavior in reactors; RELAP5 and TRAC-M. TRAC-M is the Consolidated Thermal-hydraulics Code developed by combining models from TRAC-B and RELAP5 into a modernized version of TRAC-P. The component models from RELAP5 and TRAC-B have been ported to TRAC-M but not the constitutive models. A suite of assessment cases are being developed to guide the constitutive model improvement process. Assessment against data will determine which constitutive relations need to be ported to TRAC-M. This paper compares the RELAP5 and TRAC-M codes against MIT pressurizer data. As water is injected into the bottom of the pressurizer the steam pressure in the top of the pressurizer rises. The pressure increase rate is controlled by wall and interface condensation rates. Both codes predict this complex compression process reasonably well. The effect of time step size and code options are explored in this paper. The benefits of using two codes to analyze thermal-hydraulic processes are evident from the results.
The presence of stratified liquid-gas interfaces in vertical flows poses difficulties to most classes of solution methods for two-phase flows of practical interest in the field of reactor safety and thermal-hydraulics. These difficulties can plague the reactor simulations unless handled with proper care. To illustrate these difficulties, the US NRC Consolidated Thermal-hydraulics Code (TRAC-M) was exercised with selected numerical bench-mark problems. These numerical benchmarks demonstrate that the use of an average void fraction for computational volumes simulating vertical flows is inadequate when these volumes consist of stratified liquid-gas interfaces. In these computational volumes, there are really two regions separated by the liquid-gas interface and each region has a distinct flow topology. An accurate description of these divided computational volumes require that separate void fractions be assigned to each region. This strategy requires that the liquid-gas interfaces be tracked in order to determine their location, the volumes of regions separated by the interface, and the void fractions in these regions. The idea of tracking stratified liquid-gas interfaces is not new. There are examples of tracking methods that were developed for reactor safety codes and applied to reactor simulations in the past with some limited success. The users of these safety codes were warned against potential flow oscillations, conflicting water levels, and pressure disturbances which could be caused by the tracking methods themselves. An example of these methods is the level tracking method of TRAC-M. A review of this method is given here to explore the reasons behind its failures. The review shows that modifications to the field equations are mostly responsible for these failures. Following the review, a systematic approach to incorporate interface tracking methods is outlined. This approach is applicable to most classes of solution methods. For demonstration, the approach to incorporate the tracking method into the field equations of TRAC-M is described in steps. The success of this approach is demonstrated by exercising TRAC-M with the same benchmark problems that were previously used to illustrate the difficulties the code suffered from in the presence of interfaces. Besides improvements to the accuracy of the code predictions, one of the benchmark problems, which simulates a strong condensation at the liquid-gas interface, shows that the code’s runtime is improved significantly where the alternative methods like water packing fails.
A deficiency has been identified with the use of boron concentration for formulating reactivity feedback while the power of a PWR in response to boron injection in the aftermath of a scram failure was simulated. The US NRC Consolidated Code (TRAC-M) was used to simulate this transient, in which the boron injection was expected to shutdown the reactor. The results of this study, which employed the point-kinetics model of TRAC-M, reveal that the use of core-average boron concentration for formulating the reactivity feedback is inadequate for transients when there is substantial coolant void inside the core. It is recommended that the macroscopic boron density be used for more accurate predictions of the boron feedback as the use of solute concentration is only adequate when no voiding occurs in the core.
A generic guideline for thermal-hydraulic (T-H) simulation of multiple bank safety relief valves (SRVs) was developed. To test the guideline, the Full Integral Simulation Test (FIST) 6PMC2 was simulated with the consolidated T-H code of USNRC, TRACE. The FIST 6PMC2 experiment simulates the response of a generic BWR/6 plant to a Main Steam-line Isolation Valve (MSIV) closure without reactor scram. During the test, the HPCS is unavailable and not used. The only inventory make-up system available is the Reactor Core Isolation Cooling (RCIC). This experiment can also be considered proto-typical of a BWR ATWS-like scenario. The simulation is largely dominated by the transient behavior of the SRVs. In this study, the experimental data was analyzed and used to check the modeling guideline for SRVs. The guideline relies on only the data available prior to an experiment or any other analysis, e.g. valve flow coefficients, inlet hydraulic diameters, etc. The study also revealed deficiencies in the “then” current valve model of the TRACE code which were subsequently corrected. The study demonstrates that the T-H models can simulate the operational behavior of SRVs very accurately while rather simple mistakes can be very damaging at the same time.
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