No abstract

TRACG code, coupling a three-dimensional neutron kinetics model for the reactor core with thermal-hydraulics based on two-fluid conservation equations, is a best-estimate (BE) code for BWRs to realistically simulate their transient and accidental behaviors. TRACG05 is the latest version and was originally developed to analyze Reactivity Initiated Accident (RIA). TRACG05 incorporates the same neutronics model of the latest core simulator with a three-group analytic-polynomial nodal expansion method. In addition to application to RIA safety analyses, TRACG05 has been planned to apply to safety analyses for Anticipated Operational Occurrences (AOOs) in BWRs by using a Best Estimate Plus Uncertainty (BEPU) methodology. To apply BEPU with TRACG05 to BWR AOOs, validations must be performed to evaluate the uncertainties of models relevant to important phenomena by comparing with appropriate test results for BWR AOOs. At first, a PIRT (Phenomena Identification and Ranking Table) was developed for each event scenario in AOOs to identify relevant physical processes and to determine their relative importance. According to the PIRT, an assessment matrix was established for separate effects tests (SETs), component effects tests (CETs), integral effects tests (IETs), and integral BWR plant start-up tests. The assessment matrix related the important phenomena to the test database, which was confirmed that all the important phenomena were covered by all tests specified in the matrix. According to the assessment matrix, comparison analyses have been specified to perform systematic and comprehensive validations of TRACG05 applicability to AOOs. The comparison analyses were done as the integrated code system with the up-stream reactor core design codes, therefore higher quality was enabled to evaluate the safety parameters. As the result, the uncertainties of important models in TRACG05 were determined so as to enable BEPU approaches for AOO safety issues. Here, as a SET, comparisons between TRACG05 and experimental data of void fraction in a bundle simulating an actual fuel bundle, which is one of the most important models in the application of TRACG05 to AOO analyses are shown. In addition, as pressurization event in AOOs, comparisons between TRACG05 and experimental data of Peach Bottom 2 Turbine Trip Test, which is one of integral tests for a BWR plant, are shown. This is the only test showing large neutron flux increase and strong coupling of neutron kinetics and thermal-hydraulics in the core due to void and Doppler feedbacks. Furthermore, a sensitivity analysis regarding a delay time of control rod (CR) insertion initiation which was the most sensitive uncertainty to the results is also shown.

No abstract

Interface-System Loss-of-Coolant Accident (ISLOCA) occurs by failure of isolation valves in piping systems. In this study, transient behavior of pressure propagation in piping systems after ISLOCA has been estimated. At first, capability of TRACG code to predict pressure propagation in a simple pipe line, which has expansion, contraction and bifurcation, has been investigated. Then, sensitivity analysis of valve opening period has been performed to investigate the behavior of pressure propagation in a simple pipe line. Finally, pressure propagation inside piping systems after ISLOCA has been simulated for Hitachi-GE standard Advanced boiling water reactor (ABWR) plant. Maximum pressure inside High Pressure Core Flooder (HPCF) systems has been less than 7.8MPa. TRACG code has been shown to be useful to predict transient behavior of pressure propagation in complicated network of piping systems after ISLOCA.

As a measure to prevent the generation of combustible hydrogen/oxygen gases expected in an ABWR accident, a flammability control system (FCS) needs to be installed. The autocatalytic FCS, which has been introduced mainly in Europe and the US, is composed of multiple passive autocatalytic recombiners (PARs) and is capable of recombining hydrogen and oxygen at room temperature (low activation energy) by catalysis. Unlike conventional forced-circulation and heating FCS, the autocatalytic FCS is a passive system that does not need operator startup; it also has a simple structure, without components requiring power supplies, such as blowers and heaters. Thus, it is superior in terms of reliability and operability. This system is considerably smaller than a conventional FCS, and can be installed in portions within the pressure containment vessel (PCV); conventional FCSs are installed inside the reactor building (outside the PCV). So autocatalytic FCS makes it possible to reduce the size of the building, and construction costs as well. Though multiple tests have been performed in Europe and the US on the autocatalytic FCS for product development and to check performance[1],[2],[3],[4], before introducing this FCS to a new ABWR plants, some additional tests have been performed with consideration for Japanese BWR accident conditions like PCV spray, influence of low oxygen condition and reaction inhibition in the collaborative research by Japanese electric power companies. These include examining the influences of low-oxygen conditions and reaction inhibitors. This paper presents the details of additional studies made after considering results of the additional tests performed in Japan to install the autocatalytic FCS instead of the thermo-reactive FCS.