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| Content Provider | The American Society of Mechanical Engineers (ASME) Digital Collection |
|---|---|
| Author | Matsui, Asuka Tamitani, Masashi Kudo, Yoshiro Takano, Sho Iwamoto, Tatsuya Nishijima, Mitsuko Kaneko, Junichi Ochi, Hitoshi Takii, Taichi Soneda, Hideo |
| Copyright Year | 2012 |
| 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. |
| Sponsorship | Nuclear Engineering Division Power Division |
| Starting Page | 355 |
| Ending Page | 364 |
| Page Count | 10 |
| File Format | |
| ISBN | 9780791844977 |
| DOI | 10.1115/ICONE20-POWER2012-54438 |
| Volume Number | Volume 3: Thermal-Hydraulics; Turbines, Generators, and Auxiliaries |
| Conference Proceedings | 2012 20th International Conference on Nuclear Engineering and the ASME 2012 Power Conference |
| Language | English |
| Publisher Date | 2012-07-30 |
| Publisher Place | Anaheim, California, USA |
| Access Restriction | Subscribed |
| Subject Keyword | Neutron flux Uncertainty Sensitivity analysis Neutrons Boiling water reactors Fuels Transients (dynamics) Design Fluids Databases Polynomials Safety Thermal hydraulics Delays Porosity Turbines Accidents |
| Content Type | Text |
| Resource Type | Article |
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