Passive Safety System [Slides]
These slides are intended as an introduction to the concept of passive safety systems in LWRs and a discussion on designs that incorporate passive safety.
22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
Engineering topics
Kile, Robby
Publications and source records attributed to Kile, Robby.
RELAP5-3D in the OECD-NEA HTGR Thermal Hydraulics Benchmark
The OECD-NEA HTGR Thermal Hydraulics benchmark uses data from the High Temperature Test Facility for a series of code-to-code and code-to-data exercises aimed at improving the state of knowledge on existing thermal hydraulics modeling and simulation tools for prismatic HTGR applications. This benchmark includes problems representing hot gas mixing in the lower plenum, the depressurized conduction cooldown accident and the pressurized conduction cooldown accident. This presentation discusses the benchmark and RELAP5-3D's role in that benchmark, including as a tool for predicting behavior in the core and as a tool for providing boundary conditions to computational fluid dynamics analysis in the lower plenum.
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RELAP5-3D Validation Using the Primary Coolant Apparatus Test (PCAT) Facility
Presentation for discussing the use for PCAT for validating RELAP5-3D
21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
University of Wyoming Light Water Reactor Slides
This slide set presents an introduction to light water reactors, including some introductory material on the nuclear fuel cycle. These slides are for a course at the University of Wyoming on nuclear fundamentals.
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Comparison of Results between the Legacy and Refined RELAP5-3D Models of the High Temperature Test Facility in Exercises 1 and 2 of the HTTF Benchmark
Work conducted in FY23 identified that RELAP5-3D was capable of reproducing trends in HTTF data during experiment PG-27 but was incapable of reproducing measured values. The primary cause of this discrepancy between RELAP5-3D results and experimental data was hypothesized to be a distortion in power density that was introduced by the radial nodalization of the model. We further hypothesized that a new model would provide better results when compared to the experiments PG-27 and PG-29. Work this FY developed a new model that is better capable of capturing local heat generation rates and contains a representation of each 1/6 azimuthal sector of the core. We used this model to develop a new set of solutions to Exercises 1 and 2 of Problems 2 and 3 in the benchmark. In this report, we present the first comprehensive comparison of the results between the two models. We see that in Exercise 1A, which is common between problems 2 and 3, the results are similar, though the results from the new model show greater detail than those from the legacy model. In Problem 2 Exercise 1B and Problem 3 Exercise 1B, we see that heat removal is slower in the new model than the legacy model. Problem 3 Exercise 1C shows temperatures that are lower in most places in the new model than the legacy model, but the area with active heat generation has higher block temperatures in the new model than the legacy model. Problem 3 Exercise 1D further shows that long-term heat removal is lower in the new model. Problem 2 Exercise 1C demonstrated that the new model observes higher temperatures in the core regions than the legacy model, justifying the need to preserve the power density in HTTF. The validation of PG-27 and PG-29 also demonstrated the improved temperature agreement in the core regions, particularly with a calibrated model that implements an effective thermal conductivity for the core material. Overall, PG-27 models show reasonable to excellent agreement for steady-state temperatures and minimal to reasonable agreement for transients. PG-29 models showed minimal to insufficient agreement with the data.
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NEA HTTR LOFC Project Test#3 Benchmark Results
In the second half of FY23, the High Temperature Engineering Test Reactor (HTTR) Loss Of Forced Cooling (LOFC)#3 data for the 9 MW test case with Vessel Cooling System (VCS) off were made available through the Nuclear Energy Agency (NEA) LOFC project framework; the neutronic model developed for the initial test (LOFC#1) achieved a satisfactory level of maturity, demonstrating its accuracy in predicting power evolution and core re-criticality, but LOFC#3 should be used primarily to investigate thermal hydraulic phenomena, as the reactor was shut down prematurely due to overheating in the upper reactor components, which prevented re-criticality; this report focuses on advancing the HTTR thermal hydraulic model to accurately simulate the LOFC#3 scenario, including simulating the LOFC#3 benchmark and generating the corresponding benchmark specifications, aiming to ensure consistency across participant models and provide essential data for future participants, including private industry stakeholders seeking to validate their computational tools.
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