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Showing 1 to 20 of 30 for “"RELAP5"”.
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Desarrollo e un simulador termohidráulico informático para PWRs basado en RELAP5/Mod3
La presente tesis consiste en el desarrollo de un proyecto de simulador termohidráulico para PC, tipo "analizador de planta" de una central PWR (C.N.Trillo) con aportaciones originales en cada uno de los hitos fundamentales que deben ser cubiertos en un trabajo de estas características: 1. Control …
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Assessment of RELAP5 MOD3.3 in predicting the flashing behavior of adiabatic steam-water flows
Submission original under an indefinite embargo labeled 'Open Access'. The submission was exported from vireo on 2023-09-01 without embargo terms
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Modeling Two-Phase Flow in the Downcomer of a Once-Through Steam Generator using RELAP5/MOD2
… (OTSG) developed by Babcock & Wilcox, using RELAP5/MOD2. While the physical model can be easily developed, several parameters are left to be adjusted to optimally model the downcomer and match data that was retrieved in a first-of-a-kind (FOAK) study conducted at Oconee Unit I in Oconee, …
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Experimental study of rapid pressure transients in a two-fluid, two-phase system and comparison to RELAP5
… These measurements are compared against a RELAP5 simulation to support the validity of the methods described in this work. Both experiments produce high fidelity measurements that match the expected behavior. The pressure measurements from either side of the interface change concurrently …
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Development, Evaluation and Improvement of Correlations for Interphase Friction in Gas-Liquid Vertical Upflow
… calculations published in the manuals of RELAP5/MOD2, RELAP5/MOD3.3, RELAP5-3D and TRACE. The findings have generally shown that the equations in the manuals provide very inaccurate approximations of the interphase friction compared to the interphase friction that was found via force …
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Investigation of the ability of RELAP5/MOD3 to model natural circulation of high pressure SF6 in the Westinghouse 1/7 scale PWR experimental facility
Thesis: M.S., Massachusetts Institute of Technology, Department of Nuclear Engineering, 1992
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Coupling of system thermal-hydraulics and Monte-Carlo method for a consistent thermal-hydraulics-reactor physics feedback
… with thermal-hydraulics safety analysis code RELAP5-3D. Thermal-hydraulics and reactor physics coupling is commonly used in deterministic methods, for example RELAP5/PARCS and TRACE/PARCS. It has been well-validated for a number of steady and transient problems. The coupling of Monte-Carlo and …
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Evaluation of the Safety Systems in the Next Generation Boiling Water Reactor
… multi-dimensional test assembly (PUMA). RELAP5 code simulations, both for the PUMA main steam line break (MSLB) case and for the simplified boiling water reactor (SBWR) MSLB case have been utilized to compare with the experiment data. The comparison shows that RELAP5 is capable to perform …
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Thermal-hydraulic analysis of innovative fuel configurations for the sodium fast reactor
… of these fuels, the capabilities of the RELAP5-3D code have been expanded to perform subchannel analysis in sodium-cooled fuel assemblies with non-conventional geometries. This expansion was enabled by the use of control variables in the code. When compared to the SUPERENERGY II code, the …
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Simulation and Validation of Two-Component Flow in a Void Recirculation System
… the amount of air in piping systems, using RELAP5/MOD3 as the computational tool, and (2) to validate the numerical results with respect to experimental results and observations.</p> <p>A one-dimensional model of the VRS was built in RELAP5, in which eight different scenarios (replicating …
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Safety analysis of a compact integral small light water reactor
… A model of the reactor is established using RELAP5/3.3gl, with reference to the features of Nuward SMR. Nuward is a compact 170 MWe Pressurized Water Reactor, whose key features include the use of Compact Steam Generators and a large water tank in which the containment submerges for passive …
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Risk-informed design changes for a passive cooling system
… fast reactor. The system was modeled using RELAP5-3D. The epistemic uncertainties in input parameters as well as the epistemic model uncertainties in the code were assessed and propagated through the model using Latin hypercube sampling. The changes in the design that we investigated reduced …
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Improved closure relations for the two fluid-model in flashing flows
… data are used to benchmark the systems code, RELAP5/MOD3.3 where the void fraction is slightly underpredicted. The benchmarking of the static correlations used by RELAP5/MOD3.3 and TRACE to calculate the interfacial area concentration shows inaccuracy, suggesting that the existing model is …
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Inverse uncertainty quantification of trace physical model parameters using Bayesian analysis
… Nuclear thermal-hydraulics codes such as RELAP5 and TRACE do not provide any information on physical model parameter uncertainties. A framework was developed to quantify input model parameter uncertainties based on Maximum Likelihood Estimation (MLE), Bayesian Maximum A Priori (MAP), and …
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Calculating failure probabilities of passive systems during transients
… fast reactor. The system was modeled using RELAP5-3D. The uncertainties in input parameters were assessed and were propagated through the model using Latin Hypercube Sampling. An important finding was the discovery that the smaller pressure loss through the DHR heat exchanger than through …
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LEU-HEU Mixed Core Conversion Thermal-hydraulic Analysis and Coolant System Upgrade Assessment for the MIT Research Reactor
… Thermal hydraulics analysis was performed using RELAP5 MOD3.3 code and STAT7 code. The MITR transitional mixed core input models were constructed to simulate the reactor primary system. Two scenarios, steady-state and loss-of-flow transient were simulated at power level of 6 MW. RELAP5 results …
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Integrating thermal energy storage and nuclear reactors: A technical and policy study
… control. A detailed nodal model was created in RELAP5-3D to assess the safety implications of hybrid PRISM-nTES. The reactor performed similarly to the reference PRISM design for bounding events that did not include a Loss of Heat Sink (LOHS). In those events where a LOHS occurred, the reactor …
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Thermal hydraulic design of a 2400 MW t̳h̳ direct supercritical CO₂-cooled fast reactor
… an extensive series of iterative calculations, RELAP5-3D was then used to evaluate the natural circulation performance of an active/passive hybrid Shutdown/Emergency Cooling System (SCS/ECS). Routes were identified by which significant post-LOCA core bypass could occur and degrade the decay heat …
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CFD Analysis of Aspirator Region in a B&W Enhanced Once-Through Steam Generator
… is important to the development of accurate RELAP5 models, steam generator level calculations, steam generator downcomer models, and flow induced vibration analyses. The OpenFOAM CFD software package was used to develop the three-dimensional model of the EOTSG aspirator region, perform the …
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Thermal hydraulics analysis of the MIT research reactor in support of a low enrichment uranium (LEU) core conversion
… against PLTEMP for steady-state analysis, and RELAP5 and temperature measurements for the loss of primary flow transient. Various fuel configurations are evaluated as part of the LEU core design optimization study. The criteria adopted for the LEU thermal hydraulics analysis for this study are …
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