Assessment of RELAP5-3D for Analysis of Very High Temperature Gas-Cooled Reactors PDF Download
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Author: Chang Oh
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Languages : en
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The RELAP5-3D© computer code is being improved for the analysis of very high temperature gas-cooled reactors. Diffusion and natural circulation can be important phenomena in gas-cooled reactors following a loss-of-coolant accident. Recent improvements to the code include the addition of models that simulate pressure loss across a pebble bed and molecular diffusion. These models were assessed using experimental data. The diffusion model was assessed using data from inverted U-tube experiments. The code's capability to simulate natural circulation of air through a pebble bed was assessed using data from the NACOK facility. The calculated results were in reasonable agreement with the measured values.
Author: Chang Oh
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Category :
Languages : en
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Book Description
The RELAP5-3D© computer code is being improved for the analysis of very high temperature gas-cooled reactors. Diffusion and natural circulation can be important phenomena in gas-cooled reactors following a loss-of-coolant accident. Recent improvements to the code include the addition of models that simulate pressure loss across a pebble bed and molecular diffusion. These models were assessed using experimental data. The diffusion model was assessed using data from inverted U-tube experiments. The code's capability to simulate natural circulation of air through a pebble bed was assessed using data from the NACOK facility. The calculated results were in reasonable agreement with the measured values.
Author: Eugene James Thomas Moore
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Languages : en
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High-temperature gas-cooled reactors (HTGR) are passively safe, efficient, and economical solutions to the world's energy crisis. HTGRs are capable of generating high temperatures during normal operation, introducing design challenges related to material selection and reactor safety. Understanding heat transfer and fluid flow phenomena during normal and transient operation of HTGRs is essential to ensure the adequacy of safety features, such as the reactor cavity cooling system (RCCS). Modeling abilities of system analysis codes, used to develop an understanding of light water reactor phenomenology, need to be proven for HTGRs. RELAP5-3D v2.3.6 is used to generate two reactor plant models for a code-to-code and a code-to-experiment benchmark problem. The code-to-code benchmark problem models the Russian VGM reactor for pressurized and depressurized pressure vessel conditions. Temperature profiles corresponding to each condition are assigned to the pressure vessel heat structure. Experiment objectives are to calculate total thermal energy transferred to the RCCS for both cases. Qualitatively, RELAP5-3D's predictions agree closely with those of other system codes such as MORECA and Thermix. RELAP5-3D predicts that 80% of thermal energy transferred to the RCCS is radiant. Quantitatively, RELAP5-3D computes slightly higher radiant and convective heat transfer rates than other system analysis codes. Differences in convective heat transfer rate arise from the type and usage of convection models. Differences in radiant heat transfer stem from the calculation of radiation shape factors, also known as view or configuration factors. A MATLAB script employs a set of radiation shape factor correlations and applies them to the RELAP5-3D model. This same script is used to generate radiation shape factors for the code-to-experiment benchmark problem, which uses the Japanese HTTR reactor to determine temperature along the outside of the pressure vessel. Despite lacking information on material properties, emissivities, and initial conditions, RELAP5-3D temperature trend predictions closely match those of other system codes. Compared to experimental measurements, however, RELAP5-3D cannot capture fluid behavior above the pressure vessel. While qualitatively agreeing over the pressure vessel body, RELAP5-3D predictions diverge from experimental measurements elsewhere. This difference reflects the limitations of using a system analysis code where computational fluid dynamics codes are better suited.
Author: C. B. Davis
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Category :
Languages : en
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Oxygen, carbon dioxide, and carbon monoxide have been added to the RELAP5-3D computercode as noncondensable gases to support analysis of high temperature gas-cooled reactors. Models of these gases are required to simulate the effects of air ingress on graphite oxidationfollowing a loss-of-coolant accident. Correlations were developed for specific internal energy, thermal conductivity, and viscosity for each gas at temperatures up to 3000 K. The existingmodel for internal energy (a quadratic function of temperature) was not sufficiently accurate atthese high temperatures and was replaced by a more general, fourth-order polynomial. Themaximum deviation between the correlations and the underlying data was 2.2% for the specificinternal energy and 7% for the specific heat capacity at constant volume. The maximumdeviation in the transport properties was 4% for oxygen and carbon monoxide and 12% forcarbon dioxide.
Author: Nolan Alan Anderson
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Languages : en
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The Very High Temperature Reactor (VHTR) system behavior should be predicted during normal operating conditions and during transient conditions. To predict the VHTR system behavior there is an urgent need for development, testing and validation of design tools to demonstrate the feasibility of the design concepts and guide the improvement of the plant components. One of the identified design issues for the gas-cooled reactor is the thermal mixing of the coolant exiting the core into the outlet plenum. Incomplete thermal mixing may give rise to thermal stresses in the downstream components. This analysis was performed by coupling a RELAP5-3D VHTR model to a Fluent outlet plenum model. The RELAP5 VHTR model outlet conditions provide the inlet boundary conditions to the Fluent outlet plenum model. By coupling the two codes in this manner, the important three-dimensional flow effects in the outlet plenum are well modeled without having to model the entire reactor with a computationally expensive code such as Fluent. The two codes were successfully coupled. The values of pressure, mass flow rate and temperature across the coupled boundary showed only slight differences. The coupling tool used in this analysis can be applied to many different cases requiring detailed three-dimensional modeling in a small portion of the domain.
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Category :
Languages : en
Pages : 15
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The capability of the RELAP5-3D computer code to perform multi-dimensional analysis of a pressurized water reactor (PWR) was assessed using data from the Loss-of-Fluid Test (LOFT) L2-5 experiment. The LOFT facility was a 50 MW PWR that was designed to simulate the response of a commercial PWR during a loss-of-coolant accident (LOCA). Test L2-5 simulated a 200% double-ended cold leg break with an immediate primary coolant pump trip. A three-dimensional model of the LOFT reactor vessel was developed. Calculations of the LOFT L21-5 experiment were performed using the RELAP5-3D computer code. The calculations simulated the blowdown, refill, and reflood portions of the transient. The calculated thermal-hydraulic response of the primary coolant system was generally in reasonable agreement with the test. The calculated results were also generally as good as or better than those obtained previously with RELAP5/MOD3.
Author: C. B. Davis
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Languages : en
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The Actinide Burner Test Reactor (ABTR) is envisioned as a sodium-cooled, fast reactor that will burn the actinides generated in light water reactors to reduce nuclear waste and ease proliferation concerns. The RELAP5-3D computer code is being considered as the thermal-hydraulic system code to support the development of the ABTR. An evaluation was performed to determine the applicability of RELAP5-3D for the analysis of a sodium-cooled fast reactor. The applicability evaluation consisted of several steps, including identifying the important transients and phenomena expected in the ABTR, identifying the models and correlations that affect the code's calculation of the important phenomena, and evaluating the applicability of the important models and correlations for calculating the important phenomena expected in the ABTR. The applicability evaluation identified code improvements and additional models needed to simulate the ABTR. The accuracy of the calculated thermodynamic and transport properties for sodium was also evaluated.
Author: American Nuclear Society
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Category : Nuclear engineering
Languages : en
Pages : 1068
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Languages : en
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The Reactor Cavity Cooling System (RCCS) is an important passive safety system being incorporated into the overall safety strategy for high temperature advanced reactor concepts such as the High Temperature Gas- Cooled Reactors (HTGR). The Natural Convection Shutdown Heat Removal Test Facility (NSTF) at Argonne National Laboratory (Argonne) reflects a 1/2-scale model of the primary features of one conceptual air-cooled RCCS design. The project conducts ex-vessel, passive heat removal experiments in support of Department of Energy Office of Nuclear Energy's Advanced Reactor Technology (ART) program, while also generating data for code validation purposes. While experiments are being conducted at the NSTF to evaluate the feasibility of the passive RCCS, parallel modeling and simulation efforts are ongoing to support the design, fabrication, and operation of these natural convection systems. Both system-level and high fidelity computational fluid dynamics (CFD) analyses were performed to gain a complete understanding of the complex flow and heat transfer phenomena in natural convection systems. This paper provides a summary of the RELAP5-3D NSTF model development efforts and provides comparisons between simulation results and experimental data from the NSTF. Overall, the simulation results compared favorably to the experimental data, however, further analyses need to be conducted to investigate any identified differences.
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Languages : en
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Book Description
The capability of the RELAP5-3D computer code to perform multi-dimensional analysis of a pressurized water reactor (PWR) was assessed using data from the LOFT L2-5 experiment. The LOFT facility was a 50 MW PWR that was designed to simulate the response of a commercial PWR during a loss-of-coolant accident. Test L2-5 simulated a 200% double-ended cold leg break with an immediate primary coolant pump trip. A three-dimensional model of the LOFT reactor vessel was developed. Calculations of the LOFT L2-5 experiment were performed using the RELAP5-3D Version BF02 computer code. The calculated thermal-hydraulic responses of the LOFT primary and secondary coolant systems were generally in reasonable agreement with the test. The calculated results were also generally as good as or better than those obtained previously with RELAP/MOD3.
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Languages : en
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Using experimental data published in the International Handbook of Evaluated Reactor Physics Benchmark Experiments for the fresh cold core of the High Temperature Engineering Test Reactor, a comprehensive validation study has been carried out to assess the performance of the SCALE code system for analysis of High Temperature Gas-Cooled Reactor (HTGR) configurations. This paper describes part of the results of this effort. The studies performed included criticality evaluations for the full core and for the annular cores realized during the fuel loading, as well as calculations and comparisons for excess reactivity, shutdown margin, control rod worths, temperature coefficient of reactivity, and reaction rate distributions. Comparisons of the SCALE results with both the experimental values and MCNP-calculated values are presented. The comparisons show that the SCALE calculated results, obtained with both multigroup and continuous energy cross sections, are in reasonable agreement with the experimental data. The agreement with the MCNP predictions is, in general, very good.
