Comparisons of RELAP5-3D Analyses to Experimental Data from the Natural Convection Shutdown Heat Removal Test Facility PDF Download
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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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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.
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.
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Languages : en
Pages : 50
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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 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.
Author: Hsun-Chia Lin
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Languages : en
Pages : 102
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Direct Reactor Auxiliary Cooling System (DRACS), a passive safety system proposed for Fluoride-salt-cooled High-temperature Reactors (FHRs), is driven by natural convection/circulation to remove the decay heat during accidents. Two test facilities were designed and built following a scaling analysis at The Ohio State University to test the thermal performance of the DRACS, including a low-temperature DRACS test facility (LTDF) and a high-temperature DRACS test facility (HTDF). RELAP5 MOD 4.0 is utilized to model the DRACS thermal performance and predict the heat removal capability of the LTDF and HTDF. The code results are benchmarked with the experimental data obtained from the LTDF, which uses water as a surrogate for both primary and secondary fluids in DRACS. Two transient scenarios have been investigated, including a startup scenario and a pump trip scenario. Comparisons between the experimental data and simulation results from the RELAP5 show good agreement. The HTDF uses FLiNaK and KF-ZrF4, respectively, as the working fluids for the scaled down DRACS primary and secondary loops. Fluid properties of FLiNaK and KF-ZrF4 have been implemented into the RELAP5 code. Startup and pump trip scenarios in the HTDF are also simulated by the code.
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The RELAP5/MOD3 5m5 code is assessed using the L9-1/L3-3 test carried out in the LOFT facility, a 1/60-scaled experimental reactor, simulating a loss of feedwater accident with multiple failures and the sequentially-induced small break loss-of-coolant accident. The code predictability is evaluated for the four separated sub-periods with respect to the system response; initial heatup phase, spray and power operated relief valve (PORV) cycling phase, blowdown phase and recovery phase. Based on the comparisons of the results from the calculation with the experiment data, it is shown that the overall thermal-hydraulic behavior important to the scenario such as a heat removal between the primary side and the secondary side and a system depressurization can be well-predicted and that the code could be applied to the full-scale nuclear power plant for an anticipated transient with multiple failures within a reasonable accuracy. The minor discrepancies between the prediction and the experiment are identified in reactor scram time, post-scram behavior in the initial heatup phase, excessive heatup rate in the cycling phase, insufficient energy convected out the PORV under the hot leg stratified condition in the saturated blowdown phase and void distribution in secondary side in the recovery phase. This may come from the code uncertainties in predicting the spray mass flow rate, the associated condensation in pressurizer and junction fluid density under stratified condition.
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Languages : en
Pages : 14
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The capability of the RELAP5-3D{copyright} computer code to perform multi-dimensional thermal-hydraulic analysis was assessed using data from steady-state flow tests conducted at Rensselaer Polytechnic Institute (RPI). The RPI data were taken in a two-dimensional test section in a low-pressure air/water loop. The test section consisted of a thin vertical channel that simulated a two-dimensional slice through the core of a pressurized water reactor. Single-phase and two-phase flows were supplied to the test section in an asymmetric manner to generate a two-dimensional flow field. A traversing gamma densitometer was used to measure void fraction at many locations in the test section. High speed photographs provided information on the flow patterns and flow regimes. The RPI test section was modeled using the multi-dimensional component in RELAP5-3D Version BF06. Calculations of three RPI experiments were performed. The flow regimes predicted by the base code were in poor agreement with those observed in the tests. The two-phase regions were observed to be in the bubbly and slug flow regimes in the test. However, nearly all of the junctions in the horizontal direction were calculated to be in the stratified flow regime because of the relatively low velocities in that direction. As a result, the void fraction predictions were also in poor agreement with the measured values. Significantly improved results were obtained in sensitivity calculations with a modified version of the code that prevented the horizontal junctions from entering the stratified flow regime. These results indicate that the code's logic in the determination of flow regimes in a multi-dimensional component must be improved. The results of the sensitivity calculations also indicate that RELAP5-3D will provide a significant multi-dimensional hydraulic analysis capability once the flow regime prediction is improved.
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The RELAP5-3D thermal-hydraulic code was assessed against Japanese Safety Experiment Loop (SEL) and Heat Transfer Loop (HTL) tests. These tests were chosen because the phenomena present are applicable to analyses of Russian RBMK reactor designs. The assessment cases included parallel channel flow fluctuation tests at reduced and normal water levels, a channel inlet pipe rupture test, and a high power, density wave oscillation test. The results showed that RELAP5-3D has the capability to adequately represent these RBMK-related phenomena.
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Languages : en
Pages : 5
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The Simplified Boiling Water Reactor (SBWR) proposed by General Electric (GE) is an advanced light water reactor (ALWR) design that utilizes passive safety systems. The PCCS is a series of heat exchangers submerged in water and open to the containment. Since the containment is inerted with nitrogen during normal operation, the PCCS must condense the steam in the presence of noncondensable gases during an accident. To model the transient behavior of the SBWR with a system code, the code should properly simulate the expected phenomena. To validate the applicability of RELAP5 MOD 3.1.1, the data from three Phase 1, Step 2 nitrogen venting tests at Toshiba's Gravity-Driven Integral Full-Height Test for Passive Heat Removal facility and RELAP5 calculations of these tests were compared. The comparison of the GIRAFFE data against the results from the RELAP5 calculations showed that it can predict condensation and gas purging phenomena occurring in the long-term decay heat rejection phase. In this phase of the transient, condensation in the PCCS is the only means to reject heat from the SBWR containment. In the two tests where the nitrogen purge vent line was at its deepest submergence in the Suppression Pool (SIP), the RELAP5 results mirrored the behavior of the containment pressures and of the water levels in the Horizontal Vent (HV) and the nitrogen purge line tube of the GIRAFFE data. However, in the test with the shallowest purge line submergence, there was appreciable direct contact condensation on the pool surface of the HV despite modeling efforts to deter these phenomena. This surface condensation, unobserved in the GIRAFFE tests, was a major cause of RELAP5 predicting early containment depressurization and the subsequent early rise in HV and nitrogen purge line water levels. The present RELAP5 MOD3.1.1 interfacial heat and mass transfer model does not properly degrade direct contact steam condensation in the presence of noncondensable gases sitting on a pool.
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Languages : en
Pages : 124
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The RELAP5/MOD3.2 computer program has been used to analyze a series of tests investigating heat-transfer from a partly uncovered VVER-1000 core in the KS test facility at the Russian Research Center ''Kurchatov Institute'' (RRC-KI). The analysis documented represents VVER Standard Problem 4 in Joint Project 6, which is the investigation of Computer Code Validation for Transient Analysis of RBMK and VVER Reactors, between the United States and Russian International Nuclear Safety Centers. The experiment facility and data, RELAP5 nodalization, and results are shown for one of the six tests defined in Standard Problem 4. Only part of the data was analyzed due to our conclusion that the available experimental data is not sufficient to allow the modeling of the actual experiment sequence. The experiment initial conditions were reached through a series of transient processes, about which no quantitative information was available. This has required the modeling of an arbitrary computational transient, with the goal of reaching initial conditions similar to those observed during the experiment. The use of an arbitrary transient introduces many degrees of freedom in the analysis, i.e. initial computational values that influence the entire sequence of events, including the loop behavior during the experiment time window. Reasonable agreement between RELAP5 and the experiment data can be obtained by manipulating a number of initial computational values, including the liquid level in the fuel assembly model, the liquid level in the annular region, the quality of the saturated vapor in the voided loop regions, etc. Our study has focused on exploring the sensitivity of results to changes in these initial values which are not based on experimental information, but are selected with the goal of matching the experimentally observed behavior during the experiment time window. We have shown that changes in several initial arbitrary values can lead to similar changes in the calculated loop behavior. Other parameters exist which can directly influence the calculated results, which have been only minimally explored, or have not been explored at all due to lack of time. Because we are not modeling the entire sequence of events which led to the initial experiment conditions, it is not possible to assume that by matching a limited set of measured parameters at the beginning of the experiment time window we have reproduced all the initial actual conditions. We conclude that the analysis of these experiments, while providing a useful demonstration of the RELAP5 capabilities to describe the experimental sequence of events, cannot be used to draw quantitative conclusions about the adequacy of specific RELAP5 models.
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In this paper, the details of the development of a demonstration problem to assess the reliability of a passive safety system are presented. An advanced small modular reactor (advSMR) design, which is a pool-type sodium fast reactor (SFR) coupled with a passive reactor cavity cooling system (RCCS) is described. The RELAP5-3D models of the advSMR and RCCS that will be used to simulate a long-term station blackout (SBO) accident scenario are presented. Proposed benchmarking techniques for both the reactor and the RCCS are discussed, which includes utilization of experimental results from the Natural convection Shutdown heat removal Test Facility (NSTF) at the Argonne National Laboratory. Details of how mechanistic methods, specifically the Reliability Method for Passive Systems (RMPS) approach, will be utilized to determine passive system reliability are presented. The results of this mechanistic analysis will ultimately be compared to results from dynamic methods in future work. This work is part of an ongoing project at Argonne to demonstrate methodologies for assessing passive system reliability.
