• Nuclear Engineering and Technology
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• v.27 no.5
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• pp.701-709
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• 1995

The realistic discharge coefficient for the critical How model of RELAP5/AOD3/KAERI are determined for the subcooled and too-phase critical flow by assessments of nine MARVIKEN Critical flew Test(CFT). The selected test runs include a high initial subcooling and large nozzle aspect rat-io(L/D). The code assessment results show that RELAP5/MOD3/KAERI over-predicts the subcooled critical flow and under-predicts the two-phase critical flow. Using these result, the realistic discharge coefficients of critical flow models are quantified by an iterative method. The realistic discharge coefficients are determined to be 0.89 for the subcooled critical How and 1.07 for the two-phase critical flow, and the associated standard deviations are 0.0349 and 0.1189, respectively. The results obtained from this study can be applied to calculate the realistic system response of Large Break Loss of Coolant Accident and to evaluate the realistic Emergency Core Cooling System performance.

• Nuclear Engineering and Technology
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• v.56 no.5
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• pp.1902-1912
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• 2024

This paper presents results from system-level modeling of a water-based reactor cavity cooling system using RELAP5-3D. The computational model is benchmarked with experimental data from a half-scale RCCS test facility at Argonne National Laboratory. The model prediction is first compared with a two-phase oscillatory baseline experimental case where mixed accuracy is obtained. The model shows reasonable prediction of mass flow rate, pressure, and temperature but significant overprediction of void fraction. The model prediction is then compared with a fault case where the inlet of the risers is gradually reduced using a throttling valve. As the valve is closed, the model is able to predict some major flow phenomena observed in the experiment such as the dampening of oscillations, the reintroduction of oscillations, as well as boiling, flashing, and geysering in the risers. However, the timeline of these events are not well captured by the model. The model is also used to investigate the evolution of flow regime in the chimney. This work highlights that the semi-empirical constitutive relations used in RELAP-3D could have a strong influence on the accuracy of the model in two-phase oscillatory flows.

• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1277-1288
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• 2018

In this work, the reactor kinetics capability is used to compute the design safety parameters in a PWR due to complete loss of coolant flow during protected and unprotected accidents. A thermal-hydraulic code coupled with a point reactor kinetic model are used for these calculations; where kinetics parameters have been developed from the neutronic SRAC code to provide inputs to RELAP5-3D code to calculate parameters related to safety and guarantee that they meet the regulatory requirements. In RELAP5-3D the reactivity feedback is computed by both separable and tabular models. The results show the importance of the reactivity feedback on calculating the power which is the key parameter that controls the clad and fuel temperatures to maintain them below their melting point and therefore prevent core melt. In addition, extending modeling capability from separable to tabular model has nonremarkable influence on calculated safety parameters.

• Nuclear Engineering and Technology
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• v.26 no.2
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• pp.265-276
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• 1994

Several improvements to the RELAP5/MOD3 reflood model hate been made. These improvement were made to correct deficiencies in the reflood model identified by the assessment of the RELAP5/MOD3 code against FLECHT-SEASET experiments. The improvements consist of modification of reflood wall heat transfer package and adjusting the droplet size in dispersed flow regime. The time smoothing of wall vaporization and level tracking of transition flow are also added to eliminate the pressure spikes and level oscillation during reflood process. Assessment of the improved model against FLECHT-SEASET experimental data and application of LBLOCA analysis for plant shows that the deficiencies have been corrected.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.159.1-159
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.591-598
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• 1997

The COBRA/RELAPS code, an integrated version of the COBRA-TF and RELAP5/MOD3 codes, has been developed for the realistic simulations of complicated, multi-dimensional, two-phase, thermal-hydraulic system transients in light water reactors. Recently, KAERA developed an unified version of the COBRA/RELAP5 code, which can run in serial mode on both workstations and personal computers. This paper provides the brief overview of the code integration scheme, the recent code modifications, the developmental assessments, and the future development plan.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.403-410
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• 1995

A loss of the residual heat removal (RHR) accident during mid-loop operation of Yong-gwang Nuclear Units 3/4 was analyzed using the RELAP5/MOD3.1.2 code. In this work the following assumptions are used; (i) initially the reactor coolant system (RCS) above the hot leg center line is filled with nitrogen gas, (ii) two 3/4-inch diameter vent valves on the reactor vessel head and the top of pressurizer in the reactor coolant system are always open, and a level indicator is connected to the RMR suction line, (iii) the two steam generators are in wet layup status and the steam generator atmospheric dump valve assemblies are removed so that the secondary side pressure remains at nearly atmospheric condition throughout the accident, and (iv) the loss of RHR is presumed to occur at 48 hours after reactor shutdown. Findings from the RELAP5 calculations are (i) the core boiling begins at ∼5 min, (ii) the peak RCS pressure is ∼3.0 bar, which implies a possibility of temporary seal break, (iii) ∼94 % of the decay heat is removed by reflux condensation in the steam generator U-tubes in spite of the presence of noncondensable gas, (iv) the core uncovery time is evaluated to be 7.2 hours. Significant mass errors were observed in the calculations.

• Nuclear Engineering and Technology
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• v.26 no.3
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• pp.355-366
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• 1994

The USNRC issued a revised ECCS rule that allows the use of best estimate computer codes for safety analysis. The rule also requires an estimation of uncertainty in calculated system response when applying the best estimate computer codes. A practical realistic evaluation methodology to evaluate the ECCS performance that satisfies the requirements of the ECCS rule has been developed and this paper describes the application of new realistic evaluation methodology to large break LOCA for, the demonstration of the new methodology. The computer code RELAP5/MOD3/KAERI, which was improved from RELAP5/MOD3.1, was used as the best estimate code in the application. The uncertainty of the code was evaluated by assessing several separate and integral effect tests, and for the application to actual plant Kori 3 & 4 was selected as the reference plant. Response surfaces for blowdown and reflood PCTs were generated from the results of the sensitivity analyses and probability distribution functions were established by random sampling or Monte-Carlo method for each response surface. Final uncertainties were quantified at 95% probability level and safety margins for large break LOCA were discussed.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.97-109
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• 2022

Multi-dimensional coupling analysis is a research hot spot in nuclear reactor thermal hydraulic study and both the full-scale system transient response and local key three-dimensional thermal hydraulic phenomenon could be obtained simultaneously, which can achieve the balance between efficiency and accuracy in the numerical simulation of nuclear reactor. A one-dimensional to three-dimensional (1D-3D) coupling platform for the nuclear reactor multi-dimensional analysis is developed by XJTU-NuTheL (Nuclear Thermal-hydraulic Laboratory at Xi'an Jiaotong University) based on the CFD code Fluent and system code RELAP5 through the Dynamic Link Library (DLL) technology and Fluent user-defined functions (UDF). In this paper, the International Standard Problem (ISP) No. 43 is selected as the benchmark and the rapid boron dilution transient in the nuclear reactor is studied with the coupling code. The code validation is conducted first and the numerical simulation results show good agreement with the experimental data. The three-dimensional flow and temperature fields in the downcomer are analyzed in detail during the transient scenarios. The strong reverse flow is observed beneath the inlet cold leg, causing the de-borated water slug to mainly diffuse in the circumferential direction. The deviations between the experimental data and the transients predicted by the coupling code are also discussed.

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.2626-2631
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• 2007

Air-water two phase natural circulation flow in the T-HERMES (Thermo-Hydraulic Evaluation of Reactor cooling Mechanism by External Self-induced flow)-1D experiment has been evaluated to verify and evaluate the experimental results by using the RELAP5/MOD3 computer code. The RELAP5 results have shown that an increase in the coolant inlet area leads to an increase in the water circulation mass flow rate. However, the water outlet area does not effective on the water circulation mass flow rate. As the coolant outlet moves to a lower position, the water circulation mass flow rate decreases. The water level is not effective on the water circulation mass flow rate. As the height increases in the air injection part, the void fraction increases. However, the void fraction in the upper part of the air injector maintains a constant value. An increase in the air injection mass flow rate leads to an increase in the local void fraction, but it is not effective on the local pressure.