Analysis of LSC phenomena of ATLAS cold leg SBLOCA tests using MARS-KS code (original) (raw)
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Progress in Nuclear Energy, 2014
Currently Appendix K of 10 CFR 50 is used in the United States to evaluate models for the emergency cooling systems of light-water reactors. To assure that these models are accurate enough to ensure that the cooling systems are satisfactory, code scalability, applicability and uncertainty methodologies are used to evaluate the uncertainty in system analysis code predictions due to the various models. One cornerstone of this methodology is the development of the Phenomenon Identification and Ranking Table (PIRT), which summarizes the thermal-hydraulic phenomenon associate with a particular accident scenario and ranks their importance in determining the effectiveness of core cooling and the parts of the accident for which the phenomenon may be important. In this paper the PIRT developed by the Institute for Nuclear Safety Systems for a small-break LOCA with loss of high-pressure emergency coolant injection is analyzed in detail and several modifications are proposed based on a mechanistic understanding of the phenomenon involved. The resulting PIRT should provide a more accurate guide for model evaluation and development in advanced thermal-hydraulic system analysis codes.
International Journal of Energy Production and Management, 2019
Thermal-hydraulic response of a reacTor core following large break loss-of-coolanT accidenT under flow blockage condiTion young seok bang & Joosuk lee korea institute of nuclear safety. absTracT since the revision of the requirements to consider the effect of fuel burnup on emergency core cooling system performance was proposed, flow blockage in reactor core has been one of the important issues in the thermal-hydraulic analysis of loss-of-coolant accident (loca). The present paper describes how much flow blockage would be expected following a large break loca based on the actual nuclear design data including the power and burnup of the fuel rods. a system thermal-hydraulic code, mars-ks, is used for calculation where the burnup specific data of the fuel rods is supported by a fuel performance code, fracon3. To recover the weakness of the system code in which the flow blockage under multiple rods configuration cannot be automatically simulated in hydraulic calculation, a special modelling scheme is developed and applied to the calculation. The effect of flow blockage on the thermal-hydraulic response of the reactor core is also discussed. To compensate for the uncertainty of the present flow blockage model, additional calculations are attempted for a wide range of the level of blockage.
Annals of Nuclear Energy, 2015
The thermal-hydraulic analysis using MARS-KS code was performed for 6-inch cold leg break test of ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation), which was the second domestic standard problem. The calculation results were compared with experimental data to assess the code capability to simulate the transient thermal-hydraulic behavior for small break loss of coolant accident (SBLOCA). The sequence of events, except for the location of loop seal clearing (LSC) and safety injection tank (SIT) injection time was predicted well. The loop seals of 1A and 2B intermediate legs were cleared at 398 s in the experiment, while that of 1A was only cleared in the calculation at the same time. The prediction showed good agreement with the experimental data for pressurizer pressure and break mass flow rate. The sudden decrease and increase of water level at the LSC time were predicted qualitatively. After LSC, there was significant water level dip at SIT injection time which was not seen in the experiment. In addition, sensitivity study to investigate the cause of core level dip at SIT injection time was performed and discussions were made for it. In conclusion, MARS-KS code has good capabilities to simulate cold leg break SBLOCA, however, including interfacial heat and mass transfer, especially condensation model needs to be improved to predict more accurate results.
Acta mechanica Malaysia, 2022
In this paper, a comparative analysis of loss of coolant accident (LOCA) in hot leg and cold leg of primary circuit in a VVER 1200 nuclear power plant is investigated. The effect of break size on the severity of the accident is observed. The break size was varied in the range 200-11350 cm 2. For all the accident scenarios, station blackout (SBO) condition is set up. Additionally, it is assumed that no ECCS (Emergency Core Cooling System) is available due to system malfunction. The whole scenario is simulated in PCTRAN (Personal Computer Transient Analyzer) software. Results reveal that with the increase in the size of the break area, the core uncovering time decreases sharply. However, for a break size of 2800 cm 2 or smaller, the water level in the core doesn't drop to zero, indicating that the core is partially uncovered throughout the accident scenario. In case of hot leg LOCA, the draining of the reactor vessel is observed to be more rapid compared to cold leg LOCA, while the core melting started earlier in case of cold leg.
2021
Korea Atomic Energy Research Institute (KAERI) has been operating an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), with reference to the APR1400 (Advanced Power Reactor 1400) for experiments for transient and design basis accidents (DBAs) simulation as shown in Figure 1 [1]. In addition, KAERI has operating the domestic standard problem (DSP) program based on the experimental data from the selected experiments in order to encourage the verification and validation of system codes. Recently sixth phase of the DSP (DSP-06) blind calculation had been proceeded, the DSP-06 aims at evaluating the physical behavior during a small break loss of coolant accident with the loss of safety injections (SBLOCA +LSI) using various system codes. In this study, SBLOCA analysis is performed using MARS-KS 1.5 [2] with improved model.
Simulation of VVER-1000 Guillotine Large Break Loss of Coolant Accident Using RELAP5/SCDAPSIM/MOD3.5
Journal of Nuclear Engineering, 2021
The safety performance of nuclear power plants (NPPs) is a very important factor in evaluating nuclear energy sustainability. Safety analysis of passive and active safety systems have a positive influence on reactor transient mitigation. One of the common transients is primary coolant leg rupture. This study focused on guillotine large break loss of coolant (LB-LOCA) in one of the reactor vessels, in which cold leg rupture occurred, after establishment of a steady-state condition for the VVER-1000. The reactor responses and performance of emergence core cooling systems (ECCSs) were investigated. The main safety margin considered during this simulation was to check the maximum value of the clad surface temperature, and it was then compared with the design licensing limit of 1474 K. The calculations of event progression used the engineering-level RELAP5/SCDAPSIM/MOD3.5 thermal-hydraulic program, which also provide a more detailed treatment of coolant system thermal hydraulics and core...
Annals of Nuclear Energy, 2015
Thermal-hydraulic behavior for a complete blockage of a single fuel channel in a generic 10 MW research reactor is studied by using the system analysis code RELAP5/MOD3.3 which is widely used in the nuclear industry. Fuel assembly geometry is lumped into a 4 channel model to model high and average power cases which are spatially discretized. Various axial power shapes coming from different control rods positions are considered in the analysis, where the minimum wall subcooled margin is found to exist for case with highest peaking for an average powered channel blockage transient. Vapor generation is observed from first and second highest peaking cases where cyclic variation of vapor inventory inside a blocked channel resulted in oscillatory behavior of the fuel temperature. Effect of a presence of an oxide layer is also tested which showed a slight increase in structure temperatures and vapor generation. Point kinetics model is utilized in the analysis code to observe the effect of reactivity feedback and consequences from different application ranges are compared. Analysis shows a consideration of assembly wise feedback results in increased feedback effect and decreased boiling which deviate from single channel wise feedback case. This calls for a detailed multi-dimensional simulation with neutronics and thermalhydraulics simultaneously considered. Analyses results show that the consideration of feedback improves the outcome in terms of fuel temperature, and its integrity is conserved for all test cases.
RELAP5 Modeling of a Siphon Break Effect on the Brazilian Multipurpose Reactor
Brazilian Journal of Radiation Sciences
This work presents the thermal-hydraulic simulation of the Brazilian Multipurpose Reactor (RMB) using the RELAP5/Mod3 code. The RMB will provide Brazil with a fundamental infrastructure for the national development on activities of the nuclear sector in the areas of social, strategic, industrial applications and scientific and technological development. A RELAP5/Mod3 code model was developed for thermal-hydraulic simulation of the RMB to analyze the phenomenology of the Siphon Breakers device (four flap valves in the cold leg and one open tube for the atmosphere in the hot leg) during a Loss of Coolant Accident (LOCA) at different points in the primary circuit. The Siphon Breaker device is an important passive safety system for research reactors in order to guarantee the water level in the core under accidental conditions. Different simulations were carried out at different location in the Core Cooling System (CCS) of the RMB, for example: LOCA before the CCS pumps with and without pump trip and LOCA after the CCS pumps and the heat exchanger. In all RELAP5/Mod3 code simulations, the Siphon Breaker device's performance after a LOCA was effective to allow enough air to enter the outlet pipe of the CCS in order to break the siphon effect and preventing the pool level from reaching the riser (chimney) and the RMB core discovering. In all cases, the reactor pool level stabilized at about 5.5 m after the end of the LOCA simulation and the fuel elements were kept underwater and cooled.
2020
In an open pool type reactor, the pool water inventory should act as a heat sink to provide emergency reactor core cooling. In the Brazilian Multipurpose Reactor – RMB, to avoid the loss of pool water inventory, all the Core Cooling System (CCS) lines penetrate at the pool top, far above the reactor core level. However, as most of CCS equipment and lines are located below the reactor core level, in the case of a Loss of Coolant Accident (LOCA), a large amount of pool water could be lost drained by siphon effect. To avoid RMB research reactor core discovering in the case of a LOCA, siphon breakers, that allow CCS line air intake, are installed in the CCS lines in order to stop the reactor pool draining due to siphon effect. As siphon breakers are important passive safety devices, their effectiveness should be verified. Several previous numerical and experimental studies about siphon break effect were found in the literature. Some of them comment about the effectiveness of the siphon ...
2007
Best-estimate thermal-hydraulic system codes are widely used to perform safety and licensing analyses of nuclear power plants and also used in the design of advance reactors. Evaluation of the capabilities and the performance of these codes can be accomplished by comparing the code predictions with measured experimental data obtained on different test facilities. OECD/NEA Commit-tee on the Safety of Nuclear Installations (CSNI) has promoted, over the last twenty-nine years, some forty-eight international standard problems (ISPs). These ISPs were performed in different fields as in-vessel thermal-hydraulic behaviour, fuel behaviour under accident conditions, fission product release and transport, core/concrete interactions, hydrogen distribution and mixing, containment thermal-hydraulic behaviour. 80 % of these ISPs were related to the working domain of principal working group no.2 on coolant system behaviour (PWG2) and were one of the major PWG2 activities for many years. A global r...