MELCOR Analysis Research Articles

Numerical Study of Natural Circulation Flow in Reactor Coolant System during a Severe Accident

The rupturing of steam generator tubes leads to serious accidents in nuclear power plants. It causes radioactive materials to leak into the secondary system and release outside the reactor containment region. Therefore, it is important to model a technique to determine whether the natural circulation within a reactor coolant system (RCS) can cause rupture. In this study, a computational fluid dynamics (CFD) analysis methodology was incorporated as a first step to establish an RCS natural circulation evaluation technique to generate RCS natural circulation input parameters for the MELCOR analysis of thermally induced steam generator tube rupture (TI-SGTR) in nuclear power plants. Benchmarking tests were conducted against existing experimental studies; the results demonstrated a difference of 9.4% or less between the experimental and CFD analysis results with respect to the main evaluation factors. Subsequently, a steam generator tube simplification modeling technique was established for application to nuclear power plants, and CFD analysis was conducted to determine its applicability. The CFD analysis results revealed that when numerous tubes are simplified into one equivalent tube, the thermal flow characteristics generated in the RCS could be distorted. The findings of this research are expected to be helpful in understanding the thermal flow characteristics of natural circulation in the RCS. Further, the findings may potentially serve as a foundation for future CFD analysis research related to the natural circulation in the RCS of nuclear power plants.

Analysis of primary side bleed and feed actions for severe accident management following total loss of feed water in a Swedish PWR

Severe accident management guidelines (SAMGs) have been developed and applied in nuclear power plants (NPPs) to provide the fourth level of defense worldwide since the TMI-2 accident which occurred in 1979. The primary objective of the SAMGs is to protect remaining safety boundaries and to limit release of fission products into the environment by specific severe accident management (SAM) actions with available safety equipment. This paper is concerned with the assessment of the primary side bleed and feed (PBF) actions in the SAMGs of a Swedish pressurized water reactor (PWR). The PBF actions are realized by operations of emergency injection pumps and power-operated relief valves (PORVs), following a severe accident induced by an event of total loss of feed water (TLOFW) and temporary station blackout. In the present study, MELCOR simulations are performed to investigate the effects of the SAM actions on severe accident progression and consequences. Different sets of available equipment and actuation time due to recovery of AC power are simulated. The interest of the MELCOR analysis is focused on how these uncertain parameters of SAM actions affect thermal–hydraulic response, hydrogen generation, core relocation, reactor pressure vessel failure and release of fission products. The results demonstrate the effectiveness of the PBF strategy to prevent RPV failure in the TLOFW accident and the effects of injection timing on accident consequences. The obtained insights are instrumental for validation and improvement of the strategies and actions in the plant specific SAMGs.

OPR1000 TI-SGTR 사고 MELCOR 해석의 자연순환유동 입력인자 생산을 위한 CFD 해석

We performed a 3-dimensional CFD analysis of a natural convective flow in the steam generator (SG) inlet plenum of an optimized power reactor 1000 MWe (OPR1000) to generate input parameters of the MELCOR code, a 1-dimensional lumped parameter code, which will be used in analyzing the temperature induced steam generator tube rupture (TI-SGTR) accident. The input parameters used in the MELCOR calculation will account for a 3-dimensional effect in the natural convective flow in the SG inlet plenum. To produce a credible CFD analysis result, we first established a CFD analysis methodology, which is developed from validation analyses against the steam generator design data of the OPR1000 and the Westinghouse 1/7 scaled-down test data, for the thermal mixing flow calculation between the SG and the reactor pressure vessel in the OPR1000. Though this CFD analysis, we produced the MELCOR input parameters, the recirculation ratio 1.96, thermal mixing fraction 0.79, discharge coefficient 0.13, and hot tube fraction 0.40, with an error range of approximately 10% to 25%. The input parameters will be used in the MELCOR analysis to evaluate a radioactive material release from broken SG tubes during the TI-SGTR accident.

CSPACE for a simulation of core damage progression during severe accidents

CSPACE (Core meltdown, Safety and Performance Analysis CodE for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling of verified system thermal hydraulic code of SPACE (Safety and Performance Analysis CodE for nuclear power plants) and core damage progression code of COMPASS (Core Meltdown Progression Accident Simulation Software). SPACE is responsible for the description of fluid state in nuclear system nodes, while COMPASS is responsible for the prediction of thermal and mechanical responses of core fuels and reactor vessel heat structures. New heat transfer models to each phase of the fluid, flow blockage, corium behavior in the lower head are added to COMPASS. Then, an interface module for the data transfer between two codes was developed to enable coupling. An implicit coupling scheme of wall heat transfer was applied to prevent fluid temperature oscillation. To validate the performance of newly developed code CSPACE, we analyzed typical severe accident scenarios for OPR1000 (Optimized Power Reactor 1000), which were initiated from large break loss of coolant accident, small break loss of coolant accident, and station black out accident. The results including thermal hydraulic behavior of RCS, core damage progression, hydrogen generation, corium behavior in the lower head, reactor vessel failure were reasonable and consistent. We demonstrate that CSPACE provides a good platform for the prediction of severe accident progression by detailed review of analysis results and a qualitative comparison with the results of previous MELCOR analysis.

Estimation of the fuel debris thermal energy at the time of the major core slumping of Fukushima Daiichi Nuclear Power Plant Unit-3 with MELCOR-2.2

To provide supportive information to understand the current debris status in Fukushima Daiichi Nuclear Power Plant Unit-3, sensitivity analyses have been carried out with MELCOR-2.2 for two scenarios, with/without direct leakage from the Reactor Pressure Vessel (RPV) to the Drywell (D/W). The particular focus of the analyses is the estimated debris thermal energy up to and at the time of the major core slumping event, which may be the key to determine the following debris cooling and failure mechanism of the lower head of the RPV. In the analysis, the debris relocation velocity from the core region to the RPV lower head (VFALL), the number of Safety Relief Valves (SRVs) opening, and the amount of core slumping at the major RPV pressure peak were tuned so that the plant data of RPV and PCV pressure from ca. 9:00 to 12:00 March 13th 2011could be reproduced in each scenario. Best-estimate case conditions are summarized for the with/without direct leakage from RPV to D/W scenarios. As a result, as a consideration common to both scenarios, MELCOR analysis tends to estimate significant core oxidation before Automatic Depressurization System (ADS) actuation, and concomitantly estimates little core oxidation from the time after ADS to the time of the major RPV pressure peak. In addition, although a remarkable difference was found in the amount of hydrogen generated in RPV during the major core slumping between the best-estimate with/without leakage cases, limited difference can be observed for the core oxidation heat generation during the major core slumping. This is because the early Zircaloy oxidation prior to ADS actuation in the current MELCOR modelling resulted that the hydrogen generation source during the major core slumping at 12:00 March 13th was primarily from oxidation of stainless steel, which was not as exothermic as that of Zr oxidation.

MELCOR Analysis of a SPARC Experiment for Spray-PAR Interaction during a Hydrogen Release

A series of experiments were performed in the SPARC (spray-aerosol-recombiner-combustion) test facility to simulate a hydrogen mitigation system with the actuation of a PAR (passive auto-catalytic re-combiner) and spray system. In this study, the SPARC-SPRAY-PAR (SSP1) experiment is chosen to benchmark the MELCOR (a lumped-parameter code for severe accident analysis) predictions against test data. For this purpose, firstly we prepared the base input model of the SPARC test vessel, and tested it by a simple verification problem with well-defined boundary conditions. The implementation of a currently used PAR correlation in MELCOR is shown to be appropriate for the simulation of a PAR actuation experiment. In an SSP1 experiment, the PAR is reacting with hydrogen, and the spray actuation starts as soon as hydrogen injection is complete. The MELCOR simulation well predicts the pressure behavior and the gas flow affected by operating both a PAR and spray system. However, the local hydrogen concentration measurement near the inlet nozzle is much higher than the volume average-value by MELCOR, since high jet flow from the nozzle is dispersed in the corresponding cell volume. The experimental reproduction of the phenomena we expect, or, conversely, the identification of phenomena we do not understand, will continue to support the verification of analytical models using experimental data and to analyze the impact of spray on PAR operations in severe accident conditions.

Potential Recovery Actions from a Severe Accident in a PWR: MELCOR Analysis of a Station Blackout Scenario

Accident-tolerant fuel (ATF) cladding materials have been a focus of recent work to provide a greater resistance to fuel degradation, oxidation, and melting in light water reactors for beyond-design accident scenarios such as a station blackout (SBO). In a previous study, researchers at The University of Wisconsin–Madison used the Surry Nuclear Plant as the pilot plant to examine the effect of ATF substitute clad materials with the short-term SBO as the postulated accident, examining the effect of a loss of auxiliary feedwater (AFW) with the MELCOR systems code. In this work, we examine the effect of recovery actions for an SBO in Surry as a follow-on topic. Specifically, we selected two kinds of core cladding materials (Zircaloy and FeCrAl), and then conducted comparative analysis of the effect of water injection; first with a delay in water injection start times into the reactor pressure vessel (RPV) and then with steam generator (SG) steam-side AFW end times. We find that alternative cladding materials (FeCrAl) can effectively delay fuel degradation and system failures for both water injection strategies. One finds that RPV water injection can prevent such severe accident effects if restored in a few hours into the SBO. Conversely, SG steam-side AFW flow with alternative cladding materials (FeCrAl) can delay the fuel degradation and system failure processes by hours. We mainly focus on analyzing the severe accident progression by different quantitative signals, such as the onset of rapid hydrogen production, hot-leg creep rupture failure, and core slump. Analyses are now underway to consider the effects of proposed coating materials on Zircaloy cladding and if such coatings can afford similar benefits.

MELCOR Analysis of OSU Multi-Application Small Light Water Reactor (MASLWR) Experiment

A Multi-Application Small Light Water Reactor (MASLWR) conceptual design was developed by Oregon State University (OSU) with emphasis on passive safety systems. The passive containment safety system employs condensation and natural circulation to achieve the necessary heat removal from the containment in case of postulated accidents. Containment condensation experiments at the MASLWR test facility at OSU are modeled and analyzed with MELCOR, a system-level reactor accident analysis computer code. The analysis assesses its ability to predict condensation heat transfer in the presence of noncondensable gas for accidents where high-energy steam is released into the containment. This work demonstrates MELCOR’s ability to predict the pressure-temperature response of the scaled containment. Our analysis indicates that the heat removal rates are underestimated in the experiment due to the limited locations of the thermocouples and applies corrections to these measurements by conducting integral energy analyses along with computational fluid dynamics simulation for confirmation. The corrected heat removal rate measurements and the MELCOR predictions on the heat removal rate from the containment show good agreement with the experimental data.

Safety analysis methodology for Chinshan nuclear power plant spent fuel pool under Fukushima-like accident condition

Abstract Chinshan nuclear power plant (NPP), a BWR/4 plant, is the first NPP in Taiwan. After Fukushima NPP disaster occurred, there is more concern for the safety of NPPs in Taiwan. Therefore, in order to estimate the safety of Chinshan NPP spent fuel pool (SFP), by using TRACE, MELCOR, CFD, and FRAPTRAN codes, INER (Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.) performed the safety analysis of Chinshan NPP SFP. There were two main steps in this research. The first step was the establishment of Chinshan NPP SFP models. And the transient analysis under the SFP cooling system failure condition (Fukushima-like accident) was performed. In addition, the sensitive study of the time point for water spray was also performed. The next step was the fuel rod performance analysis by using FRAPTRAN and TRACE's results. Finally, the animation model of Chinshan NPP SFP was presented by using the animation function of SNAP with MELCOR analysis results.

Safety research on fusion DEMO in Japan: Toward development of safety strategy of a water-cooled DEMO

This paper reports the current status of a safety research on water-cooled fusion DEMO in Japan. A basic strategy of development of the safety guidelines is described for DEMO based on a water-cooled solid pebble bed blanket. Clarification of safety features of the DEMO in accident situations is a key issue to develop the guidelines. Recent achievements in understanding of the safety features of the water-cooled DEMO are reported. The MELCOR analysis has clarified the temperature histories of the DEMO components in a complete loss of decay heat removal event. The transient behavior of the first wall temperature is found to be essentially different from that of ITER. The pressure load to the tokamak cooling water system vault (TCWSV) is analyzed based on a simple model equation of the energy conservation. If the amount of the primary coolant is the same as that of Slim-CS, the previous small Japanese DEMO, the discharged water does not damage the TCWSV with the volume and pressure-tightness similar to those of pressurized light water reactors. It is shown that implementation of a pressure suppression system to the small TCWSV is effective to suppress the pressure load to the second confinement barrier.

Severe accident analysis of plant-specific spent fuel pool to support a SFP risk and accident management

The Fukushima accident that happened on March 11, 2011 has raised the possibility of severe accidents in an SFP under beyond design basis external events (BDBEEs) that might lead to a loss of all safety functions and need a way to cope effectively with such an event through a SFP risk and accident management. From a SFP safety point of view, the low decay heat of the fuel assemblies and large water inventory in an SFP may make the accident processes slow compared to an accident in the reactor core, but a huge number of fuel assemblies stored inside it and no containment in a SFP building (especially for PWR types) might be exposed to much greater risk. A quantitative analysis for SFP accidents can give more insights into which aspects play dominant roles in an accident progression. To date, however, few studies have been made for severe plant-level SFP accidents. This paper provides the plant-specific MELCOR analysis results for potential severe accidents that can be expected in a typical PWR. The SFP of the reference plant is located adjacent to the primary containment (outside primary containment), and its bottom is placed above the plant grade. Three representative accident scenarios (loss of cooling accident, loss of coolant inventory, and complete loss of coolant accident) and two different reactor operating modes (normal operation and refuelling mode) have been considered for the analysis. From the risk and accident management point of view, phenomenological analysis was mainly focused on (a) the level of the coolant, (b) the cladding temperature, (c) the mass of hydrogen generated during accident progressions, and (d) MCCI (Molten Corium Concrete Interaction).

シビアアクシデント時の格納容器貫通部リークパスでのFP エアロゾル捕集効果(II)<br/>貫通部での除染係数と実機への適用

Fission product (FP) aerosol trapping along the leakage paths of degraded containment penetrations during a severe accident was investigated. Prior to this aerosol trapping test, several actual electrical penetration assemblies (EPAs) and hatch flange gaskets were degraded under prototypical severe accident conditions in a reactor containment. Then, the aerosol trapping test at the leakage path was conducted. Cesium iodide (CsI) was used as a FP aerosol in the test. The aerosol trapping effect along the leakage paths was expressed as a decontamination factor (DF) and the average values were about 740 for the low-voltage module and about 14 for the flange gaskets. The overall DF value of a typical BWR was estimated at about 450, using the average DF value and a number of containment penetrations of each type. These results were reflected in the severe accident analysis by the MELCOR code. The total amount of FP leaked from containment was reduced to around 50 to 70% compared with that determined by the original MELCOR analysis 1 day after a typical BWR severe accident.

MELCOR Analysis of Steam Generator Tube Creep Rupture in Station Blackout Severe Accident

A pressurized water reactor steam generator tube rupture (SGTR) is of concern because it represents a bypass of the containment for radioactive materials to the environment. In a station blackout accident, tube integrity could be threatened by creep rupture, particularly if cracks are present in the tube walls. Methods are developed herein to improve assessment capabilities for SGTR by using the severe-accident code MELCOR. Best-estimate assumptions based on recent research and computational fluid dynamics calculations are applied in the MELCOR analysis to simulate two-dimensional natural circulation and to determine the relative creep-rupture timing in the reactor coolant pressure boundary components. A new method is developed to estimate the steam generator (SG) hottest tube wall temperature and the tube critical crack size for the SG tubes to fail first. The critical crack size for SG tubes to fail first is estimated to be 20% of the wall thickness larger than by a previous analysis. Sensitivity studies show that the failure sequence would change if some assumptions are modified. In particular, the uncertainty in the countercurrent flow limit model could reverse the failure sequence of the SG tubes and surge line.

SOMBRERO LOVA Analysis Using CFC NB31 Oxidation Data

Carbon fiber composites (CFCs) are often suggested as armor material for the first wall of a fusion plasma chamber because of carbon’s low atomic number, high thermal conductivity, and high melting point. However, carbon is chemically reactive in air and readily absorbs tritium. Accordingly, it is believed that during a loss-of-vacuum accident (LOVA), the CFC armor will react with the air ingress and release its absorbed tritium. The mobilization of this tritium and the carbon monoxide produced by the CFC-air chemical reaction are both safety concerns. This paper discusses the MELCOR thermal-hydraulic analysis of a simulated LOVA for the SOMBRERO fusion design. The MELCOR analysis is important because it included data from recent oxidation experiments that studied the advanced CFC NB31. A previous MELCOR analysis of a simulated SOMBRERO LOVA event suggested that the ingress of air would aggressively oxidize the CFC. While the current analysis revealed initial first-wall temperatures that exceed those of the prior analyses, the trend reversed 10 h after the onset of the LOVA. The calculated wall temperatures at the back of the blanket for the current analysis were consistently lower than those previously calculated using the older data. Accordingly, the conclusion is that a LOVA event for a fusion design similar to SOMBRERO may not be as grave as once predicted.

Numerical Analysis on Ingress of Coolant Event in Vacuum Vessel of Fusion Reactor Using Modified TRAC-BF1

The semi three-dimensional numerical analysis model of the FDR-ITER to simulate the in-vessel LOCA or the ICE was constructed using the modified TRAC-BF1 code in which the vacuum vessel, the pressure suppression tank and the relief pipe header are modeled using one VESSEL component. Analytical results obtained from the modified TRAC-BF1 code were compared with those from the MELCOR code concerning pressure in the vacuum vessel when a cooling pipe of 0.6 m2 flow area breaks in it. The mass flow rate of the injected water into the vacuum vessel and the initial wall temperature in the modified TRAC-BF1 input data were the same as those calculated in the MELCOR analysis. The maximum pressure in the vacuum vessel obtained from the modified TRAC-BF1 code is 10% higher than that from the MELCOR code, but it is still below the design pressure 0.5 MPa of the FDR-ITER vacuum vessel. The semi three-dimensional analysis using the modified TRAC-BF1 obtained that the maximum delay in opening time among the four rupture-disks is within 3 ms. The maximum pressure in the vacuum vessel does not exceed the design pressure, even if one of rupture-disks is not opened.

Numerical Analysis on Ingress of Coolant Event in Vacuum Vessel of Fusion Reactor Using Modified TRAC-BF1.

The semi three-dimensional numerical analysis model of the FDR-ITER to simulate the in-vessel LOCA or the ICE was constructed using the modified TRAC-BF1 code in which the vacuum vessel, the pressure suppression tank and the relief pipe header are modeled using one VESSEL component. Analytical results obtained from the modified TRAC-BF1 code were compared with those from the MELCOR code concerning pressure in the vacuum vessel when a cooling pipe of 0.6 m2 flow area breaks in it. The mass flow rate of the injected water into the vacuum vessel and the initial wall temperature in the modified TRAC-BF1 input data were the same as those calculated in the MELCOR analysis. The maximum pressure in the vacuum vessel obtained from the modified TRAC-BF1 code is 10% higher than that from the MELCOR code, but it is still below the design pressure 0.5 MPa of the FDR-ITER vacuum vessel. The semi three-dimensional analysis using the modified TRAC-BF1 obtained that the maximum delay in opening time among the four rupture-disks is within 3 ms. The maximum pressure in the vacuum vessel does not exceed the design pressure, even if one of rupture-disks is not opened.

MELCOR Analysis of the Three Mile Island Unit 2 Accident

The MELCOR computer code has been used to analyze the first 174 min of the Three Mile Island Unit 2 (TMI-2) accident. MELCOR is being developed at Sandia National Laboratories for the U.S Nuclear R...