Break Loss Of Coolant Accident Research Articles

Scaling analysis of nuclear power plant (APR1400) utilizing ATLAS integral effect test result of A5.1 (Counterpart test for LSTF 1% cold-leg break LOCA)

ATLAS is the thermal hydraulic integral effect test loop that can simulate the same pressure and temperature condition with nuclear power plant. It was designed with a height ratio of 1/2 and a volume ratio of 1/288 with reference to APR1400. The counterpart test against LSTF which was named as A5.1 test was performed within the OECD/NEA ATLAS international joint research project to compare major thermal hydraulic phenomena between two test facilities. The main scenario for the A5.1 test is the LSTF SB-CL-32 test. The SB-CL-32 test is a 1% horizontal small break loss of coolant accident (SBLOCA) at the cold leg with cold leg injection of the emergency core cooling system (ECCS) and accident management (AM) action.From the counterpart test results, the significant differences were found in the viewpoint of the thermal–hydraulic phenomena. It was concluded that the different result were caused by the different design characteristic of each test facility because they referred the different nuclear power plant. Thus, the applicability of the ATLAS test results to its reference nuclear power plant was evaluated by utilizing ATLAS A5.1 test results which was the counterpart test with the LSTF.The transient scenario and initial and boundary conditions of the A5.1 test were adopted to APR1400 after applying scaling ratios of ATLAS. The MARS-KS code was utilized to analyzed thermal hydraulic behaviors of APR1400 and the analysis results were compared with the scaled A5.1 test result. From the result of comparison, it was concluded that the ATLAS test results predicted the thermal hydraulic behavior of the nuclear power plant well. Thus, it was confirmed that ATLAS is a well-designed integral effect test facility which is designed with an appropriate scaling method.

Multi-physics analysis of SBLOCA by coupled Neutronics/Thermal-Hydraulics full plant model

With the development of computational tools, the accident analysis would be performed by best estimate approaches. The employment of multi-physics codes would be considered as new calculation procedure to obtain high valuable results by improving calculations with specialized codes. The small break loss of coolant accident (SBLOCA) is an accident that has typically been analyzed using system codes with a limited number of Thermal-Hydraulics (TH) channels in the core region, neglecting the use of multi-physics calculations. This approach results in insufficient accuracy and resolution in both neutronics and TH calculations within the core zone. To address this issue, this study proposes the implementation of a Neutronics/Thermal-Hydraulics coupling with one-to-one mapping in the core zone, achieved through the development of a comprehensive plant TH model. This approach aims to enhance the accuracy and resolution of calculations, enabling the identification of local phenomena, parameters, and feedback during accident simulations, and the incorporation of these effects on the macroscopic cross-sections at each time step. The results of SBLOCA simulation reveals the effects of conservative/best estimate approach on the sequence of events, the initiation time of the safety systems, the critical controlled parameters of the power plant, and the precise understanding of the accident. Also, it was found that the radial power distribution becomes disrupted and loses its symmetry immediately after the SCRAM leading to different characteristics of steam generators (SGs) and natural circulation flow rate in each loop of primary side.

Fractional scaling analysis of cladding temperature in large break loss of coolant accident

In integral test facilities, some phenomena relevant in the separate effect test are usually neglected due to scaling limitations and design compromises. The typical heat transfer phenomenon in reactor core in large break loss of coolant accident (LBLOCA) scenario is analyzed with fractional scaling analysis (FSA) method in this work. A conservation form of cladding temperature is derived on the basis of energy conservation. Then the scaling criteria and their importance are evaluated. It is found that the similarities of the criteria for convection heat transfer and heat accumulation should be guaranteed since their priorities are high. The contribution of initial pressure and distribution of coolant inventory to cladding temperature in LBLOCA scenario is too slight. This work is encouraging and may shed a light on the coupling application of hierarchical two-tiered scaling (H2TS) and FSA method in engineering application.

Transient Analysis of LBLOCA with Subsequent Loss of Offsite Power and Loss of All AC Power in VVER-1200 Reactor

The generation III+ VVER-1200 (AES-2006) reactor is equipped with improved safety features with component redundancy compared to the previous models. This paper is concerned with the performance of active and passive safety features during the transient of a major beyond design basis accident (BDBA) scenario of the reactor. The study is done considering large break loss of coolant accident (LB LOCA) with loss of offsite power (LOOP) and LB LOCA with loss of all AC power (LOACP) sources namely station blackout (SBO) using VVER-1200 NPP simulator. It was found that the reactor is capable of mitigating such BDBA scenarios by the effectiveness of the active and passive safety systems. Moreover, LB LOCA with LOACP is observed to be more severe compared to LBLOCA with LOOP. This comparative analysis is compatible with IAEA safety guidelines.

Application of uncertainty and sensitivity analysis in dose assessment during a postulated LBLOCA for VVER-1000 nuclear reactor

A key element in the process of accident analysis for the event of a severe nuclear emergency such as the large break loss of coolant accident (LBLOCA) is to properly quantify safety margins using reliable tools. Assessing the consequences of radioactive material release and the equivalant dose after the LBLOCA can provide clarity about the accident. Reliable safety analysis methods for nuclear power plants (NPPs) can be used to obtain relatively realistic outcomes considering the complex uncertainties of NPP events. In this study, an appropriate technique for sensitivity and uncertainty analysis is employed to develop a methodology for performing dose assessment within a postulated LBLOCA including uncertainties. Considering the approved reliability over the years and safety aspects, VVER reactor is among the most widely used nuclear reactors, which is selected as the case study in this article. The main problem that the study aims to solve is to illustrate whether the inordinate deviation of the input parameters from their reference values lead to violation of the allowed TEDE value by conducting uncertainty analysis. For performing calculations of total effective dose equivalent (TEDE), the environmental contamination with radioactive materials has been simulated using HotSpot code and the obtained results were compared with the solutions of JRODOS program. To obtain the most influential parameters on the obtained TEDE values in different distances from the NPP, an appropriate sensitivity analysis has been carried out, following which the Wilks’ statistical method is used to generate 59 data samples with uncertainty propagation and 95 % confidence. The resulting uncertainty bands demonstrate that the maximum TEDE values of calculated data lie within about 7 to 15 mSv in the postulated LBLOCA scenario, which are below the threshold limit of 50 mSv recommended by International Atomic Energy Agency.

SBLOCA analysis of a floating nuclear reactor under ocean conditions using MARS-KS moving reactor model

With increasing demands to reduce greenhouse gas emissions from the maritime industry, nuclear power has emerged as a potential solution. Among these options, offshore floating reactors have been demonstrated and developed by various organizations and countries. The offshore floating reactors offer advantages such as a long refueling period for enhanced safety and high mobility, enabling access to remote areas. However, when operating on a moving platform, offshore floating reactors are exposed to external forces that can affect the thermal-hydraulic properties within the reactor core and various hydraulic components. To address the safety concerns associated with offshore floating reactors, evaluating their thermal-hydraulic performance under different ocean motions is necessary. Therefore, this study focused on conducting accident analyses in an offshore floating reactor called BANDI-60, specifically investigating a Small Break Loss of Coolant Accident caused by a double-ended guillotine break in the Direct Vessel Injection line using the system analysis code MARS-KS. Various ocean conditions were considered, including static inclination, combined inclination with heaving, and rolling motion. Among these conditions, static inclination, especially when the break was aligned downward with the inclination direction, was identified as the most significant factor affecting the integrity of the reactor core.

Safety analysis of pressurized water reactors with annular fuel during different transients

The dual-cooled annular fuel has two channels, with a shorter heat transfer path and a larger heat transfer area than solid fuel, which can effectively improve the safety of the reactor. It has excellent potential to improve the economy of the pressurized water reactor (PWR) and is considered to have a good development prospect. In this paper, the safety of the reactor with annular fuel assemblies under different transient conditions is analyzed using the system code NUSOL-SYS and the subchannel code NACAF, with the CPR1000 as the reference power plant. The analysis mainly includes the large break loss of coolant accident (LBLOCA), main steam line break accident (MSLB), loss of flow accident (LOFA), and rod ejection accident (REA) under nominal power and higher power density. The flow blockage accident and the combined accidents of LBLOCA and the flow blockage accident are also evaluated. The results show that during LBLOCA, the peak cladding temperature (PCT) of the core is only 833 K, much lower than that of the solid fuel. In addition, the departure from nucleate boiling ratio (DNBR) of annular fuel is larger during MSLB and LOFA, which means a more negligible risk of fuel damage. In REA, the average temperature of annular fuel is below 900 K, and the average enthalpy is below 200 kJ/kg, which is much lower than that of solid fuel. The results at 150 % power show that the performance of the core with annular fuel is still better than that of solid fuel at nominal power, except that the PCT during LBLOCA is slightly higher than solid fuel at nominal power. During the flow blockage accident, the results with different blockage areas show that a small proportion of blockage of the inner channel does not have a significant effect on the cladding temperature and that there is a risk of rupture when the inner channel is completely blocked. Finally, the results of the assumed combined accident show that the blockage of the inner channel before scram will significantly increase the PCT of the core and the time to complete the reflood during LBLOCA.

Thermal-Hydraulic comparative analysis of G-III + BWR and PWR passive safety features under Loss of coolant accident

As one of symbolic design features of Generation-III and III+ (G-III and G-III + ) nuclear reactors, the passive safety system has been implemented on majority of G-III and III + PWRs and BWRs. The passive safety systems in those reactors are designed differently yet adhere to identical fundamental principles. It would be interesting to examine the similarities and differences between the passive safety systems of G-III + PWR and BWR. System code simulations for Small Break and Large Break Loss of Coolant Accident for the Economic Simplified Boiling Water Reactor (ESBWR) and China’s advanced PWR had been conducted, respectively. The results showed that the passive safety systems of ESBWR and China’s Advanced PWR were comparable in design, and when dealing with the similar type of SBLOCA and LBLOCA, the passive safety components with the same functionality could all contribute to the core's cooling with stability and reliability, and the injection flow rate values were nearly equivalent. Separating the injection lines of ESBWR's safety systems guarantees that the action sequence could well remain stable. This design could prevent the injection flow rate and action sequence from being impacted by the interaction of safety injection systems. This work could serve as a reference for the design of subsequent new reactors, thereby contributing to the design of reactors with enhanced economics and safety.

Optimization strategy for SAM in nuclear power plants based on NSGA-II

Abstract The Severe Accident Management Guide (SAMG) is an important component of nuclear safety regulations. Many studies are being conducted to optimize severe accident management (SAM) strategies. To ensure the safety of nuclear power plants, decision makers need to monitor multiple parameters with security threats. Therefore, it is particularly important to search optimal SAM strategies under different numbers of mitigation targets. The Non-dominated Sorting Genetic Algorithm-II (NSGA-II) is an evolutionary algorithm that does not require derivative differentiation and is capable of population search. In this study, a nuclear power plant accident optimization strategy is developed using the Modular Accident Analysis Program (MAAP) in conjunction with NSGA-II. The strategy enables decision makers to consider multiple mitigation objectives in a complex decision environment. Focusing on the CPR1000, this study applies the optimization strategy to automatically search for optimal mitigation strategies for small break loss of coolant accident (SBLOCA) and station blackout hot leg creep rupture accidents (SBOHLCR). Comparing the optimization results with the basic accident sequence, it is found that the reactor pressure vessel (RPV) failure time is delayed from 72,702 s to 128,730 s under SBLOCA and from 23,828 s to 28,363 s under SBOHLCR. This study has also verified that the optimal SAM strategy obtained by the strategy through dual objective optimization has better mitigation effects than a strategy that only considers one objective. This optimization strategy has the potential to be applied to other types of severe accident management studies in the future.

3D Analysis of Hydrogen Distribution and Its Mitigation Using Passive Autocatalytic Recombiners (PARs) Inside VVER-1000 Containment

Hydrogen is a flammable gas that can generate thermal and mechanical loads which could jeopardise the containment integrity upon combustion inside nuclear power plants containment. Hydrogen can be generated from various sources and disperses into the containment atmosphere, mixing with steam and air following a loss of coolant accident and its progression. Therefore, the volumetric hydrogen concentration should be examined within the containment to determine whether a flammable mixture is formed or not. Codes with 3D capabilities could serve this examination by providing detailed contours/maps of the hydrogen distribution inside containment in view of the local stratification phenomenon. In this study, a 3D VVER-1000 as-built containment model was sketched in AutoCAD and then processed into GOTHIC nuclear containment analysis code for hydrogen evaluation. The model was modified to a great extent by installing 80 passive autocatalytic recombiners and locating hydrogen sources to evaluate the performance of the hydrogen removal system inside the containment on maintaining the hydrogen concentration below the flammability limit during a large break loss of coolant accident. 2D profiles and 3D contours of volumetric hydrogen concentration with and without PARs are presented as the simulation outcome of this study. The results were validated against the results of the Final Safety Analysis Report, which also demonstrates the effectiveness of the hydrogen removal system as an engineered safety feature to keep the containment within a safe margin. Detailed 3D contours of hydrogen distribution inside containment can be employed to evaluate the local hot spots of hydrogen, rearranging and optimising the number and location of PARs to avoid the hydrogen explosion inside containment.

Investigating the effect of spray system on the containment condition of VVER1000/V446 NPP during LBLOCA and TLOFW accidents

The containment system serves as the last barrier against the release of fission products. However, over pressurization or hydrogen detonation can compromise its effectiveness, as evidenced by the Fukushima nuclear power plant (NPP) disaster. This study describes the efficacy of the spray system in reducingcontainment pressure andtemperature, as well as,itsimpacton hydrogen mitigation during large break loss of coolant accident (LBLOCA) and total loss of feed water (TLOFW) accident in the VVER-1000/V446NPP. The evaluation is conductedusing theMELCOR 1.8.6code.The results of the LBLOCA analysis demonstrate that the spray system significantly reduces the pressure and temperature of the containment. It also enhances hydrogen mitigation by increasing the amount of recombined hydrogen by around 4% and reducing the onset time of hydrogen recombination by 90,000 s. In the TLOFW accident scenario, the spray system is effective in reducing the containment pressure from 0.41 MPa to 0.27 MPa. The investigation of hydrogen behavior in the reactor containment atmosphere reveals that the simultaneous operation of the spray and the hydrogen recombiner systems is the most effective strategy for hydrogen mitigation. However, activating the spray system without the passive autocatalytic recombiners (PARs) function can lead to an increase in the hydrogen mole fraction in all containment compartments, thereby increasing the likelihood of hydrogen combustion.

Safety considerations related to boron redistribution during Loss-of-Coolant-Accident initiated beyond design basis events for the NuScale power module

The authors, as staff at the U.S. Nuclear Regulatory Commission (NRC), evaluated the potential consequences of boron dilution under beyond design basis conditions in the NuScale Power Module to inform NRC staff’s safety evaluation report for the NuScale design certification review. The authors identified the small break loss of coolant accident with coincident failure of control rods to insert as the scenario with the most adverse boron dilution and reactivity insertion consequences. The authors evaluated the sequence of events for the scenario to determine which conditions or phenomena could produce the potential for a core flow surge to yield core damage because of a super-prompt reactivity excursion. The authors screened the conditions and phenomena that could potentially cause a core flow surge based on the dynamic speed of the incursion compared to key reactivity feedback mechanisms: thermal–hydraulic coupling via rod heat transfer and convective mixing. Based on the screening process, the authors were able to identify two processes that could potentially lead to a power excursion. However, the authors’ analysis relies heavily on assumptions regarding convective mixing, making this an important phenomenon worth careful consideration in any evaluation of boron redistribution.

Uncertainty Assessment in the Prediction of Selected Fission Products in the Case of a Postulated Severe Accident Scenario at VVER-1000

The aim of the paper is to present the results and insights gained from the analysis of the postulated severe accident scenario for the Nuclear Power Plant (NPP) equipped with a VVER-1000, focusing on the uncertainty assessment in the prediction of selected Fission Products (FPs). To this end, the simulation of the Large Break Loss of Coolant Accident (LB LOCA) along with a Station Blackout (SBO) scenario was carried out. The capabilities of the ASTECv2.2.0.1 [1] severe accident (SA) code along with the uncertainty quantification tool – SUNSET [2] in the coupled mode was used. The analysis was carried out for the eight input variables and ten output variables. The uncertainty quantification method incorporated within the SUNSET statistical tool was used in order to conduct the present analysis. Based on the Wilks’ formula, the number of ASTEC code runs was determined. The main statistical data are obtained for the selected fission products as figures of merit. The main steps of the analysis are outlined. Furthermore, a brief overview of the SUNSET/ASTECv2.2.0.1 coupling procedures [3] is also presented and discussed. This work demonstrates the application of the uncertainty quantification methodology in the analyses of the severe accidents in the Nuclear Power Plants and provides the specific recommendations of its use. The main outcomes from the assessment provided will be useful and may be a base for the future analyses related to the evaluation of the safety of the VVER-1000.

Uncertainty and sensitivity analysis of iodine release in severe accidents of advanced pressurized water reactors based on the Latin Hypercube method and Grey Correlation Coefficients

Based on an HPR1000 pressurized water reactor (PWR), a large break loss of coolant accident (LBLOCA) model was created in ASTEC software to simulate several accident scenarios. By varying the initial conditions, the release of iodine was evaluated to analyse the influencing factors. It was shown that the break size contributes to the release rate of iodine, while in-vessel retention (IVR) and the containment spray system (CSS) play significant roles in inhibiting the release of iodine into the environment, and more than 90% of the iodine is distributed in the containment when a break occurs in a hot leg or surge pipeline. With the use of the Latin hypercube method, grey correlation coefficients were calculated to quantitatively estimate the input parameters. The results showed that the main factors affecting iodine release vary in the different phases, and the temperature of the cooling water in external vessel reactor cooling (EVRC) displays the highest correlation coefficients, with values greater than 0.85. Overall, IVR is an important measure to prevent iodine release into containment and environmental settings.

The large-break LOCA uncertainty analysis in a VVER-1000 reactor using TRACE and DAKOTA

An uncertainty analysis is inevitable in simulating reactor performance, particularly in the accident analysis for the estimation of safety borders. In this paper, the application of the uncertainty analysis of a Large Break Loss of Coolant Accident (LBLOCA) from main coolant pipeline at the reactor inlet and outlet, and also pressurizer (PRZ) connecting pipe are presented. First, the model developed in TRACE is verified in steady-state conditions and LBLOCA using the Final Safety Analysis Report (FSAR) of Bushehr nuclear power plant. By the verified model, the uncertainty analysis is made respecting uncertainty in boundary conditions. Finally, the analysis is achieved using the surrogate models and Wilks’ method in DAKOTA. The 95/95 uncertainty band of the peak cladding temperature (PCT) and main reactor parameters are obtained and analyzed. The results indicate that the PCT uncertainty in LBLOCA at the reactor inlet is large, the PCT varies in the range of 910 to 970 °C. The drop temperature time of fuel cladding is in the range of 270 to 600 s for LBLOCA at the reactor inlet. The PCT uncertainty in LBLOCA at the reactor outlet and PRZ connecting pipe is low.

Multiphysics analysis of fuel fragmentation, relocation, and dispersal susceptibility–Part 3: Thermal hydraulic evaluation of large break LOCA under high-burnup conditions

Increasing the peak rod average burnup of pressurized water reactor (PWR) fuel beyond 62 GWd/tU may increase fuel fragmentation, relocation, and dispersal (FFRD) susceptibility during a large break loss of coolant accident (LBLOCA). TRACE thermal hydraulic (TH) LBLOCA analyses were performed for a realistic 24-month high-burnup PWR equilibrium cycle, to inform subsequent transient BISON high-burnup FFRD susceptibility evaluations. Realistic LBLOCA systems behavior was first established by configuring to and comparing with the BEMUSE OECD LBLOCA benchmark. Fuel and operating conditions were then applied from high-burnup VERA depletion calculations. LBLOCA simulations were performed for 281 selected high-burnup rods, for which transient TH boundary conditions were collected for later use in BISON. The TRACE results indicated that rod linear heat rate (rather than burnup) is the main predictor of peak cladding temperature (PCT) during the event. PCT typically occurred at a local burnup lower than the rod-average burnup, especially for twice-burned fuel.

The effects of cladding thermo-physical and thermo-chemical properties on the coping time during a PWR unmitigated LB-LOCA

To evaluate quantitatively the effects of individual thermo-physical and thermo-chemical cladding properties on the coping time, an unmitigated large break loss of coolant accident (LB-LOCA) was simulated using Modular Accident Analysis Program (MAAP) code. For this purpose, hypothetical claddings were created to decompose the influence of each property enhancement. The variations in the cladding properties could not affect the main conditions of the reactor coolant system such as pressure, reactor core water level, etc., but led significantly different variations in the cladding temperature. When the coping time was defined as the time elapsed between departure from normal operation and the moment at which cladding melts, the specific heat capacity (enthalpy) and melting temperature of cladding materials were identified as properties extending the coping time most. The improved oxidation resistance could extend moderately the coping time. On the other hand, the cladding thermal conductivity had no effect on the coping time.

The modeling of passive containment cooling system and sensitivity analysis using the SPRUCE code

Passive Containment Cooling Systems (PCCSs) have attracted wide attention to improve the inherent safety of nuclear power plants. This study summarizes a system analysis method of a closed-loop natural circulation of the PCCS via separately evaporation and condensation. A lumped-parameter model is built using SPRUCE, a code of system simulation of severe accidents, to investigate the natural circulation characteristics and heat-removal efficiency of the PCCS. The analysis of heat-removal capacity of the PCCS is carried out under typically thermal conditions of the containment, such as pressure from 0.28 to 0.4 MPa, and the results are validated by comparison with the experimental data obtained from a single-loop testing facility of the PCCS in full scale. The SPRUCE model agrees with the data, despite slightly underestimating the rate of condensation heat transfer. It conservatively satisfies the requirement of design capacity. To further verify the effectiveness of mitigation on containment overpressure, a representative scenario of LBLOCA (Large Break Loss of Coolant Accident) is implemented using the code. Meanwhile, the effects of non-condensable gas are also presented, as a reference for optimization of PCCS design.

Investigation of the source term and radiological consequences of small break LOCA with failure of the emergency core cooling system in Bushehr nuclear power plant

In the present study, the in-containment source term derivation, environmental dispersion and deposition as well as radiological consequences of Bushehr Nuclear Power Plant (BNPP) due to Small Break Loss of Coolant Accident (SB-LOCA) have been studied. First, the calculation of in-containment release and distribution of the source terms due to the severe accident of SB-LOCA with failure of the active part of the Emergency Core Cooling System (ECCS) has been carried out. In this calculation, two scenarios, including with and without operator control actions, have been considered. Then, the environmental leakage of radionuclides as well as their dispersion and deposition considering two meteorological scenarios have been investigated. Consequently, public exposure has been calculated for each meteorological scenario and the obtained results have been reported and discussed. For two accidents with the same meteorological conditions, at a distance of 400 m from the release point, the received dose in the first 24 h is 18 and 342 mSv without and with operator action, respectively. The results show that due to the intervention of the operator and the use of emergency cooling supply systems, more water enters the reactor. Therefore, if the operator's actions do not prevent the core from melting, more source terms have entered the containment after the core melts and worse conditions than the baseline scenario without operator action occurred.

Assessment of spray system capabilities in reduction of containment pressure and deposition of fission products during a severe accident in VVER-1000

In this study, we have investigated the spray system's capabilities to reduce fission products in the containment atmosphere and control the containment pressure in a VVER-1000 nuclear power plant. Large Break Loss of Coolant Accident (LB-LOCA) without access to the active part of the Emergency Core Cooling System (ECCS), as a worst-case severe accident, has been investigated as a case study. Only containment is simulated via the MELCOR code, while leakage of various materials from the nuclear power plant's core and the primary circuit is considered as the boundary conditions. Following the Fukushima accident in 2011, many nuclear safety regulatory organizations focused on using portable equipment that is essential for decay heat removal in nuclear power plants during severe accidents. Based on the literature studies, the accident management scenarios have changed recently since severe accidents inside nuclear power plants have worried many scientists and the public. Evaluating the current spray system in a severe condition based on which this system was not designed is vital and novel. We considered that the spray system would be fully restored and perform its safety function using the existing portable equipment to depressurize the containment and deposition of the fission products. To calculate the maximum allowed delay time of the spray system recovery, the accident was modeled without the operation of the spray system while a safety criterion was considered to avoid reaching the ultimate pressure of containment. For a more detailed evaluation of the spray system, two different operating modes including the activation of one spray train and the activation of both spray trains have been selected. According to the obtained results, if the spray system is revived before 4,000s from the beginning of the accident, this system would be able to reasonably mitigate the accident consequences. Moreover, without the need for any external water supply source for the spray system, this system can be permanently fed from the containment sump for at least 800,000s after the initiation of the accident, requiring the relevant heat exchanger of the spray system to be recovered for this period.