Small Break Loss Of Coolant Accident Research Articles

Study on the effects of accident-tolerant fuels on the safety of CPR1000 nuclear power plants

Abstract In this paper, CPR1000 Nuclear power plant was taken as the reference unit. According to CPR1000 Probabilistic Safety Analysis (PSA), the Large Break loss of coolant accident (LOCA), Medium Break LOCA, Small Break LOCA, Station Blackout (SBO), Total Loss of Feed water (TLOFW) and Anticipated transient without trip (ATWT) with Loss of Main Feed water (LOMF) were selected as the representative accident scenarios. The house holding deterministic programs LOCUST were used to analyze each of the above selected scenarios for five different accident tolerant fuels (ATF). The effects on accident process, reactor core damage time, system success criteria and personnel response time were analyzed under the above scenarios. The above results for the five ATF materials were compared to those for standard UO2-Zr in the same scenarios, then some useful conclusions were obtained which can be used to support ATF material selection. In the end, based on the deterministic calculations, the effects of different ATF materials on CPR1000 PSA results such as core damage frequency (CDF) were evaluated in detail with traditional PSA software Risk Spectrum. According to the results, ATFs have some advantages on core damage time, success criteria or operators’ action time window. But the CDF was changed little compared to standard UO2-Zr system. Through the studies, it is found the most important thing is that due to the much longer time windows some new plant modifications can be provided to mitigate the accidents which can contribute to the plant safety and decrease risk more.

Improvement of macroscopic turbulence model for subchannel analysis in the rod bundle array

The PRIUS program was established to generate an experimental database for the 6 × 4 and 12 × 6 rod bundle geometry. The database will be used to address the subchannel and CFD code analysis required for modeling and validation. This is necessary because Small Break Loss of Coolant Accident (SBLOCA) and Intermediate Break Loss of Coolant Accident (IBLOCA) present three-dimensional phenomena in the core due to the radial power profile, crossflow, and diffusion-dispersion. Therefore, specific experimental programs are required, especially during core reflooding, to investigate the large-scale three-dimensional effects. However, validating each sensitive model of the code separately in the presence of 3D effects is not possible due to the inability to implement instrumentation at high pressure and temperature steam-water flow conditions. The PRIUS test program uses a single-phase flow test to simulate a non-homogeneous velocity distribution and provide information on crossflow with radial mixing effects between subchannels. The CUPID code, which uses a macroscopic turbulence model, has been validated using the PRIUS-II experimental database. Existing macroscopic turbulence models were also validated for their prediction capabilities with different inlet flow conditions. However, the validation revealed significant errors in the shear region between subchannels. An improved macroscopic turbulence model showed promising results in predicting turbulence kinetic energy in porous media analysis.

Thermal safety characteristics analysis of helium xenon gas cooled small reactor system under LOCA accidents

The thermal safety characteristics of the system after a small break loss of coolant accident (SB-LOCA) in a helium xenon gas-cooled small reactor will determine whether the reactor will experience a meltdown and the duration of the meltdown. At present, research on SB-LOCA both domestically and internationally is mostly focused on water-cooled reactors, lacking relevant research on helium xenon gas-cooled reactors (He-Xe GCR). Therefore, it is of great significance to research the thermal safety characteristics of He-Xe GCR after a breach loss of coolant accident. This article improves the gas turbine model based on the original SB-LOCA analysis program for pressurized water reactors. And added a calculation model for the characteristics of helium xenon mixed gas, a heat transfer model for helium xenon mixed airflow, and a compressor model. Finally, a Brayton Cycle Reactor System Analysis program (BRESA) was developed for SB-LOCA analysis of He-Xe GCR. Explore the impact of the size and location of fractures on the thermal safety characteristics of the system using BRESA. Research has found that as the size of the rupture gradually increases, the decrease in reactor core power is greater, and the reactor power is shut down earlier. Comparative analysis of the impact of different breach positions on the residual heat removal characteristics reveals that a breach accident on the high-pressure side of the system poses a greater threat to system safety. Therefore, when setting the safety threshold for the Brayton cycle system, the focus should be on the hazards of high-voltage side breaches. This research achievement assists with the safe operation strategy of He-Xe GCR in the future.

Study on the lower head failure in severe accidents Part II: Thermal-mechanical coupling behavior analysis on CAP1400 reactor lower head

Under severe accident conditions with core meltdown, large amounts of molten material relocate to the lower head at high temperatures. Significant thermal and mechanical loads may lead to failure of the reactor pressure vessel (RPV) lower head, spreading radioactive fission products. Lower head failure time and location are crucial for accident mitigation and consequence assessment. This study developed a thermodynamic failure analysis model for the lower head by investigating material failure mechanisms, coupling this into an Integrated Severe Accident Analysis code (ISAA), and conducting experimental validation and applied research. Part I mainly introduces the thermodynamic analysis module development, validation against OLHF experiments, and preliminary failure criteria analysis. Results indicate the model accurately predicts lower head failure behavior, and Larson-Miller life fraction and Hedl-Dorn parameter criteria accurately assess lower head integrity. Part II presents applied research after validation and preliminary analysis. Using the improved ISAA, melt pool behavior and lower head thermal–mechanical response were analyzed for a CAP1400 small break loss-of-coolant accident (SBLOCA), assuming multiple safety system failures. Results indicate the bottom of the lower head melt pool didn’t form a hard shell due to high heat release rate, only an oxide shell layer around the second layer. Decay heat from melt pool components cannot be dissipated, leading to lower head failure. Further External Reactor Vessel Cooling (ERVC) heat transfer model analysis and research are needed in ISAA. The two developed mechanical analysis models predict material failure at high temperatures, and the timing and location of lower head failure are consistent, at approximately 16,000 s at 25.38°±5.47° on the lower head. The predicted failure mode is similar to Part I validation, indicating both Larson-Miller life fraction and Hedl-Dorn parameter criteria provide consistent lower head integrity judgments.

Analysis of AP1000 Small-Break Loss-of-Coolant Accident Using Reactor Transient Simulator

The Westinghouse Electric Company’s Advanced Passive Reactor (AP1000) is characterized by the incorporation of passive safety systems (PSSs) designed to ensure core cooling during transient events. The assessment of PSSs requires evaluation of their performance through a combination of experiments and simulations employing various thermal-hydraulic codes. In addition, detailed evaluation of PSSs for a specific reactor system transient analysis such as loss-of-coolant-accident analysis supports understanding representative integral effects test facility development and the further evolution model development and assessment process. Developing a reactor system code is a complex and time-consuming process that requires significant engineering expertise and effort. It can take several months to even years to complete in the early stages of reactor system design and analysis. However, this process can be expedited through the use of transient simulator models for similar reactor systems, which can be used for lesson learning and training purposes. This study uses the Personal Computer Transient Analyzer (PCTRAN) code. The main advantage of PCTRAN is its ease of use and ability to run faster than real time. This study presents the results obtained for a small-break loss-of-coolant accident (SBLOCA) for two breaks using the full version (licensed) of PCTRAN. The purpose of this investigation is to evaluate the overall system behavior during the postulated SBLOCA event as well as assess the capability of the PCTRAN code to reproduce the system response during transient events. The obtained results were compared with the Westinghouse NOTRUMP system code. The PCTRAN code proved to be reliable in predicting the qualitative behavior of the system in both transient cases. As for the system response, it was found that it is contingent on the activation time of the PSSs. The differences in reactor coolant system pressure between the two codes were attributed to the critical flow model and simplification of mass and energy balance. Despite PCTRAN’s limitations, it can still provide a reasonable prediction of various reactor parameters such as pressure, mass flow rate, and void fraction during a SBLOCA scenario. It is worth noting that PCTRAN currently employs a bulk approach similar to that of the Modular Accident Analysis Program (MAAP) and MELCOR codes. However, the upcoming version of PCTRAN will include an artificial intelligence–based detection and accident prevention system, as well as different models for different reactor components. Consequently, PCTRAN has the potential to be upgraded to match the system thermal-hydraulic codes of the U.S. Nuclear Regulatory Commission and become more widely used in cybersecurity to safeguard nuclear power plants from cyberattacks.

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.

Design and analysis of KA-130, the marine-based SMR for icebreakers

Designed for icebreakers, the KA-130 integral SMR provides a carbon-free solution for navigating the North Pole Route with nuclear propulsion. As an integral reactor, all primary components are integrated into the reactor pressure vessel, eliminating large-break loss-of-coolant accidents and enhancing reactor safety. The study evaluates its load following capability and demonstrates efficient load changes within two minutes. The safety analysis during small break loss-of-coolant accidents (SBLOCA) incorporates ship movements into the minimum critical heat flux ratio (MCHFR) calculations and includes a sensitivity analysis to identify significant scenarios. The most limiting case was the guillotine break at the safety injection line with low initial pressure, high core exit temperature and bottom-peaked core. The analysis shows that core flooding is maintained, and decay heat removal is successfully performed. Acceptance criteria, aligned with regulatory and KA-130 specific standards, ensure safety during and after SBLOCA, meeting MCHFR and collapsed water level conditions above the core.

Evaluating direct vessel injection accident-event progression of AP1000 and key figures of merit to support the design and development of water-cooled small modular reactors

The passive safety systems (PSSs) within water-cooled reactors are meticulously engineered to function autonomously, requiring no external power source or manual intervention. They depend exclusively on inherent natural forces and the fundamental principles of reactor physics, such as gravity, natural convection, and phase changes, to manage, alleviate, and avert the release of radioactive materials into the environment during accident scenarios like a loss-of-coolant accident (LOCA). PSSs are already integrated into such operating commercial reactors as the Advanced Pressurized Reactor-1000 MWe (AP1000) and the Water-Water Energetic Reactor-1200 MWe (WWER-1200) are adopted in most of the upcoming small modular reactor (SMR) designs. Examples of water-cooled SMR PSSs are the passive emergency core-cooling system (ECCS), passive containment cooling system (PCCS), and passive decay-heat removal system, the designs of which vary based on reactor system-design requirements. However, understanding the accident-event progression and phases of a LOCA is pivotal for adopting a specific PSS for a new SMR design. This study covers the accident-event progression for direct vessel injection (DVI) small-break loss-of-coolant accident (SB-LOCA), associated physics phenomena, knowledge gaps, and important figures of merit (FOMs) that may need to be evaluated and assessed to validate thermal-hydraulics models with an available experimental dataset to support new SMR design and development.

Overview of separate effect and integral system tests on the passive containment cooling system of SMART100

SMART100 has a containment pressure and radioactivity suppression system (CPRSS) for passive containment cooling system (PCCS). This prevents overheating and over-pressurization of a containment through direct contact condensation in an in-containment refueling water storage tank (IRWST) and wall condensation in a CPRSS heat exchanger (CHX) in an emergency cool-down tank (ECT). The Korea Atomic Energy Research Institute (KAERI) constructed scaled-down test facilities, SISTA1 and SISTA2, for the thermal-hydraulic validation of the SMART100 CPRSS. Three separate effect tests were performed using SISTA1 to confirm the heat removal characteristics of SMART100 CPRSS. When the low mass flux steam with or without non-condensable gas is released into an IRWST, the conditions for mitigation of the chugging phenomenon were identified, and the physical variables were quantified by the 3D reconstruction method. The local behavior of the non-condensable gas was measured after condensation inside heat exchanger using a traverse system. Stratification of non-condensable gas occurred in large tank of the natural circulation loop. SISTA2 was used to simulate a small break loss-of-coolant accident (SBLCOA) transient. Since the test apparatus was a metal tank, compensations of initial heat transfer to the material and effect of heat loss during long-term operation were important for simulating cooling performance of SMART100 CPRSS. The pressure of SMART100 CPRSS was maintained below the design limit for 3 days even under sufficiently conservative conditions of an SBLOCA transient.

Analysis of control rod driving mechanism nozzle rupture with loss of safety injection at the ATLAS experimental facility using MARS-KS and TRACE

Korea Atomic Energy Research Institute (KAERI) has operated 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 tests for transient and design basis accidents simulation. A test for a loss of coolant accident (LOCA) at the top of the reactor pressure vessel (RPV) had been conducted at ATLAS to address the impact of the loss of safety injections (LSI) and to evaluate accident management (AM) actions during the postulated accident. The experimental data has been utilized to validate system analysis codes within a framework of the domestic standard problem program organized by KAERI in collaboration with Korea Institute of Nuclear Safety. In this study, the test has been analyzed by using thermal-hydraulic system analysis codes, MARS-KS 1.5 and TRACE 5.0 Patch 6, and a comparative analysis with experimental and calculation results has been performed. The main objective of this study is the investigation of the thermal-hydraulic phenomena during a small break LOCA at the RPV upper head with the LSI as well as the predictability of the system analysis codes after the AM actions during the test. The results from both codes reveal that overall physical behaviors during the accident are predicted by the codes, appropriately, including the excursion of the peak cladding temperature because of the LSI. It is also confirmed that the core integrity is maintained with the proposed AM action. Considering the break location, a sensitivity analysis for the nodalization of the upper head has been conducted. The sensitivity analysis indicates that the nodalization gave a significant impact on the analysis result. The result emphasizes the importance of the nodalization which should be performed with a consideration of the physical phenomena occurs during the transient.

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.

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.

The reliability analysis of pressurized water reactor safety injection system based on GO-FLOW

The safety injection system (RIS)is a large and complex fluid system in a nuclear power plant. Its central role is to ensure the safety of the core in the event of a breach event in the first circuit of the nuclear power plant. In order to investigate the timing of the safe injection system at the time of a Small-Break Loss-Of-Coolant Accident (SBLOCA), this paper uses the GO-FLOW method to model seven time points in the three operational phases of the safe injection system. The calculation is completed in conjunction with the failure rate of each component to give the reliability of the RIS at each point in time at SBLOCA. The final analysis yielded low reliability at the beginning of the direct cold section injection phase, which is of great reference for subsequent equipment maintenance and the safe operation of nuclear power.

Experimental Investigations and Numerical Studies of Two-Phase Countercurrent Flow Limitation in a Pressurized Water Reactor: A Review

Gas–liquid two-phase countercurrent flow limitation (CCFL) phenomena widely exist in nuclear power plants. In particular, the gas–liquid countercurrent flow limitation phenomena in a pressurized water reactor (PWR) during a loss-of-coolant accident (LOCA) or a small-break loss-of-coolant accident (SBLOCA) play an important role in nuclear reactor safety research. Over several decades, a series of experimental investigations and numerical studies have been carried out to study the CCFL phenomena in a PWR. For the experimental investigations, numerous experiments have been conducted, and different CCFL mechanisms and CCFL characteristics have been obtained in various test facilities simulating different scenarios in a PWR. The CCFL phenomena are affected by many factors, such as geometrical characteristics, liquid flow rates, and fluid properties. For the numerical studies, more and more numerical models were presented and applied to the calculations of two-phase countercurrent flow over the past several decades. It is considered that the computational fluid dynamics (CFD) tools can simulate most of the two-phase flow configurations encountered in nuclear power plants. In this paper, the experimental investigations and the numerical studies on two-phase countercurrent flow limitation in a PWR are comprehensively reviewed. This review provides a further understanding of CCFL in a PWR and gives directions regarding future studies. It is found that relatively fewer investigations using steam–water under high system pressures are performed due to the limitation of the test facilities and test conditions. There are a number of numerical studies on countercurrent two-phase flow in a PWR hot leg geometry, but the simulations in other flow channels were relatively rare. In addition, almost all of the numerical simulations do not include heat and mass transfer. Thus, it is necessary to investigate the effects of heat and mass transfer experimentally and numerically. Furthermore, it is of significance to perform numerical simulations for countercurrent two-phase flow with a fine computational grid and suitable models to predict the formation of small waves and the details in two-phase flow.

Interpretable time series forecasting of NPP parameters in accident scenarios

Nuclear power plant operators are limited in their capacity to analyze large amount of data in a short period, during a transient or accident scenario. Due to recent advances in machine learning methods, advanced algorithms have become available, which can support the operator with real-time analysis and forecast of important NPP parameters. The aim of this study was to provide short-term forecasts of key NPP parameters, which operators use to diagnose reactor states. The study focused on forecasting pressurizer level and subcooling margin, measured between the pressurizer and RCS hot leg, by using a state-of-the-art supervised learning algorithm, the Temporal Fusion Transformer. The model was trained with data from various transient and accident scenarios. The data was generated using an efficient basic principles PWR simulator called the BasicSim-PWR. The models demonstrated the capacity to differentiate between the small break LOCA and SGTR by predicting the safety trip of the SGTR scenario. The interpretable nature of Temporal Fusion Transformer gave insight into model behavior and potentially useful correlations for managing accident scenarios by observing few key inputs. The future research is directed towards advancing the scenario identification capability and efficiency evaluation of operator actions by forecasting the power plant response.

MELCOR 2.2-ASTEC V2.2 crosswalk study reproducing SBLOCA and CSBO scenarios in a PWR1000-like reactor part II: Analysis of containment thermal-hydraulics and ex-vessel phenomena

The Accident Source Term Evaluation Code (ASTEC) and Methods of Estimation of Leakages and Consequences of Releases (MELCOR) system codes are both used by Tractebel to perform Severe Accident (SA) analyses for the Belgian Nuclear Power Plants (NPPs).This paper contains the second part of the comparative study between these two integral software focused on the SA progression in the containment after the Vessel Failure (VF).The main phenomena characterize this phase of the accident usually called ex-vessel phase are the Molten Corium Concrete Interaction (MCCI) and ex-vessel corium/debris coolability, which determine the containment Thermal-Hydraulics (TH) response and can cause its late failure. Despite their importance from a safety point of view, chemical and physical processes behind these phenomena are not yet well understood, partly because of their extreme complexity and partly because of the limitations of current experimental capabilities in reproducing them. As result, models developed and implemented in the codes to describe the MCCI and debris coolability are extremely simplified because based on limited data availability.The benchmark study consists of a comparison of the results calculated by the ASTEC and MELCOR codes during the ex-vessel phase following two different scenarios a Complete Station Blackout (CSBO) and a Small Break Loss Of Coolant Accident (SBLOCA).For each scenario, two different Severe Accident Management Strategies (SAMS) are adopted, the first one is to ensure that the corium, once fully relocated in the cavity, has top-flooding conditions throughout the transient, while the second one is to keep the reactor cavity completely dry. Furthermore, for every case reproduced the reference and best-practice models are investigated.The results show that in case of dry cavity conditions, despite the different VF timing calculated by the two codes, the final results provide very similar indications even though some dissimilarity is observed regarding the amount of fission product (FP) released through the Containment Filtering Venting System (CFVS).In contrast, in the case of wet cavity conditions, the accident progression predicted is significantly different regarding number of venting operations, TH containment conditions, and amount of concrete ablated.The discrepancies observed, should be sought in the corium and debris coolability models, which given their parametric nature and oversimplification can lead to these results.

Investigation of small break LOCA with failure of the emergency core cooling system accident management in VVER-1000 reactor type

In this study, the thermohydraulic condition of the power plant is estimated as a result of the Small Break Loss of Coolant Accident (SB-LOCA) for Bushehr Nuclear Power Plant (BNPP) and the effect of control measures in postponing severe events as well as increasing the time available to enter control measures has been investigated. Two scenarios including control measures with and without operator action have been considered, considering the progress of the accident and the lack of effective intervention of control measures, both scenarios lead to the melting of the reactor core. The critical times for entering additional measures and restoring systems has been obtained from the progress of the accident for two scenarios that is used to develop an accident management strategy. The thermohydraulic calculation in this article deals with the estimation of time of critical point for accident management measures at a different phase of SB-LOCA.

Investigation of condensation with non-condensable gas in natural circulation loop for passive safety system

The system-integrated modular advanced reactor 100 (SMART100), an integral-type pressurized water small modular reactor, is based on a novel design concept for containment cooling and radioactive material reduction; it is known as the containment pressure and radioactivity suppression system (CPRSS).There is a passive cooling system using a condensation with non-condensable gas in the SMART CPRSS. When a design basis accident such as a small break loss of coolant accident (SBLOCA) occurs, the pressurized low containment area (LCA) of the SMART CPRSS leads to steam condensation in an in-containment refuelling water storage tank (IRWST). Additionally, the steam and non-condensable gas mixture passes through the CPRSS heat exchanger (CHX) submerged in the emergency cooldown tank (ECT) that can partially remove the residual heat. When the steam and non-condensable gas mixture passes through the CHX, the non-condensable gas can interrupt the condensation heat transfer in the CHX and it degrades CHX performance.In this study, condensation heat transfer experiments of steam and non-condensable gas mixture in the natural circulation loop were conducted. The pressure, temperature, and effects of the non-condensable gas were investigated according to the constant inlet steam flow rate with non-condensable gas injections in the loop.

Fracture mechanics analysis of a closure head of a PWR reactor pressure vessel LEFM-based SIF

In this study, fracture mechanics analysis of a closure head of a PWR reactor pressure vessel has been performed under accidental conditions. For the purpose, a wide range of elliptical surface cracks starting from minute cracks to worst-case cracks was postulated at the highest stress concentration locations of the head. The analyzed accidental conditions are small break loss of coolant accident (SB-LOCA) and Rancho-Seco transient (RST). The maximum linear elastic fracture mechanics-based stress intensity factors of the worst-case crack ‘a = 0.25 th; a = 0.33 c’ under SB-LOCA and RST are 104.37 MPa√m and 137.12 MPa√m respectively, which are less than the fracture toughness value of the material i.e. 155.6 MPa√m. Hence, the accidental conditions do not threaten the structural integrity of the proposed closure head.