Emergency Core Cooling System Research Articles

NPP Krško 3 inch Cold Leg Break LOCA Calculation using RELAP5/MOD 3.3 and MELCOR 1.8.6 Codes

NPP Krško input deck developed at Faculty of Electrical Engineering and Computing (FER) Zagreb, for severe accident code MELCOR 1.8.6 is currently being tested. MELCOR is primarily used for the analyses of severe accidents including in-vessel and ex-vessel core melt progression as well as containment response under severe accident conditions. Accurate modelling of the plant thermal-hydraulic behaviour as well as engineering safety features, e.g., Emergency Core Cooling System, Auxiliary feedwater system and various containment systems (e.g., Passive Autocatalytic Recombiners, Fan Coolers and Containment spray) is necessary to correctly predict the plant response and operator actions. For MELCOR input data verification, the comparison of the results for small break (3 inch) cold leg Loss of Coolant Accident (LOCA) for NPP Krško using MELCOR 1.8.6 and RELAP5/MOD 3.3 was performed. A detailed RELAP5/MOD 3.3 model for NPP Krško has been developed at FER and it has been extensively used for accident and transient analyses. The RELAP5 model has been upgraded and improved along with the plant modernization in the year 2000. and after more recent plant modifications. The results of the steady state calculation (first 1000 seconds) for both MELCOR and RELAP5 were assessed against the referent plant data. In order to test all thermal-hydraulic aspects of developed MELCOR 1.8.6 model the accident was analysed, and comparison to the existing RELAP5 model was performed, with all engineering safety features available. After initial fast pressure drop and accumulator injection for both codes stable conditions were established with heat removal through the break and core inventory maintained by safety injection. Transient was simulated for 10000 seconds and overall good agreement between results obtained with both codes was found.

An application of fault tree analysis for computing the bounds on system failure probability through qualitative data in intuitionistic fuzzy environment

AbstractFor the safety assessment of complex engineering systems, fault tree analysis (FTA) is a widely used method and can be performed only when all its basic events (BEs) have their corresponding failure data. However, in many applications, obtaining precise quantitative failure data of all the components of the system is not feasible due to changing environment, insufficient data or new components. The present research develops a novel FTA to identify the best estimate and bounds of system failure probability through qualitative data processing. The failure data of BEs are uncertain therefore expert elicitation method is used incorporating intuitionistic fuzzy (IF) set theory to evaluate the top event failure probabilities. The novelty of the proposed work comes with the use of linguistic values as failure possibilities of the system components and utilizing triangular intuitionistic fuzzy numbers (TIFNs) to quantify the linguistic terms into mathematical form to assess the best‐estimated value of the system failure probability, and also the lower bound and the upper bound of the system failure probability. To exemplify the appropriateness of the proposed study, the case studies Combustion Engineering Reactor Protection System (CERPS) and Emergency Core Cooling System (ECCS) are carried out and a comparison between computed results and outcomes from the previous methods is demonstrated. The results are exceptionally near to the outcomes of conventional FTA, which confirms that the proposed procedure is a real substitute in the IF environment when components have imprecise quantitative failure data.

Evaluation of advanced accumulator in a VVER-1000 reactor in loss of coolant accident

The advanced accumulator is a passive system which gradually decreases the high flow into a low flow. The role of the fluidic device is to obtain an additional safety margin by increasing golden time to cope in the LOCA. In this paper, the effect of using the safety injection tank and the fluidic device, called advanced accumulator, in a wide spectrum of LOCA break sizes without emergency core cooling systems (ECCS) of a VVER-1000 is evaluated. Firstly, the main component of primary and secondary loops of Bushehr nuclear power plant (BNPP) is modeled by RELAP5. Secondly, the performance of the model is assessed in steady-state and LOCA conditions by comparing the obtained results with the plant’s FSAR. Thirdly, a model is developed in RELAP5 to simulate the behavior of advanced accumulator and validated by blowdown test in a test facility and MARS-KS code. Finally, the performance of advanced accumulator in the spectrum of 200–300 mm break sizes of LOCA without ECCS. In this range of breaks, the core damage times (CDT) developed from peak cladding temperature (PCT) of 1204 °C have been calculated. In accordance with the results, the advanced accumulators can increase the CDT in the whole range of breaks.

A PCTRAN BASED ANALYSIS ON THE EFFECT OF BREAKSIZE AND COMPARATIVE STUDY BETWEEN HOT AND COLD LEG LOSS OF COOLANT ACCIDENTS IN VVER 1200 POWER REACTOR

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 cm2. 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 cm2 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.

Investigation of Small Break Loca with Failure of the Emergency Core Cooling System Accident Management in Vver-1000 Reactor Type

Investigation of Small Break Loca with Failure of the Emergency Core Cooling System Accident Management in Vver-1000 Reactor Type

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

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

Calculation of dose rates in loss of coolant accident due to double ended rupture of the experimental tangential irradiation beam tube of MTR reactor.

A MTR reactor is an open water pool reactor type. The water pool serves as a shield from radioactive radiations. The most serious accident in this type of reactor is the Loss of Coolant Accident (LOCA) due to rupture either of a primary coolant pipe or of any experimental beam tube. In the present work it has been assumed that pool water drains out due to double ended rupture of the tangential irradiation beam tube (TIC) which has a diameter 150 mm. For an operating power level of 22 MW, the equilibrium core would enter into melting conditions if the pool drain time is less than one hour. It was also assumed that Emergency Core Cooling System (chimney water injection system and siphon effect breaker were not working. The reactor interiors have been modeled using the Monte Carlo N-Particle Transport code MCNPX 2.7.0. The source term has been determined using the ORIGEN-2 code. The doses and dose rates calculations in different places of operator (as phantom of Tissue-Equivalent Material) inside of the reactor building were determined by using Monte Carlo N-Particle Transport (MCNPX).The results show that the dose rate in the control room would be 5.24557 SV/h, the dose rate in the reactor hall above the pools on the gate would be7.20137 SV/h and the dose rate in the Emergency control room would be 2.68239 SV/h. Those dose rates are extremely high and would lead to fatal doses in short time.

Calculation of dose rates in loss of coolant accident due to double ended rupture of the experimental tangential irradiation beam tube of MTR reactor

The doses and dose rates following a LOCA in MTR reactor have been studied. A MTR reactor is an open pool type. The water pool serves as a shield from radioactive radiations.The most serious accident in this type of reactor is the Loss of Coolant Accident (LOCA) due to rupture either of a primary coolant pipe or of any experimental beam tube. In the present work it has been assumed that pool water drains out due to double ended rupture of the tangential irradiation beam tube (TIC) which has a diameter 150 mm. For an operating power level of 22 MW, the equilibrium core will enter into melting conditions if the pool drain time is less than 1 h.It was also assumed that Emergency Core Cooling System (chimney water injection system and siphon effect breaker) were not working. Therefore, conservatively a severe damage (∼80%) is expected to occur to the core, either by a core uncover situation following extended boiling operation, or by a core covered situation with extended boiling.The reactor interiors have been modeled using the Monte Carlo N-Particle Transport code MCNPX 2.7.0. The source term has been determined using the ORIGEN-2 code. The doses and dose rates calculations in different places of operator (as phantom of Tissue-Equivalent Material) inside of the reactor building were determined by using Monte Carlo N-Particle Transport (MCNPX). The results show that the dose rate in the control room is 5.24 Sv/h, the dose rate in the reactor hall above the pools on the gate is 7.20Sv/h and the dose rate in the Emergency control room is 2.68Sv/h. Those dose rates are extremely high and would lead to fatal doses in short time.

Effect of ECCS cold-leg injection angle on thermal hydraulic characteristics and core recovery during LBLOCA in a PWR

The Emergency Core Cooling System (ECCS) is actuated following Loss of Coolant Accidents (LOCA) to inject coolant into the primary loop through a T-junction configuration in nuclear power plants. The injection angle; i.e., the T-junction angle, in the cold leg injection system in a pressurized water reactor is an important parameter during a postulated LOCA. The amount of the break flow rate and the time at which the refill period initiates impact the thermal-hydraulic behavior of the reactor coolant system and eventually affect the integration of reactor nuclear core. In the current work, RELAP5 code is used to study the effect of the cold leg injection angle on LOCA scenario in a standard four-loop Westinghouse Pressurized Water Reactor (PWR). Injection angles investigated in the study are 45°, 90°, and 135°. The whole reactor system is simulated using RELAP5 code whilst the injection junction is numerically developed using ANSYS-FLUENT commercial code. The break flow, core recovery, and thermal-hydraulic parameters are calculated for the different injection angles. Research results are validated with published data available in the literature. The results manifest that the injection angle of 45° gives the highest and the most uniform core level during the core refilling phase, the best temperature behavior for both fuel centerline and maximum cladding temperatures, where both temperature distributions reach a stable value earlier than that of the other angles. ANSYS-FLUENT code is utilized to explain the complex mixing phenomena of fluids during the analysis.

Influence of Passive Containment Cooling System on containment thermal hydraulic transients during a postulated severe accident

The nuclear reactor containment is an important safety system that acts as an ultimate barrier to prevent the release of radioactive fission products to the environment in case of a Severe Accident (SA). The containment system of advanced nuclear reactors is expected to be designed to meet the challenges posed by severe accidents. The response of containment to pressure and temperature build-up during a SA with core melt needs to be assessed to ascertain its structural integrity and leak tightness. The focus of this paper is on analysing the influence of a natural circulation based conceptual Passive Containment Cooling System (PCCS) on the containment pressure and temperature transients in a 220 MWe Pressurized Heavy Water Reactor (PHWR) and highlighting its role as an additional energy management feature. The scenario considered is a Large Break Loss of Coolant Accident (LBLOCA) along with failure of Emergency Core Cooling System (ECCS), moderator and calandria vault water cooling systems. A validated, lumped parameter computer code called Thermal Hydraulics of Containment (THYCON) containing a PCCS model is used for the analysis. The mathematical formulations relevant to the present analysis, solution methodology, validation and benchmarking exercises of THYCON and PCCS model with available experimental and numerical works is presented. It is then applied to analyse the influence of the conceptual PCCS on the long-term containment transients during a postulated SA. Parametric studies are conducted to elucidate the influence of location and operating conditions of PCCS. The key findings are presented and discussed in this paper.

Heat transfer characterization under radial jet and falling film induced rewetting

Abstract Empirical correlations are developed for rewetting velocity and maximum heat transfer coefficient during rewetting phase of single hot vertical Fuel Pin Simulator (FPS) by using radial jet impingement and falling film. Emergency Core Cooling System (ECCS) has been designed for Advance Heavy water Reactor (AHWR) to rewet the hot fuel pin under the loss of coolant accident. Coolant injection takes place from a water rod which is located at the center of the fuel bundle in form of jets to rewet hot surface of fuel pin under loss of coolant accident. This kind of design to reflood the fuel bundle is different than bottom and top spray reflooding practiced in PWR and BWR type of nuclear reactors. There are two different kinds of rewetting found during radial jet induced cooling. The first one is due to radial jet impingement and the second one is due to falling film which is below the jet impingement point. Rewetting velocity has been predicted along the length of fuel pin due to radial jet impingement cooling. Temperature of FPS has been varied from 400°C to 700°C with help of different powers supply, simulating decay heat of reactor. A variation of coolant radial jet mass flow rate is from 0.5 lpm to 1.8 lpm. It is considered during ECCS injection. It has been observed from the experiments that rewetting velocity decreases with increasing the clad surface temperature and increases with increasing the coolant mass flow rate. The rewetting velocity in falling film is found to be nearly 1.8 times higher than rewetting velocity predicted in circumferential direction. Further, it is found that maximum heat transfer coefficient increases with increasing the radial jet coolant mass flow rate. The maximum heat transfer coefficient in case of radial jet impingement is found to be nearly 1.5 times the falling film rewetting. Developed correlation predicts the maximum heat transfer coefficient with experimental data well within the error band of ±10%.

Experimental study on thermally assisted sagging deflection and interaction of multiple coolant channels

A reactor shutdown would typically initiate with a trip signal following a postulated Loss of Coolant Accident (LOCA). The Emergency Core Cooling System (ECCS) assures fuel cooling by flooding the core and recirculating coolant inside the channels for long-term decay heat removal from the fuel. When the primary flow is interrupted, and simultaneously the ECCS is severely impaired or failed, fuel cooling tends to degrade, leading to the heat-up of fuel and coolant channels. Analysis indicates that further progression of such accidents in Pressurized Heavy Water Reactors (PHWRs) is primarily delayed due to the huge water mass in the calandria vessel around the reactor core. However, the horizontal orientation introduces several phenomenological complexities introducing complexity in formulating the Severe Accident Management Guidelines (SAMGs). For a postulated unmitigated low-pressure accident scenario, the accident progresses with the channels heating up, and creep sagging under thermal load and channel and fuel weight. The sagged channel contacts the channel below, transferring heat and its load. With further progress, it is postulated that the supporting channel would fail, and leading to a cascading core collapse.The present paper discusses the experimental studies carried out on multiple channel sagging and disassembly phenomena envisaged during the mid-phase of a severe accident. It is challenging to perform full-length experiments; hence, scaled-down channels were used in the experiments. The postulated accident condition is a significant break LOCA with loss of ECCS and loss of all heat sinks. The experimental setup consists of three channels placed vertically in a fixed lattice pitch forming three distinct channel rows. The paper compares two possible situations or test cases: 1) all three channels are heated, 2) the top two channels are heated, and the lower channel is unheated. The lower channel is an unheated (dummy) channel supporting the upper deforming channels. Finally, a multi-channel sagging deformation mode has been postulated based on the experimental observations. The contact propagation along the contacting surface is found to be a time-dependent event, indicating a channel's rate-dependent elongation. The middle CT is compressed between the upper and lower channels, compromising its structural support for middle PT.

Behavior of Emergency Core Cooling System (ECCS) during the early stage of Loss of Coolant Accident (LOCA) for APR 1400 with Flownex software

The behavior of Emergency Core Cooling System (ECCS) in an Advanced Pressurized Reactor 1400 (APR) during the early stage of Loss of Coolant Accident (LOCA), was analyzed by using Flownex software. The LOCA has been considered in the primary loop in one of the cold legs of APR1400. Through evaluation of the varying break sizes, sensitivity analysis as well as an uncertainty analysis was performed on the Safety Injection System (SIS) where its performance was tested to observe its capability to mitigate a LOCA. The sensitivity analysis was performed on five out puts including core mass flow rate, break mass flow rate, pressure, temperature, and SIS mass flow rate. Ten different break sizes were considered, and sensitivity coefficients were obtained and used to calculate the uncertainty of all outputs. It was observed that all five output responses were sensitive to changes in the input parameter, and both high levels of uncertainty as well as low were observed.

Analysis and Estimation of Core Damage Frequency of Flow Blockage and Loss of Coolant Accident: A Case Study of a 10 MW Water-Water Research Reactor-PSA Level 1

Fault trees (FT) and event trees (ET) are widely used in industry to model and evaluate the reliability of safety systems. This work seeks to analyze and estimate the core damage frequency (CDF) due to flow blockage (FB) and loss of coolant accident (LOCA) due to large rupture of primary circuit pipe with respect to a specific 10 MW Water-Water Research Reactor in Ghana using the FT and ET technique. Using FT, the following reactor safety systems: reactor protection system, primary heat removal system, isolation of the reactor pool, emergency core cooling system (ECCS), natural circulation heat removal, and isolation of the containment were evaluated for their dependability. The probabilistic safety assessment (PSA) Level 1 was conducted using a commercial computational tool, system analysis program for practical coherent reliability assessment (SAPHIRE) 7.0. The frequency of an accident resulting in severe core damage for the internal initiating event was estimated to be 2.51e − 4/yr for the large LOCA as well as 1.45e − 4/yr for FB, culminating in a total core damage frequency of 3.96e − 4/yr. The estimated values for the frequencies of core damage were within the expected margins of 1.0e − 5/yr to 1.0e − 4/yr and of identical sequence of the extent as found for similar reactors.

The ETH Zurich curated nuclear events database: Layout, event classification, and analysis of contributing factors

We present an open database of nuclear events focused on worldwide safety significance with potentials for precursors. Explaining our events collection method and classification approach, each of the 1250 events in the database has been subjugated to coherent breakdown of features such as significance, origin, operating conditions, failure chains, contributing factors, severity of failures, and others. The events have been analyzed by experts and researchers in nuclear technology and safety, and are accessible using a custom-made user interface, making the database the largest open, comprehensive, curated, and user-friendly database in the world. We find that the majority of events (52%) have originated outside the nuclear island compared to within (48%). The most commonly affected components are related to the emergency power and emergency core cooling systems (ECCS). Design residuals are the major contributor to systems’ unreliability, with an occurrence frequency of more than 20%. Finally, the importance of vigilance by the plant staff and regulators is highlighted, as the contribution of human/organizational factors is found to be similar to that of technical factors.

A proposed improvement for the design of safety injection system in VVER-1000/V446 reactor

If the Total Loss of Feedwater (TLFW) accident occurs in a VVER reactor, the primary side Feed & Bleed (F&B) recovery strategy prevents core damage by reducing the reactor's pressure by opening the pressurizer's safety valves and establishing core cooling by the Emergency Core Cooling System (ECCS). Although the strategy saves the core, the leaked coolant together with the containment spray system pollutes the containment environment, deactivates electrical devices, and jeopardizes the containment integrity by the release of the generated hydrogen during the core re-flooding. Further, Steam Generators (SGs) would be exposed to excessive thermal stress because of drying out, making use of recovered feedwater too risky because of the primary to secondary side leakage possibility. ECCS in VVER-1000/V446 has low-pressure Hydro accumulators (HAs) that could carry out secondary side F&B through SGs. This feasibility study uses Deterministic Safety Analysis (DSA) to show that the secondary side F&B recovery strategy can provide 6 h after losing Emergency Feedwater (EFW) which is a significant time to recover the failed safety systems. This strategy is more significant, especially if the TLFW accident is followed by the loss of A.C. power, which makes the primary side F&B ineffective, leading to the core meltdown within 2 h. The Standardized Plant Analysis Risk-Human Reliability Analysis (SPAR-H) is used to calculate Human Error Probability (HEP) for the primary and secondary side F&B recoveries. Probabilistic Safety Assessment (PSA) reveals that the proposed recovery strategy reduces the Core Damage Frequency (CDF) by one order of magnitude.

Investigation of ex-vessel core catcher for SBO accident in VVER-1000/V528 containment using MELCOR code

To mitigate consequences of core melting, an ex-vessel core catcher is investigated in this study. Instructions should be obeyed to cool down the corium caused by core melting. The corium destroys the reactor containment and causes radioactive materials to be released into the environment if it does not cool down well. It is important to build a core catcher system for the reception, localization, and cool down of the molten corium during a severe accident resulting from core melting. In this study, the role of a core catcher in the VVER-1000/v528 reactor containment during a station black out (SBO) accident is evaluated using the MELCOR1.8.6 code. In addition, parametric analyses of the SBO for (i) SBO accidents with emergency core cooling system (ECCS) operation, and (ii) without ECCS operation are performed. Furthermore, thermal–hydraulic analyses in dry and wet cavities with/without water are conducted. The investigations include the reduction of gases resulting from molten–corium–concrete interactions (H2, CO, CO2). Core melting, gas production, and the pressure/temperature in the reactor containment are assessed. Additionally, a full investigation pertaining to gas release (H2, CO, CO2) and the pressure/temperature of the core catcher is performed. Based on MELCOR simulations, a core cavity and a perimeter water channel are the best options for corium cooling and a lower radionuclide release. This simulation is also theoretically investigated and discussed herein. The simulation results show that the core catcher system in addition to an internal sacrificial material reduces the containment pressure from 689 to 580 kPa and the corresponding temperature from 394 to 380 K. Furthermore, it is observed that the amount of gases produced, particularly hydrogen, decreased from 1698 to 1235 kg. Moreover, the presence of supporting systems, including an ECCS with a core catcher, prolonged the core melting time from 16,430 to 28,630 s (in an SBO accident) and significantly decreased the gases produced.

An old issue and a new challenge for nuclear reactor safety

Nuclear reactor safety (NRS) and the branch accident analysis (AA) constitute proven technologies: these are based on, among the other things, long lasting research and operational experience in the area of water cooled nuclear reactors (WCNR). Large break loss of coolant accident (LBLOCA) has been, so far, the orienting scenario within AA and a basis for the design of reactors. An incomplete vision for those technologies during the last few years is as follows: Progress in fundamentals was stagnant, namely in those countries where the WCNR were designed. Weaknesses became evident, noticeably in relation to nuclear fuel under high burn-up. Best estimate plus uncertainty (BEPU) techniques were perfected and available for application. Electronic and informatics systems were in extensive use and their impact in case of accident becomes more and more un-checked (however, quite irrelevant in case of LBLOCA). The time delay between technological discoveries and applications was becoming longer. The present paper deals with the LBLOCA that is inserted into the above context. Key conclusion is that regulations need suitable modification, rather than lowering the importance and the role of LBLOCA. Moreover, strengths of emergency core cooling system (ECCS) and containment need a tight link.

Numerical Simulation Study on the Air/Water Counter- current Flow Limitation in Nuclear Reactors

After a loss-of-coolant accident (LOCA) in a Pressurized Water Reactor (PWR), the temperature of the fuel elements cladding increases dramatically due to the heat produced by the fission products decay, which is not adequately removed by the vapor contained in the core. In order to avoid this sharp rise in temperature and consequent melting of the core, the Emergency Core Cooling System is activated. This system initially injects borated water from accumulator tanks of the reactor through the inlet pipe (cold leg) and the outlet pipe (hot leg), or through the cold leg only, depending on the plant manufacturer. Some manufacturers add to this, direct injection into the upper plenum of the reactor. The penetration of water into the reactor core is a complex thermofluidodynamic process because it involves the mixing of water with the vapor contained in the reactor, added to that generated in the contact of the water with the still hot surfaces in various geometries. In some critical locations, the vapor flowing in the opposite direction of the water can control the penetration of this into the core. This phenomenon is known as Countercurrent Flow Limitation (CCFL) or Flooding, and it are characterized by the control that a gas exerts in the liquid flow in the opposite direction. This work presents a proposal to use a CFD to simulate the CCFL phenomenon. Numerical computing can provide important information and data that is difficult or expensive to measure or test experimentally. Given the importance of computational science today, it can be considered a third and independent branch of science on an equal footing with the theoretical and experimental sciences.

Analysis of loss of coolant accident without ECCS and DHRS in an integral pressurized water reactor using RELAP/SCDAPSIM

Small Modular Reactors (SMRs) are currently being developed and licensed in many countries. The innovations in the designs provide many significant features, such as decreased construction cost and time, enhanced safety, modularity, possibility of factory production etc. Several new SMR designs are based on the concept of an integral Pressurized Water Reactor. One of such designs, namely the NuScale SMR was selected for this analysis. An input model was developed from scratch for the RELAP/SCDAPSIM code using publicly available data from the NuScale Final Safety Analysis Report provided by the U.S. NRC and some validation tests have already been conducted. An extremely severe event is analyzed and discussed in the current work. It is an inside containment Loss of Coolant Accident followed by failure of both the Emergency Core Cooling System and the Decay Heat Removal System. The simulation was conducted for 86400 s (24 h) with the core damage starting after ~4.8 h, which is comparable to the results for similar scenarios presented in the references.