High Pressure Safety Injection Research Articles

RPV top and bottom break SBLOCA with passive emergency core cooling system using ATLAS test facility

A SBLOCA test, named as C2.1 in the frame of the OECD/NEA ATLAS3 project, using a passive emergency core cooling system (PECCS) was performed by Korea Atomic Energy Research Institute (KAERI). In the test, two simultaneous 2-inch equivalent breaks occurred at top and bottom nozzles of the reactor pressure vessel of ATLAS test facility. The main purpose of the test is to investigate thermal hydraulic transient behavior in a reactor coolant system (RCS) during a SBLOCA scenario and to evaluate the effectiveness of PECCS. From the observation on thermal hydraulic phenomena during the test, the test period was divided into three characteristic phases such as pressure plateau, safety injection, and clad temperature re-increase phases. Continuous inventory loss of the primary loop through two break nozzles led to a primary system pressure decrease and a clad temperature increase. Even though high-pressure safety injection tanks (HPSITs) were actuated in an early stage of the pressure plateau phase, safety injection from HPSITs, however, were not actually injected. Safety injection flows from the HPSITs and SITs were practically injected to the downcomer only after the opening of automatic depressurization valves (ADVs). With a continuous decrease of a primary system pressure, low-pressure (LP) injection was started. However, LP injection could not change the increasing trend of the clad temperatures. Post-test calculations were performed using MARS-KS 1.4, and calculation results were compared with those of the test.

Safety margins improvement by means of the passive second stage hydroaccumulators in a VVER-1000/V320 reactor

Since the Fukushima accident, there has been a new surge in interest in passive safety systems that ensure the core integrity in accidental sequences involving a total loss of AC power. In this sense, the majority of the GEN-III reactors incorporate advanced passive safety systems. One such system is the Second Stage Hydroaccumulators (HA-2) system, which is a passive safety injection system included in some advanced VVER reactors. The goal of this paper is to analyse the impact of the HA-2 on the events of a LOCA sequence, with and without SBO. For this purpose, a VVER-1000/V320 thermal hydraulic model for TRACEV5P5 code has been modified to include the HA-2 system. By analysing the results, it is found that the single performance of this passive safety system along with the accumulators, without considering any other management action, is enough to avoid the core damage in medium/large LOCA along with SBO during 24 h, moreover this system allows also to relax the success criteria of the active High and Low Pressure Safety Injection Systems in medium/large LOCA sequences without SBO conditions.

A Pragmatic Approach to Update Support Studies and Human Reliability Analyses for a PSA Model

A probabilistic safety assessment (PSA) is used to estimate potential risks associated with a nuclear power plant. Event trees (ETs) and fault trees (FTs) describe accident scenarios following initiating events. They allow for identifying and quantifying the consequence frequencies. Based on success criteria, support studies are used to determine each scenario’s consequence. They are also used to define the time available for the operators to carry out the actions involved in the scenario, and consequently, the corresponding human failure probability. In this paper, we use a dynamic PSA toolbox to optimize support studies for a classical (ET/FT) PSA model. We analyze medium-break loss-of-coolant accident (LOCA) initiating events with the possible failure of the high-pressure safety injection (HPSI) system for an internal event Level 1 PSA for a pressurized water reactor plant. In order to prevent core uncovery and subsequent core damage, operators must depressurize the primary circuit to reach the low-pressure safety injection set point. This scenario represents a significant contribution to the core damage frequency. The dominant cut sets corresponding to this scenario involve HPSI failure to run over 24 h, while the thermal-hydraulic analysis supporting the human reliability assessment (HRA) analysis for this sequence assumes the unavailability of HPSI pumps at the beginning of the accident. We update the thermal-hydraulic assumptions in the support study by performing simulations for different values of the HPSI failure time. We find that even a short HPSI operation time buys significant (from the HRA point of view) available time for operators and drives significant improvement in estimating the human error probability (HEP). We postprocessed the minimal cut-set probability by integrating the obtained HEP. This result allows for more realistic quantification of the contribution of the medium-break LOCA in the core damage frequency.

Safety Benefits Assessment for Accident Tolerant Fuels in Consideration of Steam Generator Tube Degradation Using Dynamic Event Tree Analysis

Accident tolerant fuel (ATF) is expected to delay or prevent core damage by providing additional coping time under accidents involving loss of core cooling. The effect of extended coping time may vary depending on the plant response to accidents. Age-related component degradation that deteriorates plant performance over time could have an impact on the actual advantages of ATF. The potential safety benefits of two near-term ATF candidates, including Cr-coated Zr cladding and FeCrAl cladding, are assessed for a 2-in. loss-of-coolant accident with failed high-pressure safety injection using the dynamic event tree (DET) approach considering possible stress corrosion cracking of steam generator (SG) tubing under aging. The DET approach allows likelihood quantification of accident sequences leading to core damage, including stochastic variation of system response and human actions during accident mitigation. The safety benefits of the selected ATF claddings in terms of additional coping time and the core damage frequency reduction rate under specified accident situations were quantitatively estimated. The results show that the deployment of the two selected ATF claddings is expected to lead to longer coping times and lower core damage frequency due to the wider safety margin to peak cladding temperature they provide. The safety advantages would be greater as SG tube degradation proceeds. Thus, the two ATF candidates would lead to less severe consequences in terms of likelihood of core damage and susceptibility to the SG tube degradation than UO2-Zr fuel.

BETHSY 9.1b test calculation with TRACE using 3D vessel component

Recently, several advanced multidimensional computational tools for simulating reactor system behavior during real and hypothetical transient scenarios were developed. One of such advanced, best-estimate reactor systems codes is TRAC/RELAP Advanced Computational Engine (TRACE), developed by the U.S. Nuclear Regulatory Commission. The advanced TRACE comes with a graphical user interface called SNAP (Symbolic Nuclear Analysis Package). It is intended for pre- and post-processing, running codes, RELAP5 to TRACE input deck conversion, input deck database generation etc. The TRACE code is still not fully development and it will have all the capabilities of RELAP5. The purpose of the present study was therefore to assess the 3D capability of the TRACE on BETHSY 9.1b test. The TRACE input deck was semi-converted (using SNAP and manual corrections) from the RELAP5 input deck. The 3D fluid dynamics within reactor vessel was modeled and compared to 1D fluid dynamics. The TRACE 3D calculation was compared both to TRACE 1D calculation and RELAP5 calculation. Namely, the geometry used in TRACE is basically the same, what gives very good basis for the comparison of the codes. The only exception is 3D reactor vessel model in case of TRACE 3D calculation. The TRACE V5.0 Patch 1 and RELAP5/MOD3.3 Patch 4 were used for calculations. The BETHSY 9.1b test (International Standard Problem no. 27 or ISP-27) was 5.08 cm equivalent diameter cold leg break without high pressure safety injection and with delayed ultimate procedure. BETHSY facility was a 3-loop replica of a 900 MWe FRAMATOME pressurized water reactor. In general, all presented code calculations were in good agreement with the BETHSY 9.1b test. The TRACE 1D calculation results are comparable to RELAP5 calculated results. For some parameters they are better, this is mostly due to better tuning of the break flow, what influences timing of the transient. When comparing TRACE 1D and TRACE 3D calculation, the latter is slightly better. One reason for comparable results is already good agreement of 1D calculations and there was not much space to further improve the results. The other reason may be that in the facility the phenomena were mostly one dimensional (for example, external downcomer was used for reactor vessel modeling). However, when 3D behavior of the heater rod temperatures was investigated, the advantage of three dimensional treatment was clearly demonstrated.

Review of Design Extension Conditions Experiments and Analyses for Non-degraded Core

The second generation nuclear power plants were designed and built to withstand without loss to the systems, structures, and components necessary to ensure public health and safety during design basis accidents (DBAs). In the transient and accident analysis the effects of single active failures and operator errors were considered. There are also accident sequences that are possible but were judged to be too unlikely and therefore were not fully considered in the design process of second generation reactors. In that sense, they were considered beyond the scope of design-basis accidents that a nuclear facility must be designed and built to withstand. Such accident sequences have been analysed in the past to fully understand the capability of a design. The requirements to analyse such sequences for existing reactors have been introduced after Fukushima Dai-ichi accident. In 2012 the design extension conditions (DECs) were introduced in the International Atomic Energy Agency (IAEA) requirements for the design of nuclear power plants (NPPs). Western European Nuclear Regulators Association (WENRA) requirements of existing reactors for DEC were introduced in 2014. The purpose of considering DEC is to further improve safety by enhancing the plant’s capability to withstand the conditions generated by accidents that are more severe than DBAs. This concept by IAEA and WENRA (WENRA definition of DEC is consistent with IAEA definition from 2012, in which DEC with prevention of core melt is called DEC A) is not completely new, since some multiple failures have already been considered in the design of existing reactors, for example anticipated transients without scram and station blackout. The research for beyond design basis accidents with non-degraded core (i.e. DEC A) for existing reactors has been already done in 80’s and 90’ of the previous century. The purpose of this paper is to review that research. The tests performed include total loss of feedwater, station blackout, small break without high pressure safety injection, steam generator tube rupture with no high pressure safety injection etc. Besides review of experiments performed on integral test facilities, examples of DEC A tests, which have been analysed at Jožef Stefan Institute using RELAP5 or TRACE computer code in the last three decades, will be presented too.

Study and Optimization of Full Flow Test Method for Core Safety Injection System

Due to the high pressure safety injection full-flow test of the nuclear power plant core safety injection system (RIS), there are problems such as the limit of the pump starting condition, the “blind adjustment” of the valve flow, the test logic defect, and the number of unqualified tests, etc. This research has improved the test principle and operation steps of the RIS high pressure safety injection full-flow periodic test, and describes in detail the comparison between the new scheme and the existing RIS high pressure safety injection full-flow test scheme, and the actual verification effect. The research results show that the optimized RIS high pressure safety injection full-flow test has the advantages of simple logic, easy operation, high qualification rate, and short test period, and it is worthy of popularization and application in this reactor type.

Investigating the Effect of Prior Distributions on Posterior Estimates of Common Cause Failure Parameters Using Bayesian Method

Abstract Quantification of common cause failure (CCF) parameters and their application in multi-unit PSA are important to the safety and operation of nuclear power plants (NPPs) on the same site. CCF quantification mainly involves the estimation of potential failure of redundant components of systems in a NPP. The components considered in quantification of CCF parameters include motor operated valves, pumps, safety relief valves, air-operated valves, solenoid-operated valves, check valves, diesel generators, batteries, inverters, battery chargers, and circuit breakers. This work presents the results of the CCF parameter quantification using check valves and pumps. The systems considered as case studies for the demonstration of the proposed methodology are auxiliary feedwater system (AFWS) and high-pressure safety injection (HPSI) systems of a pressurized water reactor (PWR). The posterior estimates of alpha factors assuming two different prior distributions (Uniform Dirichlet prior and Jeffreys prior) using the Bayesian method were investigated. This analysis is important due to the fact that prior distributions assumed for alpha factors may affect the shape of posterior distribution and the uncertainty of the mean posterior estimates. For the two different priors investigated in this study, the shape of the posterior distribution is not influenced by the type of prior selected for the analysis. The mean of the posterior distributions was also analyzed at 90% confidence level. These results show that the type of prior selected for Bayesian analysis could have effects on the uncertainty interval (or the confidence interval) of the mean of the posterior estimates. The longer the confidence interval, the better the type of prior selected at a particular confidence level for Bayesian analysis. These results also show that Jeffreys prior is preferred over Uniform Dirichlet prior for Bayesian analysis because it yields longer confidence intervals (or shorter uncertainty interval) at 90% confidence level discussed in this work.

Preliminary optimization of emergency core cooling system with accident-tolerant fuels

Based on the excellent performance of accident-tolerant fuels (ATFs), the optimization schemes for the safety system of Chinese advanced three-loop pressurized water reactor during loss of coolant accident (LOCA) were proposed and corresponding feasibility analysis was carried out in this study. The modified system analysis code was used to safety evaluation and four ATF systems were investigated. Several modifications were made to the rod geometries in partial ATF systems to make up the poor neutron economic performance of FeCrAl and MoMAX. The emergency core cooling system (ECCS) was optimized during LOCA with different break sizes. The results show that it is feasible for all ATF systems to cancel the high pressure safety injection (HPSI) or cancel both HPSI and low pressure safety injection (LPSI) but add an extra accumulator (ACC). Also the safety of reactors fueled with ATFs can be guaranteed with only 0.35 times the original flowrate of LPSI during large break LOCA (LB-LOCA). In the scheme of replacing the original ECCS with two extra ACCs during LB-LOCA, reactors loaded with partial ATFs are safe, while UO2-Zr systems are in danger as a contrast. This work could provide reference for the design of reactor safety system in the future.

Analysis of Reliability-Critical Hydraulic Impact Conditions at WWER-1000 NPP Active Safety Systems

An original method to determine the onset of conditions for reliabilitycritical hydraulic impacts for reliability analysis of active safety systems of nuclear power installations is proposed. The suggested method is based on determining the effect of head-flow characteristic delay onto hydraulic impact preconditions during the changes of pipeline system hydrodynamic parameters under transitional modes (e.g. in pump start-up). The delay time of responses to change in the hydrodynamic system parameters embodies the determining factor of head-flow characteristic’s inertance, depending on both design and technical parameters of system components (including pumps), and the hydrodynamic parameters change rate under transitional modes. Using the proposed method, the analysis of conditions for critical hydraulic impacts is performed for the primary high-pressure safety injection system of serial WWER-1000 nuclear power plants. The analysis results allow a conclusion that for this system the conditions leading to hydraulic impact due to the pump start are not reached. The developed method can be applied to any thermal and nuclear power facilities’ pipeline systems equipped with pumps.

Variability of plant risk due to variable operator allowable time for aggressive cooldown initiation

Abstract Recent analysis results with realistic assumptions provide the variability of operator allowable time for the initiation of aggressive cooldown under small break loss of coolant accident or steam generator tube rupture with total failure of high pressure safety injection. We investigated how plant risk may vary depending on the variability of operators’ failure probability of timely initiation of aggressive cooldown. Using a probabilistic safety assessment model of a nuclear power plant, we showed that plant risks had a linear relation with the failure probability of aggressive cooldown and could be reduced by up to 10% as aggressive cooldown is more reliably performed. For individual accident management, we found that core damage potential could be gradually reduced by up to 40.49% and 63.84% after a small break loss of coolant accident or a steam generator tube rupture, respectively. Based on the importance of timely initiation of aggressive cooldown by main control room operators within the success criteria, implications for improvement of emergency operating procedures are discussed. We recommend conducting further detailed analyses of aggressive cooldown, commensurate with its importance in reducing risks in nuclear power plants.

Cooldown procedure success criteria map for the full break size spectrum of SBLOCA

A success criteria map defining successful implementation of a cooldown procedure to mitigate small-break loss of coolant accident (SBLOCA) with failed high pressure safety injection system for the OPR1000 is generated for the full break size spectrum from 0.5 in to 2.5 in diameter breaks. The success criteria map is a multi-dimensional response surface that gives the expected peak clad temperature (PCT) as a function of break size and operator action time to start the 55.6 °C/hr reactor coolant system cooldown and depressurization involving secondary side steam dump through manual manipulation of the atmospheric dump valves (ADV). The PCT response surface is constructed through a regression analysis using a Gaussian process model (GPM) coupled to an adaptive sampling procedure that greatly reduces the computational cost and analysis time by limiting the number of required best estimate simulations of the SBLOCA with the MARS code to approximately 100. Break flow rates and times to core heat up each span several orders of magnitude when considering the full break size spectrum so an innovative solution of transforming the break size and ADV actuation time input variables using exponential and logarithm functions was implemented such that the regression could be performed on the scaled and transformed input space.

Fire Safety Evaluation for HPSI Pump Room Using CDF Code

Although fire accidents in nuclear power plants rarely occurs, it may cause a stoppage of a nuclear reactor and damage of safety devices and cause massive damage such as radiation leakage. Therefore, nuclear power plants require specialized fire safety design conditions. In this study, CFD code is used to predict fire behavior and verify safety for HPSI (High Pressure Safety Injection) Pump Room passing through cables of important pump among various pump rooms of nuclear power plant. Fire simulation was performed based on NUREG-1934, NUREG / CR-6850, etc. using Fire Dynamics Simulator (FDS) as a fire model for fire safety evaluation of HPSI Pump Room. As a result, the fire damper closing time of each exhaust was checked and the damage time of the cable was evaluated. This will provide information for safety assessment and assurance of nuclear power plants. Keywords: HPSI Pump Room, Fire Dynamics Simulator, Fire Damper, Cable Damage

Quantification of LOCA core damage frequency based on thermal-hydraulics analysis

A loss-of-coolant accident (LOCA) has always been significantly considered one of the most important initiating events. However, most probabilistic safety assessment models, up to now, have undoubtedly adopted the three groups of LOCA, and even an exact break size boundary that used in WASH-1400 reports was published in 1975. With an awareness of the importance of a realistic PSA for a risk-informed application, several studies have tried to find the realistic thermal-hydraulic behavior of a LOCA, and improve the PSA model. The purpose of this research is to obtain realistic results of the LOCA core damage frequency based on a success criteria analysis using the best-estimate thermal-hydraulics code. To do so, the Korea Standard Nuclear Power Plant (KSNP) was selected for this study. The MARS code was used for a thermal hydraulics analysis and the AIMS code was used for the core damage quantification. One of the major findings in the thermal hydraulics analysis was that the decay power is well removed by only a normal secondary cooling in LOCAs of below 1.4in and by only a high pressure safety injection in LOCAs of 0.8–9.4in. Based on the thermal hydraulics results regarding new break size boundaries and new success criteria, five new event trees (ETs) were developed. The core damage frequency of new LOCA ETs is 5.80E−07 (/y), which is 12% less than the conventional PSA ETs. In this research, we obtained not only thermal-hydraulics characteristics for the entire break size of a LOCA in view of the deterministic safety assessment, but also a more realistic core damage frequency of the LOCAs using updated information. The difference between the results of the present study and the conventional knowledge can be used to modify the current emergency operator procedures and to design another nuclear power plant type.

Effectiveness and adverse effects of in-vessel retention strategies under a postulated SGTR accident of an OPR1000

ABSTRACTDuring a steam generator tube rupture (SGTR) accident, direct release of radioactive nuclides into the environment is postulated via bypassing the containment building. This conveys a significant threat in severe accident management (SAM) for minimization of radionuclide release. To mitigate this risk, a numerical assessment of SAM strategies was performed for an SGTR accident of an Optimized Power Reactor 1000 MWe (OPR1000) using MELCOR code. Three in-vessel mitigation strategies were evaluated and the effect of delayed operation action was analyzed. The MELCOR calculations showed that activation of a prompt secondary feed and bleed (F&B) operation using auxiliary feed water and use of an atmospheric dump valve could prevent core degradation. However, depressurization using the safety depressurization system could not prevent core degradation, and the injection of coolant via high-pressure safety injection without the use of reactor coolant system (RCS) depressurization increased fission product release. When mitigation action was delayed by 30 minutes after SAMG entrance, a secondary F&B operation failed in depressurizing the RCS sufficiently, and a significant amount of fission products were released into the environment. These results suggest that appropriate mitigation actions should be applied in a timely manner to achieve the optimal mitigation effects.

Numerical analysis of fluid mixing characteristics in cold leg and downcomer under high pressure safety injection

Numerical analysis of fluid mixing characteristics in cold leg and downcomer under high pressure safety injection

Investigation on transient flow of a centrifugal charging pump in the process of high pressure safety injection

In order to investigate the transient flow characteristics of the centrifugal charging pump during the transient transition process of high pressure safety injection (HPSI) from Q=148m3/h to Q=160m3/h, numerical simulation and experiment are implemented in this study. The transient flow rate, which is the most important factor, is obtained from the experiment and works as the boundary condition to accurately accomplish the numerical simulation in the transient process. Internal characteristics under the variable operating conditions are analyzed through the transient simulation. The results shows that the absolute error between the simulated and measured heads is less than 2.26% and the absolute error between the simulated and measured efficiency is less than 2.04%. Pressure fluctuation in the impeller is less influenced by variable flow rate in the HPSI process, while flow pattern in the impeller is getting better and better with the flow rate increasing. As flow rate increases, fluid blocks on the tongue of the volute and it strikes in this area at large flow rate. Correspondingly, the pressure fluctuation is intense and vortex occurs gradually during this period, which obviously lowers the efficiency of the pump. The contents of the current work can provide references for the design optimization and fluid control of the pump used in the transient process of variable operating conditions.

Reduction of core damage frequency via new design for emergency core cooling system in a typical PWR

The purpose of this study is probabilistic safety assessment of two different designs for emergency core cooling system (ECCS) in a typical PWR. The first design which is indeed the current operating design for ECCS includes i) high pressure safety injection system which comprises two identical trains that each of them includes two 50% capacity pumps, ii) low pressure safety injection system comprises two identical trains that each of them includes one 100% capacity pump, and iii) Two 100% capacity passive accumulators. The second design, in other words, the improved design namely, safety injection system, includes two trains that each of them comprises one 100% capacity pump and four 50% capacity passive accumulators. The assessment of two designs was completed in four steps. First, identify system components. Second, do failure mode and effect analysis. Third, perform fault tree analysis considering uncertainty and common cause failure and finally, calculate the core damage frequency via event tree analysis. By considering large break loss of coolant accident as an initiating event the final value of core damage frequencies are obtained 6.5 × 10−9/Reactor.Yr and 2.69 × 10−9/Reactor.Yr for the first and second design scheme of ECCS, respectively. It is found that there is a significant reduction in core damage frequency utilizing the second design for ECCS. All calculations are carried out by using Risk Spectrum software.

Effects of RCP trip when recovering HPSI during LOCA in a Westinghouse PWR

Current Westinghouse Emergency Operating Procedures (EOPs) indicate initially that the operator must keep the reactor coolant pumps (RCPs) running during a Small Break Loss of Coolant Accident (SBLOCA) if there is unavailability of high pressure safety injection (HPSI) system in order to cool the core by forced convection. However, the crew must follow different EOPs along the transient depending on its evolution. In these EOPs there are several conditions which indicate the necessity of tripping one or more RCPs when HPSI is recovered. In this paper the occurrence of a SBLOCA with unavailability of HPSI has been analyzed with a model of Almaraz Nuclear Power Plant (Westinghouse 3 Loop) for TRACE code V5.0 patch 1. Two different approaches have been considered: the first one, taking into account Optimal Recovery Guidelines (ORGs) and in the second approach, the transition to Function Restoration Guidelines (FRGs) due to inadequate core cooling (ICC) conditions is considered. Results of this paper lead to the implementation of an improvement in current EOPs regarding how many RCPs should be tripped during SBLOCA sequences.

Feasibility analysis of aggressive cooldown in OPR-1000 nuclear power plants

Aggressive cooldown is the action taken by main control room (MCR) operators to rapidly cool down and depressurize the reactor coolant system at the maximum allowed cooldown rate during a small break loss of coolant accident (SBLOCA) or a steam generator tube rupture accident with failure of high-pressure safety injection (HPSI) in a pressurized water reactor. We examined the feasibility of aggressive cooldown in OPR-1000 nuclear power plants by establishing success criteria based on experimental results and thermal–hydraulic code simulations, reviewing emergency operating procedures, and estimating the necessary time for MCR operators to initiate aggressive cooldown based on experimental results with licensed nuclear power plant operators. For an SBLOCA with a break size of 0.02ft2 combined with total failure of HPSI, we found that aggressive cooldown can generally be performed successfully, but there is potential for failure to initiate aggressive cooldown in a timely manner. We discuss the potential effects of introducing the Ultimate Procedure on reducing the core damage potential. Detailed analysis and further discussion are necessary to reduce uncertainties associated with aggressive cooldown and its positive effect on the safety of nuclear power plants.