Loss Of Residual Heat Removal Research Articles

TPTF horizontal flow prediction by SYS-TH codes – Recent analyses made within the FONESYS network

The horizontal stratification occurrence in water cooled nuclear reactors (WCNR) plays an important role in many transients such as loss of coolant accidents and loss of residual heat removal by influencing the liquid mass repartition, the natural circulation and reflux condenser modes, and the break or leak flowrate. A benchmark activity on horizontal flow prediction conducted in the framework of FONESYS is here presented. FONESYS is a network among code developers who share the common objective to strengthen current technology. The aim of the network is to highlight the capabilities and the robustness as well as the limitations of current SYS-TH codes to predict the main phenomena during transient scenarios in nuclear reactors for safety issues.One-hundred-twenty-six runs performed with the Two-Phase Test Facility (TPTF) are considered. Tests were selected aiming to support and complement an ongoing activity on the scalability of codes’ closure laws, currently focusing on the horizontal stratification criteria, the onset of droplet entrainment and entrainment fraction criteria. Therefore, experimental tests performed with the 4-inch and 8-inch tests sections, at pressure ranging from 30 bar to 118 bar, in different flow regimes are considered. Focus is put on the capability of ATHLET, CATHARE 3, MARS-KS, RELAP5, RELAP5-3D, SPACE and TRACE in predicting void fractions, the transition to slug flow and the onset of droplet entrainment.Overall, the benchmark results show reasonably good capabilities of codes in predicting horizontal flow. However, some noticeable limitations emerged related to flow regime transition criteria, interfacial friction, and to the prediction of non-established flow. The scalability of closure laws and possible ways for improving them are discussed.

Study of transient events during the full-system chemical decontamination operation of the Kori-1 nuclear power plant

The Kori-1 nuclear power plant (NPP), a Westinghouse 2-loop pressurized water reactor in Korea, permanently ceased operations on June 18, 2017. Before the main decommissioning activities, full chemical decontamination of primary systems will be performed to secure the safety of workers and minimize the quantity of radioactive waste. In this study, the steady and transient flow characteristics for the chemical decontamination operations were analyzed via a simulation using the Multidimensional Analysis of Reactor Safety, KINS Standard version. Then, based on the accumulated flow mass released from the reactant cooling systems during transient events, the radioactivity levels were estimated using the oxide layer removal ratio obtained from the chemical decontamination experiment. The loss of residual heat removal, steam generator tube rupture (SGTR), power-operated relief valve open (PORV Open), and combination of SGTR and PORV Open simulations were investigated in terms of the transient events. Finally, a strategy for managing transient events during the full-system chemical decontamination operation of the Kori-1 NPP is proposed.

Flow Characteristics Analysis for the Chemical Decontamination of the Kori-1 Nuclear Power Plant

Chemical decontamination of primary systems in a nuclear power plant (NPP) prior to commencing the main decommissioning activities is required to reduce radiation exposure during its process. The entire process is repeated until the desired decontamination factor is obtained. To achieve improved decontamination factors over a shorter time with fewer cycles, the appropriate flow characteristics are required. In addition, to prepare an operating procedure that is adaptable to various conditions and situations, the transient analysis results would be required for operator action and system impact assessment. In this study, the flow characteristics in the steady-state and transient conditions for the chemical decontamination operations of the Kori-1 NPP were analyzed and compared via the MARS-KS code simulation. Loss of residual heat removal (RHR) and steam generator tube rupture (SGTR) simulations were conducted for the postulated abnormal events. Loss of RHR results showed the reactor coolant system (RCS) temperature increase, which can damage the reactor coolant pump (RCP)s by its cavitation. The SGTR results indicated a void formation in the RCS interior by the decrease in pressurizer (PZR) pressure, which can cause surface exposure and tripping of the RCPs unless proper actions are taken before the required pressure limit is achieved.

Validation of interfacial area concentration model for simulating bubbly and cap/slug flow behaviors in large diameter pipes using LSTF, ATLAS, and horizontal pipe experiment data

This study discusses Interfacial Area Concentration (IAC) models developed for bubbly and cap/slug flow regimes in a large diameter pipe incorporating recently developed small bubble fraction correlations in various flow conditions. The main focus of this study was the robust simulation of the behaviors of bubbly and cap/slug flows in large diameter pipes observed in Large Scale Test Facility (LSTF), Advanced Thermal-hydraulic test Loop for Accident Simulation (ATLAS), and horizontal flow test facility. Since the IAC was one of the key parameters to achieve this, an IAC model for large diameter pipes was first derived based on the existing Hibiki-Ishii IAC model for bubbly flow, and the Shen-Hibiki IAC model for cap bubbly-to-churn flow. The volume fraction of group-1 bubbles, i.e., small spherical bubbles, in the IAC model, which have been developed based on the assumption that the group-1 bubbles exponentially decay in the slug flow regime, was refined. As a result, a more physical model of the group-1 bubble number density that remains constant in the cap/slug flow regime was imposed for large diameter pipes. The validity of the IAC model for large diameter pipes in predicting the interfacial drag force between gas and liquid phases in large pipes of multiple test facilities was then examined. The validation of the proposed IAC model was conducted through simulations of 1) the loss of Residual Heat Removal (RHR) during the mid-loop operation of the LSTF, 2) the 6-inch Small Break Loss of Coolant Accident (SBLOCA) of the ATLAS, and 3) the two-phase flow behaviors under a wide range of fluid conditions in the horizontal pipe test facility. Improved results were obtained with the proposed model as it indicated that there was good agreement between the calculated results and the experimental data throughout the whole transient.

Application of BEPU method to loss of RHR event during mid-loop operation: Uncertainty quantification of RELAP5 model

In recent years, best estimate plus uncertainty (BEPU) method has become more important, and the guidelines on the process have been developed. However, there are not many specific reports on the probability distribution of the model uncertainty that greatly affects the evaluation result. Also, there are not many application cases of the BEPU method for the events of the PWR mid-loop operation which is suggested to be important in terms of probabilistic risk assessment (PRA). Therefore, this report describes the results of the quantification of the RELAP5/MOD3.2 model uncertainties related to the loss of residual heat removal (RHR) event during the PWR mid-loop operation. Based on the experimental literatures, the probability distributions of the model uncertainty were developed, which contributes to statistical analyses and research on the model reliability.

ASTEC V2.0 computational evaluation of source term and its chemical forms under accidental conditions during mid-loop operation

The possibility of an accident or component failure during mid-loop operation has been identified in probabilistic safety studies as a major contributor to core melt frequency and source term risk. The fission products release and transport to the containment has been analyzed during mid-loop operation of a reference PWR 1000 MWe reactor using the severe accident integral code ASTEC V2.0. The analyses have been performed considering the loss of residual heat removal (RHR) system at various times after reactor shutdown for the reactor vessel configuration with the removed upper head (open reactor). In this configuration, the possible air ingress can have an impact on safety such as accelerated oxidation and increased volatility of certain FPs (particularly iodine and ruthenium). Sensitivity calculations have been performed in terms of air ingress simulation with a different intensity. Besides equilibrium chemistry model, most of the calculations have also used a limited kinetics model. The study has shown that without air ingress the only predicted gaseous form of iodine is HI (≤7.4% of the total mass of iodine released from core) and no gaseous RuO 4 is created. Sensitivity calculations have illustrated that the gross fraction of gaseous iodine (I 2 + HOI + HI) has an increased trend with growth of air ingress intensity and with the duration of sequence evolution. In most oxidative atmosphere the gross iodine gaseous fraction could increase by a factor form of two to several times as compared to the corresponding case without air ingress (particularly due to I 2 persistence). Creation of gaseous RuO 4 is sensitive to carrier gas temperature; therefore a considerable fraction (≤3%) is predicted only in the sensitivity cases with the shortest time of loss of RHR after reactor scram.

Thermal hydraulic behaviors during loss of RHR system at mid-loop operation of Chinese 300 MWe PWR NPP

Thermal hydraulic behavior of nuclear power plant (NPP) is analyzed by using mechanistic computer code for loss of residual heat removal (RHR) system during mid-loop operation of Chinese 300 MWe two-loop pressurized water reactor is presented. In the absence of recovery of RHR or other accident management measures, the reactor core will be uncovered for a long term resulting in core heat-up, degradation and relocation to the lower plenum. The effectiveness of available mitigate measures, such as safety injection system, gravity feed from refueling water storage tank (RWST) and steam generator (SG) reflux-condensation, are investigated. Coolant injection is highly effective in halting the accident progression and make the core recovered. The cooling capability of SG reflux-condensation has a relationship with different availabilities of steam generators and decay heat power. 6 days after shutdown, 2SG operation can keep the water level at mid-line of hot leg. 12 days after shutdown, both 2SG operation and 1SG operation can keep the water level at mid-line of hot leg. The analyses also indicate that the cooling mechanism of safety injection system is more effective than gravity feed from RWST and SG reflux-condensation. Through confirming the success criteria of SG reflux-condensation, time windows can be devided. Then, event trees for loss of RHR system under mid-loop operation are built with considering the analysis results and abnormal procedure.

Accident management following loss of residual heat removal during mid-loop operation in a Westinghouse two-loop PWR

The possibility of an accident or component failure during mid-loop operation has been identified in probabilistic studies as a major contributor to core melt frequency and source term risk during shutdown conditions. The wide range of plant states encountered and the unavailability of certain safety features make it difficult to guarantee that safety systems operation will always be sufficient to terminate the accident evolution. In this context analyses are performed using MELCOR 1.8.5 for loss of residual heat removal (RHR) at various times during mid-loop operation of a Westinghouse two-loop PWR. In the absence of recovery of RHR or other accident management (AM) measures, the sequences necessarily lead to a long term core uncovery, heat-up and degradation, loss of geometry and eventual failure of the reactor pressure vessel (RPV). The results show an extensive time window before uncovery and additionally before core damage, which increase progressively with increasing time after shutdown at which loss of RHR occurs. Significant oxidation of the cladding may result in concentrations of hydrogen sufficient for deflagration. The slow evolution implies an opportunity for the plant operators to initiate AM measures even after core uncovery has started. The analyses indicate a substantial time window during the uncovery within which the injection can recover the core without damage. The upper end of the window is determined by the temperature at which heat from cladding oxidation becomes a dominant factor, marking a critical point for the effectiveness of this recovery mode. The results provide confidence in the inherent robustness of the plant with respect to accident sequences of this type.

Evaluation of Reflux Condensation Heat Transfer of Steam-Air Mixtures under Gas-Liquid Countercurrent Flow in a Vertical Tube

Reflux condensation in steam generator (SG) is one of the major heat removal mechanisms in a loss of residual heat removal (RHR) system event in mid-loop operation during a pressurized water reactor (PWR) plant outage. In order to evaluate the effectiveness of reflux condensation in SG U-tubes, the condensation heat transfer characteristics in the presence of a non-condensable gas must be clarified. Local temperature data were previously measured for steam-air mixtures under gas-liquid countercurrent flow, in a vertical tube of inner diameter 19.3 mm. In this study, local reflux heat transfer coefficients were calculated by evaluating steam flow rate profile along the tube at low heat fluxes assuming saturated steam conditions and empirical correlations were derived. The correlations are valid over a range 2–9,000W/m2-K for 0.1–0.4MPa, 0.014–0.2 air mass fraction. The axial distributions of the steam—air mixture temperatures calculated using the correlation agreed well with the measured results and validity of evaluation methods and the correlation were verified.

Evaluation of Reflux Condensation Heat Transfer of Steam-Air Mixtures under Gas-Liquid Countercurrent Flow in a Vertical Tube

Reflux condensation in steam generator (SG) is one of the major heat removal mechanisms in a loss of residual heat removal (RHR) system event in mid-loop operation during a pressurized water reactor (PWR) plant outage. In order to evaluate the effectiveness of reflux condensation in SG U-tubes, the condensation heat transfer characteristics in the presence of a non-condensable gas must be clarified. Local temperature data were previously measured for steam-air mixtures under gas-liquid countercurrent flow, in a vertical tube of inner diameter 19.3 mm. In this study, local reflux heat transfer coefficients were calculated by evaluating steam flow rate profile along the tube at low heat fluxes assuming saturated steam conditions and empirical correlations were derived. The correlations are valid over a range 2–9,000W/m2-K for 0.1–0.4MPa, 0.014–0.2 air mass fraction. The axial distributions of the steam—air mixture temperatures calculated using the correlation agreed well with the measured results and validity of evaluation methods and the correlation were verified.

Assessment and Application of the ROSE Code for Reactor Outage Thermal-Hydraulic and Safety Analysis

The currently available tools, such as RELAP5, RETRAN, and others, cannot easily and correctly perform the task of analyzing the system behavior during plant outages. Therefore, a medium-sized program aiming at reactor outage simulation and evaluation, such as midloop operation (MLO) with loss of residual heat removal (RHR), has been developed. Important thermal-hydraulic processes involved during MLO with loss of RHR can be properly simulated by the newly developed reactor outage simulation and evaluation (ROSE) code. The two-region approach with a modified two-fluid model has been adopted to be the theoretical basis of the ROSE code.To verify the analytical model in the first step, posttest calculations against the integral midloop experiments with loss of RHR have been performed. The excellent simulation capacity of the ROSE code against the Institute of Nuclear Energy Research Integral System Test Facility test data is demonstrated. To further mature the ROSE code in simulating a full-sized pressurized water reactor, assessment against the WGOTHIC code and the Maanshan momentary-loss-of-RHR event has been undertaken. The successfully assessed ROSE code is then applied to evaluate the abnormal operation procedure (AOP) with loss of RHR during MLO (AOP 537.4) for the Maanshan plant. The ROSE code also has been successfully transplanted into the Maanshan training simulator to support operator training. How the simulator was upgraded by the ROSE code for MLO will be presented in the future.

Code Development and Assessment for Reactor Outage Thermal-Hydraulic and Safety Analysis – Midloop Operation with Loss of Residual Heat Removal

Although only a few percent of residual power remains during plant outages, the associated risk of core uncovery and corresponding fuel overheating has been identified to be relatively high, particularly under midloop operation (MLO) in pressurized water reactors. However, to analyze the system behavior during outages, the tools currently available, such as RELAP5, RETRAN, etc., cannot easily perform the task. Therefore, a medium-sized program aiming at reactor outage simulation and evaluation, such as MLO with the loss of residual heat removal (RHR), was developed. All important thermal-hydraulic processes involved during MLO with the loss of RHR will be properly simulated by the newly developed reactor outage simulation and evaluation (ROSE) code. Important processes during MLO with loss of RHR involve a pressurizer insurge caused by the hot-leg flooding, reflux condensation, liquid holdup inside the steam generator, loop-seal clearance, core-level depression, etc. Since the accuracy of the pressure distribution from the classical nodal momentum approach will be degraded when the system is stratified and under atmospheric pressure, the two-region approach with a modified two-fluid model will be the theoretical basis of the new program to analyze the nuclear steam supply system during plant outages. To verify the analytical model in the first step, posttest calculations against the closed integral midloop experiments with loss of RHR were performed. The excellent simulation capacity of the ROSE code against the Institute of Nuclear Energy Research Integral System Test Facility (IIST) test data is demonstrated.

98/03823 Application of nonuniform-line theory to the simulation of electromagnetic transients in power systems

Loss of residual heat removal has occurred in several PWRs during mid-loop operation after plant shutdown, and is now recognized as an accident situation whose relevant physical phenomena require improved understanding. Several cases have been selected for investigation in the frame of the BETHSY program, with the main purpose of providing an experimental basis for the assessment of safety codes which have been so far especially involved with accident transients initiated during full-power operation.The major results from four experiments are presented in this paper, which addresses two main kinds of physical problem: mechanisms for, and rate of, primary mass inventory reduction in the case of manways opened at various locations in the primary coolant system; cooling of the plant when the primary is only half-open through vent paths and one steam generator is available to remove decay heat in reflux condenser mode with the presence of non-condensable gas.

Investigation of System Responses of the Maanshan Nuclear Power Plant to the Loss of Residual Heat Removal During Midloop Operation Using a RELAP5/MOD3 Simulation

Because several abnormal incidents involving the loss of the residual heat removal (RHR) system during refueling and maintenance outages have occurred in pressurized water reactors, the importance of investigating the physical phenomena with respect to these events has been recognized. RHR cooling is a major means to remove core decay heat after a reactor power plant shutdown. If the RHR system is lost and an alternate means for heat removal cannot be established in time, the core will boil off, and the primary system will be pressurized, which potentially results in uncovering of the fuel rods and failure of the temporary boundaries. The objective of this paper is to simulate the Maanshan nuclear power plant (MNPP) responses to the loss of the RHR system during midloop operation under variable outage conditions. Without gravity feed, the current investigation concentrates on the effects of different liquid levels, the existence of vents, and the number of active steam generators. Based on the simulated results, the total heat removal capability of one active steam generator and the pressurizer venting process is sufficient to remove core decay heat of 11.1 MW, which corresponds to the power level 3 days after plant shutdown, in the event that RHR cooling fails during midloop operation. The primary system will be stabilized, and the pressure throughout the transient will not exceed the design pressure of the nozzle dams or the temporary seals. The heat removal capability of the pressurizer vent plays a crucial role in system pressurization during loss of RHR and in the severity of this event, as shown by the calculated results of the open and closed MNPP, respectively. If only one or two active steam generators serve as an alternate cooling means, the increased pressure will exceed the design criteria of the temporary low-pressure boundaries. Then, for the closed conditions of MNPP, the loss-of-RHR event during midloop operation has the potential to induce another loss-of-coolant accident and to cause more serious consequences.

Investigation of mid-loop operation with loss of RHR at INER integral system test (IIST) facility

A series of seven experiments was conducted at the INER integral system test (IIST) facility to investigate the thermal hydraulics of the PWR mid-loop operations with loss of residual heat removal. The measured results of the controlling phenomena for several different plant configurations and core powers are exhibited in explicit curves as a function of time. It is found that in a closed system with higher core decay heat, reflux condensation as well as upflow side flooding and both-side liquid holdup may occur in the steam generator U-tubes. A detailed description of these mechanisms is discussed in this paper.

An Estimation of Core Damage Frequency of a Pressurized Water Reactor During Midloop Operation Due to Loss of Residual Heat Removal

The core damage frequency caused by loss of residual heat removal (RHR) events was assessed during midloop operation of a Westinghouse-designed three-loop pressurized water reactor. The assessment considers two types of outages (refueling and drained maintenance) and uses failure data collected specifically for shutdown condition. Event trees were developed for five categories of loss of RHR events. Human actions to mitigate the loss of RHR events were identified and human error probabilities were quantified using the human cognitive reliability (HCR) and the technique for human error rate prediction (THERP) models. The results showed that the core damage frequency caused by loss of RHR events during midloop operation was 3.4 {times} 10{sup {minus}5} per year. The results also showed that the core damage frequency can be reduced significantly by removing a pressurizer safety valve before entering midloop operation. The establishment of reflux cooling, i.e., decay heat removal through the steam generator secondary side, also plays an important role in mitigating the loss of RHR events during midloop operation.

RELAP5/MOD3 Simulation of the Loss of the Residual Heat Removal System During a Midloop Operation Experiment Conducted at the ROSA-IV Large-Scale Test Facility

A simulation of the loss of residual heat removal (RHR) system during midloop operations was performed using the RELAP5/MOD3 thermal-hydraulic code. The experiment was conducted at the Rig of Safety Assessment (ROSA)-IV/Large-Scale Test Facility. The experiment involved a 5% cold-leg break along with the loss of the RHR system. The transient was simulated for 3040 s. Core boiling and subsequent primary system pressurization occurred after the initiation of the transient. There was a good agreement between the measured and the calculated data until the loop seal clearing (LSC). It was found that the steam condensation was underpredicted in the calculations. This caused the calculated data after the LSC to differ from that of the measured data. The core rod surface temperature excursion around the occurrence of the LSC was not calculated. Overall, there was good qualitative agreement between the measured and the calculated data. The calculations, performed on the CRAY-YMP supercomputer, took over 60 h of CPU time for a transient of 51 min.

Loss of residual heat removal during mid-loop operation: BETHSY experiments

Loss of residual heat removal has occurred in several PWRs during mid-loop operation after plant shutdown, and is now recognized as an accident situation whose relevant physical phenomena require improved understanding. Several cases have been selected for investigation in the frame of the BETHSY program, with the main purpose of providing an experimental basis for the assessment of safety codes which have been so far especially involved with accident transients initiated during full-power operation. The major results from four experiments are presented in this paper, which addresses two main kinds of physical problem: mechanisms for, and rate of, primary mass inventory reduction in the case of manways opened at various locations in the primary coolant system; cooling of the plant when the primary is only half-open through vent paths and one steam generator is available to remove decay heat in reflux condenser mode with the presence of non-condensable gas.

Simulation of Loss of RHR During Midloop Operations and the Role of Steam Generators in Decay Heat Removal Using the RELAP5/MOD3 Code

Loss of residual heat removal during midloop operations was simulated for a typical four-loop pressurized water reactor operated under reduced inventory level using the RELAP5/MOD3 thermal-hydrauli...

Core Boiling During Midloop Operation

The reactor coolant system (RCS) water level is reduced during each refueling at some plants. Decreasing the level below the top of the loop piping (midloop operation) may be necessary to work on unisolable RCS loop components. A loss of residual heat removal (RHR) under these conditions can be serious due to the reduced water inventory, air in the RCS, and openings in the RCS loops. Under certain conditions, a loss of RHR could lead to rapid core uncovery and potential fuel damage.Core boiling due to a loss of RHR during midloop operation has received little attention until recently. The transient involves complex phenomena induced by core boiling, such as inventory loss from RCS openings and differences between the downcomer and upper plenum water levels, with the reactor vessel acting like a manometer. These phenomena cannot be easily evaluated without a versatile thermal-hydraulic computer code such as RETRAN.Yankee Atomic Electric Company’s RETRAN analysis of these phenomena reveals that the time to core uncovery is shortened by the loss of coolant through RCS openings and the manometer behavior of the reactor vessel water level. This analysis points out some limitations in applying the RETRAN code to this transient. However, the results are confirmed by a Westinghouse report issued after the completion of this analysis.