Low Pressure Safety Injection Research Articles

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.

Simulation of Branch Pipes Fracture Condition in Low-Pressure Safety Injection System of PWR

In the low-pressure injection system (LHSI) of the PWR unit, two branch pipes of the recirculation pipeline are vulnerable to fracture during minimal-flow operation. Flowmaster was used to create a hydraulic model for the LHSI minimal-flow operation and to simulate the branch double-end shear fracture condition in order to predict the leakage flow volume and the water level drop rate of the refueling water storage tank (RWST) in order to analyze the impact of branch pipe fracture on the water leakage of the RWST. According to the study results, there is very little deviation between the operating condition of the system as it actually is and the Flowmaster model, and the leakage volume following branch pipe fracture is considerably less than that of the RWST. Leakage volume per minute is around 0.025% of the RWST’s overall volume, which has negligible influence on the unit’s safety. Based on the results of the simulation, this study suggests LHSI system operation recommendations.

Simulation research on passive safety injection system of marine nuclear power plant based on compressed gas

In case of a large or medium LOCA accident of marine nuclear power plant, the loop pressure drops rapidly. In case of power failure of the whole ship at this time, the active safety injection system will lose its function because the safety injection pump needs AC power, and the core is at risk of melting. A passive marine safety injection system based on high-pressure compressed gas is designed for the superposition of large and medium LOCA accidents and the whole ship's power failure accidents, including the safety injection subsystem of medium and low pressure and the safety injection subsystem of recirculation. The energy required for safety injection is provided by the potential energy of compressed gas, the pressure head is provided by the technology of gas-liquid booster pump, and the collection signals and valves are controlled by the technology of intelligent fluid control unit with small power consumption. The power supply used in the system is provided by DC 24 V battery. The mathematical and physical models of accumulator, gas storage tank, gas-liquid booster pump, reactor body and its pipeline are established for the designed passive safety injection system. The operation and safety characteristics of medium pressure safety injection, low pressure safety injection and recycling safety injection under LOCA accident conditions are simulated. The safety injection time and flow rate of three safety injection pressures are simulated by medium pressure safety injection, the safety injection time and peak temperature are simulated by low pressure safety injection, and the pressure and system temperature are simulated by recirculation safety injection. The simulation results show that the marine passive safety injection system designed under LOCA accident conditions can meet the requirements of core cooling.

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.

Preliminary neutronic/thermal-hydraulic evaluation and safety system optimization of PWR loaded with fully ceramic microencapsulated fuel

As one of the Accident Tolerant Fuels (ATFs), Fully Ceramic Microencapsulated (FCM) presents advantageous safety performance in the severe accidents compared with UO2 owing to its high melting point, superior heat transfer ability and proliferation resistance capability. In this study, preliminary neutronic and thermal-hydraulic analysis of Chinese improved three-loop Pressurized Water Reactor (PWR) with FCM pellet and SiC cladding (FCM-SiC) were conducted to demonstrate the fuel performance of ATFs. First, a three dimensional full core model loaded with FCM-SiC was developed to acquire the cycle length, power distribution and isotopic inventories. And low reactor power and high fuel enrichment were used to satisfy the annual reload requirement. Then, the reactor thermal-hydraulic safety evaluation was carried out by the modified subchannel and system analysis codes, which depicts lower Fuel Centerline Temperature (FCT) and Peak Cladding Temperature (PCT) in FCM-SiC reactor compared to the reactor with traditional fuels. The Minimum Departure from Nucleate Boiling Ratio (MDNBR) of cladding is much higher than the criterion of 1.3 as well. Furthermore, the reactor safety system was simplified and optimized by taking full advantage of the preponderance of ATFs, and the results show that the reactor of FCM-SiC can operate safely with only 10% initial flowrate of Low Pressure Safety Injection (LPSI) or without Accumulators (ACCs) during Large Break Loss of Coolant Accident (LBLOCA), while the conventional zircaloy is failed under the same conditions. This work could provide guidelines for the optimized design of the reactor safety system in the future.

Steam condensation simulation in a scaled IRWST-ADS simulator with GOTHIC 8.1

The AP1000® advanced reactor passive safety systems are based on natural phenomena to ensure the containment integrity during an accident. One of the most important passive systems is the Passive Core Cooling System (PXS) which includes the In-containment Refueling Water Storage Tank (IRWST), a pool that serves as a heat sink for the Passive Residual Heat Removal Heat Exchanger (PRHR) and the Automatic Depressurization System (ADS).The thermal-hydraulic behavior of the pool is quite complex under accidental conditions: the pool receives steam at high velocity and temperature from the Reactor Cooling System (RCS), the pool is the heat sink of the PRHR, and eventually, it is the source for the low pressure safety injection. In order to study the behavior of the AP1000 IRWST during a steam discharge from the ADS at different stages of a SBLOCA, an experimental test rig was constructed in cooperation with the Technical University of Munich (TUM). The results obtained in the IRWST-ADS experiment were used to test the capability of GOTHIC 8.1 to simulate the phenomena observed experiments such as thermal stratification, direct contact condensation, and flashing. The GOTHIC model is based on two control volumes, one represents the channel test and the other the laboratory room. These control volumes are connected with a 3D-Connector between each other and with flow-paths to the boundary conditions. Pressure boundary conditions has been applied to the laboratory room control volume to maintain a constant pressure and temperature. Simulations were able to reproduce major occurring phenomena. Thermal stratification and flashing were predicted plausibly, however the direct contact condensation of steam and void fraction distribution are not perfectly simulated probably because of the selected mesh size. A sensitivity analysis has been conducted to the discretization scheme, the turbulence model, and the heat transfer correlation in order to test their effect upon GOTHIC model simulation.

Modeling and analysis of selected organization for economic cooperation and development PKL-3 station blackout experiments using TRACE

Abstract A series of tests dedicated to station blackout (SBO) accident scenarios have been recently performed at the Primarkreislauf-Versuchsanlage (primary coolant loop test facility; PKL) facility in the framework of the OECD/NEA PKL-3 project. These investigations address current safety issues related to beyond design basis accident transients with significant core heat up. This work presents a detailed analysis using the best estimate thermal–hydraulic code TRACE (v5.0 Patch4) of different SBO scenarios conducted at the PKL facility; failures of high- and low-pressure safety injection systems together with steam generator (SG) feedwater supply are considered, thus calling for adequate accident management actions and timely implementation of alternative emergency cooling procedures to prevent core meltdown. The presented analysis evaluates the capability of the applied TRACE model of the PKL facility to correctly capture the sequences of events in the different SBO scenarios, namely the SBO tests H2.1, H2.2 run 1 and H2.2 run 2, including symmetric or asymmetric secondary side depressurization, primary side depressurization, accumulator (ACC) injection in the cold legs and secondary side feeding with mobile pump and/or primary side emergency core coolant injection from the fuel pool cooling pump. This study is focused specifically on the prediction of the core exit temperature, which drives the execution of the most relevant accident management actions. This work presents, in particular, the key improvements made to the TRACE model that helped to improve the code predictions, including the modeling of dynamical heat losses, the nodalization of SGs' heat exchanger tubes and the ACCs. Another relevant aspect of this work is to evaluate how well the model simulations of the three different scenarios qualitatively and quantitatively capture the trends and results exhibited by the actual experiments. For instance, how the number of SGs considered for secondary side depressurization affects the heat transfer from primary side; how the discharge capacity of the pressurizer relief valve affects the dynamics of the transient; how ACC initial pressure and nitrogen release affect the grace time between ACC injection and subsequent core heat up; and how well the alternative feeding modes of the secondary and/or primary side with mobile injection pumps affect core quenching and ensure stable long-term core cooling under controlled boiling conditions.

Development of a feed-and-bleed operation strategy with hybrid-SIT under low pressure condition of PWR

While safety functions in current nuclear power plants are mainly provided by active safety systems, recently passive safety systems are being combined with the active systems to strengthen accident mitigation capability and therefore enhance overall plant safety. To this end, securing long-term cooling of the core is of particular importance. This study considers the hybrid safety injection tank (H-SIT), a passive injection system, as a target component to develop a long-term cooling strategy using active and passive systems concurrently.In the feed-and-bleed (F&B) operation, one of the important long-term cooling strategies to maintain core safety in pressurized water reactors, low pressure safety injection (LPSI) pumps are typically considered inoperable as depressurization is first required, which leads to core dry-out before reaching LPSI operable pressure. This study investigates whether H-SITs, with the important design feature of passive coolant injection under any pressure condition of the primary coolant system, can make up the core during depressurization thereby allowing LPSI pumps to be used in F&B operation as an additional means of long-term cooling. The effectiveness of the H-SITs is verified using thermal-hydraulic simulations, and based on the results a novel F&B operation strategy with H-SITs and LPSI pumps is developed. A probabilistic safety assessment (PSA) model is then developed in order to assess the risk effect of the suggested strategy. PSA results demonstrate that the proposed strategy lowers core damage frequency in the target plant by 5.0 percent in the case of a small break loss of coolant accident. Total core damage frequency of the plant decreases by 4.8 percent compared to the reference model, and also lowers the number of cutsets by around 13 percent from the original value.

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.

Simulation of advanced accumulator and its application in CPR1000 LBLOCA analysis

The advanced accumulator is designed to improve the safety and reliability of CPR1000 by providing a small injection flow to keep the reactor core in flooded condition. Thus, the core still stays in a cooling state without the intervention of low pressure safety injection and the startup grace time of the low pressure safety injection pump can be greatly extended.A new model for the advanced accumulator, which is based on the basic conservation equations, is developed and incorporated into RELAP5/MOD 3.3. The simulation of the advanced accumulator can be carried out and results show that the behavior of the advanced accumulator satisfied its primary design target. There is a large flow in the advanced accumulator at the initial stage. When the accumulator water level is lower than the stand pipe, a vortex appears in the damper, which results in a large pressure drop and a small flow. And then the sensitivity analysis is performed and the major factors which affected the flow rate of the advanced accumulator were obtained, including the damper diameter, the initial volume ratio of the water and the nitrogen and the diameter ratio of the standpipe and the small pipe.Additionally, the primary coolant loop cold leg double-ended guillotine break LBLOCA in CPR1000 with advanced accumulator is analyzed. The results show that the criterion for maximum cladding temperature limit (1477K) (NRC, 1992) can be met ever with 200s after the startup of the low pressure safety injection. From this point of view, passive advanced accumulator can strive a longer grace time for LPSI. Thus the reliability, safety and economy of the reactor system can be improved.

Large Break Loss-of-Coolant Accident Analysis of VVER-1000 Reactor Using CATHARE Code

In the safety analysis of nuclear power plants, large break loss-of-coolant accidents (LB LOCAs) continue to be one of the major issues. In this study, the latest version of the French best-estimate computer code CATHARE2 V2.5_1 Mod6.1 was used in order to predict the thermal-hydraulic phenomena in the VVER-1000 reactor during LB LOCA. This type of reactor is in operation and under construction in several countries including Russia, Ukraine, Bulgaria, Czech Republic, China, and India.The paper first presents the CATHARE modeling of a VVER-1000 reactor, including the core, the vessel, the primary and secondary circuits with the pressurizer, the main circulation pumps, the horizontal steam generators, and the steam lines. The emergency core cooling system (ECCS) is presented also with hydroaccumulators and high- and low-pressure safety injection pumps. The break, located in the cold leg and close to the reactor, is represented by the RUPTURE module of CATHARE, which modelizes a double-ended guillotine type of break.In the current calculations, the bottom-up and top-down reflooding play a very important role, compared to some western pressurized water reactors, because of the ECCS injection into the downcomer and upper plenum of the VVER-1000. That is why special attention is paid to the validation of the CATHARE reflooding model, based on the REWET-II experiment devoted to VVER. In CATHARE, a special two-dimensional reflooding module has been developed, which takes into account the radial and axial conduction in the vicinity of the quench front. The assessment based on the REWET-II facility shows the ability of the code to predict the progression of two simultaneous quench fronts.Then, the paper presents the results of the LB LOCA analysis for VVER-1000 (reference case and sensitivity study), describing the blowdown, refill, and reflood phases of the accident. Problems related to the countercurrent flow limitation (CCFL) phenomena at the core outlet are also considered. The results of the calculations show reasonable prediction of the basic parameters important to the safety of the plant, such as quench front progression, fuel-cladding temperatures, break flows, pressure behavior, etc. As expected, the temperatures—despite the conservative character of the modeling—remain below the safety criteria. It should be noted that the calculated fuel wall temperatures are very sensitive to the proper selection of the reflooding parameters and CCFL and need skillful choice of the reflooding conditions.

DETAILED EVALUATION OF THE IN-VESSEL SEVERE ACCIDENT MANAGEMENT STRATEGY FOR SBLOCA USING SCDAP/RELAP5

As part of an evaluation for an in-vessel severe accident management strategy, a coolant injection into the reactor vessel under depressurization of the reactor coolant system (RCS) has been evaluated in detail using the SCDAP/RELAP5 computer code. A high-pressure sequence of a small break loss of coolant accident (SBLOCA) has been analyzed in the Optimized Power Reactor (OPR) 1000. The SCDAP/RELAP5 results have shown that safety injection timing and capacity with RCS depressurization timing and capacity are very effective on the reactor vessel failure during a severe accident. Only one train operation of the high pressure safety injection (HPSI) for 30,000 seconds with RCS depressurization prevents failure of the reactor vessel. In this case, the operation of only the low pressure safety injection (LPSI) without a HPSI does not prevent failure of the reactor vessel.

An evaluation of the severe accident management strategies for CANDU-6 type plants

This paper provides an evaluation of the mitigation effects for the severe accident management strategies of the Wolsong plants which are typical CANDU-6 type reactors. The evaluation includes the effect of the following six mitigation strategies: (1) injection into the primary heat transport system (PHTS), (2) injection into the calandria vessel, (3) injection into the calandria vault, (4) reduction of the fission product release, (5) control of the reactor building condition, (6) reduction of the reactor building hydrogen. The tested scenario is a loss of coolant accident with a small out-of-core break, and the thermal hydraulic and severe accident phenomenological analyses were implemented by using the ISAAC computer program. The calculation results show that the most effective means for a primary decay heat removal is a low pressure safety injection, that for a calandria vessel integrity is an end-shield cooling injection, and that for a reactor building integrity is a pressure control via local air coolers. Besides the above, the usefulness of each safety component was evaluated in this analysis.

Detailed evaluation of coolant injection into the reactor vessel with RCS depressurization for high pressure sequences

A coolant injection into the reactor vessel with depressurization of the reactor coolant system (RCS) has been evaluated as part of the evaluation for a strategy of the severe accident management guidance (SAMG). Two high pressure sequences of a small break loss of coolant accident (LOCA) without safety injection (SI) and a total loss of feedwater (LOFW) accident in Optimized Power Reactor (OPR)1000 have been analyzed by using the SCDAP/RELAP5 computer code. The SCDAP/RELAP5 results have shown that only one train operation of a high pressure safety injection at 30,000 s with indirect RCS depressurization by using one condenser dump valve (CDV) at 6 min after implementation of the SAMG prevents reactor vessel failure for the small break LOCA without SI. In this case, only one train operation of the low pressure safety injection (LPSI) without the high pressure safety injection (HPSI) does not prevent reactor vessel failure. Only one train operation of the HPSI at 20,208 s with direct RCS depressurization by using two SDS valves at 40 min after an initial opening of the safety relief valve (SRV) prevents reactor vessel failure for the total LOFW.

Commissioning of the ATLAS thermal-hydraulic integral test facility

KAERI recently constructed a new thermal-hydraulic integral test facility for advanced pressurized water reactors (PWRs) – ATLAS. The ATLAS facility has the following characteristics: (a) 1/2-height&length, 1/288-volume, and full pressure simulation of APR1400, (b) maintaining a geometrical similarity with APR1400 including 2(hot legs) × 4(cold legs) reactor coolant loops, direct vessel injection (DVI) of emergency core cooling water, integrated annular downcomer, etc., (c) incorporation of specific design characteristics of OPR1000 such as cold leg injection and low-pressure safety injection pumps, (d) maximum 10% of the scaled nominal core power. The ATLAS will mainly be used to simulate various accident and transient scenarios for evolutionary PWRs, OPR1000 and APR1400: the simulation capability of broad scenarios including the reflood phase of a large-break loss-of-coolant accident (LOCA), small-break LOCA scenarios including DVI line breaks, a steam generator tube rupture, a main steam line break, a feed line break, a mid-loop operation, etc. The ATLAS is now in operation after an extensive series of commissioning tests in 2006.

Loss-of-Coolant Accident Aspects of the Combustion Engineering Advanced Light Water Reactor—System 80+

Combustion Engineering’s advanced light water reactor, System 80+, is an evolutionary upgrade of the proven System 80® nuclear steam supply system design. While both plants are rated at 3817 MW(thermal), System 80+ incorporates a number of design enhancements, including direct vessel injection for the safety injection (SI) system and other changes to the reactor cooling system.The results of a best-estimate small-break loss-of-coolant accident (LOCA) study that addresses utility investment protection concerns is presented. Specifically, the size piping break that can be tolerated without the liquid or two-phase fluid level falling below the top of the active core is addressed. Using best-estimate analytical procedures, and assuming no single failure, the active core remains covered with substantial margin for breaks up to 0.254-m (10-in.) diameter. This reduces the possibility of core damage due to a small LOCA.A large-break, cold-leg LOCA licensing analysis is also presented that addresses the reflood capability after the end of SI tank discharge without credit for a low-pressure SI pump system. This analysis confirms that the improved high-pressure SI system provides adequate reflood capability to satisfy the U.S. Nuclear Regulatory Commission LOCA licensing criteria.