Small Break Loss Of Coolant Accident Research Articles

Progression of Severe Accidents, Containment Performance, and Estimates of Releases of Radionuclides-Level-2 Probabilistic Safety Assessment IPHWR

Probabilistic safety assessment (PSA) of nuclear power plants is performed to yield insights into the safety, design, and performance of the plants and their potential environmental effects. This includes the identification of dominant risk contributors, determination of the vulnerabilities of plant and containment systems, and comparison of options for risk reduction. Three levels of PSA are recognized. Level-1 addresses the identification of plant failures leading to core damage and their frequencies of occurrence. Level-2 addresses the assessment of containment response leading together with level-1 results to the determination of containment release frequencies. A level-2 PSA analyses the challenges to the containment, the possible containment responses and their estimated probabilities, and an assessment of the consequent releases to the environment. Level-3 is the assessment of off-site consequences leading, together with the results of level-2 analysis, for estimation of public risks. A comprehensive level-2 PSA study of a 220 MWe Indian Pressurized Heavy Water Reactor (IPHWR) is performed to assess the challenges to the containment, the possible containment responses and their estimated probabilities, and consequent releases to the environment. The dominating sequences consist of small-break loss of coolant accident (SBLOCA) and station black out (SBO) followed by containment isolation failure. The results of this are used as an input for developing the severe accident management guidelines (SAMG) measures. All the SAMG measures incorporated in this study have been found as beneficial and resulted in reduced large early release frequency (LERF).

Comparison and analysis on transient characteristics of integral pressurized water reactors

In the present work, the similarities and differences of representative IPWRs (integral pressurized water reactor) are studied, and two typical reactor design schemes are summarized. To get a comprehensive understanding of their transient characteristics, SBO (station blackout) and SBLOCA (small break LOCA) are simulated and analyzed respectively by using Relap5/Mod3.2. The calculation results show that, both designs are effective in keeping reactor safe. However, the transient features of the two designs show significant differences. In the primary side passive safety system (PSS) connection design, PRHRS (passive residual heat removal system) shows a roughly congruent performance in removing residual heat under various accidents. While in secondary side PSS connection design, the capability of PRHRS is closely related to primary coolant circulation condition. In SBLOCA analysis, different design approach shows different primary coolant water inventory change trend. And primary PSS connection design could potentially keep reactor core well covered for a longer time.

Knowledge-based software design for Defense-in-Depth risk monitor system and application for AP1000

ABSTRACTAs part of the new risk monitor system, the software for the plant Defense-in-Depth (DiD) risk monitor system was designed based on the state-transition and finite-state machine, and then the knowledge-based software was developed by object-oriented method utilizing the Unified Modeling Language (UML). Currently, there are mainly two functions in the developed plant DiD risk monitor software that are knowledge-base editor which is used to model the system in a hierarchical manner and the interaction simulator that simulates the interactions between the different actors in the model. In this paper, a model for playing its behavior is called an Actor which is modeled at the top level. The passive safety AP1000 power plant was studied and the small-break loss-of-coolant accident (SBLOCA) design basis accident transient is modeled using the plant DiD risk monitor software. Furthermore, the simulation result is shown for the interactions between the actors which are defined in the plant DiD risk monitor system as PLANT actor, OPERATOR actor, and SUPERVISOR actor. This paper shows that it is feasible to model the nuclear power plant knowledge base using the software modeling technique. The software can make the large knowledge base for the nuclear power plant with small effort.

Simulation and Analysis of Small Break LOCA for AP1000 Using RELAP5-MV and Its Comparison with NOTRUMP Code

Many reactor safety simulation codes for nuclear power plants (NPPs) have been developed. However, it is very important to evaluate these codes by testing different accident scenarios in actual plant conditions. In reactor analysis, small break loss of coolant accident (SBLOCA) is an important safety issue. RELAP5-MV Visualized Modularization software is recognized as one of the best estimate transient simulation programs of light water reactors (LWR). RELAP5-MV has new options for improved modeling methods and interactive graphics display. Though the same models incorporated in RELAP5/MOD 4.0 are in RELAP5-MV, the significant difference of the latter is the interface for preparing the input deck. In this paper, RELAP5-MV is applied for the transient analysis of the primary system variation of thermal hydraulics parameters in primary loop under SBLOCA in AP1000 NPP. The upper limit of SBLOCA (10 inches) is simulated in the cold leg of the reactor and the calculations performed up to a transient time of 450,000.0 s. The results obtained from RELAP5-MV are in good agreement with those of NOTRUMP code obtained by Westinghouse when compared under the same conditions. It can be easily inferred that RELAP5-MV, in a similar manner to RELAP5/MOD4.0, is suitable for simulating a SBLOCA scenario.

SMART-ITL 1 계열 피동안전계통을 이용한 유동분사기 성능에 대한 실험연구

노심보충탱크 상부에 설치되는 유동분사기 형상에 따른 냉각수 주입특성 및 탱크 내에서의 열수력 현상 변화를 파악하기 위한 안전주입배관 2인치 파단 소형냉각재상실사고(SBLOCA) 모의시험이 잔열 및 피동잔열제거계통(PRHRS) 모의 없이 수행되었다. 두 가지 형상의 유동분사기를 설치하고 수행한 각각의 시험은 거의 유사한 초기 및 경계조건에서 수행되었으며, 이로 인해 반복시험에 대한 재현성이 충족되었다고 판단된다. 시험결과는 유동분사기의 종류(본 시험에서는 구멍의 개수에 해당)에 관계없이 유사한 열수력학적 거동을 보였으며, 초기 주입유량 관점에서는 구멍의 개수가 2배인 B형이 A형에 비해 좀 더 우수한 주입 성능을 보였다. 노심보충탱크 격리 밸브가 개방된 후 압력평형배관을 통해 유입되는 고온의 원자로냉각재는 상부 헤더에서 상대적으로 저온인 <TEX>$50^{\circ}C$</TEX> 물과 혼합되면서 증기 응축과 같은 상변화에 의한 압력 변동을 동반하는 다차원 열유동 현상을 일으키게 된다. 이로 인해 초반부 노심보충탱크 주입 유량은 상온운전 조건에서 보다는 작게 되고, 일정시간 경과 후에는 유사한 주입유량 특성을 보였다. In order to estimate the effect of flow distributors connected to an upper nozzle of CMT(Core Makeup Tank) on the thermal-hydraulic characteristics in the tank, a simplified 2 inch Small Break Loss of Coolant Accident(SBLOCA) was simulated by skipping the decay power and Passive Residual Heat Removal System(PRHRS) actuation. The CMT is a part of safety injection systems in the SMART (System Integrated Modular Advanced Reactor). Each test was performed with reliable boundary conditions. It means that the pressure distribution is provided with repeatable and reproducible behavior during SBLOCA simulations. The maximum flow rates were achieved at around 350 seconds after the initial opening of the isolation valve installed in CMT. After a short period of decreased flow rate, it attained a steady injection flow rate after about 1,250 seconds. This unstable injection period of the CMT coolant is due to the condensation of steam injected into the upper part of CMT. The steady injection flow rate was about 8.4% higher with B-type distributor than that with A-type distributor. The gravity injection during hot condition tests were in good agreement with that during cold condition tests except for the early stages.

Comparative study on aerosol removal by natural processes in containment in severe accident for AP1000 reactor

Focusing on aerosol removal by naturally occurring processes in containment in severe accident for AP1000, integral severe accident code MELCOR and rapid assessment method mentioned in NUREG/CR-6189 are utilized to study aerosol removal by natural processes, respectively. Three typical severe accidents, induced by large break loss of coolant accident (LBLOCA), small break loss of coolant accident (SBLOCA) and steam generator tube rupture (SGTR), respectively, are selected for the study. The results obtained by two methods were further compared in the following several aspects: efficiency of aerosol removal by natural processes, peak time of aerosol suspended in containment atmosphere, peak amount of aerosol suspended in containment atmosphere, time when aerosol removal efficiency by natural processes is up to 99.9%. It was further concluded that results obtained by rapid assessment with shorter calculation process are more conservative. The analysis results provide reference to assessment method selection of severe accident source term for AP1000 nuclear emergency.

Reflux condensation behavior in SBLOCA tests of ATLAS facility

The behavior of the reflux condensation heat transfer in a hot side steam generator (SG) U-tubes during a cold leg (CL) pipe and a direct vessel injection (DVI) line break in small break loss-of-coolant accident (SBLOCA) tests of the ATLAS facility was investigated including MARS code calculations. Among the SBLOCA tests, a 6″-CL pipe and 50%-DVI line break SBLOCA test were selected to investigate the behavior of the reflux condensation. A reflux condensation heat transfer seemed to occur from the time the SG U-tubes were half-empty to near the loop seal clearing (LSC). It was found that a transition regime existed between the reflux condensation heat transfer and reverse heat transfer. The remaining reflux condensate in SG U-tubes owing to the counter-current flow limit (CCFL) phenomenon and a separating effect of liquid carry-over and/or entrainment with steam moisturizing seemed to affect the thermal-hydraulic behavior of the transition regime. It was also found that the steam flowrate of the loop pipings and SG U-tubes seemed to have a strong effect on the duration time of the transition regime, e.g., a larger steam flowrate results in a longer duration. From a comparison of the reflux condensation behavior between the ATLAS tests and MARS code calculations, overall qualitative agreements were found between the two cases. The largest discrepancies were found in the SG inlet plenum water level between the two cases, and the authors suggest that the combination effects of the remaining reflux condensate in SG U-tubes and a separating effect of liquid carry-over and/or entrainment with steam moisturizing are the reasons for the discrepancies in the water levels of the SG inlet plenum, which seems to be difficult to simulate in the current version of the MARS-KS code.

Development of a LOCA safety analysis evaluation model for the Westinghouse Small Modular Reactor

The Westinghouse Small Modular Reactor (W-SMR) is an integral pressurized water reactor (PWR) with all major components located inside the reactor vessel, with a passive safety system that was derived from the AP1000® PWR passive safety design. The integral design of the reactor coolant system (RCS) effectively removes the postulated large break loss-of-coolant accident (LOCA). Therefore, small break LOCA (SBLOCA) is the most limiting design basis accident for the W-SMR. The LOCA safety analysis of the W-SMR requires an evaluation model (EM) that is adequate for a reactor design with passive safety systems. Development of the EM follows the Evaluation Model Development Assessment Process (EMDAP) by the United States Nuclear Regulatory Commission (US NRC). The first step in this process is the development of a SBLOCA Phenomena Identification and Ranking Table (PIRT), which identifies the figures of merit for the LOCA safety analysis and key thermal hydraulic phenomena pertinent to the passive safety design of the W-SMR. Next, an assessment matrix that consists of experiments from the AP600™ and AP1000 PWR development programs, University of California Berkeley (UCB) tube condensation tests, and the integral effects and separate effects tests for the W-SMR testing program is established to assess the evaluation model. The systems code used in the W-SMR LOCA evaluation model is the WCOBRA/TRAC-TF2 systems code, which has been extensively validated for previous LOCA applications. WCOBRA/TRAC-TF2 is being further developed to address the W-SMR EM PIRT. Among key thermal hydraulic phenomena, the assessment of the condensation on the containment wall and natural circulation/draining through the CMT are presented here as examples to demonstrate the W-SMR SBLOCA EM development, while all important phenomena are expected to be assessed using the WCOBRA/TRAC-TF2 systems code. Finally, the reference SBLOCA transient of W-SMR is presented to demonstrate the ability of the systems code to properly capture the key phenomena identified in the PIRT. It is expected that the development of an evaluation model for safety analysis of the Westinghouse SMR will facilitate its licensing and eventual commercialization.

Comparative Experiments to Assess the Effects of Accumulator Nitrogen Injection on Passive Core Cooling During Small Break LOCA

Abstract The accumulator is a passive safety injection device for emergency core cooling systems. As an important safety feature for providing a high-speed injection flow to the core by compressed nitrogen gas pressure during a loss-of-coolant accident (LOCA), the accumulator injects its precharged nitrogen into the system after its coolant has been emptied. Attention has been drawn to the possible negative effects caused by such a nitrogen injection in passive safety nuclear power plants. Although some experimental work on the nitrogen injection has been done, there have been no comparative tests in which the effects on the system responses and the core safety have been clearly assessed. In this study, a new thermal hydraulic integral test facility—the advanced core-cooling mechanism experiment (ACME)—was designed and constructed to support the CAP1400 safety review. The ACME test facility was used to study the nitrogen injection effects on the system responses to the small break loss-of-coolant accident LOCA (SBLOCA) transient. Two comparison test groups—a 2-inch cold leg break and a double-ended direct-vessel-injection (DEDVI) line break—were conducted. Each group consists of a nitrogen injection test and a nitrogen isolation comparison test with the same break conditions. To assess the nitrogen injection effects, the experimental data that are representative of the system responses and the core safety were compared and analyzed. The results of the comparison show that the effects of nitrogen injection on system responses and core safety are significantly different between the 2-inch and DEDVI breaks. The mechanisms of the different effects on the transient were also investigated. The amount of nitrogen injected, along with its heat absorption, was likewise evaluated in order to assess its effect on the system depressurization process. The results of the comparison and analyses in this study are important for recognizing and understanding the potential negative effects on the passive core cooling performance caused by nitrogen injection during the SBLOCA transient.

Fracture analysis of the set-in nozzle of a PWR reactor pressure vessel – Part 1: Determination of critical crack

In this study, fracture mechanics analysis of the set-in nozzle of a reactor pressure vessel (RPV) is being presented. The selected RPV belongs to a 300MW pressurized water reactor (PWR). For the purpose, a wide range of elliptical surface cracks was analyzed at the nozzle-cylinder intersection under two accidental conditions of the reactor: small-break loss of coolant accident (SB-LOCA), and Rancho-Seco transient (RST). It has been demonstrated that at the location being investigated ‘a=0.05t’ is the critical crack depth; where ‘a’ is the depth/minor axis of the crack and ‘t’ is the thickness of the wall.

Simulation of steam generator plugging tubes in a PWR to analyze the operating impact

A number of nuclear power plants (NPPs) with pressurized water reactors (PWR) in the world have replaced their steam generators (SG) due to degradation of the SG tubes caused by different problems. Several methods were attempted to correct the defects of the tubes, but eventually the only permanent solution was to plug them. The consequences of plugging the tubes are the decrease of heat transfer surface, the reduction of the flow area and subsequent reduction of the primary system mass flow and for a fraction of plugged tubes higher than a given value, the reduction of reactor output and economic losses. The objective of this paper is to analyze whether steam generator tube plugging has an impact in the effectiveness of accident management actions. An analysis with Relap5 Mod 3.3 patch03 for the Spanish reactor Ascó-2, a 3-loop 2940.6MWth Westinghouse PWR, in which plugging of steam generator tubes are simulated, is presented in order to find the limit for the adequate operation of the plant. Several steady state calculations were performed with different fractions of plugged SG tubes, by modeling the reduction of the primary to secondary heat transfer surface and the reduction of the primary coolant mass flow area in the tubes as well. The results of the analysis yield that plugging 12% of the SG tubes is around the limit for optimal reactor operation. To complete the study two events, in which the steam generators are used to cooldown the plant, were simulated to find out if the plugging of SGs tubes could influence the efficiency of the operator actions described in the emergency operating procedures (EOPs) to handle this kind of events. The selected events have been the rupture of one tube of a SG (SGTR) and a small break LOCA (SBLOCA). Two transient cases with no plugging at all (0%) and 12% SG tube plugging were performed. The actions of the corresponding EOPs for SG tube rupture and SBLOCA were coded in the input deck. The results show no significant impact on operator actions.

Experimental simulation of liquid entrainment in ADS-4 depressurization line in AP1000

The fourth stage Automatic Depressurization System is an important passive safety feature in Westinghouse AP1000 which enables controlled depressurization of reactor coolant system in small break LOCA. However, large amount of coolant can be carried to the containment via the ADS-4 branch entrainment and the upper plenum entrainment during the depressurization process, which poses great threats to core uncovering and melting. The automatic Depressurization and Entrainment TEst Loop (ADETEL) modeled after AP1000 with a scaling ratio of 1:5.6 was constructed to investigate the entrainment and depressurization behavior after the actuation of ADS-4 valves. The entrainment and depressurization features were investigated under different initial pressure, mixture liquid level in the pressure vessel and heating power. The entrainment deposition effect of the reactor internals was also investigated. The test data reveals that large amount of water are entrained through the ADS-4 branch line within a short period of time. The liquid entrainment rate and the reduced rate of the mixture liquid level in the pressure vessel increase with the initial system pressure. It is notable that the core uncovery was experienced when the initial pressure was set to 0.5 MPa in current experimental conditions. The reactor internals have little effect on the entrained mass and the mixture liquid levels in the pressure vessel.

Uncertainty in RELAP5/MOD3.2 calculations for interfacial drag in downward two-phase flow

RELAP5/MOD3.2 is a thermal-hydraulic system analysis code used to predict the response of nuclear reactor coolant systems in the event of certain accident scenarios. It is important that RELAP and other system analysis codes are able to accurately predict various two-phase flow phenomena, particularly the interfacial transfers between the liquid and gas phases. It is also important to understand how much uncertainty exists in these predictions due to uncertainties in the constitutive relations used to close the two-fluid model.In this paper, the uncertainty in the interfacial drag calculated by RELAP5/MOD3.2 due to errors in the drift-flux models used to close the model is evaluated and compared to the correlation developed by Goda et al. (2003). The case of downward flow is considered due to the importance of co-current and counter-current downward flow for predicting behavior in the downcomer of reactor systems during small-break Loss of Coolant Accidents (LOCAs) in nuclear reactor systems. The overall uncertainty in the interfacial force calculations due to error in the distribution parameter models were found to have a bias of +8.1% and error of 20.1% for the models used in RELAP5, and a bias of −30.8% and error of 23.1% for the correlation of Goda et al. (2003). However this analysis neglects the effects of compensating errors in the drift-flux parameters, as the drift velocity is assumed to be perfectly accurate. More physically meaningful results could be obtained if the distribution parameter and drift velocity were calculated directly from local phase concentration and velocity measurements, however no studies were available which included all of this information.

Analysis of the Core Exit Temperature and the Peak Cladding Temperature During a SBLOCA: Application to a Scaled-Up Model

During loss-of-coolant accidents (LOCAs), operators may start accident management (AM) actions when the core exit temperature (CET) measured by thermocouples exceeds a certain value. However, a significant time delay and temperature discrepancy in the superheat detection were observed in several facilities. This work is focused on clarifying CET thermocouple responses versus peak cladding temperature (PCT) and studying if the same physical phenomena are reproduced in two TRACE5 models with different geometry (a large-scale test facility (LSTF) and a scaled-up LSTF) during a pressure vessel (PV) upper head small break LOCA (SBLOCA). Results obtained show that the delay between the core uncover and the CET excursion is reproduced in both cases.

An experimental study of pool entrainment in high gas flux region

Pool entrainment is an important phenomenon in LOCA transient in both conventional and passive PWR plants. Kataoka and Ishii's pool entrainment model is widely used in small break LOCA and reactor transient analysis, but has limited validation against experimental data in high gas flux region. This paper is to report an experimental study of air–water pool entrainment with prototypic gas flux conditions of AP1000. The test section is 2200 mm long and 380 mm in diameter, and built by transparent material for visualization. The range of air flow superficial velocity is 0.98–5.41 m/s. The two phase mixture level is measured by Guided Wave Radar (GWR), and the water entrainment is recorded by weight measure after it is separated from entrainment separator. A correlation of pool entrainment in high gas flux region is proposed based on experimental data obtained in this study. The phenomenon of entrainment saturation is found in high gas flux region. The visualization indicates that pool entrainment in high gas flux region is in transition to jet/fountain flow, which might have strong exit effect.

Analyses of ACME integral test results on CAP1400 small-break loss-of-coolant-accident transient

CAP1400 has been developed as a new passive safety and generation III + reactor based on AP600/1000 technology. Like previous reactors, the small-break loss-of-coolant-accident (SBLOCA) is still one of the most important safety problems that have to be considered before it's widely applications. Although extensive theoretical and experimental results of SBLOCA have been reported in previous studies, the dominant SBLOCA transient mechanisms and the important thermal hydraulic phenomena have not been clearly explained, especially in the passive safety system, due to the complexity of the passive safety nuclear reactor geometry coupled with the transient two-phase fluid interactions. In this study, a new constructed thermal hydraulic integral effect test facility, named as advanced core-cooling mechanism experiment (ACME), which is designed and constructed for supporting CAP1400 safety review, was used to investigate the SBLOCA transient in passive safety plants. 1-inch, 2-inch and 8-inch cold leg breaks and a double-ended direct-vessel-injection (DEDVI) line break were selected as the critical and essential SBLOCA tests. The system pressure, injection flow rate, core level and fluid temperature were considered as the safety related parameters to investigate the system responses to the SBLOCA. In addition, the conditions that affect the SBLOCA chronology were investigated as well. Also, a few important thermal hydraulic phenomena observed in the ACME SBLOCA test were reported, and their effects on the transient were analyzed. The dominant SBLOCA transient mechanisms in the passive safety system were presented. This study not only helps to improve the understanding of SBLOCA mechanisms and important thermal hydraulic phenomena in passive safety systems, but also sheds some lights on SBLOCA analysis, test facility scaling, safety performance evaluation, code simulation and even the design optimization for the passive safety PWRs.

Post-test analysis of the ROSA/LSTF and PKL counterpart test

Experimental facilities are scaled models of commercial nuclear power plants, and are of great importance to improve nuclear power plants safety. Thus, the results obtained in the experiments undertaken in such facilities are essential to develop and improve the models implemented in the thermal-hydraulic codes, which are used in safety analysis. The experiments and inter-comparisons of the simulated results are usually performed in the frame of international programmes in which different groups of several countries simulate the behaviour of the plant under the accidental conditions established, using different codes and models. The results obtained are compared and studied to improve the knowledge on codes performance and nuclear safety.Thus, the Nuclear Energy Agency (NEA), in the nuclear safety work area, auspices several programmes which involve experiments in different experimental facilities. Among the experiments proposed in NEA programmes, one on them consisted of performing a counterpart test between ROSA/LSTF and PKL facilities, with the main objective of determining the effectiveness of late accident management actions in a small break loss of coolant accident (SBLOCA). This study was proposed as a result of the conclusion obtained by the NEA Working Group on the Analysis and Management of Accidents, which analyzed different installations and observed differences in the measurements of core exit temperature (CET) and maximum peak cladding temperature (PCT). In particular, the transient consists of a small break loss of coolant accident (SBLOCA) in a hot leg with additional failure of safety systems but with accident management measures (AM), consisting of a fast secondary-side depressurization, activated by the CET. The paper presents the results obtained in the simulations for both installations using TRACE, observing, in general, a good agreement with the experiments. However, ROSA/LSTF calculations underestimated the maximum PCT value, what might be explained by the higher core level predicted in the simulation compared with the experiment. In PKL calculations, PCT maximum value is slightly higher than in the experiment, and the core level predicted is lower. In the comparison of the evolution of both installations a different timing in the transient events is observed, due to the difference in the pressure vessel design. Thus, when PKL vessel is modified with some of the ROSA/LSTF features, the evolution of the new PKL model behaviour is closer the one observed in ROSA/LSTF calculations.

Analysis of the small break loss of coolant accident in the VVER-1000/V446 reactor

Abstract In this paper, the analysis of a Small Break Loss of Coolant Accident (SBLOCA) in the VVER-1000/V446 nuclear power plant is presented. For a conservative analysis of the accident, the loss of power to the NPP and failure of one accumulator, and also of two emergency core cooling systems (ECCS) in loops 2 and 3 of the primary and secondary circuits are considered when SBLOCA has occurred. The RELAP5/MOD3.2 computer code has been used in performing the analyses. Two cases of accident scenarios as 25 mm and 100 mm breaks are analyzed. The results are in good agreement with those reported in the plant's FSAR. The results of liquid velocity show that in both cases, the flow of hot legs after the break is reversed, which provides the potential for reflux condensation phenomena. Furthermore, in the 25 mm break, the flow rate in the broken and intact side downcomer remains in the downward motion while in the 100 mm break, the broken and intact side flow rate changes to the reversed state alternatively.

Analytical studies of long-term IRWST injection core cooling under small break LOCA in passive safety PWR

As an important mechanism for passive core cooling in the AP1000 and CAP1400 small-break loss-of-coolant-accident (SBLOCA) scenarios, the IRWST gravity injection provides the largest inventory for the long-term core decay heat removal. This study focuses on the analytical analyses of the long-term IRWST injection core cooling process. By recognizing the quasi-steady-state nature of the process from its phenomenology analysis, a simple 1-D thermal hydraulic analytical model for analyzing the system behavior has been developed. The model is based on the fundamental conservation laws and well-established correlations for two-phase flow. Using a numerical solution approach for the model, important parameters such as the injection flow rate, core level and ADS-4 venting mass flow rate for characterizing the long-term IRWST injection core cooling have been determined. The model results have also been compared with the ACME integral test data. The tests were carried out under different typical SBLOCA conditions, including 2-in. and double-ended DVI breaks of various boundary conditions. The comparison shows good agreements between the model results and the experimental data for the selected SBLOCA cases. Finally, the conditions that affect the analysis results, the asymmetric ADS-4 discharge flow distribution and the model applicability are discussed.

Comparison of critical flow models for SBLOCA tests of the ATLAS facility

A comparison of critical flow models between the Trapp–Ransom and Henry–Fauske models for all SBLOCA (small break loss of coolant accident) scenarios of the ATLAS (advanced thermal–hydraulic test loop for accident simulation) facility was performed using the MARS-KS code. For the comparison of the two critical models, the accumulated break mass was selected as the main parameter for the comparison between the calculations and tests. Four cases showed the same respective discharge coefficients between the two critical models, e.g., 6″ CL (cold leg) break and 25%, 50%, and 100% DVI (direct vessel injection) breaks. In the case of the 4″ CL break, no reasonable results were obtained with any possible Cd values. In addition, typical system behaviors, e.g., PZR (pressurizer) pressure and collapsed core water level, were also compared between the two critical models. From the comparison between the two critical models for CL breaks, the Trapp–Ransom model predicted quite well with respect to the other model for the smallest and larger breaks, e.g., 2″, 6″, and 8.5″ CL breaks. In addition, from the comparison between the two critical models for DVI breaks, the Trapp–Ransom model predicted quite well with respect to the other model for the smallest and larger breaks, e.g., 5%, 50%, and 100% DVI breaks. In the case of 50% and 100% breaks, the two critical models predicted the test data quite well.