Effect Test Facility Research Articles

Droplet Entrainment in Steam Supply System of Water-Cooled Small Modular Reactors: Experiment and Modeling Approaches

Droplet entrainment in steam-flow is a prominent phenomenon that needs adequate safety and risk analysis of postulated transient and accident scenarios—including experimental investigation and representative modeling and simulation (M&S)—for small modular reactor (SMR) system design and demonstration. This study identifies knowledge gaps by evaluating experimental and computational fluid dynamics modeling approaches to support early-stage reactor system design, testing, and model evaluation. Previous studies reported in the literature for steam-flow entrainment primarily focused on gigawatt capacity pressurized water reactor (PWR) systems. However, entrainment phenomena are even more prominent for PWR-type SMRs due to their more compact integrated designs, which need further research and development. To fill the research gaps, this study provides insight by specifying the phenomena of interest by leveraging the lessons learned from past research, adopting advanced M&S techniques and advanced instrumentation and control. The findings and recommendations are applicable for evaluating steam-flow entrainment models and for designing integral effect test and separate effect test facilities for gaining reactor design approvals.

Separate effect experimental test and theoretical investigate of pressure wave propagation and pressure rising process upon SGTR accident in advanced LFRs

Steam Generator Tube Rupture accident (SGTR) accident is one of the main safety events to the industrial application of Lead-cooled Fast Reactors (LFRs). This paper presents a separate effect test facility, called Lead-bismuth Eutectic Steam generator tube rupture Test facility (LEST), for testing high-pressure subcooled water jetting into a Lead Bismuth Eutectic (LBE) pool, and analyzes experimental data of maximum initial pressure peak, dynamic pressure attenuation together with the pressure rising process upon SGTR. The experimental analyses put forward a general equation form together with a dimensionless number of the initial pressure peak associated with a high-pressure Newtonian fluid injecting into another low-pressure Newtonian fluid, and the derivation of the general equation form is specifically applied to match a further semi-empirical equation for an initial maximum pressure peak with 5–10 MPa water injection in SGTR. The attenuation curve of dynamic pressure wave in each experiment is matched, and more weakening conditions can change the pressure peak from concentrated and non-period peak group to approximate period 1.4–3.7 s pulse group. The size relationship between pipe length L and nozzle diameter D, which L/D=12, is set to divide SGTR into small wall rupture and tube shear fracture, further three stages of pressure rising are distinguished in the tube shear fracture. This work provides experimental data to help determining the severity of SGTR and assessing the consequences of this event as part of the safety case for the industrial application of LFRs.

RELAP5-3D validation studies based on the High Temperature Test facility

In the spring and summer of 2019, experiments were conducted at the High Temperature Test Facility (HTTF) that form the basis of an upcoming high-temperature gas-cooled reactor (HTGR) thermal hydraulics (T/H) benchmark. HTTF is an integral effects test facility for HTGR T/H modeling validation. This paper presents RELAP5-3D models of two of those experiments: PG-27, a pressurized conduction cooldown (PCC); and PG-29, a depressurized conduction cooldown (DCC). These models used the RELAP5-3D model of HTTF originally developed by Paul Bayless as a starting point. The sensitivity analysis and uncertainty quantification code, RAVEN was used to perform calibration studies for the steady-state portion of PG-27. We developed four PG-27 calibrations based on steady-state conditions. These calibrations all used an effective thermal conductivity equal to 36 % of the measured thermal conductivity, but they differed with respect to the frictional pressure drops and radial conduction models. These models all captured the trends in steady-state temperature distributions and transient temperature behavior well. All four calibrations show room for improvement in predicting the transient temperature rise. The smallest error in temperature rise during the transient was a 21 % underprediction, and the largest was a 48 % underprediction. The errors in transient temperature rise are largely a result of a mismatch in power density between the RELAP5-3D model and the experiment due to the location of active heater rods along the boundary between heat structures in the model. The best of these calibrations was applied to PG-29 to model the DCC. Once again, temperatures during the transient were underpredicted but trends in temperature were captured. The RELAP5-3D model captured trends in the data but could not reproduce measured temperatures exactly. This result is not attributed to deficiencies in the experimental data or to RELAP5–3D itself. Rather, this result likely arises due to the some of the assumptions and decisions made when the RELAP5-3D model was first developed, prior to the execution of HTTF experiments. An agreement in prediction of temperature trends but challenges reproducing HTTF temperatures within measurement uncertainty is consistent with previous analyses of HTTF in the literature. Future RELAP5-3D validation activities centered around HTTF may be able to provide greater insight into the code’s capabilities for HTGR modeling with a more finely nodalized model.

Analysis of AP1000 Small-Break Loss-of-Coolant Accident Using Reactor Transient Simulator

The Westinghouse Electric Company’s Advanced Passive Reactor (AP1000) is characterized by the incorporation of passive safety systems (PSSs) designed to ensure core cooling during transient events. The assessment of PSSs requires evaluation of their performance through a combination of experiments and simulations employing various thermal-hydraulic codes. In addition, detailed evaluation of PSSs for a specific reactor system transient analysis such as loss-of-coolant-accident analysis supports understanding representative integral effects test facility development and the further evolution model development and assessment process. Developing a reactor system code is a complex and time-consuming process that requires significant engineering expertise and effort. It can take several months to even years to complete in the early stages of reactor system design and analysis. However, this process can be expedited through the use of transient simulator models for similar reactor systems, which can be used for lesson learning and training purposes. This study uses the Personal Computer Transient Analyzer (PCTRAN) code. The main advantage of PCTRAN is its ease of use and ability to run faster than real time. This study presents the results obtained for a small-break loss-of-coolant accident (SBLOCA) for two breaks using the full version (licensed) of PCTRAN. The purpose of this investigation is to evaluate the overall system behavior during the postulated SBLOCA event as well as assess the capability of the PCTRAN code to reproduce the system response during transient events. The obtained results were compared with the Westinghouse NOTRUMP system code. The PCTRAN code proved to be reliable in predicting the qualitative behavior of the system in both transient cases. As for the system response, it was found that it is contingent on the activation time of the PSSs. The differences in reactor coolant system pressure between the two codes were attributed to the critical flow model and simplification of mass and energy balance. Despite PCTRAN’s limitations, it can still provide a reasonable prediction of various reactor parameters such as pressure, mass flow rate, and void fraction during a SBLOCA scenario. It is worth noting that PCTRAN currently employs a bulk approach similar to that of the Modular Accident Analysis Program (MAAP) and MELCOR codes. However, the upcoming version of PCTRAN will include an artificial intelligence–based detection and accident prevention system, as well as different models for different reactor components. Consequently, PCTRAN has the potential to be upgraded to match the system thermal-hydraulic codes of the U.S. Nuclear Regulatory Commission and become more widely used in cybersecurity to safeguard nuclear power plants from cyberattacks.

A review of natural circulation flow in integral effect test facilities

The proposed NCFM map was reviewed using natural circulation data identified in the PWR simulated integral effect test (IET) facility with a height (length) scale ratio of 1/1. Based on this, the characteristics of natural circulation flow in the ATLAS and FESTA facilities, which are major IET facilities in Korea, are compared. It was confirmed that the natural circulation flow characteristics were well maintained within the NCFM area even at the half-height test facility of ATLAS. It was also confirmed that the natural circulation flow characteristics were well maintained within the NCFM area at the full-height test facility of FESTA, and the natural circulation flow characteristics were also maintained outside the NCFM region for a relatively small amount of coolant inventories.

Scaling analysis of nuclear power plant (APR1400) utilizing ATLAS integral effect test result of A5.1 (Counterpart test for LSTF 1% cold-leg break LOCA)

ATLAS is the thermal hydraulic integral effect test loop that can simulate the same pressure and temperature condition with nuclear power plant. It was designed with a height ratio of 1/2 and a volume ratio of 1/288 with reference to APR1400. The counterpart test against LSTF which was named as A5.1 test was performed within the OECD/NEA ATLAS international joint research project to compare major thermal hydraulic phenomena between two test facilities. The main scenario for the A5.1 test is the LSTF SB-CL-32 test. The SB-CL-32 test is a 1% horizontal small break loss of coolant accident (SBLOCA) at the cold leg with cold leg injection of the emergency core cooling system (ECCS) and accident management (AM) action.From the counterpart test results, the significant differences were found in the viewpoint of the thermal–hydraulic phenomena. It was concluded that the different result were caused by the different design characteristic of each test facility because they referred the different nuclear power plant. Thus, the applicability of the ATLAS test results to its reference nuclear power plant was evaluated by utilizing ATLAS A5.1 test results which was the counterpart test with the LSTF.The transient scenario and initial and boundary conditions of the A5.1 test were adopted to APR1400 after applying scaling ratios of ATLAS. The MARS-KS code was utilized to analyzed thermal hydraulic behaviors of APR1400 and the analysis results were compared with the scaled A5.1 test result. From the result of comparison, it was concluded that the ATLAS test results predicted the thermal hydraulic behavior of the nuclear power plant well. Thus, it was confirmed that ATLAS is a well-designed integral effect test facility which is designed with an appropriate scaling method.

Scaling methodologies and similarity analysis for thermal hydraulics test facility development for water-cooled small modular reactor

Small modular reactors (SMRs) represent a promising option for providing clean and sustainable energy due to their potential for enhanced safety, reduced capital costs, and increased siting flexibility. However, new reactor systems require the development and operation of representative scaled-down test facilities to support the verification and validation of system computer codes and models. This study reviews the research on scaling methodologies and similarity principles pivotal in developing non-nuclear integral effects test and separate effects test facilities for water-cooled SMRs. The study focuses on a review of the scaling methods, similarity approaches, and possible challenges posed by the unique and compact design features of integral-pressurized water reactor-type SMRs, and their representative test facilities. This study also reviews previous research related to scaling and similarity methodologies and provides insights into design considerations for achieving prototypic conditions in test facilities. The findings and recommendations emphasize the broader impact of appropriate scaling and similarity principles to ensure meaningful and transferable results from non-nuclear test facilities to accelerate the safe and efficient deployment of next-generation water-cooled SMRs.

Analysis of control rod driving mechanism nozzle rupture with loss of safety injection at the ATLAS experimental facility using MARS-KS and TRACE

Korea Atomic Energy Research Institute (KAERI) has operated an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), with reference to the APR1400 (Advanced Power Reactor 1400) for tests for transient and design basis accidents simulation. A test for a loss of coolant accident (LOCA) at the top of the reactor pressure vessel (RPV) had been conducted at ATLAS to address the impact of the loss of safety injections (LSI) and to evaluate accident management (AM) actions during the postulated accident. The experimental data has been utilized to validate system analysis codes within a framework of the domestic standard problem program organized by KAERI in collaboration with Korea Institute of Nuclear Safety. In this study, the test has been analyzed by using thermal-hydraulic system analysis codes, MARS-KS 1.5 and TRACE 5.0 Patch 6, and a comparative analysis with experimental and calculation results has been performed. The main objective of this study is the investigation of the thermal-hydraulic phenomena during a small break LOCA at the RPV upper head with the LSI as well as the predictability of the system analysis codes after the AM actions during the test. The results from both codes reveal that overall physical behaviors during the accident are predicted by the codes, appropriately, including the excursion of the peak cladding temperature because of the LSI. It is also confirmed that the core integrity is maintained with the proposed AM action. Considering the break location, a sensitivity analysis for the nodalization of the upper head has been conducted. The sensitivity analysis indicates that the nodalization gave a significant impact on the analysis result. The result emphasizes the importance of the nodalization which should be performed with a consideration of the physical phenomena occurs during the transient.

Hydrodynamic design of the Separate Effect test facility for Flow-Accelerated Corrosion and Erosion (SEFACE) studies in liquid lead

Flow-accelerated corrosion and erosion (FACE) phenomena can be crucial for performance of structural elements in heavy liquid metal (HLM) cooled reactor systems. Existing experimental observations indicate that turbulent flow characteristic can affect FACE, but there is no quantitative data that can be used for model development and validation. Main recirculation pump impellers, which operate at high relative velocities and rotational flow conditions can be especially vulnerable to FACE. For comparison, the core internals operate at lower velocities and in axial flow conditions, but at higher temperatures and neutron fluence. Hence, systematic experimental data is needed to improve our knowledge on FACE phenomena. The Separate Effect Test Facility for Flow-Accelerated Corrosion and Erosion (SEFACE) is designed to obtain such experimental data including high relative velocities (up 20 ms−1) and high temperatures (400 to 550 °C) of liquid lead. This article focuses on the hydrodynamic design of SEFACE. The aim of the design is to achieve well defined flow conditions for experiments and ensure safe operation of the facility. First, we examine three design concepts (i.e., forced convection loop, rotating cylinder, and rotating disk) and motivate the choice of the rotating disk approach for SEFACE. Second, we discuss different design options, i.e., a confined rotor–stator test chamber and the unconfined rotating disk configuration. We used Reynolds-Averaged Navier Stokes (RANS) calculations to identify and solve the issues stemming from the high rotational speed. These include, for instance, lead free surface deformation, radial pressure buildup, and axial bending forces due to asymmetric test chamber. The CFD-derived torque and power predictions in rotor–stator and rotating disk systems are verified with selected empirical turbulent friction factor correlations or/and DNS calculations. We demonstrate that the developed hydrodynamic design of SEFACE solves identified issues and enables obtaining experimental data under well-defined flow conditions. The findings are deemed to also be applicable to the design of rotating disk-type FACE installations for other liquid mediums.

The LIFUS5 separate effect test facility experimental programme

The LIFUS5 facility is a separate effect test facility aimed at investigating the heavy liquid metal-water interaction. It has been designed and constructed to withstand high pressure and temperature (i.e. up to 200 bar and up to 500 °C) and to record the fast pressure transients in heavy liquid metal melt. This type of transient is typical of a Steam Generator Tube Rupture event in a pool type Gen IV Heavy Liquid Metal cooled Fast Reactor system, and should be investigated from the safety point of view because it could potentially induce, beyond the damaging of the internal structures (HX tube bundle, above core structures, Fuel Assembly, Control Rods, etc.) several negative effects on the operation of the reactor. These include an insertion of positive reactivity into the system or reduced cooling efficiency due to steam dragging into the core. It will also have an effect on the chemistry control of the cooling. All these effects compromise the safety and the reliability of the system. The twenty years experimental programme and the different LIFUS5 configurations are presented in the paper, as well as the experimental campaigns and test matrix characterizing the heavy liquid metal-water interaction phenomena and data for codes validation.

Progress towards the validation of SIMMER-III code model for lead-lithium water chemical interaction

• SIMMER-III code validation for WCLL in-box LOCA assessment • Lead-lithium eutectic and water chemical interaction • Numerical Analysis of Experimental Test E5.2 • Execution of the Series E experimental campaign employing the facility, LIFUS5/Mod3 Research activities are continuing between the University of Pisa and ENEA Brasimone Research Center to better understand the phenomena and processes that occur during a postulated in-box LOCA in the Water-Cooled lead-lithium Breeding Blanket (WCLL BB) and during the system's safety response. Various activities are also going on to strengthen the reliability of numerical tools and validate computer models, codes, and procedures for their implementations. The study presented here aims to assist in the validation process of the SIMMER-III code model, to simulate the lead-lithium and water interaction under conditions like those expected for the WCLL BB during accident conditions. In particular, the current work has been performed by numerically reproducing the experimental Test E5.2, executed in the separate effect test facility LIFUS5/Mod3 (installed at ENEA Brasimone Research Center).The code used for the simulation was SIMMER-III computational code, in a version modified by the University of Pisa in order to take into account also the chemical interaction between the two fluids. A comparison of significant parameters computed by the SIMMER code with those obtained in the experimental test during the transient has been carried out and it is presented and described in this paper.

Integral and Separate Effects Test Facilities To Support Water Cooled Small Modular Reactors: A Review

This study reviews previous experimental facilities and test programs relevant to water-cooled reactor system design and analysis to meet regulatory compliances. This study aims to find the best solution for designing the required experiments, obtaining necessary test data, and verifying the developed computer code/models to support the new reactor design and development while minimizing cost and time while leveraging experiences from previous facilities to minimize. Nuclear reactor licensing requires supportive design, analysis, and experimental results to ensure the safety of the full-scale prototype reactor in regular operation, as well as during postulated accident scenarios. These reactor design analyses are generally performed using system codes and other associated simulation tools that require assessment, verification and validation using an appropriate experimental dataset. Experimental facilities used for reactor system safety analysis and system code assessments are categorized into integral effects test (IET) and separate effects test (SET) facilities. These IET and SET experiments and studies use geometrically scaled systems to reproduce the prototype system behavior at a reasonable cost, albeit with some scaling-related distortions. The design challenge of these model facilities is to identify and minimize scaling distortions while reproducing the most important operational phenomena in steady-state operation and in postulated accident scenarios. Lessons learned from previous experimental facilities, models, and correlations can support the development of new multipurpose, scaled, hybrid, integrated, and modular experimental facilities for advanced light water-cooled small modular reactors (SMRs). Successful operation of these facilities can significantly reduce upfront reactor development and demonstration costs and time to deployment.

Experimental Results of PbLi-Water Reaction Performed in LIFUS5/Mod3 Separate Effect Test Facility

A major safety concern addressed during the design of the water-cooled lead-lithium (PbLi) breeding blanket (BB) is represented by an in-box loss-of-coolant accident, where high-pressure water is supposed to interact with PbLi inside the BB. Code development activities are being carried out to create the needed tools for the safety analysis of these systems in the case of incidental scenarios, which are supported by extensive experimental campaigns that aim at providing data for SIMMER code verification and validation (V&V). In this regard, the present work aims at presenting the dataset generated during the Series D experimental campaign performed at the LIFUS5/Mod3 facility operated at the ENEA Brasimone Research Centre. This is a separate effect facility able to simulate the mixing of water and PbLi alloy in conditions of temperature and pressure similar to the ones encountered by the system during nominal operation and to acquire significant data on all the relevant thermochemical parameters. Moreover, a preliminary analysis of the data has been performed to critically determine the quality of the data and to identify possible issues in the experimental process. In the end, the foreseen extension of the experimental work is described, as well as the foreseen application of the acquired data in the code V&V activities.

Analysis of 4-inch cold leg top-slot break LOCA in ATLAS experimental facility using MARS-KS

Abstract Korea Atomic Energy Research Institute (KAERI) has operated an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS) with reference to the Advanced Power Reactor 1400 MW (APR1400) for transient and design basis accidents (DBAs) simulation. An experiment for a 4-inch cold leg top-slot break was performed at ATLAS to resolve a safety issue that the loop seal reformation (LSR) of APR1400 could lead to the increase of peak cladding temperature. In addition, the experimental data has been utilized to validate system codes within a framework of domestic standard problem (DSP) program organized by KAERI in collaboration with Korea Institute of Nuclear Safety (KINS). In this study, the experiment has been analyzed by thermal-hydraulic system analysis code, MARS-KS 1.5 and a comparison with experimental and calculation results has been performed. Since the top-slot break is not a typical break geometry for safety analyses, this study aims at examining the applicability of MARS-KS to the top-slot break accident where the LSR occurs repeatedly. The results revealed that overall physical behavior during the accident was predicted by the code, appropriately. MARS-KS showed the excursion of the peak cladding temperature because of the LSR as in experiment. It has been confirmed that the core integrity was maintained because the temperature excursion by the LSR was not large enough to alter the acceptance criteria. In addition, it is presented the results of the sensitivity analysis of parameter that affect the figure of merits.

OECD/NEA/CSNI/WGAMA PERSEO benchmark: Main outcomes and conclusions

In the framework of the OECD/NEA/CSNI/WGAMA, an activity on the “Status report on thermal–hydraulic passive systems design and safety assessment” has been conducted. Within this activity, a benchmark exercise, based on the experimental data developed in the full scale PERSEO (in-Pool Energy Removal System for Emergency Operation) component separate effect test facility, built at SIET (Piacenza, Italy), has been proposed and carried out. An “OPEN” benchmark exercise, hosted by ENEA, has been conducted. Twelve results from eleven Organizations were submitted. PERSEO is a full-scale separate effect test facility designed to study a new passive decay heat removal system operating in natural circulation. Test 7 is a full pressure test (7 MPa) and investigates the system stability and the system operation. The accuracy of the calculated results has been evaluated both qualitatively and quantitatively. The latter has been conducted adopting the Fast Fourier Transform Based Method. The present paper summarizes the main features of the PERSEO facility and Test 7 and discusses the main results and outcome of the benchmark exercise.

Modeling of horizontal steam injection in a water pool

A comprehensive understanding of the direct contact condensation (DCC) phenomena of steam discharged into a subcooled water pool is essential for the nuclear engineering field. DCC is the most rapid and efficient way to condense high-pressure and high-temperature steam with simple engineering systems. However, the pressure oscillation due to the rapid condensation and small length and time scales of turbulent two-phase flow make the investigation of DCC phenomena challenging either by experiments or with numerical simulations.This paper presents the computational fluid dynamics (CFD) simulations of the small-scale separate effect test facility (SEF-POOL) experiment of Lappeenranta-Lahti University of Technology (LUT University). In the SEF-POOL test, the steam was injected using the orifice of a diameter of 16 mm. All simulations were performed by employing the compressible two-phase solver of OpenFOAM which was based on the Eulerian–Eulerian two-fluid approach. The interfacial heat transfer between steam and water was modeled by using the Nusselt number formulation of Coste (2004). The Rayleigh–Taylor interfacial (RTI) area model of Pellegrini et al. (2015) was applied for the interfacial area modeling. In this work, the formation and collapse of the steam bubbles are studied using the extended pattern recognition based image analysis algorithm. The pattern recognition (PR) algorithm is based on the video material recorded during the SEF-INF2 experiment of the SEF-POOL test facility.The presented study show that if the two-phase interface is rough and smear into the subgrid scale, the interfacial area density calculation solely with the volume fraction gradient is inadequate for predicting rapid condensation rates. Results indicate that rapid and high interfacial accelerations in DCC trigger interfacial instabilities which increased the interface roughness and total interfacial area consequently. Thus, interfacial area modeling with the RTI increased the total interfacial heat transfer rate in the simulations. The achieved results exhibit that the selected modeling approach captured the rapid bubbling condensation oscillation mode of the SEF-POOL test. Presented results demonstrate that image analysis and pattern recognition have great potential in thermal-hydraulic research that can be beneficial for direct contact condensation and phase-interface dynamics model development.

Development of fault diagnosis for nuclear power plant using deep learning and infrared sensor equipped UAV

Fault component detection is necessary for safety and maintenance in large-scale industrial fields including nuclear power plants. Therefore, this study proposes a method for diagnosing a power plant composed of numerous components based on deep learning using a UAV with an IR sensor and a camera. The proposed method could diagnose the components and recognize the fault component in real time. In this study, a thermal–hydraulic integral effect test facility, which is a scaled-down nuclear power plant, is utilized considering the nuclear power plant. The database for the application of deep learning was performed by combining an IR intensity map and general image to enhance the performance of component classification and fault detection. Deep learning was applied using object detection and classification methods based on convolutional neural networks (CNNs) that are effective for image processing. As a result, this technology can diagnose the multi-component by a single measurement instrument. The optimal performance of component classification and fault detection was 55.9 ms per 16 batches, demonstrating a mean average precision (mAP) of 0.9913. This technology could be applied to various industries as a comprehensive component condition monitoring method for operating efficiency and safety.

Deep-learning-based system-scale diagnosis of a nuclear power plant with multiple infrared cameras

Comprehensive condition monitoring of large industry systems such as nuclear power plants (NPPs) is essential for safety and maintenance. In this study, we developed novel system-scale diagnostic technology based on deep-learning and IR thermography that can efficiently and cost-effectively classify system conditions using compact Raspberry Pi and IR sensors. This diagnostic technology can identify the presence of an abnormality or accident in whole system, and when an accident occurs, the type of accident and the location of the abnormality can be identified in real-time. For technology development, the experiment for the thermal image measurement and performance validation of major components at each accident condition of NPPs was conducted using a thermal-hydraulic integral effect test facility with compact infrared sensor modules. These thermal images were used for training of deep-learning model, convolutional neural networks (CNN), which is effective for image processing. As a result, a proposed novel diagnostic was developed that can perform diagnosis of components, whole system and accident classification using thermal images. The optimal model was derived based on the modern CNN model and performed prompt and accurate condition monitoring of component and whole system diagnosis, and accident classification. This diagnostic technology is expected to be applied to comprehensive condition monitoring of nuclear power plants for safety.

Similarity evaluation of the pump simulation loop in STELLA-2 for conservation of mechanical sodium pump characteristics

The STELLA-2 is a large-scale sodium thermal-hydraulic integral effect test facility and supports the development of PGSFR. The facility adopted Pump Simulation Loop System (PSLS) concept for the mechanical sodium pump in the reference reactor to control and to measure the primary sodium flow. Since the component (mechanical pump) is replaced by the loop, it is very important to evaluate the similarity between the pump and the loop. In this paper, to simulate the characteristic of the mechanical sodium pump, the pressure loss along the various options of the loop was evaluated and the comprehensive validity of each design options was analyzed. Using the similarity criteria based on the Richardson number and Euler number conservation, the PSLS design was finalized and the result was within the acceptable error range. Finally, the result of this study was used for construction of the overall facility, STELLA-2.

RELAP5 validation for core makeup tank analysis based on separate effect tests

The RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during accident and has been widely used in the field of nuclear thermal-hydraulics and safety analysis. The core make-up tank (CMT) is a crucial device of the passive core cooling system in the AP series pressurized water reactor nuclear power plants. Since the CMT not only provides full-pressure injection but also triggers auto depressurization system by its level signal during small-break loss of coolant accident. Therefore, it is necessary to carry out CMT separate effect tests to verify the feasibility and accuracy of RELAP5/MOD 3.4 to simulate CMT. The previous studies found that the lower version of the RELAP5 program has problems in simulating CMT, and the applicability of the latest RELAP5/MOD3.4 has not been evaluated. In this paper, a CMT separate effect test facility were presented and then RELAP5/MOD 3.4 was used to establish a CMT test bench model to evaluate the applicability, and obtain the general rules and methods of RELAP5/MOD 3.4 to simulate CMT. It’s found that RELAP5/MOD 3.4 can accurately simulate most of the CMT operating conditions. However, large uncertainty is expected in the simulation of delayed time and the flowrate of delayed drainage stage is not affected by the heat transfer coefficient of steam direct contact condensation. The simulation of delay time depends on experimental data and the CMT delayed drainage experiment cannot verify the accuracy of the RELAP5/MOD 3.4 direct contact condensation model.